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{{#Wiki_filter:MPS-3 FSARMillstone Power Station Unit 3 Safety Analysis Report Chapter 3 MPS-3 FSAR 3-i Rev. 30CHAPTER 3 - DESIGN OF STRUCTURE S, COMPONENTS, EQUIPMENT, AND SYSTEMS Table of ContentsSection Title Page3.1CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA...................3.1-13.1.1Conformance with Single Failure Criterion................................................3.1-13.1.1.1Single Failure Criterion..............................................................................3.1-13.1.1.2Definitions of Terms Used in Single Failure Criterion...............................3.1-13.1.1.3Application of Single Failure Criterion......................................................3.1-23.1.2Criterion Conformance...............................................................................3.1-63.1.2.1Quality Standards and Records (Criterion 1)..............................................3.1-63.1.2.2Design Bases for Protection against Natural Phenomena (Criterion 2)......3.1-63.1.2.3Fire Protection (Criterion 3).......................................................................3.1-83.1.2.4Environmental and Missile Design Bases (Criterion 4).............................3.1-93.1.2.5Sharing of Structures, System s, and Components (Criterion 5)...............3.1-103.1.2.6(Criterion 6)..............................................................................................3.1-113.1.2.7(Criterion 7)..............................................................................................3.1-113.1.2.8(Criterion 8)..............................................................................................3.1-113.1.2.9(Criterion 9)..............................................................................................3.1-113.1.2.10Reactor Design (Criterion 10)...................................................................3.1-113.1.2.11Reactor Inherent Protection (Criterion 11)...............................................3.1-123.1.2.12Suppression of Reactor Power Os cillations (Criterion 12).......................3.1-123.1.2.13Instrumentation and Control (Criterion 13)..............................................3.1-133.1.2.14Reactor Coolant Pressure Boundary (Criterion 14)..................................3.1-143.1.2.15Reactor Coolant System Design (Criterion 15)........................................3.1-143.1.2.16Containment Design (Criterion 16)..........................................................3.1-153.1.2.17Electric Power Systems (Criterion 17).....................................................3.1-153.1.2.18Inspection and Testing of Electric Power Systems (Criterion 18)............3.1-163.1.2.19Control Room (Criterion 19)....................................................................3.1-173.1.2.20Protection System Functions (Criterion 20).............................................3.1-173.1.2.21Protection System Relia bility and Testability (Criterion 21)...................3.1-183.1.2.22Protection System Independence (Criterion 22).......................................3.1-193.1.2.23Protection System Failure Modes (Criterion 23)......................................3.1-193.1.2.24Separation of Protection a nd Control Systems (Criterion 24)..................3.1-203.1.2.25Protection System Requirements fo r Reactivity Control Malfunctions (Criterion 25)............................................................................................3.1-203.1.2.26Reactivity Control System Redundancy and Capability (Criterion 26)...3.1-213.1.2.27Combined Reactivity Control System Capability (Criterion 27)..............3.1-223.1.2.28Reactivity Limits (Criterion 28)...............................................................3.1-223.1.2.29Protection against Anticipated Operational Occurrences (Criterion 29)..3.1-233.1.2.30Quality of Reactor Coolant Pressure Boundary (Criterion 30).................3.1-233.1.2.31Fracture Prevention of Reactor Coolant Pressure Boundary (Criterion 31).
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3.1-MPS-3 FSARCHAPTER 3 -DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYS-TEMS Table of Contents (Continued) 3-ii Rev. 30 243.1.2.32Inspection of Reactor Coolant Pressure Boundary (Criterion 32)............3.1-253.1.2.33Reactor Coolant Makeup (Criterion 33)...................................................3.1-263.1.2.34Residual Heat Removal (Criterion 34).....................................................3.1-263.1.2.35Emergency Core Cooling (Criterion 35)..................................................3.1-273.1.2.36Inspection of Emergency Core Cooling System (Criterion 36)................3.1-283.1.2.37Testing of Emergency Core Cooling System (Criterion 37)....................3.1-283.1.2.38Containment Heat Removal (Criterion 38)...............................................3.1-283.1.2.39Inspection of Containment Heat Removal System (Criterion 39)............3.1-293.1.2.40Testing of Containment Heat Removal System (Criterion 40).................3.1-293.1.2.41Containment Atmosphere Cleanup (Criterion 41)....................................3.1-303.1.2.42Inspection of Containment Atmosphere Cleanup Systems (Criterion 42)3.1-313.1.2.43Testing of Containment Atmosphere Cleanup Systems (Criterion 43)....3.1-313.1.2.44Cooling Water (Criterion 44)....................................................................3.1-313.1.2.45Inspection of Cooling Water System (Criterion 45).................................3.1-323.1.2.46Testing of Cooling Water System (Criterion 46)......................................3.1-333.1.2.47(Criterion 47)............................................................................................3.1-333.1.2.48(Criterion 48)............................................................................................3.1-333.1.2.49(Criterion 49)............................................................................................3.1-333.1.2.50Containment Design Basis (Criterion 50).................................................3.1-343.1.2.51Fracture Prevention of Containment Pressure Boundary (Criterion 51)..3.1-343.1.2.52Capability for Containment Leakage Rate Testing (Criterion 52)...........3.1-353.1.2.53Provisions for Containment Testing and Inspection (Criterion 53)..........3.1-353.1.2.54Piping Systems Penetrating Containment (Criterion 54)..........................3.1-363.1.2.55Reactor Coolant Pressure Boundary Penetrating Containment (Criterion 55)....
3.1-363.1.2.56Primary Containment Isolation (Criterion 56)..........................................3.1-373.1.2.57Closed System Isolation Valves (Criterion 57)........................................3.1-383.1.2.58(Criterion 58)............................................................................................3.1-383.1.2.59(Criterion 59)............................................................................................3.1-383.1.2.60Control of Releases of Radioactive Materials to the Environment (Criterion 60) 3.1-393.1.2.61Fuel Storage and Handling and Radioactive Control (Criterion 61)........3.1-403.1.2.62Prevention of Criticality in Fuel St orage and Handling (Criterion 62)....3.1-413.1.2.63Monitoring Fuel and Waste Storage (Criterion 63)..................................3.1-413.1.2.64Monitoring Radioactivity Releases (Criterion 64)....................................3.1-423.1.3Reference for Section 3.1..........................................................................3.1-433.2CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS..3.2-13.2.1Seismic Classification.................................................................................3.2-1 MPS-3 FSARCHAPTER 3 -DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYS-TEMS Table of Contents (Continued) 3-iii Rev. 303.2.2System Quality Group Classification.........................................................3.2-13.2.3Quality Assurance Categories.....................................................................3.2-63.2.4Other Classification Systems......................................................................3.2-63.2.5Tabulation of Codes and Classifications....................................................3.2-73.3WIND AND TORNADO LOADINGS...............................................................3.3-1 3.3.1Wind Loadings............................................................................................3.3-13.3.1.1Design Wind Velocity................................................................................3.3-13.3.1.2Determination of Applied Forces...............................................................3.3-13.3.2Tornado Loadings.......................................................................................3.3-13.3.2.1Applicable Design Parameters....................................................................3.3-13.3.2.2Determination of Forces on Structures.......................................................3.3-23.3.2.3Effect of Failure of Structures or Components not Designed for Tornado Loads 3.3-43.3.3References for Section 3.3..........................................................................3.3-43.4WATER LEVEL (FLOOD) DESIGN.................................................................3.4-1 3.4.1Flood Protection..........................................................................................3.4-13.4.1.1Flood Protection Measures for Se ismic Category I Structures...................3.4-13.4.1.2Permanent Dewatering System...................................................................3.4-33.4.2Analytical and Test Procedures..................................................................3.4-33.4.3Reference for Section 3.4............................................................................3.4-33.5MISSILE PROTECTION....................................................................................3.5-13.5.1Missile Selection and Description..............................................................3.5-13.5.1.1Internally Generated Missiles (Outside Containment)...............................3.5-13.5.1.2Internally Generated Missiles (Inside Containment)..................................3.5-33.5.1.2.1Missile Selection and Description..............................................................3.5-33.5.1.2.2Missile Protection Provided........................................................................3.5-63.5.1.3Turbine Missiles.........................................................................................3.5-73.5.1.3.1Turbine Placement and Orientation............................................................3.5-73.5.1.3.2Missiles Identification and Characteristics.................................................3.5-73.5.1.3.3Target Description......................................................................................3.5-73.5.1.3.4Probability Analysis....................................................................................3.5-73.5.1.3.5Turbine Overspeed Protection....................................................................3.5-83.5.1.3.6Turbine Valve Testing................................................................................3.5-83.5.1.4Missiles Generated by Natural Phenomena................................................3.5-83.5.1.5Missiles Generated by Events Near the Site...............................................
3.5-9 MPS-3 FSARCHAPTER 3 -DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYS-TEMS Table of Contents (Continued) 3-iv Rev. 303.5.1.6Aircraft Hazards..........................................................................................3.5-93.5.2Structures, Systems, and Components to be Protected from Externally Generat-ed Missiles..................................................................................................3.5-93.5.3Barrier Design Procedures..........................................................................3.5-93.5.3.1Concrete Barriers........................................................................................3.5-93.5.3.2Steel Barriers.............................................................................................3.5-113.5.3.3Design Evaluation.....................................................................................3.5-113.5.3.4Secondary Missiles...................................................................................3.5-123.5.4References for Section 3.5........................................................................3.5-123.6PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIA TED WITH THE POSTULATED RUPTURES OF PIPING..........................................................3.6-13.6.1Postulated Piping Failures in Fluid Systems Inside and Outside of Containment 3.6-13.6.1.1Design Bases...............................................................................................3.6-13.6.1.1.1Design Basis Protection Criteria.................................................................3.6-13.6.1.1.2Design Basis Pipe Break/Crack Criteria.....................................................3.6-23.6.1.1.3Essential Systems, Components, and Structures........................................3.6-23.6.1.1.4Design Approach........................................................................................3.6-23.6.1.2Description..................................................................................................3.6-33.6.1.3Safety Evaluation........................................................................................3.6-43.6.1.3.1Operability of Essential Systems and Components....................................3.6-43.6.1.3.2Failure Mode and Effects............................................................................3.6-53.6.1.3.3Pipe Break/Crack Analysis.........................................................................3.6-63.6.2Determination of Break Locations and Dynamic Effects Asso ciated with the Postulated Rupture of Piping....................................................................3.6-183.6.2.1Criteria Used to Define Break and Crack Location and Configuration....3.6-183.6.2.1.1Criteria for Inside Containment................................................................3.6-193.6.2.1.2Criteria for Outside Containment.............................................................3.6-203.6.2.1.3Design Basis Break/Crack Types and Orientation...................................3.6-243.6.2.1.4Conformance with Regulatory Guide 1.46...............................................3.6-253.6.2.2Analytical Methods to Define Forcing Functions and Response Models3.6-263.6.2.2.1Introduction...............................................................................................3.6-263.6.2.2.2Time Dependent Blowdown Force...........................................................3.6-273.6.2.2.3Simplified Blowdown Analysis................................................................3.6-333.6.2.2.4Lumped-Parameter Dynamic Analysis.....................................................3.6-343.6.2.2.5Energy Balance Analysis..........................................................................3.6-353.6.2.2.6Local Pipe Indentation..............................................................................3.6-383.6.2.2.7Concrete Barrier Impact............................................................................3.6-393.6.2.3Dynamic Analysis Methods to Verify Integrity and Operability.............
3.6-39 MPS-3 FSARCHAPTER 3 -DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYS-TEMS Table of Contents (Continued) 3-v Rev. 303.6.2.3.1Pipe Rupture Restraints............................................................................3.6-413.6.2.4Guard Pipe Assembly Design Criteria......................................................3.6-423.6.3References for Section 3.6........................................................................3.6-423.7SEISMIC DESIGN..............................................................................................3.7-13.7B.1Seismic Input............................................................................................3.7B-23.7B.1.1Design Response Spectra..........................................................................3.7B-23.7B.1.2Design Time History.................................................................................3.7B-23.7B.1.3Critical Damping Values..........................................................................3.7B-23.7B.1.4Supporting Media for Seismic Category I Structures...............................3.7B-33.7B.2Seismic System Analysis..........................................................................3.7B-33.7B.2.1Seismic Analysis Methods........................................................................3.7B-33.7B.2.2Natural Frequencies and Response Loads................................................3.7B-43.7B.2.2.1Containment Summary.............................................................................3.7B-43.7B.2.2.2Main Steam Valve Building Summary.....................................................3.7B-53.7B.2.2.3Emergency Generator Enclosure Summary..............................................3.7B-53.7B.2.3Procedures Used for Analytical Modeling................................................3.7B-63.7B.2.4Soil-Structure Interaction..........................................................................3.7B-73.7B.2.5Development of Floor Response Spectra..................................................3.7B-73.7B.2.6Three Components of Earthquake Motion................................................3.7B-83.7B.2.7Combination of Modal Responses............................................................3.7B-83.7B.2.8Interaction of Non-Category I Structures with Seismic Category I Structures...
3.7B-83.7B.2.9Effects of Parameter Variations on Floor Response Spectra....................3.7B-83.7B.2.10Use of Constant Vertical Static Factors....................................................3.7B-83.7B.2.11Method Used to Account for Torsional Effects........................................3.7B-83.7B.2.12Comparison of Responses.........................................................................3.7B-93.7B.2.13Methods for Seismic Analysis of Category I Dams.................................3.7B-93.7B.2.14Determination of Seis mic Category I Structure Overturning Moments...3.7B-93.7B.2.15Analysis Procedure for Damping............................................................3.7B-103.7B.3Seismic Subsystem Analysis..................................................................3.7B-123.7B.3.1Seismic Analysis Methods......................................................................3.7B-123.7B.3.1.1Equipment and Components...................................................................3.7B-123.7B.3.1.2Piping Systems........................................................................................3.7B-193.7B.3.2Determination of Number of Earthquake Cycles...................................3.7B-203.7B.3.2.1Equipment and Components...................................................................3.7B-203.7B.3.2.2Piping Systems........................................................................................3.7B-203.7B.3.3Procedures Used for Modeling...............................................................3.7B-213.7B.3.3.1Equipment and Components...................................................................3.7B-213.7B.3.3.2Piping Systems........................................................................................
3.7B-21 MPS-3 FSARCHAPTER 3 -DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYS-TEMS Table of Contents (Continued) 3-vi Rev. 303.7B.3.4Basis for Selection of Frequencies..........................................................3.7B-213.7B.3.4.1Equipment and Components...................................................................3.7B-213.7B.3.4.2Piping Systems........................................................................................3.7B-213.7B.3.5Use of Equivalent Static Load Method of Analysis...............................3.7B-213.7B.3.5.1Equipment and Components...................................................................3.7B-213.7B.3.5.2Piping Systems........................................................................................3.7B-253.7B.3.6Three Components of Earthquake Motion..............................................3.7B-263.7B.3.6.1Equipment and Components...................................................................3.7B-263.7B.3.6.2Piping Systems........................................................................................3.7B-263.7B.3.7Combination of Modal Responses..........................................................3.7B-263.7B.3.8Analytical Procedures for Piping Systems..............................................3.7B-283.7B.3.8.1Natural Frequencies and Mode Shapes...................................................3.7B-293.7B.3.8.2Dynamic Response.................................................................................3.7B-293.7B.3.8.3Response Spectrum Modal Analysis......................................................3.7B-303.7B.3.9Multiply Supported Equipment and Co mponents with Distinct Inputs..3.7B-313.7B.3.10Use of Constant Vertical Static Factors..................................................3.7B-313.7B.3.10.1Equipment and Components...................................................................3.7B-313.7B.3.10.2Piping Systems........................................................................................3.7B-313.7B.3.11Torsional Effects of Eccentric Masses....................................................3.7B-323.7B.3.12Buried Seismic Category I Piping Systems............................................3.7B-323.7B.3.12.1Seismic Wave Effect in the Free Field...................................................3.7B-323.7B.3.12.2Effects of Differential Movements Between Structure and Adjacent Soil Due to Seismic Motion.......................................................................................3.7B-363.7B.3.12.3Effects of Diff erential Movements Due to Structural Settlement...........3.7B-373.7B.3.12.4Accommodations for Buried Piping Structural Penetrations..................3.7B-373.7B.3.13Interaction of Other Systems (Piping and Equipment) with Seismic Category I Systems (Piping and Equipment)............................................................3.7B-373.7B.3.14Seismic Analysis for Reactor Internals...................................................3.7B-383.7B.3.15Analysis Procedure for Damping............................................................3.7B-383.7NSEISMIC DESIGN.......................................................................................3.7N-1103.7N.1Seismic Input.......................................................................................3.7N-1103.7N.1.1Design Response Spectra.....................................................................3.7N-1103.7N.1.2Design Time History............................................................................3.7N-1103.7N.1.3Critical Damping Values.....................................................................3.7N-1103.7N.1.4Supporting Media for Seismic Category I Structures..........................3.7N-1113.7N.2Seismic System Analysis.....................................................................3.7N-1113.7N.3Seismic Subsystem Analysis...............................................................3.7N-1113.7N.3.1Seismic Analysis Methods...................................................................3.7N-1113.7N.3.1.1Dynamic Analysis - Mathematical Model...........................................
3.7N-111 MPS-3 FSARCHAPTER 3 -DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYS-TEMS Table of Contents (Continued) 3-vii Rev. 303.7N.3.1.2Modal Analysis....................................................................................3.7N-1133.7N.3.1.3Response Spectrum Analysis...............................................................3.7N-1163.7N.3.2Determination of Number of Earthquake Cycles................................3.7N-1163.7N.3.3Procedure Used for Modeling..............................................................3.7N-1163.7N.3.4Basis for Selection of Frequencies.......................................................3.7N-1163.7N.3.5Use of Equivalent Static Load Method of Analysis............................3.7N-1173.7N.3.6Three Components of Earthquake Motion...........................................3.7N-1173.7N.3.7Combination of Modal Responses.......................................................3.7N-1173.7N.3.8Analytical Procedures for Piping.........................................................3.7N-1193.7N.3.9Multiply Supported Equipment Components with Distinct Inputs.....3.7N-1193.7N.3.10Use of Constant Vertical Static Factors...............................................3.7N-1193.7N.3.11Torsional Effects of Eccentric Masses.................................................3.7N-1193.7N.3.12Buried Seismic Category I Piping Systems and Tunnels....................3.7N-1193.7N.3.13Interaction of Other Piping with Seismic Category I Piping...............3.7N-1193.7N.3.14Seismic Analyses for Reactor Internals...............................................3.7N-1193.7N.3.15Analysis Procedure for Damping.........................................................3.7N-1203.7N.4Seismic Instrumentation......................................................................3.7N-1203.7.4.1Comparison with Regulatory Guide 1.12...............................................3.7-1223.7.4.2Location and Description of Instrumentation.........................................3.7-1223.7.4.3Control Room Operator Notification......................................................3.7-1233.7.4.4Comparison of Measured and Predicted Responses...............................3.7-1243.7.5References for Section 3.7......................................................................3.7-1243.8DESIGN OF CATEGORY I STRUCTURES.....................................................3.8-1 3.8.1Concrete Containment................................................................................3.8-13.8.1.1Description of the Containment..................................................................3.8-13.8.1.1.1Base Foundation.........................................................................................3.8-23.8.1.1.2Cylindrical Wall..........................................................................................3.8-23.8.1.1.3Dome...........................................................................................................3.8-33.8.1.1.4Steel Liner and Penetrations.......................................................................3.8-33.8.1.1.5Ring Girder.................................................................................................3.8-73.8.1.2Applicable Codes, Standards, and Specifications.......................................3.8-73.8.1.2.1General........................................................................................................3.8-73.8.1.2.2Structural Specifications.............................................................................3.8-83.8.1.2.3Steel Liner and Penetrations.....................................................................3.8-103.8.1.3Loads and Loading Combinations............................................................3.8-113.8.1.3.1Containment Mat, Shell, and Dome..........................................................3.8-113.8.1.3.2Steel Liner and Penetrations.....................................................................3.8-143.8.1.4Design and Analysis Procedures...............................................................3.8-153.8.1.4.1Containment Structure..............................................................................
3.8-15 MPS-3 FSARCHAPTER 3 -DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYS-TEMS Table of Contents (Continued) 3-viii Rev. 303.8.1.4.2Steel Liner and Penetrations.....................................................................3.8-193.8.1.5Structural Acceptance Criteria..................................................................3.8-193.8.1.5.1Containment Structure..............................................................................3.8-193.8.1.5.2Steel Liner and Penetration.......................................................................3.8-203.8.1.6Materials, Quality Control, and Special Construction Techniques..........3.8-203.8.1.6.1Concrete....................................................................................................3.8-203.8.1.6.2Reinforcing Steel......................................................................................3.8-233.8.1.6.3Structural Steel..........................................................................................3.8-263.8.1.6.4Waterproofing Membrane.........................................................................3.8-263.8.1.6.5Steel Liner and Penetrations.....................................................................3.8-273.8.1.6.6Backfill Around Containment Structure...................................................3.8-283.8.1.7Testing and Inservice Surveillance Requirements....................................3.8-283.8.1.7.1Concrete Containment..............................................................................3.8-283.8.2Steel Containment.....................................................................................3.8-303.8.3Concrete and Structural Steel Internal Structures of Steel or Concrete Contain-ments.........................................................................................................3.8-303.8.3.1Description of Internal Structures.............................................................3.8-303.8.3.2Applicable Codes, Standards, and Specifications.....................................3.8-313.8.3.3Loads and Loading Combinations............................................................3.8-313.8.3.4Design and Analysis Procedures...............................................................3.8-323.8.3.5Structural Acceptance Criteria..................................................................3.8-323.8.3.6Materials, Quality Control, and Special Construction Techniques..........3.8-333.8.3.7Testing and Inservice Surveillance Requirements....................................3.8-343.8.4Other Seismic Category I Structures (and Major Nons afety Related Structures) 3.8-343.8.4.1Description of the Structures....................................................................3.8-343.8.4.2Applicable Codes, Standards, and Specifications.....................................3.8-433.8.4.3Loads and Loading Combinations............................................................3.8-433.8.4.4Design and Analysis Procedures...............................................................3.8-443.8.4.5Structural Acceptance Criteria..................................................................3.8-443.8.4.6Materials, Operating Control, and Special Construction Techniques......3.8-463.8.4.7Testing and Inservice Surveillance Requirements....................................3.8-463.8.4.8Masonry Walls..........................................................................................3.8-463.8.5Foundations...............................................................................................3.8-463.8.5.1Description of the Foundations.................................................................3.8-463.8.5.2Applicable Codes, Standards, and Specifications.....................................3.8-473.8.5.3Loads and Loading Combinations............................................................3.8-483.8.5.4Design and Analysis Procedures...............................................................3.8-483.8.5.5Structural Acceptance Criteria..................................................................3.8-493.8.5.6Materials, Quality Control, and Special Construction Techniques..........3.8-493.8.5.7Testing and Inservice Surveillance Requirements....................................
3.8-50 MPS-3 FSARCHAPTER 3 -DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYS-TEMS Table of Contents (Continued) 3-ix Rev. 303.8.6References for Section 3.8........................................................................3.8-503.9MECHANICAL SYSTEMS AND COMPONENTS..........................................3.9-1 3.9B.1Special Topics for Mechanical Components............................................3.9B-23.9B.1.1Design Transients.....................................................................................3.9B-23.9B.1.2Computer Programs Used in Analysis......................................................3.9B-33.9B.1.3Experimental Stress Analysis...................................................................3.9B-33.9B.1.4Consideration for the Evaluation of the Faulted Conditions....................3.9B-33.9B.1.4.1Loading Conditions...................................................................................3.9B-33.9B.1.4.2Evaluation of Reactor Coolant Loop and Supports for Faulted Loading Condi
-tion............................................................................................................3.9B-43.9B.1.4.3Reactor Coolant Loop Models and Methods............................................3.9B-53.9B.1.4.4Primary Component Supports Models and Methods................................3.9B-83.9B.1.4.5Equipment and Components.....................................................................3.9B-93.9B.2Dynamic Testing and Analysis...............................................................3.9B-103.9B.2.1Preoperational Vibration and Dynamic Effects Testing on Piping.........3.9B-103.9B.2.2Seismic Qualification Testing of Safety Related Mechanical Equipment..3.9B-113.9B.3ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures...................................................................................3.9B-123.9B.3.1Loading Combinations, Design Transients, and Stress Limits...............3.9B-123.9B.3.1.1ASME III Class 1 Components..............................................................3.9B-123.9B.3.1.2ASME III Class 2 and 3 Components.....................................................3.9B-123.9B.3.2Pump and Valve Operability Assurance.................................................3.9B-133.9B.3.2.1Pump Operating Program.......................................................................3.9B-133.9B.3.2.2Valve Operability Program.....................................................................3.9B-163.9B.3.3Design and Installation Details for Mounting of Pressure Relief Devices.3.9B-183.9B.3.3.1Open Relief System................................................................................3.9B-193.9B.3.3.2Closed Relief System..............................................................................3.9B-203.9B.3.4Component Supports...............................................................................3.9B-203.9NMECHANICAL SYSTEMS AND COMPONENTS.....................................3.9N-273.9N.1Special Topics for Mechanical Components.........................................3.9N-273.9N.1.1Design Transients..................................................................................3.9N-273.9N.1.2Computer Programs Used in Analyses..................................................3.9N-423.9N.1.3Experimental Stress Analysis................................................................3.9N-423.9N.1.4Considerations for the Evaluation of the Faulted Condition.................3.9N-433.9N.1.4.1Loading Conditions................................................................................
3.9N-43 MPS-3 FSARCHAPTER 3 -DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYS-TEMS Table of Contents (Continued) 3-x Rev. 303.9N.1.4.2Analysis of Primary Components..........................................................3.9N-433.9N.1.4.3Dynamic Analysis of Reactor Pressure Vessel for Postulated Loss-of-Coolant Accident.................................................................................................3.9N-443.9N.1.4.4Stress Criteria for Reactor Coolant System Components......................3.9N-473.9N.2Dynamic Testing and Analysis..............................................................3.9N-473.9N.2.1Preoperational Vibration and Dynamic Effects Testing on Piping........3.9N-473.9N.2.2Seismic Qualification Testing of Safety Related Mechanical Equipment.3.9N-473.9N.2.3Dynamic Response Analysis of React or Internals Under Operational Flow Transients and Steady State Conditions.................................................3.9N-473.9N.2.4Preoperational Flow-Induced Vibration Testing of Reactor Internals...3.9N-493.9N.2.5Dynamic System Analysis of the Reactor Internals Under Faulted Conditions..
3.9N-513.9N.2.6Correlations of Reactor Internals Vibration Tests with the Analytical Results...
3.9N-553.9N.3ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures..................................................................................3.9N-553.9N.3.1Loading Combinations, Design Transi ents, and Stress Limits for Class 2 Components...........................................................................................3.9N-553.9N.3.1.1Design Loading Combinations..............................................................3.9N-553.9N.3.1.2Design Stress Limits..............................................................................3.9N-553.9N.3.2Pumps and Valve Operability Assurance..............................................3.9N-563.9N.3.2.1ASME Code Class Pumps.....................................................................3.9N-563.9N.3.2.2Valve Operability...................................................................................3.9N-563.9N.3.2.3Pump Motor and Valve Operator Qualification....................................3.9N-583.9N.3.3Design and Installation Details for Mounting of Pressure-Relieving Devices....
3.9N-583.9N.3.4Component Supports..............................................................................3.9N-583.9N.3.4.1Component Supports for Tanks and Heat Exchangers..........................3.9N-583.9N.3.4.2Component Supports for Pumps............................................................3.9N-593.9N.4Control Rod Drive System (CRDS).......................................................3.9N-593.9N.4.1Descriptive Information of CRDS.........................................................3.9N-593.9N.4.2Applicable CRDS Design Specifications..............................................3.9N-643.9N.4.3Design Loads, Stress Limits, and Allowable Deformations..................3.9N-653.9N.4.3.1Pressure Vessel Assembly.....................................................................3.9N-653.9N.4.3.2Drive Rod Assembly..............................................................................3.9N-663.9N.4.3.3Latch Assembly and Coil Stack Assembly............................................3.9N-663.9N.4.4CRDS Performance Assurance Program...............................................3.9N-683.9N.5Reactor Vessel Internals........................................................................3.9N-693.9N.5.1Design Arrangements............................................................................3.9N-693.9N.5.2Design Loading Conditions...................................................................
3.9N-74 MPS-3 FSARCHAPTER 3 -DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYS-TEMS Table of Contents (Continued) 3-xi Rev. 303.9N.5.3Design Loading Categories....................................................................3.9N-753.9N.5.4Design Bases..........................................................................................3.9N-763.9N.6Inservice Testing of Pumps and Valves.................................................3.9N-77 3.9.7Inservice Testing of Pumps and Valves....................................................3.9-903.9.7.1Inservice Testing of Pumps.......................................................................3.9-903.9.7.2Inservice Testing of Valves......................................................................3.9-903.10SEISMIC QUALIFICATION OF SEISMIC CATEGOR Y I INSTRUMENTATION AND ELECTRICAL EQUIPMENT.................................................................3.10-13.10B.1Seismic Qualification Criteria................................................................3.10B-23.10B.2Methods and Procedures for Qualifying Electrical Equipmen t and Instrumenta-tion..........................................................................................................3.10B-23.10B.3Methods and Procedures of Analysis or Testing of Supports of Electrical Equip-ment and Instrumentation.......................................................................3.10B-33.10B.4Replacement Items..................................................................................3.10B-33.10B.5Operating License Review......................................................................3.10B-43.10NSEISMIC QUALIFICATION OF SEISMIC CATEGOR Y I INSTRUMENTATION AND ELECTRICAL EQUIPMENT..............................................................3.10N-53.10N.1Seismic Qualification Criteria...............................................................3.10N-53.10N.1.1Qualification Standards..........................................................................3.10N-53.10N.1.2Performance Requirements for Seismic Qualification..........................3.10N-53.10N.1.3Acceptance Criteria................................................................................3.10N-63.10N.2Methods and Procedures for Qualifying Electrical Equipmen t and Instrumenta-tion.........................................................................................................3.10N-63.10N.2.1Seismic Qualification by Type Test.......................................................3.10N-63.10N.2.2Seismic Qualification by Analysis.........................................................3.10N-73.10N.3Method and Procedures for Qualifying Supports of Electrical Equipment and Instrumentation......................................................................................3.10N-83.10N.4Operating License Review.....................................................................3.10N-83.10.1Replacement Items....................................................................................3.10-9 3.10.2References for Section 3.10......................................................................3.10-93.11ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRICAL EQUIPMENT....................................................................................................3.11-13.11B.1Equipment Identification and Environmental Conditions......................3.11B-33.11B.1.1Environmental Conditions......................................................................3.11B-33.11B.1.2Equipment Identification........................................................................
3.11B-4 MPS-3 FSARCHAPTER 3 -DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYS-TEMS Table of Contents (Continued) 3-xii Rev. 303.11B.2Qualification Tests and Analyses...........................................................3.11B-43.11B.2.1Regulatory Guides..................................................................................3.11B-43.11B.2.2Safety-Related (Class 1E) Equipmen t and Component Qualifications..3.11B-53.11B.3Qualification Test Results.......................................................................3.11B-83.11B.3.1Nuclear Steam Supply System Equipment Qualification Program........3.11B-83.11B.3.2Qualification of Safety-Related Equipment (BOP) not covered by Section 3.11.
3.11B-83.11B.4Loss of Ventilation..................................................................................3.11B-83.11B.5Chemical and Radiation Environment....................................................3.11B-93.11B.5.1Radiation Environment...........................................................................3.11B-93.11B.5.2Chemical Environment.........................................................................3.11B-103.11NENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRICAL EQUIPMENT...............................................................................................3.11N-113.11N.1Equipment Identification and Environmental Conditions...................3.11N-113.11N.2Qualification Tests and Analysis.........................................................3.11N-113.11N.2.1Environmental Qualification Criteria..................................................3.11N-113.11N.2.2Performance Requirements for Environmental Qualification.............3.11N-113.11N.2.3Methods and Procedures for Environmental Qualification.................3.11N-123.11N.3Qualification Test Results....................................................................3.11N-123.11N.4Loss of Ventilation...............................................................................3.11N-123.11N.5Estimated Chemical and Radiation Environment................................3.11N-123.11.6References for Section 3.11....................................................................3.11-13APPENDIX 3A -COMPUTER PROGRAMS FOR DYNAMIC AND STATIC ANALYSIS OF SEISMIC CATEGORY I STRUCTURES, EQUIPMENT, AND COMPO-NENTS.....................................................................................................3A.1-03A.1STRUCTURES........................................................................................3A.1-13A.1.1STRUDL II..............................................................................................3A.1-13A.1.2SHELL 1..................................................................................................3A.1-2 3A.1.3STRUDL-SW...........................................................................................3A.1-33A.1.4ASAAS (Asymmetric Stress Analysis of Axisymmetric Solids)............3A.1-33A.1.5TAC2D (A General Purpose Two-Dimensional Heat Transfer Computer Code) 3A.1-53A.1.6Time History Program.............................................................................3A.1-73A.1.7PLAXLY..................................................................................................3A.1-83A.1.8MAT5 (Circular Mat with Axisymmetric Loading)................................3A.1-83A.1.9ANSYS....................................................................................................3A.1-8 3A.1.10MEMBRANE (Membrane Stress Analysis)............................................3A.1-93A.1.11SBMMI (Single Barrier Mass Missile Impact)........................................3A.1-93A.1.12GHOSH-WILSON.................................................................................3A.1-10 MPS-3 FSARCHAPTER 3 -DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYS-TEMS Table of Contents (Continued) 3-xiii Rev. 303A.1.13References for Appendix 3A.1..............................................................3A.1-113A.2EQUIPMENT AND COMPONENTS.....................................................3A.2-13A.2.1ASAAS (Asymmetric Stress Analysis of Axisymmetric Solids)............3A.2-23A.2.2Limita 25 - 2D Nonlinear Transient Dynamic Analysis..........................3A.2-33A.2.3MISSILE..................................................................................................3A.2-53A.2.4SLOSH.....................................................................................................3A.2-6 3A.2.5LION (ME-112).......................................................................................3A.2-73A.2.6LIMITA 3................................................................................................3A.2-73A.2.7TAC2D (A General Purpose, Two-Dimensional Heat Transfer Computer Code) 3A.2-93A.2.8SHELL 1..................................................................................................3A.2-93A.2.9Vessel Penetration Analysis.....................................................................3A.2-93A.2.10DINASAW (Dynamic Inelastic Nonlinear Analysis by Stone and Webster).....
3A.2-103A.2.11LIMITA 2..............................................................................................3A.2-11 3A.2.12STARDYNE..........................................................................................3A.2-143A.2.13ASYMPR (ME-171)..............................................................................3A.2-153A.2.14LIDOP (ME-184)...................................................................................3A.2-16 3A.2.15Dynamic Load Factors (DLF ME-185).................................................3A.2-173A.2.16References for Appendix 3A.2..............................................................3A.2-193A.3PIPING SYSTEMS..................................................................................3A.3-13A.3.1NUPIPE II................................................................................................3A.3-13A.3.2PITRUST.................................................................................................3A.3-33A.3.3PILUG......................................................................................................3A.3-3 3A.3.4SAVAL....................................................................................................3A.3-43A.3.5STEHAM.................................................................................................3A.3-53A.3.6WATHAM...............................................................................................3A.3-5 3A.3.7PSPECTRA..............................................................................................3A.3-63A.3.8STRUDL-SW...........................................................................................3A.3-73A.3.9STRUDL-II (ICES)..................................................................................3A.3-9 3A.3.10APEN.....................................................................................................3A.3-103A.3.11PITRIFE.................................................................................................3A.3-113A.3.12PITAB....................................................................................................3A.3-12 3A.3.13CHPLOT................................................................................................3A.3-123A.3.14LOADCOMB.........................................................................................3A.3-133A.3.15BEARST................................................................................................3A.3-143A.3.16ANCCOMB...........................................................................................3A.3-143A.3.17BENDCORD.........................................................................................3A.3-16 3A.3.18NUPIPE-SWPC.....................................................................................3A.3-163A.3.19PC-PREPS.............................................................................................3A.3-173A.3.20PILUG-PC.............................................................................................3A.3-17 MPS-3 FSARCHAPTER 3 -DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYS-TEMS Table of Contents (Continued) 3-xiv Rev. 303A.3.21References for Appendix 3A.3..............................................................3A.3-18 APPENDIX 3B -ENVIRONMENTAL DESIGN CONDITIONS..................................3B.B-0 MPS-3 FSAR 3-xv Rev. 30CHAPTER 3-DESIGN OF STRUCTURES , COMPONENTS, EQUIPMENT, AND SYSTEMS List of Tables Number Title3.2-1List Of QA Category I and Seismic Category I Structures, Systems, and Components3.2-2Safety Classes 2 and 3 In strument Tubing Requirements3.2-3Instrument Tubing Exam ination and Testing Program3.2-4Comparison of Proposed Tubing Examin ation and Testing with ASME III Requirements3.3-1Structural Panels Subject to Tornado Pressure Drop3.5-1Safety Related Structures, Systems, and Components Outside Containment Required for Safe Reactor Shutdown*3.5-2Safety Related Structures, Systems, and Components Inside Containment Required for Safe Reactor Shutdown*3.5-3Summary of Control Rod Drive Mechanism Missile Analysis3.5-4Valve-Missile Characteristics*3.5-5Piping Temperature Element Assembly - Missile Characteristics3.5-6Characteristics of Other Missiles Postulated within Reactor Containment3.5-7Turbine-Target Distances and Impact Areas3.5-8Deleted by FSARCR 02-MP3-133.5-9Deleted by FSARCR 02-MP3-133.5-10Deleted by FSARCR 02-MP3-133.5-11Deleted by FSARCR 02-MP3-133.5-12Deleted by FSARCR 02-MP3-133.5-13Postulated Tornado-Generated External Missiles3.5-14Barrier Deflection and Ductility Ratios for Tornado-Borne Missiles Plus 360 MPH Tornado Wind3.5-15Barrier Deflection and Du ctility Ratio for a Beam-Column Plus 360 MPH Tornado Wind3.6-1High-Energy Systems Remote from E ssential Structures, Systems, and Components MPS-3 FSAR List of Tables (Continued)
Number Title 3-xvi Rev. 303.6-2High-energy Systems in Proximity to Essential Structures, Systems, and Components3.6-3Moderate-Energy Systems in Remote fro m Essential Structures, Systems, and Components3.6-4Moderate-Energy Systems Proximate to Essential Structures, Systems, and Components3.6-5Essential Structures, Systems, and Components Required for Safe Reactor Shutdown3.6-6Postulated Breaks Main Steam System3.6-7Pipe Whip Effect s Main Steam System3.6-8Jet Impingement Effects - Main Steam System3.6-9Postulated Breaks Main Feedwater System3.6-10Pipe Whip Effects -
Main Feedwater System3.6-11Jet Impingement Effects - Main Steam System3.6-12Postulated Breaks - Reactor Coolan t System - Loop Stop Valve Bypass Piping, Excess Letdown Piping, Loop Fill Piping, Loop Drain Piping. Letdown Line, and Normal Charging3.6-13Pipe Whip Effects - React or Coolant System - Loop Stop Valve Bypass Piping, Excess Letdown Piping, Loop Fill Piping, Loop Drain Pipi ng, Letdown and Normal Charging3.6-14JET Impingement Effects - Reactor Co olant System - Loop Stop Valve Bypass Piping, Excess Letdown Piping, Loop Drai n Piping, Loop Fill Piping, Letdown Line, and Normal Charging3.6-15Postulated Breaks - Reactor Coolan t System - Pressuri zer Cubicle Piping3.6-16Pipe Whip Effects - R eactor Coolant System - Pr essurizer Cubicle Piping3.6-17Jet Impingement Effects - Reactor Cool ant System - Pressurizer Cubicle Piping3.6-18Postulated Pipe Breaks - Reactor Coolant System - Low Pressure Safety Injection, High Pressure Safety Injection, and Residual Heat Removal Piping3.6-19Pipe Whip Effects - Reactor Coolant System - Low Pressure Safety Injection, High Pressure Safety Injection, a nd Residual Heat Removal Piping3.6-20Jet Impingement Effects - Low Pressure Safety Injection, Hi gh Pressure Safety Injection, and Residual Heat Removal Piping MPS-3 FSAR List of Tables (Continued)
Number Title 3-xvii Rev. 303.6-21Postulated Breaks - Chemical Volu me Control System - Normal Charging3.6-22Pipe Whip Effects - Ch emical Volume Control System - Normal Charging3.6-23Jet Impingement Effects - Chemical Vo lume Control System - Normal Charging3.6-24Postulated Breaks - Chemical Vo lume Control System - Letdown Line3.6-25Pipe Whip Effects -
Chemical Volume Control System - Letdown Line3.6-26Jet Impingement Effects -
Chemical and Volume Contro l System - Letdown Line Piping3.6-27Postulated Breaks - Se al Water Injection System3.6-28Pipe Whip Effects - Chemical Volume Control System - Seal Water Injection Line3.6-29Jet Impingement Effects - Chemical Volume Control System - Seal Water Injection Line3.6-30Postulated Breaks - Steam Generator Blowdown System3.6-31Pipe Whip Effects - Stea m Generator Blowdown System3.6-32Jet Impingement Effects - St eam Generator Blowdown System3.6-33Energy Absorbing Capacity of a 4 Inch Schedule 80 Pipe3.6-34Postulated Breaks Auxiliary Feedwater System3.6-35Pipe Whip Effects A uxiliary Feedwater System3.6-36Jet Impingement Effects A uxiliary Feedwater Systems3.7-1Seismic Instrumentation3.7B-1Damping Factors3.7B-2Methods of Seismic Analysis Us ed for Seismic Category I Structures3.7B-3Containment and Internal Structur es Significant Modal Frequencies and Participation Factors Uncracked Model3.7B-4Containment and Internal Structur es Significant Modal Frequencies and Participation Factors Cracked Model3.7B-5Containment and Internal Structures Significant Mode Shapes (Normalized Eigenvectors) Uncracked Model3.7B-6Containment and Internal Structures Significant Mode Shapes (Normalized Eigenvectors) Cracked Model MPS-3 FSAR List of Tables (Continued)
Number Title 3-xviii Rev. 303.7B-7Containment and Internal Structures SRSS Accelerations a nd Displacements for Safe Shutdown Earthquake Uncracked Mo del, Response Spectrum Analysis3.7B-8Containment and Internal Structures SRSS Accelerations a nd Displacements for Safe Shutdown Earthquake Cracked M odel, Response Spectrum Analysis3.7B-9Containment and Internal Structures Degrees of Freedom3.7B-10Containment and Internal Structure E nveloped Accelerations and Displacements Horizontal-motion *3.7B-11Containment and Internal Structure E nveloped Accelerations and Displacements Vertical-Motion *3.7B-12Main Steam Valve Building Significan t Modal Frequencies and Participation Factors3.7B-13Main Steam Valve Building Signi ficant Mode Shapes (Eigenvectors)3.7B-14Main Steam Valve Building CSM* Acce lerations for Safe Shutdown Earthquake from Response Spectrum Analysis3.7B-15Main Steam Valve Building CSM* Displacements for Safe Shutdown Earthquake from Response Spectrum Analysis3.7B-16Main Steam Valve Building ABS
* of Safe Shutdown Earthquake Responses3.7B-17Main Steam Valve Building Degrees of Freedom3.7B-18Emergency Generator Enclosure Accele rations and Displacem ents Horizontal-Motion*3.7B-19Emergency Generator Enclosure Accele rations and Displacem ents Vertical -
Motion3.7B-20Comparison of Response Spectra and Time History Analysis Results Containment and Internal Structures (Uncracked Properties)3.7B-21Comparison of Response Spectra  and Time History Analysis Results Containment and Internal Structures (Cracked Properties)3.7B-22Comparison of Response Spectra and Time History Analysis Results Main Steam Valve Building SSE Accelerations*3.7B-231 G Flat Response3.7B-24Modal Density, n*3.7B-25Amplified Response Dynamic Factor Study3.7B-26Piping System Seismic Design and Analysis Criteria MPS-3 FSAR List of Tables (Continued)
Number Title 3-xix Rev. 303.7N-1Damping Values Used for Seismic Analysis for Westinghous e Supplied Equipment3.8-1Loading Conditions - Liner Plate and Access Openings3.8-2Loading Conditions, Penetrations3.8-3Loads and Loading Combinations3.8-4Load Combinations for ASME III Cl ass 2 Penetrations except Quench, Recirculation, and Safe ty Injection Piping3.8-5Load Combinations for ASME III Class 2 Penetrations for the Quench Spray, Recirculation Spray, and Sa fety Injection Systems3.8-6Load Combinations for ASME II I Class MC Sleeved Penetrations3.8-7Nomenclature for Tables 3.8-4 through 3.8-63.9B-1List of Input Documents Describing Design Transient for Fatigue Analysis of RCP and Associated Class 1 Piping3.9B-2List of MPS-3 Documents Describi ng Components Requiring Inelastic Analysis3.9B-3Preoperational Tests (1)3.9B-4Omitted3.9B-5Stress Limits for ASME Section III Class 1 (NB)
Seismic Category I Components (Elastic Analysis)3.9B-6Comparison of Class 1 Re quirements Regulatory Guide 1.48 vs. Tables 3.9B-5 and 3.9B-103.9B-7Stress Limits for ASME Section III Class 2 and 3 Com ponents (Elastic Analysis)3.9B-8Comparison of Classes 2 and 3 Requi rements Regulatory Guide 1.48 vs Table 3.9B-73.9B-9Stress Limits for ASME Section II I Class 1, 2, and 3 Component Supports *3.9B-9A Load Combinations for ASME Sectio n III Class 1, 2, and 3 Component Supports3.9B-10 ASME III Class 1 Stress and Fa tigue Analysis Requi rements per NB36503.9B-11ASME III Class 2 and 3* Stress An alysis Requirements per NC3650 and ND36503.9B-12Loadings Applicable to Piping Systems3.9B-13Active Pumps and Valves3.9B-14Omitted3.9B-15Postulated Primary Loop Breaks MPS-3 FSAR List of Tables (Continued)
Number Title 3-xx Rev. 303.9B-16Steam Generator and Reactor Coolant Pump Support Snubber Embedment Loads (1) (KIPS)3.9B-17Steam Generator and Reactor Cool ant Pump Support Column Embedment Loads(1)(KIPS)3.9B-18Factors of Safety for Primary Members of Steam Generators and Reactor Coolant Pump Supports3.9B-19Minimum Design Margins for Pressurizer Support3.9B-20Design Margins for the Primary Members of the Pressurizer Safety Valve Support3.9B-21Design Margins for Primary Members of Reactor Pressure Vessel Support3.9B-22Design Margins for Primary L oop Bumper Support Structure3.9N-1Summary of Reactor Coolant System Design Transients3.9N-2Loading Combinations for R eactor Coolant System Components3.9N-3Allowable Stresses for Reac tor Coolant System Components3.9N-4Design Loading Combinations for AS ME Code Class 2 and 3 Components and Supports3.9N-5Stress Criteria fo r Safety Code Class 2 (1) and Class 3 Tanks3.9N-6Stress Criteria for ASME Code Class 2 Tanks (1)3.9N-7Stress Criteria for ASME Code Class 2 and Class 3 Inactive Pumps3.9N-8Stress Criteria for Safety Related AS ME Code Class 2 and Class 3 Active and Inactive Valves3.9N-9Stress Criteria for ASME Code Class 2 and 3 Piping3.9N-10Design Criteria for Active Pumps3.9N-11Active Pumps3.9N-12Active Valves3.9N-13Maximum Deflections Allowed for Reactor Internal Support Structures3.10N-1Seismic Category I Instrumentation a nd Electrical Equipment in Westinghouse NSSS Scope of Supply3.11N-1Safety-Related Equipment in Westinghouse NSSS Scope of Supply (Historical, not subject to future updating)3.A.1.2-1Thin-Wall Cylinder, Pertinent Parameters MPS-3 FSAR List of Tables (Continued)
Number Title 3-xxi Rev. 303.A.1.2-2Exact and Computer Stre sses for Thin-wall Cylinder3.A.1.4-1Infinitely Long Solid Cy linder, Pertinent Parameters3.A.1.5-1Input Thermal Parameter Func tions for TAC2D Sample Problem3.A.1.6-1Peak Acceleration and Displacement3.A.1.6-2Horizontal Amplif ied Response Spectra3.A.1.10-1Pertinent Parameters of a Cylindrical Shell3.A.1.10-2Summary of Results3.A.1.11-1Test Problem Data3.A.1.11-2A Comparison of Hand a nd Computer Program Results3.A.2.1-1Infinitely Long Solid Cy linder, Pertinent Parameters3.A.2.4-1Comparison of Results of SLOSH vs AEC Analysis*3.A.2.6-1Comparison of Experimental Data wi th Analytical Data Using LIMITA 33A.3-1Comparison of Support Reaction Due to Thermal, Anchor Movement, and External Force Loading3A.3-2Comparison of Deflections and Rotations Due to Thermal, Anchor Movement, and External Force Loading3A.3-3Comparison of Stress Due To Thermal, Anchor Movement, and External Force Loading3A.3-4Comparison of Internal Forc es Due To Deadweight Analysis3A.3-5Comparison of Deflections and Ro tation Due to Deadweight Analysis3A.3-6Comparison of Stresses Due to Deadweight Analysis3A.3-7Comparison of Natural Frequencies3A.3-8Comparison of Natural frequencies3A.3-9Comparison of Class 1 Pipe Stress Analysis3A.3-10Individual Pair Usage Factor for Point No. 303A.3-11Comparison of PITRUST With Franklin Institute Program CYLNOZ and Hand Calculation3A.3-12Comparison of PITRUST With Reference 8 Results3A.3-13Comparison of PILUG Computer Pr ogram Output with Hand Calculations MPS-3 FSAR List of Tables (Continued)
Number Title 3-xxii Rev. 303A.3-14Summary of Comparison of SAVAL Com puter Output with Hand Calculation As-Designed Condition3A.3-15Summary of Comparison of SAVAL Computer Output with Hand Calculation Reinforced Condition (1 1/4" Pad)3A.3-16Nodal Force Comparison3A.3-17Input Data for WATHAM3A.3-18Comparison of NODAL Force Calculation at Time = 2.34 Seconds3A.3-19Comparison of PITRIFE Computer Pr ogram Output With STRUDL-II Output3A.3-20Comparison of PITRIFE Computer Pr ogram Output With Hand Calculations3A.3-21BENDCORD Program--Verification Problem3A.3-22Pipe 1, Segment 1 Force Versus Time for Run Number S2807011 MPS-3 FSARNOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.
3-xxiii Rev. 30CHAPTER 3 - DESIGN OF STRUCTURE S, COMPONENTS, EQUIPMENT, AND SYSTEMS List of Figures Number Title3.2-1Line Designation, Sa fety Class Boundaries3.3-1Dynamic Load Factor Curve for Tornado Pressure Drop3.4-1Circulating and Service Water Pumphouse 3.5-1Turbine Placement and Orient ation for Three-Unit Site3.5-2Turbine Placement and Orientation Profile3.5-3Deleted 3.5-4Deleted3.5-5Deleted3.5-6Missile Barrier Beam Column 3.6-1Fluid Systems Subjected to Break and Crack Analysis3.6-2Relative Locations of Safety Relate d Equipment, Elevation 3 feet 8 inches3.6-3Relative Locations of Safety Relate d Equipment, Elevation 24 feet 6 inches3.6-4Relative Locations of Safety Relate d Equipment, Elevation 51 feet 4 inches3.6-5Relative Locations of Safety Related Equipment, Section 1-13.6-6Relative Locations of Safety Related Equipment, Section 2-23.6-7Relative Locations of Safety Rela ted Equipment, Sections 3-3, 4-43.6-8Main Steam System Notes to Figure 3.6-8: Data for Pipe R upture Restraints - Main Steam System3.6-9Main Steam System, (Continued)3.6-10Feedwater System Postulat ed Pipe Rupture Locations3.6-11Feedwater System Pipe Rupture Restraints Notes to Figure 3.6-11: Data for Pipe Rupture Restraints - Main Feedwater System3.6-12Reactor Coolant System - Primary Coolant Piping Notes to Figure 3.6-12: Data for Pipe Rupture Restraints - Reactor Coolant System - Primary Coolant Piping3.6-13Reactor 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 Pressu re Safety Injection, and Residual Heat MPS-3 FSAR List of Figures (Continued)NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.
Figure Title 3-xxiv Rev. 30 Removal Piping3.6-14Reactor Coolant System -
Pressurizer Cubicle Piping Notes to Figure 3.6-14: Data for Pipe R upture Restraints Reactor Coolant System Pressurizer Cubicle Piping3.6-15Chemical and Volume Control System - Charging Line Piping Notes to Figure 3.6-15: Data for Pipe Rupture Restraints Chemical Volume Control System - Normal Charging3.6-16Chemical and Volume Control System - Letdown Line Piping Notes to Figure 3.6-16: Data for Pipe Rupture Restraints Chemical Volume Control System - Letdown Line3.6-17Chemical and Volume Control System - Seal Water Injection and Piping3.6-18Steam Generator Blowdown System Notes to Figure 3.6-18: Data for Pipe Rupture Restraints Steam Generator Blowdown System3.6-19Pipe Rupture Analysis Flow Chart 3.6-20Variation of Steady State Steam Blowdown vs Friction3.6-21Pipe Restraint Intermediate Stru cture System, Mathematical Model3.6-22Laminated Strap Restraint for Small Lines 3.6-23Feedwater System Short Loop Strap near the Steam Generator, NET Shear Force at Node 103.6-24Feedwater System Short Loop Strap near the Steam Generator, NET Shear Force at Node 143.6-25Feedwater System Short Loop Strap near the Steam Generator, NET Moment Force at Node 143.6-26Force Acting on the Strap Node 83.6-27Feedwater System Short Loop Strap near the Steam Generator, Energy3.6-28Energy Balance Analysis Model 3.6-29Pipe Crush Bumper 3.6-30Pipe Crush Bumper3.6-31Omni Directional Pi pe Rupture Restraint3.6-32Auxiliary Feedwater System 3.6-33Reactor Coolant System Small Bore Piping in Steam Generator Cubicles MPS-3 FSAR List of Figures (Continued)NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.
Figure Title 3-xxv Rev. 303.7B-1Horizontal Design Response Spectra 0.17g SSE3.7B-2Horizontal Design Response Spectra 0.09g 1/2 OBE 3.7B-3Horizontal Time History Respons e Spectra 0.5 Percent Damping Value3.7B-4Horizontal Time History Respons e Spectra 1.0 Percent Damping Value3.7B-5Horizontal Time History Respons e Spectra 2.0 Percent Damping Value3.7B-6Horizontal Time History Respons e Spectra 5.0 Percent Damping Value3.7B-7Horizontal Time History Respons e Spectra 7.0 Percent Damping Value3.7B-8Horizontal Time History Respons e Spectra 10 Percent Damping Value3.7B-9Dynamic Model of the Containment Structure3.7B-10Dynamic Model of the Main Steam Valve Building3.7B-11Dynamic Model of the Emerge ncy Generator Enclosure N-S View3.7B-12Dynamic Model of the Emerge ncy Generator Enclosure E-W View3.7B-13Reactor Containment Mat Elevation -37 feet 3 inches Amplified Response Spectra N-S Excitation SSE3.7B-14Reactor Containment Mat Elevation -37 feet 3 inches Amplified Response Spectra E-W Excitation SSE3.7B-15Reactor Containment Mat Elevation -37 feet 3 inches Amplified Response Spectra Vertical Excitation SSE3.7B-16Reactor Containment Steam Generators Support Slab Elevation 3 feet 0 inches Amplified Response Spectra N-S Excitation SSE3.7B-17Reactor Containment Steam Generators Support Slab Elevation 3 feet 0 inches Amplified Response Spec tra E-W Excitation SSE3.7B-18Reactor Containment Steam Generators Support Slab Elevation 3 feet 0 inches Amplified Response Spectr a Vertical Excitation SSE3.7B-19Reactor Containment Top of Primary Sh ield Wall Elevation 24 feet 6 inches Amplified Response Spectra N-S Excitation SSE3.7B-20Reactor Containment Top of Primary Shie ld Wall El 24 feet 6 inches Amplified Response Spectra E-W Excitation SSE3.7B-211 Reactor Containment Top of Primary Sh ield Wall Elevation 24 feet 6 inches Amplified Response Spectr a Vertical Excitation SSE3.7B-22Reactor Containment Operation Floor Elevation 50 feet 10 inches Amplified MPS-3 FSAR List of Figures (Continued)NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.
Figure Title 3-xxvi Rev. 30 Response Spectra N-S Excitation SSE3.7B-23Reactor Containment Operating Floor Elevation 50 feet 10 inches Amplified Response Spectra E-W Excitation SSE3.7B-24Reactor Containment Operating Floor Elevation 50 feet 10 inches Amplified Response Spectra Vertical Excitation3.7B-25Reactor Containment T op of Crane Wall Elevation 109 feet 1 inch Amplified Response Spectra N-S Excitation SSE3.7B-26Reactor Containment T op of Crane Wall Elevation 109 feet 1 inch Amplified Response Spectra E-W Excitation SSE3.7B-27Reactor Containment T op of Crane Wall Elevation 109 feet 1 inch Amplified Response Spectra Vertical Excitation SSE3.7B-28Reactor Containment Springline Eleva tion 104 feet 0 inches Amplified Response Spectra N-S Excitation SSE3.7B-29Reactor Containment Springline Eleva tion 104 feet 0 inches Amplified Response Spectra E-W Excitation SSE3.7B-30Reactor Containment Springline Eleva tion 104 feet 0 inches Amplified Response Spectra Vertical Excitation3.7B-31Reactor Containment Dome Apex Elevation 163 feet 4 inches Amplified Response Spectra N-S Excitation SSE3.7B-32Reactor Containment Dome Apex Elevation 163 feet 4 inches Amplified Response Spectra E-W Excitation SSE3.7B-33Reactor Containment Dome Apex Elevation 163 feet 4 inches Amplified Response Spectra Vertical Excitation SSE3.7B-34Main Steam Valve Building Mat Elevati on 9 feet -0 inches Amplified Response Spectra N-S Excitation SSE3.7B-35Main Steam Valve Building Mat Elevat ion 9 feet 0 inches Amplified Response Spectra E-W Excitation SSE3.7B-36Main Steam Valve Building Mat Elevat ion 9 feet 0 inches Amplified Response Spectra Vertical Excitation SSE3.7B-37Main Steam Valve Building Floor Slab Elevation 41 feet 0 inches Amplified Response Spectra N-S Excitation SSE3.7B-38Main Stem Valve Building Floor Slab Elevation 41 feet 0 inches Amplified Response Spectra E-W Excitation SSE MPS-3 FSAR List of Figures (Continued)NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.
Figure Title 3-xxvii Rev. 303.7B-39Main Steam Valve Building Floor Slab Elevation 41 feet 0 inches Amplified Response Spectra Vertical Excitation SSE3.7B-40Main Steam Valve Building Roof Slab Elevation 85 feet 4 inches Amplified Response Spectra N-S Excitation SSE3.7B-41Main Steam Valve Building Roof Slab Elevation 85 feet 4 inches Amplified Response Spectra E-W Excitation SSE3.7B-42Main Steam Valve Building Roof Slab Elevation 85 feet 4 inches Amplified Response Spectra Vertical Excitation SSE3.7B-43Emergency Generator Enclosure Transfer Function of Ba se of Structure Elevation 9 ft-0 in for Horizontal Acceleration N-S Excitation SSE3.7B-44Emergency Generator Enclosure Transfer Function at Floor Slab Elevation 51 feet 0 inches for Horizontal Acceleration N-S Excitation SSE3.7B-45Emergency Generator Enclos ure Transfer Function at R oof Slab Elevation 66 feet 0 inches for Horizontal Acceleration N-S Excitation SSE3.7B-46Emergency Generator Enclosure Transfer Function at Base of Structure Elevation 9 feet 0 inches for Vertical A cceleration Vertical Excitation SSE3.7B-47Emergency Generator Enclosure Transfer Function at Floor Slab Elevation 51 feet 0 inches for Vertical Acceler ation Vertical Excitation SSE3.7B-48Emergency Generator Enclos ure Transfer Function at R oof Slab Elevation 66 feet 0 inches for Vertical Acceler ation Vertical Excitation SSE3.7B-49Emergency Generator Enclosure Transfer Function at Base of Structure Elevation 9 feet 0 inches for Horizontal Acceleration E-W Excitation SSE3.7B-50Emergency Generator Enclosure Transfer Function at Floor Slab Elevation 51 feet 0 inches for Horizontal Ac celeration E-W Excitation SSE3.7B-51Emergency Generator Enclos ure Transfer Function at R oof Slab Elevation 66 feet 0 inches for Horizontal Ac celeration E-W Excitation SSE3.7B-52Emergency Generator Enclosure Floor Sl ab Elevation 24 feet 6 inches Amplified Response Spectra N-S Excitation SSE3.7B-53Emergency Generator Enclosure Floor Sl ab Elevation 24 feet 6 inches Amplified Response Spectra E-W Excitation SSE3.7B-54Emergency Generator Enclosure Floor Sl ab Elevation 24 feet 6 inches Amplified Response Spectra Vertical Excitation SSE3.7B-55Emergency Generator Enclosure Floor Sl ab Elevation 51 feet 0 inches Amplified MPS-3 FSAR List of Figures (Continued)NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.
Figure Title 3-xxviii Rev. 30 Response Spectra N-S Excitation SSE3.7B-56Emergency Generator Enclosure Floor Slab Elevation 51 ft-0 in Amplified Response Spectra E-W Excitation SSE3.7B-57Emergency Generator Enclosure Floor Sl ab Elevation 51 feet 0 inches Amplified Response Spectra Vertical Excitation SSE3.7B-58Emergency Generator Enclosure Roof Sl ab Elevation 66 feet 0 inches Amplified Response Spectra N-S Excitation SSE3.7B-59Emergency Generator Enclosure Roof Sl ab Elevation 66 feet 0 inches Amplified Response Spectra E-W Excitation SSE3.7B-60Emergency Generator Enclosure Roof Sl ab Elevation 66 feet 0 inches Amplified Response Spectra Vertical Excitation SSE3.7B-61Emergency Generator Enclosure Diesel Generator Mat Elevation 24 feet 0 inches Amplified Response Spectra N-S Excitation SSE3.7B-62Emergency Generator Enclosure Diesel Generator Mat Elevation 24 feet 6 inches Amplified Response Spectr a Vertical Excitation SSE3.7B-63Emergency Generator Enclosure Horizontal Acceleration Time History at Bedrock N-S Excitation SSE3.7B-64Emergency Generator Enclos ure Horizontal Acceleration Time History at Base of Structure N-S Excitation SSE3.7B-65Typical Mathematical Model of a Piping System3.7B-66Representation of Family of Peak Responses Curves within Broadened Resonant Peak3.7B-67Hypothetical vs Actual Response of Multiple Modes within Broadened Response Peak3.7B-68Justification of Static Load Factor3.7B-69Typical Amplified Response Spectra3.7B-70Model Beams 3.7B-71Damping Value for Seismic Analysis of Piping 3.7N-1Multi-Degree of Freedom System3.8-1Containment Struct ure Foundation Detail3.8-2Typical Detail of Dome Cylinder Junction 3.8-3Knuckle Plate MPS-3 FSAR List of Figures (Continued)NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.
Figure Title 3-xxix Rev. 303.8-4Mat Reinforcement with Radial Shear Bar Assembly3.8-5Reinforcement Around Ma in Steam Penetrations3.8-6Reinforcing Details Sections thro ugh Ring Beam of Personnel Access Lock3.8-7Reinforcing Details Personnel Access Lock Opening3.8-8Typical Reinforcing Details Sections through Ring Beam of Equipment Access Hatch3.8-9Reinforcing Details Equipment Access Hatch Opening 3.8-10Base Plate Details Contai nment Enclosure Structure3.8-11Ring at Apex of Containment Structure Dome3.8-12Elevation Dome Maintenance Truss 3.8-13Liner Knuckle Plate3.8-14Basic Liner Dimensions3.8-15Typical Piping Penetrations 3.8-16Typical Sleeved Pi ping Penetration3.8-17Multiple Pipe Pe netration Assembly3.8-18Unsleeved Piping Penetration 3.8-19Electrical Penetration3.8-20Fuel Transfer Tube3.8-21Personnel Hatch 3.8-22Equipment Hatch3.8-23Containment Structure Founda tion Detail with Ring Girder3.8-24Detail of Ring Girder at ESF Building 3.8-25Containment Structure Mat Moment and Shear Diagrams3.8-26Containment Structure Mat Moment and Shear Diagrams3.8-27Containment Structure Mat Moment and Shear Diagrams 3.8-28Containment Structure Mat Moment and Shear Diagrams 3.8-29Containment Structure Mat Moment and Shear Diagrams3.8-30Containment Structure Mat Moment and Shear Diagrams MPS-3 FSAR List of Figures (Continued)NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.
Figure Title 3-xxx Rev. 303.8-31Containment Structure Mat Moment and Shear Diagrams3.8-32Load Plot Nomenclature 3.8-33Containment Structure Wall and Dome Force Resultants3.8-34Containment Structure Wall and Dome Force Resultants3.8-35Containment Structure Wall and Dome Force Resultants3.8-36Containment Structure Wall and Dome Force Resultants3.8-37Containment Structure Wall and Dome Force Resultants3.8-38Containment Structure Wall and Dome Force Resultants3.8-39Containment Structure Wall and Dome Force Resultants3.8-40Containment Structure Wall and Dome Force Resultants3.8-41Containment Structure Wall and Dome Force Resultants3.8-42Containment Structure Wall and Dome Force Resultants3.8-43Containment Structure Wall and Dome Force Resultants3.8-44Containment Structure Wall and Dome Force Resultants3.8-45Containment Structure Wall and Dome Force Resultants3.8-46Containment Structure Wall and Dome Force Resultants3.8-47Containment Structure Wall and Dome Force Resultants3.8-48Containment Structure Wall and Dome Force Resultants3.8-49Containment Structure Wall and Dome Force Resultants3.8-50Containment Structure Wall and Dome Force Resultants3.8-51Containment Structure Wall and Dome Force Resultants3.8-52Containment Structure Wall and Dome Force Resultants3.8-53Containment Structure Wall and Dome Force Resultants3.8-54Containment Structure Wall and Dome Force Resultants3.8-55Containment Structure Wall and Dome3.8-56Details of Wall Diagona ls at Containment Mat3.8-57Details of Waterproof Membrane 3.8-58Displacement Profile for Containment Shell Under Pr essure Test (Predicted)
MPS-3 FSAR List of Figures (Continued)NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.
Figure Title 3-xxxi Rev. 303.8-59Containment Structure, Plan Views3.8-60Containment Structur e, Elevation Views3.8-61Containment Enclosure Building3.8-62Auxiliary Building3.8-63Fuel Building 3.8-64Control Room Area3.8-65Cable Tunnel3.8-66Emergency Generator Enclosure Notes to Figure 3.8-66: Data for Pipe Rupture Restraints Steam Generator Blowdown System3.8-67Engineered Safety Features Building3.8-68Main Steam Valve Building 3.8-69Circulating and Service Water Pumphouse3.8-70Recombiner Building3.8-713.8-71 Circulating Discha rge Structure and Tunnel3.8-72Service Building3.8-73Turbine Building3.8-74Solid and Liquid Waste Disposal Building 3.8-75Warehouse No. 53.8-76Auxiliary Boiler Enclosure3.8-77Arrangement of Ho rizontal Shear Keys3.8-78Typical Shear Key Detail3.8-79Fuel Pool Temperature Transients3.8-80Fuel Pool--Fuel Building Temperature Conditions 3.8-81Fuel Pool Temperature Transi ents--for Higher Enrichment Fuel3.8-82Fuel Pool Temperature Tran sients--Full Core Offload3.8-83Millstone Stack Elev ations and Sections3.8-84Millstone Stack -
Mathematical Model3.9B-1Reactor Coolant PPG MPS-3 FSAR List of Figures (Continued)NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.
Figure Title 3-xxxii Rev. 303.9B-2Dynamic Model - Typical RCS Loop3.9B-3Dynamic Model - Center Section 3.9B-4X2 View - Rotated Axes3.9B-5Snubber Orientation3.9N-1Reactor Pressure Vessel and Internals System Model3.9N-2Deleted by PKG FSC 07-MP3-0393.9N-3RPV Support Model for Westi nghouse Internals Response Analysis3.9N-4Control Rod Drive Mechanism 3.9N-5Control Rod Drive Mechanism Schematic3.9N-6Nominal Latch Clearance at Minimum and Maximum Clearance3.9N-7Control Rod Drive Mechanism Latch Clearance Thermal Effect3.9N-8Lower Core Support Assembly (Core Barrel Assembly)3.9N-9Upper Core Support Assembly3.9N-10Plan View of Uppe r Core Support Assembly3.9N-11Neutron Shield Panel Design3.9N-12Neutron Shield Panel Design Details3A.1.2-1One Hundred-Element Ideali zation of Thin-Wall Cylinder3A.1.4-1Element Plot3A.1.4-2Harmonic Axisymmetric Plain Strain3A.1.5-1TAC2D Sample Problem Thermal Model 3A.1.5-2Transient Temperatures in a Right Circular Cylinder-Comparison of TAC2D Results with Series Solution3A.1.6-1Structural Model3A.1.7-1Finite Element Grid fo r PLAXLY-FLUSH Comparison3A.1.7-2Horizontal Amplified Response Spec tra at Top Mass in Beam Structure3A.2.3-1Target for Low Trajectory Missile 3A.2.11-1Cantilever Pipe Used in MIT Experiment3A.3-1Mathematical Mode l for Flexibility An alysis Verification MPS-3 FSAR List of Figures (Continued)NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.
Figure Title 3-xxxiii Rev. 303A.3-2NUPIPE Program Force Time-History Verification3A.3-3Mathematical Model for Class 1 Stress Verification3A.3-4Test Problem - SAVAL3A.3-5Sudden Discharge of a Gas fr om a Pipeline through a Nozzle3A.3-6Sudden Discharge of a Gas fr om a Pipeline through a Nozzle3A.3-7Comparison of Pressure Response at the Closed End3A.3-8Comparison of Pressure Response at the Open End3A.3-9Comparison of Pressure Res ponses by STEHAM and Experiment3A.3-10Hydraulic Network for Verification Problem3A.3-11Hydraulic Network fo r WATHAM Verification3A.3-12Head versus Time Plot for Junction J3A.3-13Head versus Time Plot at Valve3A.3-14Design Input to a Problem3A.3-15Pipe 1 Segment 1 Force-Time History MPS3 UFSAR3.1-1Rev. 30CHAPTER 3 - DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIAThis section evaluates the design bases of Millstone 3 as measured against the NRC General Design Criteria for Nuclear Power Plants, Appendix A to 10 CFR 50, as amended through October 27, 1978. The General Design Criteria (GDC) consist of the single failure definition and 55 individual criteria, and are inte nded to establish the basic requi rements for the principal design criteria of nuclear power plants.The General Design Criteria were not written specifically for the pressurized water reactor; rather, they were intended to guide the design of all water-cooled nuclear power plants. As a result, the criteria are generic in nature and subject to interpretation. For this reason, there are some cases where conformance to a particular criterion is not directly assessable. In these cases, the conformance of plant design to the interpretation of the criteria is discussed. For the single failure definition and each of th e 55 criteria, a specific assessment of the plant design is made and references are included, where necessary, to identify where detailed design information pertinent to each criterion is treated in this FSAR. For the purpose of this report, the terms "important to safety," "safety related," and "safety systems" are synonymous.Based on the contents herein, the Applicants conclude that Millstone 3 satisfies and complies with the GDC, with the exceptions as noted.
 
====3.1.1 CONFORMANCE====
WITH SI NGLE FAILURE CRITERION 3.1.1.1 Single Failure CriterionAppendix A to 10 CFR 50, as ame nded through October 27, 1978, defi nes single fail ure criterion as follows:"A single failure means an occurrence which results in the loss of capability of a component to perform its inte nded safety functions. Multiple failures resulting from a single occurrence are considered to be a si ngle 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 (assu ming active component s function properly) results in a loss of the capability of the system to perform its safety functions."
3.1.1.2 Definitions of Terms Used in Single Failure Criterion Active Component Failure An active component is one in which mechanical movement must occur to complete the component's intended function. An active component failure is failure of the component to complete its intended function upon demand.
MPS3 UFSAR3.1-2Rev. 30 Spurious action of a powered component originati ng within its actuation system shall be regarded as an active component failure unless specific design features or operating restrictions preclude such spurious action.Examples of active component failures include the failure of a powered, manual, or check valve (except ECCS check valves within the NSSS vendor original scope of design) to move to its correct position; the failure of an electrical breake r or relay to respond; and the failure of a pump, fan, or emergency diesel generator to start.
Passive Component Failure A passive component is one in wh ich mechanical movement need not occur for the component to perform its intended function. A passive component failure is the structural failure of the component or the blockage of a process flowpath by the passive component so that the component does not perform its in tended function. For fluid pressure boundary components, a passive component failure is a break of, or crack in, the pressure boundary.
Passive components include pi ping, cables, and valve bodies.Short Term The short term is defined as the first 24 hours following the start of an incident.Long Term The long term is defined as the period following the short term (i.e., greater than 24 hours), during which time the system safety function is still required as defined in the individual equipment qualification documentation described in Section 3.11.
Related Service Systems Related service systems are those systems which provide the services necessary to a fluid system to enable that system to complete its intended safety function.Examples of related service systems for the emergency core cooling system (ECCS) include the safety injection pumps cooling system, charging pumps cooling system, service water system, electric power supply system, protection systems, and the ECCS equipment area ventilation systems.3.1.1.3 Application of Single Failure CriterionFor incidents of moderate frequency (Chapter 15) that can result in automatic reactor or turbine trip when Millstone 3 is generating power for offsite transmission, Millstone 3 is designed to maintain capability for cold safe shutdown assuming a single failure.
MPS3 UFSAR3.1-3Rev. 30 The aggregate of fluid systems provi ded to mitigate the consequences of an infr equent incident or limiting fault (Chapter 15) are designed to tolerate a single failure in addition to the incident which requires their function, wit hout loss of safety function to Millstone 3. Exampl es of systems in this category are: the ECCS, the containment depres surization systems, th e supplementary leak collection and release systems, the auxiliary feedwater system, and the cooling water systems used in combination with these systems.
The 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 conse quences of the initiating event, and3.Loss of offsite electric power if the initiating event results in a trip of either the turbine generator or reactor protection system.For fluid systems, the single failure is limited to an active component failure during the short term, and assuming no prior failure during the short term, the single failure is either an active or a passive component failur e during the long term.High energy piping ruptures and moderate energy leakage cracks are considered separately as single postulated initial events occurring during normal plant operation. A single active failure is assumed in systems used to mitigate the consequences of the piping failure and to shut down the reactor. The single active failure is assumed to occur in addition to the postulated piping failure and any direct consequences of the piping failure.
Where the initial postulated piping failure is assumed to occur in one of two or more redundant trains of a dual-purpose moderate-energy essential system (i.e., one required to operate during normal plant conditions as well as to shut down the reactor and mitigate the consequences of the piping failure), single failures of components in the other train or trains of that system are not assumed, provided that the system is designed to Seismic Category I standards, is powered from both offsite and onsite sources, and is constructed, operated, and inspected to QA Category I and to the testing and inservice inspection standards a ppropriate for nuclear safety systems. Examples of systems that qualify as dual-purpose essential systems are the service water system, the reactor plant component cooling system, and the residual heat removal systems.
A single active failure is not assumed in a dual-purpose safe shutdown system only if:*the initial failure event occurs from a sour ce external to the du al-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 MPS3 UFSAR3.1-4Rev. 30*the safe shutdown system's components are not required to be actuated (i.e.; change state upon demand) to achieve cold shutdown.Under these circumstances, continued operation of the plant is controlled by plant Technical Specifications.A single active failure is not assumed in a safety-related system if the initial failure event is postulated in one of two redundant trains of that system whose function is not required during normal plant operation or to achieve cold shutdown if the failure does not require an automatic protective action to mitigate the consequences of the failure. Under these circumstan ces, continue operation of the plant is controlled by plant Technical Specifications.For electrical and protection systems, the single failure can be either an active or passive failure during the short or long term. This is in accordance with IEEE-308 and Regulatory Guide 1.53 (Section 7.1.2).For the purpose of single failure criterion, the system subject to the criterion includes the safety system itself and its related service systems, such as reactor plant component cooling, service water, electric power, etc., required for the safety system to perform its safety function. Even though each system indicated above is designed to the single failure criterion, only one single failure and its consequences are assumed to take place in the aggregate of safety systems and related service systems in the plant.
Nonsafety related systems are designed such that their failure does not cause safety related systems to lose their safety function.If the proper active function of a component is demonstrated de spite any reasonable postulated condition, that component is considered especially qualified for service and exempt from active component failure. Examples of such component function exemptions include opening of code safety valves and ECCS swing check valves wi thin the NSSS vendor original scope of design.This exemption from single activ e failure consideration applies only to the ECCS system design philosophy employed by the NSSS vendor. As active components (described in FSAR Section 3.9), ECCS check valves are subject to stringent design criteria a nd their operational readiness is periodically verified in accordance with applicable plant technical specifications and the MP3 Inservice Testing Program.In the case of ECCS check valves, the differential pressures required to open the check valve in the forward direction, and the reverse flow velocity required to close the check valve, are very small in comparison to the differential pressures and reverse flow velocities which will be generated if the check valve does not change position immediately as required. Therefore, failure of the check valve to open or close is likely to be momentary, and not of sufficient duration to impact the safety function of the system. In the design of the ECCS, therefore, the failure of a safety-related check valve to open when required to pass flow in the forward direction, or the failure of a safety-related check valve to seat to prohibit flow in the reverse direction is not considered to be a credible event.
MPS3 UFSAR3.1-5Rev. 30Passive component failure assumed in failure analysis of a system is defined by review of each component in the system, considering conditions of operation and possible failure or leakage modes. As an example, review of systems involving piping, heat exchangers, valves, flange joints, and system interface barriers results in a definition of a design leak rate for passive component failure evaluation based on maximum flow through a failed packing, a mechanical seal, or a piping crack (having an opening in size equal to half the pipe diameter in length and half the pipe wall thickness in width for non-ECCS systems). Passiv e component failure is considered for system functional design adequacy and effects from flooding only.As an exception, for ECCS systems containing recirculated sump fluid during the long term post LOCA mode of operation, only limited passive failures, with leakage rates up to 50 gpm, are postulated to occur in piping components or as pump mechanical seal failures. In some areas of the ESF building and pipe tunnel, these long term failures are precluded by conservative piping system design (refer to Section 3.6 for pa ssive failure exclusion design criteria).Means are provided to detect a nd isolate limited passive failures of up to 50 gpm, from ECCS components, within approximately 30 minutes in the ESF building and in approximately 60 minutes in the Auxiliary building (refer to Section 6.3 for ECCS failure modes and effects analysis and a description of leakage de tection and isolation design features).If proper passive functions of a component can be demonstrated despite any reasonable postulated conditions, that component is considered especially qualified for service and exempt from passive component failure. One example of such a case is not assuming a passive component failure in ductwork.In lieu of a review of all the components in a system, complete severance of a line is assumed as the passive component failure.Passive component failure or leakage in excess of limits defined in the Technical Specifications (Chapter 16) is not considered in the primary reactor containment vessel boundary and the containment isolation systems.Conservative piping design criteria have been applied in conjunction with dual isolation valves located outside containment as an acceptable alternative to General Design Criterion 56 for certain containment penetrations described in Sect ion 6.2. Passive failures are not assumed in the line outside containment (between the containment and the outboard isolation valve), provided that the stresses are low, augmented inservice inspection is performe d, and protection from postulated missiles is assured.For design, the mass of fluid discharged through the crack or break prior to effective isolation is considered. Leakage duration is conservatively determined consistent with leakage detection, location, and isolation mechanisms. As a guide, a nominal 30-minute leakage duration is considered conservative for manual operat or action outside of the control room.
MPS3 UFSAR3.1-6Rev. 30The plant design is such that all active components of the designated safety systems and related service systems can be proved operational by scheduled periodic operational tests or operational status indications.
 
====3.1.2 CRITERION====
CONFORMANCE 3.1.2.1 Quality Standards and Records (Criterion 1)
Criterion"Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recogniz ed codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their functions. Appropriate records of the design, fa brication, erection, and testing of structures, systems, and components important to safety shall be maintained by, or under the control of, the nuclear power unit li censee throughout the life of the unit."
Design Conformance Structures, systems, and components important to safety are designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.Quality standards applicable to safety related structures, systems, a nd components are generally contained in codes such as the ASME Boiler and Pressure Vessel Code. The applicability of these codes is specifically identifi ed throughout this report and is summarized in Section 3.2.5.
Chapter 17 provides direct reference to the Quality Assurance Program established to provide assurance that safety related structures, systems, a nd components satisfactorily perform their intended safety functions. The procedures for generating and maintaining appropriate design, fabrication, erection, and testing records are contained within the referenced documents.
3.1.2.2 Design Bases for Protection agains t Natural Phenomena (Criterion 2)
Criterion "Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floo ds, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect: (1) appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident MPS3 UFSAR3.1-7Rev. 30conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed."
Design Conformance Those features of plant facilities that are essential to the prevention of accidents that could affect the public health and safety or to the mitiga tion of accident conse quences are designed to:1.Quality standards that reflect the impor tance of the function to be performed.
Approved design codes are used when a ppropriate to the nuclear application.2.Performance standards that enable the f acility to withstand, without loss of the capability to protect the public, the addi tional forces imposed by the most severe earthquake, flooding condition, wind, ice, or other natural phenomena for the site, and credible combinations of the effect s of normal and accident conditions with the effects of the natural phenomena.Features of the facility essential to accident prevention and mitigation of accident consequences, which are designed to withstand the effects of natural phenomena, are:1.The reactor coolant pressure boundary and containment barriers2.The controls and emergency cooling syst ems whose functions are to maintain the integrity of these barriers3.Systems which depressurize the containment following a loss of coolant accident (LOCA)4.Power supply and essential services5.Reactivity systems, monitoring systems, and fuel systems6.The components used to store and cool spent reactor fuelAll piping, components, and supporting structures of the reactor and safety related systems are designed to withstand a specified seismic disturbance and credible combinations of effects of normal and accident conditions coincident with the effects of natural phenomena. Plant design criteria specify that there is to be no loss of function of such equipment in the event of the safe shutdown earthquake (SSE) ground acceleration acting in the horizontal and vertical directions simultaneously. The dynamic response of Seismic Category I structures to ground acceleration, based on an envelope of characteristics of the site foundation soils and on the critical damping of the foundation and structures, is included in the design analysis.Design of structures for protection against natural phenomena is described in Section 3.8. Safety related structures have sufficient capacity to accept a combination of normal operating loads, MPS3 UFSAR3.1-8Rev. 30functional loads due to the design basis accident (DBA), and the loadings imposed by the maximum wind velocity, or those due to the SSE, whichever is the larger.The emergency onsite power sources are not subject to interruption due to earthquake, windstorm, floods, or to disturbances on the ex ternal power transmission system.
Power cabling, motors, and other equipment required for operation of the engineered safety features are suitably protected against the effects of the design basis accident (DBA) and from severe external weather conditions, as applicable.
Unit design criteria which ensu re protection against natura l phenomena are described in Section 3.2 (Classification of Structures, Systems, and Components), Section 3.3 (Wind and Tornado Loadings), Section 3.4 (Water Level Design), and Section 3.7 (Seismic Design).
3.1.2.3 Fire Protection (Criterion 3)
Criterion "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. Noncombustible and heat-resistant materials shall be used wherever practical throughout the unit, particularly in locations such as the containment and control room. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Fire fighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components."
Design Conformance The design of Millstone 3 minimizes the probability and effect of fires and explosions on structures, systems, and components important to safety. Nonc ombustible and heat-resistant materials are used wherever practical throughout the unit. Fire detection and fire suppression systems of sufficient capacity and capability minimize the adverse effects of fires on structures, systems, and components important to safety. Fire suppression systems are designed to assure that rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.Section 9.5.1 and Millstone 3 Fire Protection Evaluation Report describe the fire protection system in detail.
MPS3 UFSAR3.1-9Rev. 30 3.1.2.4 Environmental and Missile Design Bases (Criterion 4)
Criterion "Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events a nd conditions outside the nuclear power unit."
Design Conformance Structures, systems, and component s important to safety are designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operating, maintenance, testing, and postulated accidents including LOCA's. These items are either protected from accident conditions or designed to withstand, without failure, exposure to the combination of temperature, pressure, humidity, radiation, and dynamic effects expected during the required operational period.Physical separation, physical protection, pipe restraints, and redundancy are included in the design of safety related systems to ensure that each such system performs its intended safety function.In a letter from B. J. Youngblood (NRC) to J. F. Opeka (NNECO) dated June 5, 1985, Millstone 3 was granted an exemption for a period of two cycles of operation from those portions of General Design Criterion 4 which require protection of structures, systems, and components from the dynamic effects associated with postulated breaks in the reactor coolant system primary loop piping.In Federal Register, Volume 51, No. 70, dated April 11, 1986, the NRC published a final rule modifying General Design Criterio n 4 to allow use of leak-befor e-break technology for excluding from the design basis the dynamic effects of postulated ruptures in primary cool ant loop piping in pressurized water reactors. This rule obvi ates the need for the above exemption.Structures, systems, and components important to safety are classified as QA Category I and are designed in accordance with the codes and classifications indicated in Section 3.2.5.
Chapter 3 provides the details of the environmental activities and dynamic effects to which the structures, systems, and components important to safety are designed.
MPS3 UFSAR3.1-10Rev. 30 3.1.2.5 Sharing of Structures, System s, and Components (Criterion 5)Criterion "Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units."
Design Conformance A system important to safety and shared by Mills tone 1, 2, and 3 is the radioactive gaseous waste system discharge stack described in Section 11.3. The sharing of the discharge stack causes no safety impairment of the three units.The following equipment may be shared and utilized by Millstone Unit 2 to meet its GDC 17 requirements for an alternate offsite source to relieve one of its emergency diesel generators and supply power to minimum post-accident loads:1.Main Transformer 15G-3X2.Normal Station Service Transformer 15G-3SA3.Reserve Station Service Transformer 15G-23SAThe sharing of this equipment does not impair its ability to perform its safety function. The transformers are adequately sized and have sufficient capacity to meet maximum postulated Unit 3 loading requirements while suppl ying Unit 2 GDC 17 minimum loads.Other facilities and systems not important to safety within the definitions of GDC 5, but which are shared by the three units are:1.Environmental monitoring systems2.Machine shops3.Offsite transmission lines and switchyard4.Office buildings5.Roadway access6.Railroad access7.General warehouses MPS3 UFSAR3.1-11Rev. 308.Fire protection system9.Potable (Waterford) water10.Warehousing facility houses the Millst one 2 condensate polishing system, the Millstone 2 and 3 (removed from service) condensate demineralizer radioactive liquid waste systems. The Unit 2 conde nsate polishing solid waste system is designed to process radioactive waste from Millstone 2 and 3.11.Alternate AC (SBO) Diesel Ge nerator (shared by Units 2 and 3)12.Security System13.Auxiliary Steam System 3.1.2.6 (Criterion 6)
Criterion 6 has not been promulgated by the NRC.
3.1.2.7 (Criterion 7)
Criterion 7 has not been promulgated by the NRC.
3.1.2.8 (Criterion 8)
Criterion 8 has not been promulgated by the NRC.
3.1.2.9 (Criterion 9)
Criterion 9 has not been promulgated by the NRC.
3.1.2.10 Reactor Design (Criterion 10)
Criterion "The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the eff ects of anticipated oper ational occurrences."
Design Conformance The reactor core and associated coolant, control, and protection systems are designed with adequate margins to:1.Assure that fuel damage is not e xpected during normal core operation and operational transients (Condition I) or any transient conditions arising from MPS3 UFSAR3.1-12Rev. 30 occurrences of moderate fr equency (Condition II). It is not possible, however, to preclude a very small number of rod failures. These are within the capability of the plant cleanup system and are cons istent with plant design bases.2.Ensure return of the reactor to a safe st ate 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 occurrence s of limiting faults (Condition IV).
Note that fuel damage as used under Item 1 is defined as pe netration of the fission product barrier (i.e., the fuel rod clad). Also note that ANSI N18.2-73 expands the definitions of the four conditions enumerated in Items 1 through 3.
Chapter 4 discusses the design bases and design evaluation of reactor components. Chapter 7gives the details of the contro l and protection systems instrumentation design and logic. This information supports the accident analyses of Chapter 15 which show that the acceptable fuel design limits are not exceeded for Condition I and II occurrences.3.1.2.11 Reactor Inherent Protection (Criterion 11)
Criterion "The reactor core and associated coolant systems shall be designe d so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity."
Design Conformance Prompt compensatory reactivity feedback effects are assured when the reactor is critical by the negative fuel temperature effect (Doppler effect) and by ensuring that the moderator temperature coefficient is maintained within the limits provided in Technical Specifica tion 3/4.1.1.3. Section 4.3.2.3 discusses these reactivity coefficients.
3.1.2.12 Suppression of Reactor Power Oscillations (C riterion 12)
Criterion "The reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and read ily detected and suppressed."
MPS3 UFSAR3.1-13Rev. 30 Design Conformance Total power oscillations of the fundamental mode are inherently stable by the negative power coefficient of reactivity.Oscillations, due to xenon spatial effects, in the radial, diametral, and azimuthal overtone modes are heavily damped due to the inherent design and due to the negative power coefficient of reactivity.Oscillations, due to xenon spatial effects, in the axial first overtone mode may occur. Assurance that fuel design limits are not exceeded by xenon axial oscillations is provided by reactor trip functions using the measured axial power imbalance as an input.
Oscillations, due to xenon spatial effects, in axial modes higher th an the first overtone are heavily damped due to the inherent design and due to the negative Doppler coefficient of reactivity.
Section 4.3 discusses xenon and samarium stability control.
3.1.2.13 Instrumentation and Control (Criterion 13)
Criterion "Instrumentation shall be provided to monitor variables and system s over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges."
Design Conformance Instrumentation and controls are provided to monitor and control neutron flux, control rod position, temperatures, pressures, flows, and levels as necessary to assure that adequate plant safety can be maintained. Instrumentation is provided in the reactor coolant system, steam and power conversion system, the containment, engineered safety features systems, and other auxiliaries. Parameters that must be provided for opera tor use under normal operating and accident conditions are indicated in proximity with the controls for maintaining the indicated parameter in the proper range.The quantity and types of process instrumentation provided ensures safe and orderly operation of all systems over the full design range of the plant. These systems are described in Chapters 6, 7, 8,9, 11, and 12.
MPS3 UFSAR3.1-14Rev. 30 3.1.2.14 Reactor Coolant Pressure Boundary (Criterion 14)
Criterion "The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture."
Design Conformance The reactor coolant system boundary is designed to accommodate the system pressures and temperatures attained under all expected modes of plant operation, including all anticipated transients, and to maintain the stresses within applicable stress limits (Section 3.9). Reactor coolant pressure boundary materials, selection, and fabrication techniques ensure a low probability of gross rupture or abnormal leakage.In addition to the loads imposed on the system under normal operating conditions, consideration is also given to abnormal loading conditions, such as seismic and pipe rupture, as discussed in Sections 3.6 and 3.7. The system is protected from overpressure by means of pressure relieving devices as required by appli cable codes (Section 5.2.2).The reactor coolant system boundary has provisions for inspection, testi ng, and surveillance of critical areas to assess the structural and leaktight integrity (Section 5.2). For the reactor vessel (Section 5.3), a material surveillance program conforming to applicable codes is provided.
3.1.2.15 Reactor Coolant System Design (Criterion 15)
Criterion "The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences."
Design Conformance The design pressure and temperature for each component in the reactor c oolant and associated auxiliary, control, and protection systems are selected to be above the maximum coolant pressure and temperature under all normal and an ticipated transient load conditions.Additionally, reactor coolant pressure boundary components achieve a large margin of safety by the use of proven ASME materials and design codes, use of proven fabrication techniques, nondestructive shop testing, and in tegrated hydrostatic te sting of assembled components. Chapter 5 discusses the reactor coolant system design.
MPS3 UFSAR3.1-15Rev. 30 3.1.2.16 Containment Design (Criterion 16)
Criterion "Reactor containment and associ ated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require."
Design Conformance A steel-lined reinforced concrete containment structure, maintained at subatmospheric pressure, encloses the entire reactor coolant system with an essentially leaktight barrier, as described in Section 6.2.1. The containment structure and the engineered safety features are designed to withstand internal and external environmental c onditions that may reasona bly be expected during the life of the unit and to ensure that the shor t and long term conditions following a LOCA do not exceed the design values. Followin g a design basis accident (DBA), the containment heat removal systems reduce the containment pressure, as described in Sections 6.1.2 and 6.2.2. Most of the leakage from the containment structure is collected and processed through the supplementary leak collection and release system, described in Section 6.2.3. This process reduces the amount of radioactivity released to the environment.
3.1.2.17 Electric Power Systems (Criterion 17)
Criterion "An onsite electric power system and an offsite electric power system shall be provided to permit functioning of structures, systems, and components important to safety. The safety function for each system (assuming the other system is not functioning) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is cooled and containment integrit y and other vital functions are maintained in the event of postulated accidents.""The onsite electric power supplies, including the batteries, and the onsite electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.""Electric power from the transmission network to the onsite electric distribution system shall be supplied by two physically independent circuits (not necessarily on separate rights of way) designed and located so as to mi nimize to the extent practical th e likelihood of their simultaneous failure under operating and postulated accident and environmental cond itions. A switchyard common to both circuits is acceptable. Each of these circuits shall be designed to be available in sufficient time following a loss of all onsite alternating current power supplies and the other offsite electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One of these circuits shall MPS3 UFSAR3.1-16Rev. 30be designed to be available within a few seconds following a loss-of-coolant accident to assure that core cooling, containment integrity, and other vital safety func tions are maintained."
"Provisions shall be included to minimize the probab ility of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission network, or the loss of power from the onsite electric power supplies."
Design Conformance Two connections to the offsite power system are provided. The preferred offsite connection is a backfeed through the main and normal station service transformers with the generator breaker open. The alternate offsite connection is through the reserve station serv ice transformers. Each offsite source has 100 percent capacity for all emergency and normal loads during all phases of operation plus the capacity to supply Millstone Unit 2 GDC 17 requirements through the NSST or RSST as an alternate offsite s ource for minimum post-accident loads.Two onsite power systems are provided. Each system has an emergency diesel generator. Each diesel generator has 100 percent capacity for the emergency loads in the event of the postulated accidents or required for reactor cooldown.
The design of the electrical system (Chapter 8) conforms to Criterion 17.
3.1.2.18 Inspection and Testing of Electric Power Systems (Criterion 18)
Criterion "Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the conditions of their components. The systems shall be designed with a capability to test periodically (1) the operability and functional performance of the components of the systems, such as onsite power sources, relays, switches, and buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems into operation, including operation of applicable portions of the protection systems, and the transfer of power among the nuclear power unit, the of fsite power system, and the onsite power system."
Design Conformance The engineered safety features power supply buses, the power supply buses for equipment required for cooldown, and associated emergency generators, all of which comprise the onsite power system, are arranged for periodic testing of each system independently. During refueling shutdowns, tests are conducted to prove the operability of th e automatic starting and load sequencing capability of the emer gency generators. Full load testing of each emergency generator is performed periodically. These tests prove the operability of the electric power systems under conditions as close to design as practical to assess the continuity of these systems and condition of MPS3 UFSAR3.1-17Rev. 30the components. The electric power systems and the testing specifications are described inSections 8.3.1.1 and 16.4.8, respectively.
3.1.2.19 Control Room (Criterion 19)
Criterion "A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe c ondition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.""Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures."
Design Conformance The control room provided is equipped to operate the unit safely unde r normal and accident conditions. Its shielding and vent ilation design permits c ontinuous occupancy of the control room for the duration of a DBA without the dose to personnel exceeding 5 rem whole body. Based on 10 CFR 50.67, the applicable dose criter ion was modified to 5 rem TEDE. The auxiliary shutdown panel located in the west switchgear room has equipment, controls, and instrumentation to accomplish, in conjunction with controls and indication located on the adjacent 4160V switchgear, a prompt hot shutdown and the capability for subsequent cold shutdown of the reactor through the use of suitable procedures. The panel is physically located outside the control room. Thus, the uninhabitability of the control room would have no effect on the availability of the auxiliary shutdown panel and adjacent controls (Section 7.4.1.3).The design of the control building (Section 3.8.4), which houses the control room and the auxiliary shutdown panel area, conforms to Cr iterion 19. Section 9.4.1 describes the control building ventilation system. Control room habitability is discussed in Section 6.4. Fire protection systems are discussed in Section 9.5.1 3.1.2.20 Protection System Functions (Criterion 20)
Criterion "The protection system shall be designed (1) to initiate automatic ally the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety."
MPS3 UFSAR3.1-18Rev. 30 Design Conformance A fully automatic protection system with appropriate redundant channels is provided to cope with transients where insufficient time is available for manual corre ctive action. The design basis for all protection systems is in accordance with IEEE Standard 279-1971 and IEEE Standard 379-1972. The reactor protection system automatically initiates a reactor trip when any variable exceeds the normal operating range. Setpoints are designed to provide an envelope of safe operating conditions with adequate margin for uncertainties to ensure that fuel design limits are not exceeded.Reactor trip is initiated by removing power to the rod drive mechanisms of all the full length rod cluster control assemblies. This causes the rods to insert by gravity rapidly reducing the reactor power output. The response and adequacy of the protection system have been verified by analysis of anticipated transients.The engineered safety features (ESF) actuation system automatically initiates emergency core cooling, and other safeguards functions, by sensing accident conditions using redundant analog channels measuring diverse variables. Manual actuation of safeguards may be performed where ample time is available for ope rator action. The ESF actuation system automatically trips the reactor on manual or automatic safety injection signal (SIS) generation.
3.1.2.21 Protection System Reliability and Testability (Criterion 21)
Criterion "The protection system shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed. Redunda ncy and independence designed into the protection system shall be sufficient to assure th at (1) no single fa ilure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum re dundancy unless the acceptable reliab ility of operation of the protection system can be otherwise demonstrat ed. The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred."
Design Conformance The protection system is desi gned for high functional reliability and inservice testability.
Compliance with this criterion is discus sed in detail in Sections 7.2.2.2.3 and 7.3.2.2.5.
MPS3 UFSAR3.1-19Rev. 30 3.1.2.22 Protection System Independence (Criterion 22)
Criterion "The protection system shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis. Design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function."
Design Conformance Protection system components are designed and arranged so that the environment accompanying any emergency situation in which the components are required to function does not result in loss of the safety function. Various means are used to accomplish this. Functional diversity has been designed into the system. The extent of this functional diversity has been evaluated for a wide variety of postulated accidents. Diverse protection functions automatically terminate an accident before intolerable consequences could occur.Section 7.1.2.1.8 provides details of ESF system diversity.
Automatic reactor trips are based upon process parameters and neutron flux measurements. Trips on process parameters include reactor coolan t loop temperature meas urements, pressurizer pressure and level measurements , and reactor coolant pump underspeed trip. Trips may also be initiated manually or by SIS. Section 7.2 descri bes all the trips and provides further details.High quality components, conservative design and applicable quality control, inspection, calibration, and tests are utilized to guard against common-mode failure. Sections 3.10 and 3.11provide details concerning qualific ation testing. Qualification testings is performed on the various safety systems to demonstrate functional operation at normal and post-accident conditions of temperature, humidity, pressu re, and radiation for specified periods, if required. Typical protection system equipment is subjected to type tests under si mulated seismic condition using conservatively large accelerations and applicable frequencies. The test results indicate no loss of the protection function.
3.1.2.23 Protection System Fail ure Modes (Criterion 23)
Criterion "The protection system shall be desi gned to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced."
MPS3 UFSAR3.1-20Rev. 30 Design Conformance The protection system is designed with due consideration of the most probable failure modes of the components under various perturbations of the environment and energy sources. Sections 7.2 and 7.3 discuss this protection system.
3.1.2.24 Separation of Protection and Control Systems (Criterion 24)Criterion "The protection system shall be separated from control systems to the extent that failure of any single control system component or channel, or failure or removal from service of any single protection system component or channel which is common to the control and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the protection system. Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired."
Design Conformance The protection system is separate and distinct from the control systems. Control systems may be dependent on the protection system in that control signals are derived from protection system measurements, where applicable. These signals are transferred to the control system by isolation devices which are classified as protection components. The adequacy of system isolation is verified by testing under conditions of postulated credible faults. Th e failure of any single control system component or channel, or failure or removal from service of any single protection system component or channel which is common to the control and protection systems, leaves intact a system which satisfies the requirements of the protection system. Distinction between channel and train is made in this discussion. The removal of a train from service is allowed only during testing of the train. Chapte r 7 gives further details.
3.1.2.25 Protection System Requirements for Reacti vity Control Malfunctions (Criterion 25)
Criterion "The protection system shall be designed to assure that sp ecified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods."
Design Conformance The protection system is designed to limit reactivity transients so that fuel design limits are not exceeded. Reactor shutdown by full length rod insertion is completely independent of the normal control function since the trip breakers interrupt power to the rod mechanisms regardless of existing control signals. Thus, in the postulated accidental withdrawal (assum ed to be initiated by a control malfunction), flux, temperature, pressure, level, and flow signals would be generated independently. Any of these signals (trip demands) would operate the breakers to trip the reactor.
MPS3 UFSAR3.1-21Rev. 30Chapter 15 discusses analyses of the effects of possible malfuncti ons. These analyses show that for postulated dilution during refueling, startup, or manual or automatic operation at power, the operator has ample time to determine the cause of dilution, terminate the source of dilution, and initiate boration before the shutdown margin is lost. The analyses show that acceptable fuel damage limits are not exceeded even in the event of a si ngle malfunction of either system.
3.1.2.26 Reactivity Control System Redundanc y and Capability (Criterion 26)
Criterion "Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods and shall be capable of reliabilit y controlling reactivity changes to assure that unde r conditions of normal operation, including anticipate d operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions."
Design Conformance Two reactivity control systems are provided. These are rod cluster control assemblies (RCCA's) and chemical shim (boric acid). The RCCA's are inserted into the core by the force of gravity.
During operation the shutdown rod banks are fully withdrawn. The rod control system automatically maintains a programmed average r eactor temperature compensating for reactivity effects associated with scheduled and transient load reductions. The rod control system cannot automatically withdraw control rods. Operator action is required to restore the plant to equilibrium conditions following load increases. The shutdown rod banks along with the control banks are designed to shut down the reactor with adequate margin under conditions of normal operation and anticipated operational occurrences, thereby ensuring that specified fuel design limits are not exceeded. The most restrictive period in core life is assumed in all analyses and the most reactive rod cluster is assumed to be in the fully withdrawn position.The chemical and volume control system maintains the reactor in the cold shutdown state independent of the position of th e control rods and can compensa te for xenon burnout transients.Chapter 4 presents details of the construction of the RCCA's and Chapter 7 discusses the operation. Chapter 9 describes the means of controlling the boric acid concentration. Chapter 15 includes performance analys es under accident conditions.
MPS3 UFSAR3.1-22Rev. 30 3.1.2.27 Combined Reactivity Control System Capability (Criterion 27)Criterion "The reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained."
Design Conformance The facility is provided with means of making and holding the core subcritical under any anticipated conditions and with appropriate margin for contingencies. Chapter 4 and 9 discuss these means in detail. Combined use of the rod cluster control system and the chemical shim control system permits the necessary shutdown margin to be maintained during long term xenon decay and plant cooldown. The single highest worth control cluster is assumed to be stuck full-out upon trip for this determination. Chapter 15 describes accident as sumptions in detail.
3.1.2.28 Reactivity Limits (Criterion 28)
Criterion "The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently distur b the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. Thes e postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition."
Design Conformance The maximum reactivity worth of control rods and the maximum rates of reactivity insertion employing control rods are limited to values that prevent rupture of the reactor coolant system boundary or disruptions of the core or vessel internals to a degree that could impair the effectiveness of emergency core cooling.The maximum positive reactivity insertion rates for the withdrawal of rod cluster control assemblies (RCCA's) and the dilution of the boric acid in the reactor coolant system are limited by the physical design characteristics of the RCCA's and of the chemical a nd volume control system. Technical Specifications on shutdown margin and on RCCA insertion limits and bank overlaps as functions of power provide additional assurance that the consequences of the postulated accidents are no more severe than those presented in the analyses of Chapter 15. Reactivity insertion rates, dilution, and withdrawal limits are also discussed in Section 4.3. The capability of the chemical MPS3 UFSAR3.1-23Rev. 30 and volume control system to avoid an inadvertent excessive rate of boron dilution is discussed in Chapter 15.Assurance of core cooling capability following Condition IV accidents, such as rod ejections, steam line break, etc., is given by keeping the reactor coolant pressure boundary stresses within faulted condition limits as specified by applicable ASME Codes. Structural deformations are checked also and limited to values that do not jeopardize the operation of necessary safety features.3.1.2.29 Protection against Anticipated Oper ational Occurrences (Criterion 29)
Criterion "The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences."
Design Conformance The protection and reactivity control systems are designed to as sure extremely high probability of performing their required safety functions in any anticipated operational occurrences. Equipment used in these systems is desi gned, constructed, operated, and maintained with a high level of reliability. Chapter 7 covers details of system design. Also refer to the discussions of GDC-20 through 28.
3.1.2.30 Quality of Reactor Coolant Pressure Boundary (Criterion 30)Criterion "Components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage."Design Conformance Reactor coolant pressure boundary components are designed, fabricated, inspected, and tested in conformance with ASME Nuclear Power Plant Components Code, Section III. All components are classified according to ANSI N18.2-73 and N18.2a-75 and are accorded the quality measures appropriate to the classification. The design bases and evaluations of re actor coolant pressure boundary components are discussed in Chapter 5.Leakage is detected by an increase in the amount of makeup water required to maintain a normal level in the pressurizer. The reactor vessel closure joint is provided with a temperature monitored leakoff between double gaskets. Leakage into the reactor containment is drained to the reactor building sump where it is monitored.
MPS3 UFSAR3.1-24Rev. 30Leakage is also detected by measuring the airborne and gaseous activity and activity of the condensate drained from the reac tor building air recirculation units. Monitoring the inventory of reactor coolant in the system at the pressurizer, volume control tank and coolant drain collection tanks make available an accurate indication of integrated leakage.
Section 5.2.5 discusses the reactor coolant pr essure boundary leakage detection system.
3.1.2.31 Fracture Prevention of Reactor Cool ant Pressure Boundary (Criterion 31)
Criterion "The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testin g, and postulated accident conditions (1) the boundary behaves in a nonbrittle manne r and (2) the probability of ra pidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws."
Design Conformance Close control is maintained over material selection and fabrication for the reactor coolant system to assure that the boundary behaves in a non-brittle manner. The reactor coolant system materials which are exposed to the coolan t are corrosion resistant stainless steel or Inconel. The NIL ductility reference temperature (RTNDT) of the reactor vessel structural steel is established by Charpy V-notch and drop weight tests in accordance with 10 CFR 50, Appendix G.As part of the reactor vessel specification, certain requirements which are not specified by the applicable ASME Codes ar e performed as follows:1.Ultrasonic Testing - In addition to c ode requirements, a 100-percent volumetric ultrasonic test of reactor vessel plat e for shear wave and a post-hydro test ultrasonic map of all full penetration fe rritic pressure boundary welds in the pressure vessel are performed. Cla dding bond ultrasonic inspection to more restrictive requirements than those spec ified in the code are also required to preclude interpretation problem s during inservice inspection.2.Radiation Surveillance Progr am - In the surveillance programs, the evaluation of the radiation damage is based on pre-irra diation and post-irra diation 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 te mperature 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.
MPS3 UFSAR3.1-25Rev. 303.Reactor vessel core region material chemistry (copper, phosphorous, and vanadium) is controlled to reduce sensitiv ity to embrittlement due to irradiation over the life of the plant.The fabrication and quality control techniques used in the fabrication of the reactor coolant system are consistent with those used for the reactor vessel. The inspections of reactor vessel, pressurizer, piping, pumps, and steam generator are governed by ASME Code requirements. (Refer to Chapter 5.)Allowable pressure-temperature relationships for plant heatup a nd cooldown rates are calculated using methods derived from the ASME Code, Section III, Appendix G, "Protection Against Non-Ductile Failure." The approach specifies that allowed stress intensit y factors for all vessel operating conditions may not excee d the reference stress intensity factor (KIR) for the metal temperature at any time. Operating specifications include conservative margins for predicted changes in the material reference temperatures (RTNDT) due to irradiation.
3.1.2.32 Inspection of Reactor Coolant Pressure Boundary (Criterion 32)
Criterion "Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of important areas and features to assess their structural and leak-tight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel."Design Conformance The design of the reactor coolant pressure boundary provides th e capability for accessibility during service life to the entire internal surfaces of the reactor vessel, certain external zones of the vessel including the nozzle to reactor coolant piping welds and the top and bottom heads, and external surfaces of the reactor coolant piping except for the area of pipe within the primary shielding concrete. The inspection capability complements the leakage detection systems in assessing the pressure boundary component's integrity. The reactor coolant pressure boundary is periodically inspected under the provisions of ASME Boiler and Pressure Vessel Code, Section XI. Section 5.2.4 gives details of the Inservice Inspection Program.
Monitoring of changes in the frac ture toughness properties of the reactor vessel core region plates forging, weldments, and associated heat treated zones are performed in accordance with 10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements." Samples of reactor vessel plate materials are retained and catalogued in case future engineering development shows the need for further testing.The material properties surveillance program includes not only the conve ntional tensile and impact tests, but also fracture mechanics specimens. The observed shifts in RTNDT of the core region materials with irradiation are used to confirm the allowable limits calculated for all operational transients. Secti on 5.3 gives further details.
MPS3 UFSAR3.1-26Rev. 30 3.1.2.33 Reactor Coolant Makeup (Criterion 33)
Criterion "A system to supply reactor coolant makeup for protection against small breaks in the reactor coolant pressure boundary shall be provided. The system safety function shall be to assure that specified acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to leakage from the reactor coolant pressure boundary and rupture of small piping or other small components which are part of the boundary. The system shall be designed to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished using the piping, pumps, and valves used to maintain coolant inventory during normal reactor operation."
Design Conformance The chemical and volume control system provides a means of reactor coolant makeup and adjustment of the boric acid concentration. Makeup is added automatically if the level in the volume control tank falls below the normal ope rating range. The high-pressure centrifugal charging pumps provided are capable of supplying the required ma keup and reactor coolant seal injection flow when power is available from either onsite or offsite electric power systems. These pumps also serve as high head safety injection pumps. Functional reliability is assured by provision of standby components assuring a safe response to probable modes of failure. Sections 6.3 and 9.3 include details of system design and Ch apter 8, details of the electric power system.
3.1.2.34 Residual Heat Removal (Criterion 34)
Criterion "A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the re actor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.""Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure."
Design Conformance The residual heat removal system , in conjunction with the steam and power conversion system, is designed to transfer the fission product decay heat and other residual heat from the reactor core within acceptable limits. The transfer of the heat removal function from the steam and power MPS3 UFSAR3.1-27Rev. 30conversion system to the residual heat removal system occurs when the reactor coolant system is at approximately 350
°F and 375 psig.Suitable redundancy at temperatures below approximately 350
°F is accomplished with the two residual heat removal pumps (located in separate compartments with means available for draining and monitoring of leakage), the two heat exchangers and the associated piping, cabling, and electric power sources. The residual heat removal system is able to operate on either onsite or offsite electrical power system.Suitable redundancy at temperatures above approximately 350
°F is provided by the steam generators and associated piping system.
Section 5.4.7 and Chapter 10 give details of the system design.
3.1.2.35 Emergency Core Cooling (Criterion 35)
Criterion "A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.""Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure."
Design Conformance An emergency core cooling system is provided to cope with any loss-of-coolant accident in the plant design basis. Abundant cooling water is available in an emergency to transfer heat from the core at a rate that clad metal-water reaction is limited to less than one percent. Adequate design provisions are made to assure performance of the required safety functions even with a single failure.Section 6.3 includes details of the capability of the systems. Chap ter 15 includes an evaluation of the adequacy of the system functions. Performance evaluations are conducted in accordance with 10 CFR 50.46 and Appendix K to 10 CFR 50.
MPS3 UFSAR3.1-28Rev. 30 3.1.2.36 Inspection of Emergency Core Cooling System (Criterion 36)Criterion "The emergency core cooling system shall be designed to permit appropriate periodic inspection of important components, such as spray rings in the reactor pressure vessel, water injection nozzles, and piping to assure the integrity and capability of the system."
Design Conformance Design provisions facilitate access to the critical parts of the injection nozzles, pipes, and valves for visual inspection and for nondestructive inspection where such techniques are desirable and appropriate. The design is in accordan ce with ASME, Section XI requirements.The components outside the co ntainment are accessible for leaktightness inspection during operation of the reactor.
3.1.2.37 Testing of Emergency Core Cooling System (Criterion 37)Criterion "The emergency core cooling system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system."
Design Conformance Active components of the emergency core cooling system can be actuated from the emergency power source at any time during unit operation to demonstrate operability.Tests are performed during refu eling shutdowns to demonstrate proper automatic operation of the emergency core cooling system. An integrated system test is performed. Sections 6.3 and 7.3 describe the above tests.
3.1.2.38 Containment Heat Removal (Criterion 38)
Criterion "A system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and te mperature following any loss-of-coolant accident and maintain them at acceptably low levels."
MPS3 UFSAR3.1-29Rev. 30"Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure."
Design Conformance Heat is removed from the containment structure following a LOCA by the containment depressurization systems, which consist of the quench spray system and the containment recirculation system (Section 6.2.2). The quench spray system, consisting of two 100-percent capacity subsystems, transfers water from the refueling water storage tank to two parallel 360-degree spray headers. The quench spray system transfers heat from the containment atmosphere to water on the containment structure floor. The cont ainment recirculation system, which consists of two 100-percent capacity subsystems (each consisting of two pumps, two coolers, and a 360-degree spray header), transfers heat from the water collected in the containment structure sump to the service water system (Section 9.2.1) via the containment recirculation coolers. The quench spray pumps and the containment recirculation pumps and coolers are located in the engineered safety features building (Section 3.8).The containment depressurization systems are designed so that no single active failure in the short term or no single active or passive failure in the long term impairs their ability to perform their safety function. Redundant components are isolated, physically and electrically. Each subsystem is connected to a separate electrical bus which can be connected to either offsite or onsite power.
3.1.2.39 Inspection of Containment Heat Removal System (Criterion 39)
Criterion "The containment heat removal system shall be designed to permit appropriate periodic inspection of important components, such as the torus, sumps, sp ray nozzles, and piping to assure the integrity and capability of the system."
Design Conformance The design of the containment depressurization systems permits appropriate periodic inspection of the important components, as described in Section 6.2.2.4.
3.1.2.40 Testing of Containment Heat Removal System (Criterion 40)Criterion "The containment heat removal system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole, and, under conditions as close to the design as practical, the performance of MPS3 UFSAR3.1-30Rev. 30the full operational sequence that brings the system into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the as sociated cooling water system."
Design Conformance The design of the containment de pressurization systems permits periodic pressure and functional testing, as described in Section 6.2.2.4.
3.1.2.41 Containment Atmosphere Cleanup (Criterion 41)
Criterion "Systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained.""Each system shall have suitable redundancy in components and features, and suitable interconnections, leak detection, is olation, and containment capabilities to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) its safety function can be accomplished, assuming a single failure."
Design Conformance The supplementary leak collection and release system (SLCRS) (Section 6.2.3.2) collects radioactive leakage from the containment to the containment enclosure and contiguous areas following a LOCA.The quench spray system (Section 6.2.2 and 6.5.2) sprays borated water into the containment atmosphere to reduce the containment pressure and remove airborne iodine. The pH in the containment sumps is controlled by the dissolution of trisodium phosphate (stored in baskets) in the sump water.
The recirculation spray system (Sections 6.2.2 and 6.5.2) sprays cont ainment sump water into the containment atmosphere to reduce containmen t pressure and remove airborne iodine.These systems are sufficiently redundant to perform their safety function assuming a single active failure in the short term or a single active or passive failure in th e long term and are operable with either onsite or offsite power.
MPS3 UFSAR3.1-31Rev. 30 3.1.2.42 Inspection of Containment Atmos phere Cleanup Systems (Criterion 42)
Criterion "The containment atmosphere cleanup systems shall be designed to permit appropriate periodic inspection of important components, such as filter frames, ducts, and piping to assure the integrity and capability of the systems."
Design Conformance The design of the supplementary leak collection and release system permit appropriate periodic inspection of the important compone nts, as described in Section 6.5.1.4.
3.1.2.43 Testing of Containment Atmosphe re Cleanup Systems (Criterion 43)
Criterion "The containment atmosphere cleanup systems shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability a nd performance of the active comp onents of the systems such as fans, filters, dampers, pumps, and valves and (3) the operability of the systems as a whole and under conditions as close to design as practical, the performance of the full operational sequence that brings the systems into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of associated systems."
Design Conformance The design of the supplementary leak collection an d release system permits periodic pressure and functional testing of components, as described in Section 6.5.1.4.
3.1.2.44 Cooling Water (Criterion 44)
Criterion "A system to transfer heat from structures, systems, and components important to safety, to an ultimate heat sink shall be provided. The system safety function shall be to transfer the combined heat load of these structures, systems, and components unde r normal operating and accident conditions."Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure."
MPS3 UFSAR3.1-32Rev. 30 Design Conformance The reactor plant component cooling water system, the charging pump cooling system, spent fuel pool cooling and purification system and the safety injection pump cooling system, transfer heat from systems containing reactor coolant to the service water system. Together, these systems transfer heat to the ultimate heat sink from structures, systems, and components important to safety during normal and accident conditions.These systems are designed with suitable redundancy in components, with leak protection, and with the capability to isolate redundant components. The systems are designed to satisfy the cooling water requirements assuming a single failure and either a loss of onsite or offsite power. Designs of these systems are descri bed in FSAR sections, as follows:
3.1.2.45 Inspection of Cooling Wa ter System (Criterion 45)
Criterion "The cooling water system shall be designed to permit appropriate periodic inspection of important components, such as heat exchangers and piping, to assure the integrity and capability of the system."
Design Conformance The service water system (Section 9.2.1), the reactor plant component cooling water system (Section 9.2.2.1), the charging pumps cooling system (Section 9.2.2.4), the safety injection pumps cooling system (Section 9.2.2.5), and the spent fuel pool cooling and purification systems (Section 9.1.3) are designed to permit appropriate periodic inspection in order to ensure the integrity of the components and the systems as a whole. In addition, as the service water, reactor plant component cooling water, charging pumps cooling, and spent fuel pool cooling and purification systems function almost continuously during normal unit operation, their capability and integrity are continuously demonstrated. The safety injection pumps cooling system is operated periodically to assure its capability and integrity.Title Section No.
AC Power Supply System8.3.1Service Water System9.2.1Reactor Plant Component Cooling Water System9.2.2.1 Charging Pumps Cooling System9.2.2.4 Safety Injection Pumps Cooling System9.2.2.5Ultimate Heat Sink9.2.5 Spent Fuel Pool Cooling and Purification9.1.3 MPS3 UFSAR3.1-33Rev. 30 3.1.2.46 Testing of Cooling Water System (Criterion 46)
Criterion "The cooling water system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and the performance of the active components of the sy stem, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation for re actor shutdown and for loss-of-coolant accidents, including operation of applicable portions of the protection system and the transfer between normal and emergency power sources."
Design Conformance The service water system (Section 9.2.1), reac tor plant component cooling water system (Section 9.2.2.1), charging pumps cooling system (Section 9.2.2.4), safety injection pumps cooling system (Section 9.2.2.5), and the spent fuel pool cooling and purification system (Section 9.1.3) are designed to permit periodic pressure and functional testing. With the exception of the safety injection pumps cooling system, these systems operate during normal operation and shutdown; thus, the structural and leaktight integrity of the system components, the operability and performance of most of the active components, and the operability of the system as a whole are continuously dem onstrated. The active components that cannot be tested during normal system operation are te sted during shutdown.The safety injection pumps cooling system, which is not normally in service, is periodically tested to assure structural and leaktight integrity of its components, the operability a nd performance of its active components, and the operability of the system as a whole.The performance of the full operational sequence for the safety related portions of the above systems that brings each system into operation for reactor shutdown, LOCA, or loss of unit power is evaluated periodically in conj unction with the applicable port ions of the protection system.Transfer between normal and emergency pow er sources is discussed in Section 8.3.
3.1.2.47 (Criterion 47)
Criterion 47 has not been promulgated by the NRC.
3.1.2.48 (Criterion 48)
Criterion 48 has not been promulgated by the NRC.
3.1.2.49 (Criterion 49)
Criterion 49 has not been promulgated by the NRC.
MPS3 UFSAR3.1-34Rev. 30 3.1.2.50 Containment Design Basis (Criterion 50)
Criterion "The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate w ithout exceeding the design leakage rate and, with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident. This margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators and as required by Paragraph 50.44 energy from metal-water and other chemical reactions that may result from degradation but not total failure of emergency core cooling functioning, (2) the limited experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters."
Design Conformance The containment structure is de signed with a leakage rate shown in Table 1.3-3. The containment is designed to withstand, by a sufficient margin, loads above those that are conservatively calculated to result from a DBA as discussed in Section 6.2.1.
3.1.2.51 Fracture Prevention of Containmen t Pressure Boundary (Criterion 51)
Criterion "The reactor containment boundary shall be designed with sufficient margin to assure that under operating, maintenance, testing, and postulated accident conditions (1) its ferritic materials behave in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the containment boundary material during operation, maintenance, testing, and postulated accident conditions, and the uncertainties in determining (1) material properties, (2) residual, steady state, and transient stre sses, and (3) size of flaws."
Design Conformance Ferritic materials for th e containment structure boundary are specified so that, when the liner, equipment latch, personnel lock, penetrations, and fluid system components, including valves required to isolate the system are exposed to postulated acci dent, test, operating and normal conditions, the corresponding and resultant stress levels are below the maximum stress level permitted by the crack arrest temperature (CAT) curve of NRL Report 6900 for each applicable correspondence temperature condition.
MPS3 UFSAR3.1-35Rev. 30Nil Ductivity Transition Temperature Figure 23 of NRL Report 6900 shows the fracture analysis diagram (FAD), which plots stress vs temperature in excess of nil ductility transition temperature (N DTT). The contai nment structure liner is designed so that no stress exceeds the CAT curve shown in this FAD. This approach is very conservative and ensures that flaws of any size are not propagated to a rapid (i.e., brittle) fracture.Uncertainties in the determination of NDTT are minimized by using the drop weight test, ASTM E208, for material 5/8 inch or thicker. Charpy V-notch tests are required for all material which form part of the containment structure boundary.
3.1.2.52 Capability for Containment Leakage Rate Testing (Criterion 52)
Criterion "The reactor containment and other equipment which may be subjected to containment test conditions shall be designed so that periodic integrated leakage rate testing can be conducted at containment design pressure."
Design Conformance The design of the containment st ructure and related equipment, which are subjected to the containment structure test conditions, as described in Section 6.2.6, allows for conducting periodic integrated leakage rate te sting of the containment structure.
3.1.2.53 Provisions for Containment Test ing and Inspection (Criterion 53)
Criterion "The reactor containment shall be designed to permit (1) appropriate periodic inspection of all important areas, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment design pressure of the leaktightness of penetrations which have resilient seals and expansion bellows."
Design Conformance The reactor containment is design ed to permit (1) appr opriate periodic inspect ion of all important areas, such as penetrations, (2) an appropriate surveillance progr am, and (3) periodic testing at containment design pressure of the leaktightness of penetrations which have resilient seals and expansion bellows, as discussed in Section 6.2.6 (Containment Leakage Testing).
MPS3 UFSAR3.1-36Rev. 30 3.1.2.54 Piping Systems Penetrating Containment (Criterion 54)
Criterion "Piping systems penetrating primary reactor containment shall be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems. Such piping systems shall be designed with a capability to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits."
Design Conformance The piping systems penetrating the containment structure are designed to minimize leakage. Containment isolation valves provide the capability to seal most penetrations redundantly; Section 6.2.4 describes the few exceptions in detail. Pressure taps provide the capability to perform a Type C (10 CFR 50 Appendix J) test to meas ure containment isolation valve leakage rates, as outlined in Section 6.2.4.
3.1.2.55 Reactor Coolant Pr essure Boundary Penetrating Containment (Criterion 55)
Criterion "Each line that is part of the reactor coolant pressure boundary and that penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:1.One locked closed isolation valve inside and one locked closed isolation valve outside containment; or2.One automatic isolation valve inside and one locked closed is olation valve outside containment; or3.One locked closed isolation valve inside and one automa tic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or4.One automatic isolation valve inside a nd one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.""Isolation valves outside containment shall be located as close to containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety."
MPS3 UFSAR3.1-37Rev. 30"Other appropriate requirements to minimize the probability or consequences of an accidental rupture of these lines or of lines connected to them shall be provided as necessary to assure adequate safety. Determination of the appropriateness of these requirements, such as higher quality in design, fabrication, and testing, additional provisions fo r inspection, protection against more severe natural phenomena, and additional isolation valves and containment, shall include consideration of the population density, use characte ristics, and physical char acteristics of the site environs."
Design Conformance All lines that are part of the reactor coolant pressure boundary and that penetrate the containment structure are provided with containment isolation valves in accordance with the above criterion, as described in Section 6.2.4. Valves outside the containment structure are located as close as practical to the containment structure. All containment penetrations and isolation valves are protected against possible environmental effects including missiles. The isolation valves are subject to periodic Type C tests (10 CFR 50, Appendix J), as outlined in Section 6.2.4, and, automatic valves, upon loss of actuating power, take the position that provides the greatest safety.
There are no additional requirements for the mechanical design of those lines that are part of the reactor coolant pressure boundary and that pe netrate the containmen t structure beyond those required by the applicable standards and codes.
3.1.2.56 Primary Containment Is olation (Criterion 56)
Criterion "Each line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:1.One locked closed isolation valve inside and one locked closed isolation valve outside containment; or2.One automatic isolation valve inside and one locked closed is olation valve outside containment; or3.One locked closed isolation valve inside and one automa tic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or4.One automatic isolation valve inside a nd one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.
MPS3 UFSAR3.1-38Rev. 30"Isolation valves outside containment shall be located as close to the containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety."
Design Conformance All lines that connect to the containment atmosphere are provided with containment isolation valves in accordance with the above criterion except for those lines identified otherwise in Section 6.2.4. Valves outside the containment structure are located as close as possible to the containment structure. The isolation valves are subject to periodic Type C tests (10 CFR 50, Appendix J) as outlined in Section 6.2.4 and, upon loss of actuating power, take the position that provides the greatest safety.Instrument line penetrations are in accordance with Regulatory Guide 1.11 (Section 6.2.4) and have a remote manual isolation valve outside of the containment structure.
3.1.2.57 Closed System Isolation Valves (Criterion 57)
Criterion "Each line that penetrates primary reactor containment and is neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere shall have at least one containment isolation valve which shall be either automatic, or locked closed, or capable of remote manual operation. This valve shall be outside containment and located as close to the containment as practical. A simple check valve may not be used as the automatic isolation valve."
Design Conformance All lines connected to closed systems are provided with at least one cont ainment isolation valve located as close as possible to the outside of the containment structure as described in Section 6.2.4. The isolation valves are subject to periodic tests (10 CFR 50, Appendix J) as outlined in Section 6.2.4. These auto matic isolation valv es, upon loss of actuating power, take the position that provides the greatest safety.
3.1.2.58 (Criterion 58)
Criterion 58 has not been promulgated by the NRC.
3.1.2.59 (Criterion 59)
Criterion 59 has not been promulgated by the NRC.
MPS3 UFSAR3.1-39Rev. 30 3.1.2.60 Control of Releases of Radioactive Ma terials to the Environment (Criterion 60)
Criterion "The nuclear power unit design shall include means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation, including anti cipated operational occurrences. Sufficient holdup capacity shall be provided for retention of gaseous and liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operational limitations upon the release of such effluents to the environment."
Design Conformance In all cases, the design for radi oactivity control is based on:1.The requirements of 10 CFR 20, 10 CF R 50, and Appendix I to 10 CFR 50 for normal operations and for any transient situation that mi ght reasonably be anticipated to occur. 2.10 CFR 50.67 dose level guidelines for pot ential accidents of extremely low probability of occurrenceAll release paths, including ven tilation and process streams, are monitored and controlled as described in Section 11.5.The activity level of the radioactive gaseous waste effluents subsequent release through the 375-foot Millstone stack are monitored (Section 11.3.2.4). Under conditions of concurrent fuel failure and steam generator tube leakage, some radioactive gas would be present and suitably controlled in the steam jet air ejector discharge in the condenser air removal system (Section10.4.2) and in the flow from the steam packing exhauster fan in th e turbine generator gland seal and exhaust system (Section 10.4.3). The steam jet air ejector discharge is directed to the Millstone stack while the seal steam packing exhauster fan discharges through the condensate polishing enclosure roof.Control of liquid waste effluents (Sections 11.2 and 11.5) is mainta ined by batch processing of all liquids, sampling before discharge, and a controlled rate of release. Liquid effluents are monitored for radioactivity and rate of flow. Radioactive liquid waste system capacities are sufficient to handle any expected transient in the processing of liquid waste.Solid wastes are prepared for offsite dispos al by either compaction or solidification (Section 11.4). Solid waste is prepared for shipment by placement in properly labeled containers that meet applicable NRC and Department of Tran sportation dose rate require ments as detailed in 10 CFR 71, 49 CFR 170-178, and Section 11.4.
MPS3 UFSAR3.1-40Rev. 30 3.1.2.61 Fuel Storage and Handling and Radioactive Control (Criterion 61)
Criterion "The fuel storage and handling, radioactive waste, and othe r systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capab ility having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage coolant inventor y under accident conditions."
Design Conformance Safety related components in the fuel storage and handling system (Section 9.1.3.4) are designed to allow periodic inspection and testing to ensure proper operation. Performance of components important to safety in the radioactive liquid and gaseous waste systems is verified by extensive process fluid analysis and continuous radiation monitoring of gaseous effluents, respectively.The new and spent fuel storage areas are designed to meet the requirements of 10 CFR 20 in providing radiation shielding for operating personnel during new and spent fuel transfer and storage. The fuel transfer canal and spent fuel pool wall thickness are sufficient to shield adjacent work areas to meet the requirements of 10 CFR 20 for personnel access during actual fuel transfer. Waste storage and processing facilities in the a uxiliary building and the waste disposal building are shielded to meet the requirements of 10 CFR 20 for operating personnel. Periodic surveys by radiation protection personnel and continuously operated radiation monitors located in areas selected to afford maximum personnel protection (Section 12.1) ensure that radiation design levels are not exceeded during the operating lifetime of the unit.New and spent fuel handling systems are designed to preclude gross mechanical failures which could lead to significant radioact ivity releases. Floor and equipmen t drains are provided to collect leakage which might occur from valve stem leakoffs, pump seals, and other equipment, and to transfer the leakage to one of the building sumps for eventual processing by the liquid waste system.Radiation gases and particulates released from components are collected by the reactor plant aerated vents system. Uncontrolled leakage of radioactive gases and particulates which may leak from spent fuel, radioactive waste, or components containing radioactive fluids is collected and treated by the respective building ventilation filtration system (Section 9.4) or supplementary leakage collection and release system (Section 6.5.1). All discharges from these systems are monitored for radioactivity.Decay heat from spent fuel is dissipated in the water of the spent fuel pool and subsequently removed by the cooling portion of the fuel pool cooling and purification system (Section 9.1.3). Redundancy of fuel pool cooling and purification system components ensures reliability in MPS3 UFSAR3.1-41Rev. 30controlling the spent fuel pool water temperature. Spent fuel pool cooling system operation is continuously monitored in the ma in control room where spent fuel pool water temperature is both indicated and alarmed. Special tests are not required because at least one pump and heat exchanger are normally in operation when spen t fuel is stored in the spent fuel pool.The piping connected to the spent fuel pool is designed so that an acceptable water level is maintained in the event of a pipe rupture. Instrumentation to annunciate spent fuel pool water level changes above or below preset levels is provided on the fuel pool control panel in the main control room. Redundancy of makeup water sources ensures adequate supply and availability of makeup to the spent fuel pool even under loss of normal electrical power.
3.1.2.62 Prevention of Criticality in Fuel Storage and Handling (Criterion 62)
Criterion "Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of ge ometrically safe configurations."
Design Conformance Criticality is prevented in the new fuel storage racks by a combination of geometry and poison material as described in Section 9.1.1 and 4.3.2.6.Criticality is prevented in the spent fuel storage area by the physical separation of fuel assemblies, limits on the enrichment, burnup and decay times of the fuel, and the use of fixed neutron poisons in Region 1 and 2. Soluble boron in the spent fuel pool water is credited for certain accident conditions. Sections 9.1.1 and 9.1.2 discuss criticality prevention in more detail.
3.1.2.63 Monitoring Fuel and Waste Storage (Criterion 63)
Criterion "Appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels a nd (2) to initiate appropriate safety actions."
Design Conformance The spent fuel pool water temperature is continuously monitored. Safety related temperature indicators are provided in the control room where alarms are also provided should the water temperature increase above a preset level. Spent fuel pool temperature is also indicated and alarmed at the fuel pool cooling and purif ication panel (FP) in the fuel building.Safety related spent fuel pool lo w water level indicating lights are displayed in the control room. Additionally, alarms are provided in the control room and at the FP in the fuel building should the level increase or decrease beyond preset levels. The spent fuel pool cooling system cooler outlet MPS3 UFSAR3.1-42Rev. 30temperature and flow rate are each monitored at the FP. Should either of these parameters or the component cooling system flow rate from the operating cooler exceed preset values, an individual alarm is actuated at the FP and the fuel pool cooling system "trouble" alarm is actuated in the control room.In the event of high temperature, low flow, or abnormal spent fuel pool level indication, administrative procedures provide for checking the operating status and integrity of the spent fuel pool and support cooling systems, inspecting for spent fuel pool leakage, and ensuring that corrective measures are taken to rest ore all system parameters to normal.The fuel pool demineralizer can be aligned to either the SFP or the RWST. The specific conductivities of the SFP and the RWST (the fuel pool demineralizer influent sources) are monitored and recorded weekly. The fuel pool demineralizer effluent specific conductivity is monitored and recorded monthly. In the event of abnormal conductivity levels in the SFP, RWST, or the effluent of the fuel pool demineralizer, administrative procedures provide for taking additional samples to assist in determining the source and/or cause of the abnormal conductivity levels and for ensuring that corrective measures are taken.Radiation levels in the area of spent fuel storage are continuously monitored by radiation detectors located around the periphery of the storage areas. Other continuously operating radiation detectors are located in the fuel and waste disposal buildings in areas best suited for alerting operating personnel of high lo cal radiation levels.
Radiation levels in excess of the preset values for either the fuel building or waste storage areas initiate alarms, both locally and in the control room.In the event of a high airborne gross activity alarm, the fuel building ventilation system exhaust (Section 9.4.2) is remote manually diverted to its exhaust filtration system. In the waste disposal building, the ventilation system exhaust (Section 9.4.5) can be remote manually diverted to the auxiliary building filtration system upon r eceipt of a high airborne activity alarm.
3.1.2.64 Monitoring Radioactivity Releases (Criterion 64)
Criterion "Means shall be provided for moni toring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents."
Design Conformance The containment atmosphere is monitored during normal and transient operations of the reactor plant by the containment structur e particulate and gas monitor located in the upper level of the auxiliary building (Section 12.3.4) or by grab sampling. Normal unit effluent discharge paths are monitored during normal plant opera tion by the ventilation particul ate samples and gas monitors in the auxiliary building and engineered safety buildings (Section 11.5). After a postulated MPS3 UFSAR3.1-43Rev. 30accident, the safety related ventilation vent monitors and the safety related Supplementary Leak Collection and Release System monitors are used to monitor the effluents from spaces contiguous to the containment structure including the areas that contain loss-of-coolant accident fluids. In addition, the service water outlet from each pair of containment recirculation coolers is monitored to ensure that any leakage of radioactive fluids into the service water system is detected (Section 11.5). Radioactivity levels in the environs are controlled dur ing normal and accident conditions by the various radiation monitoring systems (Sections 11.5 and 12.3.4) and monitored by the collection of samples as part of the offsite radiological monitoring program.
 
====3.1.3 REFERENCE====
FOR SECTION 3.1 MPS3 UFSAR3.2-1Rev. 30
 
===3.2 CLASSIFICATION===
OF STRUCTURES, SYSTEMS, AND COMPONENTS
 
====3.2.1 SEISMIC====
CLASSIFICATION Seismic Category I structures, systems, and components are those necessary to ensure:1.The integrity of the reactor coolant pressure boundary2.The capability to shut down the reactor and maintain it in a safe shutdown condition, or3.The capability to prevent or mitigate th e consequences of ac cidents which could result in potential off site exposures comparable to the guideline exposure of 10 CFR 100.Seismic Category I structures, sy stems, and components are designed to rema in functional during a safe shutdown earthquake (SSE). Strain limits in excess of yi eld are allowed provided safety functions are maintained. Seismic Category I structures, systems, and components are also designed to be well within the el astic limit for a vibratory motion at 50 percent of the SSE. This requirement is called the oper ating basis earthquake (OBE).
The SSE and the OBE are described in Section 2.5. The seismic design of Seismic Category I systems and components is described in Sections 3.7 and 3.10 and of structures in Section 3.8.
Seismic Category I structures, systems, and components are listed in Table 3.2-1.The seismic classification of structures, systems, and components complies with Regulatory Guide 1.29.Structures, systems, and com ponents designated Seismic Category I in accordance with NRC Regulatory Guide 1.29 are listed in Table 3.2-1.
 
====3.2.2 SYSTEM====
QUALITY GROUP CLASSIFICATION The containment structure and safety related fluid systems, listed in Table 3.2-1, are classified according to the classes listed below. Supports are in the same safety class as the components for which they provide support if failure of the support could cause a loss of a safety function associated with the supported component.
Safety Class 1 Safety Class 1 (SC-1) applies to reactor coolant pressure boundary com ponents whose failure during normal reactor operation would prevent orderly reactor shutdown and cooldown. The reactor coolant pressure boundary is defined in Paragraph 50.2 of 10 CFR 50 as being all those pressure-containing components "... such as pressure vessels, piping, pum ps, and valves, which are:
MPS3 UFSAR3.2-2Rev. 301.Part of the reactor coolant system, or2.Connected to the reactor coolant system up to and including any and all of the following:a.The outermost containment isolat ion valve in system piping which penetrates the primar y 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.According to Section 50.55a of 10 CFR 50, components which are connected to the reactor coolant system and are part of the reactor coolant pressure boundary may be downgraded in safety class 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, or2.The component is or can be isolated from the reactor coolant system by two valves (both closed, both open, or one closed a nd the other open). Each open valve must be capable of automatic act uation 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.""Shut down and cooled down in an orderly manner" is taken as an action where the pressurizer does not empty, assuming normal makeup is available.
Safety Class 2 Safety Class 2 applies to:1.Pressure containing components of th e reactor coolant pressure boundary not covered in Safety Class 12.The reactor containment, including those valves and components of systems used to affect isolation of the reactor containment3.Components necessary to the system safety function of the following:a.Residual heat removal system MPS3 UFSAR3.2-3Rev. 30b.Those portions of the reactor coolant auxiliary systems that form the reactor coolant letdown and makeup loopc.Reactor containment cooling spray systemsd.Emergency core cooling system in cluding injection and recirculation modese.Containment air purification and cl eanup 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 out side the containment structuref.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 systemh.Hydrogen recombiner system Safety Class 3 Safety Class 3 applies to:1.Components necessary to the safety system function of the following:a.Portions of the reactor auxiliary sy stems that provide boric acid for the reactor coolant letdown and makeup loopb.Portions of the auxiliary feedwater systemc.Portions of the reactor plant compone nt 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; al so cooling systems for the control room and safety related electrical equi pment. If a heat transfer component is classified as Safety Class 2 or 3 ju st 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 MPS3 UFSAR3.2-4Rev. 30e.On site emergency power supply support systems external to the emergency generators (the emerge ncy generators are defined in IEEE-387-1977)f.Portions of the main steam system that supply steam to the turbine drive of the steam generator a uxiliary feedwater pumpg.Air purification and cleanup systems used to clean up the atmosphere after leakage from the containment struct ure and other air purification and cleanup systems used after accidents2.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 fo llowing systems, the failure of which would result in uncontrollable release to the environmen t 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 2b.Portions of the radioactive wa ste processing and handling systems Nonnuclear Safety Class This class applies to portions of the unit not covered in Safety Classes 1, 2, or 3. This class includes most of the steam and power conversion systems, radio active liquid waste system, and portions of the boron recovery sy stem containing degassed liquid.
Quality Group Classification System The quality group classification system and its relation to i ndustrial codes conform with Regulatory 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 safe ty (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 acc eptable as the boundary between the reactor coolant pressure boundary and lowe r safety class or NNS line.
Millstone 3 has constructed components in safety-related systems to the ASME Boiler and Pressure Vessel Code, Section III, Nuclear Po wer Plant Components, Division I, as follows:
MPS3 UFSAR3.2-5Rev. 301.Quality Group A (Section III, Class 1) components within the reactor coolant pressure boundary comply with Section 50.55a, 10 CFR 502.Quality Group B (Section III, Class 2) components comply with the requirements of Subsection NA-2130 of the code3.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.The boundary of jurisdiction of ASME Code Section III, Classes 2 and 3 process piping extends to and includes the root valve. The appropriate safety class extends from the root valve to the sensing instrument. Seismic Category I supports are employed for Safety Classes 2 and 3 instrument tubing. The tubing used is one-half inch or less in diameter. The requirements for Safety Classes 2 and 3 instrument tubing are listed in Tables 3.2-2, 3.2-3 and 3.2-4.The safety classes of safety-related fluid systems are given in Table 3.2-1. In addition, the safety class boundaries are shown on the various piping and instrumentation diagrams (P&IDs) located throughout this safety analysis report. The following line designations are used on P&IDs to indicate these boundaries:Safety Class 1, SC-1 (Quality Group A) = line designator, - 1Safety Class 2, SC-2 (Quality Group B) = line designator, - 2Safety Class 3, SC-3 (Quality Group C) = line designator, - 3
 
Nonnuclear Safety Class, NNS (Qua lity Group D) = line designator - 4The QA classification process for the identification of structures, systems, and components as nuclear safety-related and non safety-related is controlled via an Engineering Design Specification. The methodology for system identifications, system interf aces, line designators, MPS3 UFSAR3.2-6Rev. 30safety class boundaries, and identification of nuclear safety related flow paths, equipment and instrumentation is defined and controlled via an Engineering De sign Specification(s).Safety class boundaries on FSAR figures (P&IDs) are extended to the first piping restraint beyond the indicated boundary.
 
====3.2.3 QUALITY====
ASSURANCE CATEGORIESTable 3.2-1 lists Millstone 3 structures, systems, and components which are classified QA Category I.
 
====3.2.4 OTHER====
CLASSIFICATION SYSTEMSTornado Design Classification The tornado design classification conforms to Regulatory Guide 1.117 (Table 1.8-1). Table 3.2 1 lists the tornado protection criteria for Cate gory I structures, systems, and components.
Section 3.8.4.1 describes those structures which are not safety-related but whose failure could reduce to an unacceptable safety level the functional capability of any plant feature included in the items listed in the appendix of Regulatory Guide 1.117.
ASME Code Classes ASME Code Classes 1, 2, 3, and MC are defined in the ASME Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Components, 1971. With regard to pumps, valves, piping, tanks, and pressure vessels, there is a direct one-for-one correlation between Code Classes 1, 2, and 3, and SC 1, 2, and 3 (Sec tion 3.2.2). Exceptions to this are noted on the engineering document. The Codes for the concrete portions of the containment structure, classified as SC 2, are specified in Section 3.8. Metal containment system s, such as the personnel access lock, are classified as Code Class MC. Paragraphs NE 1100 and NE 1140 of ASME Section III delineate the portions of the containment structure classified as Code Class MC.
The code classes of structures, systems, and components are given in Table 3.2 1.
Engineered Safety Features Engineered safety features (ESF) are those systems used to directly mitigate the consequences of a major loss-of-coolant accident, up to and including the design basis accident, which is the double ended rupture displacement of the largest pipe in the reactor coolant pressure boundary. The following types of systems are classified as ESF:*containment systems, including containm ent structure and containment enclosure building;*ESF actuation systems; MPS3 UFSAR3.2-7Rev. 30*emergency core cooling system;*containment heat removal systems;*containment combustible gas control systems;*containment isolation systems; and*supplementary leak collection and release system.It should be noted that systems supporting the above (e.g., cooling systems and electrical systems) are not classified as ESF, but are safety-related (QA Category I).
IEEE Classification Systems IEEE Std-308-1971, IEEE Standard Criteria for Class 1E Electric Systems for Nuclear Power Generating Stations, delineates three classes of structures and equipment (Classes I, II, and III) and one class of electric systems (Class 1E). Class 1E electric systems provide the electric power used to shut down the reactor and limit the release of radioactive material following a design basis event (i.e., postulated events used in the design to establish the performance requirements of the structures and systems). Class II and III electric equipment may be supplied from Class 1E electric systems.
 
====3.2.5 TABULATION====
OF CODES AND CLASSIFICATIONSThis subsection provides a concise compilation of the safety classes, codes, and design classifications of the structures, systems, and components in Table 3.2-1 that are QA Category I.QA Category I structures, systems, and components are defined in Table 3.2-1. Seismic Category I structures, systems, and components are defined in Section 3.2.1 and are designed in accordance with the seismic design criteria of Sections 3.7, 3.8, and 3.10. Becaus e the definitions of Seismic Category I and QA Category I are different, the various structures, systems, and components falling in one category do not necessarily fall in the other. QA Category I items which are not Seismic Category I are pointed out in the notes column of Table 3.2-1.The safety class, as defined in Section 3.2.2, is indicated for each QA Category I structure, system, and component to which it applies. The codes applicable to the QA Category I components are also presented; in addition, Tabl e 3.2-1 indicates which structures, systems, and components are designed for tornado resistance and are protected from tornado effects.
MPS3 UFSARMPS3 UFSAR3.2-8Rev. 30TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (Symbols and references are defined at the end of this Table) ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotesSTRUCTURESContainment Structure Reinforced Concrete Substructure2ACI 318-71N/ACSP Reinforced Concrete Superstructure2ACI 318-71N/ACSDContainment Enclosure BuildingN/AACI 318-71N/ACSDContainment Enclosure BuildingN/AAISC Steel Construction Manual, 7th EditionN/ACEBN/AReinforced Concrete Interior Shields and Walls2ACI 318-71N/ACSPContainment Structure Liner2ASME IIIMCCSP(Note 2)Piping and Duct Penetrations2ASME III2CSP(Note 2)Electrical Penetrations2IEEE-317N/ACSPPersonnel Access Lock2ASME IIIMCCSP(Note 2)
Equipment Hatch2ASME IIIMCCSP(Note 2) (Note 3)
Containment Dome Closure2ASME IIIMCCSD(Note 2)Containment Sump Strainer2AISC 6th EditionN/ACSP MPS3 UFSARMPS3 UFSAR3.2-9Rev. 30Circulating Water Discharge Tunnel N/AACI 318-71N/AOYDDemineralizer Water Storage Tank Enclosure N/AACI 318-71N/AOYDCable Tunnel from Auxiliary Building to Control Building N/AACI 318-71N/ASBDMain Steam Valve Building N/AACI 318-71N/AMSVD Engineered Safety Features Building N/AACI 318-71N/AESBD Auxiliary Building Reinforced Concrete StructureN/AACI 318-71N/AABDMCC and Rod Control AreaN/AACI 318-71N/AABP
 
Hydrogen Recombiner Building N/AACI 318-71N/AHRBD Fuel Building Reinforced Concrete StructureN/AACI 318-71N/AFBD (Note)Capital new fuel pool, spent fuel pool, pipe tunnel below grade, and the fuel pool cooler and pump area only will be Seismic
 
Category I and protected from tornado missiles.TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-10Rev. 30Steel Roof StructureN/AACI 318-71N/AFBD (Note)Not missile protected. Designed to withstand tornado winds only (Section 3.8.4.3).
Spent Fuel Pool Liner (including Gates)N/AASME VIIIN/AFBPSpent Fuel Pool Racks3ASME III3FBP Control BuildingN/AACI 318-71N/ACBDEmergency Generator EnclosuresN/AACI 318-71N/AEGEDCirculating and Service Water PumphouseN/AACI 318-71N/ACSPD (Note)Se rvice water pump cubicles only.West Retaining WallN/AACI 318-71N/ACSPDEmergency Generator Fuel Oil Transfer Pump VaultN/AACI 318-71N/AOYD SYSTEMS Reactor Coolant SystemReactor Vessel1ASME III1CSPTABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-11Rev. 30Reactor Internals, Core Supports1N/AN/ACSPReactor core supports and internals are designed to the intent of Subsection NG of ASME III.Reactor Head Vent Piping and Valves1ASME III1CSPFuel Assemblies1N/AN/ACSP Full Length Control Rod Drive Mechanism Housing1ASME III1CSP Part Length Control Rod Drive Mechanism Housing1ASME III1CSPCRDM Pressure Vessel1ASME III1CSPCRDM Latch AssemblyNNSN/AN/ACSPCRDM Drive Rod AssemblyNNSN/AN/ACSP Control Rods1N/AN/ACSPSteam GeneratorTube Side1ASME III1CSP Shell Side2ASME III2CSPReactor Coolant Stop Valves1ASME III1CSPPressurizer1ASME III1CSPTABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-12Rev. 30Pressurizer Heaters (Groups A&B)1No CodeN/ACSPPressurizer Spray Valves and Piping1ASME III1CSP Reactor Coolant Hot and Cold Leg Piping, Fittings, and Fabrication1ASME III1CSPSurge Pipe, Fittings, and Fabrication1ASME III1CSPLoop Bypass Line1ASME III1CSPReactor Coolant Thermowell1ASME III1CSP Reactor Coolant Thermowell Boss1ASME III1CSPSafety Valves1ASME III1CSPRelief Valves (PORV)1ASME III1CSPActuators are qualified to IEEE-323-74.Power Operated Block Valves1ASME III1CSPActuators are qualified to IEEE-323-74.Valves to Reactor Coolant System Boundary1ASME III1CSP Control Rod Drive Mechanism Head Adapter Plugs1ASME III1CSP Reactor Coolant PumpReactor Coolant Pump Casing1ASME III1CSPTABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-13Rev. 30Main Flange1ASME III1CSPThermal Barrier1ASME III1CSPNo. 1 Seal Housing1ASME III1CSPNo. 2 Seal Housing2ASME III1CSPN
: o. 2 Seal Housing is permitted to be Code Class 2; however, it is supplied as Code Class 1 by Westinghouse.No. 3 Seal Housing2ASME III2CSPPressure Retaining Bolting1ASME III1CSPReactor coolant pump seal bolting is NSS.Reactor Coolant Pump SealsN/AN/AN/ACSPSpecial requirements are included in the specifications.Reactor Coolant Pump MotorRCP motor is not safety-related.Shaft Coupling2NEMA MG1N/ACSPSpool Piece2NEMA MG1N/ACSPArmature2No CodeN/ACSP Flywheel2No CodeN/ACSPMotor Bolting2No CodeN/ACSP Upper Oil CoolerTABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-14Rev. 30Tube Side-Component Cooling Water3ASME III3CSShell Side-oil2ASME III2CSP Lower Oil CoolerTube Side-Component Cooling Water3ASME III3CSPShell Side-oil2ASME III2CSPAir-Water Coolers3ASME III3CSPPiping and Valves***See NoteCSPAll 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 boundary is ASME III 2 or NNS.
Instrumentation and Controls required to perform safety function in QA Category I portions of system***N/AIEEE-279-71N/ACS / CR / IRR / MRC P Inadequate Core Cooling InstrumentationN/AIEEE-323-74N/ACS / CRPTABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-15Rev. 30 IEEE-344-75Subcooled/Superheat Margin Monitor Core Exit Thermocouple
 
Heated Junction Thermocouple Core Exit Thermocouple Pressure Boundary/
Reactor Internal Modifications for Inadequate Core Cooling Instrumentation1ASME III1CSP Supports for QA Category I Components*Same as componen t being supported.Chemical and Volume Control System Regenerative Heat Exchanger**Tube Side2ASME III2CSPShell Side2ASME III2CSP Letdown Heat Exchanger**Tube Side2ASME III2ABPShell Side3ASME III3ABPMixed Bed Demineralizer***3ASME III3ABPTABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-16Rev. 30 Cation Bed Demineralizer**3ASME III3ABPReactor Coolant Filter**2ASME III2ABPVolume Control Tank**2ASME III2ABPCentrifugal Charging Pump**2ASME III2ABP Seal Water Injection Filter**2ASME III2ABPLetdown Orifices**2ASME III2CSP Excess Letdown Heat Exchanger**Tube Side2ASME III2CSPShell Side3ASME III3CSPSeal Water Return Filter**2ASME III2ABP Seal Water Heat Exchanger**Tube Side2ASME III2ABPShell Side3ASME III3ABPLetdown Filter3ASME III3ABPBoric Acid Tanks *3ASME III3ABPBoric Acid Transfer Pump**3ASME III3ABP Boric Acid Blender**3ASME III3ABPBoric Acid Filter**3ASME III3ABPTABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-17Rev. 30Boric Acid Tank Orifice*3ASME III3ABPInstrumentation and ControlsSee NoteAB / CSPPressure retaining portions of inline instruments are the same as connecting piping.
Moderating Heat Exchanger**Tube Side3ASME III3ABPShell Side3ASME III3ABP Letdown Chiller Heat Exchanger**Tube Side3ASME III3ABPShell SideNNSASME VIIIN/AABP Letdown Reheat Heat Exchanger**Tube Side2ASME III2ABPShell Side3ASME III3ABP Thermal Regeneration Demineralizer**3ASME III3ABPPiping and Valves inside RCPB*
No. 1 Seal Water Injection Lines1ASME III1CSPFigures 9.2-5 (P&ID 105), 9.3-7 (P&ID 103), and 9.3-8 (P&ID 104) delineate SC boundaries.TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-18Rev. 30Piping and Valves outside RCPB*See NoteSame references as inside RCPB.Letdown and charging lines2ASME III2AB / CSPCharging pump alternate minimum flow lines
 
downstream of isolation valves 3CHS*MV8512 A/B to the RWST are ANS Safety Class 4.Seal water injection and return lines2ASME III2AB / CSP Mixed-bed and cation-bed lines3ASME III3ABPBoric acid lines3ASME III3ABP Thermal regeneration lines3ASME III3ABPEmergency boration2ASME III2ABP Supports for QA Category I Components*Same as componen t being supported.Residual Heat Removal System Residual Heat Removal Pump**2ASME III2ESBPResidual Heat Exchanger**Tube Side2ASME III2ESBPTABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-19Rev. 30Shell Side3ASME III3ESBPPiping and Valves inside RCPB*1ASME III1CS / ESBPFigure 5.4-5 delineates SC boundaries.Piping and Valves outside RCPB*2ASME III2CS / ESBPFigure 5.4-5 (P&ID 112) delineates SC boundaries.Valve interlocks for over-pressurization protectionN/AIEEE-279-71N/ACS / ESB /
IRR / ESR /
 
CRPSection 7.6.2.2 for criteria comments.Supports for QA Category ISame as components being supported.Emergency Core Cooling System Accumulators**2ASME III2CSPSafety Injection Pumps**2ASME III2ESBPPiping and Valves inside RCPB*1ASME III1CSPFigures 5.4-5 (P&ID 112), 6.3-2 (P&ID 113), and 9.2-4 (P&ID 114) delineate SC boundaries.Piping and Valves outside RCPB*F igures same as inside RCPB.
High and low pressure safety injection lines2ASME III2CS / ESBP Supports for QA Category I Components *Same as componen t being supported.TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-20Rev. 30 Hydrogen Recombiner System
*Hydrogen Recombiner Blower2ASME III2HRBPPreheater Coil2ASME III2HRBPAir Cooler2ASME III2HRBP Thermal Recombiner2ASME III2HRBPPiping and Valves2ASME III2CS / HRBPFigure 6.2-36 (P&ID 115) delineates SC boundaries.
Instrumentation and Controls required to perform safety functionN/AIEEE-279-71N/AHRBP Supports for QA Category I Components *Same as componen t being supported.
Quench Spray System
*Refueling Water Storage Tank (RWST)2ASME III2OYN/A Quench Spray Pumps2ASME III2ESBN/APiping and Valves, excluding RWST recirculation lines and test piping (NNS)2ASME III2ESB / CSN/AFigure 6.2-36 (P&ID 115) delineates SC boundaries.TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-21Rev. 30 Instrumentation and Controls required to perform safety function in QA Category I portions of systemN/AIEEE-279-71N/AOY / ESB /
IRR / ESR /
CR P IEEE-323-74 IEEE-336-71 Supports for QA Category I Components*Same as componen t being supported.
Containment Recirculation System
*Containment Recirculation Pumps2ASME III2ESBP Containment Recirculation CoolersTube Side3ASME III3ESBP Shell Side2ASME III2ESBPPiping and Valves, excluding test piping2ASME III2ESB / CSPFigure 5.4-5 (P&ID 112) delineates SC boundaries.
Instrumentation and Controls required to perform safety function in QA Category I portions of systemN/AIEEE-279-71N/ACS / E /
IRR / ESR /
ESB / CR P IEEE-323-74 IEEE-336-71 Supports for QA Category I ComponentsSame as componen t being supported.TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-22Rev. 30Emergency Generator Fuel Oil System
*Emergency Generator Fuel Oil Day Tanks3ASME III3EGEPEmergency Generator Fuel Oil Transfer Pumps3ASME III3OYPLocated underground in emergency generator fuel oil transfer pump vault.Emergency Generator Fuel Oil Storage Tanks3ASME III3OYPUnderground vault.Piping and Valves3ASME III3OY / EGEPFigure 9.5-2 (P&ID 117) delineates SC boundaries.
Instrumentation and Controls required to perform safety function.N/AIEEE-279-71N/AEGE / OY /IRR /  ESR / CR P IEEE-323-74 IEEE-336-71 Supports for QA Category I Components*Same as componen t being supported.Emergency Diesel Engine Air Start System
*Air Receiver Tanks3ASME III3EGEPPiping and Valves3ASME III3EGEPTABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-23Rev. 30 Instrumentation and Controls required to perform safety functionN/AIEEE-279-71N/AEGE /
OY / IRR /
ESR / CR P IEEE-323-74 IEEE-336-71 Supports for QA Category I Components*Same as componen t being supported.Emergency Diesel Engine Jacket Water Cooling System*Fresh Water Expansion Tank3ASME III3EGEPPiping and Valves3ASME III3EGEP Instrumentation and Controls required to perform safety functionN/AIEEE-279-71N/AEGEP IEEE-323-74 IEEE-336-71 Supports for QA Category I Components*Same as componen t being supported.Emergency Diesel Engine Exhaust and Combustion Air System
*Air Piping3ASME III3EGEP Supports for QA Category I Components*Same as componen t being supported.TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-24Rev. 30Emergency Diesel Engine Lube Oil System
*Piping3Mfgr StdSee NoteEGEPEquivalent to ANSI B31.1. See NRC Question Q430.73.Valves3ASME III3EGEP Lube Oil Heat Exchanger3ASME III3EGEP 3EGO*E3A,BLube Oil Strainers 3ASME III See Note3EGEPAdditional strainer to be purchased/installed ASME III, or if not available, ASME VIII.
3EGO*STR1A,B 3EGO*STR2A,BLube Oil Filter3ASME III3EGEP3EGO*FLT1AHeater3ASME VIIIN/AEGEP 3EGO*H1A,BLube Oil and Pre-Lube Pump3Mfgr StdN/AEGEPASME material 3EGO*P3A,B 3EGO*P4A,B Rocker Arm Lube Oil System
*TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-25Rev. 30 Rocker Arm Lube Oil and Prelube Pump3Mfgr StdN/AEGEP 3EGO*P2A,B 3EGO*P1A,BPiping3Mfgr StdSee NoteEGEPEquivalent to ANSI B31.1. See NRC Question Q430.73.Valves3ANSI B16.5N/AEGEP Filter3Mfgr StdN/AEGEP3EGO*FLT2A,B Fuel Pool Cooling and Purification System
*Fuel Pool Cooling Pumps3ASME III3FBPFuel Pool Coolers3ASME III3FBPPiping and Valves required for cooling3ASME III3FBPFigure 9.1-6 (P&ID 111) delineates SC boundaries.Service Water Piping for Emergency Makeup to Fuel Pool3ASME III3FBPFigure 9.1-6 (P&ID 111) delineates SC boundaries.
Instrumentation and Controls required to perform safety function in QA Category I portions of systemN/AIEEE-279-71N/AFB / CR /
ESRPPressure retaining portions of inline instruments are the same as connecting piping.
IEEE-323-74 IEEE-336-71TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-26Rev. 30 Supports for QA Category I ComponentsSame as componen t being supported.
Containment Isolation System
*Containment isolation valves and associated piping for all system penetrating containment structure.2ASME III2CS / FB / MSV / AB /
ESB / HRBPIndividual fluid system figures delineate SC boundaries.
Instrumentation and Controls required to perform safety functionN/AIEEE-279-71N/ACS / FB / MSV / AB /
CR / IRR / ESB / HRB P IEEE-323-74 IEEE-336-71SupportsSame as compon ent being supported.Service Water System
*Service Water Pumps3ASME III3CSPPService Water Strainer3ASME III3CSPP Piping and Valves supplying cooling water to QA Category I equipment3ASME III3 AB / ESB /
EGE / OY (Buried) / CR
 
/ CSPPFigure 9.2-1 (P&ID 133) delineates SC boundaries.
Instrumentation and Controls required to perform safety function in QA Category I portions of system.N/AIEEE-279-71N/ACSP / ESB / EGE / CR /
 
IRR / ESR P IEEE-323-74 IEEE-336-71TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-27Rev. 30 Supports for QA Category I ComponentsSame as componen t being supported.Reactor Plant Component Cooling Water Subsystem
*Component Cooling Pumps3ASME III3ABPComponent Cooling Surge Tank3ASME III3ABPComponent Cooling Heat Exchangers3ASME III3ABP Piping and Valves supplying cooling water to QA Category I equipment3ASME III3AB / CS /
ESB / FBPFigure 9.2-2 (P&ID 121) delineates SC boundaries.
Instrumentation and Controls required to perform safety function in QA Category I portions of systemN/AIEEE-279-71N/AAB / CS /
CR / IRR /
ESB /  ESR / FB P IEEE-323-74 IEEE-336-71 Supports for QA Category I ComponentsSame as componen t being supported.Neutron Shield Tank Cooling System
*Neutron Shield Tank1N/AN/ACSPClassified as SC-1 because portions support reactor vessel, not because of shielding function.Charging Pumps Cooling System
*TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-28Rev. 30Charging Pumps Cooling Pumps3ASME III3ABPCharging Pumps Coolers3ASME III3ABPCharging Pumps Surge Tank3ASME III3ABPPiping and Valves3ASME III3ABP
 
Instrumentation and Controls required to perform safety functionN/AIEEE-279-71N/AAB / CR /
IRR P IEEE-336-71 IEEE-323-74 Supports for QA Category I Components*Same as componen t being supported.Safety Injection Pumps Cooling System
*Safety Injection Pumps Coolers3ASME III3ESBP Safety and Injection Pumps Cooling Pumps3ASME III3ESBPSafety Injection Pumps Surge Tank3ASME III3ESBPPiping and Valves3ASME III3ESBP Instrumentation and Controls required to perform safety functionN/AIEEE-279-71N/AESB / CR / IRR P IEEE-323-74 IEEE-336-71TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-29Rev. 30 Supports for QA Category I Components*Same as componen t being supported.
Fuel Handling System (Note 3)Reactor Vessel Head Lifting Device1 & NNSN/AN/ACSPPortions that furnish support to Control Rod Drive Mechanism are SC-1.Refueling MachineNNSCMMA Spec #70N/ACSPSpent Fuel Shipping Cask TrolleyNNSCMMA Spec #70N/AFBPSpent Fuel Assembly Handling ToolNNSN/AN/AFBPSpent Fuel Pit Bridge and Hoist3N/AN/AFBPFuel Transfer Tube and Flange2ASME IIIN/AFB / CSPPortions of Containment BoundaryFuel Basket3N/AN/AFB / CSPProtects fuel during transportation from damage.Drive Mechanism and ControlsNNSN/AN/AFB / CSP Supports for QA Category I Components*Same as componen t being supported.Main Steam System TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-30Rev. 30Main Steam Piping and Valves from steam generators up to and including main steam isolation trip valves and isolation valves in main steam supply
 
lines to auxiliary feed pump turbines*2ASME III2CS / ESB /
MSVPFigure 10.3-1 (P&ID 123) delineates SC boundaries.Main Steam Piping and Valves from isolation valves to the steam generator
 
auxiliary feed water pump turbine *3ASME III3ESBPFigure 10.3-1 (P&ID 123) delineates SC boundaries.Main Steam Safety Valves2ASME III2MSVPMain Steam Pressure Relieving Valves and Bypass Valves2ASME III2MSVPActuators are qualified to IEEE-323-74.Main Steam Piping from main steam isolation valves to turbine building3ASME III3MSVPFigure 10.3-1 (P&ID 123) delineates SC boundaries.Main Steam Flow Restrictors2ASME III2CSP Instrumentation and Controls required to perform safety function in QA
 
Category I portions of system***N/AIEEE-279-71N/AMSV / CR / IRR P IEEE-323-74
 
IEEE-336-71 Supports for QA Category I Components*Same as componen t being supported.
Auxiliary Feedwater System
*TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-31Rev. 30Demineralized Water Storage Tank (DWST)3ASME III3OYPSteam Generator A uxiliary Feedwater Pumps (Turbine- and Motor-Driven)3ASME III3ESBPPiping and Valves supplying auxiliary feedwater from DWST to steam generator auxiliary feedwater isolation
 
valves3ASME III3OY / ESBPFigure 10.4-6 (P&ID 130) delineates SC boundaries.Piping and Valves from steam generator auxiliary feedwater isolation
 
valves to steam generator feedwater lines2ASME III2ESB / CSPFigure 10.4-6 (P&ID 130).
Instrumentation and Controls required to perform safety function in QA Category I portions of systemN/AIEEE-279-71N/AESB / CR / IRR / OY /
CS P IEEE-323-74 IEEE-336-71 Supports for QA Category I ComponentsSame as componen t being supported.
Feedwater System TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-32Rev. 30Steam Generator Feedwater Piping and Valves inside containment structure up to and including the first restraint beyond isolation valve outside
 
containment structure *2ASME III2MSV / CSPFigure 10.4-6 (P&ID 130) delineates SC boundaries.Feedwater Piping and Valves from the first restraint beyond isolation valve outside containment to turbine building wall3ASME III3MSVPFigure 10.4-6 (P&ID 130) delineates SC boundaries.
Instrumentation and Controls required to perform safety function in QA Category I portions of system***N/AIEEE-279-71N/AMSV / CR /
CS P IEEE-323-74 IEEE-336-71 Supports for QA Category I Components*Same as componen t being supported.Steam Generator Blowdown System Piping and Valves from steam generator to containment isolation valves2ASME III2CS / MSVPFigure 10.3-1 (P&ID 123) delineates SC boundaries.
Reactor Plant Sampling Systems
*TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-33Rev. 30Piping and Valves for reactor coolant loop, pressurizer, and safety injection accumulator sampling up to and including first isolation valve outside
 
containment structure2ASME III2CS / ABPFigure 9.3-2 (P&ID 144) delineates SC boundaries.
Instrumentation and Controls in QA Category I portions of systemN/AIEEE-279-71N/ACS / AB/
CR / IRR P IEEE-323-74
 
IEEE-336-71 Sample lines originating from safety-related components, up to and
 
including remotely operated sample selection valve or second manual isolation valves2 or 3ASME III2 or 3AB / ESBPFigure 9.3-2 (P&ID 144) delineates SC boundaries.
Supports for QA Category I Components*Same as componen t being supported.
Reactor Plant Aerated Drains Instrumentation and Control (sump level indication) required to provide leak detectionN/AIEEE-279-71N/AAB / ESBP IEEE-336-71 IEEE-323-74TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-34Rev. 30 Porous Concrete Groundwater Sump, Piping & Containment Recirculation Cubicle Sumps3ANSI B31.1N/AESBPFigure 9.3-6 (P&ID 106D) delineates SC boundaries Containment Monitoring System Instrumentation and Controls required to monitor hydrogen concentrationN/AIEEE-279-71N/ACSP IEEE-336-71 IEEE-323-74Heating, Ventilation, and Air Conditioning System
*ESF BuildingAll air-conditioning systems3SMACNA ARI, AMCAASME IIIESBPASME III, Class 3 for service water side of refrigeration condenser.Emergency Ventilation System for mechanical equipment room and auxiliary feedwater pump
 
room3SMACNA, AMCAN/AESBPEmergency Generator EnclosureVentilation (except normal exhaust fan)3SMACNA, AMCAN/AEGEPTABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-35Rev. 30 Control BuildingAll air-conditioning systems, including control building chilled water system3SMACNA, AMCAASME IIICBPASME III, Class 3 for chilled water and service water pressure-containing
 
components.
Battery rooms ventilation and chiller room ventilation (excluding the control room kitchenette and toilet exhausts)3SMACNA, AMCAN/ACBPControl Room PressurizationControl room post-accident dose analyses do not credit this system or its components.Piping3ANSI B31.1N/ACBPTanks3ASME VIIIN/ACBPValves3ANSI B31.1N/ACBPControl Room Emergency Ventilation3Note AN/ACBPSee ESF Filter Systems, below.
Auxiliary BuildingCharging pumps and service water pumps cubicles ventilation (excl uding auxiliary building exhaust filtration
 
system)3SMACNA, AMCAN/AABPTABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-36Rev. 30 Auxiliary building exhaust filtration3Note AABPSee ESF Filter Systems, below.MCC and rod control area air-conditioning3SMACNA, AMCAASME IIIABPASME III, Class 3 for service water coils.
Hydrogen Recombiner Building Hydrogen recombiner cubicles ventilation3SMACNA, AMCAN/AHRBPMain Steam Valve BuildingMain Stream Valve building ventilation (except normal
 
exhaust fans)3SMACNA, AMCAN/AMSVPYard StructuresService Water Pumphouse ventilation3SMACNA, AMCAN/ACSPP Fuel BuildingFuel Building Filtration System3Note AABPA. See ESF Filter Systems, below. This system is not credited for post-accident analyses.Supplementary Leak Collection and Release System (SLCRS)3Note AAB(Note B)B. Only portions inside tornado protected buildings.TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-37Rev. 30 Instrumentation and Controls required to perform safety function for QA Category I ventilation systemsN/AIEEE-279-71N/AESB / EGE / CB / CSP / ABPOnly portions required to mitigate effects of LOCA will be Seismic Category I.
IEEE-336-71 Supports for QA Category I Components*Same as componen t being supported.
Fuel BuildingFuel Building Exhaust Filtration3See NoteSee ESF Filter Systems, below.
This system is not credited for post-accident analyses.
ESF Filter Systems and SLCRS Fans3AMCAN/ANote CNote CC. For location and tornado criteria, see individual system listing.Motors3NEMA, IEEE-323, 344, 334N/ANote CNote CFilter Trains3ANSI N509N/ANote CNote CDuct3SMACNAN/ANote CNote CDampers3AMCAN/ANote CNote CTABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-38Rev. 30 Filter segment drainage up to and including the isolation valves3ANSI B31.1N/AABNote CAppli cable only to Auxiliary Building Ventilation Filters, Fuel Building Ventilation Filters and SLCRs.Electrical Systems
*Emergency Generators, including AuxiliariesN/AIEEE-308-74N/AEGEP IEEE-323-74Unit Batteries and ChargersN/AIEEE-308-74N/ABRP IEEE-323-74Vital Bus and InvertersN/AIEEE-308-74N/AESRP IEEE-323-74Emergency Unit SubstationsN/AIEEE-308-74N/AESR /
MAC P IEEE-323-74Emergency Station Service SwitchgearN/AIEEE-308-74N/AESRP IEEE-323-74Emergency Motor Control CentersN/AIEEE-308-74N/AESR / MRC / EGE / CSP / ESB P IEEE-323-74Electric Motors-PumpsN/AIEEE-334-74N/AMiscP(Note 1)TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-39Rev. 30Control PanelboardsN/AIEEE-308-74N/A Main control board and panels with safety related functionsCRPDiesel engine driven emergency generators panelRCRPRadiation monitor panelCRPAuxiliary shutdown panelESRP Air-conditioning control panelCRPHydrogen recombiner control panelHRBPTrays, Conduits, and Ducts Carry Safety Related WiringN/AN/AN/ACS / CSP /
AB / CB / FB / EGE / ESB / HRB OY / MSV /
MRC P Class 1E AC Instrumentation, Control and Power Cables - Essential Buses (orange/purple trains)N/AIEEE-323-741ECS /CSP AB / ESF / CB / FB / HRB / OY / MSV/
MRC / EGE P IEEE-383-74TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-40Rev. 30Class 1E DC Power CablesN/AIEEE-323-741E CS/ CSP / AB / ESF / CB / FB /
MSV / EGE
 
/ HRB P IEEE-383-74Class 1E dc Switchgear, Distribution Panels, and Protective RelaysN/AIEEE-323-741ECB / CSPP IEEE-344-75Emergency Lighting Battery Pack SupportsN/AMfgr StdSee NoteCS / AB CSP / ESF /
FB / HRB /
MSV / CB /
 
EGEPSeismic Category I only.Reactor Trip System
***All portions of reactor trip system which must operate to safely shut down reactor to hot subcritical conditionN/AIEEE-279-71N/ACS / CR / IRRPIncludes instrumentation and control components from and to turbine impulse pressure transmitters.Incore Instrumentation System
***Instrumentation and Conduit Tubes1ASME III1 (Note)CSPDesign Basis allowed use of Subsection NC 3600 Bottom Mounted Instrumentation Thimble Tubes2Mfgr. Std.N/ACSPTABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-41Rev. 30 Engineered Safety Features Actuation System ***N/AIEEE-279-71N/ACS / CRP Miscellaneous
*Containment Structure Polar CranN/AN/AN/ACSPDesigned for earthquake in unloaded condition.
LEGEND General Symbols Location Symbols N/A - Not applicableAB - Auxiliary buildingSC  - Safety classBR - Batt ery 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 boundaryCS - Containment structure* - SWEC Scope of SupplyCEB -
Containment enclosure building** - WNES Scope of SupplyCSP - Circ ulating and service water pumphouse
*** - Scope of Supply shared between WNES and SWECEGE - Emergency generator enclosure ESB - Engineered safety features buildingTornado Criteria Symbols ESR - Emergency switchgear room (in control building)TABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-42Rev. 30NOTES: Note 1. The mechanical system components sa tisfy the codes and addenda (A SME Section III, Division 1) in effect at the time of component order.
Note 2. There was no applicable code for th e design of concrete containment structure liners at th e 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 missile shield blocks is not required during Modes 5 and 6.Note 4. This FSAR table identifies safety-rel ated pumps for a given system. Unless ot herwise indicated, motors for these safety-related pumps are also safety-related and included under the same safety class.P - Protected from tornado effects by a structure or because below grade FB - Fuel buildingD -Designed to withstand tornado effectsHRB -Hydrogen recombiner building IRR -Instrument rack r oom (in control building)
MRC -MCC and rod control ar ea (in auxiliary building)
MSV -Main steam valve building OY -Outside, yard
 
SB -Service buildingTB -Turbine buildingWDB -Waste disposal buildingTABLE 3.2-1 LIST OF QA CATEGORY I AND SEISMIC CATEGORY I STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)
ANS Safety Class Code (Note 1)
Code ClassLocationTornado CriterionNotes MPS3 UFSARMPS3 UFSAR3.2-43Rev. 30TABLE 3.2-2 SAFETY CLASSES 2 AND 3 INSTRUMENT TUBING REQUIREMENTS ASME III (All Ref. Summer 1973)QA Category I ProgramOrganizationRequiredSame TrainingRequiredSameDesign SpecificationPressure boundary integrity for SSE and dead load, thermal Same, except for code references to certification in specification (1)Engineering, Design, and Document ControlCategory ISameProcurement ControlASME-approved suppliersASME III design and materials, Category I supplier, no N-stamp required Receiving, Inspection, Identification, Storage, and Handling Control Physical inspection and review of documentation; ANSI storage and material identification SameFabrication and Installation ControlCont rol Drawing Package, FQC, and ANI review and established holdpoints; material traceability Same, except no mandato ry holdpoints; no third party documentation review; no individual packages per drawing; normal Category I IR System (Table 3.2-3 and 3.2-4)Field Welding and Brazing ControlASME III Procedures - Weld data package each weld; ASME IX welders (2)Same (Table 3.2-4)Bolted and Other Mechanical JointsData sheet for special bolted jointsNo special bolted joints; mechanical fittings installed to MFG requirements; documented on inspection reportTubing SupportsDesign to AISC, 7th Edition (Refer to NRC Question 210.36 for Details)
Same MPS3 UFSARMPS3 UFSAR3.2-44Rev. 30Heat Treatment and Special Operations and RepairsNot ApplicableSameFabrication and Installation InspectionFQ C, ANI, ASME acceptance; material traceability required of selected components to specific point of installation Same, except limited third party surveillance; Category I material marking or exclusive purchase of Category I material (Tables 3.2-3 and 3.2-4)Nondestructive TestingDye penetrant for Class 2, visual for Class 3; traceability (2)SameNonconformancesN&DSameControl of Measuring and Test EquipmentRequiredSame Authorized Nuclear Inspector and Code CertificationN-stampNot Applicable Quality Assurance Audit ProgramSWEC, ASME, ANI, NUSCOSWEC, NUSCO Company Quality Assurance and Control ManualSWEC QA Program ManualS WEC QA Program Manual SWEC ASME III Control ManualFinal DocumentationAs-built data package FQC-ANI Certification Documented on inspection reports; FQC acceptanceCertificate Holder (Installation Subcontractor)Not ApplicableSamePressure Testing1.25 tim es design pressure FQC/ANI to witness Same SWEC's Responsibiliti es when Owner's Designee Prepare code data forms, N5, N3, ANI witness N-stampNot ApplicableTABLE 3.2-2 SAFETY CLASSES 2 AND 3 INSTRUMENT TUBING REQUIR EMENTS (CONTINUED)ASME III (All Ref. Summer 1973)QA Category I Program MPS3 UFSARMPS3 UFSAR3.2-45Rev. 30NOTES: 1. For tubing which serves a nonsafety-relate d function attached to a Category I proc ess pipe, functional ca pability is not req uired; therefore, the allowable stress values for ASME equation 9 util ize the faulted allowable of 2.4 Sh for all loading conditions. The faulted allowable for this tubing will en sure that the pressure boundary is maintaine d, thereby protecting the safety-related function of the process piping.2. Compression fittings are used exclusively for tubing installa tion except where transition from pipe to tubing is required. T he pipe-to-tubing transition is controlled via piping specificatio n requirements as part of the ASME III piping system SWEC Operations under ASMI Section XIGoverns repair of componentsNot ApplicableTABLE 3.2-2 SAFETY CLASSES 2 AND 3 INSTRUMENT TUBING REQUIR EMENTS (CONTINUED)ASME III (All Ref. Summer 1973)QA Category I Program MPS3 UFSAR3.2-46Rev. 30NOTE: 1. For Class 2, LP is required, for Cl ass 3, visual inspection is required.TABLE 3.2-3 INSTRUMENT TUBING EXAMINATION AND TESTING PROGRAMA.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 ND E, 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).B.Safety Classes 2 and 3 Compression Fittings 1.One hundred percent of fitting make-up by construction prior to release to FQC using vendor's recommended practices and inspection tools (document via construction checklist).2.One hundred percent of fitting make-up by FQC us ing 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 inspec tion of fittings (document via Pressure Test Report).
MPS3 UFSAR3.2-47Rev. 30NOTE: 1. For Class 3, LP is not required by ASME III.TABLE 3.2-4 COMPARISON OF PROPOSED TUBING EXAMINATION AND TESTING WITH ASME III REQUIREMENTSASME Program Classes 2 and 3 QA Category I Exam/Testing Program 1.One hundred percent visual inspection.1.Same2.One hundred percent LP inspection (1)2.Same (1)3.Surveillance by ASME and ANI; approximately 10 percent in process activities3.Surveillance by a third party inspector; approximately 10 percent in-process activities - includes welding and weld
 
hydros4.Hydro - 100 percent inspection by FQC and ANI at 1.25 times design pressure4.Pressure test (Hydro or Pneumatic) -100 percent inspection by FQC at 1.25 times
 
design pressure; ANI to witness.5.Surveillance inspection performed by FQC; i.e., In process Welding, Weld
 
material Control, Material Control5.Same6.All inspection performed, with the exception of Item 5, are documented in the weld data packages; i.e., Weld Data Sheets6.Same7.Welders and procedures to be qualified to ASME IX7.Same MPS3 UFSAR3.3-1Rev. 30 3.3 WIND AND TORNADO LOADINGS 3.3.1 WIND LOADINGS This section discusses the design wind load on Seismic Category I structures.
3.3.1.1 Design Wind VelocityThe maximum wind experienced in the vicinity of the Millstone site was associated with a 1960 hurricane. According to the prel iminary site evaluation for Millstone 1 (TRC 1965), the Montauk Point Coast Guard Station on the tip of Long Island, across Long Island Sound from Millstone Point, recorded a wind speed of 115 mph with gusts to 140 mph. This wind speed is greater than the 1,000 year mean recurrence interval wind speed for the New Haven area (Hollister 1970). Velocity profiles are as give n in ANSI A58.1 - 1972 (ANSI 1972).
All Seismic Category I structures are designed to remain elastic under the 115 mph basic wind speed. The gust factors are included in the determination of effective velocity pressures as described in ANSI A58.1. Because the site is considered to be flat, open country, the 1/7 power law for vertical velocity distribution is used.
3.3.1.2 Determination of Applied ForcesANSI A58.1-1972 (ANSI 1972) is used to develop wind pressures and distributions on structures. The effective external velocity pressures for structures, q f, and for portions thereof, q p, at various heights above ground are in accordance with Tables 5 and 6 of ANSI A58.1 for exposure C. Effective velocity pressures for calculating internal pressure (q m) are in accordance with ANSI A58.1 Table 12. External pressure coefficients (C p) are in accordance with ANSI A58.1 Tables 7 and 10 and internal pressure coefficients (C pi) are in accordance with ANSI A58.1 Table 11.A step function of pressure with height is used. The specifie d resultant design wind pressure at a given height is applied over a height zone defined by one-half the difference in adjacent heights for which the design wind pressures are specified.
The resultant design wind pressure acts normal to the surface of the structure being considered.
 
====3.3.2 TORNADO====
LOADINGS Seismic Category I structures requiring tornado design are listed in Table 3.2-1.
3.3.2.1 Applicable Design ParametersTornado design parameters are in conformance with Regulatory Guide 1.76 (Section 1.8) as described below:Maximum wind speed360 mphRotational speed290 mph MPS3 UFSAR3.3-2Rev. 30Translational speed70 mph (max)Radius of maximum rotational speed150 ftPressure drop3.0 psiRate of pressure drop2.0 psi/secThe maximum wind speed is the sum of the maximum rotational speed component and the maximum translational speed component.Table 3.5-13 lists the tornado generate d missiles and thei r characteristics.
3.3.2.2 Determination of Forces on StructuresTotal tornado load is determined as a result of tornado wind (W w), differential pressure (W p), and missile (W m) loadings.Tornado Wind Load (W w) Tornado wind load is determined as follows:
W w = q C p where: q = External effective velocity pressure C p = External wind pressure coeffici ent for structure being consideredThe external effective velocity pressure (q) is determined in accordance with ASCE Paper No.
3269 (ASCE 1961) as follows:
q = 0.00256 v 2 psf where: v = Applicable tornado wind speed in mphThe external wind pressure coefficient (C p) defines the pressure acting on the surface of the structure. The external pressure coefficients for rectangular shape structures are in accordance with ANSI A58.1. The coefficients for cylindrical shape structures are based on Table 4(f) of ASCE Paper No. 3269 (ASCE 1961). The coefficients for the containment structure dome are based on Wind Stresses in Domes (Gondikas and Salvadori 1960). Coefficients for structural steel shapes are based on ASCE Paper No. 3269 (ASCE 1961).Tornado Differential Pressure Load (W p)
MPS3 UFSAR3.3-3Rev. 30For all reinforced concrete structures, with the exception of the fuel building and the emergency generator enclosure, the exterior walls and roofs were designed for nonvented conditions (full 3 psi pressure drop).For the fuel building, the section of metal roof at elevation 55 feet-3 inches and between column lines G.5 and H is capable of venting the building. For this reason, the design is based on two conditions. The first condition considers the structure to be vented through the area covered by this metal roof. Those sections of walls and floors of interior cubicles which are affected by the venting of the building through the roof area are designed for the full 3 psi pressure drop. For the second condition, it is assumed that this roof does not vent the building. Therefore, the exterior walls and roofs of the building are designe d for the full pressure drop (3 psi).For the emergency generator enclosure, the diesel generator muffler cubicle above elevation 51 feet-0 inch is capable of venting. Therefore, the design of this building is also based on two conditions. The first condition considers the diesel generator muffler cubicle above elevation to be vented by the large openings in the exterior walls of the cubicle. The portions of walls and floors of interior cubicles which are affected by venting of the muffler c ubicle are designed for the full 3 psi pressure drop. For the second condition, it is assumed that this cubicle does not vent and the exterior walls and the roof of the structure affected are designed for the full 3 psi pressure drop.A dynamic load factor of 1.0 is applied to the pressure dr op since all wall and floor panels subject to pressure drop have frequencies greater than 4 Hz as shown in Table 3.3-1. Figure 3.3-1 presents the pressure drop time history and the calculated dynamic load fact or curve for the pressure drop. Comparing the two, it is evident that the Millstone 3 structural elements experience no dynamic amplification during the pressure drop condition.Tornado Missile Load (W m) Section 3.5.3 describes the methods of analysis to determine the impact effects of tornado missiles.Total Tornado Load (W t) The total tornado load to be applied to Seismic Category I structures is determined from the most adverse combination of individual tornado loadi ngs. The loading combinations considered are:W = W wW = W pW = W mW = W w + 0.5 W p W = W w + W m MPS3 UFSAR3.3-4Rev. 30W = W w + 0.5 W p + W m The total tornado load is then combined with other loads as specified in Sections 3.8.1.3, 3.8.3.3, 3.8.4.3, and 3.8.5.3.
3.3.2.3 Effect of Failure of Structures or Components not Designed for Tornado LoadsStructures which do not require Category I design are either located so that structural failure does not affect the ability of safety related structures or systems to perf orm their intended design function, or designed so that they do not collapse under tornado wind load. The metal siding and roofing of the service, turbine, waste disposal, containment enclosure buildings, and portions of the fuel building are assumed to blow off under tornado wind load. Th e resulting siding and roofing missiles are less severe than the missiles described in Section 3.5. The structural steel framing of these structures is designed to withstand tornado wind loads so as not to compromise the integrity of any safety relate d structure, system, or component.
3.
 
==3.3 REFERENCES==
FOR SECTION 3.33.3-1American National Sta ndards Institute (ANSI) 1972. A58.1 1972, Building Code Requirements for Minimum Design Loads in Buildings and Other Structures. New York, N.Y.3.3-2American 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.3.3-3Gondikas, P. and Salvadori, M.C. 1960. Wind Stresses in Domes. Engineering Mechanics Division, American Soci ety of Civil Engineers (ASCE).3.3-4Hollister, S. C. 1970. The Engineering Interpretation of Weather Bureau Records for Wind Loading on Structures. National Bureau of Standards (U.S.), Building Sciences Service, 30.3.3-5TRC Service Corporation 1965. Millstone Unit No. 1, Pr eliminary Site Evaluation.
MPS3 UFSAR3.3-5Rev. 30TABLE 3.3-1 STRUCTURAL PANELS SUBJECT TO TORNADO PRESSURE DROPPANELSIZE (FT)
THICKNESS (FT)FUNDAMENTAL FREQUENCY (HZ)REMARKSMSVB North Wall36x7326.5Simple SupportsAux. Bldg. Roof49.5x6324.6Simple Supports Aux. Bldg. Roof126.5x26.529.0Simple Supports BeamFuel Bldg. Roof59.5x66.524.0Simple Supports Control Bldg.
Roof57.5x10224.2Partial Fixity MPS3 UFSAR3.4-1Rev. 30
 
===3.4 WATER===
LEVEL (FLOOD) DESIGNThis section discusses the flood and the highest groundwater level design for Seismic Category I structures and components. Internal flooding has been addressed as part of equipment qualification requirements and is described in Section 3.11.
 
====3.4.1 FLOOD====
PROTECTION 3.4.1.1 Flood Protection Measures for Seismic Category I StructuresThe design basis flood (maximum combination of storm surge and wave runup) established for Millstone 3 is elevation +23.8 feet msl and the maximum still water level is elevation +19.7 feet (Section 2.4.5). All unit safety-related structures and equipment, except the circulating and service water pumphouse, are protect ed from flooding by the site gr ade of elevation +24 feet msl.Each pair of service water pumps and pump motors is located at elevation +14.5 feet msl inside individual watertight cubicles in the seismically designed pumphous e (Figure 3.4 1). The walls of these cubicles are watertight up to elevation +25.5 feet msl protecting the pump motors and associated electrical equipment from wave action a nd probable maximum hurricane (PMH) surge.All accesses to safety-related structures and facilities are at an elevation of 24 feet-6 inches above the nominal site grade elevation of 24 feet-0 inch and are consequently protected from flooding due to groundwater, storm surge, and direct rainfall, except for the doors that are discussed in Section 2.4.2.3. The two access openings to the service water cubicles inside the pumphouse which are below the flood protection level of +23.8 feet msl are fitted with watertight steel doors capable of withstanding the maximum hydrostatic load occurring at their respective location. Pumphouse roof ventilators are weatherproof and located above the ma ximum accumulation of snow resulting from the occurrence of the 100-year accumulated snow depth of 52 inches.
Equipment access openings on the pumphouse roof over the service wa ter cubicles ar e fitted with watertight covers. During normal plant operation, the service water cubicles have open drain lines installed in the cubicle sump to enable the service water pump seal water leak off to drain directly into the intake structure pump bay. During severe weather or flooding conditi ons, the drain lines are isolated and the service water cubicle sump s are drained using sump pumps 3PBS-P1A and 3PBS-P1B.Foundations of safety-related structures are constructed of reinforced concrete. All subgrade joints between walls and slabs are seal ed with waterstops cast in concrete.The storm drain system uses catch basins and underground conduits and/or drainage ditches to convey runoff to Niantic Bay. Roof and site storm drain syst ems are designed for a maximum precipitation of 6.5 inch per hour. The probable maximum precipitation of 70.4 inch per hour for a 5-minute period (Hansen, et al., 1982) would result in temporary flooding of the site area in the vicinity of the safety-related bui ldings as described in Section 2.4.2.3.
MPS3 UFSAR3.4-2Rev. 30In addition to the storm drains, surface runoff from the higher ground to the north and east of plant perimeter roads is intercepted by open ditches and drained into Niantic Bay. Further discussion on the effects of local intense precipi tation can be found in Section 2.4.2.3.The seaward wall of the intake structure is constructed of reinforced concrete, designed to withstand the forces of a standing wave, or clapotis, with a maximum crest elevation of +41.2 feet msl. Further discussion on the effects of PMH storm surge and wave action, including the resultant pressure distribution on the inta ke wall, can be found in Section 2.4.5.Combinations of the maximum surge level with a coincident wave and the maximum wave height with a coincident surge for three different speed PMH were examined to determine the maximum uplift pressure on the pumphouse floor. The most critical combination is the maximum wave height of 16.2 feet and a surge level of elevation 19.7 feet msl associated with the slow speed PMH.The calculated maximum uplift pressure on the pumphouse floor due to the most critical combination of wave action and storm surge during PMH conditions is 863 lb/sq ft. The floor is designed to withstand more than this pre ssure, precluding the possibility of failure.
The water level fluctuations within the pumphouse, resulting from storm surge and wave action, are dampened by the energy lost in passage through the restricted openings in the trash racks, traveling screens, and operating deck. Internal water level fluctuations are further attenuated because water must enter the structure through a submerged opening (elevation -7 to -30 feet msl) through which the pressure response factor is less than unity.The discharge outfall structure is also designed to withstand maximum wave forces induced by the most critical combination of wave action and storm surge during PMH conditions. The maximum horizontal pressure is determined for the combination of a maximum wave height of 20.1 feet and a maximum surge level of 19.7 feet msl. The maximum vertical pressure on the upper outfall structure is obtai ned for a minimum water submergence at the surge level of elevation 9.46 feet msl and a coincident breaking wave of 14.7 feet. The calculated maximum horizontal pressure and the maximum vertical pressure on the discharge outfall structure for the above-mentioned combination events are 2,325 lb/sq ft and 585 lb/sq ft, respectively.
Section 2.5.5.1 discusses shoreline protect ion in the vicinity of the pumphouse.Flood protection complies with Regulatory Guide 1.102, Flood Pr otection for Nuclear Power Plants, as follows:1.C1 Flood protection is accomplished by th e unit's location on a "Dry Site," with the exception of the circulating a nd service water pumphouse which utilizes "Incorporated Barriers."2.C2 Not applicableRefer to Section 1.8 for clarification to Position C1.
MPS3 UFSAR3.4-3Rev. 30 3.4.1.2 Permanent Dewatering SystemThere is no safety-related dewatering system for lowering groundwater levels for Millstone 3. This system is not applicable. Removal of water which bypasses the membrane installed below the Containment Structure is collected in the Engineered Safety Features Building porous concrete groundwater sump and is removed by the Underdrain System (see Section 9.3.3).
 
====3.4.2 ANALYTICAL====
AND TEST PROCEDURESGround levels of all Category I structures, excep t for the circulating a nd service water pumphouse and the discharge structure, are located above the design basis flood (DBF) level. This level is based on the maximum combination of storm surge due to the PMH and associated wave run-up (Section 3.4.1). Structures located above this level are designed for the hydrostatic effects of uplift and lateral water pressure resulting from the DBF or normal groundwater, whichever is more severe. Groundwater levels are based on piezometric readings taken at the site (Figure 2.5.4-37).The circulating water discharge structure and discharge tunnel and the circulating and service water pumphouse are located below the DBF level.The circulating water discharge structure and discharge tunnel are designed for the hydrostatic and dynamic effects of the DBF as described in Section 3.4.1.
The circulating and service water pumphouse is designed laterally for a standing wave and for uplift on the operating floor due to confined wave action within the pumphouse (Section 3.4.1).
Foundation loadings used in the design reflect saturated soil conditions, where applicable.
The design wind loading described in Section 3.3.1 is applied concurrently with the hydrostatic and dynamic effects of the DBF (Sections 3.8.1.3, 3.8.3.3 and 3.8.4.3). Tornado loading is not applied concurrently with the DBF.
 
====3.4.3 REFERENCE====
FOR SECTION 3.43.4-1Hansen, 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.
MPS3 UFSAR3.5-1Rev. 30
 
===3.5 MISSILE===
PROTECTION
 
====3.5.1 MISSILE====
SELECTION AND DESCRIPTION Systems and components located both inside and outside the containment have been examined to identify and classify potential missiles. Two broad categories of systems and components are reviewed to determine the potential for generating missiles; pressurized components and high speed rotating machinery. Only designs where a single failure could lead to missile ejection are considered. The basic approach to ensure missile protection of syst ems and components both inside and outside of containment involves the following considerations:1.examination of systems in order to identi fy and classify potential missile sources;2.evaluation of the design adequacy of equipment to preclude generation of missiles; and3.evaluation of the effects of the generation of missiles where the potential exists and provisions for protection against them.The objective is to ensure design adequacy against generation of missiles and means of protecting essential structures, systems, and co mponents should a missile be generated.
3.5.1.1 Internally Generated Missil es (Outside Containment)The design bases consider missiles generated outside the containment but internal to the plant site. These shall not cause damage that may affect the safe shutdown or cause radiation release during operating conditions, and postulated accident conditions associated with the effects of missile formation. Table 3.5-1 identifies the safety-related structures, systems, and components outside the containment required for safe shutdown of th e reactor under all conditions of plant operation.Valves in high energy fluid systems are evaluated as potential missile sources. Valves are typically designed with parts which are removable for maintenance. It is these removable parts which present the most significant potential for missile producing failures. Valves provided with back-seated stems are not considered credible sources of missiles. This design feature effectively eliminates the possibility of ejecting valve stems even if the stem threads fail. Valve bonnets are considered credible sources for missiles in cases where the bonnet is bolted to the body. The bonnet and the connection bolts are postulated to be ejected. Valve bonnet missiles are not considered credible where the bonnet is welded to the body, the bonnet is integral with the body, or the bonnet bolts are torqued in a controlled manner. Credit may be taken for air and motor operators which interfere with th e ejection of valve stems and bonnets. The Applicant's review of valves located outside containment has indicated that all the valves in the high energy systems can be categorized as follows.1.Valves that are isolated or enclosed from safety related equipment. These valves do not impose any danger.
MPS3 UFSAR3.5-2Rev. 302.Valves that are not enclosed or isolated. These valves are reviewed for a credible failure mechanism which could eject a miss ile. If a credible failure mechanism is 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.Centrifugal pumps and fans located outside the containment in areas containing safety related components have been evaluated for missiles caused by overspeed or failure. The maximum no-load speed of these centrifugal pumps and fans is equivalent to the ma ximum operating speed of their motors. Consequently, no overspeed is expected and missiles associated with overspeed conditions in centrifugal pumps and fans out side the containment are not postulated.Fans are further evaluated for missile generation under normal operating speeds due to fatigue failure or manufacturing defects. Fan fragments are postulated only where a credible single failure mechanism results in fragmentation. Such fragments have been shown either to lack sufficient energy to penetrate the fan housing or to result in acceptable interactions with essential targets. In this assessment, the fragments are assumed to be unimpeded by any flexib le connections between the fan housing and attached ducting.
The auxiliary feedwater pump turbine is equipped with redundant overspeed detection devices and a regularly tested turbine trip valve, as such overspeed in the turbine is considered credible only up to the trip setting at 10 percent over rated speed. At this speed there exist substantial margins between the energy available in the fragments generated from the turbine-driven pump and the energy required to escape the pump casing; therefore, missiles are not postulated from the pump component. Similarly, fragments generated from the turbine component lack the penetrating geometry and the energy to perforate its casing. As such, neither the pump nor the turbine driver is considered a source of internally generated missiles a nd no essential systems or components can be adversely affected. Nonetheless, the auxiliary feedwater pump turbine is located and oriented within a concrete cubicle to prevent any generated missiles from affecting other safety systems, such as the motor-driven auxiliary feedwater pumps, in adjacent cubicles.The motor-generator that provides power to the control rod drive mechanisms (CRDM) is located outside the containment. The flywheel on this component has been evaluated as a potential missile. The fabrication specifications of the motor-generator-set-flywheel control the material to meet ASTM-A533-70, Grade B, Class I with inspections per MIL-I-45208A and flame cutting and machining operations governed to prevent flaws in the material. Nondestructive testing consisting of nilductility (ASTM-E-208), Charpy V-notch (ASTM-A593), ultrasonic (ASTM-A577 and A578), and magnetic particle (ASME Section III, NB2545) is performed on each flywheel material lot. In addition to these requirements, stress calculations are performed consistent with guidelines of ASME Section III, Appendix A, to show the combined stresses due to centrifugal forces and the shaft interference fit shall not exceed one-third of the yield strength at normal operating speed (1,800 rpm) and likewise, shall not exceed two-thirds of the yield strength at 25 percent overspeed. However, no overspeed is expected for the following reason: the flywheel weighs approximately 1,300 pounds and has dimensions of 35.36 inches in diameter and MPS3 UFSAR3.5-3Rev. 304.76 inches in width. The flywheel, mounted on the generator shaft and directly coupled to the motor shaft, is driven by a 200 hp, 1,800 rpm synchronous motor. The torque developed by the motor is insufficient for overspeed. Therefore, there are no credible missiles from the CRDM motor-generator flywheel.Evaluation of missiles being generated from the emergency gene rator enclosure concluded that there is no need to evaluate missile generation. Safety related emergency generators are located in a structure designed for tornado missile protection; consequently, missiles from the diesel engines are considered unable to penetrate this structure. The essential diesel generator systems are redundant and separated so that a missile generated by one diesel engine will not affect the other. Doors are offset from the generators' axes precluding the possibility of missiles exiting from the doorway of the structure.
3.5.1.2 Internally Generated Missiles (Inside Containment)The design bases are such that missiles generated within the reactor containment will not cause loss of function in any redundant engineered safety feature nor radiation release or damage the containment boundary.
In addition, a missile accident which is not caused by a LOCA shall not initiate a LOCA.
Table 3.5-2 identifies the structures, systems, and components inside the containment whose failure could lead to offsite radiological consequences or which are required for safe plant shutdown to a cold condition assumi ng an additional single failure.
Equipment inside the containment has been evaluated for potential missile generati on. As a result of this review, the following information concerns potential missile sources and systems which require protection from internally generated missiles inside the containment.
3.5.1.2.1 Missile Selection and DescriptionFailure of the reactor vessel, steam generators, pressurizer, and reactor coolant pump casings leading to missile generation are not considered credible because of the combination of material characteristics, inspections, quality control during fabrication, erection, and operation, conservative design, and prudent operation as applied to the particular component.
The reactor coolant pump flywheel is not considered a source of missiles for the reasons discussed in Section 5.4.1. Nuts and bolts are of negligible concern because of the small amount of stored elastic energy.
Centrifugal pumps, fans, and air compressors (centrifugal and axial) located inside the containment have been evaluated for missiles associated with overspeed failure. The maximum no-load speed of these cen trifugal pumps, fans, and air compressors is equi valent to the operating speed of their motors. Therefore, no overspeed is expected and missiles associated with overspeed conditions in centrifugal pumps, fans, or air compressors within the containment are not postulated.
MPS3 UFSAR3.5-4Rev. 30Fans are further evaluated for missile generation under normal operating speeds due to fatigue failure or manufacturing defects. Fan fragments are postulated only where a credible single failure mechanism results in fragmentation. Such fragments have been shown either to lack sufficient energy to penetrate the fan housing or to result in acceptable interactions with essential targets. In this assessment, the fragments are assumed to be unimpeded by any flexib le connections between the fan housing and attached ducting.The following nuclear steam supply system components are considered to have a potential for missile 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; and4.pressurizer heaters.Gross failure of a control rod mechanism housing, sufficient to allow a control rod to be rapidly ejected from the core, is not consider ed 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 comp leted reactor coolant system.3.Control rod drive mechanism housings are made of Type 304 stainless steel. This material exhibits excellent notch t oughness at all temperatures that are encountered.However, it is postulated that the top plug on the control rod drive mechanism could become
 
loose 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 approximate ly 12 feet, the rod cluster control spider hits the underside of the upper support plate.4.Upon impact, the flexure arms in the coupl ing join the drive shaft and the control cluster fracture freeing the drive shaft from the control rod cluster.
MPS3 UFSAR3.5-5Rev. 305.The control cluster is completely stopped by the upper support plate. However, the drive shaft continues to accelerate upw ard and strikes the control rod drive mechanism missile shield provided.The control rod drive mechanism (CRDM) missiles are summarized in Table 3.5-3. The missile velocities have been calculated by balancing the forces due to the water jet. No spreading of the water jet has been assumed. CRDM missile impact velocity, kinetic energy, and penetration are considered in the design and layout of the missile shield.Valve stems in motor-operated (MOV) or air-operated (AOV) valves are not considered credible sources of missiles. All the isolation valves installed in the reactor coolant system have stems with a back seat. This effectively eliminates the possibility of ejecting valve stems even if the stem threads fail.Valves within the reactor coolant pressure boundary have been reviewed to identify potential missiles. This review identified no credible failures that could result in missile generation, except for valve missiles in the region where the pressuri zer extends above the operating floor. Valves in this region are the pressurizer safety valves, the motor-operated isolation valves in the relief line, the air-operated relief valves, and the air-operated spray valves. Although failure of these valves is not considered credible, failure of the valve bonnet body bolts is, neve rtheless, postulated and the integrity of the containment liner and safety related equipment from the resultant bonnet missile is assured by th e pressurizer cubicle wa lls and roof which act as a missile barrier.The missile characteristics of the valves in the region where the pressurizer extends above the operating deck are given in Table 3.5-4.The only credible sources of jet-propelled missiles from the re actor coolant piping and piping systems connected to the reactor coolant system are the temperature and pressure sensor assemblies. The resistance temperature sensor assemblies are of two types: "with well" and "without well." Two rupture locations have been postulated: around the weld (or thread) between the temperature element assembly and the boss of the "without well" element, and the weld (or thread) between the well and the boss for the "with well" element.The missile characteristics of the piping temperature sensor assemblies are given in Table 3.5-5. A 10-degree half-angle expansion water or steam jet has been assumed. The missile characteristics of the piping pressure elem ent assemblies are less severe than those of Table 3.5-5.Temperature and pressure sensors are installed on the reactor coolant pumps close to the radial bearing assembly. A hole is drilled in the gasket and sealed on the internal end of a steel plate. In evaluating missile potential, it is assumed that this plate breaks and the pipe plug on the external end of the hole becomes a missile.The missile characteristics of the reactor coolant pump temperature sensor, the instrumentation well of the pressurizer, and the pressurizer heaters are given in Table 3.5-6. A 10-degree expansion jet has been assumed.
MPS3 UFSAR3.5-6Rev. 30In addition, it is assumed that the welding between the instrumentation well and the pressurizer wall could fail causing the well and sensor assembly to become a jet-propelled missile.The initial flight direction of these missiles has been determined and only low kinetic energies are possible. Only fragile components would be damaged by this missile category.
Protection was implemented by keeping such components as electrical equipment and cables out of the primary flight path.Pressurizer heaters could loosen and become jet-propelled missiles. The integrity of safety related equipment from postulated heater vessels is as sured by the pressurizer cubicle floor and walls.All valves have been designed against bonnet-to-body connection failure and subsequent bonnet ejection by means of:1.Compliance with the ASME Code, Section III2.Control of load during the bonnet-to-body connection stud tightening processThe proper stud torquing procedures limit the stress of the studs to the allowable limits established in the ASME Code. This stress level is far below the material yield. The valves are hydrotested per the ASME Code, Section III. The bodies and bonnets are volumetrically and surface tested to verify soundness. Critical valves are also designated for inservice inspection to ASME Code, Section XI, requirements (Section 6.6).Non-ASME III valves were examined for potential missiles. Provi sion for protection against them was 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.
Missiles generated due to a seismic event are addressed in Section 3.7B.3.13.
3.5.1.2.2 Missile Protection ProvidedSafety related structures, systems, and components whose safety function might be impaired, are protected from postulated missiles by:1.Locating the systems or components in individual missile proof structures MPS3 UFSAR3.5-7Rev. 302.Physically separating redundant systems or components of the system3.Providing special localized pr otective shields or barriers.The ability of structures or barriers to withstand the effects of potential internally generated missiles is discussed in Section 3.5.3.
3.5.1.3 Turbine Missiles 3.5.1.3.1 Turbine Placement and OrientationFigure 3.5-1 shows the turbine pla cement and orientation for the three unit site. This figure also indicates the
+/-25-degree missile ejection zone with respect to the low pressure turbine wheels for each turbine unit "within reach" of plant structures. Figure 3.5-2 shows an elevation view of Millstone 3 struct ures within the
+/-25-degree missile ejection zone.
3.5.1.3.2 Missiles Identification and CharacteristicsA bounding value of 1.0E-2 per year for the probability of unacceptable damage in the event that a turbine missile is generated by wheel failure, is used.
3.5.1.3.3 Target DescriptionHeavy zones on Figure 3.5-1 identify target areas for low trajectory missiles. Table 3.5-7 provides applicable elevations over which those wall areas are considered as targets for each heavy line. Areas other than those shown by the heavy lines have been excluded as targets because they are either outside the ejection zone for low trajectory missiles or they meet the criteria that the probability of the missile strike damaging its target is zero, P 3 = 0.Target areas for high trajectory missiles are the roof areas of all Cate gory I structures. These targets are shown as shaded areas on Figure 3.5-1.
3.5.1.3.4 Probability AnalysisRegulatory Guide 1.115 Revision 1, "Protection Against Low-Trajectory Turbine Missiles" states that the NRC staff considers a hazar d due to low trajectory turbine ge nerated missiles of less than 1.0E-7 per year an acceptable risk rate for the loss of an essential system for a single event. The methodology is based on maintenance and inspection activities wh ich minimize the potential for missile generation. Usi ng the missile damage probability of 1.0E-2 (for "unfavorable orientation"), the probability of turbine failure, P 1, must be shown to be less than 1.0E-5 in order to meet overall R.G. 1.115 acceptable risk rate.The turbine system maintenance program is esta blished based on manufacturer's calculations of missile probabilities. The program ensures that the 1.0E-5 probability is maintained. Higher probabilities of turbine failure are acceptable for short duration's (for example, to support on-line MPS3 UFSAR3.5-8Rev. 30maintenance and surveillance activities) based on criteria reviewed and accepted by the NRC (NUREG-1048, Supplement No. 6, Appendix U).
3.5.1.3.5 Turbine Overspeed ProtectionThe turbine control system is an electrohydraulic control (EHC) system that includes both digital and analog circuitry, electronic servo hardware, and hydraulic valve actuators.
The EHC provides a normal oversp eed protection system and an emergency overspeed protection system to limit turbine overspeed. These two systems are essentially separate and independent. The normal overspeed protection system is part of the turbine load and speed control system and is designed to limit turbine overspeed without a turbine trip under all load conditions. The emergency overspeed protection system is part of the emergency trip system and is designed to trip the turbine if the turbine speed exceeds 110 percent of rated speed (Section 10.2).
3.5.1.3.6 Turbine Valve TestingThe main turbine generator control, main stop, intercept, and reheat stop valves are routinely tested (Section 10.2.3.6).Control and main stop valves are tested one at a time, and as each test is completed, the valve is returned to its original position before the next valve is tested. Intercept and reheat stop valves are interlocked so that a pair of thes e valves in one crossover pipe is tested together. For this test, one pair of pipe is tested and the valves returned to the open position before the next pair is tested.
3.5.1.4 Missiles Generated by Natural PhenomenaThe only credible missiles gene rated by natural phenomena are those genera ted by a design basis tornado. Those Category I structur es designed to withstand the effects of to rnado missiles and the systems and components thus protected are identified in Tables 3.5-1 and 3.5-2.A minimum of 2 feet thickness of reinforced concrete having a minimum strength of 3,000 psi (28 day compressive strength) was used for walls, roofs, and floors designated as missile protection. The minimum reinforcing steel each way in each face of any square foot of wall or slab providing missile protection is 1.85 square inches.
Postulated missiles generated by the design basis tornado (Section 3.3.2) are listed in Table 3.5-13, which includes all parameters necessary to determine missile penetration. These missiles are considered capable of striking in any orientation.Ventilation openings in the various facility buildings housing es sential shutdown equipment are protected by reinforced concrete labyrinths.There are no other design basi s missiles resulting from flood or any othe r natural phenomena described in Section 2.2.3.
MPS3 UFSAR3.5-9Rev. 30 3.5.1.5 Missiles Generated by Events Near the SiteNo missiles of any significance are expected to be generated by events near the site due to distances from nearby transportation routes. The possibility of missiles from a Providence & Worcester (P&W) railroad tank car explosion was considered; however, it was determined that such a tank car explosion would not generate significant missiles at the pl ant site. For a detailed description, see Section 2.2.3.
3.5.1.6 Aircraft HazardsA study of the probability of aircraft which use the nearby airport and airways colliding with the safety related structures of the Millstone site has been conducted. Due to the conservatism of the analysis, the values derived by the analytical model are believed to be substantially higher than the true probability. The study concludes that the aircraft accident probability would be less than 1.3 x 10-7 per year for a number of years since no increase in air traf fic is projected in the vicinity of the site.
3.5.2 STRUCTURES, SYSTEMS, AND COMP ONENTS TO BE PROTECTED FROM EXTERNALLY GENERATED MISSILES Missiles to be considered in this section are identified in Section 3.5.1.As discussed in Section 3.5.1, all plant systems and components must be protected whose failure can lead to unacceptable offsite radiological conse quences or which are requi red to shut down the reactor and maintain it in a safe condition, assuming an additional single failure. Safety related structures, systems, and components that are required for a safe shutdown of the reactor are identified in Tables 3.5-1 and 3.5-
: 2. All components contai ning radioactive fluids are protected or fall within the results of the design basis offsite radiological analyses which are presented in Sections 15.7.1, 15.7.2, and 15.7.3.
 
====3.5.3 BARRIER====
DESIGN PROCEDURESMissile barriers are designed to withstand the effects of missiles described in Section 3.5.1 without compromising plant safety.Most internally generated missiles have low to moderate energy. Consequently most barriers erected specifically for protection against internal missiles are rather light and usually protect only a limited area. Small steel barriers, rather than concrete structures are generally most suitable. There are a few exceptions, such as the control rod drive missiles, where sufficient energy exists to make a concrete barrier more practical.
3.5.3.1 Concrete Barriers Missile barrier design requirements include:
MPS3 UFSAR3.5-10Rev. 301.Prevent perforation of the missile into safety related areas with one or more barriers2.Scabbing particles are not generated or are limited to energy levels which still permit safe shutdown of the plant3.Structural response to missile impact perm its 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 safe ty function of a Category I system.For concrete barriers, the requirement to stop the missile is fulfilled by showing that perforation of the barrier, or the final barrier in a multiple barrier design, does not occur. This is established based either on test data (Nusbaum et al., 1976; Stephenson 1976; and Rotz 1975), or an empirical evaluation of these test data.Determination of whether scabbing occurs is also based on procedures set forth in Stone & Webster Engineering Corporation Topical Report, SWEC-7703, Appendix B (SWEC 1977).The overall structural response to missile impact is based on dynamic time history calculations for elastic-plastic behavior of the missile and the barrier. The general method for calculating overall structural response is described in "Introduction to Structural Dynamics" (Biggs 1964). The ultimate load capacity of concrete barriers is determined by yield line theory. The barriers are designed so that the calculated ductility ratio of the barriers for any load combination is less than the maximum allowable ductility ratio.Ductility ratio is defined as the ratio of maximum acceptable displacement (Xm) to the displacement at the effective yiel d point (Xy) of the structure.For analytical procedures for analyzing concrete barriers for overall structural response to tornado missiles refer to Stone & Webster Engineering Corporation Topical Report SWEC 7703, Appendix C (SWEC 1977).If the concrete barrier is required to carry lo ads during and after missile impact, the maximum allowable ductility ratio is limited to a factor of 10. In particular, fo r beam-column members, where the compressive load is equal to or less than one-third of that which produces balanced strain conditions for the cross section, the allowa ble ductility ratio is 10.Refer to Table 3.5-14 for examples of maximu m barrier elastic defl ection, maximum barrier deflection, and ductility ratio as a function of missile and barrier span for 24 inch concrete barriers.For an example of maximum barrier elastic deflection, maximum barrier deflecti on, and ductility ratio for a beam-column, refer to Table 3.5-15.If a concrete barrier is not required to carry other loads during and af ter impact, the maximum allowable rebar elongation is limited to 5 percent.
MPS3 UFSAR3.5-11Rev. 30 3.5.3.2 Steel BarriersSteel barriers are designed with consideration given to local perforation, overall response, and support loads.The thickness of the barrier plate required to prevent perforation is determined from the Stanford equation (Gwaltney 1968). The BRL fo rmula for steel perforation is used if the Stanford equation is not applicable. However, impacts near a support are most limiting with respect to perforation and where these are postulated, the Stanford equation is used with the artificial, but conservative, assumption that the support spacing equals the missile diameter.The preferred design is a single plate supported at the corners and, if necessary, reinforced by bolting around the periphery to channels or other rolled shap es. The overall res ponse is evaluated considering a central impact and equating the missile energy to the energy absorbing capacity of the barrier plate. Plastic limit theory and an allowable of 50% ultimate uniform strain are used to determine this capacity.The load delivered to the supports is dynamic and equal to the plastic limit load of the plate. Since the impact may occur near one support, each support is sized accordingly. When the resulting load is too high for founding structure, soft supports are used. In this case the supports, whether elastic or plastic, are designed to accept the available energy:
where E o = Initial missile energy,m = Missile mass, andM = Barrier mass participating in missile momentum transfer 3.5.3.3 Design EvaluationOnly one missile is postulated at a time and is assumed to strike the barrier e nd-on. The missiles have two effects on structures, walls, or any other barrier: local effects and overall response. The local effects include penetrat ion, perforation, and spalling or scabbing. The overall response includes the flexural and shear effects.Reinforced concrete external roofs and walls of Seismic Category I structures form barriers against tornado-generated missiles. Buried underground, safety related duct runs and piping are also protected.Local Effect: The estimate of missile penetration in concrete barrier is based on the modified Petry formula (Amirikian 1950). Sufficient thickness of concrete is provided to prethickness EE o m Mm+---------------
=
MPS3 UFSAR3.5-12Rev. 30determined for an infinitely thick slab is provided. Determination of penetration by missiles into steel plates is in accordance with the Stanford Research Institute formula (Gwaltney 1968).Overall Barrier Response: The overall response of structural barriers to missile impact is determined by time history analysis of the slab missile system. The ductility ratio so determined must be less than the allowable ductility ratio. This ductility ratio is a function of th e controlling nature of the structural behavior. For beams, walls, and slabs where flexural controls design, the permissible ductility ratios are given in Section 3.5.3.1 and are based on Stone & Webster Engineering Corporation Topical Report, SWEC 7703 (SWEC 1977). Flexural strength is determined from an ultimate strength theory with the limitations on ductility.An example of a missile shield that has been de signed using these procedures is the control rod drive mechanism missile shield.The shield is provided over the control rod drive mechanisms to block any missiles which might be associated with a fracture of the pressure housing of any mechanism. This shield is constructed of reinforced concrete with a steel facing plate, and is located above, and as near as possible, to the housing. This limits the velocity of the ejected missile and prevents missiles from bypassing it. The shield is designed using procedures given in ORNL-NSIC-5 (Greenstreet et al., 1965).
3.5.3.4 Secondary MissilesConcrete impacts produce secondary missiles from the front face of the barrier. These missiles have considerably less energy than the primary missile. Concrete secondary missiles were only considered for a postulated primary missile which has more than 4 ft-kips of kinetic energy. Our review showed that scabbing does not damage safety related equipments.Secondary missiles from impacts are limited to components that may be torn loose by the primary missile. The energy dissipated in creating secondary missiles cause the remaining energy in the primary missile to be negligible. Only concrete walls were found to exhibit this phenomena of loosing parts or scabbings. Nevertheless, all secondary missiles were found passive.
3.
 
==5.4 REFERENCES==
FOR SECTION 3.53.5-1Amirikian, A. 1950. Design of a Protective Structure. NAVDOCKS, P-51, Bureau of Yards and Docks, Dept. of the Navy.3.5-2Biggs, J.M. 1964. Introduction to Structural Dynamics. McGraw-Hill, New York, New York.3.5-3Bush, S.H., Probability of Damage to Nuclear Components Due to Turbine Failures, Nuclear Safety. Vol. 14, No. 3, May - June, 1973.3.5-4General Electric Co., March 1973. Hypothetical Turbine Missiles - Probability of Occurrence, Memo Report. Turbine Department.
MPS3 UFSAR3.5-13Rev. 303.5-5Greenstreet, B.L.; Salmon, M.A.; and Weil, N.A. 1965. US Reactor Containment Technology. Vol. I, Section 6.6, ORNL-NISC-5, Ed. W.B. Cottrell and A.W. Savolaineu, Oak Ridge National Laboratory and Bechtel Corp., US AEC Technical Information Center, TD-4500.3.5-6Gwaltney, R.C. 1968. Missile Generation and Protection in Light Water-Cooled Power Reactor Plants. ORNL-NSIC-22.3.5-7Nusbaum, M.S., Welch, R.E., and Fox x, C.E. 1976. One Quarter Size Reinforced Concrete Missile Barrier Tests. ITT Research Institute Report.3.5-8Rotz, 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.3.5-9Sihweil, I. 1976. Inte rim Position on Ductility Ratios of Reinforced Concrete Structural Elements. Memorandum, USNRC, (SEB:306).3.5-10Stephenson, A.E. 1946. Full-Scale Tornado - Missile Impact Tests. Electric Power Research Institute, In terim Report EPRI NP-148.3.5-11Stevenson, 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.3.5-12Stone & Webster Engineering Corporati on (SWEC) 1977. Missile Barrier Interaction, Topical Report.
MPS3 UFSAR3.5-14Rev. 30TABLE 3.5-1 SAFETY RELATED STRUCTUR ES, SYSTEMS, AND COMPONENTS OUTSIDE CONTAINMENT REQUIRED FOR SAFE REACTOR SHUTDOWN*StructuresLocation**FSAR SectionFSAR FigureContainment penetrationsAB, EF, MS, HR3.8.1 3.8-18, 3.8-19 Containment hatches (personnel and equipment) outside containmentAB, HR3.8.1 3.8-21, 3.8-22 Containment reinforced concrete external superstructureCS3.8-1 thru 3.8-14 and
 
3.8-19 3.8.1 Cable tunnel from containment structure to control buildingCB, AB, SB3.8.4 3.8-62 thru 3.8-67 and
 
8.3-1 Building reinforced concrete walls surrounding safety related systems and
 
componentsAB, EF, FB, SB3.8.4 3.8-62 thru 3.8-67 Main steam and feedwater valve AreasMS3.8.4 3.8-62 thru 3.8-67 Engineered safety features areasEF3.8.4 3.8-62 thru 3.8-67Control buildingCB3.8.4 8.3-1 Service water pumphouseSP3.8.4 3.8-69 and 3.4-1 Diesel generator buildingDG3.8.4 3.8-68, 3.8-69
 
Diesel generator fuel oil pump HouseOY3.8.4, 3.8.5 3.8-68, 3.8-69 Systems Chemical and Volume Control System 9.3.4 9.3-7 and 9.3-8 Charging pumpsABSeal water injection filterAB Boric acid tanksABBoric acid transfer pumpsAB Piping, valves, and portions of the system required for charging, seal water injection and boration AB MPS3 UFSAR3.5-15Rev. 30 Instrumentation, cable, and controls required to perform a safety function in the above portions of system AB, CR, SB Residual Heat Removal System 5.4 5.4-4 Residual heat removal pumpsEFResidual heat removal exchangerEFTube side Shell side Piping and valves required to cool and maintain the RCS in a cold shutdown
 
conditionEF, AB Instrumentation, cable, and controls required to perform a safety function in the above
 
portions of systemEF, AB, SB, CREmergency Core Cooling System 6.3 6.3-1 thru 6.3-4High-head safety injection pumps (charging pumps)AB Low-head safety injection pumps (residual heat removal pumps)
EF Piping and valves required for injection to the RCSEF, AB Instrumentation, cable and controls required to perform a safety function in the above portions of systemEF, AB, SB, CRQuench Spray System 6.2.2 6.2-36 Refueling water storage tank, quench spray pumps, piping and valves required to
 
perform intended safety functionsEF, OYInstrumentation, cables and controls required to perform a safety function in the above portions of systemEF, OY, CREmergency Generator Fuel Oil Storage and Transfer Systems 9.5.4 9.5-2 TABLE 3.5-1 SAFETY RELATED STRUCTUR ES, SYSTEMS, AND COMPONENTS OUTSIDE CONTAINMENT REQUIRED FOR SAFE REACTOR SHUTDOWN*StructuresLocation**FSAR SectionFSAR Figure MPS3 UFSAR3.5-16Rev. 30Emergency generator fuel oil day TanksDGEmergency generator fuel oil transfer pumpsDG fuel oil transfer Pump houseEmergency generator fuel oil storage tanksOY (Underground)
Piping and valves required to perform a safety function OY/DG Instrumentation, cable, and controls required to perform a safety function in the above
 
portions of system DB / OY IR /
EREmergency Generator Cooling Water System 9.5.5 9.5-3 Emergency generator cooling water exchanger DG Cooling water electric immersion heater pressure housing DGEmergency generator cooling water expansion tank DGMotor driven circulating water pumpDG Piping and valves required to perform a safety function DG Instrumentation and controls required to perform a safety function in the above portions of system DGEmergency Generator Starting System 9.5.6 9.5-3 Emergency generator air starting storage tanks DG Piping and valves required to perform a safety function DG Instrumentation, cable, and controls required to perform a safety function in the above portions of system DGEmergency Generator Lubrication System 9.5.7 9.5-3 TABLE 3.5-1 SAFETY RELATED STRUCTUR ES, SYSTEMS, AND COMPONENTS OUTSIDE CONTAINMENT REQUIRED FOR SAFE REACTOR SHUTDOWN*StructuresLocation**FSAR SectionFSAR Figure MPS3 UFSAR3.5-17Rev. 30Emergency generator lubrication oil coolerDGEmergency generator lubrication oil filterDG Piping and valves required to perform a safety function DG Instrumentation and controls required to perform a safety function in the above portions of system DGEmergency Generato r Air Intake and Exhaust System 9.5.8 9.5-4 Air filterDGAir intake silencerDGExhaust silencerDGPiping required to perform a safety functionDG Service Water System 9.2.1 9.2-1 Service water pumpsSPPiping and valves suppl ying cooling water to safety related equipmentAB, EF, OY (buried)Instrumentation, cable, and controls required to perform a safety function in the above
 
portions of systemEF, DG, CB Reactor Plant Component Cooling Water System 9.2.2.1 9.2-2 Reactor plant component cooling pumpsABReactor plant component cooling surge tankAB Reactor plant components cooling heat exchangers ABPiping and valves suppl ying cooling water to safety related equipment required for shutdown AB, SB Instrumentation, cable and controls required to perform a safety function in the above portions of systemAB, CBTABLE 3.5-1 SAFETY RELATED STRUCTUR ES, SYSTEMS, AND COMPONENTS OUTSIDE CONTAINMENT REQUIRED FOR SAFE REACTOR SHUTDOWN*StructuresLocation**FSAR SectionFSAR Figure MPS3 UFSAR3.5-18Rev. 30Steam Generator Blowdown System 10.4.8 10.3-1 Piping and valves from containment up to and including the blowdown isolation valves MS Instrumentation, cabling, and controls required to trip close the isolation valves MSMain Steam System 10.3 10.3-1,2,3,4 Main steam piping and valves from containment up to and including main steam isolation stop valves and all piping in the valve house MS Main steam piping and valves from main steam lines to steam generator auxiliary feedwater pump turbine MS Instrumentation, cable, and controls required to trip close the MS stop valves and bypass valvesMS / CR / IRSteam generator atmospheric relief valves (can be operated manually)
MS Auxiliary Feedwater System 10.4.9 10.4-9 Demineralized water storage tankOY Auxiliary feedwater pumps (turbine- and motor-driven)
EF Piping and valves supplying auxiliary feedwater from ST to containment penetrationOY, EF Instrumentation, cable and controls required to perform a safety function in the above portions of system Feedwater System 10.4.7 10.4-8 Feedwater piping and valves from containment up to and including the feedwater isolation valves MSTABLE 3.5-1 SAFETY RELATED STRUCTUR ES, SYSTEMS, AND COMPONENTS OUTSIDE CONTAINMENT REQUIRED FOR SAFE REACTOR SHUTDOWN*StructuresLocation**FSAR SectionFSAR Figure MPS3 UFSAR3.5-19Rev. 30 Instrumentation, cable, and controls required to trip close the feedwater isolation valves and to trip the main feedwater pumps Air Conditioning, Heating, Cooling and Ventilation Systems
 
===9.4 Emergency===
generator enclosure ventilationEGE9.4.6 9.4-3 Engineered safety features area unit cooler and ductworkEF / AB9.4.3 9.4-2 Control building heating and ventilation, including control building chilled water systemCB9.4.1 9.4-1 Service water pumphouse ventilationSP9.4.8 9.4-3 Instrumentation and control required to perform a safety related function for the above portions of system EF / EGE  / CB / AB Lighting System 9.5.3 NoneEmergency lightingAB/CRElectrical Systems 8.3 8.3-1 thru 8.3-3Emergency generatorsEGEBatteries, chargers, and battery switchboardsERVital bus panels and invertersER Essential unit substationsEREssential metalclad switchgearEREssential motor control centersER Control panelboards:CR Main control board and panelsEmergency generator panelERAuxiliary shutdown panelERControl room air-conditioning control panelCRTABLE 3.5-1 SAFETY RELATED STRUCTUR ES, SYSTEMS, AND COMPONENTS OUTSIDE CONTAINMENT REQUIRED FOR SAFE REACTOR SHUTDOWN*StructuresLocation**FSAR SectionFSAR Figure MPS3 UFSAR3.5-20Rev. 30NOTES: *The applicable seismic categor y and quality group classificati on for the equipment listed in this table is delineated in Table 3.2-1
.** Location Symbols Cabling and raceway supports for safety related equipment required for safe shutdownAll locations outside containmentMotors for safety related equipment required for safe shutdown Same as components Instrumentation and cables required for the safety related portion of the above electrical systemsAll locations outside containmentReactor Trip Systems All portions of the reactor trip and safeguards actuation channels including sensors, circuitry, and proces sing equipment required to control and safety shut down the reactor to
 
the cold shutdown condi tion (the protection circuits used to trip the reactor on undervoltage, underfrequency, and turbine
 
trip are excluded)CR / IR7.2, 7.4 7.2-1 thru 7.2-4 Controls for defeating automatic safety injection actuation during a cooldown and
 
depressurizationCR / IR7.3, 7.4 None Indication of the following plant parameters should be available to the operator:7.4, 7.5, 7.7 7.7-1, 7.7-4, 7.7-5, 7.7-6 Indication of plant parameters required for safe shutdown (see Table 7.5-1)
CR/IR Containment Isolation System Containment isolation valves and associated piping for all systems penetrating containment structureAB, EF, MS6.2.4 6.2-47 Instrumentation, cable and controls required to perform a safety function in the above portions of system ABTABLE 3.5-1 SAFETY RELATED STRUCTUR ES, SYSTEMS, AND COMPONENTS OUTSIDE CONTAINMENT REQUIRED FOR SAFE REACTOR SHUTDOWN*StructuresLocation**FSAR SectionFSAR Figure MPS3 UFSAR3.5-21Rev. 30 AB - Auxiliary building CB - Control building CR - Control room (in control building)
CS - Containment structureEGE - Emergency generator enclosure EF - Engineered safety features buildingER - Emergency switchgear room (in control building)
IR - Instrument rack room (in control building)
MS - Main steam valve enclosures SP - Service water pumphouse OY - Outside, yardAT - Auxiliary feedwater storage tank HR - Hydrogen recombiner building SB - Service building FB - Fuel building MPS3 UFSAR3.5-22Rev. 30TABLE 3.5-2 SAFETY RELATED STRUCTUR ES, SYSTEMS, AND COMPONENTS INSIDE CONTAINMENT REQUIRED FOR SAFE REACTOR SHUTDOWN*Structures Location **FSAR Section Containment Structure3.8.1, 3.8.3 Reinforced concrete interior substructureCSReinforced concrete interior shields and wallsCS Containment structure linerCSPiping and duct penetrationsCS Electrical penetration/assembliesCSPersonnel access lockCSEquipment hatchCSSystems Reactor Coolant System 5.1, 5.2, 5.3, 5.4 Reactor vesselCSControl rod drive mechanismsCS Steam generatorsCSPressurizerCSPiping, fitting and valves within RCPBCS Reactor coolant thermowellCSReactor coolant thermowell bossCSSafety valvesCS Relief valvesCSControl rod drive mechanism head adapter plugsCSIsolation valves for reactor coolant system branch line CSReactor coolant pumpCSReactor coolant pump casingCS Main flangeThermal barrierSystems Location **FSAR Section No. 1 seal housing MPS3 UFSAR3.5-23Rev. 30 No. 2 seal housingBoltingImpellerDiffuserReactor coolant pump motorCSRotor FlywheelShaft Shaft coupling BearingsUpper oil coolerCS- Tube side - ccw- Shell side - oilLower oil coolerCS- Tube side - ccw- Shell side - oil Instrumentation cable and controls required to perform a safety function in the above portion of system CSChemical and Volume Control System 9.3.4 Piping, valves, and portions of the system required for charging, seal water injection, and boration CS Instrumentation, cable, a nd controls required to perform a safety function in the above portion of system CS Residual Heat Removal System 5.4 Piping and valves required to cool and maintain the RCS in a cold shutdown condition CS Instrumentation, cable, a nd controls required to perform a safety related function in the above portions of system CS 6.3TABLE 3.5-2 SAFETY RELATED STRUCTUR ES, SYSTEMS, AND COMPONENTS INSIDE CONTAINMENT REQUIRED FOR SAFE REACTOR SHUTDOWN*
MPS3 UFSAR3.5-24Rev. 30ESF pumpsCS Piping and valves required for injection to the RCSCSSystems Location**FSAR Section Instrumentation, cable and controls required to perform a safety related function in the above
 
portions of system CS Containment Isolation System 6.2.4 Containment isolation valves and associated piping for all systems penetrating containment structure CS Instrumentation, cable and controls required to perform a safety related function in the above
 
portions of system CSReactor Plant Component Cooling Water System 9.2.2.1 Piping and valves supplying cooling water to safety related equipmen t required for shutdown CS Instrumentation, cable and controls required to perform a safety related function in the above
 
portions of system CS Auxiliary Feedwater System 10.4.9 Piping and valves supplying auxiliary feedwater to the steam generators CS Feedwater System 10.4.7 Piping and valves supplying auxiliary feedwater to the steam generator CSAir-Conditioning, Heating, Cooling, and Ventilation Systems 9.4.7 Containment recirculat ion unit coolers and associated equipment excluding chillers CS Instrumentation, cable, a nd controls required to perform a safety related function for the above air-conditioning, heating, c ooling and ventilation systems CSElectrical Systems8.3 TABLE 3.5-2 SAFETY RELATED STRUCTUR ES, SYSTEMS, AND COMPONENTS INSIDE CONTAINMENT REQUIRED FOR SAFE REACTOR SHUTDOWN*
MPS3 UFSAR3.5-25Rev. 30NOTES: *The applicable seismic category and quality group classification for the equipment is listed in Table 3.2-1
.**Location SymbolsCS - Containment StructureCabling and raceways for safety related equipment required for safe shutdown CS Motors for safety related equipment required for safe shutdown CSSystems Location **FSAR Section Reactor Trip SystemSee Table 3.5-1 Those portions of the reactor trip system listed in Table 3.5-1 that are located inside containment CSTABLE 3.5-2 SAFETY RELATED STRUCTUR ES, SYSTEMS, AND COMPONENTS INSIDE CONTAINMENT REQUIRED FOR SAFE REACTOR SHUTDOWN*
MPS3 UFSAR3.5-26Rev. 30TABLE 3.5-3
 
==SUMMARY==
OF CONTROL ROD DRIVE MECHANISM MISSILE ANALYSISTypical Examples of Postulated MissilesMissile Weight (lb)Impact Velocity (ft/sec)Kinetic Energy (ft-lb)Penetration (in)Assumptions1.Mechanism housing cap11901,3800.05Plug becomes loose and is accelerated by the water jet2.Mechanism top cap and drive rod assembly
 
impacting on the same missile shield spot13315046,7570.08Drive shaft further pushes the plug into the shield3.Drive shaft latched to mechanism1,50012.11,4900.057-MPS3 UFSAR3.5-27Rev. 30NOTE:* Typical values based upon manufacturer design dataTABLE 3.5-4 VALVE-MISSILE CHARACTERISTICS*Missile DescriptionWeight (lb)Flow Discharge Area (in 2)Thrust Area (in 2)Weight to Impact Area (in 2)Impact Area Ratio (psi)Velocity (fps)Safety valve bonnet3502.86802414.6110
 
(3 inches x 6 inches) or 6 inches x 6 inches3-inch motor-operated isolation valve bonnet (plus motor and stem)4005.51132814.1135 3-inch solenoid-operated relief valve bonnet (plus stem)751.820203.751154-inch air-operated spray valve2009.350504190 MPS3 UFSAR3.5-28Rev. 30TABLE 3.5-5 PIPING TEMPERATURE ELEMENT ASSEMBLY - MISSILE CHARACTERISTICS1.For a tear around the weld between the boss and the pipe:2.For a tear at the junction between the temp erature element assembly and the boss for the "without well" element and at th e junction between the boss and the well for the "with well" element.Characteristics Without Well With Well Flow Discharge Area (in 2)0.110.60Thrust Area (in 2)7.19.6Missile Weight (lb)11.015.2 Area of Impact (in 2)3.143.14Missile WeightImpact Area (psi)3.54.84Velocity (ft/sec)20.0120.0 Characteristics Without Well With Well Discharge Area (in 2)0.110.60Thrust Area (in 2)3.143.14Missile Weight (lb)4.06.1
 
Area of Impact (in 2)3.143.14Missile WeightImpact Area (psi)1.271.94Velocity (ft/sec)75.0120.0 MPS3 UFSAR3.5-29Rev. 30TABLE 3.5-6 CHARACTERISTICS OF OTHER MISSILES POSTULATED WITHIN REACTOR CONTAINMENT Characteristics Reactor Coolant Pump Temperature SensorInstrument Well of PressurizerPressurizer HeatersWeight (lb)0.255.515Discharge Area (in 2)0.500.4420.80Thrust Area (in 2)0.501.352.4 Impact Area (in 2)0.501.352.4Missile WeightImpact Area (psi)0.54.16.25Velocity (ft/sec)26010055 MPS3 UFSARMPS3 UFSAR3.5-30Rev. 30TABLE 3.5-7 TURBINE-TARGET DISTANCES AND IMPACT AREASHigh Trajectory TargetsTarget Area (sq ft)  R * (ft) y ** (ft) 1.Control Building11,73021615.5 2.Emergency Generator Enclosure3,999303-13.03.Fuel Oil Storage Tanks1,736301-44.54.Auxiliary Building15,97526524.5 5.Main Steam Valve Building1,85617114.96.Fuel Building9,70134637.07.Engineered Safeguards Building8,24133324.0 8.Demineralizer Water Storage Tank1,13342311.09.Refueling Water Storage Tank2,64140687.010.Hydrogen Recombiner Buildings720307-17.7 11.Containment Structure8,721256108.312.Service Water Pumps2,750570-45.0 Low Trajectory TargetsTarget Area (sq ft)R * (ft)y ** min (ft)y ** max (ft) 1.Auxiliary Building Wall ***023414.924.52.Main Steam Valve Building Wall1,560157-12.014.93.Refueling Water Storage Tank Wall816383-17.718.0 4.Hydrogen Recombiner Building Wall819295-45.0-17.75.Containment Wall - Section 12,65718814.986.96.Containment Wall - Section 22,372188-45.086.9 MPS3 UFSAR3.5-31Rev. 30NOTES:
* R = average distance from turbine
** y = height above spin axis of turbine
*** Auxiliary building wall can be hit by missiles from only one wheel MPS3 UFSAR3.5-32Rev. 30TABLE 3.5-8 DELETED BY FSARCR 02-MP3-13 MPS3 UFSAR3.5-33Rev. 30TABLE 3.5-9 DELETED BY FSARCR 02-MP3-13 MPS3 UFSAR3.5-34Rev. 30TABLE 3.5-10 DELETED BY FSARCR 02-MP3-13 MPS3 UFSAR3.5-35Rev. 30TABLE 3.5-11 DELETED BY FSARCR 02-MP3-13 MPS3 UFSAR3.5-36Rev. 30TABLE 3.5-12 DELETED BY FSARCR 02-MP3-13 MPS3 UFSAR3.5-37Rev. 30NOTES:* Missiles 1 through 6 are considered at all elevations. Maximum trajectory height of Missile 7 is 25 feet above grade.** Vertical velocities of 70 pe rcent of horizontal velocities are used, except for Missile 3 which has the same speed in all directions.TABLE 3.5-13 POSTULATED TORNADO-GENERATED EXTERNAL MISSILESMissile Description *Weight (lbs)Horizontal Velocity ** (mph)1. 4 x 12 in plank, 12 ft long at 50 lb/ft 3200210 2. Utility Pole 13.5 inch diameter, 35 feet long at 43 lb/ft 31,4961203. 1 inch solid steel rod, 3 feet long81304. 3 inch schedule 40 pipe, 10 feet long781405. 6 inch schedule 40 pipe 15 feet long2851206. 12 inch schedule 40 pipe, 15 feet long743110
: 7. Automobile frontal area 20 ft 24,00050 MPS3 UFSARMPS3 UFSAR3.5-38Rev. 30Barrier Information
: Reinforcement: #11 at 10 inches GR 40 each way, each face Concrete: 3,000 psi Barrier Depth: 24 inchesTABLE 3.5-14 BARRIER DEFLECTION A ND DUCTILITY RATIOS FOR TORNADO-BORNE MISSILES PLUS 360 MPH TORNADO WIND Missile DescriptionMass (slugs)Velocity (fps)Momentum (lb sec)Barrier Span (ft)Maximum Elastic Barrier Deflection (Xy) (ft)Barrier Deflection (Xm) (ft)Barrier Ductility 12 inch schedule 40 pipe 15 feet long 743 lb23.07161.333721.965.0.00380.00691.7910.0.01540.02661.7320.0.06150.03980.6530.0.13830.04890.354,000 lb auto124.273.339108.5.0.00380.02957.6810.0.01540.02021.32 20.0.06150.04140.6730.0.13830.08430.61 13.5 inch diameter utility pole 35 feet long 1,496 lb46.4617681775.0.00380.00310.8010.0.01540.01220.7920.0.06150.04990.8130.0.13830.08710.63 MPS3 UFSARMPS3 UFSAR3.5-39Rev. 30TABLE 3.5-15 BARRIER DEFLECTION AND DUCTILITY RATIO FOR A BEAM-COLUMN PLUS 360 MPH TORNADO WIND Barrier DescriptionMissile Description Mass (slug)Velocity (fps)Momentum (lb sec)Maxim Barrier Elastic Deflection (ft)Maximum Barrier Deflection (ft)Barrier Ductility Ratio  Barrier Compression Load (#)Maximum Allowable Ductility Ratio Refer to Figure 3.5-613.5 inch diameter utility pole 35 ft
 
long46.4617681770.030.0310.Barrier dead load only 10.
MPS3 UFSAR3.6-1Rev. 30
 
===3.6 PROTECTION===
AGAINST DYNAMIC E FFECTS ASSOCIATED WITH THE POSTULATED RUPTURES OF PIPINGThis section addresses the subj ect of a postulated pipe break/
crack inside and outside the containment. It also presents the results of analyses initiate d in response to NRC Regulatory Guide 1.46 for inside containment and the Giam busso letter of December 18, 1972, for outside containment. The methods of evaluation, how ever , reflect the a pproach and methodology contained in the Branch Technical Positio ns ASB 3-1 and MEB 3-1 as qualified in Sections 3.6.1and 3.6.2.3.6.1 POSTULATED PIPING FAILURES IN FL UID SYSTEMS INSIDE AND OUTSIDE OF CONTAINMENT This section describes the design criteria and bases for protecti ng essential equipment from the ef fects 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 pipi ng failures are preserved.2.The plant can be safely shut down to a cold shut down condi tion without of fsite power.3.Offsite doses in excess of a pplicable guidelines do not occur
.As used in this section, essent ial equipment is defined as the structures, systems, portions of systems, and components required to mitigate the ef fects of the pos tulated 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 CriteriaWithin the plant there are high and moderate en er gy piping systems which are postulated to fail. Failures of these systems are evaluated to determine their deleterious effects on essential equipment and those essential syst ems, components, and structures that are susceptible to the effects of these failures are provided protection. The criteria for protection agains t pipe rupture inside the containment are contained in NRC Regulatory Guide 1.46, as described in Section 1.8 , the criteria for protection agains t pipe rupture outside containmen t conform to the NRC Technical Branch Position ASB 3-1.
Section 3.6.1.1.4 discusses design feat ures recommen ded for protection.
MPS3 UFSAR3.6-2Rev. 30 3.6.1.1.2 Design Basis Pipe Break/Crack Criteria The design basis pipe breaks are postulated to occur in high ener gy 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 oper ates during normal plant operatin g 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 gr aphically, including the diff erences between inside and outside the containment.
Maximum operating temperature and pressure are the maximum temp erature and pressure in the fluid systems during occurrence s which are expected during nor mal plant conditions. Normal plant conditions are defined as startup, operation at power, hot standby, or reac tor cooldown to cold shutdown conditions.
Through-wall leakage cracks are postulated to occu r in moderate ener gy sy stems (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 cond itions but which has bot h 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 StructuresTable 3.6-5 lists all essential equipment required to shut down th e 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 eq uipment within the plant. Thus, a de termination can be made of which high and moderate ener gy 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 mitiga te 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 sepa ration criteria, so that the ef fects of postulated pipe breaks at any lo cation are isolated or physically remote from essential structures, sy stems, and components.
Redundant essential equipment is locat ed in separate cubicles so that a pipe failure in one cubicle cannot af fect the backup equipment. Inside each redundant cubicle, high energy piping is located with as much separation as possibl e from the essential equipment so that the effects of pipe whip or jet impingement are minimized.
MPS3 UFSAR3.6-3Rev. 30Whenever possible, safety related electrical a nd control system equipmen t is located in areas which do not have piping (e.g., the emergency switchgear ar ea). 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 instru mentation lines. For all safety components where physical separation is impossible, one or more of the design features de scribed below are provided to ensure that essential equipment remains operable.
3.6.1.1.4.2 Enclosures Where remote physical se paration is not practical , protective enclosures are used. For example, each char ging pump is located in a separate cubicle with c oncrete walls between redundant pumps. Redundant electrical and c ontrol systems in the area of high energy piping are located as far as possible from pos tulated pipe break locat ions so that potential damage is minimized.
Section 3.8.4 discusses the concrete de sign of these enclosures.
3.6.1.1.4.3 Restraints Where neither of the above design approaches is feasible, protecti on is provided by pipe restraints to limit pipe movement fo llowing a rupture to restrict the area affected by the failure. Placement of restraints is based on postula ted pipe break locations. The design of the restraints for these locations limits pipe whip to acceptable levels and reduces je t 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 separa tion criteria by locating high and moderate ener gy systems physically remote from essential equipm ent. Where plant arrangement and/or piping layout does not pe rmit complete physical separa tion, protective enclosures are provided. When these methods are not practical, restraints are in stalled 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 build ing where no essential equipment is located, thus satisfying the physical separation criterion.
Pipe breaks are only postulated for the main steam line in this ar ea due to its proximity to the control building.Table 3.6-2 summarizes all high energy syst ems that are located in proxi mity to essential systems, components, and structures. The majority of essential systems, and components are found in the Engineered Safety Features, Auxi liary Building, and the Containm ent Structure. Pipe breaks and through wall leakage cracks are po stulated in systems that are proximate to these essential systems, and components.Table 3.6-3 contains all moderate energy systems outside the containm ent separated from essential systems, components, and structures by plant a rrangement and/or piping layout.
Therefore, these systems are not evaluated for pipe rupture.
MPS3 UFSAR3.6-4Rev. 30Table 3.6-4 includes all moderate energy systems loca ted outside containmen t in proximity to essential systems, comp onents, and structures. Through-wall cracks are postulated in these systems or portions of systems to evaluate the ef fects of fluid spray and flooding.
Jet impingement effects from cracks are not evaluated.
Each system listed in Table 3.6-2 is referenced by pi pe break location draw ing (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 ef fe cts 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 co mponent following a postula ted piping failure may be dependent on one or more of the following assumptions in a ddition to consideration of the ef fects of the failure.
Single Failure Criterion A single active component failure is assumed to occur in essent ial 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 postulate d 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 criter ion and its applications.Loss of Offsite Power Offsite power is assumed unavailable if a trip of the turbine generator syst em 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 conseque nces of a postulated event may be taken only for those systems and components designed to Seis mic Category I requirements, ex cept for the High Ener gy Line Break (HELB) isolation systems for the auxiliary steam and th e hot water heating systems.
Although the electrical sensing a nd isolation devices are Category I and located in a Category I building, the valves are not Category I as they are installe d in a non-Category I piping system located in a non-Category I structure. The locati on of the isolation valv es 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 even
: t. Judging the availability of sy stems includes consideration of MPS3 UFSAR3.6-5Rev. 30 the postulated failure and its direct consequences (e.g., unit trip and loss of of fsite power) and the assumed single active component fa ilure plus its direct conseque nces. 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. Re gulatory Guide 1.62 provides guidance in evaluating the feasibility of operator action.
3.6.1.3.2 Failure Mode and EffectsAn 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 comprise s the pipe which is postulated to fail and the resulting effects of the failure. The target comprises structures, systems, and components considered essent ial for shutting down the plant safely, maintaining the safe shutdown, and mitigating the effects of the postulated pipe failure.
Interactions between sour ces and tar gets 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 determinin g break locations in piping systems.
Section 3.6.2.2 describes the pipe whip an alysis for defining forcing fu nctions and pipe responses. The effects of a particular pipi ng failure are evaluated for each ta r get required to mitigate the consequences of the failure. Th is evaluation is based on the loca tion 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 protect ed 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 determine d, for each piping failure which c ould 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 brea ks are established as de sign basis environmental conditions for essential equipm ent affected by these pipe breaks and are included as environmental qualification criteria for the essential equipment. Section 3.11 describes the environmental qualification of e ssential equipment. The follow ing environmental effects are considered:1.Fluid spray - Essential equipment locate d proximate to high and moderate ener gy fluid systems are either de signed to withstand the mo st severe environmental conditions resulting from fluid spray without loss of safety function or the most MPS3 UFSAR3.6-6Rev. 30severe environmental conditions resulting from fluid spray which are humidity of 100 percent, maximum expected temperature due to the r uptured fluid system, and water spraying essential equipment.2.Flooding - Compartments a nd areas containing essentia l equipment are examine d 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 th e 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 re quired pressure differential design or the vent area required for build ing compartments. Maximum calculated temperatures are used to determine esse ntial equipment quali fication requirements.
3.6.1.3.3 Pipe Break/Crack Analysis Each pipe break and crack postulated in the high and moderate ener gy 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 para graphs describe in detail the interactions between the tar gets and the postulated breaks and cracks in each of the piping systems identified herein:1.Main Steam System2.Feedwater System3.Reactor Coolant System 4.Chemical and Volume Control System5.Steam Generator Blowdown System6.Auxiliary Feedwater SystemMain Steam System Break LocationsPipe breaks in the main steam pi ping inside the containment are pos tulated 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 containmen t penetrations are mi rror-images of each other; therefore, pipe breaks are postulated at identical locations for both lines. Similarly, the other pair of main st eam lines originating from steam ge nerators 3RCS*SG1B and 3RCS*SG1C MPS3 UFSAR3.6-7Rev. 30are symmetrical and the pipe break locations in the individual line are identi cal. Only one line of each pair is evaluated for the ef fects of the postulated piping failure; however, the results are also applicable to the ot her identical lines.
The portion of main steam piping between the containment penetrat ion (inboard side) and the first restraint outboard of the isolat ion valve is considered a brea k 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(1.2S h + S A)(3.6.1)The remainder of the main steam piping is physically located re mote from essential systems, components, and structures excep t for a relatively short sec tion 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 contai nment penetratio ns and the opera bility of the isolation valves in the event th at pipe rupture should occur at any of the postulated locations.
Consideration is given to potential damage to the contro l building wall which may adversely affect the safety related e quipment in the control room.
Figures 3.6-8 and 3.6-9 show all postulated pipe breaks inside a nd outside of the containment. Table 3.6-6 lists the postulated break locations.No single active failure has been assumed concurrent with brea ks which are postulated in the break exclusion zone since the break is postulated to evalua te environmental effects only.
Separation As mentioned above, the main st eam piping outside the containm ent, extending from the main steam valve building wall (Col. F , Figure 3.6-8) to the turbine buildi ng satisfy the physical separation criterion. Pipe breaks, however , are postulated in this re gion to assure that occurrence of such breaks do not jeopardize the integrity of the containment pe netrations and operability of the isolation valves and essential equi pment located in the control building.Pipe Whip EffectsTable 3.6-7 summarizes the results of pipe whip analys is inside and outside of containment. T o prevent whipping of the main steam piping inside the containment, whip restraints are provided to protect the crane wall and the ad joining containment wall. Similar ly , no whipping is permitted for the main steam piping outside containment to prevent damage to cont ainment penetrations, isolation valves, and control building wall by pl acing restraints at specified locations shown on Figure 3.6-8. Details of these restraints ar e 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 r upture locations.
Notes to Figure 3.6-8 provide location, t ype, and function for each pipe rupture restraint.
MPS3 UFSAR3.6-8Rev. 30Jet Impingement EffectsJet impingement effects are addressed in Table 3.6-8
.Environmental Effects Refer to Section 3.11B for information.
Conclusion Adequate protection and/or sepa ration is provided for essentia l systems and components required to mitigate the consequences of main steam piping ruptures.Essential systems, structures, and components rema in functional subsequent to a postulated main steam line break.
Feedwater System Break Locations Pipe breaks in the feedwater pi ping inside the containment are pos tulated at terminal ends and intermediate locations in accordance with Section 3.6.2.1.1
.Pipe breaks are excluded in the main steam va lve building between the containment penetration (inboard side) and the first restraint on the 20-inc h feedwater lines and between the tap into the 20-inch line and the first restra int on the 8-inch bypass line, inas much 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 ups tream of the break ex clusion zone are postulated at terminal ends and intermediate loca tions in accordance with Section 3.6.2.1.2. Figures 3.6-10 and 3.6-11 show all the terminal and intermediate br eaks inside and outside of containment. Table 3.6-9 lists the postulated break locations.
Separation The feedwater piping upstream of the break exclusion z one, extending from the main steam valve building wall (Col. F) to the tu rbine building fully meet the physi cal separation cr iterion, thereby precluding potential damage to es sential equipment. Pipe break s, however , are postulated to evaluate the effects of these breaks on the break ex clusion 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 MPS3 UFSAR3.6-9Rev. 30Table 3.6-10 summarizes the results of pipe whip analysis inside and outside containment.
Whipping of feedwater pipi ng inside the containmen t is not allowed except in the area where the whipping pipe cannot potentially damage an essential system, co mponent, 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 approxi mately 40 feet at their neares t approach. A whipping feedwater pipe can not impact the wall.
Pipe Restraints The pipe rupture restraints on the Ma in 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 EffectsJet impingement effects are addressed in Table 3.6-11
.Environmental Effects Refer to Section 3.11 for information.
Conclusion Adequate protection and/or sepa ration is provided for essentia l systems and components required to mitigate the consequenc es of piping ruptures.Reactor Coolant System Determination of the design basis break locations in the reactor c oolant system are based on the criteria defined in Regulatory Guide 1.46. It ta kes into account certa in 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 effect s fall into three separate areas: 1. primary coolant piping (including Loop Stop Valve Bypa ss Piping, Excess Letdow n 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 pr imary coolant piping at postulated locations defined in Section 3.6.2.1. Each break location and orie ntation is determined on the basis of stress and fatigue analysis. Table 3.6-12 summarizes the postulate d break locations. Only one loop (loop A) of the four identical loops is analyzed; how ever , results of analysis are applicable to all loops by virtue of their symmetry.
MPS3 UFSAR3.6-10Rev. 30The mechanistic effects of pipe breaks in the RCS hot leg, cold le g, and crossover leg need not be addressed due to the GDC4 exempt ion for primary loop pipe breaks.
Separation Protection criteria dictat e physical separation of each of the reactor cool ant loops in order to assure the continued inte grity and operability of essential systems, com ponents, and structures required to safely shut down the pl ant in the event of a loss of c oolant accident, ma in steam, or feedwater break. This protection requirement is achieved by isol ating each reactor coolant loop within a cubicle.
Section 3.8-1 discusses the design of the cubicle.Pipe Whip EffectsTable 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 integrit y and operability of the safe-shutdown targets identified in the table. To meet the relevant prot ection 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 C oolant System Primary C oolant Piping are shown on Figure 3.6-12. The postulated pipe rupt ure 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 orig inally installed to prevent prim ary loop pipe whip have been deactivated due to the GDC4 exemption.Jet Impingement EffectsJet impingement effects are addressed in Table 3.6-14
.Environmental Effects Refer to Section 3.11 for information.
Conclusion Adequate protection and/or sepa ration is provided for essentia l systems and components required to mitigate the consequences of RCS loop (i ncluding 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 MPS3 UFSAR3.6-11Rev. 30 Break Locations Figure 3.6-14 shows the design basis break s postulated in accordance wi th the rules specified in Section 3.6.2.1. Table 3.6-15 summarizes the postu lated break locations.
Separation The pressurizer tank located in proximity to the re actor coolant loop is encl osed in a cubicle to satisfy the protection crite rion. The pressurizer sur ge line, however , penetrates the cubicle wall as it connects to the hot leg of th e reactor coolant loop (Loop B). Terminal and intermediate breaks are postulated in the surge line to assess their effects (p ipe whip, jet impingement and environment) on the integrity of the pressurizer tank, pressurizer cubicle, and the reactor coolant system components located within the imme diate vicinity of the postulated breaks.Pipe Whip Effects Results of pipe whip analysis are summarized in Table 3.6-16
.Pipe RestraintsThe pipe rupture restraints on the Reactor Coolant System Pres s urizer 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 EffectsJet impingement effects are addressed in Table 3.6-17
.Environmental Effects Refer to Section 3.11 for information.
Conclusion Adequate protection and/or sepa ration is provided for essentia l systems and components required to mitigate the consequences of the Reactor Coolant System - Pressurizer Cubicle Piping ruptures.3.Safety Injection SystemThe 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).
MPS3 UFSAR3.6-12Rev. 303.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 interconn ect with the reactor coolant sy stem which operate following a loss-of-coolant accident. Where the safety inject ion 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 remai nder of the connecting systems excluded from the effects of pipe rupture.
Break Locations Figure 3.6-13 shows all postulated break locations with in the reactor coolant pressure boundary except for the accumulator piping where brea k postulation is extende d beyond the pressure boundary up to the accumulator tank dischar ge nozzle. Note that each pair of the accumulator tanks is located 180 degrees apar t 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.
SeparationThe safety injection accumulator tanks are conveniently located at the basement level of the containment structure, physically remote fr om high ener gy 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 EffectsTable 3.6-19 summarizes the pipe whip analysis.
Pipe Restraints The pipe rupture restraints on the Sa fety 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 EffectsJet impingement effects are addressed in Table 3.6-20
.Environmental Effects MPS3 UFSAR3.6-13Rev. 30 Refer to Section 3.11 for information.
Conclusion Adequate protection and/or sepa ration is provided for essentia l 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 char ging, letdown, and seal water system; chemical control, purification, and makeup sy stem; and boron thermal regeneration system.Only the charging, letdown, and seal water system is systematically analyz ed for pipe rupture, since it qualifies as a high ener gy 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.
SeparationThe charging lines inside and outsi de 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 contai nment 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 e quipment. Similarly, the charging lines in the auxiliary building are routed in an area whic h precludes potential damage to essential equipment.Pipe Whip Effects Since the separation cr iterion is fully met in the regenerative heat exchanger cubicle, the annulus area, and in the auxiliary building, prevention of pipe whip to mitig ate its consequential ef fects 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.
MPS3 UFSAR3.6-14Rev. 30 Pipe Restraints The pipe rupture restraints on th e CVCS char ging li ne are shown on Figure 3.6-15 along with the postulated pipe r upture locations.
Notes to Figure 3.6-15 provide location, type, and function for each pipe rupture restraint.Jet Impingement EffectsJet 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 postulat ed at terminal ends and intermediate locations (Figure 3.6-16). However, breaks are excluded in the portion of letdow n piping from the containment penetration to and in cluding the containment isolation va lves to ensure that Letdown Line breaks may be isolated assuming the most limiting single ac tive failure as discussed in Chapter 15, Section 15.6.2. The stresses within th is portion of the letdown line satisfy the break exclusion requirement (Equation 3.6-1).
Separation The letdown line inside and outside the contai nmen t 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 containm ent into the auxiliary building.Pipe Whip EffectsThe effects of pipe whip have been mitigat ed by the installation of rupture restraints. Table 3.6-25 provides the results of pipe whip ef fects 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 r upture locations.
Notes to Figure 3.6-16 provide location, type, and function for each pipe rupture restraint.Jet Impingement EffectsJet impingement effects are addressed in Table 3.6-26
.
MPS3 UFSAR3.6-15Rev. 30Environmental Effects Refer to Section 3.11 for information.
Conclusion Adequate protection and/or sepa ration is provided for essentia l systems and components required to mitigate the consequences of CHS charging and letdown piping ruptures.c.CHS Seal Water Injection Line Break LocationsPipe breaks in the CHS seal water injection lines are postulated at terminal ends and intermediate locations and these postulated pi pe break locations are shown on Figure 3.6-17 and listed in Table 3.6-27
.SeparationThe seal water injection line, insi de and outside containment, is located with as much separation as possible from essential equi pment by routing it in the annul us area as it penetrates the containment to the auxiliary building.Pipe Whip EffectsTable 3.6-28 indicates pipe whip effects.
Pipe whip of the seal water in jection lines, subsequent to a postulated pipe break, is not sustained since these lines conne ct the char ging pumps to the reactor coolant pumps. The flow in the seal water injection line is limited.
Pipe RestraintsBased upon the pipe whip effects noted above , pipe whip restraints are not required.Jet Impingement EffectsJet impingement effects are addressed in Table 3.6-29
.Environmental Effects Refer to Section 3.11 for information.
Conclusion Essential systems, structures, a nd components remain functional s ubsequent to a postulated pipe rupture of the CHS seal water injection system.
MPS3 UFSAR3.6-16Rev. 30Steam 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 penetrati on (outboard side) up to th e isolation valves to ensure the integrity of a minimum of two intact and func tioning Steam Generators. The stre sses along this section of the steam generator blowdown piping fully meet th e break exclusion requi rements (Equation 3.6-1).
Pipe breaks are postulated at term inal 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 bl owdown piping in areas physically remote from essential equipment to minimize the consequential ef f ects of pipe rupture.Pipe Whip EffectsTable 3.6-31 summarizes the pipe whip effects resu lting from the postulated piping failure.
Pipe Restraints The pipe rupture restraints on the stea m generator blowdown system are shown on Figure 3.6-18 along with the postulated pipe rupture locati ons and data for these rupture restraints.Jet Impingement EffectsJet impingement effects are addressed in Table 3.6-32
.Environmental Effects Refer to Section 3.11 for information.
Conclusion Adequate protection and/or sepa ration is provided for essentia l systems and components required to mitigate the consequences of steam generator blowdow n piping ruptures.Auxiliary Feedwater System Break Locations MPS3 UFSAR3.6-17Rev. 30Based upon application of the criteria specified in Section 3.6.2.1.1 , pipe breaks on the auxiliary feedwater system inside the c ontainment are postulated at term inal ends and intermediate locations.
Based upon applicat ion of the criteria specified in Section 3.6.2.1.2 , pipe breaks in the auxiliary feedwater system outside the containment are postu lated 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 sepa ration 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 Auxi liary Feedwater System piping from each Motor Driven Pump is separated fro m the piping of the other Motor Driven Pump by the ESF Building walls. This prev ents 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 Feed water 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 Pu mps 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 nom inal pipe size and wall thickne ss 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 feed water system is n ot sustained si nce backflow from each main feedwater line is prevented by dual check valves installed in the system and flow from each auxiliary feedwate r pump is limited. Table 3.6-35 provides the results of pipe whip effects for the Auxiliary Feedwater System.
Pipe RestraintsPipe whip restraints are not required for postulated pipe brea ks on the Auxiliary Feedwater System.Jet Impingement Effects MPS3 UFSAR3.6-18Rev. 30Jet impingement effects are addressed in Table 3.6-36
.In general, jet impingement fo rces, subsequent to postulate d 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, a nd components remain functional s ubsequent to a postulated pipe rupture of the auxiliary feedwater system.
 
====3.6.2 DETERMINATION====
OF BREAK LOCATIONS AND DYNAMIC EFFECTS ASSOCIATED WITH THE POST ULATED RUPTURE OF PIPING This section describes the design bases used for defining postulated pipe break and crack locations in high- and moderate-ener gy piping syst ems inside and outside of the containment, the methods of analysis used to evaluate the jet re action forces at the break locations, and the jet impingement effects and loading effects on adjacent essentia l systems, components, and structures.
3.6.2.1 Criteria Used to Define Break a nd Crack Location and ConfigurationPipe breaks and cracks are postu lated in those high- and modera te-ener gy piping systems located in proximity to essential system s, components, and structures re quired for the safe shutdown of the plant. All postulated breaks and cracks are evaluated for potential damage to essential systems, components, and struct ures due to pipe whip, jet im pingement, and environmental effects. If the damage is unacceptable, protective measures are provided either by rerouting of piping, relocation of e ssential equipment, or providing enclosur es. Where this is not feasible, pipe whip restraints and/or jet impi ngement shields are installed to protect the essential systems, components, and structures.
An unrestrained whipping pipe is considered cap able of causing circumferential and longitudinal breaks, individually , in impacted pipes of smaller nom inal 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 c onsidered inconseque ntial and is not evaluated. In all cases, the effect s of jet impingement are less se vere 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.) Ther efore, jet impingement on pipes of equal and larger nominal pipe size and wall thic kness also is considered inconsequential and is MPS3 UFSAR3.6-19Rev. 30not evaluated. In all other cases, the jet interactio n is either avoided or the tar get pipe and its supports are evaluated for the jet lo ad 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 Sect ion III Code Class 1 Piping Breaks in ASME Section III C ode Class 1 high-ener gy piping ar e 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 r un, except where the branch run is classified as part of a main run in the stress analysis and is shown to have a significant ef fect on the main run behavior.2.At intermediate locations between termin al ends where either of the following criteria are exceeded:a.At any intermediate locations betw een terminal ends where the maximum stress intensity ranges, for normal an d upset plant conditions, and for a 1/2 safe shutdown earthquake (OBE) even t transient exceed 2.4Sm (the design stress intensity as specified in Secti on III of the ASME Bo iler and Pressure Vessel Code), calculated by Equati on 10 and either Equation 12 or Equation 13 in Paragraph NB-3653 of the ASME Code, Section III, orb.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, C ode Classes 2 and 3, hi gh ener gy piping system s 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-1 120 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 e ach piping run or branch run:
MPS3 UFSAR3.6-20Rev. 301.At terminal ends of the pre ssurized portions of the runs, and2.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 ra nges 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 (1.2S h + S A).S A 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. S A is the allowable stress range for expansion stress ca lculated by the rules of NC 3600 of ASME Code Section III, 1971 editi on up to and including the summer 1973 addenda.
3.6.2.1.2 Criteria for Outside ContainmentHigh-Energy Piping Systems 1.Piping Systems Separated from Essential Structures, Systems and Components - A primary objective in the piping layout and pl ant arrangement is to have adequate separation, so that the effect s of postulated pipe breaks at any location are isolated or physically remote from essential stru ctures, systems, and components. Pipe breaks are not postulated in these separated high-energy piping systems.2.Piping Systems in Containm ent Penetration Areas Break Exclusion Zone - Breaks are not postulated in certain high ener gy piping systems in the containment penetration areas. Breaks are not postulated in portions of high ener gy 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 E ngineered 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 containm ent 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 pipi ng failure outboard of the first restraint beyond the containment isolation valves. These restraints are
 
located as close as practical to the containment isolation valves.
MPS3 UFSAR3.6-21Rev. 30The 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 ex clusion zone are designed to meet the requirements of ASME Code Section III, Subarticle NE-1 120 and the following additional design requirements:a.The following design stress and fati gue limits should not be exceeded for Class 2 piping:
 
*The maximum stress ranges as calc ulated 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 (1.2S h + S A).*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 pi ping do not exceed 1.8S
: h. 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 conn ections are minimized.d.The length of these portions of piping is reduced to the minimum length practicab le.e.The design of pipe anchors or restra ints (e.g., connecti ons to containment penetrations and pipe whip restraints) does not require welding directly to the outer surface of the pi ping (e.g., flued integrally fo r ged 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 contai nment isolation valves.f.For these portions of high-ener gy piping, inservice examination is performed in accordance with the re quirements specified in ASME Code, Section XI, with the exception that examination of ci rcumferential and MPS3 UFSAR3.6-22Rev. 30 longitudinal pipe welds between and including the boundaries of the pipe break exclusion zones is performed in accordance with the risk-informe d methodology established in WCAP
-14572, Revision 1-NP-A, Addendum
: 1. The examination includes the va lve and pipe pressure boundaries.
Details of containment pe netration, identification of pipe welds, access for inservice inspection, and points of fixi ty 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-ener gy piping are postulated at the following locations in each piping and branch run outboard of break exclusion zones (except those porti ons of piping systems identified in Section 3.6.2.1.2 , Items 1 and 2):  a.At terminal ends of the pr essurized portions of the runs.b.At the extremity of the break exclusion zone (if applicable), andc.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 S h + S A).Breaks in nonnuclear safety clas s high-ener gy piping are postulated at the following locations in each piping run or branch run: 1.At terminal ends of the pre ssurized portions of the runs, and2.At intermediate locations selected by either of the following criteria: a.At each pipe fitting, weld ed attachment, and valve, orb.At each location where the stress ra nges associated with normal and upset plant conditions and a 1/2 SSE event, calculated by Equations 9 and 10, paragraph NC-3652 of ASME Se ction III exceed 0.8 (1.2 S h + S A).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 MPS3 UFSAR3.6-23Rev. 30 through-wall leakage crac ks 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 calc ulated by Equations 9 and 10, Paragraph NC-3652 of Section III of the ASME Code) does not exceed 0.4 (1.2S h + S A).Crack Exclusion Zones are invoked for the Hydr ogen Recombiner (HCS)
System Suction piping, the Containment Atmosphere Monitoring (C MS) System Pump Suction piping, and the Containment Vacuum (CVS) Sy stem Pump Suction piping. Th e 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 excepti on to GDC-56 by placing two is olation 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-ener gy 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 (1.2S h + S A). For nonnuclear, nonseismic piping system s, throughwall leakage cracks are postulated at any location that has the worst consequences for essential structures, sy stems, or components. Only environmental effects due to fluid spray and flooding are considered.
Cracks are not postulated in m oderate-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 prev ious paragraph are applied.
Through-wall leakage cracks instead of breaks are postulated in the piping of those fluid systems that qualify as high-ener gy systems for only short operational peri ods, but qualify as moderate-energy systems for the major operational period.
An operational period is considered short if the fr action of time that the system operates within the pressure-temperature conditions specified for high-energy syst ems 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 system s operated during reactor startup, hot standby, or shutdown, qualify as high-energy systems.
MPS3 UFSAR3.6-24Rev. 30 3.6.2.1.3 Design Basis Break/Crack Types and Orientation Circumferential Pipe Breaks The following circumferential breaks are postulated in high-ener gy 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-ener gy piping runs and branch runs exceeding a nominal pipe size of 1 inch. When the maximum stress range or usage factor exceeds the limits sp ecified for break postulation, and if it is determined by detailed stress analysis, that the maxi mum 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 se lected at pipe fittings wi thout the benefit of stress calculations, breaks are postulated at the piping weld to each fitting, valve, or welded attachment. If detailed stress anal yses 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 displace ment of the ruptur ed piping sections unless physically limited by piping restraints, structural members, or piping stif fness as may be demonstrated by inelastic analysis.4.The dynamic force of the jet discharge at the break location is based on the ef fective cross-sectiona l 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-ener gy piping at the locations of each circumferential break sp ecified under Circumferent ial Pipe Breaks in this section, except as noted:1.Longitudinal breaks in piping runs and bran ch runs are postulated in nominal pipe sizes 4 inches and larger. However, wh en the maximum stress range or usage factor exceeds the limits sp ecified for break postulation and if it is determined by detailed stress analysis that the maximum stress range in the axial direction is at MPS3 UFSAR3.6-25Rev. 30 least 1.5 times that in the circumferentia l direction, then onl y a circumferential break is postulated.2.Longitudinal breaks are not pos tulated at terminal ends.3.Longitudinal breaks are assume d to result in an axial sp lit without pipe severance.
Splits are located (but not concurrently) at two diam etrically opposed points on the piping circumferen ce 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 stre ss as determined by detailed stress analysis.4.The dynamic force of the fluid jet discharg e is based on a circul ar break area equal to the ef fective cross-secti onal flow area of th e pipe at the break location, and on a calculated fluid pressure modified by an analytically or experimentally determined thrust coefficient as determined for a ci rcumferential break at the same location.
Line restrictions, flow l imiters, positive pump-controlled flow, a nd 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 th e directions defined by the stif fness 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 leakag e cracks are postulated in moderate-ener gy 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-ener gy 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 lengt h 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 compar tment, with conse quent flooding in the compartment and communicating compartm ents. Flooding ef fects are determined on the basis of a conservative ly 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.
MPS3 UFSAR3.6-26Rev. 30 3.6.2.2 Analytical Methods to Define Forcing Functions and Response Models 3.6.2.2.1 Introduction Pipe rupture analyses include cal culations to determine the fluid forces generate d 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 diamet er 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 ruptur e 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 stre ss 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 th at 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 devel oped after a postulated rupture, account for the effects of the following:a.Mass, inertia, and stiffne ss properties of the system.b.Impact and rebound (if any) as perm itted by gaps between piping and restraint.c.Elastic and inelastic deformat ion of piping and/or restraint.d.Support boundary conditions.5.An allowable design strain limit of 0.5 ulti mate 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.
MPS3 UFSAR3.6-27Rev. 30b.The design limit for pipes crushed uni formly along their length is the lesser of:1.one half of the pipe diameter, or2.the maximum flattening limits as prescribed by ASTM A530.
The above deformation limits assure that the area unde r the essentially flat portion of the materials force-deflection curve is not exceeded.6.A 10-percent increase of mini mum specified yield strength (S Y) 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 ar e based on methods suggested by Moody (1973) and include consideration of the transient pressures, velo cities, and other therm odynamic properties of the fluid. T o provide the time history of pressure, velocity, etc., the me thod of characteri stics is used to solve the continuity and momentum equations simultaneously. A gene ral description of the method can be found in gas dynamics te xtbooks (De Haller 1945, R udinger 1969, Owzarek 1968). For these one dimensional fluid mechanics analyses, the pipe r un is treated as an equivalent section of straight pipe. The calculated mome ntum 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 moment um. It is assumed that pipe bends and elbows neither attenuate the travel ing pressure waves nor cause reflections.
Immediately following the ruptur e 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 initia l blowdown condition is maintained until a return signal from the pressure reservoir reaches th e break. At this time, repeated wave reflections between the reservoir and break prevail until a steady state flow condition is established. B oundary conditions that govern the flow at the break are considered.
 
Fluid momentum changes will result in dynami c forces being exerted on pipe segments. The forcing function is calculated using Sh apiro Volume I Equation 4.20 as follows:  (3.6.2.2-1) where: P = Local static pressure (psia)FPP a-()A m*u g-------+PP a-()Ru 2 144 g-----------
-+A==
MPS3 UFSAR3.6-28Rev. 30 P a = Ambient pressure (psia)
R = Fluid density (lb m/ft 3)u = Velocity of blowdown fluid (fps)A = Local flow cross-s ectional area (in 2)g = Dimensional Constant
= 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 stea dy state condition. A decompression wave travels upstream toward the pressure source. The initia l blowdown velocity can be calcu lated by applying the momentum equation across the wave:
RCu = gP(3.6.2.2-2) (u-o) = 144g(P o-P a)/RC (3.6.2.2-3) where: C = The speed of sound (fps)
P o = The initial stagnation pressure (psia)
Therefore, the initial blowdown thrust for non-flashing water is: (3.6.2.2-4)
For transient flow analysis, th e one dimensional equations of conservation of momentum and continuity are solved simultaneously to obtain transient pressures and velo cities, which are used to calculate the transient blow down thrust. The solutions are su bject to the following boundary conditions:
U (,t)  O      at t = 0 P (,t)  P o        at t = 0 32.2ftlb m -lb fsec 2-----------------------
-m*F Ru 2 g 144 ()----------------
-A gP o P a-()144 ()RC 2----------------------------------------
P o A==
MPS3 UFSAR3.6-29Rev. 30 The initial blowdown flow remains constant unt il the decompression wave, which is reflected from the pressure source, reaches the break end. It is then refl ected again, causin g 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:  (3.6.2.2-5) which is derived by applying Ber noulli's equation across the pipe and by using the expression for the forcing function calculation, where:
L e = Total equivalent length of pipe friction f = Friction factor (Reynolds number and pipe surface roughness depend ent), obtained from Moody's chart, (1961).D = Pipe inside diameter. The friction para meter is calculated by considering all the possible losses on the line.The friction parameter fL e/D is calculated by considering all the possible losses on the line.Transient Flow With Friction With friction losses taken into consideration, tr ansient pressures and velocities of a subcooled nonflashing water line blowdown ca n 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 th e 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 al ong 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 reache d, the corresponding characteristic equation and boundary co ndition are used. Th e process proceeds until a steady state is reached.
Pt , ()P o Ru 2t , ()2 g 144 ()----------------------
at  -0==F 2 P o P a-()P o--------------------------
1 1 fL e D-------+----------------
-P o A=
MPS3 UFSAR3.6-30Rev. 30 The friction losses are expressed in terms of press ure 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 calcul ate 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 P o A for an extremely short time interval, which allows decompression waves to move out of the break regi on 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: (3.6.2.2-6) where: A ~ The break opening area of the split 3.6.2.2.2.2 Steamline BlowdownTransient Flow without Friction Steam is treated as an ideal, single-phase gas with a constant specific h eat 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 characterist ic method (Jonssen et al., 1973, Ha rtree 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 equati ons. The transient pressu re, 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 t oward the pressure reservoir. The fluid in front of the wave is at a state:
U 1 = 0 C 1 = C F 2 P o P a-()A 1 fL e D-------+------------------------------
=
MPS3 UFSAR3.6-31Rev. 30 where: U 1 = Velocity of fluid C o = Speed of sound in fluid The fluid state at the exit is at the soni c condition (Shapiro 1953 V olume 1 Equation 4.6a). (3.6.2.2-7)
The blowdown force can be calculated us ing Shapiro V olume I Equation 4.20 as: (3.6.2.2-8) where: The pressure ratio across the wave is: (3.6.2.2-9) where:T = Temperature The density ratio is: (3.6.2.2-10)Therefore, the blowdown force can be reformulated as:
U e C o------C e C o------2 K 1+------------
-0.5 0.9325===F P e P o------R e C 2 e gP o 144 ()-----------------------
+P oA==P e P o------R e R o------C e C o------2  R o C 2 o gP o 144 ()-----------------------
+P oA C o 2 KgP o 144 ()R---------------------------
-=P e P o------T e T o-----k k 1------------
C e C o------
2 k k 1------------
2 k 1+-----------
-k k 1------------
0.55====R e R o------P e P o------1 k---C e C o------
2 k k 1------------
1 k---
2 K 1+------------
-1/k 1-()0.628====
MPS3 UFSAR3.6-32Rev. 30 (3.6.2.2-11)
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 compre ssion wave. The boundary condition at the reservoir lies on the steady stat e ellipse. From, Shapiro V olume II Equation 24.36. (3.6.2.2-12)which is the energy equation applying across th e vessel-pipe inlet.
The boundary condition for this case is from Shapiro V olume II Page 963:
(3.6.2.2-13) where: C p = The constant pressure specific heat of a fluid 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 b oundary condition at the break alwa ys lies on the sonic line. For example: (3.6.2.2-14)
Then from the critical flow conditi on from Shapiro V olume I Equation 4.6a: (3.6.2.2-15) where:* = The critical flow condition Then, the steady state blowdown force is:
F (1 K)2 K 1+------------
-2 K/K-1 ()P o A+0.685 P o A==C i C o------2 K 1-2------------
-U i C o------
2+1=T o T i U 2 i 2 C p g 778 ()--------------------------
-+=Btu lb m°F---------------------
U* C o-------C* C o-------=U* C o-------C* C o-------2 k 1+------------0.9325===
MPS3 UFSAR3.6-33Rev. 30 (3.6.2.2-16)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 descri bes the variation of st eady state steam blowdown force versus friction parameter fL e/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 equa tions of mass, momentum , and ener gy simultaneous ly 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 integrat e the fluid variables which represent the pressure, velocity, and entropy along the characteristic line s and the path line. The equati ons are solved using computer program ME-143 One-Dimensiona l Unsteady Flow of a Compre ssible Fluid with Friction (UFLOW). These transient pressure s, 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 forc ing function may be used for calculations based on the ener gy balance method. The function has a magnitude ofT = KPA(3.6.2.2-17) where: P = System pressure prior to pipe break A = Pipe break areaK = Thrust coefficient (theoretical maximum)K values are 1.26 for saturated steam, water, and steam/water mixtures and 2.00 for nonflashing subcooled water
.F P* P o-------R*U*()2 P o g 144 ()-----------------------
+P o A= 1 k+()2 k 1+-----------
-
k/k-1 ()P o A1.255 P o A==
MPS3 UFSAR3.6-34Rev. 30 An amplification factor between 1.1 and 1.2 has been demons trated by testing.
To account for rebound, it is applied to the above force. Alternatively, the maximum flui d 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 math ematically as a seri es 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 pi pe in the elastic range and approximate the plastic behavior above yield.
Before a rupture, pipe is stresse d by internal pressure and is in static equilibrium. When initial conditions have an ef fect on the parameters being calcula ted, such as stresses in break exclusion regions or loads on attached components, this effect is considered.As a circumferential brea k propagates, the load-car rying metal area of the pi pe 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-depende nt blowdown force derived (Section 3.6.2.2.2). Similarly, for a longitudinal split, the crack propagation speed li mits 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 ruptur e, the inelastic system response is analyzed by the use of an elastic-plastic lumped-mass b eam element computer code su ch as DINASAW or LIMITA (Appendix Sections 3A.2.6 , 3A.2.10 , and 3A.2.11. The analysis consider s 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 includ es the restraint or barrier , and sometimes a member simulati ng the local crush resistance of the pipe. Rebound ef fects are considered by automatically c onnecting and disconnecting that member for impact and rebound, respectively.
 
3.6.2.2.4.1 Sample Pipe Rupture Dynamic Analysis Pipe rupture restraint 3FWS-PRR 5S limits the motion of the f eedwater 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 in ch strainable width and 1 inch thickness. The initial clear ance between the hot pi pe 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 co mputer code for the ma thematical model of the pipe, restraint, and intermediate structure (Figure 3.6-21). The elastic-plasti c characteristics of the pipe, strap, and intermediate st ructure were used in the computer solution in terms of bilinear engineering stress-strain relati onships with the corresponding allo wable stress not exceeding half MPS3 UFSAR3.6-35Rev. 30the uniform ultimate strain. Strain ra te sensitivities for materials below 400
°F were considered.
The fluid forcing function depicted on Figure 3.6-21 is applied at the el bow in terms of normal pressures per unit tangential length.
The restraint and intermediate structure react ion load are shown on Figures 3.6-23 through 3.6-25 which illustrate the interm ediate structure loads and Figure 3.6-26 which illustrates the restraint reaction load. The depletion of the kinetic en er gy 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 AnalysisThe energy balance technique for analyzing pipe impact equate s the work done by the escaping fluid to the ener gy absorbed in deforming the ruptured pipe and the impacted target. A steady state blowdown force is used for the energy balance analysis. Th e 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 fo rce in the direction of the displacement.The input energy is: (3.6.2.2-18) where: F b = Component of blowdown force in direction of pipe displacement L h = Length from break to plastic hinge L = Length from break to restraintg = Pipe-target gap d = Restraint deflectionThe strain energy absorbed during pipe whip and impact consists of the ener gy absorbed by pipe bending, E b, the energy absorbed by pipe crush during impact, E c, and the energy absorbed by deformation of the target, E t.To determine post-impact target deformation a nd the peak reaction for ce, the input ener gy 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 displaceme nt integral of the appropriate force-deformation curves.
EF b gd+()L h L h L---------------
-=
MPS3 UFSAR3.6-36Rev. 30Sample Energy Balance Analysis Analyze the impact of a 4-in ch, schedule 80 pipe into a pi pe 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 th rough the distance moved by the ruptured end of the pipe. The input ener gy is: (3.6.2.2-19) where: F b = Fluid blowdown force g = Acceleration gap d = Restraint deflection
 
L h = Length from break to plastic hinge L = Length from break to restraint The ratio L h/(L h-L) represents the increased pipe di splacement at the break, compared to displacement at the restraint, due to the assumed pipe rotation about a plastic hinge.The fluid force is calculated:
F b = K r K f P o A = 13.4 kips (3.6.2.2-20) where: Kr = Rebound factor (1.2)
Kf = Thrust coefficient (0.88)
P o = Initial pressure (1,106.7 psi)A = Pipe flow area (11.497 square inches)The maximum thrust coefficient in the period when the ener gy balance occurs is 1.0. However, this drops to 0.88 as soon as the deco mpression 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 in terval, 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.
E in F b gd+()L h L h L---------------
-=
MPS3 UFSAR3.6-37Rev. 30Energy may be absorbed in plastic bending of the pipe and in crush of the restraint. The ener gy absorbed by bending at the plastic hinge is:
E b = M p = M p (g + d)/(L h-L)(3.6.2.2-21) where: M p = Plastic moment = Hinge rotation The value of M p may be obtained from rigid, perfect-plastic limit theory , but, for this application, a strain hardening moment (Gerbe r 1974) is appropriate. This requi res an estimate of the hinge rotation (i.e., g, d, L h and L must be determined). The acceleration gap, g, is 1.63 inches. Let the bumper pipe have the same diameter as the proces s pipe. Set L at 20 inches to place the restraint on the straight pipe. Using the common expression for plastic hinge length, L h = 3 M p/F b , and the method described by Gerber (1974), an iterative solution shows that M equals 244.9 in-kips and L h equals 119.6 inches.Input energy: (3.6.2.2-22)Energy absorbed by the process pipe bending: (3.6.2.2-23)The difference between E in and E b is the energy absorbed by th e crush bumper and equals 22.3 +
13.6 d in-kips.Size the bumper pipe thickness subject to the following constraint:  (3.6.2.2-24) where: t b = Bumper pipe wall thickness t p = Process pipe wall thickness r b = Mean radius of bumper pipe E in F b gd+()x L h L h L-(-----------------
-26.316.1  d  in-kips
+=E b M p gd+()/L h L-()  =  42.5  d  in-kips
+=t b0.75  t p r b r p----0.1310.237 in for schedule 40
=
MPS3 UFSAR3.6-38Rev. 30 r p = Mean radius of process pipe The above identity assures that the bumper pipe will crush wit hout 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 displacemen t and impact force (Peech et al., 1977). Interpolate to find the exact point of energy balance:
d = 2.34 in E p = 54.4 in-k F = 32.4 kips The impact force is thus 2.4 ti mes 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 fl uid force did not exceed 0.88 P 0A during the energy balance event: (3.6.2.2-25) 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 stif fness of the pipe is usually considered where other energy-absorbing mechanisms are not av ailable at the point of impact. Examples include impacts into rigid displacement-limiting bumpers, concrete walls, and the omni directional restraint weldment (the latter interposes a significant mass between the impacting pipe and the energy absorbers).
Experimentally derived cr ush stif fnesses have been used. Thes e 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 sensit ivity to material strength, dynami cs, and variations in loading geometry. Computer Program LIDOP (ME-184) Local Indentation of Piping (Appendix Section 3A.2.14) augments the static crush tests.
t 2 mL h 3 3 L n L-()--------------
---------gd+()2 g-------------------
1 F b L h M p--------------------------=
MPS3 UFSAR3.6-39Rev. 30 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 crushe s against the wall, and the force tr ansmitted to the impact point by the more gradual deceleration of th e adjacent run of pipe. After impact, the pi pe 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 in to 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 incl udes the impulse of th e initial impact, as well as the subsequent fluid blowdow n force and other concurrent loads.
Four basic parameters must be determined to de fine the equivalent missile: the kinetic ener gy (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 ener gy (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 im pact of piping systems with complex geometries and can even consider multiple impact points. As an alternative to the kinetic energy, the impact force hi story (impulse) can be computed.
Regardless of which analysis met hod is used, the crush resistance of the equivalent missile and the bearing area are derived from th e experimental da ta 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 Ve rify Integrity and Operability Required pipe rupture loads to de termine 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 br eak exclusion regions are presented in Sections 3.9 and 3.6.2.1 , respectively. Criteria for ruptur e restraints are presented in Section 3.6.2.3.1
.Jet impingement loadings ar e determined as follows:1.Jet forces are represented by time-dependent forcing f unctions. The ef fects 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)
MPS3 UFSAR3.6-40Rev. 30 where: P = System pressure prior to pipe break A = Pipe break areaK = Jet coefficient (theoretical maximum)
The following K values are:a.1.26 for saturated steam, saturated water , and steam/w ater mixtures blowdown (presented in 3.6.2.2.2).b.2.00 for nonflashing subcooled wate r blowdown (presented in 3.6.2.2.2).3.In calculating the jet impingement load on an object or tar get, the retarding action of the surrounding air along the jet path is neglected. The jet imp ingement 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 blowdow n of steam or steam-water mixtures, Moody's asymptotic jet expansion mode l is utilized (Moody 1969). Jet expansion is not applicable to cases involving satu rated water or subcooled water blowdown which are below the saturation temperatur e at the corresponding ambient pressure beyond the break.4.The proportion of the total jet force act ing on a tar get 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 calculati ng the total jet impingement load. For a 20-degree divergence angle and target at a distance x, the pressure intensity at the target is:  (3.6.2.3-2) where: P 1 = Pressure intensity at the source (psi) d 1 = Diameter of the assumed circular break area (inch) x = Normal distance between th e break and the tar get (inch)
PP 1 d d 1 d 1 2 x 10 tan+----------------------------------
-2=
MPS3 UFSAR3.6-41Rev. 30 Since the jet impingement force is a dynamically applied load, the tar get is analyzed either by static methods using an appropriate dynamic load factor, or dynamically usi ng elastic or inelastic structural response codes (Section 3.8.3.3). The load combinations and design al lowables 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 RestraintsTwo basic restraint types are used, elastic and en er gy-absorbing. Elastic restraints are generally used where displacements subsequent to a postulate d pipe rupture must be minimized to restrict the break opening area, limit load s in the broken piping run, limit the pipe movement to protect some equipment, or to limit pr essure buildup to minimize exte rnal 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 di splacements of the broke n 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 de signed as a pipe support.
If an elastic restraint contacts the pipe following a ruptur e, it is designed according to the criteria for structural steel (Section 3.8.3) which in effect limits st resses 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 ar e used, with a design li mit of 50 percent of unifo rm ultimate strain. In compression, honeycomb panels or crushable pipes are used. Th e design limit for crushable ener gy absorbing components is defined in Section 3.6.2.2.1 , Item 5.Elastic intermediate structures of energy-absorb ing 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. Subs equent 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 mechanica lly attached to its support by welding or bolting (Figures 3.6-29 and 3.6-30).2.Laminated Strap Restraint - The laminated strap restra int is capable of absorbing impact loads in the outward dire ction from the supporting structure
 
MPS3 UFSAR3.6-42Rev. 30 (Figure 3.6-22). The energy-absorbing component is a "U" shaped strap consisting of multiple strips (numbe r and geometry depending on ener gy to be absorbed) of highly ductile material.
This laminated design exhibits great flexibility in application. The design minimizes bending strains, permitting th e 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 di rection 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 ener gy in tension. The crushable pipe absorbs energy from impact loads acting in an inward direction. Side load impacts are absorbed by the comb ined action of the studs and crushable pipe.Combinations of pipe crus h bumpers and laminated stra ps are used to ach ieve 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 3.63.6-1De Haller, P. 1945. The Application of Gra phical Method to Some Dynamic Problems in Gases. Sulzer T echnical Review No. 1, p 6 24.3.6-2Gerber, T. L. 1974. Plastic Deformation of Piping Due to Pipe Whip Loading. ASME Paper 74 NE 1.3.6-3Hartree, D. R. 1952. Some Pract ical Methods of Using Charact eristics in the Calculation of Non-Steady Compressible Flow. Los Alamos Report LA HU 1, p a 44, Los Alamos, N. Mex.3.6-4Jonssen, V. K.; Matthews, L.; and Spaldi ng, D. B. 1953. Numerical Solution Procedure for Calculating the Unsteady, One-Dimensi onal Flow of Compressible Fluid with Allowance for the Effect of Heat Transf er and Friction. ASME Paper 73 FE 23, Report AECU 2713, USAEC, p 1 5.3.6-5Moody, J. F. 1961. Pipe Friction Manual (chart), 3rd Edition. Hydraulic Institute, New York, N.Y.3.6-6Moody, J. F. 1973. Time Dependent Pipe Forces Caused by Blowdown and Flow Stoppage. ASME Paper 73 FE 23.
MPS3 UFSAR3.6-43Rev. 303.6-7Owzarek, J. A. 1968. Fundamentals of Ga s Dynamics. International Textbook Company, Second Printing, Scranton, Pa.3.6-8Peech, 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, Transacti ons of the Fourth International Conference on Structural Mechanics in Reactor Technology, San Francisco, Calif.3.6-9Rudinger, G. 1969. Nonsteady Duct Flow-W ave Diagram Anal ysis. Dover Publications, Inc., New York, N.Y.3.6-10Shapiro, A. H. 1953. The Dynamics and Th ermodynamics of Compressible Fluid Flow , Vol. I and II. Ronald Press, New York, N.Y.3.6-11Streeter, V. L. and Wylie, E. G. 1967.
Hydraulic Transients. McGraw-Hill Book Co., New York, N.Y., p 32 37.3.6-12Letter from the NRC to Northeast Nuclear Ener gy Compa ny (NNECO) approving NNECO's Request for Exemption from a Port ion of General Desi gn 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, 1985.3.6-13Weigand, G. G., Thompson, S. L., Tomasko, D. - "Two Phase Jet Loads" - Sandia National Laboratories, Albuquerque, NM - January 1983 for the USNRC, NUREG/CR -
2913.3.6-14WCAP-14572, Revision 1-NP-A, Addendum 1, Addendum to "W estinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Topical Report" to Address Changes to Augmen ted Inspection Requirements.
MPS3 UFSARMPS3 UFSAR3.6-44Rev. 30TABLE 3.6-1 HIGH-ENERGY SYSTEMS REMOTE FROM ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS SYSTEMCODELOCATIONMAXIMUM OPERATING TEMPERATURE (1) Approximately in Degrees Fahrenheit MAXIMUM OPERATING PRESSURE (1) Approximately in psigAuxiliary Boiler BlowdownABDAuxiliary Boiler Room, Turbine Building Condensate Polishing Area2287Auxiliary Boiler CondensateABFAuxiliary Boiler Room228165Auxiliary Boiler SteamABMAuxiliary Boiler Room366150 Chemical Feed Condensate (2)CNCTurbine Building 94 550Chemical Feed Steam GeneratorSGFTurbine Building 446 1262 Cold ReheatCRSTurbine Building 374 183 Condensate Demineralizer Mixed BedCNDCondensate Polishing Area 94 550Extraction SteamESSTurbine Building 454 424Feedwater Heater Relief Vents & DrainsSVHTurbine Building 442 375 Feedwater Pump Recirculation and Balance Drum LeakoffFWRTurbine Building 367 1632Feedwater Pump Turbine Steam &
 
ExhaustTFMTurbine Building 557 1092 High Pressure Fe edwater Heater DrainsHDHTurbine Building 450 407 MPS3 UFSARMPS3 UFSAR3.6-45Rev. 30 Hot ReheatHRSAuxiliary Boiler Room, Turbine Bldg.
510 106Hot Water HeatingHVHCondensate Polishing Area Turbine Building Service Building, Warehouse 270 185 Condenser Air RemovalARCTurbine Building 212 125 Main FeedwaterFWSTurbine Building 446 1401Radioactive Liquid Waste (3)LWSTurbine Building 266 150Waste Disposal Building, Yard Instrument AirIASTurbine Building 350110Low Pressure Feedwater Heater DrainsHDLTurbine Building 374 519 Condensate Demineralizer Liquid WasteLWCWarehouse 5 250 63Turbine Plant Miscellaneous DrainsDTMTurbine Building 556 1092Turbine Plant SampleSSTTurbine Building 5561106 Main Condensate (4)CNMTurbine Building 366 551 Nitrogen SupplyGSNTurbine Building, Yard Ambient 950TABLE 3.6-1 HIGH-ENERGY SYSTEMS REMOTE FROM ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)
SYSTEMCODELOCATIONMAXIMUM OPERATING TEMPERATURE (1) Approximately in Degrees Fahrenheit MAXIMUM OPERATING PRESSURE (1) Approximately in psig MPS3 UFSARMPS3 UFSAR3.6-46Rev. 30NOTES:(1) Maximum Operating Temperature and Maxi mum Operating Pressure are tabulated fo r the Normal Plant Conditions. These are approximate values presented to classify a system as either high or moderate energy.(2) The Chemical Feed Condensat e High Ener gy piping extends from the Condensate Chemical Addition Feed Pump Discharge to the Condensate Lines.(3) The High Energy portion of the Radioact ive Liquid Waste piping consists of the Re boiler 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 SystemCondensate Make-Up & Draw OffCNSTurbine Building, Yard 250 531Moisture Separator Reheater Vents & DrainsDSRTurbine Building 536 936 Fire Protection Low Pressure CO 2FPLAuxiliary Boiler Room, Turbine Bldg.60300Reactor Plant Aerated VentsVASWaste Disposal Building25015Moisture Separator Reheater Relief ValveDischarge and Bonnet VentHRSTurbine Building510106Moisture Separator Vents & DrainsDSMTurbine Building373518Turbine Generator Gland Seal & Exhaust SteamTMETurbine Building5571107TABLE 3.6-1 HIGH-ENERGY SYSTEMS REMOTE FROM ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)
SYSTEMCODELOCATIONMAXIMUM OPERATING TEMPERATURE (1) Approximately in Degrees Fahrenheit MAXIMUM OPERATING PRESSURE (1) Approximately in psig MPS3 UFSARMPS3 UFSAR3.6-47Rev. 30TABLE 3.6-2 HIGH-ENERGY SYSTEMS IN PROXIMITY TO ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS HIGH ENERGY SYSTEMCODELOCATIONMAXIMUM OPERATING TEMPERATURE (1) Approximately in Degrees Fahrenheit MAXIMUM OPERATING PRESSURE (1) Approximately in psigReactor Coolant - Normal ChargingRCSContainment5572485Normal Charging (2)CHSContainment, Auxiliary 190 (3)2485Reactor Coolant - Normal LetdownRCSContainment 557 2485 Normal Letdown (2)CHSContainment, Auxiliary 290 400Seal Water Injection (2)CHSContainment, Auxiliary 130 2485Reactor Coolant - Excess LetdownRCSContainment 557 2485Reactor Coolant - Loop FillRCSContainment 557 2485Reactor Coolant - Loop DrainsRCSContainment 557 2485 Reactor Coolant - High Pressure Safety Injection (4)RCSContainment 557 2485 High Pressure Safety Injection (5)SIHAuxiliary Building Ambient2485Reactor Coolant - Low Pressure Safety Injection (4)RCSContainment 557 2485Low Pressure Safety Injection Accumulator DischargeSILContainmentAmbient 695Reactor Coolant - Primary CoolantRCSContainment 557 2485 Reactor Coolant - Residual Heat Removal (6)RCSContainment 557 2485 MPS3 UFSARMPS3 UFSAR3.6-48Rev. 30Reactor Coolant - Accumulator DischargeRCSContainment5572485Reactor Coolant - Pressurizer SurgeRCSContainment5572485Reactor Coolant - Pressurizer SprayRCSContainment5572485 Reactor Coolant - Pressurizer SafetyRCSContainment5572485Reactor Coolant - Pressurizer ReliefRCSContainment5572485Reactor Coolant - Loop Stop Valve BypassRCSContainment5572485Reactor Coolant - Loop Stop Valve
 
Bypass - 1.5 inch Bypass LineRCSContainment5572485Steam Generator BlowdownBDGContainment Main Steam Valve Building (MSVB)543975Auxiliary SteamASSEngineered Safety Features (ESF) Auxiliary Building366150Hot Water HeatingHVHAuxiliary Building Fuel Building Main Steam Valve
 
Building (MSVB)270185 Boron Recovery (7)BRSAuxiliary Building 312 165Containment Vacuum Pump DischargeCVSAuxiliary Building Containment 268 1.4TABLE 3.6-2 HIGH-ENERGY SYSTEMS IN PROXIMITY TO ESSENTIAL STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)HIGH ENERGY SYSTEMCODELOCATIONMAXIMUM OPERATING TEMPERATURE (1) Approximately in Degrees Fahrenheit MAXIMUM OPERATING PRESSURE (1) Approximately in psig MPS3 UFSARMPS3 UFSAR3.6-49Rev. 30 Auxiliary CondensateCNAA uxiliary Building 366 150Reactor Plant Gaseous DrainsDGSContainment Auxiliary Building 225110Radioactive Gaseous Waste (8)GWSAuxiliary Building 350 300Turbine Plant Miscellaneous DrainsDTMMain Steam Valve Building (MSVB) Engineered Safety Features (ESF) 575 1262 Control Building Air Conditioning (9)HVCControl Building1152450Main SteamMSSContainment, Engineered Safety Features (ESF),
Main Steam Valve Building (MSVB)557 1092 Auxiliary FeedwaterFWAContainment, ESF 100 1658 Main FeedwaterFWSContainment, MSVB 446 1980 Nitrogen Supply (10)GSNAuxiliary Building Containment90950Emergency Diesel Generator Air Start (11)EGAEmergency Generator Enclosure (EGE)Ambient500TABLE 3.6-2 HIGH-ENERGY SYSTEMS IN PROXIMITY TO ESSENTIAL STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)HIGH ENERGY SYSTEMCODELOCATIONMAXIMUM OPERATING TEMPERATURE (1) Approximately in Degrees Fahrenheit MAXIMUM OPERATING PRESSURE (1) Approximately in psig MPS3 UFSARMPS3 UFSAR3.6-50Rev. 30NOTES:(1) Maximum Operating Pressure and Maximum Operating Temper ature 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 pipi ng extends from the Reactor Coolan t System isolation valves to the Char ging 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.Main Steam Safety Valve Vents & DrainsSVVMain Steam Valve Building (MSVB)510741 Fire Protection Low Pressure CO 2FPLEmergency Generator Enclosure (EGE) Control
 
Building60300Emergency Diesel Generator Exhaust and Combustion (11)EGDEmergency Generator Enclosure (EGE)8200Post Accident SamplingSSPContainment6172485Reactor Plant SamplingSSRAuxiliary Building and Containment6802485Reactor Plant Aerated VentsVASAuxiliary Building25015Steam Generator Chemical FeedSGFContainment, Main Steam Valve Building (MSVB)4461262TABLE 3.6-2 HIGH-ENERGY SYSTEMS IN PROXIMITY TO ESSENTIAL STRUCTUR ES, SYSTEMS, AND COMPONENTS (CONTINUED)HIGH ENERGY SYSTEMCODELOCATIONMAXIMUM OPERATING TEMPERATURE (1) Approximately in Degrees Fahrenheit MAXIMUM OPERATING PRESSURE (1) Approximately in psig MPS3 UFSARMPS3 UFSAR3.6-51Rev. 30(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 followi ng a Loss of Coolant Accident (LOCA) thus these systems are excluded from pipe rupture analysis. Portions of piping include d within the Reactor Coolant System pressu re boundary are analyzed for pipe rupture effects (i.e. pipe whip and fluid jet impingement).(5) The High Pressure Safety Injection (S IH) between the Charging Pump Discharge Header and the SIH Valv es 3SIH*MV8801A, and 3SIH*MV8801B is High Energy.(6) The Residual Heat Removal (RHS) Syst em, although having pressure and temperatur e above the High Ener gy Threshold is classified as a Moderate Energy System as delineated in FSAR Section 3.6.2.1.2.3. Porti ons of piping included within the Reacto r 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 Di stillation piping from the Boron Evaporator to the Boron Distillate Cooler and Boron Evaporator Bottom Cooler.(8) Radioactive Gaseous Waste (GWS) System High Ener gy 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 Ener gy components consists solely of the Control Room Air Bottle Supply.(10) Portion of Nitrogen Gas System from Pr essurized Nitrogen Gas T ubes to the Safety Injection Accumulator Tanks is High Energ y piping. Pipe breaks are excluded in this line since the line is 1 inch diameter piping.(11) The Emergency Diesel Generator (EGD)
System and related auxiliaries is consider ed a single system for evaluating the ef fec ts of pipe break. A single failure is not postulated in the redundant system as discussed in FSAR Section 3.1.1.
MPS3 UFSARMPS3 UFSAR3.6-52Rev. 30TABLE 3.6-3 MODERATE-ENERGY SYSTEMS IN REMOTE FROM ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS MODERATE ENERGY SYSTEMCODELOCATIONMAXIMUM OPERATING TEMPERATURE (1) Approximate Values TabulatedMAXIMUM OPERATING PRESSURE (1) Approximate Values TabulatedAuxiliary Boiler Fuel OilFOAAuxiliary Boiler Room100
°F 48 psigCondensate Makeup & DrawoffCNSTurbine Building, Yard 120°F 136 psig Condenser Air RemovalARCTurbine Building 170°F 125 psigCondenser Tube CleaningCWATurbine Building 95°F 29 psigDomestic WaterDWSAuxiliary Boiler, Turbine Building, Service Building 140°F 80 psigFeedwater Pump & Drive Lube OilFWLTurbine Building 145°F 20 psigFire Protection - WaterFPWAuxiliary Boiler Room, Turbine Building, Service Building, Waste Disposal Building, Transformer Area110°F 122 psig Generator Hydrogen (H
: 2) & Generator Carbon Dioxide (CO 2)GMHTurbine Building55
°F 70 psigInstrument AirIASAuxiliary Boiler Room, Turbine Building, Service Building, Waste Disposal Building115°F110 psig Boron Recovery (2)BRSWaste Disposal Building, Yard 170°F 80 psig MPS3 UFSARMPS3 UFSAR3.6-53Rev. 30 Main Condensate (3)CNMTurbine Building 200°F Hydrogen (H
: 2) GasGSHGas Storage Area, Turbine Building, Yard 90°F 100 psig Nitrogen (N
: 2) GasGSNNitrogen Storage Pad, Auxiliary Boiler Room, Service Building Ambient 200 psigPrimary Grade WaterPGSPrimary Grade Water Pump House, Yard 100°F 122 psigHot Water Pre-HeatingHVGService Building 200°FService AirSASAuxiliary Boiler Room, Turbine Building, Service Building, Waste Disposal Building, Circulating Water Pump House (CWPH), Warehouse 5115°F110 psigReactor Plant Aerated VentsVASWaste Disposal Building 200°F 15 psigReactor Plant Gaseous VentsVRSWaste Disposal Building 200°F 75 psigReactor Plant Aerated DrainsDASWaste Disposal Building 200°F 70 psigService WaterSWPAuxiliary Boiler Room, Turbine Building, Service Building 80°F 66 psigTABLE 3.6-3 MODERATE-ENERGY SYSTEMS IN REMOTE FROM ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)MODERATE ENERGY SYSTEMCODELOCATIONMAXIMUM OPERATING TEMPERATURE (1) Approximate Values TabulatedMAXIMUM OPERATING PRESSURE (1) Approximate Values Tabulated MPS3 UFSARMPS3 UFSAR3.6-54Rev. 30Turbine Generator Lube OilTMLTurbine Building 100°F0 psigWaste Oil DisposalWOSTurbine Building 100°F 15 psigWaste Water TreatmentWTWTurbine Building AmbientAtmosphericWater TreatingWTSTurbine Building, Yard 60°F 128 psigStation Vacuum PrimingVPSTurbine Building115°F 20 psigReactor Plant Solid WasteWSSWaste Disposal Building 100°F 140 psigChemical Feed ChlorinationWTCCirculating Water Pump House 80°F 248 psigTurbine Plant Component Cooling WaterCCSTurbine Building 95°F 125 psig Auxiliary Feedwater (4)FWAYard 100°F 60 psig Condensate Demineralizer Component CoolingCCDCondensate Polishing Area115°F 91 psig Quench SprayQSSYard 75°F 158 psigTraveling Screen Wash & DisposalSWTCirculating Water Pump House 80°F 135 psigYard Vacuum PrimingVPSCirculating Water Pump House115°F 30 psigCirculating WaterCWSTurbine Building, Yard, Circulating Water Pump House 80°F 26 psigTABLE 3.6-3 MODERATE-ENERGY SYSTEMS IN REMOTE FROM ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)MODERATE ENERGY SYSTEMCODELOCATIONMAXIMUM OPERATING TEMPERATURE (1) Approximate Values TabulatedMAXIMUM OPERATING PRESSURE (1) Approximate Values Tabulated MPS3 UFSARMPS3 UFSAR3.6-55Rev. 30 Note (1) Maximum Operating Pressure and Maximum Operating Temp erature for the Normal Plant C onditions are tabulated. These valu es are approximate and are presented to classify the system as high or moderate energy.Note (2) Boron Recovery High Energy piping c onsists of the Boron Evaporator Recirculation and Distillate piping from the Boron Evaporator to the Boron Distillate C ooler. Remainder of the Boron Recovery (BRS) System is Moderate Energy piping.Note (3) Moderate Energy piping for the Main Condensate (CNM) Syst em 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 Nonr eturn Valves, 6th Point Drain Loop Seal, Condenser Vacuum Breaker Valves and Loop Seals.
Note (4) Auxiliary Feedwater and Recirculat ion (FWA) System is Moderate Energy piping from the Demineralized Water Storage Tank (DWST) to the suction side of the Aux iliary Feedwater pumps. This Moderate Energy piping include s the Recirculation piping.Low Pressure Safety InjectionSILYard Ambient 30 psigContainment Recirculation SprayRSSYard115°F 225 psigTABLE 3.6-3 MODERATE-ENERGY SYSTEMS IN REMOTE FROM ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)MODERATE ENERGY SYSTEMCODELOCATIONMAXIMUM OPERATING TEMPERATURE (1) Approximate Values TabulatedMAXIMUM OPERATING PRESSURE (1) Approximate Values Tabulated MPS3 UFSARMPS3 UFSAR3.6-56Rev. 30TABLE 3.6-4 MODERATE-ENERGY SYSTEMS PROXIMATE TO ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS MODERATE ENERGY SYSTEMCODELOCATIONMAXIMUM OPERATING TEMPERATURE (1) Approximate ValuesMAXIMUM OPERATING PRESSURE (1)Approximate Values Auxiliary Feedwater & Recirculation (2)FWAEngineered Safety Features Building 100°F 48 psig Residual Heat Removal (3)RHSEngineered Safety Features Building, Containment 350°F 575 psigService WaterSWPAuxiliary Building, Engineered Safety Features Building, Circulating Water Pump House, Emergency Generator Enclosure, Control Building 95°F 66 psigFire Protection - WaterFPWAuxiliary Building, Emergency Generator Enclosure, Control
 
Building, Fuel Building, Main Steam Valve Building, Containment110°F 122 psigComponent Cooling WaterCCPA uxiliary Building, Engineered Safety Features Building, Fuel
 
Building, Containment 137°F 186 psig Control Building - Chilled WaterHVKControl Building 55°F 70 psig MPS3 UFSARMPS3 UFSAR3.6-57Rev. 30Instrument AirIASAuxiliary Building, Engineered Safety Features Building, Fuel Building, Emergency Generator Enclosure, Control
 
Building, Containment, Hydrogen Recombiner Building, Main Steam Valve
 
Building115°F110 psigRadioactive Solid WasteWSSAuxiliary Building, Fuel Building 100°F 140 psigCharging Pump CoolingCCEAuxiliary Building 105°F 58 psig Fuel Pool Cooling & PurificationSFCAuxiliary Building, Engineered Safety Features Building, Containment, Fuel Building 150°F 135 psigChemical & Volume Control (4)CHSAuxiliary Building 190°F 220 psigReactor Plant Aerated DrainsDASAuxiliary Building, Engineered Safety Features Building, Containment, Fuel Building 200°F 70 psigDomestic WaterDWSAuxi liary Building, Control Building 140°F 80 psig Fire Protection - HalonFPGC ontrol Building Ambient 36 psigTABLE 3.6-4 MODERATE-ENERGY SYSTEMS PROXIMATE TO ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)MODERATE ENERGY SYSTEMCODELOCATIONMAXIMUM OPERATING TEMPERATURE (1) Approximate ValuesMAXIMUM OPERATING PRESSURE (1)Approximate Values MPS3 UFSARMPS3 UFSAR3.6-58Rev. 30Containment Instrument AirIACContainment 120°F110 psig Containment Atmosphere MonitoringCMSAuxiliary Building, Containment 100°FAtmospheric Safety Injection Pump CoolingCCIEngineered Safety Features Building110°F 35 psigChilled Water, Containment Structure VentilationCDSAuxiliary Building, Containment, Engineered
 
Safety Features Building 90°F 140 psigPiping for Containment Purge AirHVUAuxiliary Building, Containment 120°F5 psig Containment Leak MonitoringLMSContainment, Auxiliary Building 90°F0 psig Hydrogen RecombinerHCSHydrogen Recombiner Building 135°F0 psig Chemical Feed Chlorination, Emergency Diesel GeneratorWTCCirculating Water Pumphouse 80°F 248 psigJacket & Intercooler WaterEGSEmergency Generator  80°F 50 psig Hydrogen GasGSHAuxiliary Building 90°F 100 psig Nitrogen SupplyGSNAuxiliary Building, Containment Ambient 200 psigTABLE 3.6-4 MODERATE-ENERGY SYSTEMS PROXIMATE TO ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)MODERATE ENERGY SYSTEMCODELOCATIONMAXIMUM OPERATING TEMPERATURE (1) Approximate ValuesMAXIMUM OPERATING PRESSURE (1)Approximate Values MPS3 UFSARMPS3 UFSAR3.6-59Rev. 30 Quench SprayQSSEngineered Safety Features Building, Containment 75°F 158 psigNeutron Shield Tank CoolingNSSContainment 135°F 20 psig SanitaryPBSControl Building 85°FStatic HeadReactor Plant Gaseous VentsVRSAuxiliary Building, Containment 200°F 75 psig Service AirSASAuxiliary Building, Engineered Safety Features Building, Fuel Building, Emergency Generator Enclosure, Control
 
Building, Hydrogen Recombiner Building115°F110 psigEmergency Generator FuelEGFEmergency Generator  20°F 32 psigLow Pressure Safety InjectionSILEngineered Safety Features Building, Containment115°F 235 psig Floor Equipment DrainageDNFCirculating Water Pumphouse, Engineered Safety Features Building, Auxiliary Building, Fuel Building AmbientAmbientTABLE 3.6-4 MODERATE-ENERGY SYSTEMS PROXIMATE TO ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)MODERATE ENERGY SYSTEMCODELOCATIONMAXIMUM OPERATING TEMPERATURE (1) Approximate ValuesMAXIMUM OPERATING PRESSURE (1)Approximate Values MPS3 UFSARMPS3 UFSAR3.6-60Rev. 30 Roof DrainageDNREngineered Safety Features Building, Auxiliary Building, Fuel Building, Main Steam Valve Building, Control
 
Building 90°FA m b i e n tReactor Coolant Pump Oil CollectionFPRContainment 160°FAtmosphericPrimary Grade WaterPGSAuxi liary Building, Engineered Safety Features Building, Fuel Building, Containment 100°F 122 psigContainment Recirculation SprayRSSEngineered Safety Features Building, Containment115°F 225 psig Boron Recovery (5)BRSAuxiliary Building 170°F 80 psigEmergency Generator Lube OilEGOEmergency Generator  60°F 150 psig Condenser Air Removal ARCAuxiliary Building 170°F 125 psigReactor Plant Gaseous DrainsDGSAuxiliary Building, Containment 165°F 104 psigReactor Plant Aerated VentsVASAuxiliary Building 120°F0 psigContainment Vacuum (6)CVSAuxiliary Building, Containment 90°F5 psigTABLE 3.6-4 MODERATE-ENERGY SYSTEMS PROXIMATE TO ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)MODERATE ENERGY SYSTEMCODELOCATIONMAXIMUM OPERATING TEMPERATURE (1) Approximate ValuesMAXIMUM OPERATING PRESSURE (1)Approximate Values MPS3 UFSARMPS3 UFSAR3.6-61Rev. 30NOTES:(1) Maximum Operating Pressure and Maximum Operating Temperat ure are tabulated for the Norm al Plant Conditions. These are approximate values presented to classify th e system as high and moderate energy.(2) The Auxiliary Feedwater and Recirculation (FWA) System Mode rate Ener gy 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 FSAR Section 3.6.2.1.2.3.(4) The Moderate Energy piping fo r the Chemical Volume Control (C HS) System extends from the Le tdown Heat Exchanger pressure reducing valve via the Mixed Bed Deminerali zer 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 Dis tillate piping from th e Boron Evaporator to the Boron Distillate Cooler. Remainder of the Boron Recovery (BRS) System is Moderate Energy piping.(6) Containment Vacuum (CVS) Sy stem Moderate Ener gy piping exte nds from Inside Containment to the Containment Vacuum Pump suction. The remainder of Containment Vacuum (CVS) System piping is High Energy.
Feedwater Lube OilFWLEngineered Safety Features Building 20°F 140 psigHot Water Pre-HeatingHVGAuxiliary Building 185°F 170 psigTABLE 3.6-4 MODERATE-ENERGY SYSTEMS PROXIMATE TO ESSENTIAL STRUCTURES, SYSTEMS, AND COMPONENTS (CONTINUED)MODERATE ENERGY SYSTEMCODELOCATIONMAXIMUM OPERATING TEMPERATURE (1) Approximate ValuesMAXIMUM OPERATING PRESSURE (1)Approximate Values MPS3 UFSARMPS3 UFSAR3.6-62Rev. 30TABLE 3.6-5 ESSENTIAL STRUCTURES , SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN SYSTEM BUILDINGREQUIR ED SAFETY FUNCTIONReactor Coolant (RCS) (1) (2)ContainmentReactor Core In tegrity and Heat RemovalReactor Vessel 3RCS*RV1ContainmentPressure B oundary integrity needed to maintain fuel within acceptable temperature limitsReactor Vessel 3RCS*RV1 Head VentContainmentProvi de Letdown for Reactivi ty & Inventory Control and Reactor Coolant Pressure Control Pressurizer 3RCS*TK1ContainmentPressure Boundary - Maintain RCS Pressure ControlSteam Generators 3RCS*SG1A, SG1B, SG1C, and SG1DContainmentRemove Heat from Core Reactor Coolant Pumps 3RCS*P1A, P1B, P1C, and P1DContainmentMaintain Pre ssure Boundary IntegrityReactor Coolant Loop Stop Valves 3RCS*MV8002A, 3RCS*MV8002B, 3RCS*MV8002C, and 3RCS*MV8002DContainmentMaintain Pressure Boundary Integrity - Primary BoundaryReactor Coolant Loop Stop Valve Bypass ContainmentPiping and Valves on the Reactor Coolant System Up to the Class 1/Class 2 Boundary as Delineated in FSAR Table 3.5-2ContainmentMaintain Pressure Boundary Integrity - Primary BoundaryPressurizer Power Operated Relief Valv esOperability - Alternate Means of RCS Depressurization(PORV's) 3RCS*PCV455A, and 3RCS*PCV456Containment Control Rod Drive MechanismContainmentInsert Control Rods to Stabilize Plant at Hot StandbyPressurizer Safety Valves 3RCS*SV8010A, B, CContainmentOperability - Ov erpressure Protection for RCS System MPS3 UFSARMPS3 UFSAR3.6-63Rev. 30PORV Block Valves 3RCS*MV8000A, and 3RCS*MV8000BContainmentOperability - Support PORV FunctionLetdown/RCS Isolation Valves 3RCS*LCV459 &
LCV460ContainmentIsolates Reactor Coolant System for Letdown Line Break Residual Heat Removal (RHS) (1) (2)ESFLow Pressure Safety Injection Path and Reactor Coolant Heat Removal Residual Heat Removal Heat Exchanger Flow Control ValvesOperability - Supports RHS Function for Controlled
 
Cooldown3RHS*FCV610, and 3RHS*FCV611ESFOperability - Supports RHS Function for Controlled Cooldown 3RHS*FCV618, and 3RHS*FCV619 ESFOperability - Supports RHS Function for Controlled Cooldown 3RHS*HCV606 and 3RHS*HCV607 ESFOperability - Supports RHS Function for Controlled Cooldown Three RHS Series Isolation Valves 3RHS*MV8701A, MV8701B, 8701C, 8702A, 8702B, and 8702CContainment & ESFOperability - Initiate Second Stage of RHS by Opening all three Isolation Valves on either TrainResidual Heat Removal (RHS) Pumps 3RHS*P1A,BESFOperability - Ta kes Suction from RCS Hot Leg & RWSTRHS Heat Exchangers 3RHS*E1A, and 3RHS*E1BESFOpera bility - Cools Reactor C oolant from Hot Leg and Discharges Reactor Coolan t Back to Cold Leg via Safety Injection System RHS Pump Suction 3SIL*MV8812A, and
 
3SIL*MV8812B ESFOperability - During Second Stage of Residual Heat Removal and Long Term Sump RecirculationTABLE 3.6-5 ESSENTIAL STRUCTURES , SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)
SYSTEM BUILDINGREQUIR ED SAFETY FUNCTION MPS3 UFSARMPS3 UFSAR3.6-64Rev. 30RHS Pump Discharge Containment Isolation Valves 3SIL*MV8809A, and 3SIL*MV8809BESFOperability - Containment Isolation and Long Term Sump Recirculation3SIL*MV8716A and 3SIL*MV8716BESFOperability Required to Support Long Term Sump Recirculation3SIL*MV8804A and 3SIL*MV8804BESFOperability Required to Support Long Term Sump RecirculationChemical & Volume Control (CHS) (1) (2)High Pressure Safety Injection, Reactivity &
 
Inventory Control Normal Charging Pumps 3CHS*P3A,B,CAuxiliaryAuxiliaryOperability - Reactiv ity & Inventory Control as well as Pressure Control, and Safety InjectionBoric Acid Transfer Pumps 3CHS*P2A,BAuxiliaryOperabilityBoric Acid Tanks 3CHS*TK5A,BAuxiliary Maintain Pressure Boundary IntegrityBoric Acid Blender 3CHS*BL1Auxiliary Maintain Pressure Boundary IntegrityCharging Pump Suction 3CHS*LCV112B, C, D, and EAuxiliaryOperability - Reactiv ity & Inventory ControlCharging Pump Discharge Containment Isolation Valves 3CHS*MV8105, and 3CHS*MV8106AuxiliaryOperability -
Containment IsolationNormal Letdown Line Containment Isolation Valve
 
3CHS*CV8152 and 3CHS*CV8160AuxiliaryOperability -
Containment IsolationLetdown Line 3CHS*AV8149A,B,C ContainmentMaintain Pr essure Boundary IntegrityTABLE 3.6-5 ESSENTIAL STRUCTURES , SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)
SYSTEM BUILDINGREQUIR ED SAFETY FUNCTION MPS3 UFSARMPS3 UFSAR3.6-65Rev. 30Reactor Coolant Pump Seal Water Injection Containment Isolation Valves 3CHS*MV8109A, 3CHS*MV8109B, 3CHS*MV8109C, and 3CHS*MV8109DAuxiliaryOperability -
Containment Isolation3CHS*MV8111A, 3CHS*MV8111B, and 3CHS*MV8111CAuxiliaryOperability -
Supports ECCS Function 3CHS*MV8512A, and 3CHS*MV8512BAuxiliaryOperability - Supports ECCS Function3CHS*MV8511A, and 3CHS*MV8511BAuxiliary Operability - Supports ECCS Function3CHS*MV8110AuxiliaryOperabilit y - Supports ECCS FunctionHigh Pressure Safety Injection (SIH) (1) (2)ESFEmergency Core Cooling System (ECCS)
High Pressure Safety Injection Pumps 3SIH*P1A,BESFOperability-Supply Borated Water to all Four RCS LoopsValves 3SIH*MV8801A, and 3SIH*MV8801BAuxiliaryOperability - Containment Isolation 3SIH*MV8802A, 3SIH*MV8802B, and
 
3SIH*MV8835 ESFOperability - Containment Isolation3SIH*MV8813, 3SIH*MV8814, and 3SIH*MV8920ESF O perability - Supports ECCS FunctionLow Pressure Safety Injection (SIL) (1) (2) Safety Injection Accumula tors 3SIL*TK1A, 1B, 1C, and 1DContainmentInject Borated Water to RCS Cold Leg and Large LOCATABLE 3.6-5 ESSENTIAL STRUCTURES , SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)
SYSTEM BUILDINGREQUIR ED SAFETY FUNCTION MPS3 UFSARMPS3 UFSAR3.6-66Rev. 30 Safety Injection Accumulator Isolation Valves 3SIL*MV8808A, MV8808B, MV8808C, and MV8808DContainmentOperability - Preclude Accumulator Injection during RCS Depressurization - Accumulators are Isolated or VentedSafety Injection Accumulator Purge Valves 3SIL*SV8875A, B, C, D, E, F, G, and HContainmentOperability - Accumulator Purge Valves3SIL*HCV943A, and 3SIL*HCV943BContainmentOperability - Purge Accumulator if Accumulator Isolation Valves Fail to CloseMain Feedwater (FWS)
(1) (2)Decay Heat Removal by Supplying Cooling Water to the Steam GeneratorsMain Feedwater Valves 3FWS*CTV41A, B, C, DMSVOperability - Containment IsolationMain Feedwater Valves 3FWS*FCV510, 520, 530, and
 
540MSVOperability - Valves Close on Feedwater Isolation Signal Main Feedwater Bypass Valves 3FWS*LV550, 560, 570, and 580MSVOperability - Valves Close on Feedwater Isolation SignalAuxiliary Feedwater (FWA)
(1) (2)Decay Heat Removal by Supplying Cooling Water to the Steam GeneratorsAuxiliary Feedwater Motor Driven Pumps 3FWA*P1A, and 3FWA*P1BESFOperability - Provide Redundant Water Supply to at Least Two Steam Generators Until Initiation of RHSAuxiliary Feedwater Turbine Driven Pump 3FWA*P2E SFOperability - Alternate Path for Water SupplyDemineralized Water Storage Tank 3FWA*TK1 Yard Pr essure Boundary - Supply for Auxiliary Feedwater Pump SuctionTABLE 3.6-5 ESSENTIAL STRUCTURES , SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)
SYSTEM BUILDINGREQUIR ED SAFETY FUNCTION MPS3 UFSARMPS3 UFSAR3.6-67Rev. 30Auxiliary Feedwater Supply Valves 3FWA*MOV35A, B, C, D and 3FWA*HV36A, B, C, D ESFOperability - Containment Isolation Auxiliary Feedwater Supply Hand Control Valves 3FWA*HV31A, B, C, and D and 3FWA*HV32A, B, C, and DESFOperability - Control Flow to Steam Generator Level to Support Cooldown Containment Recirculation Spray (RSS) (1) (2)ESFRecirculation of Emerge ncy Core Cooling System (ECCS) Water following a Design Basis Accident (DBA)Containment Recirculati on Spray Pumps 3RSS*P1A, 3RSS*P1B, 3RSS*P1C, and 3RSS*P1D ESFOperability Containment Recirculation Spray Pumps are started on receipt of a RWST Low-Low signal
 
coincident with a CDA Signal Containment Recirculation Spray Coolers 3RSS*E1A, 3RSS*E1B, 3RSS*E1C, and 3RSS*E1D ESFOperability Containment Recirculation Water Flows Shell side of the Heat ExchangersContainment Recirculation Spray Valves 3RSS*MOV20A, B, C, and D ESFOperability - Containment Isolation Containment Recirculation Spray Suction Isolation Valves 3RSS*MOV23A, B, C, D ESFOperability - Containment Isolation 3RSS*MOV8837A, and B a nd 3RSS*MOV8838A, and B ESFOperability - Switchover to Long Term Sump RecirculationQuench Spray (QSS) (1) (2)ESFMaintain Integrity of th e Containment by Removing Heat from Containment Atmosphere following a LOCA or a Main Steam Line or Feedwater Line BreakQuench Spray Pumps 3QSS*P3A, and 3QSS*P3BESF Operability - Provide Borated Water to QSS HeadersTABLE 3.6-5 ESSENTIAL STRUCTURES , SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)
SYSTEM BUILDINGREQUIR ED SAFETY FUNCTION MPS3 UFSARMPS3 UFSAR3.6-68Rev. 30Refueling Water Storage Tank 3QSS*TK1Yard Provide Chilled Water Supply to QSS Pump Suction and suction for the RHS, SIH, and CHS Pumps for the ECCS FunctionQuench Spray Valves 3QSS*MOV34A,B ESFOpera bility - Containment Isolation and Open subsequent to a CDA Signa l to Supply QSS HeadersSteam Generator Blowdown (BDG)Maintain Pressu re Boundary Integrity to preserve Heat Removal via Steam GeneratorsSteam Generator Blowdown Containment Isolation Valves 3BDG*CTV22A, CTV22B, CTV22C, and CTV22D MSVOperability - Containment IsolationMain Steam (MSS)
(1) (2)Maintain Pressure Boundary of Steam Generators &
Control Steam Release for Heat RemovalMain Steam Isolation Valves 3MSS*CTV27A,B,C,DMSVOperability - Containment IsolationMain Steam Safety Valves 3MSS*RV22A, B, C, DMSVOperability & Over Pressure ProtectionMain Steam Relief Valves 3MSS*RV23A, B, C, and DMSVOperability & Over Pressure Protection3MSS*RV24A,B,C and D 3MSS*RV25A, B, C and D and 3MSS*RV26A, B, C and DMSVOperability & Over Pressure ProtectionMain Steam to Turbine Driven Auxiliary Feedwater Pump 3MSS*MOV17A,B,D and 3MSS*AOV31A, 3MSS*AOV31B, and 3MSS*AOV31DESFOperability - Containment IsolationOperability - Support Turbine Driven Auxiliary
 
Feedwater Pump OperationMain Steam Pressure Relieving Bypass Valves
 
3MSS*MOV74A, B, C, DMSVOperability - Provides Means to Reduce Steam Pressure during First Stage of RCS Cool DownTABLE 3.6-5 ESSENTIAL STRUCTURES , SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)
SYSTEM BUILDINGREQUIR ED SAFETY FUNCTION MPS3 UFSARMPS3 UFSAR3.6-69Rev. 30Main Steam Pressure Relieving Valves 3MSS*PV20A, 20B, 20C, and 20DMSVOperability - Primary Means to Reduce Steam Pressure during First Stage of RCS Cool DownMain Steam Pressure Relieving Block Valves 3MSS*MOV 18A, B, C, DMSVOperability - Isolation 3MSS*HV28A, 3MSS*HV28B, 3MSS*HV28C, and 3MSS*HV28DMSVOperability - Containment IsolationReactor Plant Component Cooling Water (CCP) (1) (2)AuxiliaryRemove Heat from Safety Components and Transfer Heat to the Ultimate Heat SinkReactor Plant Component Cooling Water Pumps 3CCP*P1A, 3CCP*P1B, and 3CCP*P1CAuxiliaryOperability - Component Cooling Water Pumps Circulate Water Through the Various Closed LoopsReactor Plant Component Cooling Water Heat
 
Exchanger 3CCP*E1A, 3CCP*E1B, and 3CCP*E1CAuxiliaryOperability - Transfer Heat to Service Water SystemResidual Heat Removal Heat Exchanger Discharge Valves 3CCP*FV66A , and 3CCP*FV66B ESFOperability - Normally Closed Valves Prevent Flow Through RHS Heat Exchanger During Normal Plant OperationValves Open Upon Residual Heat Removal System Initiation to Provide Component Cooling Water to RHS Heat ExchangerReactor Plant Component Cooling Water Surge Tank 3CCP*TK1 AuxiliaryPressure Boundary Integrity - Provide Sufficient NPSH to CCP PumpsReactor Plant Component Cooling Water Valves
 
3CCP*MOV45A, and 3CCP*MOV45BAuxiliaryOperability -
Containment IsolationValves 3CCP*MOV48A, and 3CCP*MOV48BAuxilia ry Operability - Containment IsolationTABLE 3.6-5 ESSENTIAL STRUCTURES , SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)
SYSTEM BUILDINGREQUIR ED SAFETY FUNCTION MPS3 UFSARMPS3 UFSAR3.6-70Rev. 30Valves 3CCP*MOV49A, and 3CCP*MOV49BAuxiliaryOperability - Containment IsolationCharging Pump Cooling (CCE)
(1) (2)Auxiliary Operability - S upports CHS Pump Function Safety Injection Pump Cooling (CCI) (1) (2)ESFOperability - Supports SIH Pump FunctionHot Water Heating (HVH)
(3)Auxiliary NoneIsolation Valves 3HVH-AOV135A, B and 3HVH-AOV136A, B ServiceOperability - A High Energy Line Break on HVH in the Auxiliary Building Trips a Pressure Switch which Closes these Valves to Isol ate Flow to the Auxiliary BuildingAuxiliary Steam (ASS)
(3)Auxiliary NoneIsolation Valves 3ASS-AOV102A and 3ASS-AOV102BTurbine Operability - A High Energy Line Break on Auxiliary Steam (ASS) System in the Auxiliary Building Trips a Pressure Switch which Closes these Valves to Isolate
 
Flow to the Auxiliary BuildingService 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 CWPHOperability Provide Cooling Water to Safety Related Components Containment Recirculation Coolers Service Water 3SWP*MOV54A, B, C, D and 3SWP*MOV57A, B, C, DESFOperability - Provide Cooling Water for Containment Recirculation Coolers 3R SS*E1A, E1B, E1C, and E1DTABLE 3.6-5 ESSENTIAL STRUCTURES , SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)
SYSTEM BUILDINGREQUIR ED SAFETY FUNCTION MPS3 UFSARMPS3 UFSAR3.6-71Rev. 30Reactor Plant Component Cooling Water Heat Exchanger Service Water 3SWP*MOV50A, and MOV50BAuxiliaryOperability - Normally Open Providing Cooling Water for CCP and Close Subsequent to a LOCA or a HELB Inside ContainmentService Water Valves 3SWP*AOV39A and 3SWP*AOV39BEGEOperability - Valves Open to Supply Cooling Water to Diesel GeneratorsEmergency Diesel Generator Fuel Oil (EGF)
(1) (2)EGEOperation of the Diesel GeneratorsEmergency Diesel Generator Fuel Oil (EGF) Storage Motor Driven Fuel Pump 3EGF*P2A, and 3EGF*P2B EGEOperability - Supports Diesel Generator FunctionEngine Driven Fuel Pump 3EGF*P3A, and 3EGF*P3BEG EOperability - Supports Di esel Generator FunctionFuel Oil Storage Tanks 3EGF*TK1A, and 3EGF*TK1BFuel Oil VaultPressure Boundary -
Supports Diesel Generator FunctionFuel Oil Day Tank 3EGF*TK2A, and 3EGF*TK2BEGE Pressure Boundary - Supports Diesel Generator FunctionFuel Oil Transfer Pumps 3EGF*P1A,1B,1C, and 1DEG EOperability - Supports Di esel Generator FunctionEmergency Diesel Generator Cooling Water (EGS) (1) (2)EGEOperator of the Diesel GeneratorsEmergency Diesel Generator Air Cooler Water Heat Exchanger 3EGS*E1A, and 3EGS*E1B EGEOperability - Supports Diesel Generator FunctionEmergency Diesel Generator Jacket Water Electric Heaters 3EGS*H1A, and 3EGS*H1B EGEPressure Boundary - Supports Diesel Generator FunctionEmergency Diesel Generator Jacket Water Circulating Water Pump 3EGS*P2A, and 3EGS*P2B EGEPressure Boundary - Supports Diesel Generator FunctionTABLE 3.6-5 ESSENTIAL STRUCTURES , SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)
SYSTEM BUILDINGREQUIR ED SAFETY FUNCTION MPS3 UFSARMPS3 UFSAR3.6-72Rev. 30Emergency Diesel Generator Fresh Water Expansion Tank 3EGS*TK1A, and 3EGS*TK1BEGEPressure Boundary - Supports Diesel Generator FunctionEmergency Diesel Generator Jacket Water Cooler 3EGS*E2A, and 3EGS*E2BEGEOperability - Supports Diesel Generator FunctionEmergency Diesel Generator Cooling Water (EGS) Lube Oil Heat Exchanger 3EGS*E3A, and 3EGS*E3BEGEOperability - Supports Diesel Generator FunctionEmergency Diesel Generator Cooling Water (EGS)
Governor Lube Oil Cooler 3EGS*E4A, and 3EGS*E4BEGEOperability - Supports Diesel Generator FunctionEmergency Diesel Generator Cooling Water (EGS)
Engine Driven Pump 3EGS*P1A and 3EGS*P1BEGEOperability - Supports Diesel Generator FunctionEmergency Diesel Generator Cooling Water (EGS) Engine Driven Intercooler Water Pump 3EGS*P3A and
 
3EGS*P3BEGEOperability - Supports Diesel Generator FunctionEmergency Diesel Generator Lube Oil (EGO) (1) (2)EGEOperation of the Diesel GeneratorsEmergency Diesel Generator Lube Oil (EGO) Lube Oil Pumps 3EGO*P3A, and 3EGO*P3B EGEOperability - Supports Diesel Generator FunctionLube Oil Pumps 3EGO*P1A, and 3EGO*P1BEGE Oper ability - Supports Dies el Generator Function Lube Oil Pumps 3EGO*P2A, and 3EGO*P2A, and 3EGO*P2B EGEOperability - Supports Diesel Generator FunctionLube Oil Pumps 3EGO*P4A, and 3EGO*P4BEGE Oper ability - Supports Dies el Generator FunctionEmergency Diesel Generator Pre Lube Oil (EGO) Filter 3EGO*FLT1A, and 3EGO*FLT1B EGEOperability - Supports Diesel Generator FunctionTABLE 3.6-5 ESSENTIAL STRUCTURES , SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)
SYSTEM BUILDINGREQUIR ED SAFETY FUNCTION MPS3 UFSARMPS3 UFSAR3.6-73Rev. 30Emergency Diesel Generator Lube Oil (EGO) Strainers 3EGO*STR1A,5A and 3EGO*STR1B,5BEGEOperability - Supports Diesel Generator FunctionEmergency Diesel Generator Lube Oil (EGO) Lube Oil Suction Strainer 3EGO*STR4A, and 3EGO*STR4BEGEOperability - Supports Diesel Generator FunctionEmergency Diesel Generator Lube Oil (EGO) Pre Lube Oil and Filter Pump Suction Strainer 3EGO*STR2A, and 2BEGEOperability - Supports Diesel Generator FunctionEmergency Diesel Generator Lube Oil (EGO) Pre Lube Oil Heater 3EGO*H1A, and 3EGO*H1BEGEOperability - Supports Diesel Generator FunctionEmergency Diesel Generator Air Start (EGA)
(2)EGEOperation of the Diesel GeneratorsEmergency Diesel Generator Air Start Air Receiver Tanks 3EGA*TK1A, 3EGA*TK1B and 3EGA*TK2A, 3EGA*TK2B EGEPressure Boundary - Air Start FunctionEmergency Diesel Generator Air Start (EGA) Air Tanks 3EGA*TK3A and 3EGA*TK3B EGEPressure Boundary - Air Start Function Diesel Generator Exhaust & Combustion (EGD) (1) (2)EGEOperation of the Diesel GeneratorsEmergency Diesel Generator Exhaust & Combustion (EGD) Exhaust Silencer (i.e. Muffler) 3EGD*SIL3A, 3B EGEPressure Boundary - Supports Diesel Generator FunctionEmergency Diesel Generator Exhaust & Combustion (EGD) Intake Air Filter a nd Silencer 3EGD*SIL1A, and 1B EGEPressure Boundary - Supports Diesel Generator FunctionTABLE 3.6-5 ESSENTIAL STRUCTURES , SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)
SYSTEM BUILDINGREQUIR ED SAFETY FUNCTION MPS3 UFSARMPS3 UFSAR3.6-74Rev. 30Emergency Diesel Generator Exhaust & Combustion (EGD) Intake Air Silencer 3EGD*SIL2A, and 3EGD*SIL2BEGEPressure Boundary - Supports Diesel Generator FunctionPost DBA Hydrogen Recombiner Control Valves 3HCS*V2, V3, and V6, and 3HCS*V9, V10, and V13RecombinerOperability - Containment Isolation Containment Isolation (1) (2)Containment and
 
Connecting Structures Maintain Containment Integrity After an AccidentContainment Isolation Valves (4)Containment and Connecting StructuresOperability - Powered Valves Required to Isolate
 
ContainmentHeating, Ventilation, and Ai r Conditioning and Cooling (HVAC) (1) (2)Containment and Connecting Structures Operability (5)Instrument and Controls (I&C) (1) (2)Containment and
 
Connecting Structures Operability (6)Electrical Equipment (1) (2)Containment and Connecting Structures Operability (7)Building Structures (1)Support and Protect Safety Related SystemsContainment StructureContainmentSecondary Barrier Providing Isolation of Reactor Coolant System and the Tertiary Barrier Between the
 
Reactor Core and the Outside AtmosphereTABLE 3.6-5 ESSENTIAL STRUCTURES , SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)
SYSTEM BUILDINGREQUIR ED SAFETY FUNCTION MPS3 UFSARMPS3 UFSAR3.6-75Rev. 30NOTES:(1) Hazards Review Program addres ses Safe Shutdown Requirements which are scenar io 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 Auxiliar y Steam (ASS) System are Non Safety Related Systems.Neutron Shield Tank 3NSS*TK2 ContainmentPrimary Barrier Provid ing Isolation of Reactor CorePrimary Shield WallContainmentPrimary Barrier Providing Isolation of Reactor Core Reinforced Concrete Internal Substructure ContainmentSecondary Barrier Providing Isolation of Reactor CoreReinforced Concrete Internal Floors, Walls, and EnclosuresContainmentSecondary Barrier Providing Isolation of Reactor Coolant SystemContainment Structure LinerContainmentProvide a Leak Tight Membrane Containment PenetrationsContainmentM aintain Pressure Boundary Integrity Auxiliary BuildingSupports Systems required for Safe Shutdown Engineered Safety Features Building Suppor ts Systems required for ECCS and Safe ShutdownMain Steam Valve Building Supports Systems required for Safe ShutdownEmergency Diesel Generator EnclosuresSupports Systems required for Emergency Power Control Building Supports System s required for Safe ShutdownCirculating Water PumphouseSupports Sy stems required for Safe Shutdown Containment Enclosure Building Supports Systems required for Post Accident MitigationTABLE 3.6-5 ESSENTIAL STRUCTURES , SYSTEMS, AND COMPONENTS REQUIRED FOR SAFE REACTOR SHUTDOWN (CONTINUED)
SYSTEM BUILDINGREQUIR ED SAFETY FUNCTION MPS3 UFSARMPS3 UFSAR3.6-76Rev. 30(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 Accide nt Mitigation and Safe Shut down are described in FSAR Chapter 7.(7) Electrical Equipment re quired to support the Systems and Co mponents delineated in this Tabl e and the Electrical Systems req uired for Accident Mitigation and Safe Shutdown are described in FSAR Chapter 8.
MPS3 UFSARMPS3 UFSAR3.6-77Rev. 30TABLE 3.6-6 POSTULATED BREAKS MAIN STEAM SYSTEM LINE DESIGNATION BREAK #BUILDINGELEVATIONBREAK TYPE (1)TOTAL ADDITIVE STRESSFIGURE 3.6-3-MSS-030-92-21Containment89'-1"CBN/A - Terminal End83-MSS-030-92-22Containment76'-6"CBN/A - Arbitrary Intermediate83-MSS-030-92-23Containment66'-3"CBN/A - Arbitrary Intermediate83-MSS-030-104-317MSVB66'-3"CBN/A - Terminal End8 3-MSS-030-25-44Turbine56'-9"CBN/A - Arbitrary Intermediate83-MSS-030-25-45Turbine56'-5"CB & LSAbove Threshold8 3-MSS-030-25-46Turbine54'-8"CBN/A - Terminal End8 3-MSS-030-67-47Turbine54'-6"CB(2)83-MSS-030-67-48Turbine54'-6"CB & LS(2)83-MSS-030-67-49Turbine54'-6"CB(2)8 28" G.E. Piping10Turbine40'-0"CB(2)828" G.E. Piping11Turbine36'-0"CB & LS(2)828" G.E. Piping12Turbine32'-6"CB & LS(2)8 28" G.E. Piping13Turbine31'-10"CB & LS(2)828" G.E. Piping14Turbine35'-4"CB & LS(2)828" G.E. Piping15Turbine72'-11"CB & LS(2)8 28" G.E. Piping16Turbine76'-5"CB(2)83-MSS-030-93-21Containment89'-11"CBN/A - Terminal End83-MSS-030-93-22Containment93'-8"CBN/A - Arbitrary Intermediate83-MSS-030-93-23Containment66'-3"CBN/A - Arbitrary Intermediate8 MPS3 UFSARMPS3 UFSAR3.6-78Rev. 303-MSS-030-105-316MSVB66'-3"CBN/A - Terminal End83-MSS-030-26-44Turbine56'-9"CB & LS Above Threshold 83-MSS-030-26-45Turbine56'-6"CB & LS Above Threshold 83-MSS-030-26-46Turbine55'-9"CBN/A - Terminal End83-MSS-030-68-47Turbine54'-6"CBN/A - Terminal End83-MSS-030-68-48Turbine54'-6"CB & LS (2)83-MSS-030-68-49Turbine54'-6"CBN/A - Terminal End828" G.E. Piping10Turbine40'-0"CB(2)828" G.E. Piping11Turbine36'-0"CB & LS (2)828" G.E. Piping12Turbine32'-6"CB & LS (2)828" G.E. Piping13Turbine31'-10"CB & LS (2)828" G.E. Piping14Turbine35'-4"CB & LS (2)828" G.E. Piping15Turbine63'-9"CB(2)83-MSS-030-94-21Containment89'-8"CBN/A - Terminal End83-MSS-030-94-22Containment93'-5"CB N/A - Arbitrary Intermediate83-MSS-030-94-23Containment66'-3"CB N/A - Arbitrary Intermediate83-MSS-030-106-318MSVB66'-3"CBN/A - Terminal End83-MSS-030-59-44Turbine66'-3"CBN/A - Arbitrary Intermediate83-MSS-030-59-45Turbine57'-0"CBN/A - Arbitrary Intermediate83-MSS-030-59-46Turbine55'-7"CBN/A - Terminal End8TABLE 3.6-6 POSTULATED BREAKS MAIN STEAM SYSTEM (CONTINUED)
LINE DESIGNATION BREAK #BUILDINGELEVATIONBREAK TYPE (1)TOTAL ADDITIVE STRESSFIGURE 3.6-MPS3 UFSARMPS3 UFSAR3.6-79Rev. 303-MSS-030-69-47Turbine54'-6"CBN/A - Terminal End83-MSS-030-69-48Turbine54'-6"CB & LS (2)83-MSS-030-69-49Turbine54'-6"CBN/A - Terminal End828" G.E. Piping10Turbine40'-0"CB(2)828" G.E. Piping11Turbine36'-0"CB & LS (2)828" G.E. Piping12Turbine32'-6"CB & LS (2)828"-G.E. Piping13Turbine31'-10"CB & LS (2)828" G.E. Piping14Turbine35'-4"CB & LS (2)828" G.E. Piping15Turbine63'-9"CB(2)83-MSS-030-95-21Containment89'-9"CBN/A - Terminal End83-MSS-030-95-22Containment74'-4"CB N/A - Arbitrary Intermediate83-MSS-030-95-23Containment66'-3"CB N/A - Arbitrary Intermediate83-MSS-030-107-319MSVB66'-3"CBN/A - Terminal End83-MSS-030-60-44Turbine65'-8"CBN/A - Arbitrary Intermediate83-MSS-030-60-45Turbine56'-8"CB & LS Above Threshold 83-MSS-030-60-46Turbine55'-6"CBN/A - Terminal End83-MSS-030-70-47Turbine54'-6"CBN/A - Terminal End83-MSS-030-70-48Turbine54'-6"CB & LS (2)83-MSS-030-70-49Turbine54'-6"CBN/A - Terminal End828" G.E. Piping10Turbine40'-0"CB(2)8TABLE 3.6-6 POSTULATED BREAKS MAIN STEAM SYSTEM (CONTINUED)
LINE DESIGNATION BREAK #BUILDINGELEVATIONBREAK TYPE (1)TOTAL ADDITIVE STRESSFIGURE 3.6-MPS3 UFSARMPS3 UFSAR3.6-80Rev. 3028" G.E. Piping11Turbine36'-0"CB & LS (2)828" G.E. Piping12Turbine32'-6"CB & LS (2)828" G.E. Piping13Turbine31'-10"CB & LS (2)828" G.E. Piping14Turbine35'-4"CB & LS (2)828" G.E. Piping15Turbine72'-11"CB & LS (2)828" G.E. Piping16Turbine76'-5"CB(2)83-MSS-003-6-21Containment93'-6"CBN/A - Terminal End83-MSS-003-6-22Containment86'-5"CB N/A - Arbitrary Intermediate83-MSS-003-6-23Containment80'-5"CB N/A - Arbitrary Intermediate83-MSS-003-6-24Containment31'-0"CBN/A - Terminal End83-MSS-003-4-21Containment66'-3"CBN/A - Terminal End83-MSS-003-4-22Containment56'-7"CBAbove Threshold 83-MSS-003-4-23Containment43'-4"CB N/A - Arbitrary Intermediate83-MSS-003-4-24Containment31'-0"CBN/A - Terminal End83-MSS-003-33-21Containment93'-6"CBN/A - Terminal End83-MSS-003-33-22Containment80'-3"CBAbove Threshold 83-MSS-003-33-23Containment71'-4"CBAbove Threshold 83-MSS-003-33-24Containment62'-4"CBAbove Threshold 83-MSS-003-33-25Containment42'-0"CB N/A - Arbitrary Intermediate83-MSS-003-33-26Containment31'0"CBN/A - Terminal End8TABLE 3.6-6 POSTULATED BREAKS MAIN STEAM SYSTEM (CONTINUED)
LINE DESIGNATION BREAK #BUILDINGELEVATIONBREAK TYPE (1)TOTAL ADDITIVE STRESSFIGURE 3.6-MPS3 UFSARMPS3 UFSAR3.6-81Rev. 30NOTES:(1) Circumferential Pipe Break (CB) and Longi tudinal Pipe Split (LS) are defined in FSAR Section 3.6.2.1.3
.(2) Pipe breaks on the Main Steam System downstream of the Ma in 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 UFSARMPS3 UFSAR3.6-82Rev. 30TABLE 3.6-7 PIPE WHIP EFFECTS MAIN STEAM SYSTEM LINE DESIGNATIONBREAK # (1)BLOWDOWN SOURCEESSENTIAL PIPE WHIP TARGETPRIMARY PROTECTION REQUIREMENT3-MSS-030-92-213-MSS-042-63-4Pol ar Crane 3MHR*CRN1 Support System 3MSS-PRR7LA3-MSS-030-92-223RCS*SG1ANone (2)3MSS-PRR5LA3-MSS-030-92-223-MSS-042-63-4None (2) 3MSS-PRR2LA3-MSS-030-92-233RCS*SG1ANone (2)3MSS-PRR3LA3-MSS-030-92-233-MSS-042-63-4None (2) 3MSS-PRR1LA3-MSS-030-104-3173RCS*SG1ANone None3-MSS-030-104-3173-MSS-042-63-4Control Building Boundary Wall3MSS-PRR4A3-MSS-030-25-443RCS*SG1ANone (2)3MSS-PRR2A & 3MSS-PRR3A3-MSS-030-25-443-MSS-042-63-4None (2)3MSS-PRR4A & 3MSS-PRR5A3-MSS-030-25-453RCS*SG1ALimit Stress to Break Exclusion Zone3MSS-PRR2A & 3MSS-PRR3A3-MSS-030-25-453-MSS-042-63-4Control Building Boundary Wall3MSS-PRR4A & 3MSS-PRR5A3-MSS-030-25-4Split(3)Limit Stress to Break Exclusion Zone3MSS-PRR2A & 3MSS-PRR3A3-MSS-030-25-463RCS*SG1AControl Building Boundary Wall3MSS-PRR5A3-MSS-030-67-473-MSS-042-63-4Control Building Boundary Wall3MSS-PRR7A & 3MSS-PRR13A3-MSS-030-67-483-MSS-042-63-4Control Building Boundary Wall3MSS-PRR7A & 3MSS-PRR13A3-MSS-030-67-493-MSS-042-63-4Control Building Boundary Wall3MSS-PRR13A3-MSS-030-67-4103-MSS-042-63-4Control Building Boundary Wall3MSS-PRR13A3-MSS-030-67-4Splits3-MSS-042-63-4None None MPS3 UFSARMPS3 UFSAR3.6-83Rev. 3028" G.E. Piping113-MSS-042-63-4None None28" G.E. Piping123-MSS-042-63-4None None28" G.E. Piping133-MSS-042-63-4None None28" G.E. Piping143-MSS-042-63-4None None28" G.E. Piping153-MSS-042-63-4None None28" G.E. Piping163-MSS-042-63-4None None28" G.E. PipingSplits3-MSS-042-63-4None None3-MSS-030-93-213-MSS-042-63-4Pol ar Crane 3MHR*CRN1 Support System 3MSS-PRR7SB3-MSS-030-93-223RCS*SG1BNone - (2)None3-MSS-030-93-223-MSS-042-63-4None - (2)3MSS-PRR6SB3-MSS-030-93-233RCS*SG1BNone - (2)3MSS-PRR5SB3-MSS-030-93-233-MSS-042-63-4None - (2)None3-MSS-030-105-3163RCS*SG1BNone None3-MSS-030-105-3163-MSS-042-63-4Control Building Boundary Wall3MSS-PRR4B3-MSS-030-26-443RCS*SG1BLimit Stress to Break Exclusion Zone3MSS-PRR2B & 3MSS-PRR3B3-MSS-030-26-443-MSS-042-63-4Control Building Boundary Wall3MSS-PRR4B3-MSS-030-26-4Split(3)Limit Stress to Break Exclusion Zone3MSS-PRR2B & 3MSS-PRR3B3-MSS-030-26-453RCS*SG1BLimit Stress to Break Exclusion Zone3MSS-PRR2B & 3MSS-PRR3BTABLE 3.6-7 PIPE WHIP EFFECTS MAIN STEAM SYSTEM (CONTINUED)LINE DESIGNATIONBREAK # (1)BLOWDOWN SOURCEESSENTIAL PIPE WHIP TARGETPRIMARY PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-84Rev. 303-MSS-030-26-453-MSS-042-63-4Control Building Boundary Wall3MSS-PRR5B3-MSS-030-26-45(3)Limit Stress to Break Exclusion Zone3MSS-PRR2B & 3MSS-PRR3B3-MSS-030-26-463RCS*SG1BControl Building Boundary Wall3MSS-PRR5B3-MSS-030-68-473-MSS-042-63-4Control Building Boundary Wall3MSS-PRR7B & 3MSS-PRR13B3-MSS-030-68-483-MSS-042-63-4Control Building Boundary Wall3MSS-PRR7B & 3MSS-PRR13B3-MSS-030-68-493-MSS-042-63-4Control Building Boundary Wall3MSS-PRR7B & 3MSS-PRR13B3-MSS-030-68-4103-MSS-042-63-4Control Building Boundary Wall3MSS-PRR13B3-MSS-030-68-4Splits3-MSS-042-63-4NoneNone28" G.E. Piping113-MSS-042-63-4NoneNone 28" G.E. Piping123-MSS-042-63-4NoneNone 28" G.E. Piping133-MSS-042-63-4NoneNone 28" G.E. Piping143-MSS-042-63-4NoneNone 28" G.E. Piping153-MSS-042-63-4NoneNone 28" G.E. PipingSplits3-MSS-042-63-4NoneNone3-MSS-030-094-213-MSS-042-63-4Pol ar Crane 3MHR*CRN1 Support System 3MSS-PRR7SC3-MSS-030-094-223RCS*SG1CNone (2)None3-MSS-030-094-223-MSS-042-63-4None (2)3MSS-PRR6SCTABLE 3.6-7 PIPE WHIP EFFECTS MAIN STEAM SYSTEM (CONTINUED)LINE DESIGNATIONBREAK # (1)BLOWDOWN SOURCEESSENTIAL PIPE WHIP TARGETPRIMARY PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-85Rev. 303-MSS-030-094-233RCS*SG1CNone (2)3MSS-PRR5SC3-MSS-030-094-233-MSS-042-63-4None (2)None3-MSS-030-106-3183RCS*SG1CNone None3-MSS-030-106-3183-MSS-042-63-4Control Building Boundary Wall3MSS-PRR4C3-MSS-030-59-443RCS*SG1CNone (2)3MSS-PRR1C & 3MSS-PRR3C3-MSS-030-59-443-MSS-042-63-4None (2) 3MSS-PRR4C3-MSS-030-59-453RCS*SG1CNone (2)3MSS-PRR1C & 3MSS-PRR3C3-MSS-030-59-453-MSS-042-63-4None (2) 3MSS-PRR5C3-MSS-030-59-463RCS*SG1CControl Building Boundary Wall 3MSS-PRR5C3-MSS-030-69-473-MSS-042-63-4Control Building Boundary Wall3MSS-PRR7C & 3MSS-PRR13C3-MSS-030-69-483-MSS-042-63-4Control Building Boundary Wall3MSS-PRR7C & 3MSS-PRR13C3-MSS-030-69-493-MSS-042-63-4Control Building Boundary Wall3MSS-PRR7C & 3MSS-PRR13C3-MSS-030-69-4103-MSS-042-63-4Control Building Boundary Wall 3MSS-PRR13C3-MSS-030-69-4Splits3-MSS-042-63-4None None28" G.E. Piping113-MSS-042-63-4None None28" G.E. Piping123-MSS-042-63-4None None28" G.E. Piping133-MSS-042-63-4None None28" G.E. Piping143-MSS-042-63-4None None28" G.E. Piping153-MSS-042-63-4None NoneTABLE 3.6-7 PIPE WHIP EFFECTS MAIN STEAM SYSTEM (CONTINUED)LINE DESIGNATIONBREAK # (1)BLOWDOWN SOURCEESSENTIAL PIPE WHIP TARGETPRIMARY PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-86Rev. 3028" G.E. PipingSplits3-MSS-042-63-4None None3-MSS-030-95-213-MSS-042-63-4Pol ar Crane 3MHR*CRN1 Support System 3MSS-PRR7LD3-MSS-030-95-223RCS*SG1DNone (2) 3MSS-PRR5LD3-MSS-030-95-223-MSS-042-63-4None (2) 3MSS-PRR2LD3-MSS-030-95-233RCS*SG1DNone (2) 3MSS-PRR3LD3-MSS-030-95-233-MSS-042-63-4None (2) 3MSS-PRR1LD3-MSS-030-107-3193RCS*SG1DNone None3-MSS-030-107-3193-MSS-042-63-4Control Building Boundary Wall 3MSS-PRR4D3-MSS-030-60-443RCS*SG1DNone (2) 3MSS-PRR1D & 3MSS-PRR3D3-MSS-030-60-443-MSS-042-63-4None (2) 3MSS-PRR4D3-MSS-030-60-453RCS*SG1DLimit Stress to Break Exclusion Zone3MSS-PRR1D & 3MSS-PRR3D3-MSS-030-60-453-MSS-042-63-4Control Building Boundary Wall 3MSS-PRR5D3-MSS-030-60-4Split(3)Limit Stress to Break Exclusion Zone3MSS-PRR1D & 3MSS-PRR3D3-MSS-030-60-463RCS*SG1DControl Building Boundary Wall 3MSS-PRR5D3-MSS-030-70-473-MSS-042-63-4Control Building Boundary Wall3MSS-PRR7D & 3MSS-PRR13D3-MSS-030-70-483-MSS-042-63-4Control Building Boundary Wall3MSS-PRR7D & 3MSS-PRR13D3-MSS-030-70-493-MSS-042-63-4Control Building Boundary Wall3MSS-PRR7D & 3MSS-PRR13D3-MSS-030-70-4103-MSS-042-63-4Control Building Boundary Wall 3MSS-PRR13DTABLE 3.6-7 PIPE WHIP EFFECTS MAIN STEAM SYSTEM (CONTINUED)LINE DESIGNATIONBREAK # (1)BLOWDOWN SOURCEESSENTIAL PIPE WHIP TARGETPRIMARY PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-87Rev. 303-MSS-030-70-4Splits3-MSS-042-63-4None None28" G.E. Piping113-MSS-042-63-4None None28" G.E. Piping123-MSS-042-63-4None None28" G.E. Piping133-MSS-042-63-4None None28" G.E. Piping143-MSS-042-63-4None None28" G.E. Piping153-MSS-042-63-4None None28" G.E. Piping163-MSS-042-63-4None None28" G.E. PipingSplits3-MSS-042-63-4None None 3-MSS-003-6-21None (4)None None 3-MSS-003-6-223-MSS-030-92-2None (2)None 3-MSS-003-6-22None (4)None None 3-MSS-003-6-233-MSS-030-92-2None (2)None 3-MSS-003-6-23None (4)None None 3-MSS-003-6-243-MSS-030-92-2None None 3-MSS-003-6-24None (4)None None 3-MSS-003-4-21None (4)None None 3-MSS-003-4-223-MSS-030-93-2None None 3-MSS-003-4-22None (4)None None 3-MSS-003-4-233-MSS-030-93-2None (2)NoneTABLE 3.6-7 PIPE WHIP EFFECTS MAIN STEAM SYSTEM (CONTINUED)LINE DESIGNATIONBREAK # (1)BLOWDOWN SOURCEESSENTIAL PIPE WHIP TARGETPRIMARY PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-88Rev. 30NOTES:(1) Repetition of Break Numbers is used to identify separate fl uid reservoirs which provide a c onstant pressure source to maint ain system pressure subsequent to the postula ted pipe break. These reservoirs maintain system blowdown during the transient event a s well as in the steady state.(2) USNRC Generic Letter 87-11 eliminates the need to evaluate th e pipe whip ef fects subsequent to Arbitrary Intermediate Break s (AIB's). Therefore pipe whip eff ects from AIB's are not evaluated.
3-MSS-003-4-23None (4)None None 3-MSS-003-4-243-MSS-030-93-2None None 3-MSS-003-4-24None (4)None None3-MSS-003-33-21None (4)None None3-MSS-003-33-223-MSS-030-95-2None None3-MSS-003-33-22None (4)None None3-MSS-003-33-233-MSS-030-95-2None None3-MSS-003-33-23None (4)None None3-MSS-003-33-243-MSS-030-95-2None None3-MSS-003-33-24None (4)None None3-MSS-003-33-253-MSS-030-95-2None (2)None3-MSS-003-33-25None (4)None None3-MSS-003-33-263-MSS-030-95-2None None3-MSS-003-33-26None (4)None NoneTABLE 3.6-7 PIPE WHIP EFFECTS MAIN STEAM SYSTEM (CONTINUED)LINE DESIGNATIONBREAK # (1)BLOWDOWN SOURCEESSENTIAL PIPE WHIP TARGETPRIMARY PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-89Rev. 30(3) The Longitudinal Split 5 on 3-MSS-030-25-4 results in blowdown from both the Steam Genera tor 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 les s 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 UFSARMPS3 UFSAR3.6-90Rev. 30TABLE 3.6-8 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM LINE DESIGNATION BREAK # (1)ESSENTIAL JET IMPINGEMENT TARGETDISTANCE TO TARGETJET INTENSITY AT THE TARGET JET LOAD ON THE TARGET PROTECTION REQUIREMENT3-MSS-030-92-21 None N/AN/A N/ANone3-MSS-030-92-22Crane Wall4 Feet340 psi1033 KipsNone (2)3-MSS-030-92-223RCS*SG1C13 Feet65 psi564 KipsNone (2)3-MSS-030-92-23 None N/AN/A N/ANone (3)3-MSS-030-92-23 None N/AN/A N/ANone (3)3-MSS-030-104-317None N/AN/A N/ANone3-MSS-030-104-317None N/AN/A N/ANone3-MSS-030-25-44 None N/AN/A N/ANone3-MSS-030-25-44 None N/AN/A N/ANone3-MSS-030-25-45 None N/AN/A N/ANone (4)3-MSS-030-25-45 None N/AN/A N/ANone (4)3-MSS-030-25-4SplitNone N/AN/A N/ANone (4)3-MSS-030-25-46Control Building (5)None (6)3-MSS-030-67-47Control Building (5)None (6)3-MSS-030-67-48Control Building (5)None (6)3-MSS-030-67-4SplitNone N/AN/A N/ANone (4)3-MSS-030-67-49 None N/AN/A N/ANone (4)
MPS3 UFSARMPS3 UFSAR3.6-91Rev. 3028" G.E. Piping10 None N/AN/A N/ANone (4)28" G.E. Piping11 None N/AN/A N/ANone (4)28" G.E. PipingSplitNone N/AN/A N/ANone (4)28" G.E. Piping12 None N/AN/A N/ANone (4)28" G.E. PipingSplitNone N/AN/A N/ANone (4)28" G.E. Piping13 None N/AN/A N/ANone (4)28" G.E. PipingSplitNone N/AN/A N/ANone (4)28" G.E. Piping14 None N/AN/A N/ANone (4)28" G.E. PipingSplitNone N/AN/A N/ANone (4)28" G.E. Piping15 None N/AN/A N/ANone (4)28" G.E. PipingSplitNone N/AN/A N/ANone (4)28" G.E. Piping16 None N/AN/A N/ANone (4)3-MSS-030-93-21 None N/AN/A N/ANone3-MSS-030-93-21 None N/AN/A N/ANone3-MSS-030-93-22 & 3None N/AN/A N/ANone (3)3-MSS-030-105-316None N/AN/A N/ANone3-MSS-030-105-316None N/AN/A N/ANone3-MSS-030-26-44 None N/AN/A N/ANone (4)TABLE 3.6-8 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM (CONTINUED)
LINE DESIGNATION BREAK # (1)ESSENTIAL JET IMPINGEMENT TARGETDISTANCE TO TARGETJET INTENSITY AT THE TARGET JET LOAD ON THE TARGET PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-92Rev. 303-MSS-030-26-44 None N/AN/A N/ANone (4)3-MSS-030-26-4SplitNone N/AN/A N/ANone (4)3-MSS-030-26-45 None N/AN/A N/ANone (4)3-MSS-030-26-45 None N/AN/A N/ANone (4)3-MSS-030-26-4SplitNone N/AN/A N/ANone (4)3-MSS-030-26-46Control Building (5)None (6)3-MSS-030-68-47Control Building (5)None (6)3-MSS-030-68-48Control Building (5)None (6)3-MSS-030-68-4SplitNone N/AN/A N/ANone (4)3-MSS-030-68-49 None N/AN/A N/ANone (4)28" G.E. Piping10 None N/AN/A N/ANone (4)28" G.E. Piping11 None N/AN/A N/ANone (4)28" G.E. PipingSplitNone N/AN/A N/ANone (4)28" G.E. Piping12 None N/AN/A N/ANone (4)28" G.E. PipingSplitNone N/AN/A N/ANone (4)28" G.E. Piping13 None N/AN/A N/ANone (4)28" G.E. PipingSplitNone N/AN/A N/ANone (4)28" G.E. Piping14 None N/AN/A N/ANone (4)TABLE 3.6-8 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM (CONTINUED)
LINE DESIGNATION BREAK # (1)ESSENTIAL JET IMPINGEMENT TARGETDISTANCE TO TARGETJET INTENSITY AT THE TARGET JET LOAD ON THE TARGET PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-93Rev. 3028" G.E. PipingSplitNone N/AN/A N/ANone (4)28" G.E. Piping15 None N/AN/A N/ANone (4)3-MSS-030-94-21 None N/AN/A N/ANone3-MSS-030-94-22 & 3None N/AN/A N/ANone (3)3-MSS-030-106-318None N/AN/A N/ANone3-MSS-030-106-318None N/AN/A N/ANone3-MSS-030-59-44 None N/AN/A N/ANone3-MSS-030-59-44 None N/AN/A N/ANone3-MSS-030-59-45 None N/AN/A N/ANone3-MSS-030-59-45 None N/AN/A N/ANone3-MSS-030-59-46Control Building (5)None (6)3-MSS-030-69-47Control Building (5)None (6)3-MSS-030-69-48Control Building (5)None (6)3-MSS-030-69-4SplitNone N/AN/A N/ANone (4)3-MSS-030-69-49Control Building (5)None (6)28" G.E. Piping10 None N/AN/A N/ANone (4)28" G.E. Piping11 None N/AN/A N/ANone (4)28" G.E. PipingSplitNone N/AN/A N/ANone (4)TABLE 3.6-8 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM (CONTINUED)
LINE DESIGNATION BREAK # (1)ESSENTIAL JET IMPINGEMENT TARGETDISTANCE TO TARGETJET INTENSITY AT THE TARGET JET LOAD ON THE TARGET PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-94Rev. 3028" G.E. Piping12 None N/AN/A N/ANone (4)28" G.E. PipingSplitNone N/AN/A N/ANone (4)28" G.E. Piping13 None N/AN/A N/ANone (4)28" G.E. PipingSplitNone N/AN/A N/ANone (4)28" G.E. Piping14 None N/AN/A N/ANone (4)28" G.E. PipingSplitNone N/AN/A N/ANone (4)28" G.E. Piping15 None N/AN/A N/ANone (4)3-MSS-030-95-21 None N/AN/A N/ANone3-MSS-030-95-21 None N/AN/A N/ANone3-MSS-030-95-22Crane Wall4 Feet340 psi1033 KipsNone (2)3-MSS-030-95-223RCS*SG1C13 Feet65 psi564 KipsNone (2)3-MSS-030-95-23 None N/AN/A N/ANone (3)3-MSS-030-107-319None N/AN/A N/ANone3-MSS-030-107-319None N/AN/A N/ANone3-MSS-030-60-44 None N/AN/A N/ANone (4)3-MSS-030-60-44 None N/AN/A N/ANone (4)3-MSS-030-60-45 None N/AN/A N/ANone (4)3-MSS-030-60-45 None N/AN/A N/ANone (4)TABLE 3.6-8 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM (CONTINUED)
LINE DESIGNATION BREAK # (1)ESSENTIAL JET IMPINGEMENT TARGETDISTANCE TO TARGETJET INTENSITY AT THE TARGET JET LOAD ON THE TARGET PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-95Rev. 303-MSS-030-60-4SplitNone N/AN/A N/ANone (4)3-MSS-030-60-46Control Building10 Feet86 psi843 KipsNone (6)3-MSS-030-70-47Control Building10 Feet86 psi843 KipsNone (6)3-MSS-030-70-48Control Building10 Feet86 psi843 KipsNone (6)3-MSS-030-70-4SplitNone N/AN/A N/ANone (4)3-MSS-030-70-49Control Building10 Feet86 psi843 KipsNone (6)28" G.E. Piping10Control Building10 Feet86 psi843 KipsNone (6)28" G.E. Piping11 None N/AN/A N/ANone (4)28" G.E. PipingSplitNone N/AN/A N/ANone (4)28" G.E. Piping12 None N/AN/A N/ANone (4)28" G.E. PipingSplitNone N/AN/A N/ANone (4)28" G.E. Piping13 None N/AN/A N/ANone (4)28" G.E. PipingSplitNone N/AN/A N/ANone (4)28" G.E. Piping14 None N/AN/A N/ANone (4)28" G.E. PipingSplitNone N/AN/A N/ANone (4)28" G.E. Piping15 None N/AN/A N/ANone (4)28" G.E. PipingSplitNone N/AN/A N/ANone (4)28" G.E. Piping16 None N/AN/A N/ANone (4)TABLE 3.6-8 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM (CONTINUED)
LINE DESIGNATION BREAK # (1)ESSENTIAL JET IMPINGEMENT TARGETDISTANCE TO TARGETJET INTENSITY AT THE TARGET JET LOAD ON THE TARGET PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-96Rev. 30NOTES:(1) Repetition of Break Numb ers is used to identify fluid reservoirs which maintain system pressure subsequent of the postulate d pipe break. The reservoirs sustain blowdown during the transient event and in the steady state.3-MSS-003-6-21Crane Wall (7)None3-MSS-003-6-22 & 3Crane Wall (7)None3-MSS-003-6-24Crane Wall (7)None3-MSS-003-6-24Crane Wall (7)None3-MSS-003-4-21Crane Wall (7)None3-MSS-003-4-22 & 3None N/AN/A N/ANone3-MSS-003-4-24 None N/AN/A N/ANone3-MSS-003-4-24 None N/AN/A N/ANone3-MSS-003-33-21Crane Wall (7)None3-MSS-003-33-22 None N/AN/A N/ANone3-MSS-003-33-23 None N/AN/A N/ANone3-MSS-003-33-24 None N/AN/A N/ANone3-MSS-003-33-25 None N/AN/A N/ANone3-MSS-003-33-26 None N/AN/A N/ANoneTABLE 3.6-8 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM (CONTINUED)
LINE DESIGNATION BREAK # (1)ESSENTIAL JET IMPINGEMENT TARGETDISTANCE TO TARGETJET INTENSITY AT THE TARGET JET LOAD ON THE TARGET PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-97Rev. 30(2) Intermediate Break Number 2 is postulated at a location proximate to an Integral Welded Attachment (IWA) on Main Steam Line s 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 re tain 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 th e need to evaluate jet impingement effects subsequent to Arbitrary Intermediate Breaks (AIB's). Therefore, jet impingement effects on the Main Steam Syst em are not evaluated for AIB's.(4) No essential systems are located in th e Turbine Building. Jet impingement on an adjacent Main Steam pipe does not cause pre ssure boundary failure for the reasons provided in FSAR Section 3.6.2.1
.(5) Fluid jet impingement subse quent to this postulated break on this tar get is enveloped by Break Number 6 on 3-MSS-030-60-4 a nd 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 fl uid jet impingement on concrete barriers described in NERM-069 Re vision 1 Hazards Review Progra m Summary for Millstone Unit 3 Section 4.4
.(7) Fluid jet impingement subseque nt to this postulated break on this target is enveloped by postulate d break number 2 on eithe r of the Main Steam System lines 3-MSS-030-92-2 or 3-MSS-030-95-2 loading the same target.
MPS3 UFSARMPS3 UFSAR3.6-98Rev. 30TABLE 3.6-9 POSTULATED BREAKS MAIN FEEDWATER SYSTEM LINE DESIGNATIONBREAK #BUILDINGELEVATION BREAK TYPE (1)TOTAL ADDITIVE STRESSFIGURE 3.6-3-FWS-020-18-21Containment64'-7"CBN/A - Terminal End10 3-FWS-020-18-22Containment64'-7"CB & LSAbove Threshold10 3-FWS-020-18-23Containment61'-0"CBN/A - Arbitrary Intermediate 10 3-FWS-018-16-39MSVB44'-3"CBN/A - Terminal End10
 
3-FWS-020-12-4 4 (2)Turbine51'-0"CBN/A - Arbitrary Intermediate 10 3-FWS-020-12-4 5 (2)Turbine51'-0"CBN/A - Arbitrary Intermediate 10 3-FWS-016-74-21Containment64'-7"CBN/A - Terminal End10 3-FWS-016-74-22Containment64'-7"CB & LSAbove Threshold10 3-FWS-016-74-23Containment64'-7"CB & LSAbove Threshold10 3-FWS-018-20-39MSVB44'-3"CBN/A - Terminal End10 3-FWS-016-73-21Containment64'-7"CBN/A - Terminal End10 3-FWS-016-73-22Containment64'-7"CB & LSAbove Threshold10 3-FWS-016-73-23Containment64'-7"CB & LSAbove Threshold10 3-FWS-018-24-39MSVB44'-3"CBN/A - Terminal End10 3-FWS-020-30-21Containment64'-7"CBN/A - Terminal End10 3-FWS-020-30-22Containment64'-7"CB & LSAbove Threshold10 3-FWS-020-30-23Containment60'-10" CB N/A - Arbitrary Intermediate 10 3-FWS-018-28-39MSVB44'-3"CBN/A - Terminal End10 MPS3 UFSARMPS3 UFSAR3.6-99Rev. 30NOTES:(1) Circumferential Pipe Break (CB) and Longi tudinal Pipe Split (LS) are defined in FSAR Section 3.6.2.1.3
.(2) The Main Feedwater piping in the Turbin e 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 AS ME Class 2 and
: 3. The criteria specified in FSAR Secti on 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-FWS-020-15-4 6 (2)Turbine51'-0"CBN/A - Arbitrary Intermediate 10 3-FWS-020-15-4 7 (2)Turbine51'-0"CBN/A - Arbitrary Intermediate 10 3-FWS-036-11-4 8 (2)Turbine51'-0"CBN/A Terminal End10 TABLE 3.6-9 POSTULATED BREAKS MAIN FEEDWATER SYSTEM (CONTINUED)
LINE DESIGNATIONBREAK #BUILDINGELEVATION BREAK TYPE (1)TOTAL ADDITIVE STRESSFIGURE 3.6-MPS3 UFSARMPS3 UFSAR3.6-100Rev. 30TABLE 3.6-10 PIPE WHIP EFFECTS - MAIN FEEDWATER SYSTEM LINE DESIGNATIONBREAK#
(1)BLOWDOWN SOURCEESSENTIAL PIPE WHIP TARGET PROTECTION REQUIREMENT3-FWS-020-18-213-FWS-036-11-4Steam Generator Cubicle Wall 3FWS-PRR12LA &
3FWS-PRR13LA3-FWS-020-18-223RCS*SG1ANoneNone3-FWS-020-18-223-FWS-036-11-4Operating Floor at Elevation 51'-4"3FWS-PRR5LA (2)3-FWS-020-18-2Split 3RCS*SG1ANone None3-FWS-020-18-23 3RCS*SG1A None (3)3FWS-PRR5LA3-FWS-020-18-233-FWS-036-11-4 None (3)3FWS-PRR3LA3-FWS-018-16-393RCS*SG1ANoneNone3-FWS-018-16-393-FWS-036-11-4NoneNone3-FWS-020-12-443RCS*SG1A None (3)None3-FWS-020-12-443-FWS-036-11-4 None (3)None3-FWS-020-12-453RCS*SG1A None (3)None3-FWS-020-12-453-FWS-036-11-4 None (3)None3FWS-016-74-213-FWS-036-11-4Steam Generator Cubicle Wall 3FWS-PRR8SB3-FWS-016-74-223RCS*SG1B3RCS*SG1B3FWS-PRR5SB3-FWS-016-74-223-FWS-036-11-4 None 3FWS-PRR8SB MPS3 UFSARMPS3 UFSAR3.6-101Rev. 303-FWS-016-74-2Split 3RCS*SG1BNone None3-FWS-016-74-23 3-FWS-036-11-4 None 3FWS-PRR8SB3-FWS-016-74-23 3RCS*SG1BNone 3-FWS-PRR5SB3-FWS-016-74-2Split 3RCS*SG1BNone None3-FWS-018-20-39 3RCS*SG1BNone None3-FWS-018-20-393-FWS-036-11-4 None None3-FWS-016-73-213-FWS-036-11-4Steam Generator Cubicle Wall 3FWS-PRR8SC3-FWS-016-73-22 3RCS*SG1C3RCS*SG1C 3FWS-PRR5SC3-FWS-016-73-22 3-FWS-036-11-4 None 3FWS-PRR8SC3-FWS-016-73-2Split 3RCS*SG1CNone None3-FWS-016-73-23 3-FWS-036-11-4 None 3FWS-PRR8SC3-FWS-016-73-23 3RCS*SG1CNone 3FWS-PRR5SC3-FWS-016-73-2Split 3RCS*SG1CNone None3-FWS-018-24-39 3RCS*SG1CNone None3-FWS-018-24-393-FWS-036-11-4 None None3-FWS-020-30-213-FWS-036-11-4Steam Generator Cubicle Wall 3FWS-PRR12LD &
3FWS-PRR13LD3-FWS-020-30-22 3RCS*SG1DNone NoneTABLE 3.6-10 PIPE WHIP EFFECTS - MA IN FEEDWATER SYST EM (CONTINUED)
LINE DESIGNATIONBREAK#
(1)BLOWDOWN SOURCEESSENTIAL PIPE WHIP TARGET PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-102Rev. 30NOTES:(1) Repetition of Break Numbers is used to identify separate fl uid reservoirs which provide a c onstant pressure source to maint ain system pressure subsequent to the postula ted pipe break. These reservoirs maintain system blowdown during the transient event a s well as in the steady state.3-FWS-020-30-22 3-FWS-036-11-4Operating Floor at Elevation 51'-4"3FWS-PRR5LD (4)3-FWS-020-30-2Split 3RCS*SG1DNone None3-FWS-020-30-233-FWS-036-11-4 None (3)None3-FWS-020-30-233RCS*SG1D None (3)None3-FWS-018-28-393RCS*SG1DNoneNone3-FWS-018-28-393-FWS-036-11-4NoneNone3-FWS-020-15-463-FWS-036-11-4 None (3)None3-FWS-020-15-463RCS*SG1D None (3)None3-FWS-020-15-473-FWS-036-11-4 None (3)None3-FWS-020-15-473RCS*SG1D None (3)None3-FWS-036-11-483RCS*SG1A & 1BNoneNone3-FWS-036-11-483RCS*SG1C & 1DNoneNoneTABLE 3.6-10 PIPE WHIP EFFECTS - MA IN FEEDWATER SYST EM (CONTINUED)
LINE DESIGNATIONBREAK#
(1)BLOWDOWN SOURCEESSENTIAL PIPE WHIP TARGET PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-103Rev. 30(2) The Main Feedwater piping in the Turbin e 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 AS ME Class 2 and
: 3. The criteria specified in FSAR Secti on 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 el iminates the requirement to eval uate pipe whip ef fects subseque nt to Arbitrary Intermediate Br eaks (AIB's). Therefore, pipe whip eff ects 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 Circ umferential Break (CB) and a Longitudinal Split (LS) are postulated but not concurrently. Th e Circumferential Break (CB) re sults in an impact of the M ain Feedwater Line 3-FWS-020-30-2 to the Operating Floor Slab at Elevation 51'-4". Unrestra ined impact would result in catastrophic failure of the Operating Floor and result in secondary missile ej ection into the Reactor Coolant System cubicles. This unrestra ined impact to the Operating Floor is prevented by the rupture restraint.
MPS3 UFSARMPS3 UFSAR3.6-104Rev. 30TABLE 3.6-11 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM LINE DESIGNATION BREAK # (1)ESSENTIAL JET IMPINGEMENT TARGETDISTANCE TO TARGETJET INTENSITY AT THE TARGETJET LOAD ON THE TARGET PROTECTION REQUIREMENT3-FWS-020-18-21None N/AN/A N/ANone3-FWS-020-18-22None N/AN/A N/ANone3-FWS-020-18-22None N/AN/A N/ANone3-FWS-020-18-2SplitNone N/AN/A N/ANone3-FWS-020-18-23 None (2)N/AN/A N/ANone3-FWS-020-18-23 None (2)N/AN/A N/ANone3-FWS-018-16-39None N/AN/A N/ANone3-FWS-018-16-39None N/AN/A N/ANone3-FWS-020-12-44 None (2)N/AN/A N/ANone3-FWS-020-12-44 None (2)N/AN/A N/ANone3-FWS-020-12-45 None (2)N/AN/A N/ANone3-FWS-020-12-45 None (2)N/AN/A N/ANone3-FWS-016-74-21None N/AN/A N/ANone3-FWS-016-74-22None N/AN/A N/ANone3-FWS-016-74-22None N/AN/A N/ANone3-FWS-016-74-2SplitNone N/AN/A N/ANone3-FWS-016-74-23None N/AN/A N/ANone3-FWS-016-74-23None N/AN/A N/ANone3-FWS-016-74-2SplitNone N/AN/A N/ANone MPS3 UFSARMPS3 UFSAR3.6-105Rev. 303-FWS-018-20-39None N/AN/A N/ANone3-FWS-018-20-39None N/AN/A N/ANone3-FWS-016-73-21None N/AN/A N/ANone3-FWS-016-73-22None N/AN/A N/ANone3-FWS-016-73-22None N/AN/A N/ANone3-FWS-016-73-2SplitNone N/AN/A N/ANone3-FWS-016-73-23None N/AN/A N/ANone3-FWS-016-73-23None N/AN/A N/ANone3-FWS-016-73-2SplitNone N/AN/A N/ANone3-FWS-018-24-39None N/AN/A N/ANone3-FWS-018-24-39None N/AN/A N/ANone3-FWS-020-30-21None N/AN/A N/ANone3-FWS-020-30-22None N/AN/A N/ANone3-FWS-020-30-22None N/AN/A N/ANone3-FWS-020-30-2SplitNone N/AN/A N/ANone3-FWS-020-30-23 None (2)N/AN/A N/ANone3-FWS-020-30-23 None (2)N/AN/A N/ANone3-FWS-018-28-39None N/AN/A N/ANone3-FWS-018-28-39None N/AN/A N/ANone3-FWS-020-15-46 None (2)N/AN/A N/ANoneTABLE 3.6-11 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM (CONTINUED)LINE DESIGNATION BREAK # (1)ESSENTIAL JET IMPINGEMENT TARGETDISTANCE TO TARGETJET INTENSITY AT THE TARGETJET LOAD ON THE TARGET PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-106Rev. 30NOTES:(1) Repetition of Break Numbers is used to identify separate fl uid reservoirs which provide a c onstant pressure source to maint ain system pressure subsequent to the postula ted pipe break. These reservoirs maintain system blowdown during the transient event a s well as in the steady state.(2)  USNRC Generic Letter 87-11 eliminates the requirement to evaluate the jet impi ngement ef fects subsequent to Arbitrary Intermediate Break (AIB's). Therefore jet impingement effects subse quent to AIB's are not evaluated.3-FWS-020-15-46 None (2)N/AN/A N/ANone3-FWS-020-15-47 None (2)N/AN/A N/ANone3-FWS-020-15-47 None (2)N/AN/A N/ANone3-FWS-036-11-48None N/AN/A N/ANone3-FWS-036-11-48None N/AN/A N/ANoneTABLE 3.6-11 JET IMPINGEMENT EFFECTS - MAIN STEAM SYSTEM (CONTINUED)LINE DESIGNATION BREAK # (1)ESSENTIAL JET IMPINGEMENT TARGETDISTANCE TO TARGETJET INTENSITY AT THE TARGETJET LOAD ON THE TARGET PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-107Rev. 30TABLE 3.6-12 POSTULATED BREAKS - RE ACTOR COOLANT SYSTEM - LOOP STOP VALVE BY PASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING. LETDOWN LINE, AND NORMAL CHARGING LINE DESIGNATIONBREAK #BUILDINGELEVATIONBREAK TYPE (1)THRESHOLD STRESSUSAGE FACTORFIGUREEQN 10EQN 12 OR 133-RCS-008-24-113Containment23'-0"CB & LSAbove Threshold3.6-12 3-RCS-008-25-114Containment31'-6"CB & LSAbove Threshold3.6-12 3-RCS-150-27-113Containment23'-0"CBTerminal End3.6-12 3-RCS-150-27-115Containment23'-0"CBN/A - Arbitrary Intermediate3.6-12 3-RCS-150-27-116Containment22'-2"CBN/A - Arbitrary Intermediate3.6-12 3-RCS-150-27-117Containment17'-6"CBTerminal End3.6-12 3-RCS-008-29-113Containment23'-0"CB & LSAbove Threshold3.6-12 3-RCS-008-30-114Containment31'-6"CB & LSAbove Threshold3.6-12 3-RCS-150-32-113Containment23'-0"CBTerminal End3.6-12 3-RCS-150-32-115Containment23'-0"CBN/A - Arbitrary Intermediate3.6-12 3-RCS-150-32-116Containment22'-2"CBN/A - Arbitrary Intermediate3.6-12 3-RCS-150-32-117Containment17'-6"CBTerminal End3.6-12 3-RCS-008-34-113Containment23'-0"CB & LSAbove Threshold3.6-12 3-RCS-008-35-114Containment31'-6"CB & LSAbove Threshold3.6-12 3-RCS-150-37-113Containment23'-0"CBTerminal End3.6-12 MPS3 UFSARMPS3 UFSAR3.6-108Rev. 303-RCS-150-37-115Containment23'-0"CB N/A - Arbitrary Intermediate3.6-12 3-RCS-150-37-716Containment22'-2"CB N/A - Arbitrary Intermediate3.6-12 3-RCS-150-37-117Containment17'-6"CBTerminal End 3.6-12 3-RCS-008-39-113Containment23'-0"CB & LS Above Threshold 3.6-12 3-RCS-008-40-114Containment31'-6"CB & LS Above Threshold 3.6-12 3-RCS-150-42-113Containment23'-0"CBTerminal End 3.6-12 3-RCS-150-42-115Containment23'-0"CB N/A - Arbitrary Intermediate3.6-12 3-RCS-150-42-116Containment22'-2"CB N/A - Arbitrary Intermediate3.6-12 3-RCS-150-42-117Containment17'-6"CBTerminal End 3.6-12 3-RCS-002-127-1TPContainment5'-5" CBN/A (2)3.6-33 3-RCS-002-127-1IPContainment4'-9"CBN/A (2)3.6-33 3-RCS-002-127-1TPContainment4'-9"CBN/A (2)3.6-33 3-RCS-002-130-1TPContainment5'-5"CBN/A (2)3.6-33 3-RCS-002-130-1IPContainment4'-9"CBN/A (2)3.6-33 3-RCS-002-130-1TPContainment4'-9"CBN/A (2)3.6-33 TABLE 3.6-12 POSTULATED BREAKS - RE ACTOR COOLANT SYSTEM - LOOP STOP VALVE BY PASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING. LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)
LINE DESIGNATIONBREAK #BUILDINGELEVATIONBREAK TYPE (1)THRESHOLD STRESSUSAGE FACTORFIGUREEQN 10EQN 12 OR 13 MPS3 UFSARMPS3 UFSAR3.6-109Rev. 303-RCS-002-135-1TPContainment5'-5" CBN/A (2)3.6-33 3-RCS-002-135-1IPContainment4'-9"CBN/A (2)3.6-33 3-RCS-002-135-1TPContainment4'-9"CBN/A (2)3.6-33 3-RCS-002-143-1TPContainment5'-5"CBN/A (2)3.6-33 3-RCS-002-143-1IPContainment4'-9"CBN/A (2)3.6-33 3-RCS-002-143-1TPContainment4'-9"CBN/A (2)3.6-33 3-RCS-002-126-1TPContainment15'-10"CBN/A (3)3.6-33 3-RCS-002-126-1IPContainment15'-2"CBN/A (3)3.6-33 3-RCS-002-126-1TPContainment10'-3"CBN/A (3)3.6-33 3-RCS-002-129-1TPContainment15'-10"CBN/A (3)3.6-33 3-RCS-002-129-1IPContainment15'-2"CBN/A (3)3.6-33 3-RCS-002-129-1TPContainment10'-3"CBN/A (3)3.6-33 3-RCS-002-134-1TPContainment15'-10"CBN/A (3)3.6-33 TABLE 3.6-12 POSTULATED BREAKS - RE ACTOR COOLANT SYSTEM - LOOP STOP VALVE BY PASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING. LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)
LINE DESIGNATIONBREAK #BUILDINGELEVATIONBREAK TYPE (1)THRESHOLD STRESSUSAGE FACTORFIGUREEQN 10EQN 12 OR 13 MPS3 UFSARMPS3 UFSAR3.6-110Rev. 303-RCS-002-134-1IPContainment15'-2"CBN/A (3)3.6-33 3-RCS-002-134-1TPContainment10'-3"CBN/A (3)3.6-33 3-RCS-002-142-1TPContainment15'-10"CBN/A (3)3.6-33 3-RCS-002-142-1IPContainment15'-2"CBN/A (3)3.6-33 3-RCS-002-142-1TPContainment10'-3"CBN/A (3)3.6-33 3-RCS-002-128-1TPContainment8'-10"CBN/A (4)3.6-33 3-RCS-002-128-1IPContainment11'-7"CBN/A (4)3.6-33 3-RCS-002-128-1TPContainment11'-7"CBN/A (4)3.6-33 3-RCS-002-131-1TPContainment8'-10"CBN/A (4)3.6-33 3-RCS-002-131-1IPContainment11'-7"CBN/A (4)3.6-33 3-RCS-002-131-1TPContainment11'-7"CBN/A (4)3.6-33 3-RCS-002-136-1TPContainment8'-10"CBN/A (4)3.6-33 3-RCS-002-136-1IPContainment11'-7"CBN/A (4)3.6-33 TABLE 3.6-12 POSTULATED BREAKS - RE ACTOR COOLANT SYSTEM - LOOP STOP VALVE BY PASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING. LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)
LINE DESIGNATIONBREAK #BUILDINGELEVATIONBREAK TYPE (1)THRESHOLD STRESSUSAGE FACTORFIGUREEQN 10EQN 12 OR 13 MPS3 UFSARMPS3 UFSAR3.6-111Rev. 303-RCS-002-136-1TPContainment11'-7"CBN/A (4)3.6-33 3-RCS-002-144-1TPContainment8'-10"CBN/A (4)3.6-33 3-RCS-002-144-1IPContainment11'-7"CBN/A (4)3.6-33 3-RCS-002-144-1TPContainment11'-7"CBN/A (4)3.6-33 3-RCS-003-137-11Containment15'-9"CBN/A Terminal End3.6-16 3-RCS-003-137-12Containment10'-1"CBAbove Threshold3.6-16 3-RCS-003-137-13Containment8'-7"CBAbove Threshold3.6-16 3-RCS-003-137-14Containment7'-0"CBAbove Threshold3.6-16 3-RCS-003-137-15Containment7'-0"CBAbove Threshold3.6-16 3-RCS-003-171-16Containment7'-0"CBAbove Threshold3.6-16 3-RCS-003-171-17Containment7'-0"CBAbove Threshold3.6-16 3-RCS-003-171-18Containment7'-0"CBAbove Threshold3.6-16 3-RCS-003-145-11Containment17'-6"CBN/A Terminal End3.6-15 3-RCS-003-145-12Containment17'-6"CBAbove Threshold3.6-15 3-RCS-003-145-13Containment17'-6"CBAbove Threshold3.6-15 TABLE 3.6-12 POSTULATED BREAKS - RE ACTOR COOLANT SYSTEM - LOOP STOP VALVE BY PASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING. LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)
LINE DESIGNATIONBREAK #BUILDINGELEVATIONBREAK TYPE (1)THRESHOLD STRESSUSAGE FACTORFIGUREEQN 10EQN 12 OR 13 MPS3 UFSARMPS3 UFSAR3.6-112Rev. 303-RCS-003-145-14Containment17'-6"CBAbove Threshold3.6-15 3-RCS-003-145-15Containment17'-2"CBAbove Threshold3.6-15 3-RCS-003-145-16Containment16'-7"CBAbove Threshold3.6-15 3-RCS-003-145-17Containment8'-0"CBAbove Threshold3.6-15 3-RCS-003-145-18Containment8'-0"CBAbove Threshold3.6-15 3-RCS-003-145-19Containment8'-0"CBAbove Threshold3.6-15 3-RCS-003-149-111Containment17'-6"CBN/A Terminal End3.6-15 3-RCS-003-149-112Containment17'-6"CBAbove Threshold3.6-15 3-RCS-003-149-113Containment17'-6"CBAbove Threshold3.6-15 3-RCS-003-149-114Containment17'-0"CBAbove Threshold3.6-15 3-RCS-003-149-115Containment16'-7"CBAbove Threshold3.6-15 3-RCS-003-149-116Containment16'-2"CBAbove Threshold3.6-15 3-RCS-003-149-117Containment7'-1"CBAbove Threshold3.6-15 3-RCS-003-149-118Containment7'-1"CBAbove Threshold3.6-15 3-RCS-003-149-119Containment7'-1"CBAbove Threshold3.6-15 TABLE 3.6-12 POSTULATED BREAKS - RE ACTOR COOLANT SYSTEM - LOOP STOP VALVE BY PASS PIPING, EXCESS LETDOWN PIPING, LOOP FILL PIPING, LOOP DRAIN PIPING. LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)
LINE DESIGNATIONBREAK #BUILDINGELEVATIONBREAK TYPE (1)THRESHOLD STRESSUSAGE FACTORFIGUREEQN 10EQN 12 OR 13 MPS3 UFSARMPS3 UFSAR3.6-113Rev. 30NOTES:(1) Circumferential Pipe Break (CB) and Longi tudinal 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 of f each Cr ossover 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 Lo op 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 Syst em. The 2% rule is defined in FSAR Section 3.6.2.1.2 under Moderate Energy Piping Systems.
MPS3 UFSARMPS3 UFSAR3.6-114Rev. 30TABLE 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 DESIGNATIONBREAK # (1)BLOWDOWN SOURCEESSENTIAL PIPE WHIP TARGETPRIMARY REQUIREMENT3-RCS-008-24-1133-RCS-275-4-1NoneNone3-RCS-008-24-1133-RCS-029-2-1 (2)3RCS-PRRBA & 3RCS-PRRFA3-RCS-008-24-1SplitHot & Cold Legs (2)3RCS-PRRCA & 3RCS-PRREA3-RCS-008-25-1143-RCS-275-4-1 (2)3RCS-PRRA1A & 3RCS-PRRA2A3-RCS-008-25-1143-RCS-029-2-1 (2)3RCS-PRR4A3-RCS-008-25-1SplitHot & Cold Legs (2)3RCS-PRRCA & 3RCS-PRREA3-RCS-150-27-113Hot & Cold LegsNoneNone 3-RCS-150-27-115Hot & Cold LegsNoneNone 3-RCS-150-27-116Hot & Cold LegsNoneNone3-RCS-150-27-117Hot & Cold LegsNoneNone3-RCS-008-29-1133-RCS-275-9-1NoneNone 3-RCS-008-29-1133-RCS-029-7-1 (2)3RCS-PRRBB & 3RCS-PRRFB3-RCS-008-29-1SplitHot & Cold Legs (2)3RCS-PRRCB & 3RCS-PRREB3-RCS-008-30-1143-RCS-275-9-1 (2)3RCS-PRRA1B & 3RCS-PRRA2B3-RCS-008-30-1143-RCS-029-7-1 (2)3RCS-PRR483-RCS-008-30-1SplitHot & Cold Legs (2)3RCS-PRRCB & 3RCS-PRREB3-RCS-150-32-113Hot & Cold LegsNoneNone 3-RCS-150-32-115Hot & Cold LegsNoneNone MPS3 UFSARMPS3 UFSAR3.6-115Rev. 303-RCS-150-32-116Hot & Cold Legs None None3-RCS-150-32-117Hot & Cold Legs None None3-RCS-008-34-1133-RCS-275-14-1None None3-RCS-008-34-1133-RCS-029-12-1 (2)3-RCS-PRRBC & 3RCS-PRRFC3-RCS-008-34-1SplitHot & Cold Legs (2)3RCS-PRRCC & 3RCS-PRREC3-RCS-008-35-1143-RCS-275-14-1 (2)3RCS-PRRA1C & 3RCS-PRRA2C3-RCS-008-35-1143-RCS-029-12-1 (2)3RCS-PRR4C3-RCS-008-35-1SplitHot & Cold Legs (2)3-RCS-PRRCC & 3RCS-PRREC3-RCS-150-37-113Hot & Cold LegsNoneNone3-RCS-150-37-115Hot & Cold LegsNoneNone3-RCS-150-37-116Hot & Cold LegsNoneNone3-RCS-150-37-117Hot & Cold LegsNoneNone 3-RCS-008-39-1133-RCS-275-19-1NoneNone3-RCS-008-39-1133-RCS-029-17-1 (2)3RCS-PRRBD & 3RCS-PRRFD3-RCS-008-39-1SplitHot & Cold Legs (2)3RCS-PRRCD & 3RCS-PRRED3-RCS-008-40-1143-RCS-275-19-1 (2)3RCS-PRRA1D & 3RCS-PRRA2D3-RCS-008-40-1143-RCS-029-17-1 (2)3RCS-PRR4D3-RCS-008-40-1SplitHot & Cold Legs (2)3-RCS-PRRCD & 3RCS-PRRED3-RCS-150-42-113Hot & Cold LegsNoneNoneTABLE 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 DESIGNATIONBREAK # (1)BLOWDOWN SOURCEESSENTIAL PIPE WHIP TARGETPRIMARY REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-116Rev. 303-RCS-150-42-115Hot & Cold Legs None None3-RCS-150-42-116Hot & Cold Legs None None3-RCS-150-42-117Hot & Cold Legs None None3-RCS-002-127-1TP3-RCS-031-3-1None None3-RCS-002-127-1IP3-RCS-031-3-1None None3-RCS-002-127-1TP3-RCS-031-3-1None None3-RCS-002-130-1TP3-RCS-031-8-1None None3-RCS-002-130-1IP3-RCS-031-8-1None None3-RCS-002-130-1TP3-RCS-031-8-1None None3-RCS-002-135-1TP3-RCS-031-13-1None None3-RCS-002-135-1IP3-RCS-031-13-1None None3-RCS-002-135-1TP3-RCS-031-13-1None None3-RCS-002-143-1TP3-RCS-031-18-1None None3-RCS-002-143-1IP3-RCS-031-18-1None None3-RCS-002-143-1TP3-RCS-031-18-1None None3-RCS-002-126-1TP3-RCS-029-2-1None None3-RCS-002-126-1IP3-RCS-029-2-1None None3-RCS-002-126-1TP3-RCS-029-2-1None None3-RCS-002-129-1TP3-RCS-029-7-1None NoneTABLE 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 DESIGNATIONBREAK # (1)BLOWDOWN SOURCEESSENTIAL PIPE WHIP TARGETPRIMARY REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-117Rev. 303-RCS-002-129-1IP3-RCS-029-7-1None None3-RCS-002-129-1TP3-RCS-029-7-1None None3-RCS-002-134-1TP3-RCS-029-12-1None None3-RCS-002-134-1IP3-RCS-029-12-1None None3-RCS-002-134-1TP3-RCS-029-12-1None None3-RCS-002-142-1TP3-RCS-029-17-1None None3-RCS-002-142-1IP3-RCS-029-17-1None None3-RCS-002-142-1TP3-RCS-029-17-1None None3-RCS-002-128-1TP3-RCS-031-3-1None None3-RCS-002-128-1IP3-RCS-031-3-1None None3-RCS-002-128-1TP3-RCS-031-3-1None None3-RCS-002-131-1TP3-RCS-031-8-1None None3-RCS-002-131-1IP3-RCS-031-8-1None None3-RCS-002-131-1TP3-RCS-031-8-1None None3-RCS-002-136-1TP3-RCS-031-13-1None None3-RCS-002-136-1IP3-RCS-031-13-1None None3-RCS-002-136-1TP3-RCS-031-13-1None None3-RCS-002-144-1TP3-RCS-031-18-1None None3-RCS-002-144-1IP3-RCS-031-18-1None NoneTABLE 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 DESIGNATIONBREAK # (1)BLOWDOWN SOURCEESSENTIAL PIPE WHIP TARGETPRIMARY REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-118Rev. 303-RCS-002-144-1TP3-RCS-031-18-1None None3-RCS-003-137-113-RCS-275-15-1None None3-RCS-003-137-123-RCS-275-15-1None None3-RCS-003-137-133-RCS-275-15-1None None3-RCS-003-137-143-RCS-275-15-1None None3-RCS-003-137-153-RCS-275-15-1None None3-RCS-003-171-163-RCS-275-15-1None None3-RCS-003-171-173-RCS-275-15-1None None3-RCS-003-171-183-RCS-275-15-1None None3-RCS-003-145-113-RCS-275-20-1None None3-RCS-003-145-123-RCS-275-20-1None None3-RCS-003-145-133-RCS-275-20-1None None3-RCS-003-145-143-RCS-275-20-1None None3-RCS-003-145-153-RCS-275-20-1None None3-RCS-003-145-163-RCS-275-20-1None None3-RCS-003-145-173-RCS-275-20-1None None3-RCS-003-145-183-RCS-275-20-1None None3-RCS-003-145-193-RCS-275-20-1None None3-RCS-003-149-1113-RCS-275-5-1None NoneTABLE 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 DESIGNATIONBREAK # (1)BLOWDOWN SOURCEESSENTIAL PIPE WHIP TARGETPRIMARY REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-119Rev. 30NOTES:(1) Repetition of Break Numbers is used to identify separate fl uid reservoirs which provide a c onstant pressure source to maint ain system pressure subsequent to the postula ted pipe break. These reservoirs maintain system blowdown during the transient event a s well as in the steady state.(2) A Large Loss of Coolan t Accident (LOCA) occurs subseque nt to this postulated break. The NS SS System Standard Design Criteri a defines the protection requirements subseque nt 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 ex ceed 20% of the flow area of the broken pi pe. 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 th e severity of the postulated LOCA.3-RCS-003-149-1123-RCS-275-5-1None None3-RCS-003-149-1133-RCS-275-5-1None None3-RCS-003-149-1143-RCS-275-5-1None None3-RCS-003-149-1153-RCS-275-5-1None None3-RCS-003-149-1163-RCS-275-5-1None None3-RCS-003-149-1173-RCS-275-5-1None None3-RCS-003-149-1183-RCS-275-5-1None None3-RCS-003-149-1193-RCS-275-5-1None NoneTABLE 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 DESIGNATIONBREAK # (1)BLOWDOWN SOURCEESSENTIAL PIPE WHIP TARGETPRIMARY REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-120Rev. 30TABLE 3.6-14 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP DRAI N PIPING, LOOP FILL PIPING,LETDOWN LINE, AND NORMAL CHARGING LINE DESIGNATIONBREAK #ESSENTIAL JET IMPINGEMENT TARGETDISTANCE TO TARGETJET INTENSITY AT THE TARGET JET LOAD ON THE TARGET PROTECTION REQUIREMENT3-RCS-008-24-113None N/AN/A N/ANone3-RCS-008-25-114None N/AN/A N/ANone3-RCS-150-27-113None N/AN/A N/ANone3-RCS-150-27-115None N/AN/A N/ANone3-RCS-150-27-116None N/AN/A N/ANone3-RCS-150-27-117None N/AN/A N/ANone3-RCS-008-29-113None N/AN/A N/ANone3-RCS-008-30-114None N/AN/A N/ANone3-RCS-150-32-113None N/AN/A N/ANone3-RCS-150-32-115None N/AN/A N/ANone3-RCS-150-32-116None N/AN/A N/ANone3-RCS-150-32-117None N/AN/A N/ANone3-RCS-008-34-113None N/AN/A N/ANone3-RCS-008-35-114None N/AN/A N/ANone3-RCS-150-37-113None N/AN/A N/ANone3-RCS-150-37-115None N/AN/A N/ANone3-RCS-150-37-116None N/AN/A N/ANone3-RCS-150-37-117None N/AN/A N/ANone3-RCS-008-39-113None N/AN/A N/ANone MPS3 UFSARMPS3 UFSAR3.6-121Rev. 303-RCS-008-40-114None N/AN/A N/ANone3-RCS-150-42-113None N/AN/A N/ANone3-RCS-150-42-115None N/AN/A N/ANone3-RCS-150-42-116None N/AN/A N/ANone3-RCS-150-42-117None N/AN/A N/ANone3-RCS-002-127-1TPNone N/AN/A N/ANone3-RCS-002-127-1IPNone N/AN/A N/ANone3-RCS-002-127-1TPNone N/AN/A N/ANone3-RCS-002-130-1TPNone N/AN/A N/ANone3-RCS-002-130-1IPNone N/AN/A N/ANone3-RCS-002-130-1TPNone N/AN/A N/ANone3-RCS-002-135-1TPNone N/AN/A N/ANone3-RCS-002-135-1IPNone N/AN/A N/ANone3-RCS-002-135-1TPNone N/AN/A N/ANone3-RCS-002-143-1TPNone N/AN/A N/ANone3-RCS-002-143-1IPNone N/AN/A N/ANone3-RCS-002-143-1TPNone N/AN/A N/ANone3-RCS-002-126-1TPNone N/AN/A N/ANone3-RCS-002-126-1IPNone N/AN/A N/ANoneTABLE 3.6-14 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP DRAI N PIPING, LOOP FILL PIPING,LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)
LINE DESIGNATIONBREAK #ESSENTIAL JET IMPINGEMENT TARGETDISTANCE TO TARGETJET INTENSITY AT THE TARGET JET LOAD ON THE TARGET PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-122Rev. 303-RCS-002-126-1TPNone N/AN/A N/ANone3-RCS-002-129-1TPNone N/AN/A N/ANone3-RCS-002-129-1IPNone N/AN/A N/ANone3-RCS-002-129-1TPNone N/AN/A N/ANone3-RCS-002-134-1TPNone N/AN/A N/ANone3-RCS-002-134-1IPNone N/AN/A N/ANone3-RCS-002-134-1TPNone N/AN/A N/ANone3-RCS-002-142-1TPNone N/AN/A N/ANone3-RCS-002-142-1IPNone N/AN/A N/ANone3-RCS-002-142-1TPNone N/AN/A N/ANone3-RCS-002-128-1TPNone N/AN/A N/ANone3-RCS-002-128-1IPNone N/AN/A N/ANone3-RCS-002-128-1TPNone N/AN/A N/ANone3-RCS-002-131-1TPNone N/AN/A N/ANone3-RCS-002-131-1IPNone N/AN/A N/ANone3-RCS-002-131-1TPNone N/AN/A N/ANone3-RCS-002-136-1TPNone N/AN/A N/ANone3-RCS-002-136-1IPNone N/AN/A N/ANone3-RCS-002-136-1TPNone N/AN/A N/ANoneTABLE 3.6-14 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP DRAI N PIPING, LOOP FILL PIPING,LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)
LINE DESIGNATIONBREAK #ESSENTIAL JET IMPINGEMENT TARGETDISTANCE TO TARGETJET INTENSITY AT THE TARGET JET LOAD ON THE TARGET PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-123Rev. 303-RCS-002-144-1TPNone N/AN/A N/ANone3-RCS-002-144-1IPNone N/AN/A N/ANone3-RCS-002-144-1TPNone N/AN/A N/ANone3-RCS-003-137-11 None N/AN/A N/ANone3-RCS-003-137-12 None N/AN/A N/ANone3-RCS-003-137-13 None N/AN/A N/ANone3-RCS-003-137-14 None N/AN/A N/ANone3-RCS-003-137-15 None N/AN/A N/ANone3-RCS-003-137-16 None N/AN/A N/ANone3-RCS-003-137-17 None N/AN/A N/ANone3-RCS-003-137-18 None N/AN/A N/ANone3-RCS-003-145-11 None None None None None3-RCS-003-145-12 None None None None None3-RCS-003-145-133RCS*V132 on 3-RCS-150-042-1 a None3-RCS-003-145-143RCS*V132 on 3-RCS-150-042-1(a)None3-RCS-003-145-15NoneNoneNoneNoneNone3-RCS-003-145-16NoneNoneNoneNoneNoneTABLE 3.6-14 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP DRAI N PIPING, LOOP FILL PIPING,LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)
LINE DESIGNATIONBREAK #ESSENTIAL JET IMPINGEMENT TARGETDISTANCE TO TARGETJET INTENSITY AT THE TARGET JET LOAD ON THE TARGET PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-124Rev. 303-RCS-003-145-173RCS*V132 on 3-RCS-150-042-1(a)None3-RCS-003-145-183RCS*V132 on 3-RCS-150-042-1(a)None3-RCS-003-145-193RCS*V132 on 3-RCS-150-042-1(a)None3-RCS-003-149-1113RCS*V25 on 3-RCS-750-110-2 b None3-RCS-003-149-1123RCS*V25 on 3-RCS-750-110-2 (b)None3-RCS-003-149-113None None None None None3-RCS-003-149-114None None None None None3-RCS-003-149-115None None None None None3-RCS-003-149-116None None None None None3-RCS-003-149-117None None None None None3-RCS-003-149-118None None None None None3-RCS-003-149-119None None None None NoneTABLE 3.6-14 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - LOOP STOP VALVE BYPASS PIPING, EXCESS LETDOWN PIPING, LOOP DRAI N PIPING, LOOP FILL PIPING,LETDOWN LINE, AND NORMAL CHARGING (CONTINUED)
LINE DESIGNATIONBREAK #ESSENTIAL JET IMPINGEMENT TARGETDISTANCE TO TARGETJET INTENSITY AT THE TARGET JET LOAD ON THE TARGET PROTECTION REQUIREMENT MPS3 UFSARMPS3 UFSAR3.6-125Rev. 30a.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 Syst em connecting to the Hot Leg. Loss of pressure boundary is admitted a nd since this connection is to the Hot Leg, the break propagation is transmitted to an unaffecte d 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 Pa ge 2 of NEU-6039. The break pr opagation for 3-RCS-150-042-1, na mely 1.4 square inches, and the given break area of the ruptured pipe , 3-RCS-003-145-1, namely 5.4 square inches ar e enveloped by the Case 1.1 analyzed in NEU-6039. The conclusion of the NSSS vendor is that when break s 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 comb ined Hot Leg/Cold Leg break cas e of 20 square inches in each loop, the total break areas is 40 square inches which corresponds to a 7.1" equi valent 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 expe cted to be less limiting th an 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 connecti on from the Hot Leg. Loss of pressure boundary is admitted a nd since this connection is to the Hot Leg, the break propagation is transmitted to an unaffecte d leg of an affected loop. This result was transmitted to the NSSS Vendor via NE S-40190 Dated November 27, 1985. A review was performed by the NSSS Vendor concerning th is propagation and the re sults are transmitted vi a 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 squa re inches. Therefore phenomena associated at the initial location are not affect ed by the propagation and calculated Emergency Core Cooling System (E CCS) performance is about the sa me as it would be without any propagation. The interaction is acceptable therefore protective hardware is not required.
MPS3 UFSARMPS3 UFSAR3.6-126Rev. 30TABLE 3.6-15 POSTULATED BREAKS - REACTOR COOLANT SYSTEM - PRESSURI ZER CUBICLE PIPING LINE DESIGNATIONBREAK #BUILDINGELEVATIONBREAK TYPE (1)THRESHOLD STRESS USAGE FACTORFIGURE 3.6-EQN 10EQN 12 OR 133-RCS-014-64-111Containment19'-14"CBN/A Terminal End 14 3-RCS-014-64-137Containment20'-2"CB & LSAbove Threshold 14 3-RCS-014-64-138Containment22'-4"CB & LSAbove Threshold 14 3-RCS-014-64-139Containment22'-4"CB & LSAbove Threshold 14 3-RCS-014-64-140Containment22'-4"CB & LSAbove Threshold 14 3-RCS-014-64-141Containment22'-4"CB & LSAbove Threshold 14 3-RCS-014-64-142Containment22'-4"CB & LSAbove Threshold 14 3-RCS-014-64-143Containment22'-4"CB & LSAbove Threshold 14 3-RCS-014-64-144Containment22'-4"CB & LSAbove Threshold 14 3-RCS-014-64-145Containment22'-4"CB & LSAbove Threshold 14 3-RCS-014-64-146Containment22'-4"CB & LSAbove Threshold 14 3-RCS-014-64-147Containment22'-4"CB & LSAbove Threshold 14 3-RCS-014-64-148Containment22'-4"CB & LSAbove Threshold 14 3-RCS-014-64-149Containment22'-4"CB & LSAbove Threshold 14 3-RCS-014-64-150Containment22'-4"CB & LSAbove Threshold 14 3-RCS-014-64-151Containment22'-4"CB & LSAbove Threshold 14 3-RCS-014-64-152Containment22'-4"CB & LSAbove Threshold 14 3-RCS-014-64-153Containment22'-4"CB & LSAbove Threshold 14 MPS3 UFSARMPS3 UFSAR3.6-127Rev. 303-RCS-014-64-154Containment22-4"CB & LSAbove Threshold14 3-RCS-014-64-155Containment22'-4"CB & LSAbove Threshold14 3-RCS-014-64-156Containment22'-4"CB & LSAbove Threshold14 3-RCS-014-64-157Containment22'-4"CB & LSAbove Threshold14 3-RCS-014-64-158Containment22'-4"CB & LSAbove Threshold14 3-RCS-014-64-159Containment22'-4"CB & LSAbove Threshold14 3-RCS-014-64-160Containment22'-4"CB & LSAbove Threshold14 3-RCS-014-64-161Containment22'-4"CB & LSAbove Threshold14 3-RCS-014-64-162Containment22'-4"CB & LSAbove Threshold14 3-RCS-014-64-163Containment24'-4"CB & LSAbove Threshold14 3-RCS-014-64-164Containment25'-11"CBN/A Terminal End14 3-RCS-004-224-11Containment78'-6"CBN/A Terminal End14 3-RCS-004-224-12Containment81'-3"CB & LSAbove Threshold14 3-RCS-004-224-13Containment82'-11"CB & LSAbove Threshold14 3-RCS-006-68-14Containment62'-3"CB & LSAbove Threshold14 3-RCS-004-61-15Containment52'-1"CB & LSAbove Threshold14 3-RCS-004-22-16Containment37'-8"CBN/A Terminal End14 3-RCS-004-22-17Containment18'-0"CBN/A - Arbitrary Intermediate14 TABLE 3.6-15 POSTULATED BREAKS - REACTOR COOLANT SYSTEM - PRESSURI ZER CUBICLE PIPING LINE DESIGNATIONBREAK #BUILDINGELEVATIONBREAK TYPE (1)THRESHOLD STRESS USAGE FACTORFIGURE 3.6-EQN 10EQN 12 OR 13 MPS3 UFSARMPS3 UFSAR3.6-128Rev. 303-RCS-004-22-18Containment17'-6"CB N/A - Arbitrary Intermediate14 3-RCS-004-22-19Containment17'-6"CBN/A Terminal End 14 3-RCS-004-60-110Containment54'-4"CB & LSAbove Threshold 14 3-RCS-004-21-111Containment52'-1"CB & LSAbove Threshold 14 3-RCS-004-21-112Containment41'-6"CBN/A Terminal End 14 3-RCS-004-21-113Containment17'-6"CB & LSAbove Threshold 14 3-RCS-004-21-114Containment17'-6"CB N/A - Arbitrary Intermediate14 3-RCS-004-21-115Containment17'-6"CBN/A Terminal End 14 3-RCS-002-170-116Containment16'-6"CBN/A Terminal End 14 3-RCS-002-150-117Containment62'-3"CBN/A Terminal End 14 3-RCS-006-65-118Containment77'-1"CBN/A Terminal End 14 3-RCS-006-65-120Containment77'-9"CB & LSAbove Threshold 14 3-RCS-003-66-121Containment75'-0"CBAbove Threshold 14 3-RCS-003-69-122Containment75'-0"CBN/A Terminal End 14 3-RCS-003-67-123Containment75'-0"CBAbove Threshold 14 3-RCS-003-70-124Containment75'-0"CBN/A Terminal End 14 3-RCS-003-69-137Containment75'-0"CBAbove Threshold 14 3-RCS-003-69-138Containment75'-0"CBAbove Threshold 14 TABLE 3.6-15 POSTULATED BREAKS - REACTOR COOLANT SYSTEM - PRESSURI ZER CUBICLE PIPING LINE DESIGNATIONBREAK #BUILDINGELEVATIONBREAK TYPE (1)THRESHOLD STRESS USAGE FACTORFIGURE 3.6-EQN 10EQN 12 OR 13 MPS3 UFSARMPS3 UFSAR3.6-129Rev. 30NOTE:(1) Circumferential Pipe Break (CB) and Longi tudinal Pipe Split (LS) are defined in FSAR Section 3.6.2.1.3
.3-RCS-003-70-139Containment75'-0"CBAbove Threshold 14 3-RCS-006-82-125Containment77'-1"CBN/A Terminal End 14 3-RCS-006-82-126Containment74'-11"CB & LSAbove Threshold 14 3-RCS-006-82-127Containment74'-11"CBN/A - Arbitrary Intermediate14 3-RCS-006-82-128Containment77'-5"CBN/A Terminal End 14 3-RCS-006-83-129Containment77'-1"CBN/A Terminal End 14 3-RCS-006-83-130Containment74'-11"CB & LSAbove Threshold 14 3-RCS-006-83-131Containment74'-11"CBN/A - Arbitrary Intermediate14 3-RCS-006-83-132Containment77'-5"CBN/A Terminal End 14 3-RCS-006-84-133Containment77'-1"CBN/A Terminal End 14 3-RCS-006-84-134Containment74'-11"CB & LSAbove Threshold 14 3-RCS-006-84-135Containment74'-11"CBN/A - Arbitrary Intermediate14 3-RCS-006-84-136Containment77'-5"CBN/A Terminal End 14 TABLE 3.6-15 POSTULATED BREAKS - REACTOR COOLANT SYSTEM - PRESSURI ZER CUBICLE PIPING LINE DESIGNATIONBREAK #BUILDINGELEVATIONBREAK TYPE (1)THRESHOLD STRESS USAGE FACTORFIGURE 3.6-EQN 10EQN 12 OR 13 MPS3 UFSARMPS3 UFSAR3.6-130Rev. 30TABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURI ZER CUBICLE PIPING Line DesignationBreak #
(1)Blowdown SourceEssential Pipe Whip TargetProtection Requirement3-RCS-014-64-111 3RCS*TK13RCS*SG1B Support System3RCS-PRR13-RCS-014-64-137 3RCS*TK13RCS*SG1B Support System3RCS-PRR13-RCS-014-64-1373-RCS-029-6-1None None3-RCS-014-64-1Split (2)None None3-RCS-014-64-1 38 3RCS*TK1 None None3-RCS-014-64-1383-RCS-029-6-1None None3-RCS-014-64-1Split (2)None None3-RCS-014-64-139 3RCS*TK1None None3-RCS-014-64-1393-RCS-029-6-1None None3-RCS-014-64-1Split (2)None None3-RCS-014-64-140 3RCS*TK1None None3-RCS-014-64-1403-RCS-029-6-1None None3-RCS-014-64-1Split (2)None None3-RCS-014-64-141 3RCS*TK1None None3-RCS-014-64-1413-RCS-029-6-1None None3-RCS-014-64-1Split (2)None None3-RCS-014-64-142 3RCS*TK1None None3-RCS-014-64-1423-RCS-029-6-1None None MPS3 UFSARMPS3 UFSAR3.6-131Rev. 303-RCS-014-64-1Split (2)None None3-RCS-014-64-143 3RCS*TK1Pressurizer Cubicle Wall (3)3RCS-PRR33-RCS-014-64-1433-RCS-029-6-1NoneNone3-RCS-014-64-1Split (2)None None3-RCS-014-64-144 3RCS*TK1Pressurizer Cubicle Wall (3)3RCS-PRR33-RCS-014-64-1443-RCS-029-6-1NoneNone3-RCS-014-64-1Split (2)None None3-RCS-014-64-145 3RCS*TK1Pressurizer Cubicle Wall (3)3RCS-PRR33-RCS-014-64-1453-RCS-029-6-1NoneNone3-RCS-014-64-1Split (2)None None3-RCS-014-64-146 3RCS*TK1Pressurizer Cubicle Wall (3)3RCS-PRR33-RCS-014-64-1463-RCS-029-6-1NoneNone 3-RCS-014-64-1Split (2)None None3-RCS-014-64-147 3RCS*TK1Pressurizer Cubicle Wall (3)3RCS-PRR33-RCS-014-64-1473-RCS-029-6-1NoneNone3-RCS-014-64-1Split (2)None None3-RCS-014-64-148 3RCS*TK1Pressurizer Cubicle Wall (3)3RCS-PRR3TABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURI ZER CUBICLE PIPING Line DesignationBreak #
(1)Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-132Rev. 303-RCS-014-64-1483-RCS-029-6-1None None3-RCS-014-64-1Split (2)None None3-RCS-014-64-149 3RCS*TK1Crane Wall (3)3RCS-PRR53-RCS-014-64-1493-RCS-029-6-1Pressurizer Cubicle Wall (3)3RCS-PRR33-RCS-014-64-1Split (2)None None3-RCS-014-64-150 3RCS*TK1Crane Wall (3)3RCS-PRR53-RCS-014-64-1503-RCS-029-6-1Pressurizer Cubicle Wall (3)3RCS-PRR33-RCS-014-64-1Split (2)None None3-RCS-014-64-151 3RCS*TK1Crane Wall (3)3RCS-PRR53-RCS-014-64-1513-RCS-029-6-1Pressurizer Cubicle Wall (3)3RCS-PRR33-RCS-014-64-1Split (2)None None3-RCS-014-64-152 3RCS*TK1Crane Wall (3)3RCS-PRR53-RCS-014-64-1523-RCS-029-6-1Pressurizer Cubicle Wall (3)3RCS-PRR33-RCS-014-64-1Split (2)None None3-RCS-014-64-153 3RCS*TK1Crane Wall -
(3)3RCS-PRR63-RCS-014-64-1533-RCS-029-6-1Pressurizer Cubicle Wall (3)3RCS-PRR4TABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURI ZER CUBICLE PIPING Line DesignationBreak #
(1)Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-133Rev. 303-RCS-014-64-1Split (2)None None3-RCS-014-64-154 3RCS*TK1Crane Wall (3)3RCS-PRR63-RCS-014-64-1543-RCS-029-6-1Pressurizer Cubicle Wall (3)3RCS-PRR43-RCS-014-64-1Split (2)None None3-RCS-014-64-155 3RCS*TK1Crane Wall (3)3RCS-PRR63-RCS-014-64-1553-RCS-029-6-1Pressurizer Cubicle Wall (3)3RCS-PRR43-RCS-014-64-1Split (2)None None3-RCS-014-64-156 3RCS*TK1Crane Wall (3)3RCS-PRR63-RCS-014-64-1563-RCS-029-6-1Pressurizer Cubicle Wall (3)3RCS-PRR43-RCS-014-64-1Split (2)None None3-RCS-014-64-157 3RCS*TK1Crane Wall (3)3RCS-PRR63-RCS-014-64-1573-RCS-029-6-1Pressurizer Cubicle Wall (3)3RCS-PRR43-RCS-014-64-1Split (2)None None3-RCS-014-64-158 3RCS*TK1Crane Wall (3)3RCS-PRR63-RCS-014-64-1583-RCS-029-6-1Pressurizer Cubicle Wall (3)3RCS-PRR73-RCS-014-64-1Split (2)None NoneTABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURI ZER CUBICLE PIPING Line DesignationBreak #
(1)Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-134Rev. 303-RCS-014-64-159 3RCS*TK1Crane Wall (3)3RCS-PRR63-RCS-014-64-1593-RCS-029-6-1Pressurizer Cubicle Wall (3)3RCS-PRR73-RCS-014-64-1Split (2)None None3-RCS-014-64-160 3RCS*TK1Crane Wall (3)3RCS-PRR63-RCS-014-64-1603-RCS-029-6-1Pressurizer Cubicle Wall (3)3RCS-PRR73-RCS-014-64-1Split (2)None None3-RCS-014-64-161 3RCS*TK1Pressurizer Cubicle Wall (3)3RCS-PRR83-RCS-014-64-1613-RCS-029-6-1Pressurizer Cubicle Wall (3)3RCS-PRR73-RCS-014-64-1Split (2)None None3-RCS-014-64-162 3RCS*TK1Pressurizer Cubicle Wall (3)3RCS-PRR83-RCS-014-64-1623-RCS-029-6-1Pressurizer Cubicle Wall (3)3RCS-PRR73-RCS-014-64-1Split (2)None None3-RCS-014-64-163 3RCS*TK1None None3-RCS-014-64-163 3-RCS-029-6-1Pressurizer Cubicle Wall (3)3RCS-PRR83-RCS-014-64-1Split (2)None None3-RCS-014-64-164 3-RCS-029-6-1Pressurizer Cubicle Wall (3)3RCS-PRR8TABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURI ZER CUBICLE PIPING Line DesignationBreak #
(1)Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-135Rev. 303-RCS-004-224-11 (4)None None3-RCS-004-224-12 (4)None None3-RCS-004-224-12 3RCS*TK1 None None3-RCS-004-224-1Split (4)None None3-RCS-004-224-13 (4)None None3-RCS-004-224-13 3RCS*TK1 None None3-RCS-004-224-1Split (4)None None3-RCS-006-68-14 (4)None None3-RCS-006-68-14 3RCS*TK1 None None3-RCS-006-68-1Split (4)None None3-RCS-004-61-15 (4)None None3-RCS-004-61-15 3RCS*TK1 None None3-RCS-004-61-1Split (4)None None3-RCS-004-22-16 (4)None None3-RCS-004-22-16 3RCS*TK1 None None3-RCS-004-22-17 3RCS*TK1 None (5)None3-RCS-004-22-173-RCS-275-10-1 None (5)NoneTABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURI ZER CUBICLE PIPING Line DesignationBreak #
(1)Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-136Rev. 303-RCS-004-22-18 3RCS*TK1 None (5)None3-RCS-004-22-183-RCS-275-10-1 None (5)None3-RCS-004-22-193RCS*TK1NoneNone3-RCS-004-60-1103RCS*TK1NoneNone3-RCS-004-60-110 (4)None None3-RCS-004-60-1Split (4)None None3-RCS-004-21-111 3RCS*TK1None None3-RCS-004-21-1113-RCS-275-10-1None None3-RCS-004-21-1Split (4)None None3-RCS-004-21-112 3RCS*TK1None None3-RCS-004-21-1123-RCS-275-5-1None None3-RCS-004-21-113 3RCS*TK1None None3-RCS-004-21-1133-RCS-275-5-1None None3-RCS-004-21-1Split (4)None None3-RCS-004-21-114 3RCS*TK1 None (5)None3-RCS-004-21-1143-RCS-275-5-1 None (5)None3-RCS-004-21-1153RCS*TK1NoneNoneTABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURI ZER CUBICLE PIPING Line DesignationBreak #
(1)Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-137Rev. 303-RCS-004-21-1153-RCS-275-5-1None None3-RCS-002-170-116 (4)None None3-RCS-002-150-117 (4)None None3-RCS-006-65-118 3RCS*TK1None None3-RCS-006-65-1Split 3RCS*TK1 None None3-RCS-006-65-120 3RCS*TK1 3-RCS-004-224-1 (6)3-RCS-PRR9063-RCS-003-66-1213RCS*TK1NoneNone3-RCS-003-69-1223RCS*TK1NoneNone3-RCS-003-67-1233RCS*TK1NoneNone3-RCS-003-70-1243RCS*TK1NoneNone 3-RCS-006-82-1253RCS*TK1NoneNone3-RCS-006-82-1263RCS*TK1NoneNone3-RCS-006-82-1Split3RCS*TK1NoneNone 3-RCS-006-82-1273RCS*TK1 None (5)None3-RCS-006-82-1283RCS*TK1NoneNone3-RCS-006-83-1293RCS*TK1NoneNone3-RCS-006-83-1303RCS*TK1NoneNone 3-RCS-006-83-1Split3RCS*TK1NoneNoneTABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURI ZER CUBICLE PIPING Line DesignationBreak #
(1)Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-138Rev. 30NOTES:(1) Repetition of Break Numbers is used to identify separate fl uid reservoirs which provide a c onstant pressure source to maint ain system pressure subsequent to a postulate d pipe break. These reservoirs maintain sy stem blowdown during the transient event as well as in the steady state.(2) Longitudinal Splits (LS) on the Pressurizer Sur ge Line result in blowdown fr om 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 wh ip of the Pressurizer Sur ge 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 unrestrai ned impact to the structure.3-RCS-006-83-131 3RCS*TK1 None (5)None3-RCS-006-83-1323RCS*TK1NoneNone3-RCS-006-84-1333RCS*TK1NoneNone3-RCS-006-84-1343RCS*TK1 3RCS-004-224-1 (6)3RCS-PRR9453-RCS-006-84-1Split3RCS*TK1NoneNone 3-RCS-006-84-1353RCS*TK1 None (5)None3-RCS-006-84-1363RCS*TK1NoneNone3-RCS-003-69-1373RCS*TK1NoneNone3-RCS-003-69-1383RCS*TK1NoneNone 3-RCS-003-70-1393RCS*TK1NoneNoneTABLE 3.6-16 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - PRESSURI ZER CUBICLE PIPING Line DesignationBreak #
(1)Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-139Rev. 30(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 th e ef fects of pipe whip for Arbi trary Intermediate Breaks (AIBs
). herefore pipe whip effects for AIBs on the Pressurizer Safety Li nes 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 r upture restraint is provided to prevent the interaction.
MPS3 UFSARMPS3 UFSAR3.6-140Rev. 30TABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSURI ZER CUBICLE PIPING Line DesignationBreak # (1)Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement3-RCS-014-64-1113-RCS-008-30-1 (2)None3-RCS-014-64-1113-RCS-006-119-1 (3)None3-RCS-014-64-1373-RCS-008-30-1 (2)None3-RCS-014-64-1373-RCS-006-119-1 (3)None3-RCS-014-64-1SplitPrimary Shield Wall2 Feet1274 Psi774 KipsNone3-RCS-014-64-1Split3-RCS-008-30-1 (4)None3-RCS-014-64-138Primary Shield Wall Less Severe than Longitudinal Split Number 37 (5)None3-RCS-014-64-138NoneN/AN/AN/ANone3-RCS-014-64-1SplitNoneN/AN/AN/ANone 3-RCS-014-64-139Primary Shield Wall Less Severe than Longitudinal Split Number 37 (5)None3-RCS-014-64-139NoneN/AN/AN/ANone 3-RCS-014-64-1SplitNoneN/AN/AN/ANone3-RCS-014-64-140Primary Shield Wall Less Severe than Longitudinal Split Number 37 (5)None3-RCS-014-64-140NoneN/AN/AN/ANone3-RCS-014-64-1SplitNoneN/AN/AN/ANone 3-RCS-014-64-141Primary Shield Wall Less Severe than Longitudinal Split Number 37 (5)None3-RCS-014-64-141NoneN/AN/AN/ANone3-RCS-014-64-1SplitNoneN/AN/AN/ANone3-RCS-014-64-142Primary Shield Wall Less Severe than Longitudinal Split Number 37 (5)None MPS3 UFSARMPS3 UFSAR3.6-141Rev. 303-RCS-014-64-142None N/AN/A N/ANone3-RCS-014-64-1SplitNone N/AN/A N/ANone3-RCS-014-64-143None N/AN/A N/ANone3-RCS-014-64-143None N/AN/A N/ANone3-RCS-014-64-1SplitNone N/AN/A N/ANone3-RCS-014-64-144North-South Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None3-RCS-014-64-144NoneN/AN/AN/ANone3-RCS-014-64-1SplitRemovable Concrete Slab at Elevation 12'-9" Less Severe than Longitudinal Split Number 48 None (6)3-RCS-014-64-145Primary Shield Wall Less Severe than Longitudinal Split Number 37 (5)None3-RCS-014-64-145Steam Generator Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None3-RCS-014-64-1SplitRemovable Concrete Slab at Elevation 12'-9"9 Feet15 psi65 Kips None (6)3-RCS-014-64-146North-South Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None3-RCS-014-64-146Steam Generator Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)NoneTABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSU RIZER CUBICLE PIPING Line DesignationBreak # (1)Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-142Rev. 303-RCS-014-64-1SplitRemovable Concrete Slab at Elevation 12'-9"9 Feet15 psi101 Kips None (6)3-RCS-014-64-147North-South Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None3-RCS-014-64-147Steam Generator Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None3-RCS-014-64-1SplitRemovable Concrete Slab at Elevation 12'-9"9 Feet15 psi125 KipsNone3-RCS-014-64-148North-South Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None3-RCS-014-64-148Steam Generator Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None3-RCS-014-64-1SplitRemovable Concrete Slab at Elevation 12'-9"9 Feet15 psi125 KipsNone3-RCS-014-64-1493-RCS-004-21-1Less Severe than Longitudinal Split Number 49None3-RCS-014-64-149North-South Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None3-RCS-014-64-1Split3-RCS-004-21-14 Feet586 psi6 Kips None (7)3-RCS-014-64-1503-RCS-004-21-1Less Severe than Longitudinal Split Number 50None3-RCS-014-64-150North-South Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)NoneTABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSU RIZER CUBICLE PIPING Line DesignationBreak # (1)Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-143Rev. 303-RCS-014-64-1Split3-RCS-004-21-14 Feet 586 psi6 Kips None (7)3-RCS-014-64-1513-RCS-004-21-1Less Severe than Longitudinal Split Number 51 None (8)3-RCS-014-64-151North-South Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None (8)3-RCS-014-64-1Split3-RCS-004-21-14 Feet 586 psi6 Kips None (7)3-RCS-014-64-1523-RCS-004-21-1Less Severe than Longitudinal Split Number 52 None (8)3-RCS-014-64-152North-South Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None (8)3-RCS-014-64-1Split3-RCS-004-21-14 Feet 586 psi6 Kips None (7)3-RCS-014-64-153Crane Wall Less Severe than Longitudinal Split Number 37 (5)None (8)3-RCS-014-64-153Removable Concrete Slab at Elevation 12'-9" Less Severe than Longitudinal Split Number 48 None (8)3-RCS-014-64-1SplitRemovable Concrete Slab at Elevation 12'-9" Less Severe than Longitudinal Split Number 48 None (6)3-RCS-014-64-154Crane Wall Less Severe than Longitudinal Split Number 37 (5)None (8)3-RCS-014-64-154Removable Concrete Slab at Elevation 12'-9" Less Severe than Longitudinal Split Number 48 None (8)TABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSU RIZER CUBICLE PIPING Line DesignationBreak # (1)Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-144Rev. 303-RCS-014-64-1SplitRemovable Concrete Slab at Elevation 12'-9" Less Severe than Longitudinal Split Number 48 None (6)3-RCS-014-64-155Crane Wall Less Severe than Longitudinal Split Number 37 (5)None (8)3-RCS-014-64-155Removable Concrete Slab at Elevation 12'-9" Less Severe than Longitudinal Split Number 48 None (8)3-RCS-014-64-1SplitRemovable Concrete Slab at Elevation 12'-9"9 Feet15 psi66 Kips None (6)3-RCS-014-64-156Crane Wall Less Severe than Longitudinal Split Number 37 (5)None (8)3-RCS-014-64-156Steam Generator Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None (8)3-RCS-014-64-1SplitRemovable Concrete Slab at Elevation 12'-9" Less Severe than Longitudinal Split Number 48 None (6)3-RCS-014-64-157Steam Generator Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None (8)3-RCS-014-64-157Removable Concrete Slab at Elevation 12'-9" Less Severe than Longitudinal Split Number 48 None (8)3-RCS-014-64-1SplitRemovable Concrete Slab at Elevation 12'-9" Less Severe than Longitudinal Split Number 48 None (6)TABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSU RIZER CUBICLE PIPING Line DesignationBreak # (1)Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-145Rev. 303-RCS-014-64-158Steam Generator Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None (8)3-RCS-014-64-158Removable Concrete Slab at Elevation 12'-9" Less Severe than Longitudinal Split Number 48 None (8)3-RCS-014-64-1SplitRemovable Concrete Slab at Elevation 12'-9" Less Severe than Longitudinal Split Number 48 None (6)3-RCS-014-64-159Steam Generator Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None (8)3-RCS-014-64-159Removable Concrete Slab at Elevation 12'-9" Less Severe than Longitudinal Split Number 48 None (8)3-RCS-014-64-1SplitRemovable Concrete Slab at Elevation 12'-9" Less Severe than Longitudinal Split Number 48 None (8)3-RCS-014-64-160Steam Generator Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None (8)3-RCS-014-64-160Removable Concrete Slab at Elevation 12'-9" Less Severe than Longitudinal Split Number 48 None (8)3-RCS-014-64-1SplitRemovable Concrete Slab at Elevation 12'-9" Less Severe than Longitudinal Split Number 48 None (6)TABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSU RIZER CUBICLE PIPING Line DesignationBreak # (1)Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-146Rev. 303-RCS-014-64-161Crane Wall Less Severe than Longitudinal Split Number 37 (5)None (8)3-RCS-014-64-161Steam Generator Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None (8)3-RCS-014-64-1SplitRemovable Concrete Slab at Elevation 12'-9" Less Severe than Longitudinal Split Number 48 None (6)3-RCS-014-64-162Crane Wall Less Severe than Longitudinal Split Number 37 (5)None (8)3-RCS-014-64-162Removable Concrete Slab at Elevation 12'-9" Less Severe than the Longitudinal Split Number 62 None (8)3-RCS-014-64-1SplitRemovable Concrete Slab at Elevation 12'-9"9 Feet15 psi131 Kips None (6)3-RCS-014-64-163Crane Wall Less Severe than Longitudinal Split Number 37 (5)None (8)3-RCS-014-64-163Steam Generator Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None (8)3-RCS-014-64-1SplitRemovable Concrete Slab at Elevation 12'-9" Less Severe than Longitudinal Split Number 37 (5)None3-RCS-014-64-164NoneN/AN/AN/ANone3-RCS-004-224-11NoneN/AN/AN/ANone3-RCS-004-224-11NoneN/AN/AN/ANone 3-RCS-004-224-12NoneN/AN/AN/ANoneTABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSU RIZER CUBICLE PIPING Line DesignationBreak # (1)Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-147Rev. 303-RCS-004-224-12None N/AN/A N/ANone3-RCS-004-224-1SplitNone N/AN/A N/ANone3-RCS-004-224-13None N/AN/A N/ANone3-RCS-004-224-13None N/AN/A N/ANone3-RCS-004-224-1SplitNone N/AN/A N/ANone3-RCS-006-68-14None N/AN/A N/ANone3-RCS-006-68-14None N/AN/A N/ANone3-RCS-006-68-1SplitNone N/AN/A N/ANone3-RCS-004-61-15None N/AN/A N/ANone3-RCS-004-61-15None N/AN/A N/ANone3-RCS-004-61-1SplitNone N/AN/A N/ANone3-RCS-004-22-16None N/AN/A N/ANone3-RCS-004-22-16None N/AN/A N/ANone3-RCS-004-22-17None N/AN/A N/ANone3-RCS-004-22-17None N/AN/A N/ANone3-RCS-004-22-18None N/AN/A N/ANone3-RCS-004-22-18None N/AN/A N/ANone3-RCS-004-22-19None N/AN/A N/ANone3-RCS-004-22-19None N/AN/A N/ANone3-RCS-004-60-110None N/AN/A N/ANoneTABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSU RIZER CUBICLE PIPING Line DesignationBreak # (1)Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-148Rev. 303-RCS-004-60-110None N/AN/A N/ANone3-RCS-004-60-1SplitNone N/AN/A N/ANone3-RCS-004-21-111None N/AN/A N/ANone3-RCS-004-21-111None N/AN/A N/ANone3-RCS-004-21-1SplitNone N/AN/A N/ANone3-RCS-004-21-112None N/AN/A N/ANone3-RCS-004-21-112None N/AN/A N/ANone3-RCS-004-21-113North-South Cubicle Wall Less Severe than Longitudinal Split Number 37 (5)None3-RCS-004-21-113NoneN/AN/AN/ANone3-RCS-004-21-1SplitNoneN/AN/AN/ANone3-RCS-004-21-114 None (9)N/AN/A N/ANone3-RCS-004-21-114None N/AN/A N/ANone3-RCS-004-21-115None N/AN/A N/ANone3-RCS-004-21-115None N/AN/A N/ANone3-RCS-002-170-116None N/AN/A N/ANone3-RCS-002-150-117None N/AN/A N/ANone3-RCS-006-65-1183-RCS-004-224-1Le ss Severe than Either L ongitudinal Split Numbers 49 Through 52 None (10)3-RCS-006-65-120None N/AN/A N/ANone3-RCS-006-65-1SplitNone N/AN/A N/ANoneTABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSU RIZER CUBICLE PIPING Line DesignationBreak # (1)Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-149Rev. 303-RCS-003-66-121None N/AN/A N/ANone3-RCS-003-69-122None N/AN/A N/ANone3-RCS-003-67-123None N/AN/A N/ANone3-RCS-003-70-124None N/AN/A N/ANone3-RCS-003-69-137None N/AN/A N/ANone3-RCS-003-69-138None N/AN/A N/ANone3-RCS-003-70-139None N/AN/A N/ANone3-RCS-006-82-125None N/AN/A N/ANone3-RCS-006-82-126None N/AN/A N/ANone3-RCS-006-82-1SplitNone N/AN/A N/ANone3-RCS-006-82-127 None (9)N/AN/A N/ANone3-RCS-006-82-128None N/AN/A N/ANone3-RCS-006-83-129None N/AN/A N/ANone3-RCS-006-83-130None N/AN/A N/ANone3-RCS-006-83-1SplitNone N/AN/A N/ANone3-RCS-006-83-131 None (9)N/AN/A N/ANone3-RCS-006-83-132None N/AN/A N/ANone3-RCS-006-84-133None N/AN/A N/ANone3-RCS-006-84-134None N/AN/A N/ANone3-RCS-006-84-1SplitNone N/AN/A N/ANoneTABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSU RIZER CUBICLE PIPING Line DesignationBreak # (1)Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-150Rev. 30NOTES:(1) Repetition of Break Numbers is used to identify separate fl uid reservoirs which provide a c onstant pressure source to maint ain system pressure subsequent to the postula ted pipe break. These reservoirs maintain system blowdown during the transient event a s well as in the steady state.(2) Terminal End break is a Large LOCA. The NSSS System Sta ndard Design Criteria requirement s are deemed essential. The requirement for a Lar ge LOCA is to prevent leg to leg propagation within an affected l oop 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 viol ate this criteria since the target pipe is loaded proximate to the Hot Leg nozzle and the remainder of the Loop Stop Valv e 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-ende d rupture since pipe rupture restraint 3RCS-PRR1 limits separati on and results in l ess than a full flow area exposed thus minimizing the jet loads on the Loop Stop Valve Bypass Line 3-RCS-008-30-1.(3) The NSSS requirement delineated in Note 1 is applicable for this interaction. T he Low Pressu re Safety Injection Line 3-RCS-006-119-1 to the Hot Leg does not lo se pressure boundary since the por tion 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 sepa ration and thus limits the amo unt 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 pos tulated split targets the Loop Stop Valve Bypa ss Line. The criteria delineated in Not e 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 pi pe rupture restraints and by the Hot Le g 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 barr ier. This conclusion is based upon generic calculations which c onclude that extremely high intensities are required to cause local barrier failure. The intensity, distance, and load for jet impingement effects on the Pressurizer C ubicle Piping is governed by the Longitudinal Split (LS) Number 37 which is the closest to the indicated barrier and produces the larg est loads.3-RCS-006-84-135 None (9)N/AN/A N/ANone3-RCS-006-84-136None N/AN/A N/ANoneTABLE 3.6-17 JET IMPINGEMENT EFFECTS - REACTOR COOLANT SYSTEM - PRESSU RIZER CUBICLE PIPING Line DesignationBreak # (1)Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-151Rev. 30(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 inte nsity 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 ca use local barrier failure.(7) This postulated Longitudinal Sp lit (LS) results in fluid jet impingement on the Pressurizer Sp ray Line associated with anot her 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 accordan ce with FSAR Section 3.9B-10 results in a stress less than 3Sm. Compliance with this stress allo wable 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 Longi tudinal Splits Numbers 37 or 48 since pipe rupt ure restraints on the Pressurizer Sur ge Line, 3
-RCS-014-64-1, limit both axial and lateral displacement. Li miting the displacement limits the am ount of flow area exposed to le ss 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 el iminates the requirement to evaluate jet impingeme nt ef fects subsequent to Arbitrary Intermedi ate Breaks (AIBs). Therefore, jet impingement effects are not ev aluated by AIBs on the Pressurizer Safety Lines.(10) This postulated break results in a Large LOCA resulting in a release of saturated steam. Th e NSSS System Standard Design Criteria does not permit propaga tion from a steam phase to liqui d phase LOCA. Liquid phase LOCA would occur if the Pressurizer Spray Line ruptures subsequent to fluid je t impingement from the Pressurizer Relief Line. This particular interaction is envelo ped by the postulated Longitudinal Splits (LS) Numbers 49 through 52 on the Pressurizer Surge Li ne, 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.
MPS3 UFSARMPS3 UFSAR3.6-152Rev. 30TABLE 3.6-18 POSTULATED PIPE BREAKS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFE TY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING Line DesignationBreak #BuildingElevationBreak Type (1)Threshold Stress Eqn 10 & Either Eqn 12 or 13Usage FactorFigure 3.6-3-RCS-010-122-110Containment18'-5"CBN/A Terminal End 13 3-RCS-010-122-1SplitContainment18'-9"LS (2)13 3-SIL-010-44-211Containment-22'-0"CBN/A Terminal End13 3-SIL-010-45-12Containment11'-10"CB & LSAbove Threshold13 3-SIL-010-45-13Containment17'-3"CB & LSAbove Threshold13 3-SIL-006-139-1TPContainment13'-7"CBN/A Terminal End13 3-RCS-010-132-110Containment18'-3"CBN/A Terminal End13 3-RCS-010-132-1SplitContainment18'-9"LS (2)13 3-SIL-010-46-211Containment-22'-0"CBN/A Terminal End13 3-SIL-010-47-12Containment11'-10"CB & LSAbove Threshold13 3-SIL-010-47-13Containment17'-3"CB & LSAbove Threshold13 3-SIL-006-140-1TPContainment13'-7"CBN/A Terminal End13 3-RCS-010-138-110Containment18'-1"CBN/A Terminal End13 3-RCS-010-138-1SplitContainment18'-9"LS (2)13 3-SIL-010-48-211Containment-22'-0"CBN/A Terminal End13 3-SIL-010-49-12Containment11'-10"CB & LSAbove Threshold13 3-SIL-010-49-13Containment17'-3"CB & LSAbove Threshold13 3-SIL-006-145-1TPContainment13'-7"CBN/A Terminal End13 MPS3 UFSARMPS3 UFSAR3.6-153Rev. 303-RCS-010-146-110Containment18'-5"CBN/A Terminal End 13 3-RCS-010-146-1SplitContainment18'-9"LS (2)13 3-SIL-010-50-211Containment-22'-0"CBN/A Terminal End13 3-SIL-010-51-12Containment11'-10"CB & LSAbove Threshold13 3-SIL-010-51-13Containment17'-3"CB & LSAbove Threshold13 3-SIL-006-146-1TPContainment13'-7"CBN/A Terminal End13 3-RCS-012-123-19Containment15'-4"CBN/A Terminal End13 3-RCS-012-123-14Containment13'-7"CB & LSAbove Threshold13 3-RCS-012-123-15Containment13'-7"CB & LSAbove Threshold13 3-RCS-012-123-16Containment13'-7"CB & LSAbove Threshold13 3-RCS-012-123-17Containment13'-7"CB N/A Terminal End13 3-RCS-006-124-18Containment9'-6"CBN/A Terminal End13 3-RCS-012-103-19Containment15'-4"CBN/A Terminal End13 3-RCS-012-103-14Containment13'-7"CB & LSAbove Threshold13 3-RCS-012-103-15Containment13'-7"CB & LSAbove Threshold13 3-RCS-012-103-16Containment13'-7"CB & LSAbove Threshold13 3-RCS-012-103-17Containment13'-7"CBN/A Terminal End13 3-RCS-006-140-18Containment9'-6"CBN/A Terminal End13 TABLE 3.6-18 POSTULATED PIPE BREAKS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFE TY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)Line DesignationBreak #BuildingElevationBreak Type (1)Threshold Stress Eqn 10 & Either Eqn 12 or 13Usage FactorFigure 3.6-MPS3 UFSARMPS3 UFSAR3.6-154Rev. 303-RCS-006-119-1TPContainment19'-1"CB N/A (3)N/A (3)13 3-RCS-006-119-1IPContainment20'-10"CB & LS N/A (3)N/A (3)13 3-RCS-006-119-1TPContainment20'-10"CB N/A (3)N/A (3)13 3-RCS-006-120-1TPContainment19'-1"CB N/A (3)N/A (3)13 3-RCS-006-120-1IPContainment20'-10"CB & LS N/A (3)N/A (3)13 3-RCS-006-120-1TPContainment20'-10"CB N/A (3)N/A (3)13 3-SIH-004-16-212Auxiliary12'-6"CBN/A Terminal End (4)3-SIH-004-22-213Auxiliary12'-6"CBN/A Terminal End (4)3-SIH-004-16-214Auxiliary12'-6"CBN/A - Arbitrary Intermediate (4)3-SIH-004-16-215Auxiliary12'-6"CBN/A - Arbitrary Intermediate (4)3-SIH-004-16-216Auxiliary18'-5"CBN/A Terminal End (4)3-RCS-003-133-1TPContainment17'-6"CB N/A (5)N/A (5)Not Shown3-RCS-003-133-1IPContainment17'-6"CB N/A (5)N/A (5)Not Shown3-RCS-150-221-1IPContainment17'-6"CB N/A (5)N/A (5)Not Shown3-RCS-150-221-1TPContainment17'-6"CB N/A (5)N/A (5)Not Shown3-RCS-003-139-1TPContainment17'-6"CB N/A (5)N/A (5)Not ShownTABLE 3.6-18 POSTULATED PIPE BREAKS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFE TY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)Line DesignationBreak #BuildingElevationBreak Type (1)Threshold Stress Eqn 10 & Either Eqn 12 or 13Usage FactorFigure 3.6-MPS3 UFSARMPS3 UFSAR3.6-155Rev. 30NOTES:(1) Circumferential Pipe Break (CB) and Longi tudinal 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 Accu mulator Dischar ge Line connecti on to each Cold Leg. A detailed stress analysis was performed as de scribed in FSAR Section 3.6.2.1.3.1 a nd the stress analysis indicates that a LS is preferential at this location3-RCS-003-139-1IPContainment17'-6"CB N/A (5)N/A (5)Not Shown3-RCS-150-222-1IPContainment17'-6"CB N/A (5)N/A (5)Not Shown3-RCS-150-222-1TPContainment17'-6"CB N/A (5)N/A (5)Not Shown3-RCS-003-147-1TPContainment17'-6"CB N/A (5)N/A (5)Not Shown3-RCS-003-147-1IPContainment16'-6"CB N/A (5)N/A (5)Not Shown3-RCS-150-223-1IPContainment9'-10"CB N/A (5)N/A (5)Not Shown3-RCS-150-223-1TPContainment9'-10"CB N/A (5)N/A (5)Not Shown3-RCS-003-121-1TPContainment17'-6"CB N/A (5)N/A (5)Not Shown3-RCS-003-121-1IPContainment17'-6"CB N/A (5)N/A (5)Not Shown3-RCS-150-220-1IPContainment16'-3"CB N/A (5)N/A (5)Not Shown3-RCS-150-220-1TPContainment16'-3"CB N/A (5)N/A (5)Not ShownTABLE 3.6-18 POSTULATED PIPE BREAKS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFE TY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)Line DesignationBreak #BuildingElevationBreak Type (1)Threshold Stress Eqn 10 & Either Eqn 12 or 13Usage FactorFigure 3.6-MPS3 UFSARMPS3 UFSAR3.6-156Rev. 30(3) Breaks are postulated at any location (I P) along the Low Pressure Safety Injection System to the Hot Leg on Loops B & C. Bo th 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 line s remain pressurized duri ng Normal Plant Operating Conditions and are reviewed for postulated pipe break.(5) Breaks are postulated at any location (IP) along the High Pressure Safety Inject ion System into each Cold Leg. Only Circumferential Breaks (CB) are postulated since the High Pressure Safety Injecti on 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 UFSARMPS3 UFSAR3.6-157Rev. 30TABLE 3.6-19 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING Line DesignationBreak #Blowdown SourceEssential Pipe Whip TargetProtection Requirement3-RCS-010-122-110 3-RCS-275-5-1 3-RCS-008-24-1 (1)3RCS-PRR6A3-RCS-010-122-1Split3-RCS-275-5-1NoneNone3-RCS-010-44-1113SIL*TK1ANoneNone 3-SIL-010-45-123SIL*TK1ANoneNone 3-SIL-010-45-1Split3SIL*TK1A (2)3SIL-PRR4A3-SIL-010-45-133SIL*TK1ANoneNone 3-SIL-010-45-1Split3SIL*TK1A (2)3SIL-PRR4A3-SIL-006-139-1TP3SIL*TK1A (2)3SIL-PRR4A3-RCS-010-132-1103-RCS-275-10-1 3-RCS-008-30-1 (1)3RCS-PRR6B3-RCS-010-132-1Split3-RCS-275-10-1NoneNone 3-RCS-010-46-1113SIL*TK1BNoneNone 3-SIL-010-47-123SIL*TK1BNoneNone 3-SIL-010-47-1Split3SIL*TK1B (2)3SIL-PRR4B3-SIL-010-47-133SIL*TK1BNoneNone 3-SIL-010-47-1Split3SIL*TK1B (2)3SIL-PRR4B3-SIL-006-140-1TP3SIL*TK1B (2)3SIL-PRR4B3-RCS-010-138-1103-RCS-275-15-1 3-RCS-008-35-1 (1)3RCS-PRR6C3-RCS-010-138-1Split3-RCS-275-15-1NoneNone MPS3 UFSARMPS3 UFSAR3.6-158Rev. 303-RCS-010-48-111 3SIL*TK1C None None 3-SIL-010-49-1 2 3SIL*TK1CNone None3-SIL-010-49-1Split 3SIL*TK1C (2)3SIL-PRR4C3-SIL-010-49-133SIL*TK1CNoneNone3-SIL-010-49-1Split3SIL*TK1C (2)3SIL-PRR4C3-SIL-006-145-1TP3SIL*TK1C (2)3SIL-PRR4C3-RCS-010-146-1103-RCS-275-20-1 3-RCS-008-40-1 (1)3RCS-PRR6D3-RCS-010-146-1Split3-RCS-275-20-1NoneNone 3-RCS-010-50-1113SIL*TK1DNoneNone 3-SIL-010-51-123SIL*TK1DNoneNone 3-SIL-010-51-1Split3SIL*TK1D (2)3SIL-PRR4D3-SIL-010-51-133SIL*TK1DNoneNone 3-SIL-010-51-1Split3SIL*TK1D (2)3SIL-PRR4D3-SIL-006-146-1TP3SIL*TK1D (2)3SIL-PRR4D3-RCS-006-119-1TP3-RCS-029-6-1NoneNone 3-RCS-006-119-1IP3-RCS-029-6-1NoneNone 3-RCS-006-119-1TP3-RCS-029-6-1NoneNone 3-RCS-006-120-1TP3-RCS-029-11-1NoneNoneTABLE 3.6-19 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)Line DesignationBreak #Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-159Rev. 303-RCS-006-120-1IP3-RCS-029-11-1 None None3-RCS-006-120-1TP3-RCS-029-11-1 None None3-RCS-003-133-1TP3-RCS-275-10-1None None3-RCS-003-133-1IP 3-RCS-275-10-1None None3-RCS-150-221-1IP 3-RCS-275-10-1None None3-RCS-150-221-1TP3-RCS-275-10-1None None3-RCS-003-139-1TP3-RCS-275-15-1None None3-RCS-003-139-1IP 3-RCS-275-15-1None None3-RCS-150-222-1IP 3-RCS-275-15-1None None3-RCS-150-222-1TP3-RCS-275-15-1None None3-RCS-003-147-1TP3-RCS-275-20-1None None3-RCS-003-147-1IP 3-RCS-275-20-1None None3-RCS-150-223-1IP 3-RCS-275-20-1None None3-RCS-150-223-1TP3-RCS-275-20-1None None3-RCS-003-121-1TP 3-RCS-275-5-1None None3-RCS-003-121-1IP 3-RCS-275-5-1None None3-RCS-150-220-1IP 3-RCS-275-5-1None None3-RCS-150-220-1TP 3-RCS-275-5-1None NoneTABLE 3.6-19 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)Line DesignationBreak #Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-160Rev. 303-RCS-012-123-193-RCS-029-1-1 Containment Liner (3)3RHS-PRS1A3-RCS-012-123-143-RCS-029-1-1 Containment Liner (3)3RHS-PRS2A3-RCS-012-123-153-RCS-029-1-1 Containment Liner (3)3RHS-PRS2A3-RCS-012-123-163-RCS-029-1-1 Containment Liner (3)3RHS-PRS2A3-RCS-012-123-173-RCS-029-1-1 Containment Liner (3)3RHS-PRS2A3-RCS-012-123-1Splits3-RCS-029-1-1NoneNone3-RCS-006-124-183-RCS-029-1-1East-West Cubicle Wall (4)3RCS-PRR8613-RCS-012-103-193-RCS-029-16-1 Containment Liner (3)3RHS-PRS1D3-RCS-012-103-143-RCS-029-16-1 Containment Liner (3)3RHS-PRS2D3-RCS-012-103-153-RCS-029-16-1 Containment Liner (3)3RHS-PRS2D3-RCS-012-103-163-RCS-029-16-1 Containment Liner (3)3RHS-PRS2D3-RCS-012-103-173-RCS-029-16-1 Containment Liner (3)3RHS-PRS2D3-RCS-012-103-1Splits3-RCS-029-16-1NoneNone 3-RCS-006-124-183-RCS-029-16-1East-West Cubicle Wall (4)3RCS-PRR8623-SIH-004-16-212 None (5)None (6)None3-SIH-004-22-213 None (5)None (6)NoneTABLE 3.6-19 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)Line DesignationBreak #Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-161Rev. 30NOTES:(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 aff ected loop cannot exceed 20% of the flow ar ea of the broken line. The Reactor Coolant System Loop Stop Valve Bypass Line identified has a nominal pipe si ze of 8" and therefore, if the 8" Loop Stop Valve Bypass Lin e is impacted pressure boundary integrity must be as sured. The pipe rupture restraint indicated prevents the impact.(2) The Terminal End (TP) break or Longit udinal 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. Th e requirement is that a Non-LOCA cannot propagate into a LOCA. If unrestrained, the broken pipe would induc e significant loading upon the Reactor Coolant System pressure boundary. The pipe rupture restraint is provided to prevent this effect.(3) A break on the Reactor Coolant System Residual Heat Remova l Piping is a Large LOCA. The NSSS System Standard design Criteria provides the acceptance criteria for Large LOCA breaks. Th e main criterion or th is break is to ensure that Containment Leak Tight Integrity is maintained. The pipe whip of the Reactor Coolant System Residual Heat Removal Pi ping 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 Cubi cle Wall would cause catastrophic failu re of the wall. Therefore, pi pe rupture restraints are provide d to prevent this impact.3-SIH-004-16-214 None (5)None (7)None3-SIH-004-16-215 None (5)None (7)None3-SIH-004-16-216 None (5)None (6)NoneTABLE 3.6-19 PIPE WHIP EFFECTS - REACTOR COOLANT SYSTEM - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)Line DesignationBreak #Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-162Rev. 30(5) The High Pressure Safety Injection (SIH) System in the Auxiliary Building is a High Energy System since the piping is maint ained pressurized during Normal Plant Operation between the Normal Charging Pumps and th e 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 a nd 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 Inje ction (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 br eak 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*M V8801A and 3SIH*MV8801B and the Reactor Coolant System Pressure Boundary is normally empty.(7) USNRC Generic Letter 87-11 el iminates 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 UFSARMPS3 UFSAR3.6-163Rev. 30TABLE 3.6-20 JET IMPINGEMENT EFFECTS - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING Line DesignationBreak #
Essential Jet Impingement TargetDistance to TargetJet Intensity at the Target Jet Load on the TargetProtection Requirement3-RCS-010-122-110NoneN/AN/AN/ANone3-RCS-010-122-1SplitNoneN/AN/AN/ANone 3-RCS-010-44-111NoneN/AN/AN/ANone3-SIL-010-45-12NoneN/AN/AN/ANone3-SIL-010-45-1SplitNoneN/AN/AN/ANone 3-SIL-010-45-13NoneN/AN/AN/ANone3-SIL-010-45-1SplitNoneN/AN/AN/ANone3-SIL-006-139-1TPNoneN/AN/AN/ANone 3-RCS-010-132-110NoneN/AN/AN/ANone3-RCS-010-132-1SplitNoneN/AN/AN/ANone3-RCS-010-46-111NoneN/AN/AN/ANone 3-SIL-010-47-12NoneN/AN/AN/ANone3-SIL-010-47-1SplitNoneN/AN/AN/ANone3-SIL-010-47-13NoneN/AN/AN/ANone 3-SIL-010-47-1SplitNoneN/AN/AN/ANone3-SIL-006-140-1TPNoneN/AN/AN/ANone3-RCS-010-138-110NoneN/AN/AN/ANone 3-RCS-010-138-1SplitNoneN/AN/AN/ANone3-RCS-010-48-111NoneN/AN/AN/ANone MPS3 UFSARMPS3 UFSAR3.6-164Rev. 303-SIL-010-49-12 None N/AN/AN/ANone3-SIL-010-49-1Split None N/AN/AN/ANone3-SIL-010-49-13 None N/AN/AN/ANone3-SIL-010-49-1Split None N/AN/AN/ANone3-SIL-006-145-1TPNoneN/AN/AN/A None3-RCS-010-146-110NoneN/AN/AN/A None3-RCS-010-146-1Split None N/AN/AN/ANone3-RCS-010-50-111NoneN/AN/AN/A None3-SIL-010-51-12 None N/AN/AN/ANone3-SIL-010-51-1Split None N/AN/AN/ANone3-SIL-010-51-13 None N/AN/AN/ANone3-SIL-010-51-1Split None N/AN/AN/ANone3-SIL-006-146-1TPNoneN/AN/AN/A None3-RCS-006-119-1TPNoneN/AN/AN/A None3-RCS-006-119-1IPNoneN/AN/AN/A None3-RCS-006-119-1TPNoneN/AN/AN/A None3-RCS-006-120-1TPNoneN/AN/AN/A None3-RCS-006-120-1IPNoneN/AN/AN/A None3-RCS-006-120-1TPNoneN/AN/AN/A NoneTABLE 3.6-20 JET IMPINGEMENT EFFECTS - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)Line DesignationBreak #
Essential Jet Impingement TargetDistance to TargetJet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-165Rev. 303-RCS-003-133-1TPNoneN/AN/AN/A None3-RCS-003-133-1IP (1)(1)None3-RCS-150-221-1IP3RCS*V984 and 3RCS*V985 (2)None3-RCS-150-221-1TP3RCS*V984 and 3RCS*V985 (2)None3-RCS-003-139-1TPNoneN/AN/AN/ANone3-RCS-003-139-1IPNoneN/AN/AN/ANone 3-RCS-150-222-1IPNoneN/AN/AN/ANone3-RCS-150-222-1TP3RCS*V979 (2)None3-RCS-003-147-1TPNoneN/AN/AN/ANone3-RCS-003-147-1IPNoneN/AN/AN/ANone3-RCS-150-223-1IPNoneN/AN/AN/ANone3-RCS-150-223-1TPNoneN/AN/AN/ANone 3-RCS-003-121-1TPNoneN/AN/AN/ANone3-RCS-003-121-1IPNoneN/AN/AN/ANone3-RCS-150-220-1IPNoneN/AN/AN/ANone 3-RCS-150-220-1TPNoneN/AN/AN/ANone3-RCS-012-103-19NoneN/AN/AN/ANone3-RCS-012-103-143-SSR-375-026-2 (3)None3-RCS-012-103-153-SSR-375-026-2 (3)NoneTABLE 3.6-20 JET IMPINGEMENT EFFECTS - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)Line DesignationBreak #
Essential Jet Impingement TargetDistance to TargetJet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-166Rev. 303-RCS-012-103-163-SSR-375-026-2 (3)None3-RCS-012-103-173-SSR-375-026-2 (3)None3-RCS-012-103-1Splits3-SSR-375-026-2 (3)None3-RCS-006-124-18NoneN/AN/AN/ANone3-RCS-012-123-19NoneN/AN/AN/ANone3-RCS-012-123-143-SSR-375-035-2 (3)None3-RCS-012-123-153-SSR-375-035-2 (3)None3-RCS-012-123-163-SSR-375-035-2 (3)None3-RCS-012-123-173-SSR-375-035-2 (3)None3-RCS-012-123-1Splits3-SSR-375-035-2 (3)None3-RCS-006-124-18NoneN/AN/AN/ANone 3-SIH-004-16-212 None (4)N/AN/AN/A None (5)3-SIH-004-22-213 None (4)N/AN/AN/A None (5)3-SIH-004-16-214 None (6)N/AN/AN/A None (5)3-SIH-004-16-215 None (6)N/AN/AN/A None (5)3-SIH-004-16-216 None (4)N/AN/AN/A None (5)TABLE 3.6-20 JET IMPINGEMENT EFFECTS - LOW PRESSURE SAFETY INJECTION, HIGH PRESSURE SAFETY INJECTION, AND RESIDUAL HEAT REMOVAL PIPING (CONTINUED)Line DesignationBreak #
Essential Jet Impingement TargetDistance to TargetJet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-167Rev. 30NOTES:(1) The NSSS System Standard Design Criteria states for a Small LOCA that break propagation to the unaffected leg of the affect ed 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 Le
: g. Loss of pressure boundary is admitted. In addition to this target, Disc Pressurization Va lves 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 i s 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 th at 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 C ooling System (ECCS) performa nce will be basically the same as it would with no propagation. Therefore, calculated Emergenc y Core Cooling System (ECCS) performa nce conservatively bounds Case 3.3 in NES-40190 and the interactions are acc eptable. Therefore, no protec tive hardware is required.(2) The NSSS System Standard Design Criteria states for a Smal l 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 ad mitted. This interaction was transmitted to the NSSS Vendor via NES-40190, dated November 27, 1985. A review was performed by the NSSS Vendor conc erning this interaction and th e results are transmitted via NEU-6039, dated December 23, 1985. This conditio n is identified as Case 3.3 in NE S-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. The break propagation areas ar e small compared to the break area of the rupt ured line. Phenomena associated with the brea k at the original location are not af fected by the propagation and calculated Emergency Core Cooling System (ECCS) performance will be basically the same as it would with no propagation. Therefore, calculated Emergenc y Core Cooling System (ECCS) performance conservatively bounds Case 3.3 in NES-40190 a nd the interaction is acceptable. Therefor e, 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 Generato r Blowdown Sampling System. Lo ss 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 result s 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 Vendor's conclusion is that heat transfer to the secondary side is unimportant for 6: and larger LOCA breaks. For Millstone 3, the li miting case is the 4" equiv alent 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 Feed water injection begins, followi ng an "S" signal mass addition t o
MPS3 UFSARMPS3 UFSAR3.6-168Rev. 30the Steam Generator secondary will exceed the mass depletion through the broken Steam Generator Blowdown Sampling line. Since the Steam Gene rator 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 Samp ling System. Therefore this break is les s limiting than the 4" equivalent diameter cold leg break and the interaction is acce ptable. 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 sin ce 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 s ubsequent 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 Pre ssure 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 maint ained pressurized during Normal Plant Operation between the Normal Charging Pumps and th e 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 a nd the Reactor Coolant System Pressure Boundary is normally empty.(6) USNRC Generic Letter 87-11 eliminates the requirement to evaluate the dynamic ef fects (i.e., fluid jet impingement) of Arbi tr ary Intermediate Breaks (AIBs). Therefore, fluid jet impingement effects subsequent to AIBs are not evaluated.
MPS3 UFSARMPS3 UFSAR3.6-169Rev. 30TABLE 3.6-21 POSTULATED BREAKS - CHEMICAL VOLUME CONTROL SYSTEM - NORMAL CHARGING Line DesignationBreak #
(1)BuildingElevationBreak Type (2)Total Additive StressFigure 3.63-CHS-003-662-210Containment-3'-6"CBN/A Terminal End15 3-CHS-003-661-220Containment-3'-6"CBN/A Terminal End15 3-CHS-003-662-221Containment-3'-6"CBN/A Arbitrary Intermediate15 3-CHS-003-662-222Containment-3'-6"CBN/A Arbitrary Intermediate15 3-CHS-003-662-223Containment-5'-10"CBN/A Terminal End15 3-CHS-002-73-224Containment-10'-1"CBAbove Threshold15 3-CHS-002-73-225Containment-10'-1"CBN/A Terminal End15 3-CHS-003-661-226Containment-2'-0"CBAbove Threshold15 3-CHS-003-661-227Containment-2'-0"CBN/A Arbitrary Intermediate15 3-CHS-003-661-228Containment-2'-0"CBN/A Terminal End15 3-CHS-003-661-229Containment-10'-2"CBAbove Threshold15 3-CHS-003-76-230Containment-9'-9"CBN/A Terminal End15 3-CHS-003-72-231Containment-1'-5"CBN/A Terminal End15 3-CHS-003-72-232Containment7'-6"CBN/A Arbitrary Intermediate15 3-CHS-003-72-233Containment11'-3"CBN/A Arbitrary Intermediate15 3-CHS-003-70-234Auxiliary12'-6"CBN/A Arbitrary Intermediate15 3-CHS-003-70-235Auxiliary15'-4"CBN/A Arbitrary Intermediate15 3-CHS-003-69-236Auxiliary16'-11"CBN/A Terminal End15 3-CHS-002-237-237Auxiliary18' - 8"CBN/A Terminal End 15 MPS3 UFSARMPS3 UFSAR3.6-170Rev. 303-CHS-002-237-238Auxiliary20'-1"CBN/A Arbitrary Intermediate15 3-CHS-002-237-239Auxiliary15'-10"CBN/A Arbitrary Intermediate15 3-CHS-002-238-240Auxiliary12'-3"CBN/A Terminal End15 3-CHS-002-238-241Auxiliary13'-0"CBN/A Arbitrary Intermediate15 3-CHS-002-238-242Auxiliary18'-7"CBN/A Arbitrary Intermediate15 3-CHS-003-72-243Containment12'-6"CBN/A Terminal End15 3-CHS-003-72-244Auxiliary12'-6"CBN/A Terminal End15
 
3-CHS-004-68-2 (3)AuxiliaryVariesCB and LSN/A (3)15 3-CHS-004-435-2 (3)AuxiliaryVariesCB and LSN/A (3)15 3-CHS-003-434-2 (3)AuxiliaryVariesCBN/A (3)15 3-CHS-004-28-2 (3)AuxiliaryVariesCB and LSN/A (3)15 3-CHS-004-30-2 (3)AuxiliaryVariesCB and LSN/A (3)15 3-CHS-004-29-2 (3)AuxiliaryVariesCB and LSN/A (3)15 3-CHS-003-65-2 (3)AuxiliaryVariesCBN/A (3)15 3-CHS-004-433-2 (3)AuxiliaryVariesCB and LSN/A (3)15 3-CHS-004-682-2 (3)AuxiliaryVariesCB and LSN/A (3)15 3-CHS-004-678-2 (3)AuxiliaryVariesCB and LSN/A (3)15 3-CHS-004-677-2 (3)AuxiliaryVariesCB and LSN/A (3)15 TABLE 3.6-21 POSTULATED BREAKS - CHEMICAL VOLUME CONTROL SYSTEM - NORMAL CHARGING Line DesignationBreak #
(1)BuildingElevationBreak Type (2)Total Additive StressFigure 3.6 MPS3 UFSARMPS3 UFSAR3.6-171Rev. 30NOTES:(1) A portion of the Chemical & Volume Cont rol System Normal Charging is part of the Reactor Coolant System. Breaks 1 through 9 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 Longi tudinal Pipe Split (LS) are defined in FSAR Section 3.6.2.1.3
.(3) Pipe breaks on the Discharge Lines of the Normal Charging Sy stem in each Char ging 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-CHS-004-679-2 (3)AuxiliaryVariesCB and LSN/A (3)15 3-CHS-003-67-2 (3)AuxiliaryVariesCBN/A (3)15 3-CHS-003-31-2 (3)AuxiliaryVariesCBN/A (3)15 3-CHS-003-32-2 (3)AuxiliaryVariesCBN/A (3)15 3-CHS-002-61-2 (3)AuxiliaryVariesCBN/A (3)15 3-CHS-002-59-2 (3)AuxiliaryVariesCBN/A (3)15 3-CHS-002-63-2 (3)AuxiliaryVariesCBN/A (3)15 TABLE 3.6-21 POSTULATED BREAKS - CHEMICAL VOLUME CONTROL SYSTEM - NORMAL CHARGING Line DesignationBreak #
(1)BuildingElevationBreak Type (2)Total Additive StressFigure 3.6 MPS3 UFSARMPS3 UFSAR3.6-172Rev. 30TABLE 3.6-22 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTRO L SYSTEM - NORMAL CHARGING Line DesignationBreak #Blowdown SourceEssential Pipe Whip TargetProtection Requirement3-CHS-003-662-210 None (1)None None3-CHS-003-661-220 None (1)None None3-CHS-003-662-221 None (1)None (2)None3-CHS-003-662-222 None (1)None (2)None3-CHS-003-662-223 None (1)None None 3-CHS-002-73-2 24 None (1)None None 3-CHS-002-73-2 25 None (1)None None3-CHS-003-661-226 None (1)None None3-CHS-003-661-227 None (1)None (2)None3-CHS-003-661-228 None (1)None None3-CHS-003-661-229 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)None3-CHS-003-72-233 None (1)None (2)None3-CHS-003-70-234 None (1)None (2)None MPS3 UFSARMPS3 UFSAR3.6-173Rev. 30 3-CHS-003-70-2 35 None (1)None (2)None3-CHS-003-69-236 None (1)None None3-CHS-002-237-237 None (1)None None3-CHS-002-237-238 None (1)None (2)None3-CHS-002-237-239 None (1)None (2)None3-CHS-002-238-240 None (1)None None3-CHS-002-238-241 None (1)None (2)None3-CHS-002-238-242 None (1)None (2)None3-CHS-003-72-243 None (1)None None 3-CHS-003-72-2 44 None (1)None None 3-CHS-004-68-2 (3)None (1)None (4)None 3-CHS-004-435-2 (3)None (1)None (4)None 3-CHS-003-434-2 (3)None (1)None (4)None 3-CHS-004-28-2 (3)None (1)None (4)None 3-CHS-004-30-2 (3)None (1)None (4)None 3-CHS-004-29-2 (3)None (1)None (4)None 3-CHS-003-65-2 (3)None (1)None (4)NoneTABLE 3.6-22 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTRO L SYSTEM - NORMAL CHARGING Line DesignationBreak #Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-174Rev. 30NOTES:(1) Sustained blowdown on the Chemical & Volume Control System Normal Char ging 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 th e requirement to evaluate the pipe whip ef fects 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 Sy stem in each Char ging Pump Cubicle are postulated at Terminal Ends (TPs), and at each Valve, Fitting, and any Integral Welded Attachments in accordance with th e criteria delineated in FSAR Section 3.6.2.1.2.3.b.
3-CHS-004-433-2 (3)None (1)None (4)None 3-CHS-004-682-2 (3)None (1)None (4)None 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 (1)None (4)None 3-CHS-003-67-2 (3)None (1)None (4)None 3-CHS-003-31-2 (3)None (1)None (4)None 3-CHS-003-32-2 (3)None (1)None (4)None 3-CHS-002-61-2 (3)None (1)None (4)None 3-CHS-002-59-2 (3)None (1)None (4)None 3-CHS-002-63-2 (3)None (1)None (4)NoneTABLE 3.6-22 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTRO L SYSTEM - NORMAL CHARGING Line DesignationBreak #Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-175Rev. 30(4) NERM-069 Revision 1, Attachment 5, Figure 10A shows the location of the Charging Pumps. NE RM-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 UFSARMPS3 UFSAR3.6-176Rev. 30TABLE 3.6-23 JET IMPINGEMENT EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - NORMAL CHARGING Line DesignationBreak #Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection  Requirement3-CHS-003-662-210 None (1)N/AN/A N/ANone3-CHS-003-661-220 None (1)N/AN/A N/ANone3-CHS-003-662-221 None (1)N/AN/A N/A None (2)3-CHS-003-662-222 None (1)N/AN/A N/A None (2)3-CHS-003-662-223 None (1)N/AN/A N/ANone3-CHS-002-73-224 None (1)N/AN/A N/ANone3-CHS-002-73-225 None (1)N/AN/A N/ANone3-CHS-003-661-226 None (1)N/AN/A N/ANone3-CHS-003-661-227 None (1)N/AN/A N/A None (2)3-CHS-003-661-228 None (1)N/AN/A N/ANone3-CHS-003-661-229 None (1)N/AN/A N/ANone3-CHS-003-76-230 None (1)N/AN/A N/ANone3-CHS-003-72-231 None (1)N/AN/A N/ANone3-CHS-003-72-232 None (1)N/AN/A N/A None (2)3-CHS-003-72-233 None (1)N/AN/A N/A None (2)3-CHS-003-70-234 None (1)N/AN/A N/A None (2)3-CHS-003-70-235 None (1)N/AN/A N/A None (2)
MPS3 UFSARMPS3 UFSAR3.6-177Rev. 303-CHS-003-69-236 None (1)N/AN/A N/ANone3-CHS-002-237-237 None (1)N/AN/A N/ANone3-CHS-002-237-238 None (1)N/AN/A N/A None (2)3-CHS-002-237-239 None (1)N/AN/A N/A None (2)3-CHS-002-238-240 None (1)N/AN/A N/ANone3-CHS-002-238-241 None (1)N/AN/A N/A None (2)3-CHS-002-238-242 None (1)N/AN/A N/A None (2)3-CHS-003-72-243 None (1)N/AN/A N/ANone3-CHS-003-72-244 None (1)N/AN/A N/ANone 3-CHS-004-68-2 (3)None (1)N/AN/A N/ANone 3-CHS-004-435-2 (3)None (1)N/AN/A N/ANone 3-CHS-003-434-2 (3)None (1)N/AN/A N/ANone 3-CHS-004-28-2 (3)None (1)N/AN/A N/ANone 3-CHS-004-30-2 (3)None (1)N/AN/A N/ANone 3-CHS-004-29-2 (3)None (1)N/AN/A N/ANone 3-CHS-003-65-2 (3)None (1)N/AN/A N/ANone 3-CHS-004-433-2 (3)None (1)N/AN/A N/ANoneTABLE 3.6-23 JET IMPINGEMENT EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - NORMAL CHARGING Line DesignationBreak #Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection  Requirement MPS3 UFSARMPS3 UFSAR3.6-178Rev. 30NOTES:(1) Sustained blowdown on the Chemical & Volu me 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 C ontrol System/Reactor Coolant Sy stem (i.e., Class 1/Cl ass 2) boundary preven t backflow from the Reactor Coolant System and the Charging Pu mps have limited capacity and cannot sust ain the system pressu re. Since sustained blowdown does not occur, dynamic analyses for the effects of fluid jet impi ngement is not performed based upon FSAR Section 3.6.2.2.1 , Item 3.(2) USNRC Generic Letter 87-11 eliminates th e need to evaluate jet impingement effects subsequent to Arbitrary Intermediate Breaks (AIBs). Therefore jet impingement effects are not evaluated for AIBs.
3-CHS-004-682-2 (3)None (1)N/AN/A N/ANone 3-CHS-004-678-2 (3)None (1)N/AN/A N/ANone 3-CHS-004-677-2 (3)None (1)N/AN/A N/ANone 3-CHS-004-679-2 (3)None (1)N/AN/A N/ANone 3-CHS-003-67-2 (3)None (1)N/AN/A N/ANone 3-CHS-003-31-2 (3)None (1)N/AN/A N/ANone 3-CHS-003-32-2 (3)None (1)N/AN/A N/ANone 3-CHS-002-61-2 (3)None (1)N/AN/A N/ANone 3-CHS-002-59-2 (3)None (1)N/AN/A N/ANone 3-CHS-002-63-2 (3)None (1)N/AN/A N/ANoneTABLE 3.6-23 JET IMPINGEMENT EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - NORMAL CHARGING Line DesignationBreak #Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection  Requirement MPS3 UFSARMPS3 UFSAR3.6-179Rev. 30(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 th e criteria delineated in FSAR Section 3.6.2.1.2.3.b.
MPS3 UFSARMPS3 UFSAR3.6-180Rev. 30TABLE 3.6-24 POSTULATED BREAKS - CHEMICAL VOLUME CONTROL SYSTEM - LETDOWN LINE Line DesignationBreak #
(1)BuildingElevationBreak Type (2)Total Additive StressFigure 3.6-3-CHS-003-1-29Containment-9'-9"CBN/A Terminal End16 3-CHS-003-416-210Containment-1'-5"CBN/A Terminal End16 3-CHS-003-416-211Containment4'-7"CBAbove Threshold16 3-CHS-003-416-212Containment4'-7"CBAbove Threshold16 3-CHS-003-416-213Containment3'-7"CBAbove Threshold16 3-CHS-002-2-214Containment3'-3"CBAbove Threshold16 3-CHS-002-590-215Containment0'-8"CBN/A Terminal End16 3-CHS-002-7-216Containment0'-8"CBN/A Terminal End16 3-CHS-002-4-217Containment4'-4"CBAbove Threshold16 3-CHS-002-589-218Containment0'-8"CBN/A Terminal End16 3-CHS-002-6-219Containment0'-8"CBN/A Terminal End16 3-CHS-002-6-220Containment0'-6"CBAbove Threshold16 3-CHS-002-3-221Containment4'-4"CBAbove Threshold16 3-CHS-002-588-222Containment0'-8"CBN/A Terminal End16 3-CHS-002-5-223Containment0'-8"CBN/A Terminal End16 3-CHS-002-5-224Containment0'-6"CBAbove Threshold16 3-CHS-025-304-225Containment7'-8"CBN/A Terminal End16 3-CHS-003-8-226Auxiliary Bldg.19'-9"CBN/A Arbitrary Intermediate16 3-CHS-003-8-227Auxiliary Bldg.20'-2"CBN/A Arbitrary Intermediate16 3-CHS-003-8-228Auxiliary Bldg.11'-2"CBN/A Terminal End16 MPS3 UFSARMPS3 UFSAR3.6-181Rev. 30 3-CHS-003-8-229Auxiliary Bldg.12'-7"CBN/A Arbitrary Intermediate16 3-CHS-003-78-230Auxiliary Bldg.32'-11"CBN/A Terminal End16 3-CHS-003-77-231Auxiliary Bldg.32'-11"CBN/A Terminal End16 3-CHS-003-9-232Auxiliary Bldg.22'-8"CBN/A Terminal End16 3-CHS-002-81-233Auxiliary Bldg.17'-3"CBN/A Arbitrary Intermediate16 3-CHS-002-81-234Auxiliary Bldg.16'-3"CBN/A Terminal End16 3-CHS-002-80-235Auxiliary Bldg.6'-0"CBN/A Arbitrary Intermediate16 3-CHS-002-80-236Auxiliary Bldg.6'-0"CBN/A Terminal End16 3-CHS-002-80-237Auxiliary Bldg.17'-2"CBN/A Arbitrary Intermediate16 3-CHS-002-80-238Auxiliary Bldg.20'-1"CBN/A Arbitrary Intermediate16 3-CHS-002-80-239Auxiliary Bldg.20'-1"CBN/A Terminal End16 3-CHS-002-80-240Auxiliary Bldg.12'-1"CBN/A Arbitrary Intermediate16 3-CHS-002-80-241Auxiliary Bldg.11'-10"CBN/A Arbitrary Intermediate16 3-CHS-002-80-242Auxiliary Bldg.-3'-8"CBN/A Terminal End16 3-CHS-002-80-243Auxiliary Bldg.9'-10"CBAbove Threshold16 3-CHS-002-80-244Auxiliary Bldg.10'-1"CBAbove Threshold16 3-CHS-002-80-245Auxiliary Bldg.13'-0"CBN/A Terminal End16 3-CHS-002-80-246Auxiliary Bldg.13'-10"CBN/A Arbitrary Intermediate16 3-CHS-002-80-247Auxiliary Bldg.14'-0"CBAbove Threshold16 3-CHS-002-80-248Auxiliary Bldg.14'-0"CBN/A Terminal End16 TABLE 3.6-24 POSTULATED BREAKS - CHEMICAL VOLUME CONTROL SYSTEM - LETDOWN LINE (CONTINUED)Line DesignationBreak #
(1)BuildingElevationBreak Type (2)Total Additive StressFigure 3.6-MPS3 UFSARMPS3 UFSAR3.6-182Rev. 30NOTES:(1) A portion of the Chemical & Volume Cont rol System Normal Letdown Line is part of the Reactor Coolant System. Breaks 1 through 8 are postulated on the Reactor Coolant System portio n of the Normal Letdown Line and are listed in FSAR Table 3.6-12.(2) Circumferential Pipe Brea k (CB) is defined in FSAR Section 3.6.2.1.3
.3-CHS-002-80-249Auxiliary Bldg.14'-0"CBAbove Threshold 16 3-CHS-002-80-250Auxiliary Bldg.14'-0"CBAbove Threshold 16 3-CHS-002-80-251Auxiliary Bldg.14'-0"CBAbove Threshold 16 3-CHS-002-80-252Auxiliary Bldg.13'-9"CBAbove Threshold 16 3-CHS-002-80-253Auxiliary Bldg.12'-10"CBN/A Terminal End 16 3-CHS-003-8-254Auxiliary Bldg.12'-6"CBN/A Terminal End 16 3-CHS-003-659-255Containment12'-6"CBN/A Terminal End 16 TABLE 3.6-24 POSTULATED BREAKS - CHEMICAL VOLUME CONTROL SYSTEM - LETDOWN LINE (CONTINUED)Line DesignationBreak #
(1)BuildingElevationBreak Type (2)Total Additive StressFigure 3.6-MPS3 UFSARMPS3 UFSAR3.6-183Rev. 30TABLE 3.6-25 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - LETDOWN LINE Line DesignationBreak #
(1)Blowdown SourceEssential Pipe Whip TargetProtection Requirement3-CHS-003-1-293-RCS-275-15-1 (2)3CHS-PRR858 & 3RCS-PRR870 3-CHS-003-416-2103-RCS-275-15-1None None 3-CHS-003-416-2113-RCS-275-15-1None None 3-CHS-002-3-2113-RCS-275-15-1None None 3-CHS-003-416-2123-RCS-275-15-1None None 3-CHS-002-4-2123-RCS-275-15-1None None 3-CHS-003-416-2133-RCS-275-15-1None None 3-CHS-002-2-2143-RCS-275-15-1None None 3-CHS-002-590-2153-RCS-275-15-1None None 3-CHS-002-7-2163-RCS-275-15-1None None 3-CHS-002-4-2173-RCS-275-15-1None None 3-CHS-002-589-2183-RCS-275-15-1None None 3-CHS-002-6-2193-RCS-275-15-1None None 3-CHS-002-6-2203-RCS-275-15-1None None 3-CHS-002-3-2213-RCS-275-15-1None None 3-CHS-002-588-2223-RCS-275-15-1None None 3-CHS-002-5-2233-RCS-275-15-1None None 3-CHS-002-5-2243-RCS-275-15-1None None 3-CHS-025-304-2253-RCS-275-15-1None None 3-CHS-003-8-2263-RCS-275-15-1 (3) (4)3CHS-PRR967 MPS3 UFSARMPS3 UFSAR3.6-184Rev. 30 3-CHS-003-8-2273-RCS-275-15-1 (4)None3-CHS-003-8-2283-RCS-275-15-1NoneNone3-CHS-003-8-2293-RCS-275-15-1 (4)None 3-CHS-003-78-2 30None None None 3-CHS-003-77-2313-RCS-275-15-1None None 3-CHS-003-9-2323-RCS-275-15-1None None 3-CHS-002-81-2 33None (4)None3-CHS-002-81-234NoneNoneNone 3-CHS-002-80-235None (4)None3-CHS-002-80-236NoneNoneNone 3-CHS-002-80-237None (4)None 3-CHS-002-80-2 38None (4)None3-CHS-002-80-239NoneNoneNone 3-CHS-002-80-240None (4)None 3-CHS-002-80-2 41None (4)None3-CHS-002-80-242NoneNoneNone 3-CHS-002-80-243NoneNoneNone
 
3-CHS-002-80-2 44None None None 3-CHS-002-80-2 45None None None 3-CHS-002-80-2 46None (4)NoneTABLE 3.6-25 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - LETDOWN LINE (CONTINUED)Line DesignationBreak #
(1)Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-185Rev. 30NOTES:(1) Repetition of Break Numbers is used to identify separate fl uid reservoirs which provide a c onstant pressure source to maint ain system pressure subsequent to the postula ted pipe break. These reservoirs maintain system blowdown during the transient event a s 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 Interm ediate Breaks (AIBs). Based upon this generic letter, pipe whip effects for these AIBs are not evaluated and no protection is necessary.
3-CHS-002-80-2 47None None None 3-CHS-002-80-2 48None None None 3-CHS-002-80-2 49None None None 3-CHS-002-80-2 50None None None 3-CHS-002-80-2 51None None None 3-CHS-002-80-2 52None None None 3-CHS-002-80-2 53None None None 3-CHS-003-8-2543-RCS-275-15-1None None 3-CHS-003-659-2553-RCS-275-15-1None NoneTABLE 3.6-25 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - LETDOWN LINE (CONTINUED)Line DesignationBreak #
(1)Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-186Rev. 30TABLE 3.6-26 JET IMPINGEMENT EFFECTS - CHEMICAL AND VOLUME CONTROL SY STEM - LETDOWN LINE PIPING Line DesignationBreak #Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement3-CHS-003-1-29NoneN/AN/AN/ANone3-CHS-003-416-210NoneN/AN/AN/ANone 3-CHS-003-416-211None N/AN/A N/ANone3-CHS-003-416-212None N/AN/A N/ANone3-CHS-003-416-213None N/AN/A N/ANone3-CHS-002-2-214None N/AN/A N/ANone3-CHS-002-590-215None N/AN/A N/ANone3-CHS-002-7-216None N/AN/A N/ANone3-CHS-002-4-217None N/AN/A N/ANone3-CHS-002-589-218None N/AN/A N/ANone3-CHS-002-6-219None N/AN/A N/ANone3-CHS-002-6-220None N/AN/A N/ANone3-CHS-002-3-221None N/AN/A N/ANone3-CHS-002-588-222None N/AN/A N/ANone3-CHS-002-5-223None N/AN/A N/ANone3-CHS-002-5-224None N/AN/A N/ANone3-CHS-025-304-225None N/AN/A N/ANone3-CHS-003-8-226 None (1)N/AN/A N/ANone3-CHS-003-8-227 None (1)N/AN/A N/ANone MPS3 UFSARMPS3 UFSAR3.6-187Rev. 303-CHS-003-8-228None N/AN/A N/ANone3-CHS-003-8-229 None (1)N/AN/A N/ANone3-CHS-003-78-230None N/AN/A N/ANone3-CHS-003-77-231None N/AN/A N/ANone3-CHS-003-9-232None N/AN/A N/ANone3-CHS-002-81-233 None (1)N/AN/A N/ANone3-CHS-002-81-234None N/AN/A N/ANone3-CHS-002-80-235 None (1)N/AN/A N/ANone3-CHS-002-80-236None N/AN/A N/ANone3-CHS-002-80-237 None (1)N/AN/A N/ANone3-CHS-002-80-238 None (1)N/AN/A N/ANone3-CHS-002-80-239None N/AN/A N/ANone3-CHS-002-80-240 None (1)N/AN/A N/ANone3-CHS-002-80-241 None (1)N/AN/A N/ANone3-CHS-002-80-242None N/AN/A N/ANone3-CHS-002-80-243None N/AN/A N/ANone3-CHS-002-80-244None N/AN/A N/ANone3-CHS-002-80-245None N/AN/A N/ANoneTABLE 3.6-26 JET IMPINGEMENT EFFECTS - CHEMICAL AND VOLUME CONTROL SY STEM - LETDOWN LINE PIPING (CONTINUED)Line DesignationBreak #Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-188Rev. 30NOTE:(1) USNRC Generic Letter 87-11 eliminates th e requirement to evaluate jet impingement effects for Arbitrary Intermediate Breaks (AIBs). Therefore, jet impingement effects subsequent to AIBs are not evaluated.3-CHS-002-80-246 None (1)N/AN/A N/ANone3-CHS-002-80-247None N/AN/A N/ANone3-CHS-002-80-248None N/AN/A N/ANone3-CHS-002-80-249None N/AN/A N/ANone3-CHS-002-80-250None N/AN/A N/ANone3-CHS-002-80-251None N/AN/A N/ANone3-CHS-002-80-252None N/AN/A N/ANone3-CHS-002-80-253None N/AN/A N/ANone3-CHS-003-8-254None N/AN/A N/ANone3-CHS-003-659-255None N/AN/A N/ANoneTABLE 3.6-26 JET IMPINGEMENT EFFECTS - CHEMICAL AND VOLUME CONTROL SY STEM - LETDOWN LINE PIPING (CONTINUED)Line DesignationBreak #Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-189Rev. 30TABLE 3.6-27 POSTULATED BREAKS - SEAL WATER INJECTION SYSTEM Line DesignationBreak #BuildingElevationBreak Type (1)Total Additive StressFigure3-CHS-003-32-21Auxiliary Bldg.27'-3"CBN/A Terminal End3.6-17 3-CHS-003-32-22Auxiliary Bldg.18'-5"CBN/A Arbitrary Intermediate3.6-17 3-CHS-003-32-23Auxiliary Bldg.15'-2"CBN/A Arbitrary Intermediate3.6-17 3-CHS-003-32-24Auxiliary Bldg.19'-6"CBN/A Terminal End3.6-17 3-CHS-002-446-25Auxiliary Bldg.21'-8"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-446-26Auxiliary Bldg.22'-6"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-34-27Auxiliary Bldg.34'-11"CBN/A Terminal End3.6-17 3-CHS-002-34-28Auxiliary Bldg.36'-4"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-34-29Auxiliary Bldg.36'-4"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-34-210Auxiliary Bldg.36'-4"CBN/A Terminal End3.6-17 3-CHS-002-33-211Auxiliary Bldg.34'-3"CBN/A Terminal End3.6-17 3-CHS-002-33-212Auxiliary Bldg.36'-4"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-33-213Auxiliary Bldg.36'-4"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-33-214Auxiliary Bldg.36'-4"CBN/A Terminal End3.6-17 3-CHS-002-447-215Auxiliary Bldg.36'-4"CBN/A Terminal End3.6-17 3-CHS-002-445-216Auxiliary Bldg.36'-4"CBN/A Terminal End3.6-17 3-CHS-003-38-217Auxiliary Bldg.18'-0"CBN/A Arbitrary Intermediate3.6-17 3-CHS-003-38-218Auxiliary Bldg.18'-5"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-39-219Auxiliary Bldg.20'-1"CBN/A Terminal End3.6-17 3-CHS-002-41-220Auxiliary Bldg.12'-9"CBN/A Arbitrary Intermediate3.6-17 MPS3 UFSARMPS3 UFSAR3.6-190Rev. 303-CHS-002-42-221Auxiliary Bldg.12'-6"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-638-222Containment11'-8"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-638-223Containment12'-6"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-638-224Containment12'-6"CBN/A Terminal End3.6-17 3-CHS-002-638-225Containment9'-3"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-638-226Containment12'-6"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-638-227Containment12'-6"CBN/A Terminal End3.6-17 3-CHS-002-638-228Containment20'-3"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-638-229Containment20'-6"CBN/A Arbitrary Intermediate3.6-17 3-CHS-003-38-233Auxiliary Bldg.18'-5"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-44-234Auxiliary Bldg.20'-0"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-44-235Auxiliary Bldg.18'-11"CBN/A Terminal End3.6-17 3-CHS-002-46-236Auxiliary Bldg.12'-6"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-637-237Containment11'-6"CBAbove Threshold3.6-17 3-CHS-002-637-238Containment11'-9"CBAbove Threshold3.6-17 3-CHS-003-38-241Auxiliary Bldg.18'-5"CBN/A Terminal End3.6-17 3-CHS-002-51-242Auxiliary Bldg.13'-4"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-51-243Auxiliary Bldg.12'-6"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-52-244Containment11'-8"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-651-245Containment12'-6"CBN/A Arbitrary Intermediate3.6-17 TABLE 3.6-27 POSTULATED BREAKS - SEAL WATER INJECTION SYSTEM (CONTINUED)Line DesignationBreak #BuildingElevationBreak Type (1)Total Additive StressFigure MPS3 UFSARMPS3 UFSAR3.6-191Rev. 303-CHS-002-651-246Containment12'-6"CBN/A Terminal End3.6-17 3-CHS-002-651-247Containment12'-9"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-651-248Containment20'-6"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-54-252Auxiliary19'-6"CBN/A Terminal End3.6-17 3-CHS-002-56-253Auxiliary12'-6"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-57-254Containment12'-6"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-650-255Containment11'-6"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-650-256Containment11'-6"CBN/A Terminal End3.6-17 3-CHS-002-650-257Containment11'-6"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-650-258Containment11'-6"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-650-259Containment11'-6"CBN/A Terminal End3.6-17 3-CHS-002-650-260Containment13'-7"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-650-261Containment19'-6"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-43-130Containment20'-6"CBN/A Arbitrary Intermediate  3.6-17 3-CHS-002-43-131Containment20'-6"CBN/A Arbitrary Intermediate 3.6-17 3-CHS-150-689-132Containment20'-6"CBN/A Terminal End 3.6-17 3-CHS-002-48-139Containment20'-6"CBN/A Arbitrary Intermediate 3.6-17 3-CHS-150-688-140Containment20'-6"CBN/A Terminal End  3.6-17 3-CHS-002-53-149Containment20'-6"CBN/A Arbitrary Intermediate3.6-17 3-CHS-002-53-150Containment20'-6"CBN/A Arbitrary Intermediate  3.6-17 TABLE 3.6-27 POSTULATED BREAKS - SEAL WATER INJECTION SYSTEM (CONTINUED)Line DesignationBreak #BuildingElevationBreak Type (1)Total Additive StressFigure MPS3 UFSARMPS3 UFSAR3.6-192Rev. 30NOTE:(1) Circumferential Pipe Brea k (CB) is defined in FSAR Section 3.6.2.1.3
.3-CHS-150-687-151Containment20'-6"CBN/A Terminal End 3.6-17 3-CHS-002-58-162Containment20'-6"CBN/A Arbitrary Intermediate 3.6-17 3-CHS-002-58-163Containment20'-6"CBN/A Arbitrary Intermediate 3.6-17 3-CHS-150-686-164Containment20'-6"CBN/A Terminal End3.6-17 3-CHS-002-42-265Auxiliary12'-6"CBN/A Terminal End3.6-17 3-CHS-002-42-266Containment12'-6"CBN/A Terminal End3.6-17 3-CHS-002-47-267Auxiliary12'-6"CBN/A Terminal End3.6-17 3-CHS-002-47-268Containment12'-6"CBN/A Terminal End3.6-17 3-CHS-002-57-269Auxiliary12'-6"CBN/A Terminal End3.6-17 3-CHS-002-57-270Containment12'-6"CBN/A Terminal End3.6-17 3-CHS-002-52-271Auxiliary12'-6"CBN/A Terminal End3.6-17 3-CHS-002-52-272Containment12'-6"CBN/A Terminal End3.6-17 TABLE 3.6-27 POSTULATED BREAKS - SEAL WATER INJECTION SYSTEM (CONTINUED)Line DesignationBreak #BuildingElevationBreak Type (1)Total Additive StressFigure MPS3 UFSARMPS3 UFSAR3.6-193Rev. 30TABLE 3.6-28 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE Line DesignationBreak #Blowdown SourceEssential Pipe Whip TargetProtection Requirement3-CHS-003-32-21 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)3-CHS-002-34-2 10 None (1)None None 3-CHS-002-33-211 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)
MPS3 UFSARMPS3 UFSAR3.6-194Rev. 30 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)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 NoneTABLE 3.6-28 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE Line DesignationBreak #Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-195Rev. 30 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)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 NoneTABLE 3.6-28 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE Line DesignationBreak #Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-196Rev. 30 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)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 NoneTABLE 3.6-28 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE Line DesignationBreak #Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-197Rev. 30NOTES:(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 el iminates the requirement to eval uate pipe whip for Arbitrary Intermediate Breaks (AIBs) theref ore pipe whip is not evaluated for AIBs.
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 NoneTABLE 3.6-28 PIPE WHIP EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE Line DesignationBreak #Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-198Rev. 30TABLE 3.6-29 JET IMPINGEMENT EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE Line DesignationBreak #Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement3-CHS-003-32-21None - See Note 1N/AN/AN/ANone3-CHS-003-32-22None - See Note 1N/AN/AN/ANone - See Note 23-CHS-003-32-23None - See Note 1N/AN/AN/ANone - See Note 23-CHS-003-32-24None - See Note 1N/AN/AN/ANone 3-CHS-002-446-25None - See Note 1N/AN/AN/ANone - See Note 23-CHS-002-446-26None - See Note 1N/AN/AN/ANone - See Note 23-CHS-002-34-27None - See Note 1N/AN/AN/ANone 3-CHS-002-34-28None - See Note 1N/AN/AN/ANone - See Note 23-CHS-002-34-29None - See Note 1N/AN/AN/ANone - See Note 23-CHS-002-34-210None - See Note 1N/AN/AN/ANone 3-CHS-002-33-211None - See Note 1N/AN/AN/ANone 3-CHS-002-33-212None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-33-213None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-33-214None - See Note 1N/A N/AN/ANone3-CHS-002-447-215None - See Note 1N/A N/AN/ANone3-CHS-002-445-216None - See Note 1N/A N/AN/ANone3-CHS-003-38-217None - See Note 1N/A N/AN/ANone - See Note 23-CHS-003-38-218None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-39-219None - See Note 1N/A N/AN/ANone MPS3 UFSARMPS3 UFSAR3.6-199Rev. 303-CHS-002-41-220None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-42-221None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-638-222None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-638-223None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-638-224None - See Note 1N/A N/AN/ANone3-CHS-002-638-225None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-638-226None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-638-227None - See Note 1N/A N/AN/ANone3-CHS-002-638-228None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-638-229None - See Note 1N/A N/AN/ANone - See Note 23-CHS-003-38-233None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-44-234None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-44-235None - See Note 1N/A N/AN/ANone3-CHS-002-46-236None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-637-237None - See Note 1N/A N/AN/ANone3-CHS-002-637-238None - See Note 1N/A N/AN/ANone3-CHS-003-38-241None - See Note 1N/A N/AN/ANone3-CHS-002-51-242None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-51-243None - See Note 1N/A N/AN/ANone - See Note 2TABLE 3.6-29 JET IMPINGEMENT EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE (CONTINUED)Line DesignationBreak #Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-200Rev. 303-CHS-002-52-244None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-651-245None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-651-246None - See Note 1N/A N/AN/ANone3-CHS-002-651-247None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-651-248None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-54-252None - See Note 1N/A N/AN/ANone3-CHS-002-56-253None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-57-254None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-650-255None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-650-256None - See Note 1N/A N/AN/ANone3-CHS-002-650-257None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-650-258None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-650-259None - See Note 1N/A N/AN/ANone3-CHS-002-650-260None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-650-261None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-43-130None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-43-131None - See Note 1N/A N/AN/ANone - See Note 23-CHS-150-689-132None - See Note 1N/A N/AN/ANone3-CHS-002-48-139None - See Note 1N/A N/AN/ANone - See Note 2TABLE 3.6-29 JET IMPINGEMENT EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE (CONTINUED)Line DesignationBreak #Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-201Rev. 30NOTES: (1) Sustained blowdown does not occur on the Ch emical & Volume Control System Seal Water Injection Line. The Seal Water Injecti on 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 effe cts are not sustained and ar e not evaluated as stated in FSAR Section 3.6.2.2.1 Item 3
.3-CHS-150-688-140None - See Note 1N/A N/AN/ANone3-CHS-002-53-149None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-53-150None - See Note 1N/A N/AN/ANone - See Note 23-CHS-150-687-151None - See Note 1N/A N/AN/ANone3-CHS-002-58-162None - See Note 1N/A N/AN/ANone - See Note 23-CHS-002-58-163None - See Note 1N/A N/AN/ANone - See Note 23-CHS-150-686-164None - See Note 1N/A N/AN/ANone3-CHS-002-42-265None - See Note 1N/A N/AN/ANone3-CHS-002-42-266None - See Note 1N/A N/AN/ANone3-CHS-002-47-267None - See Note 1N/A N/AN/ANone3-CHS-002-47-268None - See Note 1N/A N/AN/ANone3-CHS-002-57-269None - See Note 1N/A N/AN/ANone3-CHS-002-57-270None - See Note 1N/A N/AN/ANone3-CHS-002-52-271None - See Note 1N/A N/AN/ANone3-CHS-002-52-272None - See Note 1N/A N/AN/ANoneTABLE 3.6-29 JET IMPINGEMENT EFFECTS - CHEMICAL VOLUME CONTROL SYSTEM - SEAL WATER INJECTION LINE (CONTINUED)Line DesignationBreak #Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-202Rev. 30TABLE 3.6-30 POSTULATED BREAKS - ST EAM GENERATOR BLOWDOWN SYSTEM Line DesignationBreak #BuildingElevationBreak Type (1)Total Additive StressFigure 3.6-3-BDG-002-45-21Containment24'-7"CBN/A Terminal End18 3-BDG-004-9-22Containment21'-1"CBN/A Arbitrary Intermediate18 3-BDG-004-79-23Containment20'-4"CBN/A Terminal End18 3-BDG-004-9-24Containment22'-7"CB & LSAbove Threshold 18 3-BDG-004-17-45MSVB20'-0"CBN/A Arbitrary Intermediate18 3-BDG-004-17-46MSVB19'-6"CBN/A Arbitrary Intermediate18 3-BDG-004-17-47MSVB19'-6"CBN/A Terminal End 18 3-BDG-002-44-28Containment24'-7"CBN/A Terminal End 18 3-BDG-002-44-29Containment24'-3"CBAbove Threshold 18 3-BDG-004-10-210Containment20'-6"CBN/A Arbitrary Intermediate18 3-BDG-004-81-211Containment20'-5"CBN/A Terminal End 18 3-BDG-004-18-412MSVB20'-0"CBN/A Arbitrary Intermediate18 3-BDG-004-18-413MSVB19'-6"CBN/A Arbitrary Intermediate18 3-BDG-004-18-414MSVB19'-6"CBN/A Terminal End 18 3-BDG-002-46-215Containment24'-7"CBN/A Terminal End 18 3-BDG-002-46-216Containment24'-2"CBAbove Threshold 18 3-BDG-004-11-217Containment21'-1"CBN/A Arbitrary Intermediate18 3-BDG-004-83-218Containment20'-3"CBN/A Terminal End 18 3-BDG-004-19-419MSVB20'-0"CBN/A Arbitrary Intermediate18 3-BDG-004-19-420MSVB19'-6"CBN/A Arbitrary Intermediate18 3-BDG-004-19-421MSVB19'-6"CBN/A Terminal End 18 3-BDG-002-47-222Containment24'-7"CBN/A Terminal End 18 MPS3 UFSARMPS3 UFSAR3.6-203Rev. 303-BDG-004-85-223Containment19'-9"CBN/A Terminal End 18 3-BDG-004-12-224Containment16'-6"CB & LSAbove Threshold 18 3-BDG-004-12-225Containment20'-0"CBN/A Arbitrary Intermediate18 3-BDG-004-20-426MSVB20'-0"CBN/A Arbitrary Intermediate18 3-BDG-004-20-427MSVB19'-6"CBN/A Arbitrary Intermediate18 3-BDG-004-20-428MSVB19'-5"CBN/A Terminal End 18 3-BDG-004-17-429MSVB20'-6"CBN/A Terminal End 18 3-BDG-004-18-430MSVB20'-6"CBN/A Terminal End 18 3-BDG-004-19-431MSVB20'-6"CBN/A Terminal End 18 3-BDG-004-20-432MSVB20'-6"CBN/A Terminal End 18 3-BDG-004-78-233Containment72'-4"CBN/A Terminal End 18 3-BDG-004-78-234Containment72'-4"CBN/A - Arbitrary Intermediate18 2" LINE TO SG1A35Containment69'-1"CBN/A - Arbitrary Intermediate18 2" LINE TO SG1A36Containment68'-10"CBN/A Terminal End 18 3-BDG-004-80-237Containment74'-7"CBN/A Terminal End 18 2" LINE TO SG1B38Containment69'-1"CBAbove Threshold 18 2" LINE TO SG1B39Containment68'-11"CBAbove Threshold 18 2" LINE TO SG1B40Containment68'-8"CBN/A Terminal End 18 3-BDG-004-82-241Containment74'-8"CBN/A Terminal End 18 3-BDG-004-82-242Containment74'-8"CBN/A - Arbitrary Intermediate18 2" LINE TO SG1C43Containment69'-7"CBN/A - Arbitrary Intermediate18 2" LINE TO SG1C44Containment69'-4"CBN/A Terminal End 18 TABLE 3.6-30 POSTULATED BREAKS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)Line DesignationBreak #BuildingElevationBreak Type (1)Total Additive StressFigure 3.6-MPS3 UFSARMPS3 UFSAR3.6-204Rev. 303-BDG-004-84-245Containment72'-4"CBN/A Terminal End 18 2" LINE TO SG1D46Containment69'-2"CBAbove Threshold 18 2" LINE TO SG1D47Containment68'-11"CBAbove Threshold 18 2" LINE TO SG1D48Containment68'-8"CBN/A Terminal End 18 3-BDG-004-55-249Containment20'-6"CBN/A Terminal End 18 3-BDG-004-54-250Containment20'-6"CBN/A Terminal End 18 3-BDG-004-56-251Containment20'-6"CBN/A Terminal End 18 3-BDG-004-57-252Containment20'-6"CBN/A Terminal End 18 3-BDG-004-17-4IPMSVB38' to 56'CB & LSN/A (2)18 3-BDG-004-18-4IPMSVB38' to 56'CB & LSN/A (2)18 3-BDG-004-19-4IPMSVB38' to 56'CB & LSN/A (2)18 3-BDG-004-20-4IPMSVB38' to 56'CB & LSN/A (2)18 3-BDG-004-21-4IPMSVB & Turbine56'-0"CB & LSN/A (2)18 3-BDG-004-22-4IPMSVB & Turbine56'-0"CB & LSN/A (2)18 3-BDG-004-23-4IPMSVB & Turbine56'-0"CB & LSN/A (2)18 3-BDG-004-24-4IPMSVB & Turbine56'-0"CB & LSN/A (2)18 3-BDG-003-49-4IPMSVB19'-1"CBN/A (2)18 3-BDG-003-48-4IPMSVB19'-2"CBN/A (2)18 3-BDG-003-50-4IPMSVB18'-10"CBN/A (2)18 3-BDG-003-51-4IPMSVB18'-9"CBN/A (2)18 TABLE 3.6-30 POSTULATED BREAKS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)Line DesignationBreak #BuildingElevationBreak Type (1)Total Additive StressFigure 3.6-MPS3 UFSARMPS3 UFSAR3.6-205Rev. 30NOTES:(1) Circumferential Pipe Break (CB) and Longi tudinal Pipe Split (LS) are defined in FSAR Section 3.6.2.1.3
.(2) Pipe breaks are postulated on the Nonnuc lear (i.e., Nonsafety and Nonseismic) porti on of the Steam Generator Blowdown Syste m in the Main Steam Valve Building at all valves, fittings, and integral welded attachments. Use of a fitting criteria is in acco rdance with FSAR Section 3.6.2.1.2.3.2.a. Break Number Designation IP is for an Intermediate Break Locati on and the Break Number Designation TP is a Terminal Point.(3) Terminal End (TP) pipe breaks ar e postulated at the normally closed valv es 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 Manifo ld, 3-BDG-008-29-4, is located in the T urbin e Building. No safety-related structure s, systems, or components are located in the Turbine Building. Steam Generator Blowdown Tank Manifold, 3-BDG-008-29-4 is therefore remote from essential structur es, system, or component. Postulated pipe break on the Steam Generator Blowdown Tank Manifold, 3-BDG-008-29-4 is not assessed.3-BDG-003-49-4TPMSVB17'-3"CBN/A (3)18 3-BDG-003-48-4TPMSVB17'-3"CBN/A (3)18 3-BDG-003-50-4TPMSVB16'-11"CBN/A (3)18 3-BDG-003-51-4TPMSVB16'-10"CBN/A (3)18 3-BDG-008-29-4NoneTurbine56'-0"NoneN/A (4)18 TABLE 3.6-30 POSTULATED BREAKS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)Line DesignationBreak #BuildingElevationBreak Type (1)Total Additive StressFigure 3.6-MPS3 UFSARMPS3 UFSAR3.6-206Rev. 30TABLE 3.6-31 PIPE WHIP EFFECTS - ST EAM GENERATOR BLOWDOWN SYSTEM Line DesignationBreak #
(1)Blowdown SourceEssential Pipe Whip Target Primary Protection RequirementFigure3-BDG-002-45-213-BDG-008-29-4Steam Generator 3RCS*SG1A Shell3BDG-PRR975A3.6-18 3-BDG-004-9-223-BDG-008-29-4East-West Concrete WallNone3.6-18 3-BDG-004-9-223RCS*SG1A Steam Generator 3RCS*SG1A Shell3BDG-PRR972A3.6-18 3-BDG-004-79-233RCS*SG1ANoneNone3.6-18 3-BDG-004-9-243RCS*SG1ANoneNone3.6-18 3-BDG-004-9-243-BDG-008-29-4NoneNone3.6-18 3-BDG-004-17-453-BDG-008-29-4NoneNone3.6-18 3-BDG-004-17-453RCS*SG1ALimit Stress to Break Exclusion Zone3BDG-PRR8863.6-18 3-BDG-004-17-463-BDG-008-29-4NoneNone3.6-18 3-BDG-004-17-463RCS*SG1ALimit Stress to Break Exclusion Zone3BDG-PRR9413.6-18 3-BDG-004-17-473-BDG-008-29-4NoneNone3.6-18 3-BDG-004-17-473RCS*SG1ALimit Stress to Break Exclusion Zone3BDG-PRR9483.6-18 3-BDG-004-17-4293RCS*SG1ALimit Stress to Break Exclusion Zone3BDG-PRR8853.6-18 3-BDG-004-17-4293-BDG-008-29-4NoneNone3.6-18 3-BDG-002-44-283-BDG-008-29-4Steam Generator 3RCS*SG1B Shell3BDG-PRR975B3.6-18 3-BDG-002-44-293-BDG-008-29-4Steam Generator 3RCS*SG1B Shell3BDG-PRR975B3.6-18 3-BDG-002-44-293RCS*SG1BNoneNone3.6-18 3-BDG-004-10-2103RCS*SG1BCrane Wall Thimble 18-1None 3.6-18 3-BDG-004-10-2103-BDG-008-29-4Pre vent Pipe Whip into Containment Liner3BDG-PRR9733.6-18 MPS3 UFSARMPS3 UFSAR3.6-207Rev. 303-BDG-004-81-2113RCS*SG1BNone None 3.6-18 3-BDG-004-18-4123-BDG-008-29-4None None 3.6-18 3-BDG-004-18-4123RCS*SG1BLimit Stress to Break Exclusion Zone3BDG-PRR8883.6-18 3-BDG-004-18-4133-BDG-008-29-4None None 3.6-18 3-BDG-004-18-4133RCS*SG1BLimit Stress to Break Exclusion Zone3BDG-PRR9423.6-18 3-BDG-004-18-4143RCS*SG1BLimit Stress to Break Exclusion Zone3BDG-PRR9503.6-18 3-BDG-004-18-4143-BDG-008-29-4None None 3.6-18 3-BDG-004-18-4303RCS*SG1BLimit Stress to Break Exclusion Zone3BDG-PRR8873.6-18 3-BDG-004-18-4303-BDG-008-29-4None None 3.6-18 3-BDG-002-46-2153-BDG-008-29-4 Steam Generator 3RCS*SG1C Shell3BDG-PRR975C3.6-18 3-BDG-002-46-2163RCS*SG1CNone None 3.6-18 3-BDG-002-46-2163-BDG-008-29-4 Steam Generator 3RCS*SG1C Shell3BDG-PRR975C3.6-18 3-BDG-004-11-2173RCS*SG1C Steam Generator 3RCS*SG1C Shell3BDG-PRR972C3.6-18 3-BDG-004-11-2173-BDG-008-29-4East-West Concrete Wall None 3.6-18 3-BDG-004-83-2183RCS*SG1CNone None 3.6-18 3-BDG-004-19-4193RCS*SG1CLimit Stress to Break Exclusion Zone3BDG-PRR8833.6-18 3-BDG-004-19-4193-BDG-008-29-4None None 3.6-18 3-BDG-004-19-4203-BDG-008-29-4None None 3.6-18 3-BDG-004-19-4203RCS*SG1CLimit Stress to Break Exclusion Zone3BDG-PRR9433.6-18 TABLE 3.6-31 PIPE WHIP EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)Line DesignationBreak #
(1)Blowdown SourceEssential Pipe Whip Target Primary Protection RequirementFigure MPS3 UFSARMPS3 UFSAR3.6-208Rev. 303-BDG-004-19-4213RCS*SG1CLimit Stress to Break Exclusion Zone3BDG-PRR9473.6-18 3-BDG-004-19-4213-BDG-008-29-4NoneNone3.6-18 3-BDG-004-19-4313RCS*SG1CLimit Stress to Break Exclusion Zone3BDG-PRR8843.6-18 3-BDG-004-19-4313-BDG-008-29-4NoneNone3.6-18 3-BDG-002-47-2223-BDG-008-29-4 Steam Generator 3RCS*SG1D Shell3BDG-PRR975D3.6-18 3-BDG-004-85-2233RCS*SG1DNoneNone3.6-18 3-BDG-004-12-2243-BDG-008-29-4NoneNone3.6-18 3-BDG-004-12-2243RCS*SG1DNoneNone3.6-18 3-BDG-004-12-2253-BDG-008-29-4Pre vent Pipe Whip into Containment Liner3BDG-PRR9743.6-18 3-BDG-004-12-2253RCS*SG1DNoneNone3.6-18 3-BDG-004-20-4263-BDG-008-29-4NoneNone3.6-18 3-BDG-004-20-4263RCS*SG1DLimit Stress to Break Exclusion Zone3BDG-PRR8813.6-18 3-BDG-004-20-4273-BDG-008-29-4NoneNone3.6-18 3-BDG-004-20-4273RCS*SG1DLimit Stress to Break Exclusion Zone3BDG-PRR9443.6-18 3-BDG-004-20-4283RCS*SG1DLimit Stress to Break Exclusion Zone3BDG-PRR9493.6-18 3-BDG-004-20-4283-BDG-008-29-4NoneNone3.6-18 3-BDG-004-20-4323RCS*SG1DLimit Stress to Break Exclusion Zone3BDG-PRR8823.6-18 3-BDG-004-20-4323-BDG-008-29-4NoneNone3.6-18 3-BDG-004-55-2493RCS*SG1ANoneNone3.6-18 TABLE 3.6-31 PIPE WHIP EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)Line DesignationBreak #
(1)Blowdown SourceEssential Pipe Whip Target Primary Protection RequirementFigure MPS3 UFSARMPS3 UFSAR3.6-209Rev. 30NOTES:(1) Repetition of Break Numbers is used to identify separate fl uid reservoirs which provide a c onstant pressure source to maint ain system pressure subsequent to the postula ted pipe break. These reservoirs maintain system blowdown during the transient event a s 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.(2) Steam Generator Blowdown (BDG) System - Wet Lay Up Piping is a High Ener gy 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-BDG-004-54-2503RCS*SG1BNone None 3.6-18 3-BDG-004-56-2513RCS*SG1CPrevent Pipe Whip into Steam Generator Support3BDG-PRR9703.6-18 3-BDG-004-57-2523RCS*SG1DNone None 3.6-18 TABLE 3.6-31 PIPE WHIP EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)Line DesignationBreak #
(1)Blowdown SourceEssential Pipe Whip Target Primary Protection RequirementFigure MPS3 UFSARMPS3 UFSAR3.6-210Rev. 30TABLE 3.6-32 JET IMPINGEMENT EFFECTS - STEAM GENERATOR BL OWDOWN SYSTEM Line DesignationBreak #
(1)Essential Jet Impingement TargetDistance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement3-BDG-002-45-21NoneN/AN/AN/ANone3-BDG-004-9-22 None (2)N/AN/AN/ANone3-BDG-004-9-22 None (2)N/AN/AN/ANone3-BDG-004-79-23 None N/AN/AN/ANone3-BDG-004-9-24 None N/AN/AN/ANone3-BDG-004-9-24 None N/AN/AN/ANone3-BDG-004-9-2Split None N/AN/AN/ANone3-BDG-004-17-45 None (2)N/AN/AN/ANone3-BDG-004-17-45 None (2)N/AN/AN/ANone3-BDG-004-17-46 None (2)N/AN/AN/ANone3-BDG-004-17-46 None (2)N/AN/AN/ANone3-BDG-004-17-47 None N/AN/AN/ANone3-BDG-004-17-47 None N/AN/AN/ANone3-BDG-004-17-429None N/AN/AN/ANone3-BDG-004-17-429None N/AN/AN/ANone3-BDG-002-44-28 None N/AN/AN/ANone3-BDG-002-44-29 None N/AN/AN/ANone3-BDG-002-44-29 None N/AN/AN/ANone3-BDG-004-10-210 None (2)N/AN/AN/ANone MPS3 UFSARMPS3 UFSAR3.6-211Rev. 303-BDG-004-10-210 None (2)N/AN/AN/ANone3-BDG-004-81-211None N/AN/AN/ANone3-BDG-004-18-412 None (2)N/AN/AN/ANone3-BDG-004-18-412 None (2)N/AN/AN/ANone3-BDG-004-18-413 None (2)N/AN/AN/ANone3-BDG-004-18-413 None (2)N/AN/AN/ANone3-BDG-004-18-414None N/AN/AN/ANone3-BDG-004-18-414None N/AN/AN/ANone3-BDG-004-18-430None N/AN/AN/ANone3-BDG-004-18-430None N/AN/AN/ANone3-BDG-002-46-215None N/AN/AN/ANone3-BDG-002-46-216None N/AN/AN/ANone3-BDG-002-46-216None N/AN/AN/ANone3-BDG-004-11-217 None (2)N/AN/AN/ANone3-BDG-004-11-217 None (2)N/AN/AN/ANone3-BDG-004-83-218None N/AN/AN/ANone3-BDG-004-19-419 None (2)N/AN/AN/ANone3-BDG-004-19-419 None (2)N/AN/AN/ANoneTABLE 3.6-32 JET IMPINGEMENT EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)Line DesignationBreak #
(1)Essential Jet Impingement TargetDistance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-212Rev. 303-BDG-004-19-420 None (2)N/AN/AN/ANone3-BDG-004-19-420 None (2)N/AN/AN/ANone3-BDG-004-19-421None N/AN/AN/ANone3-BDG-004-19-421None N/AN/AN/ANone3-BDG-004-19-431None N/AN/AN/ANone3-BDG-004-19-431None N/AN/AN/ANone3-BDG-002-47-222None N/AN/AN/ANone3-BDG-004-85-223None N/AN/AN/ANone3-BDG-004-12-224None N/AN/AN/ANone3-BDG-004-12-224None N/AN/AN/ANone3-BDG-004-12-2Split None N/AN/AN/ANone3-BDG-004-12-225 None (2)N/AN/AN/ANone3-BDG-004-12-225 None (2)N/AN/AN/ANone3-BDG-004-20-426 None (2)N/AN/AN/ANone3-BDG-004-20-426 None (2)N/AN/AN/ANone3-BDG-004-20-427 None (2)N/AN/AN/ANone3-BDG-004-20-427 None (2)N/AN/AN/ANone3-BDG-004-20-428None N/AN/AN/ANone3-BDG-004-20-428None N/AN/AN/ANoneTABLE 3.6-32 JET IMPINGEMENT EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)Line DesignationBreak #
(1)Essential Jet Impingement TargetDistance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-213Rev. 303-BDG-004-20-432None N/AN/AN/ANone3-BDG-004-20-432None N/AN/AN/ANone3-BDG-004-55-249None N/AN/AN/ANone3-BDG-004-54-250None N/AN/AN/ANone3-BDG-004-56-251None N/AN/AN/ANone3-BDG-004-57-252None N/AN/AN/ANone3-BDG-004-78-233 None (3)N/AN/AN/ANone3-BDG-004-78-234 None (3)N/AN/AN/ANone2" LINE TO SG1A35 None (3)N/AN/AN/ANone2" LINE TO SG1A36 None (3)N/AN/AN/ANone3-BDG-004-80-237 None (3)N/AN/AN/ANone2" LINE TO SG1B38 None (3)N/AN/AN/ANone2" LINE TO SG1B39 None (3)N/AN/AN/ANone2" LINE TO SG1B40 None (3)N/AN/AN/ANone3-BDG-004-82-241 None (3)N/AN/AN/ANone3-BDG-004-82-242 None (3)N/AN/AN/ANone2" LINE TO SG1C43 None (3)N/AN/AN/ANone2" LINE TO SG1C44 None (3)N/AN/AN/ANoneTABLE 3.6-32 JET IMPINGEMENT EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)Line DesignationBreak #
(1)Essential Jet Impingement TargetDistance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-214Rev. 303-BDG-004-84-245 None (3)N/AN/AN/ANone2" LINE TO SG1D46 None (3)N/AN/AN/ANone2" LINE TO SG1D47 None (3)N/AN/AN/ANone2" LINE TO SG1D48 None (3)N/AN/AN/ANone3-BDG-004-17-4IPNone N/AN/AN/ANone3-BDG-004-17-4IPNone N/AN/AN/ANone3-BDG-004-17-4Split None N/AN/AN/ANone3-BDG-004-18-4IPNone N/AN/AN/ANone3-BDG-004-18-4IPNone N/AN/AN/ANone3-BDG-004-18-4Split None N/AN/AN/ANone3-BDG-004-19-4IPNone N/AN/AN/ANone3-BDG-004-19-4IPNone N/AN/AN/ANone3-BDG-004-19-4Split None N/AN/AN/ANone3-BDG-004-20-4IPNone N/AN/AN/ANone3-BDG-004-20-4IPNone N/AN/AN/ANone3-BDG-004-20-4Split None N/AN/AN/ANone3-BDG-004-21-4IPNone N/AN/AN/ANone3-BDG-004-21-4IPNone N/AN/AN/ANone3-BDG-004-21-4Split None N/AN/AN/ANoneTABLE 3.6-32 JET IMPINGEMENT EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)Line DesignationBreak #
(1)Essential Jet Impingement TargetDistance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-215Rev. 30NOTES:(1) Repetition of Break Numbers is used to identify separate fl uid reservoirs which provide a c onstant pressure source to maint ain system pressure subsequent to the postula ted pipe break. These reservoirs maintain system blowdown during the transient event a s well as in the steady state.3-BDG-004-22-4IPNone N/AN/AN/ANone3-BDG-004-22-4IPNone N/AN/AN/ANone3-BDG-004-22-4Split None N/AN/AN/ANone3-BDG-004-23-4IPNone N/AN/AN/ANone3-BDG-004-23-4IPNone N/AN/AN/ANone3-BDG-004-23-4Split None N/AN/AN/ANone3-BDG-004-24-4IPNone N/AN/AN/ANone3-BDG-004-24-4IPNone N/AN/AN/ANone3-BDG-004-24-4Split None N/AN/AN/ANone3-BDG-003-49-4IPNone N/AN/AN/ANone3-BDG-003-48-4IPNone N/AN/AN/ANone3-BDG-003-50-4IPNone N/AN/AN/ANone3-BDG-003-51-4IPNone N/AN/AN/ANone3-BDG-003-49-4TPNone N/AN/AN/ANone3-BDG-003-48-4TPNone N/AN/AN/ANone3-BDG-003-50-4TPNone N/AN/AN/ANone3-BDG-003-51-4TPNone N/AN/AN/ANoneTABLE 3.6-32 JET IMPINGEMENT EFFECTS - STEAM GENERATOR BLOWDOWN SYSTEM (CONTINUED)Line DesignationBreak #
(1)Essential Jet Impingement TargetDistance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-216Rev. 30(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 Ge nerators so any fluid jet impingement in one Steam Generator C ubicle does not da mage 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 Genera tor 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 UFSARMPS3 UFSAR3.6-217Rev. 30TABLE 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)0.00.00.00.526.6018.151.0517.6823.691.5731.1827.74 2.0946.4530.752.6263.4334.32 MPS3 UFSARMPS3 UFSAR3.6-218Rev. 30TABLE 3.6-34 POSTULATED BREAKS AUXILIARY FEED WATER SYSTEM Line DesignationBreak #BuildingElevationBreak Type (1)Total Additive StressFigure3FWA*P1A Discharge Flange1ESF26'-5"CBN/A Terminal End3.6-32 3-FWA-006-2-32ESF26'-5"CBN/A Arbitrary Intermediate3.6-32 3-FWA-006-2-33ESF23'-2"CBN/A Arbitrary Intermediate3.6-32 3-FWA-006-2-34ESF22'-5"CBN/A Terminal End3.6-32 3-FWA-006-8-35ESF22'-6"CBN/A Arbitrary Intermediate3.6-32 3-FWA-003-32-36ESF25'-11"CBN/A Arbitrary Intermediate3.6-32 3-FWA-003-39-37ESF25'-11"CBN/A Terminal End3.6-32 3-FWA-003-39-38ESF25'-1"CBN/A Arbitrary Intermediate3.6-32 3-FWA-004-121-39ESF22'-6"CBN/A Arbitrary Intermediate3.6-32 3-FWA-004-121-310ESF22'-6"CBN/A Terminal End3.6-32 3-FWA-004-42-213Containment43'-0"CBN/A Arbitrary Intermediate3.6-32 3-FWA-004-42-214Containment43'-0"CBN/A Arbitrary Intermediate3.6-32 3-FWA-006-135-215Containment58'-4"CBN/A Terminal End 3.6-32 3-FWA-003-33-316ESF25'-11"CBN/A Terminal End 3.6-32 3-FWA-003-33-317ESF25'-1"CBN/A Arbitrary Intermediate3.6-32 3-FWA-004-123-318ESF22'-6"CBN/A Arbitrary Intermediate3.6-32 3-FWA-004-123-319ESF22'-6"CBN/A Terminal End 3.6-32 3-FWA-004-36-222Containment43'-0"CBN/A Arbitrary Intermediate3.6-32 3-FWA-004-36-223Containment43'-0"CBN/A Arbitrary Intermediate3.6-32 3-FWA-006-100-224Containment58'-5"CBN/A Terminal End 3.6-32 3-FWA-004-125-325ESF22'-6"CBN/A Terminal End 3.6-32 MPS3 UFSARMPS3 UFSAR3.6-219Rev. 303-FWA-004-120-228Containment20'-0"CBN/A Arbitrary Intermediate3.6-32 3-FWA-004-120-229Containment20'-0"CBN/A Arbitrary Intermediate3.6-32 3-FWA-004-120-230Containment20'-0"CBN/A Terminal End 3.6-32 3-FWA-004-120-231Containment20'-0"CBN/A Arbitrary Intermediate3.6-32 3-FWA-004-120-232Containment21'-6"CB & LSAbove Threshold 3.6-32 3-FWA-004-120-233Containment21'-6"CBN/A Terminal End 3.6-32 3-FWA-004-54-234Containment21'-6"CBN/A Arbitrary Intermediate3.6-32 3-FWA-006-102-235Containment52'-11"CBN/A Arbitrary Intermediate3.6-32 3-FWA-006-138-236Containment58'-4"CBN/A Terminal End 3.6-32 3-FWA-004-127-337ESF22'-6"CBN/A Terminal End 3.6-32 3-FWA-004-119-240Containment20'-5"CBN/A Arbitrary Intermediate3.6-32 3-FWA-004-119-241Containment21'-0"CBN/A Arbitrary Intermediate3.6-32 3-FWA-004-119-242Containment21'-0"CBN/A Terminal End 3.6-32 3-FWA-004-48-243Containment21'-0"CBN/A Arbitrary Intermediate3.6-32 3-FWA-006-101-244Containment21'-9"CBN/A Arbitrary Intermediate3.6-32 3-FWA-006-137-245Containment58'-4"CBN/A Terminal End 3.6-32 3-FWA-006-79-346ESF22'-6"CBN/A Terminal End 3.6-32 3-FWA-004-117-247Containment43'-0"CBN/A Terminal End 3.6-32 3-FWA-004-118-248Containment43'-0"CBN/A Terminal End 3.6-32 3-FWA-004-119-249Containment20'-0"CBN/A Terminal End 3.6-32 3-FWA-004-120-250Containment20'-0"CBN/A Terminal End 3.6-32 3-FWA-004-41-251ESF43'-0"CBN/A Terminal End 3.6-32 TABLE 3.6-34 POSTULATED BREAKS AUXILIARY FEEDWATER SYSTEM (CONTINUED)Line DesignationBreak #BuildingElevationBreak Type (1)Total Additive StressFigure MPS3 UFSARMPS3 UFSAR3.6-220Rev. 30NOTE:(1) Circumferential Pipe Break (CB) and Longi tudinal Pipe Split (LS) are defined in FSAR Section 3.6.2.1.3
.3-FWA-004-35-252ESF43'-0"CBN/A Terminal End 3.6-32 3-FWA-004-47-253ESF20'-0"CBN/A Terminal End 3.6-32 3-FWA-004-53-254ESF20'-0"CBN/A Terminal End 3.6-32 3-FWA-004-122-355ESF12'-0"CBN/A Terminal End 3.6-32 3-FWA-003-66-356ESF26'-0"CBN/A Terminal End 3.6-32 3-FWA-004-124-357ESF12'-0"CBN/A Terminal End 3.6-32 3-FWA-003-71-358ESF26'-0"CBN/A Terminal End 3.6-32 3-FWA-006-8-359ESF26'-6"CBN/A Terminal End 3.6-32 3-FWA-004-126-360ESF22'-0"CBN/A Terminal End 3.6-32 3-FWA-003-56-361ESF26'-0"CBN/A Terminal End 3.6-32 3-FWA-004-128-362ESF22'-0"CBN/A Terminal End 3.6-32 3-FWA-003-61-363ESF26'-0"CBN/A Terminal End 3.6-32 3-FWA-008-14-364ESF22'-6"CBN/A Terminal End 3.6-32 3-FWA-008-14-365ESF26'-5"CBN/A Terminal End 3.6-32 3-FWA-004-160-366ESF25'-5"CBN/A Terminal End 3.6-32 TABLE 3.6-34 POSTULATED BREAKS AUXILIARY FEEDWATER SYSTEM (CONTINUED)Line DesignationBreak #BuildingElevationBreak Type (1)Total Additive StressFigure MPS3 UFSARMPS3 UFSAR3.6-221Rev. 30TABLE 3.6-35 PIPE WHIP EFFECTS AUXILIARY FEEDWATER SYSTEM Line DesignationBreak #Blowdown SourceEssential Pipe Whip TargetProtection Requirement3FWA*P1A Discharge Flange1None (1)None None3-FWA-006-2-3 2None (1)None None3-FWA-006-2-3 3None (1)None None3-FWA-006-2-3 4None (1)None None3-FWA-006-8-3 5None (1)None None3-FWA-003-32-3 6None (1)None None3-FWA-003-39-3 7None (1)None None3-FWA-003-39-3 8None (1)None None3-FWA-004-121-3 9None (1)None None3-FWA-004-121-3 10None (1)None None3-FWA-004-42-2 13None (1)None None3-FWA-004-42-2 14None (1)None None3-FWA-006-135-2 15None (1)None None3-FWA-003-33-3 16None (1)None None3-FWA-003-33-3 17None (1)None None3-FWA-004-123-3 18None (1)None None3-FWA-004-123-3 19None (1)None None MPS3 UFSARMPS3 UFSAR3.6-222Rev. 303-FWA-004-36-2 22None (1)None None3-FWA-004-36-2 23None (1)None None3-FWA-004-36-2233-FWS-020-18-2None None3-FWA-006-100-2 24None (1)None None3-FWA-004-125-3 25None (1)None None3-FWA-004-120-2 28None (1)None None3-FWA-004-120-2 29None (1)None None3-FWA-004-120-2 30None (1)None None3-FWA-004-120-2 31None (1)None None3-FWA-004-120-2 32 SplitNone (1)None None3-FWA-004-120-2 32None (1)None None3-FWA-004-120-2 33None (1)None None3-FWA-004-54-2 34None (1)None None3-FWA-004-54-2343-FWS-020-26-2None None3-FWA-006-102-2 35None (1)None None3-FWA-006-102-2353-FWS-020-26-2None None3-FWA-006-138-2 36None (1)None None3-FWA-004-127-3 37None (1)None NoneTABLE 3.6-35 PIPE WHIP EFFECTS AUXIL IARY FEEDWATER SYST EM (CONTINUED)Line DesignationBreak #Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-223Rev. 303-FWA-004-119-2 40None (1)None None3-FWA-004-119-2 41None (1)None None3-FWA-004-119-2 42None (1)None None3-FWA-004-48-2 43None (1)None None3-FWA-004-48-2433-FWS-020-22-2None None3-FWA-006-101-2 44None (1)None None3-FWA-006-101-2443-FWS-020-22-2None None3-FWA-006-137-2 45None (1)None None3-FWA-006-79-3 46None (1)None None3-FWA-004-117-2 47None (1)None None3-FWA-004-118-2 48None (1)None None3-FWA-004-119-2 49None (1)None None3-FWA-004-120-2 50None (1)None None3-FWA-004-41-2 51None (1)None None3-FWA-004-35-2 52None (1)None None3-FWA-004-47-2 53None (1)None None3-FWA-004-53-2 54None (1)None None3-FWA-004-122-3 55None (1)None NoneTABLE 3.6-35 PIPE WHIP EFFECTS AUXIL IARY FEEDWATER SYST EM (CONTINUED)Line DesignationBreak #Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-224Rev. 30NOTE:(1)Sustained blowdown of the Auxiliary Feedwater System does not o ccur. Dual inline check valves prevent backflow from the Main Feedwater System and the Auxiliary Feedwater Pumps have limited capacity to sustain th e system pressure. Therefore, pipe whip does not occur.3-FWA-003-66-3 56None (1)None None3-FWA-004-124-3 57None (1)None None3-FWA-003-71-3 58None (1)None None3-FWA-006-8-359None (1)None None3-FWA-004-126-3 60None (1)None None3-FWA-003-56-3 61None (1)None None3-FWA-004-128-3 62None (1)None None3-FWA-003-61-3 63None (1)None None3-FWA-008-14-3 64None (1)None None3-FWA-008-14-3 65None (1)None None3-FWA-004-160-3 66None (1)None NoneTABLE 3.6-35 PIPE WHIP EFFECTS AUXIL IARY FEEDWATER SYST EM (CONTINUED)Line DesignationBreak #Blowdown SourceEssential Pipe Whip TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-225Rev. 30TABLE 3.6-36 JET IMPINGEMENT EFFECTS AUXILIARY FEEDWATER SYSTEMS Line DesignationBreak #Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement3FWA*P1A Discharge Flange1None (1)N/AN/A N/ANone3-FWA-006-2-3 2None (1)N/AN/A N/A None (2)3-FWA-006-2-3 3None (1)N/AN/A N/A None (2)3-FWA-006-2-3 4None (1)N/AN/A N/ANone3-FWA-006-8-3 5None (1)N/AN/A N/A None (2)3-FWA-003-32-3 6None (1)N/AN/A N/A None (2)3-FWA-003-39-3 7None (1)N/AN/A N/ANone3-FWA-003-39-3 8None (1)N/AN/A N/A None (2)3-FWA-004-121-3 9None (1)N/AN/A N/A None (2)3-FWA-004-121-3 10None (1)N/AN/A N/ANone3-FWA-004-42-2 13None (1)N/AN/A N/A None (2)3-FWA-004-42-2 14None (1)N/AN/A N/A None (2)3-FWA-006-135-2 15None (1)N/AN/A N/ANone3-FWA-003-33-3 16None (1)N/AN/A N/ANone3-FWA-003-33-3 17None (1)N/AN/A N/A None (2)3-FWA-004-123-3 18None (1)N/AN/A N/A None (2)3-FWA-004-123-3 19None (1)N/AN/A N/ANone MPS3 UFSARMPS3 UFSAR3.6-226Rev. 303-FWA-004-36-2 22None (1)N/AN/A N/A None (2)3-FWA-004-36-2 23NoneN/AN/A N/A None (2)3-FWA-006-100-2 24None (1)N/AN/A N/ANone3-FWA-004-125-3 25None (1)N/AN/A N/ANone3-FWA-004-120-2 28None (1)N/AN/A N/A None (2)3-FWA-004-120-2 29None (1)N/AN/A N/A None (2)3-FWA-004-120-2 30None (1)N/AN/A N/ANone3-FWA-004-120-2 31None (1)N/AN/A N/A None (2)3-FWA-004-120-2 32 SplitNone (1)N/AN/A N/ANone3-FWA-004-120-2 32None (1)N/AN/A N/ANone3-FWA-004-120-2 33None (1)N/AN/A N/ANone3-FWA-004-54-2 34NoneN/AN/A N/A None (2)3-FWA-006-102-2 35NoneN/AN/A N/A None (2)3-FWA-006-138-2 36None (1)N/AN/A N/ANone3-FWA-004-127-3 37None (1)N/AN/A N/ANone3-FWA-004-119-2 40None (1)N/AN/A N/A None (2)3-FWA-004-119-2 41None (1)N/AN/A N/A None (2)TABLE 3.6-36 JET IMPINGEMENT EFFECTS AUX ILIARY FEEDWATER SY STEMS (CONTINUED)Line DesignationBreak #Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-227Rev. 303-FWA-004-119-2 42None (1)N/AN/A N/ANone3-FWA-004-48-2 43NoneN/AN/A N/A None (2)3-FWA-006-101-2 44NoneN/AN/A N/A None (2)3-FWA-006-137-2 45None (1)N/AN/A N/ANone3-FWA-006-79-3 46None (1)N/AN/A N/ANone3-FWA-004-117-2 47None (1)N/AN/A N/ANone3-FWA-004-118-2 48None (1)N/AN/A N/ANone3-FWA-004-119-2 49None (1)N/AN/A N/ANone3-FWA-004-120-2 50None (1)N/AN/A N/ANone3-FWA-004-41-2 51None (1)N/AN/A N/ANone3-FWA-004-35-2 52None (1)N/AN/A N/ANone3-FWA-004-47-2 53None (1)N/AN/A N/ANone3-FWA-004-53-2 54None (1)N/AN/A N/ANone3-FWA-004-122-3 55None (1)N/AN/A N/ANone3-FWA-003-66-3 56None (1)N/AN/A N/ANone3-FWA-004-124-3 57None (1)N/AN/A N/ANone3-FWA-003-71-3 58None (1)N/AN/A N/ANoneTABLE 3.6-36 JET IMPINGEMENT EFFECTS AUX ILIARY FEEDWATER SY STEMS (CONTINUED)Line DesignationBreak #Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSARMPS3 UFSAR3.6-228Rev. 30NOTES:(1) Sustained blowdown of the Auxiliary Feed water 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 th e requirement to evaluate the jet impingeme nt ef fects of Arbitrary Intermediate Breaks (AIBs). Therefore, jet impingement effects subsequent to AIBs are not evaluated.3-FWA-006-8-359None (1)N/AN/A N/ANone3-FWA-004-126-3 60None (1)N/AN/A N/ANone3-FWA-003-56-3 61None (1)N/AN/A N/ANone3-FWA-004-128-3 62None (1)N/AN/A N/ANone3-FWA-003-61-3 63None (1)N/AN/A N/ANone3-FWA-008-14-3 64None (1)N/AN/A N/ANone3-FWA-008-14-3 65None (1)N/AN/A N/ANone3-FWA-004-160-3 66None (1)N/AN/A N/ANoneTABLE 3.6-36 JET IMPINGEMENT EFFECTS AUX ILIARY FEEDWATER SY STEMS (CONTINUED)Line DesignationBreak #Essential Jet Impingement Target Distance to Target Jet Intensity at the Target Jet Load on the TargetProtection Requirement MPS3 UFSAR3.7-1Rev. 30
 
===3.7 SEISMIC===
DESIGNSections whose identification number includes the letter B contain material within balance-of-plant (BOP) scope, while sections whose identification number includes the letter N contain material within the nuclear steam supply system (NSSS) scope (including material pertaining to the reactor coolant loop piping out to the first weld of any connecting piping).
MPS3 UFSAR3.7B-2Rev. 30 3.7B.1 SEISMIC INPUT 3.7B.1.1 Design Response SpectraThe horizontal design response spectra used for seismic analysis are shown on Figures 3.7B-1 and 3.7B-2. The spectra for the safe shutdown eart hquake (SSE) correspond to a maximum rock acceleration of 0.17g, and the spect ra for the operating basis earthquake (OBE) correspond to a maximum rock acceleration of 0.09g. These spec tra are based on seismic analysis data determined during the construction permit stage. The vertical design response spectra are taken to be two-thirds of the horizontal design response spectra. Regulat ory Guide 1.60 does not apply as described in Section 1.8.
3.7B.1.2 Design Time History The horizontal design time hi stories produce spectra which envelop the horizontal design response spectra. These spectra are shown on Figures 3.7B-3 through 3.7B-8. The vertical design time histories are taken to be two-thirds of the horizontal desi gn time histories for both the SSE and the OBE.
Where soil structure interaction analysis is performed, the time history is applied at bedrock in the free field. Where analysis is performed using subgrade springs to represent rock stiffness, the time history is applied at the foundation level.
 
The horizontal design time history is checked by comparing its spectral values with those of the site design response spectra at 250 oscillator periods between 0.0167 second and 5.0 seconds. The oscillator periods are distri buted logarithmically accordi ng to the following expression:
T i = Ti-1 (3.7B-1) = (T 250/T 1)1/249 = (5/0.0167) 1/249 = 1.023 where: Ti-1 and T i = Any two consecutive oscillator periods T i/Ti-1 = The period ratio (ratio between the i th period, T i , and the (i-1) period, Ti-1).3.7B.1.3 Critical Damping Values The values of the percentage of critical damp ing used in the analysis of Seismic Category I structures, systems, and components depends on the stress levels resulting from the seismic input motion (SSE or OBE) used in the analysis.The values of damping used for the input motion are listed in Table 3.7B-1. Structural damping is assigned to be 2.0 percent for th e OBE and 5.0 percent for the SSE.The higher damping values (Table 3.7B-1) are used only where ju stified by detailed study (either testing or calculated stress levels). Damping values utilized in th e analysis of Seismic Category I MPS3 UFSAR3.7B-3Rev. 30 structures, systems, and components meet the intent of Regulatory Guide 1.61 which is not applicable as described in Section 1.8.
3.7B.1.4 Supporting Media for Seismic Category I Structures The founding materials for major plant structures are listed in Table 2.5.4-14. Most of the major safety related structures are founded on bedroc k, with the exception of the control building, emergency diesel generator build ing, and the hydrogen recombiner building. The control building is founded on 1 to 4 feet of compacted structural backfill overlying basal till of thickness varying between 1 foot on the east side and 15 feet on the west. The emergency diesel generator building is founded on basal till varying in thickness from less than 10 feet to 30 feet overlying bedrock.
The hydrogen recombiner is founded on concrete fill overlying bedrock.A large portion of the circulating water discharg e tunnel and the service water intake lines are founded on bedrock. However, some sections are founded on soil, particularly near the intake and discharge points in the vicinity of Niantic Bay. When soil was encountered as a founding material, all unsuitable overburden was removed to sound ablati on or basal till. In the event that the invert elevation was higher than the excavated grade, compacted structural backfill was placed in thin lifts to the subgrade elevation in accordance with procedures described in Section 2.5.4.5.2.
A description of each of the founding materials at the site is presented in Section 2.5.4.2. The seismic velocities and moduli va lues of these materials are presented in Section 2.5.4.4.3. The properties of structural backfi ll materials are described in Section 2.5.4.5.2. The use of these properties in the analysis of soil-structure interaction is discussed in Section 3.7B.2.4.
3.7B.2 SEISMIC SYSTEM ANALYSISSeismic systems discussed in this section are those major Category I structures that are considered in conjunction with foundation media in forming a soil-structure interaction model. Other systems and components are consid ered seismic subsystems (Sections 3.7B.3 and 5.4.14).
3.7B.2.1 Seismic Analysis MethodsGenerally, analysis of seismic systems were performed using eith er the modal analysis response spectra method or the modal analysis time hist ory method. Response spectrum analysis uses the natural frequencies, mode shapes, and weight ed modal dampings to determine the maximum seismic response of multi-degree-of-freedom systems with lumped masses and elastic connecting members. Time history modal analysis uses the same free vibration char acteristics and damping factors as the spectrum analysis, primarily to de termine acceleration respons e spectra at selected locations within the structure. Those buildings that required finite element soil-structure interaction were analyzed using the frequenc y domain time history method (Section 3.7B.2.4). The methods used for seismic analysis of particular Seismic Category I structures are summarized in Table 3.7B-2.Mathematical models of three buildings representative of Se ismic Category I structures are discussed in detail. Models selected are the containment structure (Figure 3.7B-9), the main steam valve building (Figure 3.7B-10), and the emergency generator enclosure (Figures 3.7B-11 and 3.7B-12). These buildings are de scribed in Sections 3.8.1 and 3.8.3.
MPS3 UFSAR3.7B-4Rev. 30Foundation effects, namely torsion, rocking, a nd translation, were included in the dynamic analysis as described in Section 3.7B.2.4. For structures founded on shallow soil, a finite element soil-structure interaction analys is was performed. In this case, a planar model was employed in each of the horizontal directi on. All other Seismic Category I structures were analyzed considering six degrees-of-fr eedom; three translat ion, two rocking, and one torsional.
Those buildings that were modeled using finite element soil structur e interaction techniques included the control building and the emergency ge nerator enclosure. The computer program for these analyses was PLAXLY. Structures founde d on rock, which are discussed in Section 3.7B.2.4, were analyzed using discrete springs s uggested by Richart et al., 1970, Whitman et al., 1967, Wass 1972, and Kausel et al., 1975. Springs asso ciated with the degrees-of-freedom in a global orthogonal coordina te system were used.
In developing the lumped-mass mode ls for Seismic Categor y I structures, an ad equate number of degrees-of-freedom was ensured by providing a lumped mass at the mat, floor, and roof elevations. Additional masses were used when dictated by changes in geometry or stiffness. In the cases of the containment shell, eight lumped masses were selected at intervals uniformly distributed along the height of the shell. Studies have been conducted which show the difference in base shear and overturning moment between a 3-mass and a 15-mass model to be less than 3 percent. Comparisons between 3-, 5-, 10-, and 15-mass models have shown a difference in natural frequency for the first two m odes of less than 1 percent.
For analysis of the models, twi ce the number of modes with frequencies less than 33 Hz were considered.
 
Each Seismic Category I building has a foundati on mat which serves as a single structural support, except the emergency ge nerator enclosure which has spread footings that serve the same purpose. Two inch (above grade) and 1 inch (below grade) shake spaces, larger than any predicted structure-to-structure displaceme nts, are provided between all Seismic Category I structures. In addition, 4 inch shake spaces ar e provided between all Seismic Category I structures and the containment structure. Relative displacements were considered for all Category I systems and components which run between structures. Seismic effects such as hydrodynamic loads and nonlinear responses were considered where appropriate (Section 2.5.4).
3.7B.2.2 Natural Frequencies and Response LoadsResponse spectrum analyses for the containment structure and the main steam valve building, are summarized below, as well as the finite elem ent soil structure interaction analysis of the emergency generator enclosure. They are repr esentative samples of Category I structures.
3.7B.2.2.1 Containment Summary The response spectrum analysis of the contai nment and internal structures included the enveloping of two discrete analyses. The dynami c model of the containm ent (Figure 3.7B-9) was analyzed with cracked and unc racked concrete properties.
The significant natural frequencies and modal participation factors result ing from the uncracked and cracked analyses are presented in Tables 3.7B-3 and 3.7B-4. Significant mode shapes resulting from these analyses are presented in Tables 3.7B-5 and 3.7B-6 and the response loads obtained by the square root of the sum of th e squares method (SRSS) are summarized in Tables 3.7B-7 and 3.7B-8. Table 3.7B-9 presents the degrees-of-freedom corresponding to x, y, MPS3 UFSAR3.7B-5Rev. 30and z direction and rotation about the x, y, and z ax is for the containment and internal structures response loads obtained by the SRSS method. Final lumped mass accelerations and displacements are obtained by taking the absolute sum of the SR SS modal responses due to each direction of excitation and enveloping the cracked and uncracked analysis (Tables 3.7B-10 and 3.7B-11).
The amplified response spectra (ARS) were produc ed using the modal analysis time history method. Typical ARS resulting fr om the enveloping of the cracke d and uncracked analyses are provided at the mat, steam generators support slab, top of the primary shield wall, operating floor and crane wall of internal structure, as well as the springline and dome apex of containment structure for horizontal and vertical safe shutdown earthquake (SSE) excitations (Figures 3.7B-13 through 3.7B-33).
3.7B.2.2.2 Main Steam Valve Building SummaryThe dynamic model of the main st eam valve building (Figure 3.7B-10) was excited separately by one vertical and two orthogonal horizontal excitations. The resu lting natural frequencies, participation factors, and mode shapes are listed in Tables 3.7B-12 and 3.7B-13. The closely spaced modes (CSM) modal responses are given in Tables 3.7B
-14 and 3.7B-15. The CSM modal combination method is explaine d in Section 3.7B.2.7. Final lu mped mass accelerations and displacements are obtained by taking the absolute sum of the CSM modal responses due to each direction of excitation (Table 3.7B-16). For the main steam valve building analysis, Table 3.7B-17 presents the relationship between the degrees-of-freedom and the motion of the lumped masses given in Ta bles 3.7B-14, 3.7B-15, and 3.7B-16.
Examples of the ARS resulting from each di rection of excitation produced using the modal analysis time history method are provided at th e mat (Figures 3.7B-34 thru 3.7B-36), the floor slab at el 41 feet 0 inch (Figures 3.7B-37 thru 3.7B
-39), and the roof slab at el 85 feet 4 inches (Figures 3.7B-40 through 3.7B-42).
3.7B.2.2.3 Emergency Generator Enclosure SummaryThe seismic analysis of the emergency generator enclosure used a lumped mass finite element model (Section 3.7B.2.4). The dynamic model shown on Figures 3.7B-11 and 3.7B-12 was excited separately by one vertical and two or thogonal horizontal excitations. Since the method used for the analysis was a frequency domain so lution rather than a modal analysis, no natural frequencies, participation factors, or mode shapes are tabulated. The solution gives transfer functions which represent the ratio of the amplitude of any partic ular mass point of the structure to the amplitude of the input motion at bedrock. Typical transfer functions (including imaginary part) for each mass point of the dynamic model are presented on Figures 3.7B-43 thru 3.7B-51 for each direction of excitation. Final absolute sum response loads are presented in Table 3.7B-18 and 3.7B-19. Also, since diesel generators are located on independent foundations, a separate analysis was performed for them.Typical ARS at elevation 24 feet 6 inches, 51 feet 0 inches, and 66 feet 0 inches, produced using the frequency domain time history method, ar e shown on Figures 3.7B-52 thru 3.7B-60. ARS resulting from the independent anal ysis of the diesel generator isolation mats are presented on Figures 3.7B-61 and 3.7B-62. Examples of the input time history at bedrock of the plant site and the resulting time history at the base of th e structure are shown on Figures 3.7B-63 and 3.7B-64.
MPS3 UFSAR3.7B-6Rev. 30 3.7B.2.3 Procedures Used for Analytical Modeling The dynamic model of a Seismic Category I struct ure is constructed to obtain a satisfactory representation of the dynamic behavi or of the structure. Major Seis mic Category I structures that are considered in conjunction with foundation media in forming a soil-structure interaction model are defined as seismic systems. Other Seismic Category I systems and components that are not designated as seismic systems are considered as seismic subsystems (R). In most cases, equipment and components come under the definiti on of seismic subsystems and are analyzed as a decoupled system from the primary structure. To define criteria for decoupling subsystems, R m , the mass ratio and R f , the frequency ratio are significant. R m and R f are defined as: (3.7B-2) (3.7B-3)The following criteria are used.1.If R m < 0.01, decoupling is acceptable for any R f ,2.If 0.01  R m  0.1, decoupling is acceptable if R f  0.8 or if R f  1.25,3.If R m > 0.1, an approximate model of the subsystem is included in the primary system model.If the subsystem is comparatively rigid and also rigidly connected to the pr imary system, only the mass of the subsystem is included at the support point in the primary system model. In the case of a subsystem supported by very flexible connect ions (e.g., a pipe supported by hangers), the subsystem is not included in the primary mode l; however additional mass is considered.
In most cases, the equipment and components wh ich come under the definition of subsystems were analyzed separately from the primary structure, and the seismic input for the subsystem was obtained from the analysis of the structure. On e important exception to this procedure is the analysis of the reactor coolant system (RCS), which was considered to be a subsystem but was analyzed using a coupled model of the RCS a nd primary structure with ground motion as the seismic input (Section 5).
The dynamic model of a seismic system consists of a set of lumped mass es, generally having six degrees-of-freedom per mass conne cted by weightless elastic members. Masses are usually lumped at floor levels and in clude the masses of the floors, walls, columns, equipment, and piping. The floors are treated as ri gid diaphragms that transfer th e earthquake inertia forces to frames and shearwalls, which in turn transfer the loads to the foundation mat and the subgrade. Beam theory, combining the effects of shear, fle xure, torsion, and axial deformation, is used to establish the stiffness characteristics of the frame-wall systems. Eccentricities between the centers of mass (CM) and centers of rigidity (CR) are considered.As an example, the lumped mass model of the containment is shown on Figure 3.7B-9. The model was constructed so that it properly represents the free vibration of a cantilevered structure in shear R m Total mass of the supported subsystem Mass that supports the subsystem       
----------------------
-----------------
------------------
------------------
-----------------
-=R f Fundamental frequency of the support subsystem Frequency of the dominant support motion       
--------------------
----------------------
---------------------
---------------------
------------------
---------------
-=
MPS3 UFSAR3.7B-7Rev. 30and flexure. The model consists of a system of lumped masses, each with six degrees-of-freedom, connected by weightless members. The base of the model is connected to the subgrade represented by springs as discus sed in Section 3.7B.2.4. Masses M 10 through M 17 represent the dome and the cylindrical portion of the containment shell and the dome. M 1 consists of the mat and portions of the walls and columns which are attached to the mat. The internal structure, consisting of equipment, primary shield wall, cubicle walls, cr ane wall, etc., were modeled by masses M 2 through M 9.3.7B.2.4 Soil-Structure Interaction The supporting media for Seismic Category I stru ctures has been discussed in Section 3.7B.1.4. As indicated in Table 2.5.4-14, Seismic Category I structures are founded directly on bedrock with a few structures founded on sh allow soil overburden over rock.Finite element soil-structure interaction analysis is required for shallowly embedded structures on shallow soil overburden over rock (NRC 1975). The control building and the emergency generator enclosure were analyzed using the finite element method.For the finite element analysis, the emerge ncy generator enclosure lumped mass model (Section 3.7B.2.3) was attached to a finite element representation of the soil subgrade (Figures 3.7B-11 and 3.7B-12). These models were then subjected to horizontal and vertical excitations by computer program, PLAXLY. The bottom boundary was taken at bedrock and the side boundaries used PLAXLY's energy transmitti ng boundary corresponding to layers of soil extending laterally to infinity. Th e material properties of the soil were assumed to be linearly viscoelastic, and the dynamic equations were so lved in the frequency domain using Fourier transformation techniques. Nonlin ear behavior in the subgrade wa s accounted for by the use of the computer program SHAKE, which determines the stain corrected soil properties.
Section 3A.1.7 of Appendix 3A give a detailed discussion of PLAXLY.
For structures founded directly on bedrock anal yses using discrete subgrade springs based on equations suggested by Richart et al. (1970), Whitman et al. (1967), Wass (1972), and Kausel et al. (1975) were perf ormed. A value of 10 percent (transla tional) and 5 perc ent (rotational) maximum subgrade damping assured the conser vatism of the disc rete spring method.
3.7B.2.5 Development of Floor Response SpectraFloor response spectra were developed separately for one vertical and two mutually perpendicular horizontal earthquake motions using the modal analysis time history method discussed in Section 3.7B.2.1 for all buildings except the emer gency generator enclosure. Floor response spectra for the emergency generator enclosure we re developed using the modal analysis method as in PLAXLY (Section 3.7B.2.4).
Combination of the three components of floor response spectra for use in subsystem seismic analysis is discussed in Section 3.7B.3.
MPS3 UFSAR3.7B-8Rev. 30 3.7B.2.6 Three Components of Earthquake MotionProcedures to combine three components of ea rthquake motion to determine the response of seismic systems meet the recommendations of Regulatory Guide 1.92, Rev. 1, to the extent described in Section 1.8.1.92.
3.7B.2.7 Combination of Modal Responses In general, for those Seismic Category I syst ems analyzed by the modal analysis response spectrum method, the influence of closely spac ed modes was considered by use of the "double sum method," described in Regulatory Guide 1.92, Rev. 1, to the extent described in Section 1.8.1.92 and in Singh et al. (1973).
3.7B.2.8 Interaction of Non-Category I Structures with Seismic Category I StructuresAdjacent Seismic Category I struct ures are separated fr om Non-Category I stru ctures by a 1-inch Shake space below grade and a 2-inch Shake space above grade.Non-Category I structures are sufficiently remote from Seismic Category I structures to preclude interaction or are designed (Section 3.8.3) to a void collapse during an SSE, thus preventing significant damage to adj acent Category I structures.
3.7B.2.9 Effects of Parameter Variations on Floor Response SpectraIn order to consider the effect s of variations of structural properties, dampings, and soil properties, peak resonant period values of the floor response spectra were spread +15 and -15 percent. For the containment structure whic h experiences substantial cracking during the structural acceptance test, floor response spec tra for the cracked and uncracked cases were enveloped.
The use of floor time histories in subsys tem analysis is disc ussed in Section 5.4.14.
3.7B.2.10 Use of Constant Vertical Static FactorsDynamic vertical responses were calculated in all seismic system analyses, precluding the need to use constant vertical static factors.3.7B.2.11 Method Used to Account for Torsional EffectsThe emergency generator building was analyzed using planar models for two horizontal directions using three de grees-of-freedom per mass. Since this building is basically symmetrical, this approach is justified. All other Seismic Category I structures we re analyzed dynamically using six degrees-of-freedom per mass. Effects of eccentricity between centers of mass and centers of stiffness or soil contact were included in determining appropriate member stiffnesses in the dynamic models.
MPS3 UFSAR3.7B-9Rev. 30 3.7B.2.12 Comparison of Responses As discussed in Section 3.7B.2.1, the results of th e response spectrum analysis were used for the design of most Seismic Category I structures and the results of the modal time history analyses provided the floor time historie s used to create floor response spectra. The maximum responses attained from the time history analysis are compared with those resulting from the response spectrum technique in Tables 3.7B-20 through 3.7B
-22 for the containment structure and main steam valve building.
The higher time history results are attributable to the fact that the artificial gr ound time history created to envelop the Millstone 3 ground response spectra is quite conservative. The extent of this conservatism is illustrated on Figures 3.7B-3 through 3.7B-8 which compares the Millstone 3 ground response spectra with the response spectra generated by using the si te artificial time history as the forcing function. However, becaus e of the method used in combining structural responses from the response spectrum analysis due to each of the three components of earthquake motion (absolute sum), a majority of the res ponse accelerations from the response spectrum method are larger than those fr om the time history analysis.
3.7B.2.13 Methods for Seismic Analysis of Category I Dams There are no Category I dams at this site.
3.7B.2.14 Determination of Seismic Category I Structure Overturning MomentsOverturning moments for Seismic Category I structures were calculated using inertia forces based on the accelerations su mmarized in Secti on 3.7B.2.2 as follows:
(3.7B-4)where OM = Overturning moment i = Mass point on structure M i = Mass at i a hi = Horizontal acceleration at i in the direction being investigated a vi = Vertical acceleration at i H i = Height of point i above bottom of foundation mat D i = Distance from point i to edge of founda tion mat in the direction of accelerationOMM i a hi H i M i a vi D i M a+i1=n+i1=n=
MPS3 UFSAR3.7B-10Rev. 30 n = Number of Mass points in structure Four directions of horizontal acceleration were considered: nor th, south, east, and west. The overturning moment was calculated for each direction of horiz ontal acceleration assuming the vertical acceleration to act either up or dow
: n. The maximum moment calculated from the different combinations of the horizontal and vertical accelerations was then divided into the resisting moment based on the deadweight of the structure to determine the factor of safety against overturning. The minimum factors of sa fety allowed are summa rized in Section 3.8.5.5.
3.7B.2.15 Analysis Procedure for Damping An equivalent viscous modal damping, reflecting the different damping rates in various portions of the structure, was computed for use in structural dynamic anal ysis. The modal damping ratio is a weighted average of member and support spring damping, based on the contribution of each to the total strain energy of the mode shape. The method is based on work by Roesset et al. (1973).
The following discussion is meant to serve as a general description. The more theoretical derivations (Roesset et al., 1973, Whitm an 1970) are found in the literature.Damping and Strain Energy MethodsAn important factor in determining structural response is the damping phenomenon. Two types of damping are generally recognize d: viscous (in which the energy dissipated per cycle is proportional to frequency) and hysteretic (in wh ich no frequency dependence is seen). Most structural elements display hysteretic behavior, while supporting soils appear to combine both hysteretic and viscous damping mechanisms.
For certain applications (e.g., pipe breaks) in which foundation motion can be neglected, only hysteretic damping need be considered. Whitman's (1973) analysis of Biggs' formula gives a useful approximation for the dampi ng of each mode when material damping varies from element to element. His expression for the equivale nt viscous modal damp ing is obtained by a strain-energy weightin g of element damping:
(3.7B-5)where: = Equivalent viscous damping ratio (fraction of critical) for structure vibrating in mode j, N = Number of elements, D i = Hysteretic damping ratio for element i, Beqv j D i E j i i1=NE j i i1=N------------
----------=Beqv j MPS3 UFSAR3.7B-11Rev. 30 j E i = Strain energy in element i when deflected into mode shape j.In particular, when damping is uniform (i.e., D i = D) then B eqv = D for all modes. When damping is not uniform, modal damping is weighted toward those elements which make the largest contribution to the energy of each mode. In ot her applications (e.g., earthquakes) in which foundation motion is significant, viscous damping also must be considered. Current practice treats the soil damping as viscous for translational motion, and hysteretic for rotational motion. Roesset et al. (1973) extended Biggs' form ula to include the viscous damping contributions of the soil:
(3.7B-6)where: N H = Number of hysteretically damped elements, N V = Number of viscously damped elements, W j = Frequency of structure m ode j (radians per second), W K = Frequency of element k (radians per second), B K = Critical damping ratio of element k at frequency k.
The square of the natural frequency w of the soil element, is equal to the ratio of the soil spring stiffness to total mass of the structure plus foundation.This formula reflects the fact that the energy dissipation per cycle by the viscous mechanism is proportional to frequency of motion. Any element which displays both hysteretic and viscous damping appears in both summations of the numerator, but is not repeated in the denominator.Each element strain energy appearing in Equation 3.7B-4 is evaluated from the element stiffness matrix and the displacement of the element's boundary joints. After comparing the Biggs and Roesset modal damping ratios cal culated from Equations 3.7B-5 and 3.7B-6, the lower value for each mode is selected for use in the dynamic an alysis, thus assuring that the composite modal damping value never exceeds the hysteretic damp ing value. In no case do modal damping values exceed 10 percent.
Beqv j D i E j i i1=NW j W k--------B k E j k k1=N v+E j i E j k i1=N v+i1=N----------------
-----------------
-----------------------------
=
MPS3 UFSAR3.7B-12Rev. 30 3.7B.3 SEISMIC SUBSYSTEM ANALYSIS 3.7B.3.1 Seismic Analysis Methods 3.7B.3.1.1 Equipment and Components Seismic Category I equipment and components are documented for seismic adequacy. The basic source of seismic design data is either the gr ound response spectra, the amplified response spectra, floor, or mat time history, derived th rough a dynamic analysis of the relevant structure (Sections 3.7B.2.5 and 3.7B.2.9).
Three principal methods, which include combinations of these methods, are used for documenting adequacy for Seismic Categor y I equipment and components:1.Static analysis2.Dynamic analysis3.TestingThe effects of supports are reflected in the input to seismically qualified equipment and components. General stress limits are given in Section 3.9B.3.1. Specific component stress allowables are given in Tables 3.9B-5, 3.9B-6, and 3.9B-7. Such lim its either conform to, or are more conservative than, those of ASME Section III, Subsection NF.
Laboratory tests are performed on Seismic Category I mechanical a nd electrical equipment, and complex instrumentation that cannot be modeled to predict response correctly. Conversely, analytical methods described in this sect ion are employed when mathematical modeling techniques are used or when equipment charac teristics (e.g., size, rati ng) preclude laboratory testing. Combinations of analysis and testing ar e also employed to assure adequacy of Seismic Category I equipment.
The safe shutdown earthquake (SSE) in combination with the faul ted loads are the load basis which assure the structural integrity of Se ismic Category I equipment. The operating basis earthquake (1/2 SSE) in conjuncti on with the operating loadings ar e used to assure continued operation of the equipment and components.
 
Equipment vendors and suppliers are required to formulate programs for qualification of Seismic Category I equipment in accordance with specif ication requirements. Documentation of the seller's qualification program is reviewed and a pproved by the Applicant.
3.7B.3.1.1.1 Static AnalysisStatic analysis is used for equipment and com ponents that can be characterized as relatively simple structures. This type of analysis i nvolves the multiplication of the equipment, or component, total weight by the specified seismi c acceleration (direction dependent loading), to produce forces that ar e applied at the center of gravity in the horizontal and vert ical directions. A stress analysis of equipment co mponents, such as supports, holddow n bolts, and other structural members, is performed to determine their adequacy.
MPS3 UFSAR3.7B-13Rev. 30 In the specification of equipment for static analysis, two or more sets of acceleration data are provided. The choice of wh ich set to use is dependent upon the equipment's fundamental natural frequency. The relevant response curves are reviewed to determine a cutoff frequency which bounds the rigid range from the resonance ra nge of the response curves. Equipment and components having fundamental natural frequencies above the cutoff frequency of the relevant response curve are analyzed to ri gid range response accelerations.
For components or equipment having a fundamental natural frequency below the cutoff frequency, the accelerations used in static analysis are 1.3 times the peak acceleration value, as indicated in the amplified res ponse curve for simply supported conditions. For equipment having multi-support points in a single plane, the res onant response acceleration is multiplied by 1.5.
Each of the three defined directions of earthqua ke input (two horizontal and one vertical taken orthogonally) is evaluated separately. Horizontal and vertic al seismic loads are combined using the square root of the sum of the squares. E quipment is designed to withstand the combined effects of normal operating loads, acting simultaneously with th e earthquake loadings, without loss of safety function or structural integrity.
3.7B.3.1.1.2 Dynamic Analysis A detailed dynamic analysis is performed when equipment complexity, or dynamic interaction, precludes static analysis, or when static analysis is too conservative. Dy namic analysis methods include:1.Response spectrum modal analysis2.Time-history by modal superposition3.Time-history by numerical integration The response spectrum modal analysis technique is used most commonly.
Modeling The lumped mass, or the consistent mass, appro ach is employed in the dynamic analysis. In the lumped and the consistent mass idealizations, the main structure is divided into substructures and the masses of these substructures are concentrated at a number of discrete points. The nature of these substructures, and the stiffness prope rties of the corresponding modeling elements, determine the minimum spacing of the mass points and the degrees of freedom to be associated with each point. In accordance with minimum spacing requirements, the analyst can then choose, for the model, particular mass points which reflect predominant masses of subcomponents which contribute significantly to the total response.
Modeling of equipment fo r dynamic analysis starts with the calculation of lumped masses at discrete stations and the evaluati on of the elastic properties (or stiffness) of connecting members. The number of discrete mass points is selected to adequately describe the dynamic characteristics and natural frequencies of the equipment. General modeling guidelines indicate that good natural frequency characteristics are obtained when the number of discrete mass or node points selected is twice the number of the highest mode of interest. Papers pr epared by Lin and Hadjian (1976),
MPS3 UFSAR3.7B-14Rev. 30 Johnson (1977), and Lin (1974), along with Sections 3.7B.2.1 a nd 3.7B.2.3, provide supplemental modeling guidance.
Response Spectrum Modal Analysis The normal mode approach is employed for seis mic analysis of equipment and components.
Natural frequencies, eigenvectors, participation factors, and modal member-end forces and moments of the undamped structure are calculate
: d. The system of equations which describes the free vibrations of an n-degree of freedom, undamped structure is:
(3.7B.3-1) where:[M] = Mass matrix for assembled system[K] = Stiffness matrix for assembled system
= Nodal acceleration vector
{X} =Nodal displacement vector The mode shapes and frequencies are solved in accordance with:
(3.7B.3-2) where:n = Natural frequency of the nth mode
[]n = Mode shape vector for the nth mode n = Number of significant modes consideredEigenvector-eigenvalue extraction techniques, such as Householder-QR, Jacobi Reduction, and Inverse Iteration, are used, de pending upon the total number of dynamic degrees of freedom and the number of modes desired.
The modal participation factor for the nth mode specific direction, i, is defined by (3.7B.3-3) where:[]T = Transpose of mode shape vector for the nth mode
[] = Earthquake direction vector referring to direction iThe modal member-end forces and moments are determined by M[]X**{}K[]X{}+0=X**{}K[]2 n M[]-() []n 0=n123N ,,,=ni[]T  M[][][]T  M[][]---------------------------------
=
MPS3 UFSAR3.7B-15Rev. 30
[F n] = [K m] []n (3.7B.3-4) where:[K m] = Member stiffness matrixFor each modal frequency, the corresponding response acceleration is determined for a given level of equipment damping from the applicable response curve.
The maximum response for each mode is found by computing:
(3.7B.3-5) where: = The modal acceleration
= The velocity
{X} = The displacement{F} = The member-end force and moment vectors R ni = The spectral acceler ation for the nth mode The basis for the combination of maximum m odal response is discussed in Section 3.7B.3.7.Time-History Methods There are two separate approaches to the solu tion of the equations of dynamic equilibrium for time-history motions. The modal superposition method involves the modal solution of the free vibration response of the system, and transformation to normal coordinates using the mode shapes of the system. This procedure uncouples the equa tions of motion so that the response of the system in each individual mode may be evaluated independently. Total system response is determined by combining responses of individua l modes oscillating simultaneously. The second method of time-history analys is is the direct integration solution that includes numerical integration of the simultaneous differential equations of dynamic equilibrium without transformation to normal coordinates. System re sponse at each time point is evaluated by this technique.
X**{}nni R ni[]n=X*{}1n------X**{}n=X{}n 12 n---------X**{}n=F{}nn R ni2 n-------------
-F n{}n=X**{}X*{}
MPS3 UFSAR3.7B-16Rev. 30Time-history solutions, due to their analytical complexity, are only used when results of other methods are too conservative and when acceptable numerical solutions are available in a computer program. Equations of motion for time-history solutions are contained in Section 3.7B.2.1. Computer progra m capabilities for time-history solutions are outlined in Appendix 3A, Section 3.2.
3.7B.3.1.1.3 Testing Equipment and components that use seismic testi ng as their qualification basis conform to the following general instructions for earthquake testing. These re quirements conform with other applicable industry standards such as I EEE 344-1975, Section 3.10, or provide guidance for testing where no such standards are available.
Equipment packages or components are shown to be seismically adequate by being tested individually, as part of a simulated structural section, or part 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 a nd anchorages) that would compromise safety related component integrity.3.No adverse or maloperation before, duri ng, or after the proposed test that could result in an improper safety action.
Equipment vendors and suppliers are required to formulate programs for qualifying the equipment in accordance with the conditions specified in the seismic design requirements contained in the equipment specifications. The vendor must submit a summary of the proposed effort for review and approval.
 
The characteristics of the test ing input at the equipment m ounting locations are defined by amplified response spectrum curves, zero period re sponse curve levels, time-history motions, or combinations of these as applicable.
The use of single and multifrequency input testing is accepted as a method of seismic qualification, based upon the particul ar plant site, structure, and floor response ch aracteristics. Structures, particularly at lower elevations, e xhibit a broad frequency range response similar to the ground motion during an earthquake. This broad range frequency motion is filtered at higher structural elevations and response becomes more sinusoidal in nature. Knowledge of the floor response characteristics of the structure and res ponse characteristics of the equipment generally dictate the requirements for testing. Periodic testi ng is applicable where periodic floor motion is indicated and, conversely, random input testing is most applicable for broad frequency range input to components. Periodic testing can be used to e nvelop multiple peak floor responses, as well as single peak, providing sufficiently high forci ng is used. For equipment exhibiting multiple response modes, single frequency i nput may be used, providing the input has sufficient intensity to envelop the floor response spectra of the individual mode s of the equipment.
The testing machine (fixture) setup is arranged so that the equipment tested is mounted to simulate, to the extent possible, the actual service mounting, with no dynamic coupling to the test item. Equipment is tested in the operating c ondition wherever possible, and functions are MPS3 UFSAR3.7B-17Rev. 30 monitored and verified both during and after testing. However, a true operating environment is sometimes not obtainable for equipment such as pumps, etc.
The in situ application of vibratory devices, to superimpose the seismic vibratory loadings on the complex active device for operabi lity testing, is acceptable when application is justifiable.
The test program may be based upon selectively testing a repres entative number of mechanical components according to type, load le vel, size, etc., on a prototype basis.
In addition to the single and multifrequency te sting programs outlined, la boratory shock results, in-shipment shock data, or adequate historical dynamic adequacy data (i.e., previous relevant test or environmental data) are also given considerat ion. The test method sel ected must demonstrate the adequacy of principal structural a nd functional capability of the equipment.General testing guidance criteria specif ied for equipment include the following:1.Single Frequency TestingTesting is performed for as much of the range between 1 and 33 Hz as practicable or justified. Input for qual ification should, as a minimum, equal the zero period response curve level.2.Sinusoidal Inputa.A frequency scan (two octaves pe r minute, maximum), at a constant acceleration level, is performed over the frequency range of interest. The objective of this test is to dete rmine the natural frequencies and amplification factors of the tested equipment, and its critical components or appurtenances, and to ensure genera l 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 sele cted if amplificat ion 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 fu rther testing. Intervals of one-half octave or less are employed for this spacing.3.Sine Beat InputA 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 MPS3 UFSAR3.7B-18Rev. 30 performed at natural frequencies, and bands of large amplification identified during the sine scan. The duration and pe ak amplitude of the beat for each 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 pr ogram consists of applying a series of sine beats at geometrically-spaced frequency interval s of one-half octave or less over the frequency range of interest. The peak am plitude of the beat employed is, as a minimum, the maximum rigid range ac celeration 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 just ified. 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 freque ncy range of 1 to 33 Hz is normally considered.5.Time-History Input An acceleration time-history of the equi pment support location, or one based on a synthesized response curve, may be us ed as testing machine input. A 15- to 30-second time history input is normall y 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 perf ormed so that the applicab le response spectrum curve is enveloped by that produced by the ta ble motion. The input is controlled by one-third octave (or less) bandwidth filt ers 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 curv e 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 conj unction with other waveforms may be employed as justified.
MPS3 UFSAR3.7B-19Rev. 307.Complex Wave Test A complex wave test may be performed by subjecting the equipment to an input motion, generated by summing a group of de caying 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 mu st be shown to envelop th e relevant response curve.
3.7B.3.1.2 Piping Systems Analyses of Seismic Category I (including all ASME Code Cla sses 1, 2, and 3 piping systems) piping are performed by the modal analysis re sponse spectra method.
Nonseismic piping is seismically analyzed when its failure could result in unacceptable damage to a Seismic Category I system. Either a modal analysis response spectra method or equivalent static load method of analysis is utilized for the latter.
 
The criteria and procedures used for mode ling are described in Section 3.7B.3.3.2. A typical mathematical model of a piping system is shown on Figure 3.7-65. Defined boundaries, such as equipment and pipe anchors, may be considered as isolating the piping system when it is undergoing earthquake excitation.
The modal analysis response spectra method is us ed for dynamic analysis of Seismic Category I piping. Input for these piping sy stems 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 broadeni ng 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 obt ained 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 differen tial piping support movements at different suppor t points are discussed in Section 3.7B.3.8
.Where a piping system is subjec ted to more than one response spectrum, as when support points are located in different pa rts of the structure or in separate structures, an enve loping procedure as well as peak broadening is applie d to generate a composite, or worst-case, spectrum for analysis.
Peak broadening of minus 15 percent and plus 15 percent of peak freque ncies is provided to account for uncertainties in the ca lculated values of structural frequencies. Accordingly, piping systems designed using those am plified response spectra having natural frequencies within +/-15 MPS3 UFSAR3.7B-20Rev. 30 percent of the peak resonant frequency are assi gned the peak response value(s). Outside this range, the amplified response spectra is used exactly as stated. The response spectra modal analysis provides peak response quantities for each mode which are then combined according to Section 3.7B.3.7. All significant dynamic modes of responses under seismic excitation with frequencies less than 50 cps or m odes less than 50, whiche ver is reached first, are included in the dynamic analysis describe d in Section 3.7B.3.8. The combined se ismic responses, together with internal forces and moments due to seismic anc hor movements, are then combined with other loadings according to ASME Section III C ode, Articles NB 3600 (Class 1 piping), NC 3600 (Class 2 piping), or ND 3600 (Class 3 piping).
Time-history modal superposition analysis is employed for fluid-induced transient dynamic problems (e.g., water hammer and steam hammer), but is not normally used for piping seismic analysis.
Small size seismic Category I piping systems (S ection 3.7B.3.5.2) are seis mically qualified, in part, by the application of standa rd span procedures. The standard span procedures are restricted to small bore piping systems (one inch and be low ASME Class 1 and two inch and below ASME Class 2, 3, and ANSI B31.1) which meet the criteria of the prequalified analysis.
No tests or empirical methods are used in lieu of analytical methods fo r all Seismic Category I piping.
Detailed descriptions of analyt ical procedures and design criter ia for Seismic Category I piping can be found in Section 3.7B.3.8. The type of seismic analysis used and the criteria used are summarized in Table 3.7B-26.
3.7B.3.2 Determination of Number of Earthquake Cycles 3.7B.3.2.1 Equipment and ComponentsASME III (NB 3112,3b) requires that the number of ea rthquake cycles to be used in the analysis of ASME Code Class 1 component s be specified as part of th e design mechanical loads. The following criteria are used for all component s 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 pe r earthquake may be obtained from the structural time-history analysis.
3.7B.3.2.2 Piping Systems All ASME Class 1 piping systems are designed fo r a minimum of 10 maximum stress cycles per seismic event in the analysis. A total of five OBE and one SSE is assumed.
MPS3 UFSAR3.7B-21Rev. 30 3.7B.3.3 Procedures Used for Modeling 3.7B.3.3.1 Equipment and Components The procedures used for modeling of e quipment and components are contained in Section 3.7B.3.1.
3.7B.3.3.2 Piping Systems The basic method of analysis used is the finite element stiffness method. In accordance with this method, the continuous piping is ma thematically idealized as an a ssembly of elastic structural members connecting discrete nodal poi nts. Nodal points ar e placed in such a manner as to isolate particular types of piping elements , such as straight runs of pipe , elbows, valves, etc., for which force-deformation characteristics can be cate gorized. Nodal points are also placed at all discontinuities, such as piping s upports, concentrated weights, bran ch lines, and changes in cross section. Inertial characteristics of the piping syst em are simulated by discrete masses of pipe and pipe components (including all concentrated and eccentric masses such as valves and valve operators) lumped at selected node points. System loads other than weights, such as thermal forces and earthquake inertial forces, are also applied at the nodal points. The stiffness matrix of the piping system is calculated based upon the elastic properties of the pipe and pipe components, to include the effects of bending, torsional, axial, and shear deformations. The stiffness of piping elbows, and certain branch connect ions, is modified to account for local deformation effects by the flexibility factors suggested in the ASME Section III Code, Articles NB 3600 (Class 1 piping), NC 3600 (Class 2 piping), and ND 3600 (Class 3 piping).
3.7B.3.4 Basis for Selecti on of Frequencies 3.7B.3.4.1 Equipment and ComponentsAmplified response spectra (floor) developed for two orthogonal horizontal and vertical direction earthquakes are the basic sources of seismic design accelerations. As noted in Section 3.7B.3.1.1, seismic accelerations are selected from the amplified response spectra based on natural frequency calculations for the equipment or component.
3.7B.3.4.2 Piping SystemsIn the seismic design and multi-mass modal an alysis of Seismic Category I piping systems (Section 3.7B.3.8), the practice of selecting piping fundamental natural fre quencies to preclude resonance is not used.
3.7B.3.5 Use of Equivalent Static Load Method of Analysis 3.7B.3.5.1 Equipment and Components Those components which are considered relatively simple or rigid are designed, by virtue of natural frequency calculations, to withstand the ef fects of amplified seis mic acceleration values MPS3 UFSAR3.7B-22Rev. 30 dependent upon frequency and amplitude ranges as sociated with the installation, location, and corresponding relevant amplified response spectrum. Analysis of components to the peak value of resonant response is considered conservative , since fundamental natural frequencies do not generally coincide with the frequency at resona nce of the relevant response curve. Components having fundamental natural frequencies within th e broadened response peak are designed to peak acceleration values, increased by a factor of 1.3, or as justified, to account for the contribution of all significant dynamic modes under a resonant condition. Generally, the vibratory characteristics of the components, qualified by res onant static analysis, are such that no possibility exists for adjacent or multiple modes to exist within th e relatively narrow peak of a typical response spectrum.The discussion which follows justifies the use of a factor of 1.3 as a conservative multiple to be applied to single or multiple degree-of-freedom systems having fundamental frequencies within the broadened resonant response peak. Multiply supported, or continuous type, span components are not part of this proof. When such cases arise, a factor of 1
.5 times the peak resonant response is used.3.7B.3.5.1.1 Single Degree-of-Freedom SystemsPeak broadening is intended to reflect a range of uncertainty in the precise location of the resonant peak of the response curve, and not to indicate that the multiple peak resonant response is possible within this broadened range. What is concluded is that there is a fairly equal chance that the peak of the curve (singular) would fall in the specified range and, thus, wh at exists, in fact, is a family of resonant response curves, each having only one point of peak resonant response (Figure 3.7B-66). If more than one system or co mponent mode of vibration falls within the broadened peak, one and only one mo de (a presumed worst case) can be presumed at an actual response peak value (Figure 3.7B
-67). All other possible modes would realistically respond to lower values. Using the simple vibration theory and some simplifying assumptions, it is shown that a factor of 1.3 is conservative.
A simple damped oscillator re sponds with a transmissibility:
(3.7B.3-6) where:n = The undamped natural circular frequency w = The frequency of the exciting force The value of TR is dependent on the damping va lue, and the ratio of exciting frequency to oscillator natural frequency. When the exciting frequency equals the oscillator natural frequency, the steady state input is amplified by the value of TR and the response amplitude is maximum. In a seismic environment, maximum response is e qual to the peak of the amplified response spectrum curve.
TR 12n------2+1n------2 2 2n------2+----------------
------------------
------------------
-------------
=
MPS3 UFSAR3.7B-23Rev. 30 If additional modes are assumed around the peak of the response curve (Figure 3.7B-67), values of TR can be determined for each mode and the s quare root of the sum of the squares (SRSS) of these values computed. It is shown that TR increases as the numb er of modes increase, and that the most conservative placement of assumed modes is with one mode at the peak and other modes centered around this peak.
Data shown on Figure 3.7B-68 justify the use of a fa ctor of 1.3 as conserva tive for all potential equipment applications. The curves are develope d for two planes representing five modes and nine modes assumed acting within the broadened resonant peak. These num bers are intended to show an upper bound for general equipment appl ication. Equipment damping values of 2.0 and
 
===3.0 percent===
are used for static an alysis. Higher damping is shown to indicate the trend and the conservatism of this method.
As further conservatism, all modes are considered participating equally. This is never the case in dynamic analysis. The higher fr equencies of the component ar e given equal weight to the fundamental resonant frequency and the modes are centered on the nominal response curve. If the fundamental frequency were place d on the peak of the nominal cu rve, the results would show even lower transmissibilities.The factor 1.3 is applicable only for those components whose fundamental natural frequency falls within the broadened response peak.
 
It has been shown that, for the range of values associated with compone nt and system static analysis, use of the 1.3 factor is conservative. In fact, for a predominant number of likely cases, a value far less than this could be justified on the basis of the data.
For example, a value of 1.1 could be justifie d for most components which present only a few significant modes of vibration within the broade ned response peak. It is further emphasized that, in reaching these conclusions, the most conser vative assumptions regarding location of the nominal response curve and the placement of re sponse modes for the arbitrary component have been made.
3.7B.3.5.1.2 Multi-Degree-of-Freedom Systems As a conclusive supplement to the previous di scussion, a study was perfor med utilizing rigorous dynamic analysis of models closely representative of typical components.This investigation consists of computing the ratio of maximum dynamic stress to maximum static stress (i.e., the factor denoted by K) for several model beams subject to a flat response and typical amplified response spectra. Since bending stress is dominant for frame/
equipment construction, the actual ratio employed equals:
(3.7B.3-7)Both SRSS and absolute (ABS) moments are computed for comparison purposes, but conclusions are based solely on SRSS moments because they most closely represent actual dynamic stress.Maximum static moment corresponds, in the case of the 1 g flat response, to 1 g static load. In the case of a typical amplified response, the ma ximum static load is based upon the following frequency relationships (Figure 3.7B-69):
f o  f p , g = g max (peak acceleration)
K maximum dynamic moment maximum static moment
---------------
------------------
-----------------
------------------
-=
MPS3 UFSAR3.7B-24Rev. 30 f o > f p , g = acceleration at f o (3.7B.3-8) where: f o = The fundamental fre quency of the model beam f p = The frequency at which the peak acceleration occursThe effect of peak spreading is investigated by using a flat response, thus giving all modes the same acceleration. This is equivalent to infinite peak spreading. The importance of the uncertainty in the location of the peak acceleration with respect to the fundamental mode of the model beams is examined by adjusting the fundamental freque ncy from well below to well above the peak resonant frequency of a t ypical response spectrum.The model beams selected for this study are shown on Figure 3.7B-
: 70. These beams are typical of the frames and equipment combinations used in nuclear power plants. All dynamic analyses were conducted using the STRUDL-SW co mputer program described in Appendix 3A. Static analyses were carried out by hand, except for the simple/fixed beam with overhang. Consistent with design practice, all mountings in this study are assumed rigid.
3.7B.3.5.1.3 Results for Flat ResponseTable 3.7B-23 summarizes the results for a 1 g flat response applied to the model beams of Figure 3.7B-70. Three K factors were computed for comparison purposes:
(3.7B.3-9)
All conclusions in this study are based on Ks/u because it most closely represents the actual ratio of dynamic moment to static moment. K a/u was not chosen because, as Table 3.7B-24 illustrates, modes are so widely spaced that no more than one modal frequency lies within a
+/-10 percent frequency band. Ks/c is shown since this is the K factor which represents a typical simplification used in component analysis (concentrated st atic loads at component center of gravity).
K s/c Maximum SRSS dynamic momentMaximum static momentfrom concentrated load
---------------------
----------------------
----------------
-=Ks/uMaximum SRSS dynamic momentMaximum static momentfrom uniform load
---------------------
---------------------
-----------------
-=Ka/u Maximum ABS dynamic moment Maximum static moment from uniform load
--------------------
---------------------
------------------
-=
MPS3 UFSAR3.7B-25Rev. 30The 1 g flat response was selected to give infinite peak spreading. As can be seen, Ks/u was never greater than unity.
3.7B.3.5.1.4 Results for Amplified ResponseTable 3.7B-25 presents the results for the simp ly supported/fixed model beam with 33 percent overhang subjected to the response spectra of Fi gure 3.7B-69. The first mode column on this table gives the fundamental frequency, f o , and response acceleration, g, at f
: o. Note that f was adjusted (by density variation) from well below to well above the peak frequency, f p , of the response spectra to determine the effect on K of the uncerta inty in the location of the peak frequency with respect to the fundamental frequency of the model beam. Since all values of Ks/u were less than unity, it is concluded that this uncertain ty has no important effects on the K factor.
3.7B.3.5.1.5 Conclusions1.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 conserva tive for simply supported systems.2.No amount of peak spreading can itself result in a K s/u factor significantly greater than unity.3.Uncertainty in the frequency at which th e 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.
3.7B.3.5.2 Piping Systems Equivalent static load method is not generally us ed in the seismic analysis of Seismic Category I piping. However, seismic and certain nonseism ic small bore piping a nd instrument tubing (Section 3.7B.3.1.2) are seismically supported at standard interv als based on maximum spans established from modal respons e spectra analysis of representative small bore piping configurations.In some cases, when deemed appropriate, the analysis of small bore piping and instrument tubing is performed using methods based on simplified dynamic engineering formulations to ascertain their code adequacy.
The simplified dynamic formulation essentially involves the applicat ion of a factor of 1.3 to the peak value of ARS when the fundamental frequenc y of the piping or tubing configuration is less than 33 Hz while accounting for the seismic eff ects. The factor of 1.0 is applied when the fundamental frequency of the pipi ng or tubing configuration is greater than or equal to 33 Hz.
MPS3 UFSAR3.7B-26Rev. 30 3.7B.3.6 Three Components of Earthquake Motion 3.7B.3.6.1 Equipment and ComponentsIn the seismic analysis of equipment and components, each of the three defined directions of earthquake input (two horizontal and one vertical taken orthogonally) is evaluated separately. The stresses resulting from the orthogona l earthquake inputs are combined by the square root of the sum of squares (SRSS).
3.7B.3.6.2 Piping Systems In the seismic analysis of piping systems, the effects of simultaneous action of three spatial components of earthquake motion are considered. When the res ponse spectrum modal analysis method is used for seismic analysis, the ma ximum piping modal responses (e.g., moments and displacements) due to the thr ee spatial components of earthqua ke motion are obtained by the modified SRSS method. In this me thod, when considering a particul ar vibration mode, responses in a particular direction due to the two horizontal direction excitations are combined first by the SRSS method and then combined with response (i n this same direction) due to the vertical direction excitation by absolute sum method. It ha s been demonstrated that this modified SRSS method always results in cons ervative answers as compared wi th the SRSS method described in position C.2.1 of Regulator y Guide 1.92 (Chang 1973).
In mathematical terms, the modified SRSS met hod of response combinati on due to three spatial components of an earthquake at a partic ular node point can be expressed as:
(3.7B.3-10) where:Once this is accomplished, (R x)N , (R y)N , and (R z)N are then combined independently for significant modes N from 1 to K by methods described in Section 3.7B.3.7, where K is the number of significant modes considered in the modal response combination.
3.7B.3.7 Combination of Modal Responses The following methods are applicable to piping system analysis employing the modal analysis response spectra method.1.If the modes are not closely spaced, the maximum responses of piping are obtained by taking the SRSS of corresponding maxi mum modal responses of the piping (R xx) (R xy) (R xz)=Response in the x, y, or z direction, respectively for mode N (R x)N=Combined response in the x direction due to seismic excitation for mode N R x ()N R xx ()2 N R xz ()2 N+R xy N+=
MPS3 UFSAR3.7B-27Rev. 30attributed to individual significant modes. This is in agreement with Position C.1.1 of Regulatory Guide 1.92, dated February 1976.2.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.
For equipment and components, the following me thods of combination of modal responses are employed for each direct ion of earthquake motion:1.When performing response spectrum modal analysis, the representative maximum value of a particular response earthqua ke is obtained by taking the SRSS of corresponding maximum values of the re sponse of the element attributed to individual significant modes of the structure, system, or component. Mathematically, this can be expressed as:(3.7B.3-11) where: R = The representative maximum value of a pa rticular response of a given element to a given component of an earthquake, R k = The peak value of the response of the element due to the kth mode, N = The number of significant modes consid ered in the modal response combination.2.When performing response spectrum m odal 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:
(3.7B.3-12) where: R lq and R mq = Modal responses, R and R m 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; RR k 2 K1=N=RR k 2 k1=N      R l q R mq mi=jli=jq1=P+12=
MPS3 UFSAR3.7B-28Rev. 30 R, R k , and N = Definition in positi on 1.1 of Regulatory Guide 1.92; P = The number of groups of closely space d modes, excluding individual separated modes.3.The method used for combination of th ree components of earthquake motion is described in Section 3.7B.3.6
.3.7B.3.8 Analytical Procedures for Piping Systems The general analytical procedure of the moda l analysis response spectra method for piping systems is described in Section 3.7B.3.1.2. Basic steps and equations used in the analytical procedure are described below.
For the dynamic analysis, the piping is represen ted by a lumped mass, multi-degree-of-freedom mathematical model. The distri buted piping mass is lumped at the system nodal points. The equation of motion for the system is: (3.7B.3-13) where: [M] = Mass matrix for assembled system[C] = Damping matrix for assembled system[K] = Stiffness matrix for assembled system
{X} = Nodal displacement vector = [X(t)]
= Nodal velocity vector =
= Nodal acceleration vector =
{F} = Applied dynamic force vector = {F(t)}, or = [M]
for seismic analysis = Seismic acceleration vector for points of pipe support Equation 3.7B.3-13 is solved for the system dyna mic response as follows: First, the frequency equation, obtained by removing the forcing and damping terms from the above equation, is solved for the system natural frequencies and mode shapes. Next, the natural mode shapes are used to effect an orthogonal transformation of the equation yielding a series of independent equations of motion uncoupled in the system modes. Then, the uncoupled equations are solved by the response spectrum method to obtain system response in e ach mode, and the individual modal results are combined to determine the total system dynamic response. The mathematical formulation of these steps is described in the following subsections.
M[]X**{}C[]X*{}K[]X{}++F{}=X*{}X*t (){}X**{}X**t (){}U**g{}U**g{}
MPS3 UFSAR3.7B-29Rev. 30 3.7B.3.8.1 Natural Frequencies and Mode Shapes First, the eigenvalues (natural frequencies) an d the eigenvectors (mode shapes) for each of the natural modes are calculated by solving the frequency equation:
(3.7B.3-14) where: W n = Natural frequency of the nth mode,{}n = Mode shape vector of the nth mode,{0} = Null vector, N = Number of significant modes considered The eigenvalues and eigenvectors are obtained using the Householder-QR algorithm, Jacobi Reduction or Inverse iteration.
3.7B.3.8.2 Dynamic ResponseNext, let {n(t)] be the generalized coordinate vector, substitute {X} = [] {n} into the equation of motion and pre-multiply by []T; an orthogonal transformation results, from wh ich the uncoupled equations of motion shown below are obtained (3.7B.3-15) where:[] = The square matrix of mode shape vectors n = Generalized coordinate for the nth mode =
nth = Damping ratio of the nth mode expres sed as percent of critical damping P n = Generalized force of the nth mode P n = {}n T {F}/M n for applied dynamic force {F}
P n = {}n T [M] /M n for seismic analysis M n = Generalized mass of the nth mode = {}n T [M] {}nSolutions to these differential equations are obt ained by the method of amplified floor response spectrum (ARS) superposition, as described below.
K[]2 n M[]-()[]n 0{}=n123N,,,,=**2nn*2 nn++P n=n123N,,,,=U**g{}
MPS3 UFSAR3.7B-30Rev. 30 3.7B.3.8.3 Response Spectrum Modal Analysis The response of a piping system to seismic exci tations is obtained usi ng the method of response spectrum superposition. At any pipe support point , seismic input is produced by a set of three ARS, one in each global coordina te direction. These ARS are gene rated from the application of time-history acceleration responses obtained fr om the structure or e quipment time-history analysis. These ARS are peak broadened (Secti on 3.7B.2.9) to reflect vari ations in structure properties. Where a piping system is subjected to more than one set of (three) ARS, such as support points located in different structures or different parts of the same structure, the enveloping and peak broadeni ng are applied to all sets of ARS at support points (Section 3.7B.3.1.2). Thus, after the enveloping and peak broadening process, a set of three ARS, one in each global coordinate di rection, results. The maximum acceleration for the nth mode of the piping system in jth global coor dinate direction is then given by (3.7B.3-16) n = 1, 2, 3, ....Nj = 1,2,3 corresponds to response in X,Y, or Z global coordinate direction, respectively.
where: = Maximum acceleration vector of mode n nj max = Maximum generalized coordi nate acceleration of mode nnj = Modal participation factor for the nth mode in jth global coordinate directionnj = {}n T [M] {e}j{e}j = A vector with components of unity in all directions parallel to jth global coordinate direction and zero otherwise (S a)nj = Spectral acceleration for the nth mode, in jth global coordinate direct ion (from enveloped and peak broadened ARS)The maximum inertia force vector for the nth mode and in jth global coordinate direction is given by: (3.7B.3-17)These inertia forces are calculated for each of the system natural modes in all three global coordinate directions and applied as static forces in the same manner as the deadweight or equivalent thermal forces, to find internal moments and forces in each mode. The total maximum responses due to seismic excitation are then obtained by combining the modal responses described in Section 3.7B.3.7. The calculated prim ary stress range due to seismic inertial X**{}nj{}n**nj max=X**{}nj{}nnj/M n=X**{}nj F{}njmaxM N{}n**nj max=F{}nj max{}nnj S a ()nj=
MPS3 UFSAR3.7B-31Rev. 30 responses for ASME Code Clas s 1 piping components are added absolutely to the secondary stresses due to seismic anchor displacement calculated in the following manner.Maximum relative displacements in two horizontal and the vert ical direction between piping supports and anchor points (i.e., between floor penetrations and equipment supports at different elevations within a building, and also betwee n buildings) are used as equivalent static displacement boundary conditions in order to calculate the se condary stresses of the piping system. Relative seismic displacements used are obtained from a dynamic analysis of the structures, and are always considered to be out-of-phase between different buildings to obtain the most conservative piping responses.
These seismic member moments and forces are then combined with loads from deadweight, pressure, thermal, and other mechanical loads to complete the stress analysis of all Seismic Category I, and some Non- Seismic Category piping. For ASME Code Class 1 piping, stress intensities, and cumulative usage factors of the piping system are computed based on the formulation specified in Subarticle NB 3600, AS ME Section III for ASME Code Class 2 and 3 piping, the formulations in Subarticle NC 3600 and ND 3600, respectively, are used.
The design criteria, loading combination, and st ress limits for BOP Seismic Category I piping systems are described in Section 3.9B.3.
3.7B.3.9 Multiply Supported Equipm ent and Components with Distinct InputsTo calculate the maximum inertial response of multiply supported subsystems, an upper bound envelope of all the individual response spectra fo r the support locations is used. In addition, the relative displacements at the support points are considered. S upport displacements are imposed statically on the subsystem in the most conservative combinations. For support locations within a Category I structure, relative displacement s are determined algebraically and imposed.
Displacements between Category I structures are considered to be out-of-phase, and the maximum relative displacements between Categor y I structures are thus determined from absolute sums of the support displacements. Th e stresses due to seismic inertia and relative displacements are added to those due to other appr opriate loadings such as deadweight, pressure, etc., and the resulting stresses are limited by allowable stresses defined in applicable codes.
3.7B.3.10 Use of Constant Vertical Static Factors 3.7B.3.10.1 Equipment and Components Constant load factors are not used for vertical floor response in the seismic design of Seismic Category I equipment and components.
3.7B.3.10.2 Piping SystemsThe method of applying constant static factors as vertical response loads, based on the assumption of vertically rigid structures, for the seismic design of Seismic Category I piping is not used. However, a simplified analysis (e quivalent static load method) usi ng constant load factors for the vertical and horizontal di rections based on the peaks of applicable amplified response spectra is used for seismic analysis of certain nonseismic piping systems (Section 3.7B.3.5.2).
MPS3 UFSAR3.7B-32Rev. 303.7B.3.11 Torsional Effects of Eccentric MassesThe effect of eccentric masses such as valve operators is considered in the seismic piping analysis described in Section 3.7B.3.1.2. These eccentric mass es are included in th e mathematical model for the piping analysis and the torsional effects ca used by them are evaluated and included in the total piping response.
3.7B.3.12 Buried Seismic Category I Piping Systems In performing stress analysis of buried Seis mic Category I piping systems, the loadings considered are:1.Internal pressure2.Soil pressure (includes dead load, and live loads due to traffic when applicable)3.Thermal expansion4.Differential movements between structures and adjacent soil due to settlement and seismic motion5.Seismic wave effectsEffects of loadings (1) and (2) are assessed by well known methods (e.g., Terzaghi 1955; King, R.C. and Crocker, C. 1967, Section 21-29 to 21-33); effects of loadings (3) and (4) are accounted for by a static analysis considering piping modeled together with soil springs. This is basically a beam on elastic foundation approach. Loadings (4) a nd (5) are discussed in detail in the sections below.The basic assumptions concerning buried piping stress analysis are:1.piping satisfies elemen tary theory of beams;2.soil is linear elastic, homogeneous and isotropic; and3.soil strain is fully transferred to the pipe; i.e., there is no slippage between the pipe and the soil.
3.7B.3.12.1 Seismic Wave Effect in the Free FieldVarious seismic waves develop during an eart hquake. There are compression waves (P-waves), shear waves (S-waves), and different kinds of su rface waves such as Ra yleigh waves (R-waves).
When seismic waves propagate through the soil , responses of buried Seismic Category I piping are calculated by making use of the analyt ical approach proposed by Goodling (1978, 1979, 1980).
Straight portions of buried piping far from the effect of extern al supports, bends and tees are assumed to move with the soil when seismic waves propagate through it. Since Rayleigh waves MPS3 UFSAR3.7B-33Rev. 30 induce the highest axial strains in buried piping, onl y these waves are considered in the analysis.
The strains of the soil have been c onservatively estab lished as follows:
(3.7B.3-18)
Bending curvature  = a m/C 2 R where: V m = peak ground velocity, in/sec a m = peak ground acceleration, in/sec 2 C R = Rayleigh wave velocity, in/sec Therefore, the stresses on the straight portions of buried piping mentioned above are given by:
(3.7B.3-20)
(3.7B.3-21) where:E = Young's modulus of pipe, psi D o = outside diameter of pipe, in F max = maximum axial force, lb A m = cross-sectional area of pipe, in 2 Seismic wave effect on bends and tees in the free field are considered separately. For a bend, the maximum stresses are determined by assuming that its longitudinal leg is in the direction of maximum soil strain and its transverse leg is in the perpendicular direction.
The longitudinal leg may terminate into another bend, an anchor, or a free end. The bend is classified accordingly and the act ual slippage length (L') along wh ich slippage between pipe and soil occurs is determined as outlined by Goodling (1978).
The net relative displacement , between soil and pipe at the bend is given by:
(3.7B.3-22)Axial strain  m V m C R--------=Axial Stress  a EV m C R-----------
-Fmax A m-----------m E===Bending Stress  b ED o a m 2C 2 R-----------------
ED o X 2---------------
==1m L'fL 2 2A m E--------------
--S 1 L'A m E-----------
--=
MPS3 UFSAR3.7B-34Rev. 30 where:mL' = theoretical unrestrained relative movement at the elbow over length L'.
S 1 L'/A m E = the pipe elongation due to friction along the soil/pipe interface.fL'2/2A m E = the pipe elongation due to friction along the soil/pipe interface.
f = frictional force per unit length of pipe, lb/in The bend legs are considered as beams on elastic foundation for which its parameter  is given by: (3.7B.3-23) where: k = soil spring constant, per unit length, lb/in 2 I = moment of inertia of pipe cross section, in 4 In case of long-transverse leg (its length is greater than 3
/4), the following equations are derived by incorporating the interdependence of forces, moments, soil deformation, and rotation of the pipe in the immediate vi cinity of the bend (Goodling 1980).
(3.7B.3-24)
(3.7B.3-25) where: M = bending moment, in-lb/in f = elbow angle, radiansR = radius of elbow, in K = t = actual pipe wall thickness, in a = outside radius of pipe, in In case of short transverse leg, the following conservative equations derived by Goodling (1978) are used:4 k 4EI---------=M1 R/K'EI---------------------
=S 1 k1 2---------M+=1 91012 tR a 2------2+---------------
-----------------
-
MPS3 UFSAR3.7B-35Rev. 30 (3.7B.3-26)
(3.7B.3-27)
(3.7B.3-28) where: C 1 , C 2 , and C 3 are coefficients given by Goodling (1978)
The maximum axial force in the longitudinal leg, F max , induced by the seismic motion is given by: F max = S 1 + fL' (3.7B.3-29)Having determined the values of S, M and F, the stresses due to local deformation at the bend can be evaluated. These stresses are superimposed on stresses caused by the curvature of the pipe during seismic wave propagation. The combined stresses at the bend are multiplied by an intensification factor (0.75 i) to account for the higher intensity of stresses at the elbow.
The following expressions for stress result:
where: Z = section modulus of pipe, in 3 Occasionally, when the conservative equations used in case of s hort transverse leg result in unacceptable stresses, the technique incorporating the passive resistance of soil, as presented by Goodling (1978), may be used.
 
A similar approach is used fo r analyzing tees (Goodling 1978).
(1)Stress at an Elbow S o1 (elbow)= 0.75i [ED o /2+M/Z] + S 1/A m (3.7B.3-30)
(2)Stress in the Longitudinal Run S o1 (long)=F max/A m + ED o /2 (3.7B.3-31)1m L'fL'2 2A m E--------------
--1KL'A m E---------------
C 3 2C 1 C 3 C 2 2-  ---------------
------------------
+----------------
-----------------
------------------
--------------
-=S 1 k1---------    C 3 2C 1 C 3 C 2 2-  ------------------
---------------
-=M k1 22---------    C 2 2C 1 C 3 C 2 2-  -----------------
----------------
-=
MPS3 UFSAR3.7B-36Rev. 30 3.7B.3.12.2 Effects of Differential Movements Between Structure and Adjacent Soil Due to Seismic MotionDuring an earthquake, differential seismic motions occur between structures and adjacent soil. These differential motions are obt ained by the seismic analysis of structures with soil-structure interaction taken into account. The effect on th e buried piping systems due to these differential motions can be evaluated by considering separately the effects of different components, namely, differential motion components transverse to the direction of the piping axis, and differential motion components parallel to the direction of the piping axis.1.Differential motion components transver se 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 differenti al motion components transverse to the direction of piping axis. The approach is based on Terzaghi's 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 pair s 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 dir ection are calculated in accordance with Terzaghi (1955). The elastic springs in the horizontal direct ion are calculated based on the method described in Audi bert and Nyman (1975). The maximum expected transverse seismic displacements at the structural penetration are used as input in the calculation. The computat ion 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 parall el to the direction of piping axis.The effect on buried long straight piping due to differ ential seismic motions along the direction of piping axis at penetrati on is assessed by considering the frictional force between pipe and soil, and the maximum axial stress due to this effect is (Yeh 1974): where:a = Stress in piping at penetration point due to axial displacement, a , of the structure at the penetration point (psi)E = Young's modulus of pipe (psi)F = Frictional force between pipe and soil (psi)a 2EFa t---------------
-=
MPS3 UFSAR3.7B-37Rev. 30 a = Axial displacement of pipe at penetration point through structure (inch) t = Pipe thickness (inch)
When bends or tees exist close to the penetration, the effect of differential motion parallel to direction of piping axis is analyzed again by the method described in Shah and Chu (1974).
3.7B.3.12.3 Effects of Differential Movements Due to Structural Settlement Settlement of a structure can happen either due to its own weight over a pe riod of time or due to an earthquake. The settleme nt of the structure where the piping is connected, as well as the soil adjacent to the structure where the piping is buried, are imposed on the buried piping, and the approach outlined in "differentia l motion components transverse to the direction of piping axis" given above is used to evalua te stresses in the buried piping.
3.7B.3.12.4 Accommodations for Buried Piping Structural PenetrationsThe resultant loadings imposed by thermal, struct ural, and seismic distortions may cause severe local stresses in buried piping at structure pene tration points. The piping, if anchored at the structural wall, may be too stiff to accommodate these distortions for such locations. In such cases, the buried piping design includes a structural penetration, consisting of a concrete box or a conduit which is not attached to the structure, and is free to move with the soil rather than with the structure. Within the box or conduit, the piping may (if necessary) be provided with expansion joints or piping loops to accommodate relative displacements in both axial and transverse directions.
3.7B.3.13 Interaction of Other Systems (Piping a nd Equipment) with Seismic Category I Systems (Piping and Equipment)
Nonseismic category systems (piping and equipmen t) are designed to be isolated from Seismic Category I systems (piping and equipment). Isolation may be accomplished by physical restraints, barriers, or separation. If it is not practical to isolate the Seis mic Category I system from the nonseismic, then the poten tial for damage due to seismic in teractions is evaluated through the following program.
The Seismic Interaction Program c onsists of three distinct tasks:*demonstrating the adequacy of equipment anchorages for nonseismic equipment in Seismic Category I buildings,*demonstrating the structural integrity of pi ping and supports for selected subsystems, and*performing walkdowns to identify swing/
sway interactions between non-Seismic Category I piping and equipment and Se ismic Category I piping and equipment.
Non-Seismic Category I equipment in Seismic Category I buildings is reviewed to ensure the seismic adequacy of its anchorag e by one of the following methods:
MPS3 UFSAR3.7B-38Rev. 301.verify that the equipment anchor age has been explicitly qualified;2.compare the anchorage detail to explicitly qualified anchorages; and3.for anchorages which are not seismically designed and are not similar to seismic anchorages, calculations are performed to demonstrate adequacy.
For equipment anchorages evaluated by Method 3, an effort is ma de to group typical anchorage details and perform bounding calculations. The ma nner in which structural integrity is demonstrated for each piece of interacting equipment is docu mented by the calculations.
Acceptance criteria for equipment anchorage evaluations is given in J.F. Opeka to B.J. Youngblood Letter B11844, dated Oct ober 31, 1985, Docket No. 50-423.
The structural integrity of pi ping and supports is addressed through a program developed and implemented by Sargent and Lundy. A set of piping subsystems was selected to be representative of pipe sizes, hanger configurat ions, and operating conditi ons. As discussed in more detail in the above-referenced letter, these selected subsystems are bounding.
Restraint loads and pipe stresses were calculated using dynamic analysis re sults. They have been compared to the failure capacities associated with each subsystem to assess the inherent margin of safety in the design. Maximum dynamic lateral di splacements are used to confirm interaction criteria utilized duri ng plant walkdowns.
Seismic interaction walkdowns are conducted to identify swing/sway interactions between non-Seismic Category I piping and equipment and Seismic Category I pipi ng and equipment. All interactions are evaluated consid ering the local flexibility of the interacting equipment/piping. These reviews address restrictions such as penetrations and interferences with structures which would limit displacements. In all cases, interactions with active Seismic Categor y I components is prevented.
Information gained from the expe rience database is used in conjunction with the above efforts.
The database shows that properly supported equi pment and piping maintains its structural integrity during strong motion earthquakes. Further, this program benefits from the database information regarding the severity of seismic in teractions and the knowle dge of configurations which have not performed well in past earthquakes.
For nonseismic category piping sy stems attached to Seismic Category I piping systems, the dynamic effects of the nonseismic category piping ar e simulated in the analysis modeling of the Seismic Category I piping. The nonseismic categor y piping is modeled in a manner consistent with the accuracy of the Category I piping analysis. The attached nonseismic category piping is also designed to ensure that, during an earthquake of SSE intensity, it does not cause a failure of the Seismic Category I piping system.
3.7B.3.14 Seismic Analysis for Reactor Internals See Section 3.7N.3.
3.7B.3.15 Analysis Procedure for Damping Damping values of equipment, components a nd piping systems are given in Section 3.7B.1.3.
MPS3 UFSAR3.7N-110Rev. 30 3.7N SEISMIC DESIGNIn addition to the steady state loads imposed on the system unde r normal operating conditions, the design of the equipment requires that consideration also be given to a bnormal loading conditions, such as earthquakes. Seismic loadings are consid ered for earthquakes of two magnitudes: safe shutdown earthquake (SSE) and operating basis ear thquake (OBE). The SSE is defined as the maximum vibratory ground motion at the plant si te that can reasonably be predicted from geologic and seismic evidence. The OBE is that earthquake which, considering the local geology and seismology, can be reasonably expe cted to occur during the plant life.
For the OBE loading condition, the reactor coolant system is designe d to be capable of continued safe operation. The design for the SSE is intended to assure:1.That the integrity of th e reactor coolant pressure boundary is not compromised2.That the capability to shut down the react or and maintain it in a safe condition is not compromised3.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 The seismic requirements for safety related instrumentation and el ectrical equipm ent are covered in Section 3.10. The safety class definitions, clas sification lists, operating conditi on categories, and the methods used for seismi c qualification of mechanical e quipment are given in Section 3.2.
3.7N.1 SEISMIC INPUT 3.7N.1.1 Design Response Spectra Refer to Section 3.7B.1.1.
3.7N.1.2 Design Time History Refer to Section 3.7B.1.2.
3.7N.1.3 Critical Damping ValuesThe damping values given in Ta ble 3.7N-1 are used in the analysis of Westin ghouse equipment.
The values for RCS components are based on te sting programs, as reported in WCAP-7921-AR, which have been accepted by the staff. This WCAP defines the We stinghouse NSSS position on Regulatory Guide 1.61.Tests on fuel assembly bundles justified conservative component damping, values of 7 percent for OBE and 10 percent for SSE to be used in the fuel assembly component qualification.
Documentation of the fuel assembly test s is found in WCAP-8236 (1973) and WCAP-8288 (1974).
MPS3 UFSAR3.7N-111Rev. 30 The damping values used in component analysis of CRDM's and their seismic supports were developed through a testing program performed by Westinghouse. The test conducted was on a full size CRDM complete with rod position indicator coils, attachment to a simulated vessel head, and variable gap between the top of the pr essure housing support plate and a rigid bumper representing the support.
The program consisted of transient vibration tests in which the CRDM was deflected a specified initial amount and suddenly releas ed. A logarithmic decrement anal ysis of the decaying transient provides the effective damping of the assembly. The effect on damping of va riations in the drive shaft axial position, upper seismic support cleara nce, and initial defl ection amplitude was investigated.The upper support clearance had the largest af fect on the CRDM damping, with the damping increasing with increasing clearance. With an upper clearance of 0.06 inches, the measured damping was approximately 8 percent. The clear ances in a typical uppe r seismic CRDM support are a minimum of 0.10 inch es. The increasing damping with incr easing clearances trend from the test results, indicated th at the damping would be greater than 8 percent for both the OBE and the SSE, based on a comparison between typical defl ections during these seismic events and the initial deflections of mechanisms in the test. Component damping values of 5 percent are, therefore, conservative for both the OBE and SSE.
3.7N.1.4 Supporting Media for Seismic Category I Structures Refer to Section 3.7B.1.4.
3.7N.2 SEISMIC SYSTEM ANALYSIS Refer to Section 3.7B.2 3.7N.3 SEISMIC SUBSYSTEM ANALYSIS This section describes the seis mic analysis performed on subsystems within Westinghouse's scope of responsibility.
3.7N.3.1 Seismic Analysis Methods Those components that must remain functional in the event of the SSE (Seismic Category I) are identified by applying the criteria of Section 3.2.1.
In general, the dynamic analyses are performed using a modal an alysis in combination with a response spectrum analysis.
3.7N.3.1.1 Dynamic Analysis - Mathematical ModelThe first step in any dynamic analysis is to model the structure or component, i.e., convert the real structure or component into a system of masses , springs, and dashpots suitable for mathematical analysis. The essence of this step is to select a model so that the displacements obtained are a good representation of the motion of the structure or component. Stated differently, the true inertia MPS3 UFSAR3.7N-112Rev. 30 forces should not be altered so as to appreciably affect the internal stresses in the structure or component. Some typical modeling tec hniques are presented in Lin (1974).
Equations of MotionConsider the multi-degree of freedom system shown in Figure 3.
7N-1. Making a force balance on each mass point r, the equations of motion can be written in the form:
(3.7N-1)where: m r = The value of the mass or mass moment of rotational inertia at mass point r
= Absolute translational or angul ar acceleration of mass point r c ri = 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. Fo rce or moment is positive in the direction of positive translational or angular velocity. = Translational or angular velocity of mass point i relative to the base K ri = Stiffness coefficient - the external fo rce (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 displacem ent (rotation).
u i = Displacement (rotation) of ma ss point i relative to the base since: (3.7N-2)where: = Absolute translational (angular) acceleration of the base = Translational (angular) acceleration of mass point r relative to the base Equation 3.7N-1 can be written as:
(3.7N-3)m r y**r c ri u*i ik ri u i i++0=y**r u*i y**u**r y**s+=y**s u**r m r u**r c ri u*i ik ri u i i++-m r y**s=
MPS3 UFSAR3.7N-113Rev. 30For a single degree of freedom system with displacement u, mass m, damping c, and stiffness k, the corresponding equation of motion is:
(3.7N-4)3.7N.3.1.2 Modal Analysis Natural Frequencies and Mode Shapes The first step in the modal analysis method is to establish the norma l modes, which were determined by eigen solution of Equation 3.7N-3.
The right hand side and the damping term are set equal to zero for this purpose, as illustrated in Biggs (1964) (Pages 83 thru 111). Thus, Equation 3.7N-3 becomes:
(3.7N-5)The equation given for each mass point r in E quation 3.7N-5 can be written as a system of equations in matrix form as M{} + K{}(3.7N-6)where:M = Mass and rotational inertia matrix
{} = Column matrix of the general displacm ent and rotation at ea ch mass point relative to the baseK = Square stiffness matrix
{}= Column matrix of general translational and angular accelerations at each mass point relative to the base, d 2 {}/dt 2 Harmonic motion is assumed and the {} is expressed as:
{}={} sin wt(3.7N-7) where:{} = Column matrix of the spatial displacement and rotation at each mass point relative to the base = Natural frequency of harmon ic motion in radians per second The displacement function and its second derivati ve are substituted into Equation 3.7N-6 and yield: mu**cu*ku++m-y**s=m r u**r k ri u i i+0=
MPS3 UFSAR3.7N-114Rev. 30[K] {} = 2 [M]{}(3.7N-8)The determinant [K] - 2 [M] is set equal to zero and is then solved for the natural frequencies.
The associated mode shapes are then obtaine d from Equation 3.7N-8. Th is yields n natural frequencies and mode shapes where n equals th e number of dynamic degrees of freedom of the system. The mode shapes are all orthogonal to each other and are sometimes referred to as normal mode vibrations. For a single degree of freedom system, the stiffness matrix and mass matrix are single terms and the determinant [K] - 2 [M] when set equal to zero yields simply:
k - 2 m = 0 (3.7N-9)where  is the natural angular fre quency in radians per second.
The natural frequency in cycles per second is therefore:
(3.7N-10)To find the mode shapes, the natural fre quency corresponding to a particular mode, n , can be substituted in Equation 3.7N-8.
1.Modal Equations The response of a structure or component is always some combination of its normal modes. Good accuracy can usually be obtained by using only th e first few modes of vibration. In the normal mode method, the mode shapes are used as prin cipal coordinates to reduce the equations of motion to a set of uncoupled differential equations that describe the motion of each mode n. These equations may be written as (Biggs 1964, pages 116 thru 125):(3.7N-11)where the modal displacement or rotation, A n , is related to the displacement or rotation of mass point r in mode n, u rn , by the equation:
u rn = A nrn (3.7N-12) where:n = Natural frequency of mode n in radians per seconds p n = Critical damping ratio of mode nn = Modal participation factor of mode n given by:k m----=f 1 2------k m----=A**2n p n A*n2 n A n++-n y**s=
MPS3 UFSAR3.7N-115Rev. 30 (3.7N-13)where:'rn = Value of rn in the direction of the earthquake The essence of the modal analysis lies in th e fact that Equation 3.7N-11 is analogous to the equation of motion for a single degree of freedom system that is developed from Equation 3.7N-4.
Dividing Equation 3.7N-4 by m gives:
(3.7N-14)The critical damping ratio of the single degree of freedom syst em, p, is defined by the equation:
(3.7N-15)where the critical damping coefficient is given by the expression:
c c = 2m(3.7N-16)Substituting Equation 3.7N-16 into Equati on 3.7N-15 and solving for c/m gives:
(3.7N-17)Substituting this expression and the expression for k/m given by Equati on 3.7N-9 into Equation 3.7N-14 gives:
(3.7N-18)Note the similarity of Equations 3.7N-11 and 3.7N-18. Thus, each mode may be analyzed as though it were a single degree of freedom system and all modes are independent of each other. By this method, a fraction of critical damping, i.e., c/c c , may be assigned to each mode and it is not necessary to identify or evaluate individual damping coefficients, i.e., c. However, assigning only a single damping ratio to each mode is not appropr iate for a slightly damp ed structure supported by a massive moderately damped structure. Ther e are several methods which can be used to incorporate damping in the model.
 
One method is to develop and analyze separate mathematical models fo r both structures, using their respective damping values. The massive mode rately damped support structure is analyzed first. The calculated response at the support points for the slightly damped structures is used as a forcing function for the subsequent detailed an alysis. Another method is to inspect the mode shapes to determine which modes correspond to th e slightly damped structure and then use the n m rrn 1 rm rrn 2 r-----------------------
=u**c m----u*k m----u++y**-s=p c c c----=c m----2p=u**2pu*2 u++y**s-=
MPS3 UFSAR3.7N-116Rev. 30 damping associated with the structure having pr edominant motion. A third method is to use the Rayleigh damping method based on computed modal energy distribution.
3.7N.3.1.3 Response Spectrum Analysis The response spectrum is a plot showing the va riation in the maximum response (Thomas et al., 1963, pages 24 thru 51) (displacement, velocity, a nd acceleration) of a si ngle degree of freedom system versus its natural frequency of vibrati on when subjected to a time history motion at its base.
The response spectrum concept can be best explained by out lining the steps involved in developing a spectrum curve. Determination of a single point on the curve requires that the response (displacement, velocity, and acceleration) of a single degree of freedom system with a given damping and natural frequency is calculated for a given base motion.The variations in response are es tablished and the maximum absolute value of each is plotted as an ordinate with the natural frequency used as the abscissa. The process is repeated for other assumed values of frequency in sufficient detail to establish the complete curve. Other curves corresponding to different fractions of critical damping are obtained in a similar fashion. Thus, the determination of each point of the curve re quires a complete dynamic response analysis; the determination of a complete spectrum may involve hundreds of such analyses. However, once a response spectrum plot is generated for the particular base motion, it may be used to analyze each structure and component with the base moti on. The spectral acceleration, velocity, and displacement are rela ted by the equation:
Response spectra developed for Millst one 3 are discussed in Section 3.7B.3.
3.7N.3.2 Determination of Number of Earthquake Cycles Where fatigue analyses of mechanical systems and components are required, Westinghouse specifies in the equipment specification the num ber of cycles of the operating basis earthquake (OBE) to be considered. The number of cycles for NSSS components is given in Table 3.9-1. The fatigue analyses are performed and presente d as part of the components stress report.
3.7N.3.3 Procedure Used for Modeling Refer to Section 3.7N.3.1 for modeling procedures for subsystems in Westinghouse's scope of responsibility.
3.7N.3.4 Basis for Selecti on of Frequencies The analysis of equipment subject ed to seismic loading involves se veral basic steps, the first of which is the establishment of the intensity of the seismic loading. Considering that the seismic input originates at the point of support, the response of the equi pment and its associated supports S a nn S v n2 n s d n==
MPS3 UFSAR3.7N-117Rev. 30based upon the mass and stiffness characteristi cs of the system determine the seismic accelerations which the equipment must withstand.
Three ranges of equipment/support behavior which affect the magnit ude of the seismic acceleration 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.
3.7N.3.5 Use of Equivalent Static Load Method of AnalysisThe equivalent static load or static analysis method involves the multiplication of the total weight of the equipment or component member by the specified seismic acceleration coefficient. The magnitude of the seismic acceleration coefficient is establishe d on the basis of the expected dynamic response characteristic s of the component. Component s which can be adequately characterized as single degree of freedom systems are considered to have a modal participation factor of one. Seismic acceleration coefficients for multi-degree of freedom systems, which may be in the resonance region of the amplified response spectra curves, are increased by 50 percent to account conservatively for the increased modal pa rticipation where the equivalent static load method is used.
3.7N.3.6 Three Components of Earthquake Motion The seismic design of the RCS equipment includes the effect of the seismic response of the supports, equipment, structures , and components. Floor response spectra are generated for two perpendicular horizontal di rections (i.e., N-S, E-W) and the vertical direction. The equipment response is determined using horizontal and vertical spectra which envelope the appropriate floor response spectra. The total se ismic response is obtained by combining the unidirectional responses in the two horizontal di rections and the vertical directio n using the square root of the sum of the squares method.Time history analysis was not used on Millstone 3.
3.7N.3.7 Combination of Modal Responses The total unidirectional seis mic response is obtained by co mbining the individual modal responses utilizing the square root of the su m of the squares method. For systems having modes with closely spaced frequencies, th is method is modified to include the possible effect of these modes. The groups of closely sp aced modes are chosen such that the difference between the MPS3 UFSAR3.7N-118Rev. 30 frequencies of the first mode and the last mode in the group does not exceed 10 percent of the lower frequency. Groups are formed starting fr om the lowest frequency and working towards successively higher frequencies. No one freque ncy is in more than one group. Combined total response for systems which have such closely spa ced modal frequencies is obtained by adding to the square root of the sum of the squares of all modes the product of the responses of the modes in each group of closely spaced modes and a coupling factor. This can be represented mathematically as:
(3.7N-20)where: R T = Total unidirectional response R i = Absolute value of response of mode iN = Total number of modes considered S = Number of groups of closely spaced modes M j = Lowest modal number associated w ith group j of closely spaced modes N j = Highest modal number associated w ith group j of closely spaced modesKl = Coupling factor with (3.7N-21)and (3.7N-22 (3.7N-23) where:K = Frequency of closely spaced mode K K = Fraction of critical dampi ng in closely spaced mode K t d = Duration of the earthquake R 2 T R i 2 i1=N2      R K R l E K llK1+=N jKM j=N j 1-j1=S+=K l 1'K'l-'KKll+()-------------------------------------
2+
1-='KK 1'K ()2-[]12='KK 2K t d-----------
-+=
MPS3 UFSAR3.7N-119Rev. 30 3.7N.3.8 Analytical Procedures for Piping Refer to Section 3.7B.3.9 3.7N.3.9 Multiply Supported Equi pment Components with Distinct InputsWhen response spectrum methods are used to evaluate reactor coolant system primary components interconnected between floors, the proc edures of the following paragraphs are used.
The primary components of the reactor coolant sy stem are supported at no more than two floor elevations.A dynamic response spectrum analysis is first made assuming no relative displacement between support points. The response spectra used in this analysis are the most severe floor response spectra.Secondly, the effect of differ ential seismic movement of co mponents interconnected between floors is considered statically in the detailed component analysis. The differential motion is evaluated as a free end di splacement in accordance with ASME III, NB-3213.19.The results of these two steps, the dynamic inertia analysis and the static differential motion analysis, are combined absolutely with due c onsideration for the ASME classification of the stresses.3.7N.3.10 Use of Constant Vertical Static Factors Constant vertical load factors are not used as the vertical fl oor response load for the seismic design of safety related components and equipment within Westinghouse's scope of responsibility.3.7N.3.11 Torsional Effects of Eccentric MassesRefer to Section 3.7B.3.11.
3.7N.3.12 Buried Seismic Category I Piping Systems and Tunnels Refer to Section 3.7B.3.12.
3.7N.3.13 Interaction of Other Piping with Seismic Category I Piping Refer to Section 3.7B.3.13.
3.7N.3.14 Seismic Analyses for Reactor Internals Fuel assembly component stresses induced by hor izontal seismic distur bances are analyzed through the use of finite element computer modeling.
The time history floor response, based on a sta ndard seismic time hist ory normalized to SSE levels, is used as the seismic input. The reactor internals and th e fuel assemblies are modeled as spring and lumped mass systems or beam elements. The component seismic response of the fuel assemblies is analyzed to determine design adequacy. A detailed discussion of the analyses MPS3 UFSAR3.7N-120Rev. 30 performed for typical fuel assemblies is contained in WCAP-8236 (1973), WCAP-8288 (1974);
and Lin (1974), ASME Paper 74-NE-7.
Fuel assembly lateral structural damping obtai ned experimentally is presented in WCAP-8236 Addendum 1 (1974) and WCAP-8288 Addendum 1 (1974) (Figure 3-4). The data indicate that no damping values less than 10 percen t were obtained for fuel assemb ly displacements greater than 0.11 inches.
The distribution of fuel assembly amplitudes decreases as one approaches the center of the core. The average amplitude for the minimum displacement fuel assembly is well above 0.11 inches for the SSE.
Fuel assembly displacement time history for th e SSE seismic input is illustrated in WCAP-8236 Addendum 1 (1974) and WCAP-8288 Addendum 1 (1974) (Figure 2-3).The CRDM's are seismically analyzed to confirm that system stresses under the combined loading conditions, as described in Section 3.9N.1, do not exceed allowable levels as defined by the ASME Code, Section III. The CRDM is mathemat ically modeled as a system of lumped and distributed masses. The model is analyzed under appropriate seis mic excitation and the resultant seismic bending moments along the length of th e CRDM are calculated. The corresponding stresses are then combined with the stresses from the other loadings required and the combination is shown to meet ASME Code, Section III requirements.
3.7N.3.15 Analysis Procedure for Damping The damping values and procedures used for Westinghouse scope of supply and analysis are discussed in Secti ons 3.7N.1.3 and 3.7N.3.1.
3.7N.4 SEISMIC INSTRUMENTATION Refer to Section 3.7.4.
MPS3 UFSAR3.7-122Rev. 303.7.4 SEISMIC INSTRUMENTATION 3.7.4.1 Comparison with Regulatory Guide 1.12 Millstone 3 complies with Regulatory Guide 1.12 (Section 1.8) with the following exceptions.
ANSI/ANS 2.2 is used instead of ANSI N18.5 referenced by the Regulatory Guide, since ANSI/
ANS 2.2 is the final issued versi on of proposed standard ANSI N18.5.
Only solid state digital instrumentation that will enable the processing of data at the plant site within 4 hours of the seismic event is used.
No peak recording instrumentation is used.
The sensor required by section C.1.c.(3) is mounted on the wall of the emergency generator enclosure approximately 3 feet a bove the mat near a corner to reduce the potential for equipment damage due to flooding. This location will provide representative indication based on the stiffness of the supporting structure.
Regulatory Position C.3 of the guide is not appl icable because the safe shutdown earthquake for Millstone 3 is less than 0.3g (Section 2.5.2.6).
3.7.4.2 Location and Description of InstrumentationTable 3.7-1 lists the location of the seismic instrumentation. Seis mic instrumentation is located in areas where it can be serviced during periods of unit shutdown. All instruments are oriented to the same azimuths used in the mathematical model to permit direct use of the data within the model. All accelerometers, recorders, and central controllers have instrument characteristics that meet or exceed the requirements in Section 5 of ANSI/ANS 2.2-1978.Triaxial Accelerogr aph (five provided)
Five triaxial time-history ac celerographs capable of measur ing and permanently recording absolute acceleration versus time are provided. The instrumentation consists of local triaxial force balance accelerometers, solid state recorders, and a central controller located in the control room.
Four accelerometers input to the two centrally located solid state dual pane l recorders. The fifth accelerometer inputs to a local solid state recorder locate d in the Auxiliary Building. Accelerometers are installed at two locations on the outside of the containment structure in the engineering safety features bu ilding. One accelerometer is m ounted on the base mat of the containment at elevation (-) 24 feet - 3 inches and the other directly above the first on the containment shell at elevation 40 feet - 6 inches.The third accelerometer is installed on the wall of the emergency generator enclosure at elevation 4 feet - 6 inches. The vault is part of the emergency generator enclosure building. This location is chosen because the emergency ge nerator enclosure building has dynamic response characteristics different from that of the containment structure. Table 2.5.4-14 describes the differing founding conditions.The fourth accelerometer is loca ted at elevation 46 feet - 6 inches near the charging pump surge tank located in the Auxiliary Building. This accelerometer replaces the response spectrum recorder required at this location by Regulatory Guide 1.12 revision 1 since alternate triaxial instrumentation may be provided at this location in accordance with Part 1b of Section 4.1.6 of ANSI/ANS 2.2-1978.
MPS3 UFSAR3.7-123Rev. 30 The fifth accelerometer is located on the elevat ion 51 foot - 4 inch slab of the containment structure internal adjacent to a steam generator support. This accelerometer inputs to the local solid state recorder located in the Auxiliary Building.Solid State Recorders Two solid state dual panel recorders capable of recording motion in all three axes are provided.
These recorders are centrally located in the control room and continuously monitor the input from the first four accelerometers described above. Each axis has an independently adjustable trigger.
If any one axis is triggered all three axes of all four sensors are recorded. Event data is stored on flash storage cards for later analysis. Each solid state recorder has its own battery to ensure the recorder can record events after a loss of AC power. The internal battery provides approximately 30 hours of power autonomy. A main control board alarm annunciates via the alarm relay panel if the containment mat accelerometer is triggered.
The alarm relay panel is not powered from the UPS such that the main control board alarm wi ll not annunciate follow ing a loss of external power. A main control board alarm annunciates on a loss of external power.The local solid state recorder located in the Auxiliary Building record s the data from the accelerometer located on the elevation 51 foot - 4 in ch slab of the containment structure internal adjacent to a steam generator support. The local so lid state recorder is similar to the solid state dual panel recorders described above except it is stand alone so it will only record data when one of its monitored axes exceeds the trigger and it will not trigger the other recorders. Operations will verify power available on a daily basis and the internal battery will ensure the recorder can record events after a loss of AC power. The internal battery provides approxi mately 30 hours of power autonomy. Data collected by the local solid stat e recorder may be evaluated using the central controller to determine the response spectrum.
Central Controller The central controller automatically retrieves da ta from the two dual panel solid state recorders shortly after the recorders are tri ggered and determines the response spectrum of the event. If the response spectrum recorded at th e containment mat exceeds the OBE criteria an LED is lit on the alarm relay panel and the central c ontroller display will indicate "OBE". The central controller is powered from a UPS such that data can be analyzed after a loss of power. The UPS is sized to power the central controller for more than 25 minutes. The alarm relay panel is not powered from the UPS and will not indicate OBE exceedance following a loss of power. A main control board alarm annunciates on a loss of external power.
The central controller may be used to determine the response spectrum of data recorded off the local solid state recorder.
3.7.4.3 Control Room Operator Notification The recorders located in the control room and the local solid state recorder located in the Auxiliary Building record the signals generated by the accelerometers on flash memory cards. If the signals from any axis of any of the first four sensors exceed the trigge r value all axes of all four sensor record. If the signa ls from any axis of the sensor mounted on the containment steam MPS3 UFSAR3.7-124Rev. 30 generator support exceed the trigger value all axes of this sensor record on the local solid state recorder. Trigger values will be set as required by ANSI/ANS 2.2-1978.
If any of the first four sensors tr igger or if power is lost to the seismic monitoring system (except the local solid state recorder) a main control board "seismic monitoring warning/trouble" alarm annunciates.If the containment mat accelerometer triggers, a main control board alarm will annunciate via the alarm relay panel. The alarm relay panel is not pow ered from the UPS such that the main control board will not annunciate following a loss of external power.The central controller will automatically retrie ve the data from the tw o solid state dual panel recorders and determine the response spectrum. If the data from the containment mat exceeds OBE criteria a LED is lit on the alarm relay panel and the central controller display will indicate "OBE". Data from the local solid state recorder must be collected and brought to the central controller or another personal computer with corr ect software installed to determine the response spectrum.The central controller will continue to ope rate on UPS supply for approximately 25 minutes following a loss of external power. The operator may determine OBE exceedance from the central controller during this time. The alarm relay pa nel is not powered from the UPS and will not indicate OBE exceedance following a loss of power.
3.7.4.4 Comparison of Measured and Predicted ResponsesThe criteria and procedures used to compare recorded data obtained from seismic instrumentation to plant design parameters are based on AN SI/ANS 2.10-1979 (Section 3.7.5). Instrumentation requirements are in accordan ce with ANSI/ANS 2.2-1978 and the supplemental provisions of Regulatory Guide 1.12, Re vision 1 (Section 3.7.4.1).
3.
 
==7.5 REFERENCES==
FOR SECTION 3.73.7-1American 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.3.7-2American Nuclear Society 1979. American National Standard for Guidelines for Retrieval, Review, Processing, and Eval uation of Records Obtained from Seismic Instrumentation. ANSI/ANS 2.10-1979, La Grange Park, Illinois.3.7-3Audibert, 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, 1975.3.7-4Biggs, J.M. 1964. Introduction to Structural Dynamics. McGraw-Hill, New York.3.7-5Chang, T.Y. 1973. Comparison of Seismic Re sponse Combination Procedures for Piping Systems. SWEC, October 1973.3.7-6Goodling, E.C., Jr. Flexibility Analysis of Buried Pipe. ASME Publication 78-PVP-82.
MPS3 UFSAR3.7-125Rev. 303.7-7Goodling, E.C., Jr. Seismic Stresses in Bu ried Elbows. Preprint 3595. ASCE National Convention, Boston, 1978.3.7-8Goodling, E.C., Jr. More on Flexibility An alysis of Buried Pi pe. ASME Publication 80-C2/PVP-67.3.7-9Hetenyi, M. 1946. Beams on Elastic Foundati on. The University of Michigan Press.3.7-10Johnson, J.J. and Kennedy, R.P. 1977. Earthqua ke Response of Nuclear Power Facilities. Presented at the ASCE Fall Convention and Exhibit, San Fr ancisco, Calif., October 17-21, 1977.3.7-11Kausel, E. and Roesset, J. 1975. Dynamic Stiffness of Circular Foundations. ASCE, Eng. Mechanics Division.3.7-12King, R.C. and Crocker, C. 1967.
Piping Handbook. McGraw-Hill Book Co.3.7-13Lin, 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 Pi ping Conference, Miami, Fla.3.7-14Lin, Y.Y. and Hanjian, A.H. 1976. Discrete Modeling of Containment Structures.
Presented at the International Symposium on Earthquake Structural Engineering, St.
Louis, Mo.3.7-15Newmark, N.M. 1972. Earthquake Response Analysis of Reactor Structures. Nuclear Engineering and Design, Vol 20, p 303-332.3.7-16Newmark, N.M. and Rosenblueth, E. 1971.
Fundamentals of Earthquake Engineering.
Prentice-Hall, Inc.3.7-17Richart, F.E., Jr.; Hall, J.R., Jr.; and Woods, R.D. 1970. Vibrations of Soils and Foundations. Prentice-Hall, Inc., Englewood Cliffs, N.J.3.7-18Roesset, J.M.; Whitman, R.V.; and Doby, R. 1973. Modal Analysis for Structures with Foundation Interaction. Journal of the Structural Divisi on, Proceedings ASCE, p 399-416.3.7-19Shah, H.H. and Chu, S.L. 1974. Seismic Analysis of Underground St ructural Elements.
Journal of the Power Division, Proceedings of ASCE, Vol 100, No. P01, p 53-62, July 1974.3.7-20Singh, A.K.; Chu, S.L.; and Singh, S.L. 1973.
Influence of Closely Spaced Modes in Response Spectrum Method of Analysis. Pres ented at the Specialty Conference on Structural Design of Nuclear Plant Facilities, Chicago, Ill.3.7-21Terzaghi, K. 1955. Evaluation of Coefficients of Subgr ade Reaction. Geotechnique, p 297-325.
MPS3 UFSAR3.7-126Rev. 303.7-22Thomas, T.H. et al., 1963. Nuclear React ors and Earthquakes.
TID-7024, U.S. Atomic Energy Commission, Washington, D.C.3.7-23U.S. Nuclear Regulatory Commission (NRC) 1975. Seismic System Analysis. Standard Review Plan, Sect ion 3.7.2, NUREG-75/087.3.7-24Wass, G. 1972. Linear Two-Dimensional Anal ysis of Soil Dynamic s in Semi-Infinite Layered Media. PhD Thesis, Univ. of Calif., Berkeley, Presented to J. Lysmer.3.7-25WCAP-7921-AR, 1974. Damping Values of Nuclear Plant Components.3.7-26WCAP-7950, 1972. Gesinski, T.L. Fuel Assembly Safety Analysis for Combined Seismic and Loss-of-Coolant Accident.3.7-27WCAP-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.3.7-28WCAP-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.3.7-29Whitman, R.V. and Richart, F.E., Jr. 1967.
Design Procedures for Dynamically Loaded Foundations. Journal of the So il Mechanics and Foundation Di vision, Proceedings of the ASCE.3.7-30Whitman, R.V. 1970. Soil Structure Inter action. In: Seismic Design for Nuclear Power Plants, MIT Press, Cambridge, Mass., p 245-269.3.7-31Yeh, 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.
MPS3 UFSARMPS3 UFSAR3.7-127Rev. 30 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 ap proximately 3 feet above the mat near a corn er to reduce the potential for equipment d amage due to flooding. This location will provide representative indication based on the stiffness of the supporting structure.TABLE 3.7-1 SEISMIC INSTRUMENTATION Instrument LocationTriaxial Force Balance AccelerometerMain Control Board AnnunciationContainment Structure - OutsideMat, elevation (-) 24 feet 3 inches X (1)X (2)Containment shell, elevation 40 feet 6 inches X (2)Containment Structure - InsideSteam Generator Support, el evation 51 feet 4 inches X (3)Emergency Generator Enclosure Located on Mat in Diesel Fuel Oil Vault, elevation 4 feet 6 inches X (1) (4)Auxiliary Building Surge Tank S upport, elevation 46 feet 6 inches X (1)
MPS3 UFSAR3.7-128Rev. 30NOTE: 1. For final reconciliation of pipe stress analys is or piping system ba ckfits, damping values as defined in ASME Code Case N-411 (Figure 3.7B-71) may be utilized for both OBE and SSE.TABLE 3.7B-1 DAMPING FACTORSStress LevelType of Condition of Structure, System or ComponentPercent of Critical Damping1.Low stress, well below proportional limit. Stresses
 
below 0.25 yield point stressSteel, reinforced concrete; no cracking and no slipping at
 
joints, piping or components 0.5 to (1)2.Working stress limited to 0.5 yield point stressa.Welded steel, well reinforced concrete (with only slight
 
cracking)2b.Bolted steel53.At or just below yield pointa.Welded steel5b.Reinforced concrete5c.Bolted steel74.At all stress levelsa.Rock (translation)10b.Rock (rotation)5 MPS3 UFSAR3.7-129Rev. 30TABLE 3.7B-2 METHODS OF SEISMI C ANALYSIS USED FOR SEISMIC CATEGORY I STRUCTURESStructure Response Spectrum AnalysisModal Time History AnalysisFrequency Domain Time History AnalysisReactor containment and internalsXXMain steam valve buildingXX Hydrogen recombiner buildingXXSafeguards areaXXAuxiliary buildingXX Fuel buildingXXControl buildingXService buildingXX Emergency generator enclosureXService water pumphouseXX Refueling water storage and chemical addition tanks XX Demineralized water storage tank and enclosure XXFuel building canopyXXTurbine buildingXXWaste disposal buildingX MPS3 UFSAR3.7-130Rev. 30TABLE 3.7B-3 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODAL FREQUENCIES AND PARTICIPATION FACTORS UNCRACKED MODEL Participation FactorsModeFrequency (Hz)HorizontalVertical N-S E-W 14.66433.88424.204.6724.751353.06-30.44-3.28 34.8543.9837.96-4.0945.565252.24-42.17-32.0056.13553.4086.74-1.22 68.2440.110.890.10710.9173.42-177.613.11810.921170.863.108.60 911.80913.83-99.2317.971012.63078.31-58.48399.461112.88110.41-117.44-228.28 1213.295157.3661.05-62.871313.6078.368.530.171415.59242.134.40-241.13 1518.75144.54-110.37-0.041618.755108.2643.751.451722.1522.60-0.14-1.10 1822.1570.480.95-0.06 MPS3 UFSAR3.7-131Rev. 30TABLE 3.7B-4 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODAL FREQUENCIES AND PARTICIP ATION FACTORS CRACKED MODEL Participation FactorsModeFrequency (Hz)Horizontal EarthquakeVertical EarthquakeN-SE-W 13.423614.97317.580.22 23.4239315.07-14.92-0.2534.75022.93277.826.2445.544295.89-48.38-32.19 55.8639.3310.31-0.5166.13457.4495.54-1.0677.76115.69155.83-0.35 87.762153.44-15.402.0299.3714.21-1.29376.49109.6950.252.02-0.03 1111.80115.39-118.1310.091212.75887.27-136.8479.361313.135190.0298.5994.36 1413.37537.78-72.22-17.731513.4284.913.73-63.011615.38844.634.92-324.04 1715.7630.881.23-0.201815.7662.090.49-29.84 MPS3 UFSAR3.7-132Rev. 30TABLE 3.7B-5 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) UNCRACKED MODELModeJointDisplacementXYZ110.0010.00.01620.0120.0060.15330.0230.0090.336 40.0160.0050.33850.0210.0030.41860.0430.0180.442 70.0540.0130.64680.0750.0110.91690.0780.0071.0100.0080.00.098110.0160.00.20120.0250.00.310 130.0350.00.425140.0440.00.536150.0510.00.621 160.0630.00.776170.0790.00.972210.0090.00.020.0310.005-0.00530.051-0.006-0.00540.0600.0-0.007 50.067-0.005-0.00660.074-0.003-0.00770.102-0.007-0.00980.143-0.013-0.01190.166-0.006-0.014100.0930.0-0.008 MPS3 UFSAR3.7-133Rev. 30110.1980.0-0.017120.3120.0-0.027130.4300.0-0.037 140.5450.0-0.04715 0.6330.0-0.05416 0.7950.0-0.068 171.00.0-0.086310.00.00.0012-0.010-0.006-0.104 3-0.021-0.007-0.2644-0.014-0.004-0.2575-0.017-0.002-0.326 6-0.038-0.014-0.3427-0.048-0.010-0.5088-0.066-0.009-0.722 9-0.066-0.006-0.776100.0080.00.083110.0170.00.188120.0280.00.303130.0380.00.421140.0490.00.538 150.0570.00.626160.0720.00.790170.0910.01.0410.0090.0-0.00220.1320.007-0.040 30.270-0.053-0.03940.327-0.016-0.063TABLE 3.7B-5 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) UNCRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-134Rev. 3050.370-0.048-0.05760.417-0.038-0.07070.590-0.062-0.093 80.858-0.094-0.14291.0-0.053-0.19110-0.002-0.0010.011-0.020-0.0010.00312-0.040-0.0010.00713-0.062-0.0010.010 14-0.082-0.0010.01415-0.098-0.0010.01616-0.129-0.0010.022 17-0.171-0.0010.028510.0040.00.00620.0100.0120.107 30.052-0.006-0.01040.1040.0020.14750.121-0.0120.10 60.0940.0100.17370.182-0.0040.29080.326-0.0180.640 90.437-0.0081.0100.0010.00.00211-0.0030.0-0.00512-0.0070.0-0.01313-0.0130.0-0.022 14-0.0180.0-0.03115-0.0220.0-0.038TABLE 3.7B-5 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) UNCRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-135Rev. 3016-0.0300.0-0.05217-0.0410.0-0.071610.00.0-0.0042-0.0810.007-0.2853-0.0360.006-0.3004-0.0240.005-0.240 5-0.0200.001-0.2356-0.0300.017-0.1937-0.0050.011-0.061 80.0890.0060.58290.1660.0051.0100.00.0-0.00411-0.0010.0-0.00512-0.0010.001-0.00513-0.0010.001-0.005 14-0.0010.001-0.00515-0.0010.001-0.00516-0.0010.001-0.004 17-0.0010.001-0.004710.0010.0-0.02820.0050.001-0.058 30.0050.004-0.05540.0030.002-0.05950.0020.003-0.05160.0030.005-0.0497-0.0010.006-0.019 8-0.0070.0070.0769-0.0110.0040.116TABLE 3.7B-5 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) UNCRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-136Rev. 30100.0040.001-0.235110.0080.001-0.429120.0100.001-0.552 130.0110.001-0.583140.0090.001-0.516150.0070.001-0.391 16-0.0010.0010.08417-0.0180.0011.081-0.027-0.0010.02-0.036-0.0020.03-0.038-0.0120.04-0.034-0.0070.0 5-0.030-0.0120.06-0.025-0.0100.07-0.002-0.0150.0 80.049-0.021-0.00190.076-0.014-0.00110-0.234-0.001-0.00411-0.429-0.002-0.00812-0.552-0.002-0.01013-0.584-0.003-0.011 14-0.517-0.003-0.01015-0.393-0.003-0.007160.083-0.0030.002171.0-0.0040.019910.0030.003-0.02420.090-0.002-0.38930.1100.025-0.337TABLE 3.7B-5 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) UNCRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-137Rev. 3040.0650.010-0.49650.0510.027-0.42360.0830.022-0.481 7-0.0120.038-0.4128-0.1070.0570.6589-0.1650.0291.0100.00.0060.00211-0.0030.0080.03312-0.0050.0110.058 13-0.0070.0130.07314-0.0070.0140.07515-0.0050.0150.065 160.0010.0160.011170.0160.017-0.1041010.0260.102-0.02020.2530.198-0.27330.4080.350-0.411 40.4180.314-0.33150.3960.365-0.32760.3550.388-0.267 70.1930.459-0.0018-0.3150.5510.2279-0.6030.4820.441100.0130.252-0.00911-0.0060.4120.00312-0.0230.5560.01513-0.0360.6810.023TABLE 3.7B-5 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) UNCRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-138Rev. 3014-0.0410.7800.02715-0.0390.8370.02516-0.0120.9130.005 170.0541.0-0.0451110.004-0.060-0.04120.058-0.077-0.542 30.077-0.070-0.78440.091-0.10-0.67050.086-0.095-0.653 60.061-0.054-0.55770.035-0.065-0.0568-0.157-0.0690.555 9-0.259-0.1161.0100.002-0.183-0.023110.0-0.3140.00112-0.002-0.4330.02413-0.003-0.5360.04214-0.004-0.6180.051 15-0.004-0.6660.049160.0-0.7290.013170.008-0.801-0.080121-0.0430.013-0.0162-0.368-0.010-0.1913-0.615-0.136-0.2504-0.609-0.087-0.2385-0.575-0.147-0.229 6-0.536-0.118-0.2107-0.255-0.189-0.065TABLE 3.7B-5 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) UNCRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-139Rev. 3080.553-0.2720.21191.0-0.1810.36110-0.0280.079-0.01111-0.0060.150-0.002120.0180.2140.007130.0370.2700.015 140.0490.3160.019150.0490.3420.019160.0190.3770.007 17-0.0640.417-0.025131-0.0410.0-0.0422-0.255-0.047-0.364 3-0.529-0.168-0.4894-0.549-0.133-0.5665-0.525-0.184-0.531 6-0.472-0.159-0.5457-0.257-0.225-0.32480.448-0.3050.636 90.831-0.2401.010-0.0300.046-0.03111-0.0100.096-0.010120.0120.1410.014130.0320.1810.034140.0440.2130.047150.0460.2320.049160.0210.2570.023 17-0.0520.286-0.051141-0.0170.073-0.002TABLE 3.7B-5 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) UNCRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-140Rev. 302-0.1420.3480.0053-0.2460.5430.0354-0.2510.5990.022 5-0.2430.6270.0226-0.2280.6590.0137-0.100.767-0.032 80.5430.909-0.09190.8541.0-0.12810-0.0170.010-0.00211-0.015-0.061-0.00112-0.008-0.1290.0130.001-0.1900.001 140.010-0.2400.002150.015-0.2690.003160.007-0.3090.003 17-0.027-0.3570.002151-0.0340.00.0842-0.0120.0020.019 30.009-0.001-0.02240.0140.0-0.02950.0170.0-0.036 60.0180.0-0.03970.0190.001-0.0368-0.0260.00.0459-0.048-0.0050.08610-0.2630.00.65011-0.4040.01.012-0.3670.00.909TABLE 3.7B-5 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) UNCRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-141Rev. 3013-0.1690.00.419140.1050.0-0.259150.2970.0-0.736 160.3290.0-0.81517-0.1940.00.4811610.0820.0010.03320.029-0.0020.0053-0.0300.003-0.0114-0.0430.003-0.014 5-0.0490.004-0.0186-0.0530.004-0.0197-0.0510.004-0.017 80.0700.0070.02090.1290.0160.037100.6490.0010.262111.0-0.00.404120.910-0.00.368130.420-0.0010.170 14-0.259-0.001-0.10415-0.736-0.001-0.29716-0.815-0.001-0.329 170.481-0.002-0.1951710.002-0.0010.02-0.002-0.017-0.00130.004-0.001-0.00140.002-0.0090.0 5-0.002-0.0020.0016-0.005-0.0040.002TABLE 3.7B-5 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) UNCRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-142Rev. 307-0.0120.0010.00280.0600.010-0.00990.0750.007-0.01110-0.0590.00.01011-0.1550.00.02712-0.2300.00.039 13-0.2300.00.03914-0.1410.00.02415-0.0200.00.004 160.2340.001-0.040171.00.001-0.1721810.00.00.0012-0.0020.001-0.00430.00.0-0.00840.0010.0-0.006 50.0010.001-0.00660.0-0.001-0.00470.0020.00.014 80.0050.00.02490.0040.00.02010-0.0100.0-0.05911-0.0270.0-0.15412-0.0390.0-0.22813-0.0400.0-0.22914-0.0240.0-0.14115-0.0040.0-0.021 160.0400.00.233170.1720.01.0TABLE 3.7B-5 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) UNCRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-143Rev. 30TABLE 3.7B-6 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) CRACKED MODELModeJointDisplacementXYZ110.00.00.00420.00.00.01030.0010.00.016 40.0010.00.01750.0010.00.01960.0010.0010.020 70.0010.00.02880.0020.00.03890.0020.00.043100.0040.00.086110.0090.00.191120.0140.00.306 130.0200.00.424140.0260.00.540150.0300.00.628 160.0370.00.791170.0470.01.0210.0040.00.020.0090.002-0.00130.013-0.001-0.001 40.0140.001-0.00150.0160.0-0.00160.0170.0-0.00170.023-0.001-0.00180.030-0.002-0.001 90.035-0.001-0.002 MPS3 UFSAR3.7-144Rev. 30100.0860.0-0.004110.1900.0-0.009120.3050.0-0.014 130.4240.0-0.020140.5400.0-0.026150.6280.0-0.030 160.7910.0-0.037171.00.0-0.047310.0010.00.00820.0120.0070.14530.0250.0100.338 40.0170.0050.33550.0210.0030.41960.0450.0180.441 70.0560.0130.65080.0790.0110.92290.0810.0081.0100.0010.00.006110.00.00.00112-0.0010.0-0.006 13-0.0010.0-0.01414-0.0020.0-0.02215-0.0020.0-0.02816-0.0040.0-0.04117-0.0050.0-0.060410.010-0.001-0.002TABLE 3.7B-6 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) CRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-145Rev. 3020.1340.008-0.04030.271-0.053-0.03840.328-0.015-0.062 50.371-0.047-0.05760.416-0.038-0.06970.590-0.061-0.091 80.858-0.094-0.13891.0-0.052-0.186100.011-0.001-0.002110.007-0.001-0.001120.002-0.0010.013-0.005-0.0010.001 14-0.013-0.0010.00215-0.019-0.0010.00316-0.033-0.0010.005 17-0.054-0.0010.009510.0060.00.00720.0520.0140.15130.119-0.0170.04340.184-0.0010.195 50.211-0.0220.14860.1950.0020.22470.322-0.0180.34580.521-0.0390.66690.659-0.0191.0100.0070.00.008110.0060.00.006TABLE 3.7B-6 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) CRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-146Rev. 30120.0030.00.00313-0.0010.0-0.00214-0.0050.0-0.007 15-0.0090.0-0.01116-0.0180.0-0.02117-0.0320.0-0.037610.0040.00.00620.0100.0120.107 30.050-0.006-0.01140.1020.0020.14650.119-0.0110.094 60.0920.0100.17270.178-0.0040.28880.320-0.0180.638 90.430-0.0071.0100.0050.00.008110.0040.00.007120.0030.00.004130.00.00.014-0.0020.0-0.004 15-0.0040.0-0.00816-0.0100.0-0.01817-0.0200.0-0.03471-0.0010.0-0.0132-0.0010.0-0.0113-0.0010.0-0.007TABLE 3.7B-6 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) CRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-147Rev. 304-0.0010.0-0.00550.00.001-0.00360.00.001 70.0010.0010.00880.0030.00.02690.0040.00.03610-0.0230.0-0.22211-0.0420.0-0.42212-0.0550.0-0.552 13-0.0590.0-0.58914-0.0530.0-0.52615-0.0410.0-0.403 160.0070.00.074170.1010.01.081-0.0120.00.0012-0.0090.00.03-0.004-0.0030.0 4-0.001-0.0020.050.001-0.003-0.00160.004-0.003-0.001 70.013-0.004-0.00180.030-0.006-0.00290.039-0.004-0.00210-0.2220.00.02211-0.4220.00.04212-0.5520.00.05613-0.5890.00.059TABLE 3.7B-6 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) CRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-148Rev. 3014-0.5260.00.05315-0.4030.00.041160.0740.0-0.007 171.00.0-0.101910.0010.0350.020.0050.043-0.00230.0090.050-0.00340.0090.050-0.003 50.0090.052-0.00360.0090.052-0.00370.0080.056-0.002 80.0070.060-0.00290.0060.060-0.002100.00.1930.0110.00.3630.0120.00.5170.013-0.0010.6510.0 14-0.0010.7590.015-0.0010.8210.0160.00.9040.0 170.01.00.0101-0.0020.0-0.0142-0.0630.002-0.4263-0.0270.006-0.378 4-0.0400.002-0.4105-0.0410.002-0.361TABLE 3.7B-6 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) CRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-149Rev. 306-0.0280.014-0.3577-0.0330.012-0.22080.0350.0120.605 90.0850.0051.010-0.001-0.001-0.010110.0-0.003-0.004120.001-0.0050.003130.001-0.0060.010140.002-0.0070.014 150.002-0.0080.015160.001-0.0090.00617-0.002-0.010-0.0181110.0040.002-0.02920.090-0.004-0.392 30.1080.023-0.34140.0640.008-0.49750.0500.024-0.424 60.0820.019-0.4807-0.0120.035-0.4068-0.1060.0540.657 9-0.1640.0251.0100.0050.001-0.039110.005-0.001-0.036120.004-0.002-0.021130.001-0.0030.001 14-0.002-0.0040.02515-0.004-0.0040.037TABLE 3.7B-6 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) CRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-150Rev. 3016-0.003-0.0050.032170.007-0.006-0.0181210.0310.022-0.05020.2700.081-0.57130.4240.211-0.840 40.4400.157-0.69650.4170.204-0.68360.3630.247-0.569 70.1930.299-0.0278-0.4100.3720.5419-0.7460.2771.0100.0770.012-0.123110.0970.001-0.158120.078-0.010-0.128 130.027-0.021-0.04614-0.036-0.0300.05615-0.077-0.0350.123 16-0.076-0.0430.121170.051-0.052-0.095131-0.061-0.023-0.0322-0.395-0.069-0.2813-0.648-0.219-0.3764-0.638-0.174-0.3465-0.601-0.241-0.334 6-0.562-0.206-0.2997-0.266-0.291-0.071TABLE 3.7B-6 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) CRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-151Rev. 3080.547-0.3920.30891.0-0.3040.53010-0.287-0.014-0.15111-0.418-0.003-0.22012-0.3680.008-0.19413-0.1590.019-0.084 140.1200.0280.063150.3120.0330.164160.3370.0410.177 17-0.1990.051-0.107141-0.0150.0050.03020.0400.022-0.10430.0770.047-0.16240.0800.039-0.148 50.0760.048-0.14560.0700.055-0.13070.0400.067-0.036 8-0.0700.0830.1269-0.1320.0650.22810-0.3280.0040.61911-0.5300.0011.012-0.495-0.0020.93313-0.239-0.0040.450140.126-0.006-0.238150.386-0.007-0.728 160.440-0.010-0.83017-0.240-0.0120.451TABLE 3.7B-6 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) CRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-152Rev. 30 1510.002-0.0220.0022-0.283-0.075-0.214 3-0.506-0.199-0.2974-0.505-0.167-0.2795-0.480-0.220-0.270 6-0.451-0.195-0.2477-0.228-0.267-0.07180.418-0.3510.257 90.777-0.2840.441100.603-0.0150.323111.0-0.0040.534120.9450.0060.505130.4660.0160.24814-0.2270.024-0.121 15-0.7240.030-0.38716-0.8350.037-0.446170.4490.0470.237161-0.0170.094-0.0022-0.1590.3670.005 3-0.2750.5530.0354-0.2790.6110.0215-0.2700.6350.0216-0.2520.6680.0127-0.1090.770-0.034 80.5660.905-0.09090.8971.0-0.127TABLE 3.7B-6 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) CRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-153Rev. 30100.0070.074-0.004110.0340.043-0.006120.0480.008-0.007 130.041-0.028-0.006140.015-0.061-0.00315-0.012-0.0830.0 16-0.048-0.1140.00817-0.105-0.1540.0351710.00.00.02-0.0010.0-0.00130.0010.0-0.002 40.0020.00.050.0020.00.060.0020.00.001 70.0020.00.00880.0020.00.00890.0020.00.01010-0.0130.0-0.04411-0.0400.0-0.13412-0.0640.0-0.215 13-0.0690.0-0.22914-0.0460.0-0.15415-0.0120.0-0.039160.0640.00.214170.2980.01.0181-0.0010.0070.0TABLE 3.7B-6 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) CRACKED MODELModeJointDisplacementXYZ MPS3 UFSAR3.7-154Rev. 302-0.0140.0220.0013-0.0190.0410.0044-0.0180.0440.003 5-0.0180.0470.0036-0.0160.0490.0027-0.0040.058-0.003 80.0580.070-0.01190.0850.077-0.01610-0.0440.0060.01311-0.1330.0040.04012-0.2140.0010.06413-0.228-0.0020.068 14-0.154-0.0040.04615-0.040-0.0060.012160.212-0.009-0.064 171.0-0.012-0.298TABLE 3.7B-6 CONTAINMENT AND INTERNAL STRUCTURES SIGNIFICANT MODE SHAPES (NORMALIZED EIGENVECTORS) CRACKED MODELModeJointDisplacementXYZ MPS3 UFSARMPS3 UFSAR3.7-155Rev. 30TABLE 3.7B-7 CONTAINMENT AND INTERNAL STRUCTURES SRSS ACCELERATIONS AND DISPLACEMENTS FOR SAFE SHUTDOWN EARTHQUAK E UNCRACKED MODEL, RESPONSE SPECTRUM ANALYSISDegree of
* FreedomAcceleration **Displacement ***HorizontalVerticalHorizontalVerticalN-SE-WN-SE-W10.0580.0260.0120.031xE-30.05xE-30.031xE-3 20.0130.0070.0480.035xE-30.034xE-30.112xE-330.0260.0630.0060.050xE-30.396xE-30.030xE-340.627xE-40.254xE-30.646xE-40.010xE-40.084xE-40.001xE-4 50.874xE-40.559xE-40.335xE-40.003xE-40.007xE-40.060.232xE-30.531xE-40.973xE-40.073xE-40.010xE-40.002xE-470.1630.0460.0630.224xE-20.047xE-20.032xE-2 80.0280.0130.0680.201xE-30.172xE-30.251xE-390.0540.1630.0800.740xE-30.382xE-20.403xE-3100.862xE-30.211xE-20.102xE-20.165xE-40.702xE-40.045xE-4110.242xE-20.421xE-20.111xE-20.280xE-40.911xE-40.047xE-4120.176xE-20.416xE-30.646xE-30.446xE-40.077xE-40.043xE-4130.1920.0530.0910.443xE-20.094xE-20.056xE-2 140.0520.0210.1010.850xE-30.287xE-30.425xE-3150.0590.2380.1180.929xE-30.834xE-20.594xE-3160.926xE-30.287xE-20.150xE-20.140xE-40.911xE-40.076xE-4 170.148xE-20.403xE-20.134xE-20.347xE-40.134xE-30.071xE-4 MPS3 UFSARMPS3 UFSAR3.7-156Rev. 30180.262xE-20.720xE-30.107xE-20.635xE-40.148xE-40.068xE-4190.2220.0560.0920.532xE-20.101xE-20.061xE-2 200.0410.0190.1010.277xE-30.150xE-30.408xE-3210.0610.2390.0980.123xE-20.839xE-20.50xE-3220.971xE-30.305xE-20.170xE-20.152xE-40.976xE-40.086xE-4 230.153xE-20.415xE-20.129xE-20.364xE-40.137xE-30.069xE-4240.285xE-20.789xE-30.115xE-20.686xE-40.160xE-40.075xE-4250.2440.0590.0890.601xE-20.116xE-20.064xE-2 260.0520.0210.1100.767xE-30.173xE-30.459xE-3270.0620.2890.0960.125xE-20.104xE-10.049xE-2280.101xE-20.321xE-20.189xE-20.162xE-40.103xE-30.095xE-4 290.159xE-20.427xE-20.124xE-20.380xE-40.141xE-30.068xE-4300.298xE-20.831xE-30.122xE-20.713xE-40.165xE-40.079xE-4310.2690.0640.0820.675xE-20.299xE-20.066xE-2 320.0480.0230.1150.620xE-30.467xE-30.481xE-3330.0710.3050.0810.144xE-20.110xE-10.042xE-2340.106xE-20.333xE-20.207xE-20.170xE-40.107xE-30.104xE-4TABLE 3.7B-7 CONTAINMENT AND INTERNAL STRUCTURES SRSS ACCELERATIONS AND DISPLACEMENTS FOR SAFE SHUTDOWN EARTHQUAK E UNCRACKED MODEL, RESPONSE SPECTRUM ANALYSISDegree of
* FreedomAcceleration **Displacement ***HorizontalVerticalHorizontalVerticalN-SE-WN-SE-W MPS3 UFSARMPS3 UFSAR3.7-157Rev. 30350.165xE-20.436xE-20.120xE-20.391xE-40.144xE-30.066xE-4360.309xE-20.870xE-30.129xE-20.735xE-40.170xE-40.083xE-4 370.3580.0540.0750.951xE-20.210xE-30.079xE-2380.0600.0260.1380.991xE-30.380xE-30.569xE-3390.0710.4330.0130.20xE-20.160xE-10.019xE-2 400.122xE-20.364xE-20.261xE-20.192xE-40.118xE-30.131xE-4410.193xE-20.481xE-20.658xE-30.456xE-40.161xE-30.046xE-4420.349xE-20.101xE-20.158xE-20.804xE-40.185xE-40.097xE-4 430.5230.1150.1110.138xE-10.305xE-20.119xE-2440.1010.0360.1660.152xE-20.402xE-30.685xE-3450.1200.6230.0780.307xE-20.228xE-10.047xE-2 460.154xE-20.416xE-20.311xE-20.210xE-40.125xE-30.156xE-4470.307xE-20.609xE-20.314xE-20.697xE-40.171xE-30.164xE-4480.414xE-20.127xE-20.212xE-20.886xE-40.272xE-40.119xE-4 490.6220.1440.1800.161xE-10.348xE-20.149xE-2500.1010.0350.1750.857xE-30.262xE-30.652xE-3510.1700.6980.1420.398xE-20.249xE-10.077xE-2TABLE 3.7B-7 CONTAINMENT AND INTERNAL STRUCTURES SRSS ACCELERATIONS AND DISPLACEMENTS FOR SAFE SHUTDOWN EARTHQUAK E UNCRACKED MODEL, RESPONSE SPECTRUM ANALYSISDegree of
* FreedomAcceleration **Displacement ***HorizontalVerticalHorizontalVerticalN-SE-WN-SE-W MPS3 UFSARMPS3 UFSAR3.7-158Rev. 30520.162xE-20.448xE-20.319xE-20.213xE-40.126xE-30.160xE-4530.312xE-20.619xE-20.328xE-20.703xE-40.172xE-30.171xE-4 540.425xE-20.133xE-20.224xE-20.895xE-40.204xE-40.122xE-4550.1290.0450.0150.268xE-20.031xE-20.027xE-3560.0240.0200.0780.093xE-30.096xE-30.270xE-3 570.0450.1210.0050.030xE-20.246xE-20.028xE-3580.233xE-30.776xE-30.829xE-40.032xE-40.258xE-40.003xE-4590.179xE-30.807xE-40.421xE-40.004xE-40.008xE-40.001xE-4 600.864xE-30.196xE-30.127xE-30.282xE-40.032xE-40.002xE-4610.2050.0440.0120.571xE-20.064xE-20.050xE-3620.0360.0330.1100.160xE-30.163xE-30.443xE-3 630.0440.1880.0030.063xE-20.501xE-20.048xE-3640.355xE-30.127xE-20.390xE-40.053xE-40.419xE-40.004xE-4650.108xE-30.938xE-40.286xE-40.006xE-40.010xE-40.001xE-4 660.147xE-20.314xE-30.789xE-40.480xE-40.054xE-40.004xE-4670.2910.0550.0130.90xE-20.100xE-20.085xE-3680.0470.0450.1330.222xE-30.224xE-30.60xE-3TABLE 3.7B-7 CONTAINMENT AND INTERNAL STRUCTURES SRSS ACCELERATIONS AND DISPLACEMENTS FOR SAFE SHUTDOWN EARTHQUAK E UNCRACKED MODEL, RESPONSE SPECTRUM ANALYSISDegree of
* FreedomAcceleration **Displacement ***HorizontalVerticalHorizontalVerticalN-SE-WN-SE-W MPS3 UFSARMPS3 UFSAR3.7-159Rev. 30690.0550.2550.0060.099xE-20.775xE-20.078xE-3700.391xE-30.166xE-20.798xE-40.070xE-40.545xE-40.005xE-4 710.179xE-30.107xE-30.537xE-40.007xE-40.011xE-40.001xE-4720.194xE-20.377xE-30.115xE-30.636xE-40.071xE-40.006xE-4730.3730.0490.0170.124xE-10.138xE-20.012xE-2 740.0590.0560.1540.276xE-30.276xE-30.736xE-3750.0490.3200.0080.136xE-20.109xE-30.106xE-1760.454xE-30.199xE-20.640xE-40.083xE-40.639xE-40.006xE-4 770.192xE-30.126xE-30.481xE-40.008xE-40.011xE-40.001xE-4780.233xE-20.455xE-30.102xE-30.751xE-40.081xE-40.007xE-4790.4580.0590.0130.157xE-10.174xE-20.015xE-2 800.0680.0650.1760.319xE-30.319xE-30.845xE-3810.0590.3810.0090.173xE-20.133xE-10.014xE-2820.508xE-30.225xE-20.425xE-40.092xE-40.704xE-40.007xE-4 830.142xE-30.113xE-30.354xE-40.008xE-40.012xE-40.002xE-4840.262xE-20.495xE-30.101xE-30.831xE-40.093xE-40.008xE-4850.5190.0650.0170.182xE-10.202xE-20.018xE-2TABLE 3.7B-7 CONTAINMENT AND INTERNAL STRUCTURES SRSS ACCELERATIONS AND DISPLACEMENTS FOR SAFE SHUTDOWN EARTHQUAK E UNCRACKED MODEL, RESPONSE SPECTRUM ANALYSISDegree of
* FreedomAcceleration **Displacement ***HorizontalVerticalHorizontalVerticalN-SE-WN-SE-W MPS3 UFSARMPS3 UFSAR3.7-160Rev. 30860.0720.0690.1880.344xE-30.343xE-30.909xE-3870.0640.4280.0090.20xE-20.154xE-10.016xE-2 880.513xE-30.238xE-20.797xE-40.096xE-40.738xE-40.007xE-4890.198xE-30.125xE-30.491xE-40.008xE-40.012xE-40.001xE-4900.277xE-20.489xE-30.140xE-30.872xE-40.097xE-40.009xE-4 910.6460.0890.0100.229xE-10.254xE-20.022xE-2920.0790.0760.2060.377xE-30.375xE-30.993xE-3930.0890.5260.0060.251xE-20.193xE-10.019xE-2 940.119xE-20.558xE-20.276xE-30.150xE-40.116xE-30.017xE-4950.168xE-30.155xE-30.462xE-40.011xE-40.017xE-40.002xE-4960.590xE-20.120xE-20.471xE-30.138xE-30.151xE-40.021xE-4 970.8320.0930.0170.288xE-10.319xE-20.029xE-2980.0930.0840.2550.415xE-30.413xE-30.109xE-2990.0920.6860.0140.316xE-20.242xE-10.025xE-21000.267xE-20.829xE-20.450xE-30.168xE-40.130xE-30.022xE-41010.290xE-30.253xE-30.780xE-40.016xE-40.020xE-40.004xE-41020.832xE-20.268xE-20.117xE-20.155xE-30.170xE-40.027xE-4TABLE 3.7B-7 CONTAINMENT AND INTERNAL STRUCTURES SRSS ACCELERATIONS AND DISPLACEMENTS FOR SAFE SHUTDOWN EARTHQUAK E UNCRACKED MODEL, RESPONSE SPECTRUM ANALYSISDegree of
* FreedomAcceleration **Displacement ***HorizontalVerticalHorizontalVerticalN-SE-WN-SE-W MPS3 UFSAR3.7-161Rev. 30NOTES:* Degrees of freedom correspondi ng to x-, y-, and z-direction translati on and rotation about the x, y, and z axis are identified in Table 3.7B-9
.** Tabulated accelerations are in units of g fo r 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.
MPS3 UFSARMPS3 UFSAR3.7-162Rev. 30TABLE 3.7B-8 CONTAINMENT AND INTERNAL STRUCTURES SRSS ACCELE RATIONS AND DISPLACEMENTS FOR SAFE SHUTDOWN EARTHQUAKE CRACKED MODEL, RESPONSE SPECTRUM ANALYSISDegree of *FreedomAcceleration **Displacement ***HorizontalVerticalHorizontalVerticalN-SE-WN-SE-W10.0680.0220.0150.296xE-30.049xE-30.024xE-320.0170.0040.0460.027xE-30.014xE-30.126xE-330.0220.0680.0040.049xE-30.284xE-30.012xE-34 ****0.418xE-40.136xE-30.284xE-40.007xE-40.065xE-40.001xE-450.882xE-40.579xE-40.869xE-40.004xE-40.007xE-40.060.168xE-30.358xE-40.761xE-40.066xE-40.007xE-40.003xE-470.1800.0540.0600.252xE-20.516xE-30.232xE-380.0230.0110.0750.20xE-30.157xE-30.263xE-390.0720.1600.0260.838xE-30.307xE-20.145xE-3100.904xE-30.206xE-20.745xE-30.184xE-40.568xE-40.026xE-4110.245xE-20.392xE-20.122xE-20.316xE-40.841xE-40.028xE-4120.201xE-20.543xE-20.502xE-30.506xE-40.089xE-40.038xE-4130.2170.0750.0650.505xE-20.102xE-20.438xE-3140.0570.0280.1030.986xE-30.284xE-30.383xE-3150.0850.2160.0390.983xE-30.699xE-20.216xE-3160.119xE-20.297xE-20.512xE-30.151xE-40.750xE-40.026xE-4170.171xE-20.405xE-20.417xE-30.386xE-40.124xE-30.034xE-4 MPS3 UFSARMPS3 UFSAR3.7-163Rev. 30180.300xE-20.948xE-30.686xE-30.726xE-40.156xE-40.056xE-4190.2520.0780.0650.608xE-20.114xE-20.502xE-3200.0390.0210.1120.325xE-30.151xE-30.408xE-3210.0830.2140.0320.134xE-20.695xE-20.204xE-3220.130xE-20.311xE-20.546xE-30.161xE-40.807xE-40.029xE-4230.176xE-20.418xE-20.405xE-30.405xE-40.127xE-30.034xE-4240.327xE-20.105xE-20.715xE-30.785xE-40.168xE-40.061xE-4250.2780.0780.0650.687xE-20.129xE-20.548xE-3260.0570.0280.1160.892xE-30.206xE-30.43xE-3270.0820.2570.0320.131xE-20.867xE-20.213xE-3280.139xE-20.321xE-20.587xE-30.178xE-40.854xE-20.032xE-4290.182xE-20.429xE-20.414xE-30.423xE-40.131xE-30.035xE-4300.341xE-20.111xE-20.750xE-30.817xE-40.175xE-40.064xE-4310.3070.0800.0650.773xE-20.161xE-20.598xE-3320.0510.0310.1220.721xE-30.415xE-30.447xE-3330.0840.2710.0300.153xE-20.914xE-20.209xE-3340.148xE-20.328xE-20.636xE-30.187xE-40.890xE-40.035xE-4350.186xE-20.438xE-20.425xE-30.435xE-40.134xE-30.036xE-4TABLE 3.7B-8 CONTAINMENT AND INTERNAL STRUCTURES SRSS ACCELE RATIONS AND DISPLACEMENTS FOR SAFE SHUTDOWN EARTHQUAKE CRACKED MODEL, RESPONSE SPECTRUM ANALYSISDegree of *FreedomAcceleration **Displacement ***HorizontalVerticalHorizontalVerticalN-SE-WN-SE-W MPS3 UFSARMPS3 UFSAR3.7-164Rev. 30360.353xE-20.116xE-20.790xE-30.842xE-40.181xE-40.066xE-4370.4130.0840.0430.109xE-10.220xE-20.794xE-3380.0700.0380.1420.115xE-20.374xE-30.515xE-3390.0770.3790.0110.208xE-20.134xE-10.236xE-3400.181xE-20.346xE-20.853xE-30.214xE-40.982xE-40.043xE-4410.212xE-20.481xE-20.302xE-30.507xE-40.150xE-30.037xE-4420.396xE-20.137xE-20.106xE-20.922xE-40.199xE-40.076xE-4430.6050.1320.1180.159xE-10.322xE-20.121xE-2440.1010.500.1730.175xE-20.435xE-30.608xE-3450.1370.5510.0370.323xE-20.191xE-10.366xE-3460.220xE-20.429xE-20.989xE-30.237xE-40.105xE-30.051xE-4470.366xE-20.656xE-20.944xE-30.777xE-40.167xE-30.065xE-4480.474xE-20.176xE-20.175xE-20.102xE-30.220xE-40.096xE-4490.7180.1750.1810.185xE-10.372xE-20.147xE-2500.0830.0380.1960.995xE-30.282xE-30.649xE-3510.1990.6200.0600.427xE-20.209xE-10.464xE-3520.228xE-20.481xE-20.105xE-20.240xE-40.105xE-30.052xE-4530.372xE-20.667xE-20.987xE-30.784xE-40.168xE-30.067xE-4TABLE 3.7B-8 CONTAINMENT AND INTERNAL STRUCTURES SRSS ACCELE RATIONS AND DISPLACEMENTS FOR SAFE SHUTDOWN EARTHQUAKE CRACKED MODEL, RESPONSE SPECTRUM ANALYSISDegree of *FreedomAcceleration **Displacement ***HorizontalVerticalHorizontalVerticalN-SE-WN-SE-W MPS3 UFSARMPS3 UFSAR3.7-165Rev. 30540.491xE-20.185xE-20.187xE-20.103xE-30.222xE-40.098xE-3550.1450.0520.0260.439xE-20.341xE-30.097xE-3560.0290.0060.0890.029xE-30.010xE-30.608xE-3570.0520.1270.0150.340xE-30.443xE-20.059xE-3580.162xE-30.733xE-30.702xE-40.030xE-40.447xE-40.003xE-4590.366xE-30.863xE-40.335xE-30.010xE-40.012xE-40.001xE-4600.740xE-30.157xE-30.169xE-30.444xE-40.030xE-40.005xE-4610.2210.0650.0340.972xE-20.707xE-30.15xE-3620.0220.0070.1410.033xE-30.009xE-30.114xE-2630.0650.2070.0200.706xE-30.980xE-20.090xE-3640.329xE-30.132xE-20.130xE-30.055xE-40.808xE-40.005xE-4650.220xE-30.106xE-30.197xE-30.018xE-40.021xE-40.001xE-4660.132xE-20.321xE-30.277xE-30.802xE-40.055xE-40.010xE-4670.2890.0650.0320.155xE-10.108xE-20.140xE-3680.0220.0050.1830.04xE-30.012xE-30.162xE-2690.0640.2860.0180.108xE-20.156xE-10.082xE-3700.458xE-30.175xE-20.193xE-30.075xE-40.109xE-30.008xE-4710.283xE-30.136xE-30.235xE-30.026xE-40.030xE-40.001xE-4TABLE 3.7B-8 CONTAINMENT AND INTERNAL STRUCTURES SRSS ACCELE RATIONS AND DISPLACEMENTS FOR SAFE SHUTDOWN EARTHQUAKE CRACKED MODEL, RESPONSE SPECTRUM ANALYSISDegree of *FreedomAcceleration **Displacement ***HorizontalVerticalHorizontalVerticalN-SE-WN-SE-W MPS3 UFSARMPS3 UFSAR3.7-166Rev. 30720.176xE-20.450xE-30.378xE-30.108xE-30.075xE-40.015xE-4730.3510.0580.0210.215xE-10.147xE-20.073xE-3740.0280.0050.2270.047xE-30.017xE-30.204xE-2750.0580.3550.0110.146xE-20.216xE-10.040xE-3760.538xE-30.211xE-20.240xE-30.090xE-40.130xE-30.011xE-4770.366xE-30.172xE-30.318xE-30.033xE-40.038xE-40.002xE-4780.213xE-20.530xE-30.450xE-30.129xE-30.090xE-40.019xE-4790.4160.0390.0100.273xE-10.185xE-20.049xE-3800.0170.0070.2630.053xE-30.220xE-30.238xE-2810.0390.4230.0060.185xE-20.275xE-10.028xE-3820.519xE-30.243xE-20.255xE-30.101xE-40.145xE-30.011xE-4830.180xE-30.198xE-30.621xE-40.039xE-40.045xE-40.002xE-4840.244xE-20.583xE-30.481xE-30.144xE-30.101xE-40.020xE-4850.4760.0610.0260.317xE-10.214xE-20.114xE-3860.0200.0080.2820.058xE-30.026xE-30.257xE-2870.0610.4840.0150.214xE-20.320xE-10.068xE-3880.584xE-30.262xE-20.248xE-30.106xE-40.153xE-30.011xE-4890.347xE-30.222xE-30.276xE-30.043xE-40.049xE-40.002xE-4TABLE 3.7B-8 CONTAINMENT AND INTERNAL STRUCTURES SRSS ACCELE RATIONS AND DISPLACEMENTS FOR SAFE SHUTDOWN EARTHQUAKE CRACKED MODEL, RESPONSE SPECTRUM ANALYSISDegree of *FreedomAcceleration **Displacement ***HorizontalVerticalHorizontalVerticalN-SE-WN-SE-W MPS3 UFSARMPS3 UFSAR3.7-167Rev. 30NOTES:* Degrees of freedom corresponding to x-and y- direction translation and rotati on about the z axis are identified in Table 3.7B-9
.** Tabulated accelerations are in units of g for translational degrees of freedom, a nd g/feet for rotational degrees of freedom
.*** Tabulated displacements are in units of feet for translati onal degrees of freedom, and rads for rotational degrees of freed om.**** 10E-4 = 10
-4 = 0.0010900.262xE-20.576xE-30.478xE-30.151xE-30.106xE-40.019xE-4910.5800.0580.0290.399xE-10.269xE-20.135xE-3920.0420.0090.3180.067xE-30.03xE-30.283xE-2 930.0580.5840.0160.268xE-20.403xE-10.077xE-3940.119xE-20.597xE-20.427xE-30.188xE-40.249xE-30.019xE-4950.444xE-30.533xE-30.200xE-40.105xE-40.123xE-40.004xE-4960.597xE-20.119xE-20.842xE-30.247xE-30.188xE-40.036xE-4970.7700.0670.0310.506xE-10.343xE-20.141xE-3980.0440.0120.3560.084xE-30.037xE-30.314xE-2990.0670.7760.0130.342xE-20.510xE-10.061xE-31000.191xE-20.761xE-20.732xE-30.225xE-40.280xE-30.034xE-41010.531xE-30.645xE-30.221xE-40.125xE-40.148xE-40.005xE-41020.773xE-20.190xE-20.132xE-20.278xE-30.225xE-40.059xE-4TABLE 3.7B-8 CONTAINMENT AND INTERNAL STRUCTURES SRSS ACCELE RATIONS AND DISPLACEMENTS FOR SAFE SHUTDOWN EARTHQUAKE CRACKED MODEL, RESPONSE SPECTRUM ANALYSISDegree of *FreedomAcceleration **Displacement ***HorizontalVerticalHorizontalVerticalN-SE-WN-SE-W MPS3 UFSAR3.7-168Rev. 30TABLE 3.7B-9 CONTAINMENT AND INTERNAL STRUCTURES DEGREES OF FREEDOMMass PointDegree of FreedomDirection of Motion (Global Coordinates) *11TranslationX12TranslationY13TranslationZ 14RotationX15RotationY16RotationZ 27TranslationX28TranslationY29TranslationZ 210RotationX211RotationY212RotationZ 313TranslationX314TranslationY315TranslationZ 316RotationX317RotationY318RotationZ 419TranslationX420TranslationY421TranslationZ 422RotationX423RotationY424RotationZ525TranslationX526TranslationY 527TranslationZ MPS3 UFSAR3.7-169Rev. 30528RotationX529RotationY530RotationZ 631TranslationX632TranslationY633TranslationZ 634RotationX635RotationY636RotationZ 737TranslationX738TranslationY739TranslationZ 740RotationX741RotationY742RotationZ 843TranslationX844TranslationY845TranslationZ 846RotationX847RotationY848RotationZ 949TranslationX950TranslationY951TranslationZ952RotationX953RotationY 954RotationZ1055TranslationXTABLE 3.7B-9 CONTAINMENT AND INTERNAL STRUCTURES DEGREES OF FREEDOMMass PointDegree of FreedomDirection of Motion (Global Coordinates)
* MPS3 UFSAR3.7-170Rev. 301056TranslationY1057TranslationZ1058RotationX 1059RotationY1060RotationZ1161TranslationX 1162TranslationY1163TranslationZ1164RotationX 1165RotationY1166RotationZ1267TranslationX 1268TranslationY1269TranslationZ1270RotationX 1271RotationY1272RotationZ1373TranslationX 1374TranslationY1375TranslationZ1376RotationX 1377RotationY1378RotationZ1479TranslationX1480TranslationY1481TranslationZ 1482RotationX1483RotationYTABLE 3.7B-9 CONTAINMENT AND INTERNAL STRUCTURES DEGREES OF FREEDOMMass PointDegree of FreedomDirection of Motion (Global Coordinates)
* MPS3 UFSAR3.7-171Rev. 301484RotationZ1585TranslationX1586TranslationY 1587TranslationZ1588RotationX1589RotationY 1590RotationZ1691TranslationX1692TranslationY 1693TranslationZ1694RotationX1695RotationY 1696RotationZ1797TranslationX1798TranslationY 1799TranslationZ17100RotationX17101RotationY 17102RotationZTABLE 3.7B-9 CONTAINMENT AND INTERNAL STRUCTURES DEGREES OF FREEDOMMass PointDegree of FreedomDirection of Motion (Global Coordinates)
* MPS3 UFSARMPS3 UFSAR3.7-172Rev. 30NOTES:* Maximum of cracked, uncracked** Tabulated accelerations are in units of gTABLE 3.7B-10 CONTAINMENT AND INTERNAL STRUCTUR E ENVELOPED ACCE LERATIONS AND DISPLACEMENTS HORIZONTAL-MOTION *MassNo.Elevation (ft)Accelerations **Displacement ***SSE ****OBESSEOBEN-SE-WN-SE-WN-SE-WN-SE-WInternal Structure 1(-) 37.250.1700.1700.0900.0900.0040.0060.0030.0042(-) 11.250.2940.2980.1680.1920.0390.0600.0250.041 33.00.3570.4140.2300.2830.0780.1180.0510.081411.270.3950.3970.2670.2720.0920.1210.0600.083518.170.4210.4470.2700.3040.1050.1450.0680.100 624.50.4510.4560.2900.3090.1250.1540.0770.105750.840.5410.5170.3490.3520.1670.2190.1080.149886.50.8550.8210.5520.5580.2430.3160.1580.215 9109.11.0701.0100.6940.6850.2840.3560.1840.243External Structure 10(-) 6.750.2230.1940.1280.1140.0580.0580.0380.038 1116.250.3210.2920.1960.1800.1270.1270.0830.0831239.250.3860.3680.2400.2300.2010.2020.1300.1311362.250.4390.4240.2940.2680.2760.2780.1800.181 1485.250.5300.4680.3570.3030.3500.3530.2280.22915104.00.6010.5600.4070.3580.4080.4100.2650.26716133.50.7450.6580.5020.4270.5130.5160.3340.336 17163.30.9420.8550.6420.5560.6500.6540.4230.425 MPS3 UFSAR3.7-173Rev. 30*** Displacements are listed in inches**** N-S applies to X-direction, E-W applies to Z-direction MPS3 UFSAR3.7-174Rev. 30NOTES:* Maximum of cracked, uncracked** Tabulated accelerations are in units of g*** Displacements are listed in inchesTABLE 3.7B-11 CONTAINMENT AND INTERNAL STRUCTURE ENVELOPED ACCELERATIONS AND DISPLACEMENTS VERTICAL-MOTION
* Mass No.Elevation (ft) Accelerations **Displacements ***SSE ****OBESSEOBEInternal Structure 1(-) 37.250.1130.0600.0020.001 2(-) 11.250.1080.0680.0070.005 33.00.1880.1210.0200.013 411.270.1720.1110.0110.007 518.170.2020.1300.0180.012 624.50.2040.1320.0190.012 750.840.2510.1620.0240.016 886.50.3240.2070.0340.022 9109.10.3170.1990.0230.015External Structure 10(-) 6.750.1240.0710.0080.005 1116.250.1790.1030.0140.009 1239.250.2250.1320.0200.013 1362.250.2680.1630.0250.016 1485.250.3080.1820.0290.019 15104.00.3300.1960.0320.021 16133.50.3690.2300.0350.023 17163.30.4310.2590.0390.025 MPS3 UFSAR3.7-175Rev. 30TABLE 3.7B-12 MAIN STEAM VALVE BUILDING SIGNIFICANT MODAL FREQUENCIES AND PARTICIPATION FACTORSModeFrequency (Hz)Participation FactorsXYZ18.380126.1010.9630.63029.4560.813-2.172-133.882313.09317.6720.5100.919 423.9917.296-46.76691.569524.478108.4175.424-6.261625.6632.749-159.477-30.236 729.68923.7915.2973.240833.83336.719-1.254-0.800934.3062.605-18.09235.379 1050.5492.403-28.776-121.3361151.756125.9390.5081.6851252.4691.052-110.17348.891 1361.22120.45926.39010.1751461.4387.329-51.449-28.9321564.4073.223-2.9010.345 1684.8531.02559.0452.653 MPS3 UFSAR3.7-176Rev. 30TABLE 3.7B-13 MAIN STEAM VALVE BUIL DING SIGNIFICANT MODE SHAPES (EIGENVECTORS)ModeJointDisplacementXYZ110.0130.0000.00020.0430.0000.00030.2320.0050.002 41.00.0050.005210.0000.0000.01720.0000.0010.053 3-0.0010.0110.3184-0.0070.0111.0310.0330.0010.00220.1060.0010.01030.2050.017 -0.03441.00.0200.091410.007-0.0360.09320.014-0.0500.16730.074-0.1341.0 40.008-0.357 -0.155510.1010.004 -0.00620.2090.005 -0.011 31.00.014 -0.0564-0.1660.0360.00761-0.0030.1250.0322-0.001-0.1730.0643-0.013-0.4580.3214-0.0131.0 -0.094710.2100.0360.02920.1670.0470.06630.2560.0660.124 MPS3 UFSAR3.7-177Rev. 3041.00.217 -0.007810.160-0.004 -0.00320.403-0.006 -0.009 31.00.010 -0.0754-0.696-0.0330.069910.009-0.0480.12420.021-0.0640.30530.053-0.1051.04-0.035-0.146 -0.776101-0.0120.1070.5992-0.0180.1351.030.0040.256 -0.350 40.0 -0.1590.0261110.6290.0020.00821.00.0010.014 3-0.3930.009 -0.00540.033-0.0060.0121-0.0070.533 -0.3152-0.0040.678 -0.3693-0.0031.00.0904-0.001-0.662 -0.0531310.8570.8290.4262-0.1931.00.09230.6190.2870.07040.186-0.4670.094141-0.1570.8290.62120.0401.00.1483-0.1170.5080.079TABLE 3.7B-13 MAIN STEAM VALVE BUIL DING SIGNIFICANT MODE SHAPES (EIGENVECTORS)ModeJointDisplacementXYZ MPS3 UFSAR3.7-178Rev. 304-0.034-0.7040.133151-0.7690.519 -0.0822-0.0550.617 -0.858 3-0.0260.1551.04-0.264-0.3870.042161-0.015-0.634 -0.03820.002-0.6460.00330.01.0-0.00240.002-0.2240.008TABLE 3.7B-13 MAIN STEAM VALVE BUIL DING SIGNIFICANT MODE SHAPES (EIGENVECTORS)ModeJointDisplacementXYZ MPS3 UFSAR3.7-179Rev. 30NOTES:* Closely spaced modes.** The relationship between the degrees of freedom and the moti on of the lumped masses is given in Table 3.7B-17
.TABLE 3.7B-14 MAIN STEAM VALVE BUIL DING CSM* ACCELERATIONS FOR SAFE SHUTDOWN EARTHQUAKE FROM RESPONSE SPECTRUM ANALYSISDegree of Freedom **  X ***  Y***  Z ***1 0.1030.0060.0152 0.0070.0590.0453 0.0150.0400.1074 0.00.0010.002 5 0.0010.00.0016 0.0020.00.07 0.1560.0030.012 8 0.0080.0720.0579 0.0120.0500.160100.0010.0020.003110.0010.00.0120.0030.0010.001130.1930.0100.018 140.0080.1150.089150.0190.0830.187160.0010.0030.004 170.0030.00.0180.0030.0020.001190.4140.0040.005 200.0140.1680.095210.0060.0220.395220.0010.0020.006 230.0030.0 0.0240.0050.0010.001 MPS3 UFSAR3.7-180Rev. 30*** Tabulated accelerations are in units of g fo r translational degrees of freedom, and g/feet for rotational degrees.
MPS3 UFSAR3.7-181Rev. 30TABLE 3.7B-15 MAIN STEAM VALVE BUIL DING CSM* DISP LACEMENTS FOR SAFE SHUTDOWN EARTHQUAKE FROM RESPONSE SPECTRUM ANALYSISDegree of Freedom **X ***Y ***Z ***
1 0.703x10-4 0.210x10-5 0.310x10-5 2 0.230x10-5 0.279x10-4 0.173x10-4 3 0.310x10-5 0.154x10-4 0.716x10-4 4 0.10x10-6 0.20x10-6 0.330x10-5 5 0.40x10-6 0.0 0.10x10-6 6 0.42x10-5 0.10x10-6 0.10x10-6 7 0.217x10-3 0.270x10-5 0.560x10-5 8 0.310x10-5 0.379x10-4 0.230x10-4 9 0.540x10-5 0.246x10-4 0.198x10-3 10 0.10x10-6 0.40x10-6 0.660x10-5 11 0.150x10-5 0.10x10-6 0.30x10-6 12 0.840x10-5 0.20x10-6 0.20x10-6 13 0.113x10-2 0.143x10-4 0.259x10-4 14 0.237x10-4 0.927x10-4 0.630x10-4 15 0.287x10-4 0.107x10-30.116x10-2 16 0.30x10-60.110x10-5 0.160x10-4 17 0.55x10-5 0.30x10-6 0.10x10-5 18 0.242x10-4 0.50x10-6 0.40x10-6 MPS3 UFSAR3.7-182Rev. 30NOTES:* Closely spaced modes.** The relationship between the degrees of freedom and the moti on of the lumped masses is given in Table 3.7B-17
.*** Tabulated displacements are in units of feet for translational de grees of freedom, and g/radians for rotational degrees.
19 0.480x10-2 0.249x10-4 0.406x10-4 20 0.289x10-4 0.199x10-30.111x10-3 21 0.379x10-4 0.463x10-3 0.340x10-2 22 0.90x10-6 0.240x10-5 0.336x10-4 23 0.274x10-4 0.20x10-6 0.140x10-5 24 0.417x10-4 0.60x10-6 0.60x10-6TABLE 3.7B-15 MAIN STEAM VALVE BUIL DING CSM* DISP LACEMENTS FOR SAFE SHUTDOWN EARTHQUAKE FROM RESPONSE SPECTRUM ANALYSISDegree of Freedom **X ***Y ***Z ***
MPS3 UFSAR3.7-183Rev. 30TABLE 3.7B-16 MAIN STEAM VALVE BUIL DING ABS
* OF SAFE SHUTDOWN EARTHQUAKE RESPONSESDegree of Freedom **Acceleration ***Displacement ****10.170 0.750x10-420.113 0.50x10-430.170 0.917x10-440.003 0.360x10-550.002 0.50x10-660.002 0.440x10-570.170 0.225x10-380.137 0.667x10-490.222 0.225x10-3100.006 0.710x10-5110.001 0.190x10-5120.005 0.880x10-5130.2210.117x10-2140.212 0.183x10-3150.290 0.129x10-2160.008 0.174x10-4170.003 0.680x10-5180.006 0.251x10-4190.424 0.487x10-2200.277 0.342x10-3210.423 0.368x10-2220.009 0.369x10-4 MPS3 UFSAR3.7-184Rev. 30NOTES:* 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.230.003 0.290x10-4240.007 0.429x10-4TABLE 3.7B-16 MAIN STEAM VALVE BUIL DING ABS
* OF SAFE SHUTDOWN EARTHQUAKE RESPONSESDegree of Freedom **Acceleration ***Displacement ****
MPS3 UFSAR3.7-185Rev. 30NOTE:* For locations of lumped masses, refer to Figure 3.7b-10
.TABLE 3.7B-17 MAIN STEAM VALVE BUILDING DEGREES OF FREEDOMMass Point*Degree of FreedomDirection of Motion11Translation X12Translation Y13Translation Z 14Rotation X 15Rotation Y 16Rotation Z 27Translation X 28Translation Y 29Translation Z 210Rotation X 211Rotation Y 212Rotation Z 313Translation X 314Translation Y 315Translation Z 316Rotation X 317Rotation Y 318Rotation Z 419Translation X 420Translation Y 421Translation Z 422Rotation X 423Rotation Y 424Rotation Z MPS3 UFSAR3.7-186Rev. 30
* N-S applies to X-direction, E-W applies to Z-direction.** Tabulated accelerations are in units of g.TABLE 3.7B-18 EMERGENCY GENERATOR ENCLOSURE ACCELERATIONS AND DISPLACEMENTS HORIZONTAL-MOTION*Accelerations **Mass No.Elevation (ft-in) SSEOBE N-S E-W N-S E-W 366-00.6600.5470.4260.357251-00.5140.4380.3410.288 124-60.3170.2850.1730.163 MPS3 UFSAR3.7-187Rev. 30
*** Displacements are listed in inches.Displacements
***Mass No.Elevation (ft-in) SSEOBE N-S E-W N-S E-W 366-00.10700.04250.06920.0278251-00.08300.03400.05510.0224124-60.03790.01700.02340.0102 MPS3 UFSAR3.7-188Rev. 30NOTES:*Tabulated accelerations are in units of g.**Displacements are listed in inchesTABLE 3.7B-19 EMERGENCY GENERATOR ENCLOSURE ACCELERATIONS AND DISPLACEMENTS VERTICAL - MOTIONMass No.Elevation ft-in Accelerations* SSE OBE 366-00.2500.148241-00.2200.128124-60.1760.100Mass No.Elevation ft-inDisplacements **SSEOBE366-00.00550.0033251-00.00480.0028 124-60.00400.0024 MPS3 UFSAR3.7-189Rev. 30NOTES:* Figure 3.7B-9 shows the location of the lumped masses.** Safe shutdown earthquake accelerations are in units of g.TABLE 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-SE-WVerticalN-SE-WVertical10.1700.1700.1130.1800.1790.140 20.2710.2980.1090.2220.2460.17030.3360.4140.1740.2760.3420.20540.3700.3970.1610.3020.3350.208 50.3920.4470.1830.3200.3700.21860.4150.4560.1860.3410.3800.22370.4860.5170.2240.4290.4830.249 80.7480.8210.3030.5980.6750.28290.9461.0100.3120.7010.7600.290100.1880.1710.1220.2260.2130.181110.2600.2350.1790.3020.2900.224120.3590.3160.2250.3830.3540.263130.4390.3760.2680.4510.4060.302 140.5300.4490.3080.5020.4620.336150.6010.5010.3300.5410.5070.355160.7450.6210.3690.6710.6200.380 170.9420.7920.4310.8390.8090.411 MPS3 UFSAR3.7-190Rev. 30NOTES:* Figure 3.7B-9 shows the location of the lumped masses.** Safe shutdown earthquake accelerations are in units of g.TABLE 3.7B-21 COMPARISON OF RESPONSE SPECTRA AND TIME HISTORY ANALYSIS RESULTS CONTAINMENT AND INTERNAL STRUCTURES (CRACKED PROPERTIES)Mass No. *SSE Accelerations **Response Spectrum Time History N-SE-WVerticalN-SE-WVertical10.1700.170 0.113 0.1780.1780.127 20.2940.258 0.108 0.2250.2420.16130.3570.340 0.188 0.3020.3380.19740.3950.329 0.172 0.3320.3390.204 50.4210.372 0.202 0.3550.3730.21160.4510.385 0.204 0.3800.3820.21570.5410.467 0.251 0.4860.4640.239 80.8550.724 0.324 0.6790.6380.27091.0700.878 0.317 0.7820.7140.281100.2230.194 0.125 0.2150.2150.159 110.3210.292 0.169 0.2720.2760.236 120.3860.368 0.210 0.3580.3610.304130.4300.424 0.260 0.4310.4350.382140.4650.468 0.287 0.4900.4980.437 150.5620.560 0.310 0.5350.5380.464160.6670.658 0.369 0.7170.7110.497170.8670.855 0.412 0.9620.9910.538 MPS3 UFSAR3.7-191Rev. 30NOTES:* Safe shutdown earthquake accelerations are in units of g.** Figure 3.7B-10 shows the location of the lumped masses.TABLE 3.7B-22 COMPARISON OF RESPONSE SPECTRA AND TIME HISTORY ANALYSIS RESULTS MAIN STEAM VALVE BUILDING SSE ACCELERATIONS*Mass No. **Response Spectrum Time History N-SE-WVerticalN-SE-WVertical10.1700.1700.1130.1740.1740.11820.1700.2220.1370.1800.1810.12030.2210.2900.2120.2500.2540.135 40.4240.4230.2770.4880.4410.159 MPS3 UFSARMPS3 UFSAR3.7-192Rev. 30NOTES:TABLE 3.7B-23 1 G FLAT RESPONSE Max Dynamic      Max Static Model Beam Fundamental FrequencySumMoment (in-lbLoad Type Moment (in-lb)LocationS/C A/U K S/UCantilever1SRSS620,000Conc *700,000Fixed End0.890.890.99ABS649,000Unif **700,000Simple-Simple1SRSS179,000Conc348,000Midspan 0.511.031.07ABS186,000Unif174,000Fixed-Fixed1SRSS103,000Conc174,000Fixed End0.590.890.97ABS112,000Unif116,000Simple-Fixed - No Overhang1SRSS152,000Conc261,000Fixed End0.580.870.97ABS169,000Unif174,000Simple-Fixed - 16%
Overhang1.34SRSS83,200Conc162,000Fixed End0.510.751.03ABS114,000Unif111,000Simple-Fixed - 33%
Overhang1.04SRSS57,000Conc77,400Fixed End0.740.741.15ABS89,000Unif77,200Simple-Fixed - 50%
Overhang0.62SRSS152,000Conc174,000Simple Support0.870.871.01ABS176,000Unif174,000 MPS3 UFSAR3.7-193Rev. 30* Conc = Concentrated load** Unif = Uniform load MPS3 UFSAR3.7-194Rev. 30NOTE:*Modal density is based on a
+/-10 percent criterion.TABLE 3.7B-24 MODAL DENSITY, n*Mode No Cantilever Frequency (Hz)Fixed-Fixed Frequency (Hz)Simple-Fixed Frequency (Hz)Simple-Simple Frequency (Hz)Simple-Fixed 33% Overhang Frequency (Hz)11.01.01.01.01.0 25.82.73.23.8 2.9315.34.96.38.26.5422.87.510.213.6 8.4 528.010.214.019.513.3643.212.917.925.417.6 MPS3 UFSARMPS3 UFSAR3.7-195Rev. 30TABLE 3.7B-25 AMPLIFIED RESPONSE DYNAMIC FACTOR STUDY(For Simple-Fixed Be am with 33% Overhang
)Model Beam First ModeDynamic LoadMaximum Dynamic Moment Sum Moment (S) (in-lb) LocationMaximum Static Moment Uniform Load (U) (in-lb)*K S/U A/UHighLowSimple-Fixed 33% g o g max g cOverhangf off cModel 6A0.102.870.33SRSS20,000Fixed222,000 s0.090.703-420ABS30,000Fixed0.13Model 6B0.102.870.33SRSS148,00Fixed222,000 s0.671.03-420ABS157,00Fixed0.71Model 6C2.872.870.33SRSS102,00Simple Support222,000 s0.453.33-420ABS118,00Simple Support 0.53Model 6D0.402.870.33SRSS22,000Fixed31,000 s0.7110.03-420ABS32,000Fixed1.03Model 6E0.332.870.33SRSS20,000Fixed25,700 s0.7820.03-420ABS27,000Fixed1.05Model 6F0.302.870.33SRSS18,000Fixed23,400 s0.7733.03-420ABS25,000Fixed1.07 MPS3 UFSAR3.7-196Rev. 30NOTE:* g = gmax, if f o < f p; g at f o  f p (where f p is the frequency at peak)
MPS3 UFSARMPS3 UFSAR3.7-197Rev. 30TABLE 3.7B-26 PIPING SYSTEM SEISMI C DESIGN AND ANALYSIS CRITERIAASME Section III Code ClassType of EarthquakeType of Seismic AnalysesCombined Stress Calculations and Stress Criteria Class 1  (Sizes 1.25 inch NPS and larger)SSEDynamic response spectraASME C ode, Section III, Subarticle NB-3600OBEDynamic response spectraASME C ode, Section III, Subarticle NB-3600 Class 1  (Sizes 1 inch NPS and below)SSESimplified analysesASME C ode, Section III, Subarticle NB-3600OBESimplified analysesASME C ode, Section III, Subarticle NB-3600 Classes 2 and 3 (Sizes 2.5 inch NPS and larger)SSEDynamic response spectraASME C ode, Section III, Subarticles NC-3600 and ND-3600OBEDynamic response spectraASME C ode, Section III, Subarticles NC-3600 and ND-3600 Classes 2 and 3 (Sizes 2 inch NPS and below)SSEDynamic response spectra or simplified analyses ASME Code, Section III, Subarticles NC-3600 and ND-3600OBEDynamic response spectra or simplified analyses ASME Code, Section III, Subarticles NC-3600 and ND-3600 MPS3 UFSAR3.7-198Rev. 30TABLE 3.7N-1 DAMPING VALUES USED FOR SEISMIC ANALYSIS FOR WESTINGHOUSE SUPPLIED EQUIPMENT        Damping (Percent of Critical)
Item Upset Conditions (OBE)Faulted Conditions (SSE, DBA)Primary coolant loop system components24Welded steel structures24 Bolted and/or riveted steel structures47Fuel assemblies710CRDMs/CRDM supports55 MPS3 UFSAR3.8-1Rev. 30
 
===3.8 DESIGN===
OF CATEGORY I STRUCTURES
 
====3.8.1 CONCRETE====
CONTAINMENTThis section provides the following information on concrete containment and on concrete portions of steel/concrete containments:1.The physical description2.The applicable design codes, standards, and specifications3.The loading criteria, including loads and load combinations4.The design and analysis procedures5.The structural acceptance criteria6.The materials, quality control progra ms, and special construction techniques7.The testing and inserv ice inspection programs 3.8.1.1 Description of the ContainmentThe containment structure is a steel-lined conventionally reinforced concrete pressurized water reactor containment structure designed to operate under s ub-atmospheric conditions. The structure has a vertical cylindrical wall and hemispherical dome supported on a flat base mat which is founded on bedrock (Figures 3.8-1 and 3.8-2).
The design, analyses, and construction of the containment structure is similar to the designs for the 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 N
: o. 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)
MPS3 UFSAR3.8-2Rev. 30 3.8.1.1.1 Base FoundationThe base foundation slab is 10 feet thick with a 158 foot diameter. The bottom reinforcement is a rectangular grid pattern and the top reinforcement consists of concentric circular bars combined with radial bars arranged to permit uniform spacing of the vertical wall reinforcing bars (rebars) which extend into the mat. Splices in adjacent parallel bars are staggered.The floor liner plate of 0.25 inch thickness is anchored to the mat by means of rebars welded to 7 inch by 0.5 inch continuous vertical plates wh ich are in turn welded to the liner seams (Figure 3.8-3). Where interior wall and column rebars are anchored to the mat, a 3 inch x 6 inch rectangular bar (bridging bar) or 1.25 inch thick plate was welded through the liner with test channels all around to ensure the continuity of the steel membrane. The vertical rebar was Cadwelded to the top of the bar and butt-welded to the bottom of these members providing continuity of rebar wit hout creating multiple penetrations of the liner.A reinforced concrete slab approximately 2 feet thick, was placed over and anchored through the mat liner to stiffen it against negative pressures a nd to protect it from heat associated with a DBA similar to other pressurized wa ter reactor containment structures (e.g., North Anna 1 and 2, Docket No. 50-338 and 50-339). his slab also serves as anchorage and support for equipment located on the lowest level of the containment.
3.8.1.1.2 Cylindrical WallThe cylindrical wall is 4 feet 6 inches thick with an inside diameter of 140 feet and a height from mat to spring line of 131 feet 3 inches. Hoop tension in the cylindrical wall of the containment structure is resisted by horizontal bars located near both the outer and inner faces of the wall. Splices in these bars are staggered where possible. Meridional tension is resisted by rows of vertical bars placed near each face of the wall. The vertical bars are placed in groups of not more than 20 bars of equal length. These groups are arranged so that no adjacent group in the same face of the wall has splices closer than 3 feet apart vertically.Radial shear in the containment structure wall resulting from the design basis accident (DBA) varies from a maximum at the base of the wall and approaches zero at some level above the top of the mat. To resist the large radial shear near the base of the wall, flat steel bars inclined at approximately 45 degrees with the horizontal are welded to the vertical reinforcement as shown in Figure 3.8-4. The welded flat bars terminate at approximately elevation (-) 18 feet where the radial shear load is reduced. Ab ove this level, radial shear is re sisted by Z type reinforcing steel. This Z reinforcing continues to elevation (-) 4 feet (-) 8 inches where the shear becomes insignificant. In the lower portion of the cont ainment structure wall, supplementary reinforcing steel, normal to the face of the wall is provided to resist potential splitting of the concrete in the plane of the vertical bars. Tangential shear (Vu) resulting from the earthquake loading is resisted by the concrete and diagonal reinforcing bars. A maximum allowable tangential shear stress is assigned to the concrete. Stresses in excess of Vc are resisted by inclined reinforcing bars anchored in the mat. No diagonal re bars are required above the spring line.
MPS3 UFSAR3.8-3Rev. 30A nominal minimum clearance of 2 inches is provided as an earthquake rattlespace at all levels between the interior steel wor k, and platforms and the containment structure liner. Actual clearances of less than 2 inches were reviewed on a case-by-case basis and were determined to provide adequate clearance to preclude interaction with the containment liner for the worst case design bases movements.Penetrations in the liner are pr ovided for piping, electrical conductors, fuel transfer tube, and purge air ducts as indicated in Section 3.8.1.1.4. Access openings in the structure are the 7 foot inside diameter personne l access lock and the 15 foot in side diameter equipment hatch (Figure 3.8-7). The reinforcement around these penetrations is shown in Figures 3.8-5 through 3.8-9.Brackets for support of internal miscellaneous small piping, and electrical conductors are attached to insert plates in the liner and/or to overlay plates on the liner. Major equipment and pipe loads, except at the penetrations, are not carried by the exterior wall. Steel anchors were used to attach the insert plates and the liner to the concrete structure. External supports on the containment wall for the enclosure structure are shown in Figures 3.8-10 and 3.8-61.
3.8.1.1.3 DomeThe inside radius of the 2 foot 6 inch thick dome is 70 feet. The internal height from base mat to the center of the dome is 201 feet 3 inches. The dome reinforcement consists of a layer of reinforcement placed meridionally extending from the vertical bars of the cylindrical wall near each face and horizontal hoop bars in similar layers. Mechanical anchorages are provided on the last set of meridional bars extending up toward the crown of the dome. All other meridional bars are terminated at a point where they are adequately developed beyond the point at which they were required for design purposes. Near the crown meridional bars are welded to a ring cast in the concrete concentric with the centerline of the containment. Figure 3.8-11 shows the rebar at the dome cylinder junction.The dome liner plate is 0.5 inch thick. Attachments to the dom e are basically for the quench and recirculation spray nozzle lines. One major connection to the dome structure is the top pivot support for the spray header main tenance truss (Figure 3.8-12). Th is structure is supported at the top of the dome and on the polar crane rail.
3.8.1.1.4 Steel Liner and Penetrations A.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.
MPS3 UFSAR3.8-4Rev. 30The liner plate is a continuously welded steel membrane supported by and anchored to the inside of the containment at sufficiently cl ose intervals with anchor studs and deformed bars so that the overall defo rmation 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 DB A conditions, the liner is under a state of biaxial compressive strain due to thermal ef fects and during the test condition, the liner plate is under a state of biaxial tensile stra in. The anchor studs prevent buckling of the liner and act as nodal points. Tests conducted at Northeastern University, Boston, Massachusetts, using 5/8-inch di ameter 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 c onducted for the stud manufacturer under the direction of Dr. I.M. Viest (TRW, Inc. 1975) indicate that, with the manufacturer's recommended depth of embedment of the stud in concrete, the ulti mate strength of the stud material can be developed in direct tension. The reinforcement ring and liner adjacent to the hatches are anchored to the concre te 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 a nd over knuckle plate to liner seams.B.Embedments Three types of embedments are used to main tain 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 reinforci ng bar continuity without crea ting multiple penetrations.
C.Insert Plates Loads from supports for piping such as th e spray headers and other miscellaneous equipment are transferred to the containment c oncrete wall through inse rt plates and their anchors. Each insert plate is designed to provide a rigid base for the attached supports and anchor studs. Sufficient insert plate anchorag e is provided so that the adjacent liner plate sees negligible stress from the applied support loads and the leak tight integrity is well maintained.
D.Penetrations
 
MPS3 UFSAR3.8-5Rev. 30Penetrations are used for personnel and equipment access, process piping, electrical service, and a mechanical fuel transfer sy stem through the containment wall (refer to Figures 3.8-15 through 3.8-22). Containment penetrations are anchored and transfer loads to the reinforced concrete containment wall. Le ak 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 cla ssified 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 th ermally 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 peri od, from exceeding 150&deg;F except at local areas around the penetr ations which are allowed to have increased temperatures not exceeding 200
&deg;F; for accident or other short-term periods, the temperatures are not to exceed 350
&deg;F for the interior surface. However, local areas are allowed to reach 650
&deg;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. Th e cooling water for the cooling jacket is supplied by the component c ooling 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 pi pe insulation and for the cooling jacket. Piping located outside the containment carrying the cooling water does not require any secondary penetration of the containm ent 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 designe d so that stress concentrations are minimized. The penetration sleeve is welded to a liner reinforcement plate (Figure 3.8-16). Loads are transferred from pipi ng to the containment structure through the forged flued head to the sl eeve 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 tors ional pipe loads. Multiple small piping penetrations pass through a forged attach ed plate which is welded to a sleeve (Figure 3.8-17
).2.Unsleeved Piping Penetrations These penetrations consist of piping inst alled through the containment wall that are thermally cold piping systems a nd 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 MPS3 UFSAR3.8-6Rev. 30 Electrical penetrations (Figure 3.8-19) are used to carry electrical cables and instrumentation leads throug h the containment wall. hey range in size from small thermocouple leads to solid rods for power circuits. Each pe netration 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 cutt ing the containment liner or sleeve. The penetrations are constructed and tested in accordance with IEEE Standard 317 dated 9-20-72, "Electrical Pene tration 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 pene tration consists of a stai nless steel pipe installed inside a stainless steel enclosure with be llows expansion joints to compensate for differential movements of the buildings. Th e 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 diff erential 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 hydrauli cally latched and hydraulically swung. Both doors are interlocke d 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 va lves are manually and automatically operated by the person entering or leav ing the personnel access lock. The personnel access lock is externally prot ected 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 MPS3 UFSAR3.8-7Rev. 30 between the "O" rings is pressurized to containment design pressure to test for leakage through the access door when it is bolted in place. The equipment hatch cover is provided with a hoist with two point suspension and a sliding rail for storage. A positive locking device is fu rnished to prevent circular swing. A removable concrete tornado missile shield protects the equipment hatch. These concrete shield blocks are removed duri ng Modes 5 and 6. Base d 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.
3.8.1.1.5 Ring GirderA reinforced concrete ring girder is provided which encircles the containment structure and is isolated from the containment wall by a compressible material (Figure 3.8-23). The ring girder is provided to prevent postulated sliding of rock wedges toward the containment wall during a seismic event. Sliding would occur when seismic forces exceeded the frictional forces on the rock joint and/or foliation planes. The ring is analyzed as a laterally loaded and supported ring, in which only inward unrestrained deflection is possible. This is because the rock itself resists lateral outward movement. The ring girder does not interact with the containment structure except that the containment mat gives vertical support for the ring gravitational forces. In the area of the ESF building, radial walls in the lower chambers act as struts between the ring and the ESF building east wall (Figure 3.8-24). Through these walls, relatively continuous support for the ring is provided around this area of the containment. Lateral displacements of the ring are calculated to be a maximum of 0.2 inch, which is negligible with respect to the 4 inches of compressible material which is provided to isolate the containment shell from any lateral loads above the mat. The ring girder is a seismic Category I structure and is designed for all applicable loadings including: dead, seismic, and rock pressure loads. Loads and load combinations are in accordance with Section 3.8.1.3.The allowable shear stress for the design of the ring girder is in accordance with ACI 318-71 because no circumferential tension exists in the ring. Materials and quality control are in accordance with Section 3.8.1.6. There are no test ing or inservice inspection requirements.
3.8.1.2 Applicable Codes, Standards, and SpecificationsStructural design, materials, the tests of material and the methods of testing, where applicable, conform to the following codes, standards, and specifications unless otherwise stated.
3.8.1.2.1 GeneralThe letters ASTM, AISC, AWS, ACI, and BOCA refer to the American Society for Testing and Materials, American Institute of Steel Construction, American Welding Society, American Concrete Institute, and Building Officials and Code Administrators, respectively.Where multiple dates are shown for structural specifications in Section 3.8.1.2.2, these reflect documents whose applicable issue has ch anged during construc tion. The engineering specifications reflect the specific issue date in effect at any given time.
MPS3 UFSAR3.8-8Rev. 30 3.8.1.2.2 Structural Specifications
: 1. ACI 211.1-70 Recommended Practi ce 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 fo r Measuring, Mixing, and Placing Concrete 4b. ACI 304-73 Recommended Prac tice 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 Requireme nts for Reinforced Concrete
: 8. ACI 347-68 Recommended Practi ce for Concrete Formwork
: 9. AISC Specification for the Desi gn, 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 MPS3 UFSAR3.8-9Rev. 30
: 17. ASTM A 490-66, 74 Specification for Quenched and Tempered Alloy Steel Bolts for Structural Steel Joints
: 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 Conc rete Compressive and Flexural Strength Test Specimens in the Field
: 21. ASTM C 33-71a Standard Specificati on for Concrete Aggregates (and 1978 Revision)
: 22. ASTM C 94-71 (1974) Specificati on 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 Sl ump of Portland Cement Concrete
: 25. ASTM C 150-73 Specificati on 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 Hardne ss of Coarse Aggregate Particles
: 29. ASTM C 260-69 (1973, 1974) Air-entrai ning Admixtures for Concrete
: 30. ASTM C 289-71 Test for Potential Reactivity of Aggregates (Chemical Method) 31. ASTM C 295-1965 (1973) Recommended Practi ce 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 fa brication of watertight, airtight, pressure resi stant, 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 MPS3 UFSAR3.8-10Rev. 30b.Reinforcing Bar Testing 1.8.1.15c.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 Inte rnational, Inc., 1970
: 37. State of Connecticut Basic Building Code, 197 3.8.1.2.3 Steel Liner and PenetrationsThere was no applicable code for the design of concrete containment structure liners at the beginning of the construction of the Millstone liner. However, ASME Section III, Divisions 1 and 2, and Section VIII were used as a guide.Design, materials, fabrication, testing, and inspection, where applicable, conform to the following codes, standards, and specifications:a.ASME Boiler and Pressure Vessel Code Sections II, III, and V, 1971 issue, including addenda up to and including th e 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 Weldi ng Qualifications, Se ction IX, issue in effect at the time of qualification.c.American National Standa rds Institute, ANSI N 101.4, Quality Assurance for Protective Coatings Applied to Nuclear Facilities, 1972 issue.d.American Society for Testing a nd 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.
MPS3 UFSAR3.8-11Rev. 30g.Institute of Electrical and Electronics Engineers, IEEE 317, Elec trical Penetration Assemblies in Containment Structures for Nuclear Fueled Power Generating 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, Surf ace Preparation Spec ifications SP1, SP2, SP5, SP6, and SP10, 1963 issue.j.Steel Structures Painting Council, Surface Preparatio n Specification, SSPC-PS8.01, Rust Preventive System wi th Thick Film Compounds, 1964 issue.k.American Society for Nondestructive Testing, Recommended Practice  for Nondestructive Testing Personnel, Qu alifications, and Certification, ASNT-TC-1A, 1971 issue.l.American Society for Testing and Mate rials (ASTM) Specif ications 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, SA 508, SA516, and SA537.
Earlier or subsequent revisions (or reissues) of the referenced codes, standards, and specifications may be used insofar as their use does not in any way degrade the safety or strength of any structure or portion of a structure.Section 3.8.1.6 describes materials and quality control procedures and any exceptions to the above listed codes.
Fire Protection is described in Section 9.5.1.
3.8.1.3 Loads and Loading Combinations The following sections discuss the loads and lo ad combinations used in the design of the 3.8.1.3.1 Containment Mat, Shell, and Dome The loadings applicable to concrete containment design include the following:a.Those loads encountered dur ing preoperational testing.b.Those loads encountered during norma l plant startup, operation, and shutdown, including dead loads, live loads, and th ermal 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.
MPS3 UFSAR3.8-12Rev. 30d.Those loads to be sustained during extreme environmenta l conditions, including those induced by the design basis torn ado and the safe shutdown earthquake.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 a bnormal plant conditions including flooding of the containment subsequent to a LOCA for fuel recovery.
Loads generated by the design basis accident (DBA) are described in Section 6.2.1. The design basis wind and tornado loadings are discussed in Section 3.3. The loadings for groundwater and flood are given in Section 3.4.
Missile loads are desc ribed in Section 3.5.1.4.
The design bases for protection against postulated rupture of piping are described in Section 3.6.The earthquake loadings which include consideration of simultaneous excitation from two (orthogonal) horizontal and one vertical earthquake motion ar e described in Section 3.7.
Normal operating temperatures are described in Section 9.4.7.
For the containment structure shel l and its foundation mat, the load combinations given in Section 9.3.2 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 co mpressive stress in concrete is 0.45 f'c, the allowable stress for steel is 0.5 fy, and sh ear 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 Test1.OD + 1.OL + 1.0P t + 1.0T t 2Construction1.OD + 1.OL + 1.OT o
MPS3 UFSAR3.8-13Rev. 30 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.Nomenclature 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.
P a = Pressure load from DBA pressure transient, including the design margin, as defined in Section 6.2.1.
P t = Test pressure.
P v = Subatmospheric minimum ope rational pressure of c ontainment structure (or atmosphere).
R a = Piping loads due to increased te mperature resulting from the DBA.
R o = Pipe loads during nor mal operating conditions.
T a = Load due to temperature gradient thro ugh the concrete shell and mat plus load exerted by the liner, based on temperature a ssociated with the DBA. This is that  3Normal1.OD + 1.OL + 1.OT o + 1.0 OBE + 1.OR o + 1.0P v Factored:  4Extreme environmental 1.OD + 1.OL + 1.OT o + 1.OW t + 1.OP v + 1.OR o 5Extreme environmental 1.OD + 1.OL + 1.OT o + 1.0SSE + 1.OR o + 1.OP v 6Abnormal1.OD + 1.OL + 1.5P a + 1.OT a + 1.OR a 7Abnormal/Severe environmental 1.OD + 1.OL + 1.25P a + 1.OT a + 1.25OBE + 1.OR a 8Abnormal/Severe environmental 1.OD + 1.OL + 1.25P a + 1.OT a + 1.25W + 1.OR a 9Abnormal/Extreme environmental1.OD + 1.OL + 1.OPa + 1.OTa + 1.0SSE + 1.ORaCase Load Category Load Combination MPS3 UFSAR3.8-14Rev. 30transient temperature which, when combined with the coincident internal pressure, produces the most adverse effect s on the containment structure.
T o = Load due to temperature gradient through the concrete shell and mat, plus load exerted by the liner based on temperatur e associated with normal operation.
T t = Load from test temperature.W = Wind load (see note).
W t = Tornado load (see note).
Y r = Reaction from pipe restraint (penetration) due to pipe break (local effects).NOTE:Wind or tornado loads are not coincident with earthquake loads.A load factor of 1.0 is assigned to loads caused by the desi gn tornado because the wind velocity given in Section 3.3.2 is a conservative estimate of the actual tornado wind velocity to be expected, and the probability of a tornado striking the unit is low (Section 2.3.1.3). A load factor of 1.0 is also used for loads caused by the SSE (Section 3.7.1) since the SSE acceleration is two times the largest acceleration expected at the site.Tornado wind effects are determin ed as described in Section 3.3.The effects of hydrostatic loading due to groundwater and the probable maximum flood level (Section 3.4.2) are considered so as to reduce the effect of the dead load (D).During construction, dewatering was maintained until the weight of the containment exceeded the postulated hydrostatic uplift forces.
Section 3.8.1.4 gives the design stresses to be used with these loads.
3.8.1.3.2 Steel Liner and Penetrations The containment structure liner plate, penetrati ons, brackets and attachments, anchors, and access openings were designed for the lo ad combinations presented in Tables 3.8-1 and 3.8-2. Sleeves of hot penetrations were considered part of the containment boundary. The maximum containment structure design pressure is 45 psig. The minimum containment structure design pressure is 8.0 psia. This pressure is equivalent to the minimum operating pressure minus the pressure drop due to the maximum, hypothetical containment cooldown situation. During this condition, the containment atmosphere pressure is assumed to be decreased below nor mal operational pressure by the inadvertent operation of the quench spray system during normal unit operation. The resulting total pressure is above the minimum containment design pressure of 8.0 psia. For a further discussion of external pressure, refer to Section 6.2.1.1.3.5.The change in barometric pressure due to tornadoes is not expected to exceed 3 psi and the change due to the maximum probable hurricane is approximately 1.1 psi. These pressure changes result in a decrease in the atmospheric pressure, which, in turn, decreases the differential between MPS3 UFSAR3.8-15Rev. 30atmospheric pressure and the containment atmosphere ambient pr essure, thereby decreasing the level of stresses in the containment structure. Penetrations were designed to take pipe rupture, thermal, and seismic loads.
Figures 3.8-25 through 3.8-55 are scaled load plots of moments, shears, and membrane forces for the containment structure. The loading conditions are listed on the plots. Compressive forces are plotted as negative values and bending mo ments are plotted on the tension side.
3.8.1.4 Design and Analysis Procedures 3.8.1.4.1 Containment StructureThe containment structure consists of a hemispherical dome, a cylindrical shell, and a circular mat supported by an elastic subgrade. The design, analyses, and construction of the containment structure is similar to and takes full advantage of SWEC experience gained in the designs for plants indicated in Section 3.8.1.1.Under design internal pressure, discontinuity forces exist at the junction of the mat and the cylindrical shell and also at the junction of the shell and the dome. The cylindrical shell is of such a length that the influence of one disc ontinuity on the other is negligible.The analysis of the mat for axisymmetric loading is accomplished using the program MAT 5 which is described in Appendix 3A. This program treats the mat as a symmetrically loaded circular plate on an elastic foundation. The general method is described by Zhemochkin et al., (1962). The program is set up so that the foundation stiffness can be formulated either through Boussinesq's approach or through a Winkler-type assumption. Zhemochkin's method is modified to account for a finite depth of elastic foundation, i.e., the distance be tween mat and underlying rock. In addition, the cylindrical containment wall, crane wall, and primary shield wall are considered as elastic constraints, which are determined by appl ying compatibility conditions at the shell mat interface.
For the purpose of calculating the elastic constraint of the containment, the base of the cylinder is assumed to be completely cracked vertically and cracked horizontally to the neutral axis of the transformed section. At this location, the cylinder has a hoop stiffness of the circumferential rebars and the meridional bending stif fness of the transformed section.A short distance above the mat, the meridional bending moment becomes so small that the entire meridional cross-section is in a state of tension. Above this plane, the cylinder is assumed to be completely cracked horizontally and vertically.Thus, the elastic constraint is determined from a cylinder which is divided into a short shell having the properties of a cross-section completely cracked vertically and partially cracked horizontally, and a long shell having the properties of a cross-section completely cracked horizontally and vertically.
MPS3 UFSAR3.8-16Rev. 30Seismic analysis of the containment structure described in Se ction 3.7B.2 provides the dynamic loads imposed on the mat as static loads. Since these loads are asymmetric, the mat is analyzed using the finite difference computer program SHELL I.The discontinuity forces at the mat-shell junction, calculated as a part of the mat analysis, are used as boundary conditions for the analysis of the cylindrical shell which is performed using the program SHELL I. The seismic analysis of the containment structure provides the accelerations to which the containment structure would be subjected. These accelerations are used for determining static loading on the shell. The tangential shear caused by the seismic loading is resisted by the concrete or by the concrete and a system of diagonal rebars.
The maximum allowable tangential shear stress carried by the concrete, (Vc) is assumed to be 40 psi. Stresses in excess of Vc are resisted by diagonal reinforcing bars. The specified design compressive strength of the concrete carrying the tangential shear is not less than 3,000 psi with coarse aggregate not smaller than size No.67 as given in ASTM C33.The diagonal reinforcement required to supplement the tangential shear capacity of the concrete consists of No. 18 rebars anchored in the mat as shown in Figure 3.8-56. The spacing between these diagonal bars is in creased as the design tange ntial shear decreases at higher wall elevations.
No diagonal rebars are required above the spring line.
The requirements for diagonal reinforcement necessary for carrying the (Vu-Vc) shear force are determined by an analysis based on a paper by Prof. M. J. Holley, Jr. (1969). The mesh of vertical, horizontal, and diagonal reinforcement is assumed to be a continuum. Forces can be determined in each type of bar for symmetric loadings such as internal pressure, dead load, vertical, and horizontal earthquake loads. The diagonal reinforcement is designed to resist force due to symmetric loadings and the entire tangential shear not carried by the concrete (Vu-Vc).Two 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.
Under normal operating conditions, the temperature gradient to produce the highest stress resultant is used.The effect of the liner temperature increase associated with the DBA condition on the concrete containment shell is determined by the following procedure. For this condition, it is assumed that the temperature of the liner increases and the concrete remains at its ambient temperature. Liner expansion is limited by the concrete shell and a pressure develops between the concrete shell and the liner. The equivalent pressure exerted by the liner on the concrete shell is given by the expression:
P e = LH*hL/RL MPS3 UFSAR3.8-17Rev. 30 where: P e = Equivalent pressure LH = Circumferential liner stress hL = Thickness of liner RL = Radius of linerAt the junction of the mat and shell, it is assumed that the mat prevents radial movement of the shell; therefore, at this location, the circumferential stress in the liner used to determine equivalent pressure is:
LH = alpha*Es*delta T where:alpha = Coefficient of ther mal expansion of the linerEs = Young's modulus of the steel liner delta T = Change in temperature due to DBAA short distance above the mat, where the effect of the mat to shell discontinuity is negligible, the liner and concrete shell expand due to the DBA pressure and temperature of the liner. Free radial displacement of the liner due to DBA temperature is larger than the displacement of the reinforced concrete shell due to the DBA pressure. Thus, the liner is constrained by the concrete shell. The resulting pressure exerted by the liner is determined from the expression for equivalent pressure given before. The liner stress is de termined 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 fr om which liner stresses are determined.The effects of creep and shrinkage of concrete are not important considerations in the analysis and design of a nonprestressed concrete containment. Shrinkage results in meridi onal and radial displacements which are the opposite of the displacements caused by the principal loads, temperature and internal pressure. Consequently, the effects of creep and shrinkage can be safely ignored.Cracking is an important consideration in the analysis and design of a reinforced concrete containment. For this reason, stiffness of the concrete is adjusted for the extent of cracking present under various design conditions. When the concrete is completely cracked, calculation of the stiffness of the structure uses only the properties of the reinforcing steel. The steel liner is assumed to make no contribution to the structural integrity of th e containment shel l except during the structural acceptance test.
MPS3 UFSAR3.8-18Rev. 30 The penetrations through the containment wall ar e grouped into the follow ing three categories for the purposes of analysis and design: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 penetrat ions are replaced by at least twice the number of bars, half of these being pla ced 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 penetrati ons. This method is consistent with established practice and pressure-tes ted at the plants previously listed.3.Personnel Access and Equipment Access Hatches Penetrations for the equipment and pers onnel access hatches ar e analyzed using the 3-dimensional finite element capabi lity 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 re inforcing bars which are placed parallel to the typical earthquake shear bars.
The ring beam is designed to resist the ax ial 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 meridion al reinforcing bars, as designed, are extended to the inner face of the ring beam, hooked a nd Cadwelded to each other, thereby providing shear resistance additional to that provided in the design.
The normal pattern of membrane fo rces 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 provi ded by the finite element computer program and extra reinfo rcement is added to areas of marked deviation from the normal pattern.
MPS3 UFSAR3.8-19Rev. 30 3.8.1.4.2 Steel Liner and PenetrationsStresses due to strain compatibility of the liner with the reinforced concrete shell due to various combinations of pressure, thermal, self-weight, and seismic load were determined using Stone & Webster's "KALNINS" program. This is a direct integration program for static analysis of multilayered thin shells of revolution. The stress analysis of a shell subjected to mechanical and thermal surface loads and edge loads, is reduced to a boundary value problem governed by a system of nonhomogeneous, linear, partial differential equations. The equations are separable with respect to the meridional and circumferential coordinates of the shell. The solution for each separable component of the loads is obtained by solving a typical two point boundary value problem governed by eight first order linear ordinary differential equations using direct integration. Local stresses due to irregular spacing of liner-headed anchor studs were determined using the ANSYS finite element program and by manual calculations. Analytical evaluation of the penetration discontinuities were modeled on the ASAAS program (Asymmetric Stress Analysis of Axisymmetric Solids). The method of analysis employed is based on a finite element idealization of an axisymmetric solid. Each element is an axisymmetric ring of a constant cross section. Since such a solid may be loaded and may deform in nonaxisymmetric modes and since the properties of the material may vary in all directions (e.g., due to temperature variations), all the dependent variables including the material properties are expressed as truncated Fourier series with the circumferential coordinate being the independent variable.Temperature profiles for the penetrations were determined using the TAC-2D program. TAC-2D is a computer program for calculating steady-state and transient temperatures in two dimensional problems by the finite differen ce method. The configuration of the body to be analyzed is described in the rectangular, cylindrical, or circular (polar) coordinate system by orthogonal lines of constant coordinate called grid lines. The grid lines specify an array of nodal elements. Nodal points are defined as lying midway between the bounding grid lines of thes e elements. A finite difference equation is formulated for each nodal point in terms of its capacitance, heat generation and heat flow paths to neighboring nodal points.
3.8.1.5 Structural Acceptance Criteria 3.8.1.5.1 Containment StructureThe containment structure is designed for the loads and load combinations presented in Section 3.8.1.3.1. Allowable stresses, unless otherwise defined, are in accordance with ACI 318-71. For the factored load combinations, design of the containm ent structure meets the broad intent of Article CC-3400 of ASME III Division 2. Details of the design conform to ACI 318-71 and the additional requirements di scussed in Section 3.8.1.4, rather th an the parallel requirements of CC-3000. Major features of the concrete design are si milar using either code.Except for test conditions, the specific limits for service loads given in CC-3430 are not addressed as acceptance criteria for ACI 318-71. However, design of the containment equals or exceeds ACI 318-71 requirements for serviceability. Predicted stresses and strains for structural acceptance tests are well within the limits stated in CC-3430.
MPS3 UFSAR3.8-20Rev. 30The tangential shear stress, Vu, resulting from earthquake loading is resisted by the concrete and by diagonal reinforcing bars. As discussed in Secti on 3.8.1.4.1, the maximum allowable tangential shear stress, Vc, carried by the concrete is assumed to be not greater than 40 psi.
3.8.1.5.2 Steel Liner and Penetration The containment structure liner plate, brackets, attachments, anchors, and access openings are designed to meet the design allowables presented in Table 3.8-1. Initial penetration sizing is performed in accordance with Ta ble 3.8-2. The final design verification is in accordance with Tables 3.8-4 through 3.8-6.To minimize the probabilities of crack propagation as the containment is exposed to the design conditions as indicated in Table 3.8-1, stress levels reached are kept within the limits given by Pellini and Loss in NRL Report 6900, Integration of Metallurgical and Fracture Mechanics Concepts of Transition Temperature Factors Relating to Fracture-Safe Design for Structural Steels.3.8.1.6 Materials, Quality Control, and Special Construction Techniques Section 3.8.1.2 contains applicable code s, standards, and specifications.
The quality assurance activities required by th is section are described in Section 17.1.
3.8.1.6.1 Concrete Materials and Quality Control ACI 301, ACI 347, and ACI 318 form the genera l basis for the concrete specifications.ACI 301 was supplemented, as necessary, with mandatory requirements relating to types and strengths of concrete, including minimum concrete densities, proportioning of ingredients, reinforcing steel requirements, joint trea tments, and testing agency requirements.Admixtures, types of cement, bonding of joints, embedded items, concrete curing, additional test specimens, additional test ing services, cement and reinforcing steel mill test report requirements, and additional concrete test require ments were specified in detail.All Portland cement conformed to requirements of ASTM C150 for Portland cement, Type II. Low alkali Portland cement was used where examinations and tests of aggregates, or of structures containing similar aggregates, indicate potential reactivity with the cement. One exception was that calcium aluminate (high-alumina) cement was used for porous concrete enclosed within the waterproof membrane under the containment mat. This substitution is made to reduce the possibility of clogging of voids by continued hydration that occurs with Portland cement. Certified copies of mill tests, showing that the cement meets or exceeds the ASTM requirements for Portland cement, were furnished by the manufacturer. An independent testing laboratory was MPS3 UFSAR3.8-21Rev. 30retained to perform tests on the cement to substantiate the mill test reports and compliance with the specifications for the cement.An air-entraining admixture was used in the c oncrete. The total air content, expressed as a percentage by volume, conforms to the requirements in Table 3.4.1 of ACI 301, Total Air Content for Various Sizes of Coarse Aggregates for Normal Weight Concrete. This admixture conforms to the requirements of ASTM C 260 when tested in accordance with ASTM C 233. The air-entraining agent was added separately to the batch in solution. The solution was batched by means of a mechanical dispenser capable of accurate measurement and was subject to periodic checks. Air-entrained cement was not used.Other admixtures to control the rate of set, reduce the water content, or improve the workability and cohesiveness of concrete were used in specific instances. Such admixtures were used only after tests were made in combination with the cement and aggregates being used and were specifically approved by the structural engineer. Calcium chloride was not used under any circumstance.Mixing water and ice were clean and free from injurious amounts of oils, acids, alkalies, salts, organic materials, or other substances which may be deleterious to concrete or steel. The mixing water and water used for making ice are periodically checked and tested for suitability by ASTM C 109 Method of Test for Compressive Strength of Hydraulic Cement Mortars (using 2-inch (50-mm) Cube Specimens).Fine and coarse aggregates conform to the requirements of ASTM C 33. Coarse aggregates contained not more than 5 percent soft fragments in accordance with ASTM C 235. Aggregates were evaluated for potential chemical alkali reactivity prior to use by ASTM C 295, Recommended Practice for Petrographic Examination of Aggregates for Concrete, and ASTM C 289, Test for Potential Reactivity of Aggregates (Chemical Method). Also, prior to use of an aggregate, one of the following tests for alkali reactivity (whichever is appropriate for the aggregate) was started: ASTM C 227, Test for Potential Reactivity of Cement Aggregate Combinations (Mortar Bar Method), or ASTM C 586, Test for Potential Alkali Reactivity of Carbonate Rocks for Concrete Aggregates. Test results were eval uated for aggregate suitability with the cement. Aggregates were free from any materials that would be deleteriously reactive in any amount sufficient to cause excessive expansion of mortar or concrete. All aggregates were tested for compliance with the above requirements by an independent testing laboratory.Proportioning of structural concrete conforms to ACI 301, Chapter 3. In general, concrete mixes were of a 28 day strength of 3,000 psi unl ess otherwise specified by the Engineers.
Concrete used for biological shielding has a density of not less than 140 lb/cu ft.Proportioning of ingredients in concrete mixes were determined and tests conducted in accordance with the methods detailed in ACI 301 and ACI 211.1 for combinat ions of materials to be established by trial mixes.
Slump of mass concrete is in accordance with ACI 301, Chapter 3.
MPS3 UFSAR3.8-22Rev. 30Concrete protection for reinforcement, preparation and cleaning of construction joints, concrete mixing, delivering, placing, and curing met or exceeded the requirements of Regulatory Guide 1.55 with the exceptions given in Section 1.8.
Batching and mixing conformed to ACI 301, Chap ter 7. Placing of concre te was by bottom dump buckets, chuting, concrete pump, or by conveyor belt. Aluminum pipe was not permitted for pumping concrete. The rate of placing concrete was controlled so that concrete was effectively placed and compacted with particular attention given around embe dded items and near the forms.Vertical drops greater than 5 feet for any concrete was not permitted, except where suitable means were provided to prevent segregation. Placing equipment and methods were reviewed for compliance with the specifications.The ACI and ASTM specifications were supplemented as necessary with mandatory requirements relating to types and strengths of concrete, minimum concrete density, proportioning of ingredients, reinforcing steel requirements, joint treatments, testing requirements, and quality control.Curing and protection of freshly deposited concrete conform to Chapter 12 of ACI 301, with the following 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 mechanic al 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, exce pt those surfaces specifically requiring water curing. Other curing compounds we re not used on surfaces to which additional concrete is bonded.Construction joints that were to be bonded or transferring shear through shear friction (per ACI-318) were properly prepared as follows.After the initial concrete set had occurred, but before the concrete had reached its final set, the surfaces of these construction joints are thoroughly cleaned to remove all laitance and to expose clean, sound aggregate. After cutting, the surface was washed and rinsed. All excess water which was not absorbed by the concrete was removed.Where, in the opinion of the field engineer, the use of an air-water jet was not advisable, then that surface was roughened by bushhammering, sand blasting, or other satisfactory means to produce the requisite clean surface. Horizontal construction joints were covered by a 0.5 inch thick layer of sand/cement grout and new concrete then placed immediately ag ainst the fresh grout.
MPS3 UFSAR3.8-23Rev. 30Concrete strength tests were performed in accordance with Chapter 16 of ACI 301, (Section 16.3/4.6.1.7). One strength test being made for each 100 cubic yards or fraction thereof for each mix design of concrete placed in any one day.The test specimens for compressive strength were 6 inch diameter by 12 inch long cylinders conforming to ASTM C 31.When required, concrete strength tests were evaluated on a statistical basis by the engineers in accordance with ACI 214, Recomme nded Practice for Evaluation of Compression Test Results of Field Concrete, and Chapter 17 of ACI 301. The strength level of the concrete was considered satisfactory if it conformed to Section 4.3.3 of ACI 318.The field tests for slump of Portland cement concrete followed ASTM C 143, "Method of Test for Slump of Portland Cement." Any batch not meeting specified requirement s was rejected. Slump tests were made frequently during concrete placement and each time concrete test specimens were made.Shop detail drawings for the reactor containment mat, shell, and dome reinforcement were checked by the designer.Special Construction Techniques No special construction techniques were used in the placing of concrete for the containment structures.
3.8.1.6.2 Reinforcing Steel Materials N14 and N18 reinforcing bars are controlled chemistry steel of 50,000 ps i minimum yield point. They conform to Grade 40 of the "Standard Specification for Deformed Billet-Steel Bars for Concrete Reinforcement" ASTM A 615, as modified, to meet the following chemical and physical requirements: Carbon 0.35 percent max Manganese 1.25 percent maxSilicon 0.15 to 0.25 percentPhosphorous 0.05 percent maxSulfur 0.05 percent maxYield strength 50,000 psi minElongation 13 percent min in an 8-in test sampleTensile strength 70,000 psi min MPS3 UFSAR3.8-24Rev. 30 Reinforcing bars smaller than N14 are grad e 40 or grade 60 conf orming to ASTM A 615.Cadweld T-Series reinforcing steel splices are full tension splices manufactured by Erico Products, Inc., Cleveland, Ohio, and are used to splice N14 and N18 reinforcing bars. In restricted areas, reinforcing bars were butt welded in a manner conforming to the requirements of AWS D12.1. Cadweld splices were made in accordance with the instructions for their use issued by the manufacturer, Erico Products, Inc.Reinforcing bars smaller than N14 were ge nerally lap spliced. Where lap splicing was impractical, splicing is accomplished by:1.Use of mechanical (Cadweld) splices as manufactured by Erico Products, Inc., Cleveland, Ohio, or equivalent, using the T-series slee ves that develop the full tensile strength of the reinforcing bars, or2.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 requi rements were rejected.
Mill tests reports showing actual chemical ladle analys is, tensile properties, bend properties, variations in weight, and co nformance 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.
MPS3 UFSAR3.8-25Rev. 30Full size test specimens of all reinforci ng bars were tested on a testing machine using an 8-inch gage length.2.Cadweld Splices Splicing complied with th e 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 pr epared 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 s ubsequent units of 100 splices.
In addition, completed reinforcing steel butt-welded splices were selected on a random basis from the containment struct ure 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 s ubsequent units of 25 splices.
Reinforcing steel bars butt welded to steel plate were te sted 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.
MPS3 UFSAR3.8-26Rev. 30CONSTRUCTION TECHNIQUES1.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 rein forcing steel for the containment structure.Tack welding of designed reinfo rcing steel was not permitted.2.Special No special construction techniques were used in the installation of reinforcing steel for the containment structure.
3.8.1.6.3 Structural Steel Material Steel specifications invoked for structural framing, brackets, and attachments are discussed in Section 3.8.1.2.2.
Quality Control All main members, columns, baseplates, bracing, trusses, girts, and bolts larger than 1 inch in diameter are traceable to a specific heat number. Traceability to a specific heat number for all clip angles, seats, stiffeners, gusset plates, bolts of 1 inch diameter and smaller, and weld filler metal is confirmed in the suppliers' shop or upon receipt at the site. The storage and issuance of these materials for construction is controlled in a manner which assures only those items procured as QA Category I are installed in QA Category I applications.Construction Techniques Structural steel material, erection, and fabrication tolerances are in accordance with the AISC Specification for the Design Fabrication, and Erection of Structural Steel for Buildings.Welding of structural steel is in accordance with AWS D 1.1-72, Revision 1-73.
3.8.1.6.4 Waterproofing MembraneA waterproofing membrane (Figure 3.8-57) was placed below the containm ent structure mat and Engineered Safety Features Building and carried up the containment wall and Engineered Safety Features Building walls to above groundwater level. Attached to and entirely enveloping the part of the containment structure and Engineered Safety Features Building below ground level, the MPS3 UFSAR3.8-27Rev. 30membrane protects the structures from the effects of groundwater and the steel liner from external hydrostatic pressure. When water penetrates or otherwise circum vents the membrane, the water drains to a layer of porous concrete directly below the mats and above the membrane. This layer of porous concrete serves as a horizontal drain under the entire containment structure and Engineered Safety Features Building. The porous layer is drained into the Engineered Safety Features Building. 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).A standpipe assembly has been installed through the waterproofing membrane extending to the floor at elevation (-) 34 feet 9 inches in the Engineered Safety Features Building. This assembly consists of a one inch diameter pipe with a ball valve, a pressure indicator and a globe valve. The purpose of this assembly is to measure the hydrostatic pressure and sample the water below the membrane. Essentially, the standpipe and valve assembly replace a small piece of the membrane.
The standpipe has been securely grouted and se aled into place to preclude membrane leakage.Core samples have been removed from the high alumina cement porous concrete layer from under the Engineered Safety Features Building basemat. The coring proc ess has disturbed the membrane at these locations. The waterproofing membrane has been replaced with grout at these locations.The surface of the containment structure steel liner in contact with concrete is not subject to corrosion because of the alkali ne nature of the concrete.
3.8.1.6.5 Steel Liner and Penetrations MaterialsLiner plate up to 1.25 inches inclusive and bridging plate are made from SA 537 Class 2 Quenched and Tempered, nil-ductility transition temperature (NDTT) test not higher than -10
&deg;F, with the exception of dome line r plate which is made from SA 537, Class 2 normalized to Class 1 practice, NDTT not higher than -10
&deg;F. All liner insert plates and embedment material greater than one inch thick was ultrasonically tested prior to installation for the purpose of detecting possible laminations.Toughness tests (Charpy V-notch) were performed on all materials which form part of the containment structure boundary. Nil-ductility Transition Temperature Tests were also performed on all ferritic steel that formed part of the pressure boundary but were not required of backing plates, test channels, ha tch bolts, and hatch nuts.Penetration sleeves are made of SA537 Grade B Q&T, SA516 Grade 60 fine grain, normalized and SA333 Grade 6 fine grain normalized, all with a NDTT of -10
&deg;F.Neutron shield tank embedment ba se and the carbon steel penetration forgings are SA508 Class 1 with a NDTT of +10
&deg;F.
MPS3 UFSAR3.8-28Rev. 30Penetration coolers, equipment hatch, personnel airlock, shear lugs, and backing plates are SA516 Grades 60 and 70 fine grain normalized with NDTT of -10
&deg;F.Bridging bars are made of SA350 Grade LF1 and SA516 Grade 70 normalized with NDTT of 0
&deg;F. Sump liners and bellows are made of Type 304 stainless steel SA240. The st ainless steel penetration forgings are ma de of types 304 and 316, SA182.Special Construction Techniques Erection of the cylindrical portion of the liner plate followed comp letion of the concrete mat. The liner plates served as the internal form for th e concrete containment dur ing construction. All liner seams are double butt welded, except for the lower 31 feet of the cylindrical shell liner, the liner fire damage repair areas, and the mat, where the plates are welded using backing plates. The liner plate is continuously anchored to the concrete sh ell with steel an chor studs and deformed bars.The maximum difference in cross-sectional diameters of the liner is in accordance with the rules shown in paragraph NB-4221.1 of Section III, ASME Boiler and Pressure Vessel Code, Nuclear Power Plant Components, 1971 Edition. The maximum misalignment between liner plates is in accordance with paragraph NB-4232 of the ASME Boiler and Pressure Vessel Code, Nuclear Power Plant Components, 1971 Edition. All measurements were taken on parent metal and not at welds. Flat spots or shar p angles were not allowed.The allowable deviation from true circular form does not affect the elastic stability of the containment liner because of the restraint provided by the anchor studs and deformed bars tying it to the reinforced concrete shell.
3.8.1.6.6 Backfill Around Containment StructureConcrete was used to backfill around the containment structure. A compressible material was used between the backfill and the containmen t structure wall to provide a rattle space.
3.8.1.7 Testing and Inservice Surveillance Requirements 3.8.1.7.1 Concrete ContainmentA structural acceptance test of the containment was pe rformed after the liner was completed, the last concrete poured, and all pene trations, sleeves, and hatches installed. The test is conducted to confirm that the design and construction of the containment are adequate to withstand the loads caused by the loss-of-coolant accident as described in Chapter 6.2. The test conforms to the requirements of Regulatory Guide 1.18, Structural Acceptance Test for Concrete Primary Reactor Containments, dated December 28, 1972. The measuring points may be varied or relocated in accordance with paragraph C.3 of the guide.The structure is surveyed, measured, and inspected for cracks prior to the test. The containment is subject to an internal pressure equal to 115 percent of the design pressure. The pressure test commences at atmospheric pressure and is raised to 115 percent of design pressure in a minimum MPS3 UFSAR3.8-29Rev. 30of four increments. The containment is depressurized in a minimum of four increments. Measurements are made at each pressurization and depressurization level. The pressure is held constant for at least 1 hour at each level before deflections are recorded. Crack patterns are measured and recorded at atmospheric pressure both before and after the test and at the maximum pressure level.
Radial deflection is measured along 6 meridians at 13 feet 6 inches above the top of mat, at mid-height between mat and springline, and at the springline of the dome. The exact locations of the measurements are indicated in the design specification. Vertical deflections are measured at the springline of the dome and the apex. Radial measurements are made using differential transducers supported from the crane wall. Radial and tangential deflections are recorded around the periphery of the equipment hatch. Verti cal measurement is made using invar tapes.Mapping of cracks is performed on exterior surfaces of the containment at locations selected prior to start of the pressure application. Mapping is on one meridian line at three locations, and one location around the equipment hatch.Testing is not conducted under ex treme weather conditions. The environmental conditions are measured and monitored to permit the evaluation of their contribution to the response of the containment. The testing sequence is repeated if the test pressure drops for unexpected reasons to or below the next lower pressure level, or if significant modification or repairs are made to the containment following the test.The anticipated deflections of the containmen t and equipment hatch are given on Figure 3.8-58. The anticipated deflections are calculated by taking into account the interaction of the liner, reinforcing and concrete including the effects of concrete cracking.A limit of 1.3 times the anticipated deflections were used for comparison to the measured deflections. This is based on a comparison of predicted and allowable stresses in the membrane hoop reinforcing at the 52 psi test pr essure level. The stress in the hoop reinforcing is predicted to be 25,600 psi at the 52 psi test pressure. When increased by 30 percent, the stress is 33,000 psi or 2/3 of yield, which is the allowable rein forcing stress for this loading condition.Cracking was expected in the portion of the containment structure shell away from the mat. (Vertical cracks at approximately 18 inches on center, and horizontal cracks at the construction joints 6 foot on center and at 2 foot to 3 foot intervals between the construction joints.) This was based on observations of previous tests.The final test report, "Report on Structural Acceptance Test of the Concrete Primary Containment Millstone Nuclear Power Station Unit 3," was prepared covering the test performed on July 10-12, 1985. This report contains the information outlined in Regulatory Guide 1.18, Regulatory Position 13. The report was compiled by Stone and Webster and presented to NUSCO in September 1985.
For liner inservice surveillance requirements and testing, including l eak rate testing, see Section 6.2.6.
MPS3 UFSAR3.8-30Rev. 30
 
====3.8.2 STEEL====
CONTAINMENT This section, as outlined in th e NRC Regulatory Guide 1.70, Rev 3, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, regarding steel containment, is not applicable to the Millstone 3 Containment Structure. A steel lined reinforced concrete containment is being used as described in Section 3.8.1. The equipment hatch, personnel hatch, and portions of the penetration sleeves that are not backed by c oncrete, and normally considered to be steel containment, are also described in Section 3.8.1.
 
====3.8.3 CONCRETE====
AND STRUCTURAL STEEL INTERNAL STRUCTURES OF STEEL OR CONCRETE CONTAINMENTS 3.8.3.1 Description of Internal Structures The containment structure interior arrangement (Figures 3.8-59 and 3.8-60) consists of heavily reinforced concrete walls and slabs which are designed to support the principal nuclear steam supply equipment. The interior concrete also provides interior biological shielding and protection from missiles resulting from postula ted component failure (Section 3.5).The reactor vessel is supported on the reactor vessel support/shield tank located within, and laterally supported by, the c oncrete primary shield wall.The refueling cavity, located above the reactor vessel, and the fuel transfer canal are stainless steel lined concrete structures.The pressurizer and each reactor coolant loop, in cluding their steam generator, reactor coolant pumps, and associated valves and piping are en closed within separate concrete cubicles.A 540-ton capacity overhead polar crane is supported by the polar crane wall. The crane is provided with earthquake restraints to preclude its dislodgment from the rails during the safe shutdown earthquake (SSE).Main steam and steam generator feedwater lines, electrical cable trays and conduits, HVAC ducts, and other miscellaneous pipes are supported by structural steel hangers and supports located on the interior structure walls and slabs.A minimum clearance, in accordance with the following, is provided at all levels between the interior structural steel work, platforms, pipe supports, condu it supports, and the containment liner and its attachments.
ElevationMinimum ClearanceAbove EL 57 feet 0 inches1.05 inches.
Between EL 13 feet 0 inches to 57 feet 0 inches (inclusive) 1 inch MPS3 UFSAR3.8-31Rev. 30No new or unique structural features are used in the design of the internal structure of the containment.
NSSS component supports ar e described in Section 3.9.3.
3.8.3.2 Applicable Codes, Standards, and SpecificationsCodes, standards, specificat ions, and NRC regulations and Regulatory Guides used in establishing design methods and material properties for concrete and steel internal structures are given in the following sections:
Section 3.8.3.6 describes materials and quality control procedures used for containment structure interior.3.8.3.3 Loads and Loading CombinationsInterior concrete structures and structural steel within the containment are designed to withstand the pressure buildup resulting from the loss of coolant accident (LOCA) discussed in Section 6.2.1. The blowdown of a postulated rupture of a main coolant pipe is assumed to be in any one of the steam generator cubicles or adjacent to the reactor vessel. Because the volume of each of these cubicles is less than the entire containment structure, initial differential pressures exist between the interior and exterior of the cubicle until full pressurization of the containment is attained. Structural components, walls, floors, a nd beams enclosing these c ubicles are designed to withstand these differential pressures.Pipe rupture may also cause bl owdown jet and reactive forces. The magnitude of a blowdown jet forcing function resulting from a pipe rupture is dependent upon th e geometry and distance of the target from its source. Critical structures and components are protected from, or designed to resist, the effects of these forces. Section 3.6 descri bes the protection provided against the dynamic effects associated with a po stulated rupture of piping.The interior concrete structures protect the containment shell from internal missiles generated by an accident. Safety features are physically protected from potential sources of missiles either by physical separation, barriers, or by providing restraints on the potential missiles. Section 3.5 describes internal missile ge neration and the design of barri ers to resist these hazards.Below EL 13 feet 0 inches0.75 inches Codes, Specifications, Design Methods, and Material Properties Section 3.8.1.2 Section 3.8.1.6 NRC Regulatory Guides 1.10, 1.15, 1.55, 1.60, and 1.61 NRC Regulatory Guide, as qualified in Section 1.8 ElevationMinimum Clearance MPS3 UFSAR3.8-32Rev. 30Containment internal structures (other than the containment structure mat, shell, and dome) are designed for the applicable loads and loading combinations in Table 3.8-3. Loads are applied to each structure as applicable.
3.8.3.4 Design and Analysis ProceduresThe interior structures generally comprise a series of frames, box type structures , and assemblies of slabs. Structural analyses ar e based on elastic behavior using commonly accepted principles of engineering mechanics a ppropriate to the geometry of the structure.Material quality control procedures, as described in Section 3.8.3.6, ensured that material strength requirements were achieved. Over strength of materials is not a factor because the strength method of design, as described in Section 3.8.3.5, is used in proportioning and reinforcing concrete sections.The amount of reinforcing steel required is determined in accordance with the procedures outlined in ACI 318 and the principa l reinforcement patterns are located in the direction of tensile stresses. Bond and anchorage requirements of ACI 318 are complied with, and where biaxial tensile fields exist, the deve lopment lengths required by Section 12.5 of ACI 318 are increased by a minimum of 25 percent.Structural steel is designed in accordance with the procedures outlined in the AISC Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings, the Structural Welding Code, AWS D1.1, and the loading combinations given in Section 3.8.3.3.The heat generated in the primary concrete shield wall due to gamma rays and neutrons emanating from the reactor core and due to gamma rays arising from neutron interactions in the primary shield wall is negligible and has no effect on the strength of the structure.Computer programs that have been used in the design and analysis of structural elements are identified in Appendix 3A.
3.8.3.5 Structural Acceptance CriteriaDesign of interior concrete structures follows ACI 318, using th e required strengt h section based on the strength design method. The basic criterion fo r concrete strength desi gn is expressed as:Required Strength  Calculated StrengthAll members and all sections of members are proportioned to meet this criterion. The required strength is expressed in terms of design loads, or their related internal moments and forces. Design loads are defined as loads which are multiplied by their appropriate load factors. Calculated strength is computed according to the provisions of ACI 318, including the appropriate capacity reduction factors. Capacity reduction factors are the same as those given in Section 9.2 of ACI 318.
MPS3 UFSAR3.8-33Rev. 30Design of interior steel structures is based either on elastic working stress design methods using normal working stress levels given in Part 1 of AISC Specification or on the plastic design methods of Part 2 of AISC.
Section 3.7.2.8 describes the variations incorporated into the seismic analysis structural model to account for a cracked and an uncracked contai nment structure shell, for variations and uncertainties in subgrade shear modulus and spring constants, for virtual mass embedment, and for contact pressure distribution. Design of the internal structures is based upon the most conservative values resulting from these vari ations in assumptions and design parameters.
Section 3.7.3.6 discusses differential seismic movement relating to interconnected components, systems, and equipment.Design for horizontal shear forces in the plane of internal structure walls is in accordance with the requirements of Section 11.6 of ACI 318. Section 11.16 incorporates the combined effects of shear and tensile stresses into the nominal permissible shear stress, Vc, allowed to be carried by the concrete.
3.8.3.6 Materials, Quality Control, and Special Construction TechniquesMaterial and quality controls used for the internal structures are as described in Sections 3.8.1.2and 3.8.1.6. The 60 day compressive strength of concrete is specified as 5,000 psi, except for the concrete ballast slab covering the floor liner, which is 3,000 psi concrete.Structural steel material, erection, and fabrication tolerances are in accordance with the "AISC Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings" (February 12, 1969), Supplement No. 1 (November 1, 1970), Supplement No. 2 (December 8, 1971), and Supplement No. 3 (June 12, 1974). In general, steel used for structural framing conforms to ASTM A 36, "Specification of Structural Steel." In areas where the design indicates that a higher strength steel is required, the steel conforms to the requirements of ASTM A 440, ASTM A 441, or ASTM A 588 as required.Certified copies of mill test reports showing actual chemical and physical properties are furnished for each heat of steel used in maki ng Seismic Category I structural steel.Welding of structural steel is in accordance with AWS D 1.1-72 and AWS D1.1-Rev 1-73 except that undercut on structural welds is limited to 0.01 inch deep when its direction is transverse to the primary tensile stress in the part that is undercut only for the following cases:1.Where cyclic fatigue is a design parameter and/or2.Where thin material is used (i.e., 0.5 inch thick and less).This requirement was implemented through notation of specific inspection criteria on the engineers drawings. In all other situations, underc ut is limited to a maxi mum depth of 1/32 inch.
MPS3 UFSAR3.8-34Rev. 30Connection bolts conform to ASTM A 325 or ASTM A 490. Anchor bolts conform to ASTM A 36 or ASTM A 193.The material, installation, and inspection of high strength bolts conform to the requirements of the Research Council on Riveted and Bolted Structural Joint Specification using ASTM A 325 or A 490 bolts.Radiation damage to concrete is insignificant for conditions of neutron fluence at least to 2 x 10 19 nvt and temperatures at least to 120 C (Clark 1958). Th e neutron fluence and the temperature levels in the primary concrete shield wall are both lower than these levels; therefore, no structural damage due to radiation or temperature effects occurs.
3.8.3.7 Testing and Inservice Surveillance RequirementsNo testing or inservice surveillance of the reactor containment interior structure is planned.
 
====3.8.4 OTHER====
SEISMIC CATEGORY I STRUCT URES (AND MAJOR NONSAFETY RELATED STRUCTURES) 3.8.4.1 Description of the StructuresThe Seismic Category I structures are indicated on the plot plan (Figur e 1.2-2). The arrangement drawings for major Seismic Category I structur es other than the reactor containment are:Title Figure Containment Enclosure Building3.8-61 (EA-42)Auxiliary Building3.8-62 (EM-6)Fuel Building3.8-63 (EM-7)
Control Building3.8-64 (EE-27E)Cable Tunnel3.8-65 (EC-7A)Emergency Generator Enclosure and Diesel Fuel Oil Tank Vault 3.8-66 (EM-13)Engineered Safety Features Building3.8-67 (EM-2)Main Steam Valve Building3.8-68 (EM-2)
Circulating and Service Water Pumphouse3.8-69 (EM-8)Hydrogen Recombiner Building3.8-70 (EM-2)Circulating Water Discharge Tunnel and 3.8-71 (EC-16C) Discharge Structure(EC-17A)Railroad Canopy(EC-38W,S,Y,Z)
MPS3 UFSAR3.8-35Rev. 30 The refueling water storage tank and the demineralized water storage tank are shown on the plot plan (Figure 1.2-2).
Arrangement drawings for major Non-Category I structures are:
Plant grade is approximately elevation 24 feet 0 inch.
Containment Enclosure Building (Seismic Category I)
The containment enclosure building, a cylindrical steel framed structure with uninsulated metal siding and builtup roofing over insulated metal roof deck, envelops the containment structure above grade. It has a diameter of 156 feet, a height above grade of approximately 166 feet, and is supported on the containment structure.The containment structure below grade is surrounded by ribbed fiberglass sheets, which extend from the top of the containment mat to above grade. This arrangem ent provides vertical ventilation channels which vent directly into the containment enclosure building or into the abutting buildings. The corrugated fiberglass sheet is covered by a 2 inch thick layer of compressible material, a 40 mil th ick waterproofing sheet membrane, and then one more layer of 2 inch thick compressible material. The waterproofing membrane encloses the containment substructure and extends beneath the mat and above plant grade. The en closure building design incorporates horizontal and vertical sliding joints to ensure that the integrity of the containment enclosure building is maintained during maximum possible pressure transients of the containment structure under DBA conditions. It assures the proper performance of the supplementary leak collection and release system, as described in Section 6.5.1.To provide the required degree of air tightness and to maintain a partial vacuum within the containment enclosure building, the metal siding, metal deck side joints, and end laps have two Millstone Stack Elevations and Sections3.8-83 Millstone Stack - Mathematical Model3.8-84 TitleFigureService Building3.8-72 (EM-5)
Turbine Building3.8-73 (EM-3)Waste Disposal Building3.8-74 (EM-9)Warehouse 5 and Mills tone 2 Condensate Polishing Facility 3.8-75 (EM-50)
Auxiliary Boiler and Condensate Polishing Building 3.8-76 (EM-38)Miscellaneous Yard Tankage1.2-2  Plot PlanTitle Figure MPS3 UFSAR3.8-36Rev. 30continuous lines of caulking at all joints. Neoprene gaskets and sheets are used to provide a flexible seal between the containment enclosure and the other buildings. The base of the containment enclosure building wall is sealed in a similar manner to th e subgrade waterproofing membrane.The structural framing, but not the siding or roofing, is designed to remain intact under tornado loading. In addition, the design of the building is based on the structural loads being transferred through the structural steel bracing and into the containment st ructure. The buildi ng is considered to be round and moderately smooth.
Auxiliary Building (Seismic Category I)
The auxiliary building is located west of the fuel building, east of the service building, and north of the containment structure. The auxiliary building, approximately 102 feet by 177 feet in plan, is supported on a reinforced concrete mat founded on rock. Rock dowels are installed along the exterior walls of the structure to resist uplifting during seis mic loading. Section 2.5.4 describes these rock dowels. The basement floor is approximately 20 feet below grade. There is one floor at grade and two above. The concrete roof is located approximately 69 feet above grade. The substructure and superstructure are reinforced concrete construction. Interior shield walls are reinforced concrete.The motor control center (MCC) and rod control area is located in the southern end of the auxiliary building. This area is open on the side adjacent to the c ontainment structure for direct access to electrical penetrations.Fuel Building (Partially Seismic Category I)The fuel building is located east of the auxiliary building, south of the waste disposal building, and north of the containment and engineered safety features structures. The building, approximately 112 feet by 92.5 feet, is s upported on compacted fill and/or rock.
The Seismic Category I portion of the building is as follows:1.Spent Fuel Pool2.Spent Fuel Shipping Cask Area3.Auxiliary-Emergency Safety Features Building Pipe Tunnel4.Canopy over Fuel Cask Shipping Area5.Demineralizer Area6.Spent Fuel Shipping Cask Washdown Area7.Spent Fuel Pool Cooler Cubicle MPS3 UFSAR3.8-37Rev. 30 The remainder of the buildi ng is non-seismic and includes:1.New Fuel Handling Area2.New Fuel Storage Vault Area3.Equipment Decontamination Area4.Equipment Storage Area5.Heat Tracing RoomThese non-seismic areas are designed to withstand seismic loading so as to prevent their collapse onto adjacent Category I areas. The fuel building has a ground floor at grade elevation and a basement 13 feet below grade. The spent fuel pool portion of the building is reinforced concrete construction from approximately 24 feet below to 28 feet above grade. The spent fuel areas are protected from tornado missiles by a reinforced concrete superstructure. The new fuel handling and equipment decontamination areas of the building have reinforced concrete walls with a steel framed roof and metal deck covered wi th 4-ply asphalt and gravel roofing.The spent fuel pool is "L" shaped, with the bottom of the pool approximately 13 feet below grade. The floor is 8 foot thick reinforced concrete. The spent fuel shipping cask storage area is 14 feet wide by 30 feet long. The bottom of this area is stepped from approximately 20 feet below grade to 1 foot above grade to 28 feet above grade to limit the lifting height of the spent fuel shipping cask. The walls of the spent fuel pool and the spent fuel shipping cask storage area are 6 foot thick reinforced concrete with the exception of the east wall of the spent fuel shipping cask storage area which is 5 feet thick and extends to approximately 28 feet above grade. Concrete dividing walls permit dewatering the spent fuel shipping cask storage area and the fuel transfer canal without dewatering the entire pool. The interior walls and floor of both the new fuel storage vault and spent fuel pool, the spent fuel shipping cask storage area, and the fuel transfer canal are lined with 0.25 inch thick stainless steel plate.The new fuel storage vault is 24 feet 0 inch long by 15 feet 9 inches wide by 18 feet 4 inches deep with the bottom approximately 9 feet 6 inches above grade.
Control Building (Seismic Category I)The control building, approximately 101 feet by 116 feet, is located north of the turbine building, south of the emergency diesel generator building, east of the technical support center and office building, and west of the service building; it is supported by a reinforced concrete mat foundation. Except for the east portion of the top level described below, all exterior walls, floor slabs, and roof slabs are reinforced concrete with interior framing and columns of structural steel.The basement level approximately 20 feet below grade houses emergency (essential) switchgear, battery chargers, battery rooms, inverters, and the emergency shutdown panel. The second floor is the cable spreading room. The third floor is the control room level which contains the control MPS3 UFSAR3.8-38Rev. 30room, instrument rack room, computer room, and facilities for personnel. The fourth floor contains the ventilation and air-conditioning equipment rooms. The roof is approximately 60 feet above grade.The east portion of the top level is nonsafety related; however, the steel superstructure is designed to withstand tornado and seismic loads.Cable Tunnel (Auxiliary Build ing to Control Building) (Seismic Category I)
The electrical cable tunnel measur es approximately 24 feet by 29 fe et in section and runs through the basement level of the service building for its full width. The tunnel is contiguous with and provides support for portions of the service building. The tunnel is reinforced concrete construction (Figure 3.8-65).Emergency Generator Enclosure and Fuel Oil Tank Vault (Seismic Category I)The emergency generator enclosure, located north of the control bui lding, is a reinforced concrete enclosure approximately 65 feet by 72 feet. The roof is located approximately 41 ft above plant grade. Its foundation consists of reinforced concrete footings placed upon glacial till. The structure is founded approximately 15 feet below grade. Each combination of emergency generator and diesel engine ha s its own foundation founde d on compacted fill. The roof and walls are reinforced concrete, a minimum of 2 feet thick. A reinforced concrete wall separates the two emergency generator units. Removable sections of the east and west walls in front of the diesel units provide for repl acement of equipment.The fuel oil tank vault, approximately 32 feet by 65 feet by 23 feet high, is located east of the main structure. The vault is below grade to provide protection from tornado missiles. A 1 foot-6 inch thick fire wall separates the two tanks.
Engineered Safety Features Buildin g (Seismic Category I)The engineered safety features building wraps around the east side of the containment structure. The structure is founded on rock at elevation 0 feet 6 inch and is approximately 140 feet long by 40 feet wide. Four pump shafts extend down to the containment mat for the containment recirculation pumps. The structure has three floor levels and extends 32 feet above grade. The entire building, including the pump shafts, is reinforced concrete construction.Main Steam Valve Building (Seismic Category I)
The main steam valve building, loca ted west of and dir ectly adjacent to the containment structure protects the main steam valves and piping from tornado missiles. The building consists of a reinforced concrete structure with a 2 foot thick wall and roof supported on the rock. The structure extends from approximately 16 feet below gr ade to approximately 63 feet above grade.
MPS3 UFSAR3.8-39Rev. 30Circulating and Service Water Pumphouse (Service Water Portion, Seismic Category I)
The circulating and service water pumphouse is located on the shoreline of Niantic Bay west of Millstone 3. Approximately 128 feet by 86 feet, it houses the circulating water pumps and the service water pumps. It is constructed of reinforced concrete founded on bedrock approximately 30 feet below mean low water. The service water pump room and its supporting elements are Seismic Category I, i.e., designed to withstand tornado and earthquake loads, and are protected against flooding to 25.5 feet above mean low water.
A retaining wall is located on the west side of the circulating and service water pumphouse and is a Category I counterfort type reinforced, concrete retaining wall. The wall is founded on bedrock and is an extension of the west wall of the circulating and se rvice water pumphouse. The function of the west retaining wall is to protect the Category I service water and electrical lines located behind the wall and to be part of the shoreline protection. For a discussion of the shoreline protection, refer to FSAR Section 2.5.5. The seawall located to the east of the circulating and service water pumphouse is not a Category I structure and therefore, is not discussed in this section.Hydrogen Recombiner Building (Seismic Category I)The hydrogen recombiner building is located adjacent to the containment structure, on the southeast side, directly below the equipment hatch. The building, approximately 56 feet by 50 feet by 27 feet high, houses the hydrogen recombiner equipment and provides access to the equipment hatch and support for the removable hatch missile shield. It is constructed of reinforced concrete, for protection of safety related equi pment, and founded on fill concrete.Circulating Water Discharge Tunnel (Seismic Category I)The circulating water discharge tunnel is a reinforced concrete structure founded entirely below grade on rock, concrete fill, or till. The portion of the circulating water discharge tunnel downstream of the service water discharge point is designed as a Seismi c Category I structure.
Railroad Canopy (Seismic Category I)
The railroad canopy is located to the east of the fuel building.The canopy structure is approximately 75 feet long by 26 feet wide with buttresses extending out an additional 20 feet. It has a mat which is about 77 feet long by 54 feet wide and 8 feet thick and is founded on concrete fill. The 50.6 and 51.2 line walls and roof are reinforced concrete 2 feet thick. The buttresses, L-line wall and west wall are all reinforced concrete, 3 feet thick.
The top of the mat is at grade. A railroad spur enters the east side of the building at ground level.
The entire canopy structure is desi gned for seismic and tornado loads.
The building protects the spent fuel pool from tornado missiles.
MPS3 UFSAR3.8-40Rev. 30Millstone StackDescription The unlined, free standing, tapere d, reinforced concrete stack has the following dimensions: Overall height above foundation 385'6"Height above adjacent grade 375'Inside diameter at top 7'Outside diameter at base 27'6"Thickness at top 7"The stack configuration is shown on Figure 3.1-7. The Millstone stack was originally designed as seismic Class I.Maximum stack stresses occur in the first 208 feet above grade, due to load combinations which include maximum wind.The stack is designed for a maximum exhaust air temperature of 150
&deg;F and a minimum ambient temperature of 0
&deg;F. Design and construction of the stack is in accordance with applicable requirements of ACI "Standard Specifications for the Design and Construction of Reinforced Concrete Chimneys" (ACI-505) as follows:Wind Pressures For normal design conditions, it was designed at specific ACI allowable unit stresses. Wind pressures on projected areas for various height zones above ground conformed to the following requirements for a basic wind pressure area of 30 psf. Tabulated wind pressures include the reduction for the circular shape of the stack. The tabular value for pressure of ACI-505 was increased for a gust velocity of 140 mph in accordance with Paragraph 400 of ACI-505.Height PressureDesign Wind Pressure0-49 ft32 psf50-99 ft39 psf100-199 ft45 psf 200-299 ft48 psf300-375 ft51 psf MPS3 UFSAR3.8-41Rev. 30The stack has a 2.94 factor of safety with regard to overturning or rocking. The maximum overturning moment due to wind on the stack is 52,500 kip feet and the stability moment for the stack is 154,000 kip feet.
Earthquake LoadingIn developing a mathematical model, for dynamic analysis, the stack was treated as a flexible cantilever system with the base fixed at the top of the foundation. (The stack base is neither rigid nor can it rotate, but is consider ed rigid as this assumption results in a shorter period for the stack. This analytical representation gives higher, more conservative response acceleration.) Forty mass points were considered to be supported my weightless elastic columns. The model is depicted by Figure 3.1-8. Subsequent to the formation of mass and stiffness matrices for the cantilever system, the periods and mode shapes were calculated, displacement and inertia force time histories were established and a time history of shears, moments, displacement s and accelerations determined.The top of the stack is at elevation 389 feet or 238 feet above the top of the Reactor Building. The stack is located 416 feet east of the Reactor Building east wall. Thus, the top of the stack could not strike the operating floor of the Reactor Building, even if toppled intact about its base, because the stack height is 385 feet from the base. The ventilation exhaust is brought through breeching which gathers the various exhaust ducts together. The stack is provided with a one foot by four foot access opening at the base, three galvanized steel balconies, an outside ladder for the full stack height, aviation obstruction light ing, lightning protection and handling of the isokinetic sampler.
Service Building (Partially Seismic Category I)The service building is located between the control and the auxi liary buildings. Approximately 100 feet by 80 feet, it is founded on bedrock and is of st eel frame construction with metal siding and builtup roofing on insulated metal deck. This building consists of one level below grade, one level at grade, and two levels above grade. The roof is located approximately 43 feet above plant grade. The below-grade level houses nonsafety related switchgear and the Seismic Category I electrical tunnel. The grade level houses offices, change rooms, radiati on protection facilities, first-aid room, and other service facilities. The first floor above grade houses the lunch room, instrument repair room, and locker area, while the second above-grade level houses mechanical equipment.
The structural framing, but not the siding and roofing, is designed to withstand tornado winds and seismic forces.Turbine Building (Nonsafety Related)The turbine building, approximately 325 feet by 115 feet, is located west of the containment structure and is supported on spread footings on basal till and compacted select granular fill. The turbine building has a basement level 10 feet below grade and a roof 107 feet above grade. The foundation walls are reinforced concrete to grade. The superstructure is steel framed with metal siding and builtup roofing on an insulated metal deck. There is an auxiliary bay of the same MPS3 UFSAR3.8-42Rev. 30construction, approximately 50 feet by 300 feet by 75 feet high above grade, on the east side of the turbine building.
The structural framing, but not the siding and roofing, is designed to withstand tornado winds and seismic forces.Waste Disposal Building (Nonsafety Related)The liquid waste disposal building is located directly north of the fuel building and east of the auxiliary building. It is a reinforced concrete structure approximately 48 feet by 114 feet with a steel framed HVAC penthouse. The building is founded on bedrock a nd basil till with a basement level 20 feet below grade. The penthouse r oof is approximately 73 feet above grade.The solid waste disposal building is located directly north of the liquid waste building. The building is approximately 38 feet by 114 feet and is founded at grade on soil backfill. The superstructure consists of a 24 foot high reinforced concrete shell and a steel framed enclosure with builtup roofing on insulated metal de ck approximately 42 feet above grade.The design equals or exceeds the requirements of Regulatory Guide 1.143.Warehouse Number 5 and Millstone 2 Condensate Polishing Facility (Nonsafety Related)The warehouse structure, approximately 98 feet by 211 feet, is located north of the Millstone 2 turbine building and south of the Millstone 3 c ondensate polishing facility and auxiliary boiler building. The structure consists of three main levels and a penthouse. The condensate polishing facility is located approximately 20 feet below grade with portions of the waste handling equipment extending to 4 feet and 26 feet above grade. Warehouse storage and Millstone 3 ultrafiltration equipment are located 4 feet above grade. Records storage is located 26 feet above grade. The penthouse, 50 feet by 46 feet, is located 40 feet above grade in the middle of the structure near the west wall. This enclosure houses the elevator machine room and building ventilation equipment.The superstructure is steel framed with the main roof approximately 40 feet above grade and the penthouse roof approximate ly 58 feet above grade.The warehouse structural framing, but not the siding and roofing, is designed to withstand tornado winds and seismic loadings.
Auxiliary Boiler and Conde nsate Polishing Facility (Nonsafety Related)The auxiliary boiler room is located south of the turbine building, east of the condensate polishing facility, and north of the warehouse. The area is approximately 66 feet by 58 feet in plan and is supported on reinforced concrete footings with a floor 0 feet 6 inches above grade. A roof with insulation and 4-ply asphalt and gravel is pr ovided at approximately 36 feet above grade.
MPS3 UFSAR3.8-43Rev. 30The condensate polishing facility is located north of Warehouse Number 5 and west of the auxiliary boiler room. The condensate polishing enclosure is a reinforced concrete two story structure, approximately 58 feet by 64 feet, desi gned for radiation protec tion. The structure is supported by spread footings. The basement floor is approximately 10 feet below grade with the second level approximately 14 feet above grade.1.Yard StructuresVacuum Priming Pumphouse (Nonsafety Related)
The vacuum priming pumphouse is a reinfo rced concrete structure located on top of the outfall structure. The area is appr oximately 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 wa ter tanks, demineralized water storage tank, refueling water storage tank, boron a nd waste test tanks, condensate storage tank, condensate surge tank and water tr eatment 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 concre te structures constructed below grade with access through manhole covers at grade.All other nonsafety related structures are located such that their failure does not damage safety related systems, structures, or components.
3.8.4.2 Applicable Codes, Standards, and SpecificationsCodes, standards, specifications, and NRC regulatory guides used in es tablishing design methods and material properties for Seismic Category I concrete and steel structures other than the containment are given in Section 3.8.1.2.
3.8.4.3 Loads and Loading CombinationsAll Seismic Category I structures other than the containment structure mat, shell, and dome are designed for the loads and load combinations in Ta ble 3.8-3. Section 3.8.4.5 describes allowable stress levels.
MPS3 UFSAR3.8-44Rev. 30For the spent fuel pool, the effects of loads imparted to the structure by the spent fuel racks as well as the effects of hydrostatic and seismically induced hydrodynamic loads are considered in the design. The historical design of the spent fuel pool walls and mat considered the thermal effects based on the temperatures indicated in  Figures 3.8-79 and 3.8-80. The analysis for classifying a full core off load as a normal evolution evaluated the thermal effects based on temperatures indicated in Figure 3.8-82. The spent fuel pool walls and mat were also investigated for the revised thermal transient effects due to the storage of higher enrichment fuels as shown in Figure 3.8-81. Utilizing the loads and load combinations in Table 3.8-3, the allowable stress levels described in Section 3.8.4.5 were satisfied.
3.8.4.4 Design and Analysis ProceduresIn general, design and analysis procedures conform to the requirements of the ACI 318-71 Code and the AISC specification, 7th Edition, for the Design, Fabrication and Erection of Structural Steel for Buildings except as noted in Section 3.8.4.3. Structural analyses are based on elastic behavior using commonly accepted principles of engineering mechanics appropriate to the geometry of the structure.The boundary conditions assumed for structural elements under design are based on the stiffness of the elements into which these elements are framed.Material quality control procedures, as referred to in Section 3.8.4.6, ensure that minimum strength material requi rements are achieved.Seismic accelerations on the structures are determined by dynamic analysis as described in Section 3.7.3.4. Forces are determined and then applied statically in the design of the structures.
The analytical techniques used to determ ine the forces are given in Section 3.7.2.A shake space, consistent with building displacements, is provided, above grade, between all independent Seismic Category I and nonseismic structures so as to prevent their interaction during a seismic event.Tornado loads (described in Section 3.3.2) include wind force loads and the loads from tornado generated missiles. Section 3.5.3 gives tornado missile impact effects.Computer programs which have been used in the design and analysis of structural elements are identified in Appendix 3A.
3.8.4.5 Structural Acceptance CriteriaSeismic Category I structures, as identified in Table 3.2-1, are designed to withstand the loading combinations given in Table 3.8-3.
Concrete structures are designed by the strength design method of ACI 318-71.The basic criterion for stre ngth design is expressed as:
MPS3 UFSAR3.8-45Rev. 30Required Strength  Calculated StrengthAll concrete members and all sections of concrete members are proportioned to meet this criterion. The required strength is expressed in terms of design loads, or their related internal moments and forces. Design loads are defined as loads that have been multiplied by their appropriate load factors. Calculated strength is that computed by the provisions of ACI 318, including the appropriate capacity reduction factors. Capacity reduction factors are taken as given in Section 9.2 of ACI 318. Allowable stresses and strains for concrete structures are within the limits specified in Section 9.2 of ACI 318.Steel structures, except as noted in this section, are designed in the elastic range to maintain actual stresses less than allowable stress given in Part 1 of the AISC "Specification for the Design, Fabrication and Erection of Structural Steel for Buildings" and also the Structural Welding Code, AWS D1.1. For members requiring tornado or seismic (SSE) design, allowable stresses are as follows:Tension F t = 0.90F y Shear F v = 0.60F y Compression Members Allowable stresses are 1.6 times the values given by Section 1.5.1.3 (p 5-64) of AISC Specification, 7th Edition, for the Design, Fabrication, and Erection of Structural Steel for Buildings.
Bending Tension and compression for compact, adequately braced members symmetrical about, and loaded in, the plane of their minor axis.
F b = 0.90F y Allowable stresses for other me mbers are 1.6 times those values given in AISC Section 1.5.1.4, but in no case greater than 0.90 F y.Connections Welds - 1.6 times values given in AISC Section 1.5.3, Table 1.5.3High Strength Bolts - 1.6 times values given in AISC Section 1.5.2, Table 1.5.2.1 MPS3 UFSAR3.8-46Rev. 30The limits of deflections are within those specified in Section 9.5 of ACI 318 for reinforced concrete construction. Stress levels, rather than de flections, are normally the criteria for structural steel design.The majority of the Millstone 3 st ructures have a relatively small height-to-width ratio; therefore, column drift due to wind is not a problem. The only building with significant height, the turbine building, has been checked for wind deflections. These deflections are limited such that overall frame stability is maintained and deflections under service loads are within common practice for the industry.
3.8.4.6 Materials, Operating Control, and Special Construction Techniques Sections 3.8.1.2 and 3.8.1.6 describe ma terial and quality control.
There are no special techniques used in constructing the structures (Section 3.8.4.1).The 60 day compressive strength of concrete for the spent fuel pool and the fuel building is specified as 5,000 psi.
3.8.4.7 Testing and Inservice Surveillance RequirementsThere are no special testing or inservice surveillance requirements for Category I structures outside the containment.
3.8.4.8 Masonry Walls Masonry walls in safety related areas in the plant comply with the requirements in Appendix A to SRP Section 3.8.4. Locations of walls ar e given in Figure 3.8-64, Sheet 1 and 3.
 
====3.8.5 FOUNDATIONS====
3.8.5.1 Description of the FoundationsFoundations for all of the major structures consist of soil or rock supported reinforced concrete mats or spread footings as described in Table 2.5.4-14. Figures 2.5.4-1 through 2.5.4-17 show plan and section views of the major foundations.To provide for independent movement of structures during a seismic event, a minimum of 1 inch thick compressible material is provided below grade between all structures. The containment is separated, below grade between all structures, by a minimum 4 inch shake space filled with a compressible material.Horizontal shear keys are provided for the control, fuel, and auxiliary buildings and for the circulating and service water pumphouse foundations. Figure 3.8-77 shows the arrangement of these horizontal shear keys and Figure 3.8-78 shows a typical detail. Rock dowels are used in the auxiliary building foundation (Section 3.8.5.5).
MPS3 UFSAR3.8-47Rev. 30Horizontal shear resulting from seismic acceleration of the containment structure is transferred through interface friction to the surrounding rock.Rock dowels are used in the exterior walls of the auxiliary building to resist uplif t during seismic loading. Section 2.5.4 describes the rock dowels and the installation program.
The containment enclosure building is supported en tirely on the containment structure and has no foundations. Figure 3.8-61 shows typical details of the interface between the enclosure and ground/adjacent structures.
Significant amounts of groundwater are not expe cted. Figure 2.5.4-37 shows the design levels for groundwater. No Seismic Category I dewatering system is required. However, the following features have been incorporated to prevent seepage of groundwater into portions of structures below the piezometric surface:1.All structures, except the containment, ha ve waterstops instal led at construction joints below grade.2.The containment substructure is encased with a waterproof membrane to elevation 25 feet 0 inches or to th e 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 c ontinued 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 connect s to a sump located in the Engineered Safety Features Building to remove groundwater which leaks through the membrane. A non safety-related pump is us ed 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, a nd Engineered Safety Features Building substructures are encased with a waterp roof 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.
3.8.5.2 Applicable Codes, Standards, and SpecificationsSection 3.8.3.2 contains the codes, standards, specifications, and NRC regulatory guides used in establishing design methods and material properties for foundatio ns and concrete supports.
MPS3 UFSAR3.8-48Rev. 30 3.8.5.3 Loads and Loading CombinationsFoundation design is based upon appropriate loading combinat ions. The loads and loading combinations given in Section 3.8.1.3 are used for the containm ent foundation design. The loads and loading combinations given in Table 3.8-3 are used for the design of all other Seismic Category I foundations.In addition to the above loads and load combinations, the following were used to check against sliding and overturning due to earthquakes, winds, tornadoes, and the design basis flood:1.D + H + OBE2.D + H + W3.D + H + SSE4.D + H + W t 5.D + F where:D, OBE, W, SSE, W t are as defined in Table 3.8-3H = The lateral earth pressure F = The buoyant and lateral force effects of the design basis flood Section 3.8.5.5 gives stability fa ctors for these conditions.
Where Seismic Category I structures extend below the surface of the finished ground grade, their external walls are designed for seismic lateral earth pressure and groundwater effects in addition to static lateral earth pressures due to soil loads, surcharge loads applied at ground surface, and lateral and buoyant force effects from groundwater or flood and loads of adjacent footings or mats.3.8.5.4 Design and Analysis Procedures Section 3.8.1.4 describes the desi gn and analysis of the reac tor containment foundation mat.
Section 3.8.4.4 describes design and analysis of Seismic Category I foundations other than the reactor containment.
Section 3.7.2.14 describes determinati on of seismic overturning moments.
MPS3 UFSAR3.8-49Rev. 30 3.8.5.5 Structural Acceptance CriteriaStructural design of the reactor containment foundation is in accordance with ACI 318, using the load criteria given in Section 3.8.1.3. Structural design of all Seismic Category I foundations other than the reactor containment foundation is in accordance with ACI 318, usi ng strength design and the load criteria given in Section 3.8.4.3.The basic criterion for stre ngth design is expressed as:Required Strength  Calculated StrengthAll members and all sections of members are proportioned to meet this criterion. The required strength is expressed in terms of design loads or their related internal moments a nd forces. Design loads are defined as loads that are multiplied by their appropriate load factors. The calculated strength for all Seismic Category I foundations is that computed by the provisions of ACI 318, including the appropriate capacity reduction factors. Capacity re duction factors are as given in Section 9.2 of ACI 318.
Sliding and overturning fact ors of stability are: No differential settlement of the reactor containment is anticipated.
3.8.5.6 Materials, Quality Control, and Special Construction TechniquesPorous concrete is used to provide subsurface drainage under and around the containment structure. This type of concrete is formed by the omission of the fine aggregate from a standard concrete mix. The mix is designed to have a minimum 28 day compressive strength of 1,000 psi.Porosity tests of porous concrete performed at Northeastern University in Sept ember 1962, used 6  inch diameter by 12 inch long cylinders. These cylinders were prepared in the laboratory by compacting the material in three layers with a standard tamping rod. Results indicated water porosities of from 28 to 47 gpm
/sq ft, depending upon the amount of compaction and resulting density of the cylinders.
LOADING CONDITION (SECTION 3.8.5.3)MINIMUM FACTORS OF SAFETYOVERTURNING SLIDINGFLOTATION1. Operating Basis Earthquake1.51.5-2. Normal Wind1.51.5-3. Safe Shutdown Earthquake1.11.1-4. Tornado1.11.1-
: 5. Design Basis Flood--1.1 MPS3 UFSAR3.8-50Rev. 30The waterproofing membrane is a butyl rubber membrane. Adhesives and tapes used for joints and seals in the membrane are the membrane manufacturer's recommended material for the applicable conditions.
Section 3.8.4.6 describes the materials and quality control used for Seismic Category I foundations and supports. No sp ecial construction techniques are used for Seismic Category I foundations.
3.8.5.7 Testing and Inservice Surveillance RequirementsThe entire reactor containment undergoes structural acceptance testing, as described in Section 3.8.1.7. Except for this test, no other testing or inservice surveillance of foundation systems is planned.There are no special testing or inservice surveillance requirements for Seismic Category I structures outside the containment.
3.
 
==8.6 REFERENCES==
FOR SECTION 3.83.8-1Clark, R.G. 1958. Radiation Damage to Concrete, US AEC Report H.W.-56195, Hanford Atomic Products Operation, March 31, 1945.3.8-2Holley, M.J., Jr. 1969. Provision of Required Seismic Resistance. MIT, Cambridge, Mass.3.8-3TRW, Inc. 1975. Embedment Properties Headed Studs. TR W Nelson Division, Lorain, Ohio.3.8-4Zhemochkin, B.N. and Sinitzin, A.P. 1962.
Practical Methods for Analysis of Beams and Plates on Elastic Foundations. In Russian, Gosstroiizdar, Moscow.
MPS3 UFSAR3.8-51Rev. 30NOTE: The normal and test load combinations are producing negligible effects.TABLE 3.8-1 LOADING CONDITIONS - LI NER PLATE AND ACCESS OPENINGSCategoryLoad Conditions Design Allowables (per ASME III Nomenclature)EmergencyD+P D+T D+SSEP m+P b+Q < 3S mTestD+1.15PP m < 0.9S y P m+P b < 1.35S y+ "CAT" curve considerationsNormal 100 cycles of PNB-3222.4 (d) or (e) 400 cycles of  T 100 cycles of 1/2-SSE Severe Operational D+Pmin +T min + 1/2-SSEP m < S m Without temperature P m+P b < 1.5S m P m+P b+Q < 3S m EmergencyD+Pmin+T D + SSESC = 0.014 ANCHORSEmergencyD+P D+T D+SSEMaximum shear < 0.425 S u Severe Operational D+Pmin+T min+ 1/2-SSEtensile
< 0.45 S u EmergencyD+Pmin+T D+SSEShear a = 0.5 u Tension  a = 0.5 u Where: D =Dead load effect of reinforced concrete structure acting on the liner plus dead load of the liner P D =Design pressure (pressure resulting from design basis accident and safety margin)
MPS3 UFSAR3.8-52Rev. 30 T D =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 earthquakeP=Differential pressure between operati ng pressure and atmo spheric pressure (100 cycles are assumed on the basis of 2.5 hour 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 s easonal 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 duri ng operation of the containment Tmin=Load due to thermal expansion resulting when the liner is exposed to the minimum pressure S y =Yield strength of the material S m=The smaller of 1/3-ultimate strength or 2/3-yield strength S u=Ultimate strength of the stud material a=Allowable displacement for liner anchors, inches u=Ultimate displacement capaci ty for liner anchors, inchesSC=Allowable liner plate compressive strain, inches/inchesWhere:
MPS3 UFSAR3.8-53Rev. 30
* For the pipe portion, refer to Section 3.7B.3.1 TABLE 3.8-2 LOADING CONDITIONS, PENETRATIONSAreas of Analysis (See Figure 3.8-15)CategoryLoad CombinationsStress Allowables per ASME III Nomenclature1DesignM p or T p or J ax or J sh P m < 0.9 S y P m + P b< 0.9 S y P concrete bearing < 2,400 psiEmergencyP d + T d + R o P m + P b+ Q < 3 S m 2 Design (1)M p or T p or J ax or J sh P m < 0.9 S y *Pm + P b < 0.9 S y *Design (2)P g + T g + Design (1)(P 1 + P b) <1.5 S m (P m+ P b + Q) < 3 S m Normal P g + T g + R e ASME III Table NC-3611.1(b) (3)-1 or Para. NB-3222.4(d) or (e)Where: M p = Yielding moment = Required bending moment to produce stresses equal to the yield strength of the pipe material T p = Yielding torque = Required torsional moment to produce stresses equal to the yield strength of the material J ax = 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 J sh = Shear jet force = Load equal to the piping design pressure times the inside area of the pipe acting transversely to the pipe P d =Containment design pressure T d =Containment design temperature P g =Piping design pressure T g =Piping design temperature R o =Piping reactions due to normal operation (including SSE effects)
R e =Piping reactions due to normal operation (including 1/2 - SSE)Design (1) - Applies to the sizing of th e sleeve and attachment plateDesign (2) - Applies to the evaluation of stresses in the area of Analysis 2 due to the given load combinations MPS3 UFSAR3.8-54Rev. 30TABLE 3.8-3 LOADS AND LOADING COMBINATIONS1.Concrete Structures (Containment Internal Structures and Category I Structures, other than the containment mat, shell, and dome).
Loads and loading combinations are based on ACI 318, and AEC Enclosure 3 - Structural Design Criteria for Evaluating Effects of High Energy Pipe Breaks on Category I Structures Outside the Containment, Structural Engi neering Branch, Directorate of Licensing.1.U = 1.4D + 1.7L2.U = 1.4D + 1.7L + 1.7H2a.U = 0.9D + 1.7H 3.U = 1.4D + 1.7L + 1.4F3a.U = 0.9D + 1.4F4.U = 0.75 (1.4D + 1.7L + 1.7W) 4a.U = 0.90D + 1.3W5.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 SSE6a.U = 0.9D + 1.1 SSE7.U = 1.1 D + 1.1 L + 1.0 Wt 7a.U = 0.9D + 1.0 Wt8.U = D + L + T a + R a + 1.5P a9.U = D + L + T a + R a + 1.25P a + 1.25 OBE + 1.0 (Y r + Y j + Y m)10.U = D + L + T a + R a + P a + SSE + 1.0 (Y r + Y j + Y m)Notes - Concrete Structures(1)U is the required section strength base d on strength design methods described in ACI 318-71.(2)In combinations 8, 9, and 10, the maximum values of P a , T a , R a , Y j , Y r , and Y m , 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 a nd stresses may be exceeded under the concentrated loads Y r , Y j , and Y m , provided there will be no loss of function of any safety related system.
MPS3 UFSAR3.8-55Rev. 30 (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.2. Steel StructuresA.Elastic Working Stress De sign Service Load Conditions1.1.0 S = D + L 1a.1.5 S = D + L + T o + R o (If T o and R o = 0 use Equation 1)2.1.0 S = D + L + OBE2a.1.5 S = D + L + T o + R o + OBE (If T o and R o = 0 use Equation 2)3.1.33 S = D + L + W3a.1.5 S = D + L + W + T o + R o (If To and Ro = 0 use Equation 3)
Factored Load Conditions4.1.6 S = D + L + T o + R o + SSE5.1.6 S = D + L + T o + R o + Wt6.1.6 S = D + L + T a + R a + P a7.1.6 S = D + L + T a + R a + P a + 1.0 (Y r+Y j+Y m) + OBE8.1.7 S = D + L + T a + R a + P a + 1.0 (Y r+Y j+Y m) + SSEB. Plastic Design Factored Load Conditions 9.0.9Y = D + L + T a + R a + 1.5 P a9a.1.0Y = D + L + T a + R a + 1.5 P a10.0.9Y = D + L + T a + R a + 1.25 P a + 1.25 OBE + 1.0 (Y r+Y j+Y m)10a.1.0Y = D + L + T a + R a + 1.25 P a + 1.25 OBE + 1.0 (Y r+Y j+Y m)TABLE 3.8-3 LOADS AND LOADING COMBINATIONS MPS3 UFSAR3.8-56Rev. 3011.0.9Y = D + L + T a + R a + P a + SSE + 1.0 (Y r+Y j+Y m)11a.1.0Y = D + L + T a + R a + P a + SSE + 1.0 (Y r+Y j+Y m)Notes - 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 th e 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 P a , T a , R a , Y r , Y j , 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 Y r , Y j , and Y m. When considering these loads, however, local section strengths may be exceeded under the effect of these concentrated loads, provided there wi ll be no loss of function of any safety related system. Furthermore, in comput ing 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).Loads, Definition of Terms, and Nomenclature1.Normal Loads - Those loads encountered during normal plant operations and shutdown. They include the following:D-Dead loads or their related intern al moments and forces including any permanent equipment loadsL-Live loads or their related intern al 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.TABLE 3.8-3 LOADS AND LOADING COMBINATIONS MPS3 UFSAR3.8-57Rev. 30H-Lateral earth pressure, or its related internal moments and forces; H is included in L for steel structures T o-Thermal loads during normal opera ting 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.R o-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 earthquakeW-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
 
W t-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:
P a=Maximum differential pressure lo ad generated by a postulated break T a=Thermal loads under accident condi tions generated by a postulated break R a=Pipe and equipment reactions under accident conditions generated by a postulated break Y r=Loads on the structure generated by the reaction on the broken high-energy pipe during a postulated break Y j=Jet impingement load on a struct ure generated by a postulated break Y m=Missile impact load on a structur e generated by or during a postulated break, such as pipe whippingTABLE 3.8-3 LOADS AND LOADING COMBINATIONS MPS3 UFSARMPS3 UFSAR3.8-58Rev. 30NOTES: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 give n by Equations (9) and (10) should not exceed 0.8 (S A + 1.2 S h).TABLE 3.8-4 LOAD COMBINATIONS FOR ASME III CLASS 2 PENETRATIONS EXCEPT QUENCH, RECIRCULATION, AND SAFETY INJECTION PIPING Plant Operating ConditionNC 3600 EquationsLoad Combinations (1)Allowable StressDesign 8P d + DS hNormal/Upset 9P p + D + E + H1.2 S h 10 (2)T + R + AS A10aS3 S c11 (2)P d + D + T + R + AS A + S hEmergency 9P p + D + H + E' + Y1.8 S hFaulted 9P p + D + H + E' + Y' + A1 2.4 S h (3) P p + D + B'2.4 S h (3)Test 8P t + DS c MPS3 UFSARMPS3 UFSAR3.8-59Rev. 30NOTES: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 give n by Equations (9) and (10) should not exceed 0.8 (S A + 1.2 S h).TABLE 3.8-5 LOAD COMBINATIONS FOR ASME III CL ASS 2 PENETRATIONS FOR THE QUENCH SPRAY, RECIRCULATION SPRAY, AND SAFETY INJECTION SYSTEMS Plant Operating ConditionNC 3600 EquationsLoad Combinations (1)Allowable StressDesign8P d + DS hNormal/Upset9P p + D + E + H1.2 S h 10 (2)T + R + AS A10aS3 S c11 (2)P d + D + T + R + AS A + S hEmergency9P p + D + H + E' + Y1.8 S hFaulted9P p + D + H + E' + Y' + A11.8 S h P p + D + B'1.8 S h 10 (2)T + R' + A' + XS A 11 (2)P d + D + T + R' + A' + XS A + S hTest 8P t + DS c MPS3 UFSARMPS3 UFSAR3.8-60Rev. 30NOTES:1. Refer to Table 3.8-7 for definition of loadings.2. Class MC allowables from Tabl e I-1.1 of ASME Code per NE 3131(d).3. The allowable cumulati ve usage factor is 1.0.4. Q excludes the thermal bending stress.5. P m <1.2S m or S y , PL <1.8S m or 1.5S y and P L + P b <1.8S m or 1.55 y (use the largest value).6. P m <S m , P L <1.5S m and P L + P b <1.5S mTABLE 3.8-6 LOAD COMBINATIO NS FOR ASME III CLASS MC SLEEVED PENETRATIONSPlant Design or Operating Condition ASME Code Reference Load Combinations (1)Allowable Stress (2)DesignPrimary Stress IntensityFig. NE-3221-1P d + D + E + HP m < S m. P L < 1.5S m and P L + P b < 1.5S m Normal/UpsetPrimary and Secondary Stress RangeNB-3222.2P O + T + R + A + E + H + LP L + P b + P e + Q < 3S m Peak Stress RangeNB-3222.4eP O + T + R + A + E + H + L P L + P b + P e + Q + F (3)Simplified Plastic-Elastic AnalysisNB-3228.3(a)P O + D + E + H + L P L + Pb + Q < 3S m (4)Expansion Stress IntensityNB-3222.3T + RP e < 3S m EmergencyFig. NB-3224-1P L + D + E' + H + Y (5)FaultedAppendix FP L + D + E' + H + Y' + A1 (6)Appendix FP L + D + E' + H + B' (6)Appendix FP L + D + E' + X + T + R' + A' (6) (7))
MPS3 UFSAR3.8-61Rev. 30 Elastic - Elastic Analysis - Use an S m value equal to the lesser of 2.4S m or 0.7OS u (8).Inelastic - Elastic Analysis - Use an S m value equal to the greater of 0.7S u or [S y + 1/3 (S u - S y)](8).7. Includes the steady-state load from B' and the stress induced in the sleeve from the liner.8. Use 85 percent of these values if Y' or B' exists in the load combination.
MPS3 UFSAR3.8-62Rev. 30TABLE 3.8-7 NOMENCLATURE FOR TABLES 3.8-4 THROUGH 3.8-6D-Sustained mechanical loads, including d eadweight of piping, components, contents, and insulation.T-Loads due to thermal expansion of the system in response to average fluid temperature.R-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.R'-Loads induced in the piping due to thermal growth of equipment and/or structures to which the piping is connected as a resu lt of plant faulted plant conditions.
Note that R' includes R.E-Inertia effects of the OBE.
E'-Inertia effects of the SSE.A1-Loads induced in the piping due to inertia effects of LOCA or displacements due to LOCA or displacements due to pipe rupture.A-Loads induced in the piping due to response of the connected equipment and/or civil structures to the OBE (commonly refe rred to as OBE anchor movements).A'-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).S-Loads induced due to building settlement effects.H-Loads resulting from occasi onal 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.Y-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.Y'-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.X-Loads induced in the piping due to pressu re response (growth) of the containment during a faulted plant condition.L-Local stress effects in piping and/or pi ping components due to sudden changes in fluid temperature. These loads are commonly referred to as thermal transient effects.B'-Loads on restraints induced by blowdow n and subsequent pipe response of a ruptured system for faulted plant conditions.
P d-Internal pressure loads due to design pressure.
P p-Internal pressure loads due to peak pressure.
MPS3 UFSAR3.8-63Rev. 30 P t-Internal pressure loads due to test pressure.
P 1-Internal pressure loads for emergency and faulted conditions as applicable.
P o-Internal pressure loads due to range of the operating pressure.TABLE 3.8-7 NOMENCLATURE FOR TABLES 3.8-4 THROUGH 3.8-6 MPS-3 FSAR1Rev. 16 3.9 MECHANICAL SYSTEMS AND COMPONENTSSections whose identification numbers include the letter B c ontain material within the balance-of-plant (BOP) scope, while sections w hose identification numbers include the letter N contain material within the nuclear steam supply system (NSSS) scope.
MPS3 UFSAR3.9B-2Rev. 30 3.9B.1 SPECIAL TOPICS FOR MECHANICAL COMPONENTS 3.9B.1.1 Design Transients The design of the reactor coolant system (RCS), RCS component supports, and reactor internals considers the following five opera ting conditions, defined in Secti on III of the ASME Boiler and Pressure 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 Modera te 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 a ny single operator error or control malfunction.b.Transients caused by a fault in a system component re quiring 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 correct ive action does not include any repair of mechanical damage. The estimate d duration of an upset condition is included in the design specific ations for each component.3.Emergency Conditions (Infrequent Inci dents) - Those devi ations from normal conditions which require shutdown for corr ection 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 lo ss of structural integrity results as a concomitant effect of any damage devel oped in the system. The total number of postulated occurrences for such events doe s not cause more than 25 stress cycles having an S value greater than that for 10 6 cycles from the applicable fatigue design curves of the AS ME Code Section III.4.Faulted Conditions (Limiting Faults) - Those combinations of conditions associated with extremely low probabilit y 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.
MPS3 UFSAR3.9B-3Rev. 305.Testing Conditions - Testing conditions are those pressure overload tests including hydrostatic tests, pneu matic tests, and leak tests speci fied. Other types of tests are classified under normal, upset, emergency or faulted conditions.To provide the necessary high degree of integrit y for the components in the RCS the transient conditions selected for component fatigue evalua tion are based upon a conservative estimate of the magnitude and frequency of the temperature a nd pressure transients resulting from various operating conditions in the plant.
The transients provided are representative of operating conditions which prudently should be consider ed to occur during plant operation and are sufficiently severe or frequent to be of possible significance to component cyclic behavior. The transients analyzed may be regard ed as a conservative representation of transients which, used as a basis for component fatigue evaluation, provide confidence that the component is appropriate for its application in acco rdance with the requireme nts of ASME Section III.
The MNPS-3 documents that provide detailed descri ptions of design transi ents used for fatigue evaluation of RCP and associated Class 1 branch piping are listed in Tabl e 3.9B-1. The applicable design transients for RPV internals are described in Section 3.9N.1.
3.9B.1.2 Computer Programs Used in Analysis Lists of computer programs that are used in the design of Seismic Category I components and piping systems within the BOP scope are pr ovided in Appendix 3A. Also included in Appendix 3A are brief descripti ons of each program, the extent of its application, and program verifications which demonstrate the appl icability and validity of each program.
3.9B.1.3 Experimental Stress AnalysisNo experimental stress analysis methods are employed. Analytical methods for the design of BOP equipment, components, and piping systems are used exclusively.
3.9B.1.4 Consideration for the Evalua tion of the Faulted Conditions 3.9B.1.4.1 Loading Conditions The structural stress an alysis performed on the reactor c oolant loop piping and other Seismic Category I ASME Code Class 1, 2, and 3 piping co nsider the loadings and load combinations specified in Tables 3.9B-10, 3.9B-1 1, and 3.9B-12. The stress limits ut ilized for the faulted plant condition are also outlined in Tables 3.9B-10 and 3.9B-11.
The loading conditions and associated stress limits defined in the above tables are applicable for elastic analysis only. Inelastic analysis is not generally used in the qua lification of piping and piping components. Where inelastic analysis is utilized in qualifying the piping and/or components, the MNPS-3 documents containing deta ils of such analysis are identified in Table 3.9B-2. The procedure used for the modeli ng and analytical met hods for evaluating the reactor coolant loop and supports for faulted loading conditions are described in Section 3.9B.1.4.2. The procedure used for mode ling of remaining Seismic Category I ASME MPS3 UFSAR3.9B-4Rev. 30 Code Class 1, 2, and 3 piping syst ems and analytical methods empl oyed for pipe stress analysis are provided in Section 3.7B.3.2. The analysis, t hus performed, complies wi th ASME Section III Subsections NB, NC, ND (1971 through Summer 1973 addenda) and Appendix F.
3.9B.1.4.2 Evaluation of Reactor Coolant Loop and Supports for Faulte d Loading Condition The faulted loading condition of reactor coolan t loop and supports considers loading due to:1.Internal pressure2.Weight3.Safe shutdown earthquake4.Loss-of-coolant accident (pipe break)5.TransientsInternal Pressure Internal pressure is identified as faulted condition pressure in accordance with ASME Section III requirements.Weight Weight consists of the weight of the piping system, insulation (i f any), and contained fluid during normal operating conditions.
Seismic The earthquake loads are part of the mechani cal loading conditions specified in the design specification for piping systems. Mechanical stresses are evaluated for SSE (safe shutdown earthquake) condition consistent with ASME Section III requirements for faulted condition loading evaluation.
Loss of Coolant Accident Mechanical loads are developed in the broken and unbroken reactor coolant loops and in the reactor vessel as a result of transient flow and pressure fluctuations following a postulated pipe break in one of the reactor coolant loops. Structural consideration of dynamic effects of postulated pipe breaks requires postulation of a finite number of break lo cations. Postulat ed pipe break locations are given in Section 3.6.Time history dynamic analysis is performed for these postulated break cases. Hydraulic models are used to generate time-dependent hydraulic forcing functions used in the analysis of the reactor MPS3 UFSAR3.9B-5Rev. 30 coolant system for each break case. For a furthe r description of the hydraulic forcing functions, refer to Section 3.6.Transients Asymmetric pressurization, resulting from the postulated pipe breaks, of the reactor cavity, steam generator, or pressurizer cubicles also impos es unbalanced mechanical loads on the RCS piping and equipment. Pressure time hi stories are developed based on release rates of energy and fluid mass from the break, which are compatible wi th the hydraulic analysis cited above. The mechanical response to as ymmetric pressure loading is calcul ated separately and then combined with the response to hydraulic forces.
Additional localized loadings due to direct impingement of a flui d jet from the postulated pipe break are also combined. Where the jet target is a support, the effect is assumed to be limited to the support and its immediate connection joints. Where the target is a major RCS component, the jet force in excess of asymmetric pressure lo adings is included in the RCS dynamic analysis (Section 6.2.1).
3.9B.1.4.3 Reactor Coolant Loop Models and Methods The analytical methods used in obtaining the solu tion consist of the transfer matrix method and stiffness matrix formulation for th e static structural analysis, th e response spectra method or time history method for seismi c dynamic analysis, and time history direct integration method for the loss-of-coolant accident dynamic analysis.
The integrated reactor coolant loop/supports system model is th e basic system model used to compute loadings on components, component suppo rts, and piping. The system model includes the stiffness and mass character istics of the reactor coolant loop piping and components, the stiffness of supports, the stiffness of auxiliary line piping which affects the system and the stiffness of piping restraints. The deflection solution of the entire system is obtained for the various loading cases from which the internal me mber forces and piping stresses are calculated.Static The reactor coolant loop/supports system model, constructed for the NUPIPE-SW computer program, is represented by an or dered set of data which numer ically describes the physical system. Figure 3.9B-1 shows an isom etric line schematic of this mathematical model for one of four reactor coolant loops. The steam generator and reactor cool ant pump vertical and lateral support members are described in Section 5.4.14.
The spatial geometric description of the react or coolant loop model is based upon the reactor coolant loop piping layout and equipment drawings. The node point coordinates and incremental lengths of the members are determined from thes e drawings. Geometrical properties of the piping and elbows along with the modulus of elasticity E, the coefficient of thermal expansion, the average temperature change from ambient temperature T, and the weight per unit length are specified for each element. The primary equipment supports are represented by stiffness matrices MPS3 UFSAR3.9B-6Rev. 30 which define restraint characteristics of the suppor ts. Due to the symmetry of the static loadings, the reactor pressure vessel nozzles are re presented by a fixed boundary in the system mathematical model. The thermal growth of the reactor vessel is considered in the construction of the model.
The model is made up of a number of sections , each having an overall transfer relationship formed from its group of elements. The linear elas tic properties of the sect ion are used to define the stiffness matrix for the section. Using the transfer relationshi p for a section, th e loads required to suppress all deflections at th e ends of the secti on arising from the ther mal and boundary forces for the section are obtained. These loads are incorporated into the overall load vector.After all the sections have been defined in this manner, the overall stiffness matrix and associated load vector to suppress the deflection of all th e network points is determined. By inverting the stiffness matrix, the flexibility matrix is determined. The flexibility matrix is multiplied by the negative of the load vector to determine the network point deflections due to the thermal and boundary force effects. Using the general transfer relationship, the deflections and internal forces are then determined at all node points in the system.
The static solutions for deadweight and therma l loading conditions are obtained by using the NUPIPE-SW computer program.
Seismic The model used in the static analysis is modifi ed for the dynamic analysis by including the mass characteristics of the piping and equipment. The effect of the equipment motion on the reactor coolant loop/supports system is obtained by modeling the mass and the stiffness characteristics of the equipment in the overall system model.
The steam generator is typically represented by fi ve discrete masses. The lower mass is located near the intersection of the centerlines of the inlet and outlet nozzles of the steam generator. A middle mass is located near the center of the st eam generator and a top mass is located at the outlet of the steam generator. The other two intermediate masses are located at the preheater section and steam separator section, respectively.The reactor coolant pump is typi cally represented by a two-discrete-mass model. The lower mass is located at the intersection of the centerlines of the pump suction and discharge nozzles. The upper mass is located near the center of gravity of the motor.
The component upper and lower lateral supports are passive during plant heatup, cooldown, and normal plant operating conditions. However, thes e restraints become active under the rapid motions of the reactor coolant loop components that occur from the dynamic loadings. Component supports are represented by stiffne ss matrices and/or individual tension or compression spring members in the dynamic model.
The solution for the seismic disturbance employs the response spectra or the time history method.
Both methods employ the lumped mass technique, li near elastic properties, and the principle of MPS3 UFSAR3.9B-7Rev. 30 modal superposition. The floor re sponse spectra or the base-mat motion histories are applied along both horizontal axes and the vertical axis simultaneously.From the mathematical de scription of the system, the overall stiffness matrix [K] is developed from the individual element stiffness matrices using the transfer matrix method. Af ter deleting the rows and columns representing rigid restraints, the stiffness matrix is revised to obtain a reduced stiffness matrix [K R] associated with mass degrees of freedom only. From the mass matrix and the reduced stiffness matrix, the natural frequencies and the normal modes are determined.
In the response spectra method, the modal particip ation factor matrix is computed and combined with the appropriate response spectra value to give the modal amplitude for each mode. The total modal amplitude is obtained by taking the squa re root of the sum of the squares of the contributions for each direction. The modal amplitudes are then converted to displacements in the global coordinate system and applied to the corresponding mass point. From these data the forces, moments, deflections, rotations, support reactions and piping stre sses are calculated for all significant modes. The total seismic response is computed by combining the contributions of the significant modes by using the me thods described in Section 3.7.
For the time history method, the seismic model de scribed above is expanded in scope to include all four reactor coolant loops, the reactor vessel and its support system, and the containment-building- model described in Section 3.8.1 with th e exception of the base-mat and soil springs.
The resulting model is coded for the STARDYNE (Appendix 3A2.5) program system.
Figure 3.9B-2 shows an isometric sketch of a typical loop of the STARDYNE model. The center section model is shown schematically in Figure 3.9B-3. The plan view of the latter figure indicates the points of attachment of the four i ndividual loop submodels to the reactor vessel. In addition, all embedment po ints of steam generator and pump supports are connected to the building model at the mass point of nearest elevation. This connectivity is shown in the computer plots of the complete STAR DYNE model (Figure 3.9B-4).
Prior to extraction of the natura l frequencies and mode shapes, a mass condensati on is performed to reduce the number of dynamic degrees-of-freedom (D-O-F) to 350. The other approximately 3,000 D-O-F are recovered by eigenvector expansion.Loss-of-Coolant Accident The mathematical model used in the time history seismic an alyses is modified for the loss-of-coolant accident analyses. To represent the severance of the reactor coolant loop piping at a postulated guillotine break location, two distinct nodes, each containing six dynamic degrees of freedom and located on each side of the break, are included in the mathematical model.
When no nonlinear elements are acti ve, i.e., for a longitudinal split or severance of the SI, RHR, PS, MS, or FW lines, the dynamic structural solu tion for the loss-of-coolant accident is obtained by using the STARDYNE subprogram DYNRE, as in the time-history seismic analysis. The natural frequencies and eige nvectors are determined from each broken model with the STARDYNE subprogram STAR/HQR.
MPS3 UFSAR3.9B-8Rev. 30 When elements of the system can be repr esented as single acting members (tension or compression members), they are considered as nonlinear elements, which are represented mathematically by the combination of a gap, a spring, and a viscous damper. The force in each nonlinear element is treated as an externally applied force in an iterative solution technique which converges when the force and displacement are compatible. Multiple nonl inear elements can be applied at the same node, if necessary, or at several nodes. The time-history solution is performed in the STARDYNE subprogram DYNRE. The input to this subprogram consists of the system masses and stiffnesses, applied forces, and nonlinear elements.
The transient applied forces are described in Section 3.6B.2.To simulate the release of the strain energy in the pipe for a postulated guillotine break, the internal forces in the system at the break loca tion due to the initial steady state hydraulic forces, thermal forces, and weight forces, are determined by a STAR static analysis. The release of the strain energy is accounted for by ap plying the negative of these inte rnal forces as a step function loading concurrent with the hydraulic forces. The initial conditions are equa l to zero because the solution is for the transient problem (the dynami c response of the system from the static equilibrium position). To be consistent with this analysis scheme, the hydraulic forces are adjusted so as to eliminate th e initial value which corresponds to the initial static loading condition.
The loss of coolant accident displacements of the reactor vessel are applied in time history form as another input to the dynamic analysis of the reactor coolant loop. The loss of coolant accident analysis of the reactor vessel includes all th e forces acting on the vessel including internals reactions, cavity pressure loads, and loop mechan ical loads. The reactor vessel analysis is described in Section 3.9.N.1.4.3.
The time-history displacement response is used in computing support loads and in performing stress evaluation of the reactor coolant loop piping. The support loads [F] are computed by multiplying the support stiffness matrix [K] and th e displacement vector [X
] at the support point.
The support loads are used in the evaluation of the supports.
The time-history displacements of the primary l oop piping are used to determine the internal forces, deflections, and stresses at each end of the piping elements. For this calculation, the displacements are treated as imposed deflections on the reactor coolant piping. The results of this solution are used in the piping stress evaluation.As part of the Stretch Power Uprate (SPU) evaluation, the LOCA analysis was revised taking into account the updated pipe break hydraulic forcing functions and LOCA RPV motions. This revised LOCA analysis was performed using the NUPIPE-SWPC computer program and the leak-before-break (LBB) criteria for primary reactor coolant loop pipe breaks.
3.9B.1.4.4 Primary Component Supports Models and MethodsThe static and dynamic structural analyses employ the matrix method and normal mode theory for the solution of lumped-parameter, multimass stru ctural models, or time-history integration MPS3 UFSAR3.9B-9Rev. 30 method. The equipment support structure models are dual-purpose since they are required to quantitatively represent the elastic restraints which the supports impose upon the loop, and to evaluate the individual support member stresse s due to the forces imposed upon the supports by the loop.Models for the STARDYNE computer program (A ppendix 3A) are constructed for the steam generator lower, steam generator upper lateral, reactor coolant pump lower and pressurizer supports. The reactor vessel suppor t is modeled using the STRUDL computer program. Structure geometry, topology and member prope rties are used in the modeling.
A description of the supports is found in Sec tion 5.4.14. Detailed models are developed using beam elements and plate el ements, where applicable.The respective computer programs are used with these models to obtain support stiffness matrices and member influence coefficients for the steam generator, reactor coolant pump, pressurizer, and reactor vessel supports. Unit force along and unit moment about each coordinate axis are applied to the models at the equipment vertical centerline joint. Stiffness analyses are performed for each unit load for each model.
Joint displacements for applied unit loads are formulated into flexibility matrices. These are inverted to obtain support stiffnes s matrices which are included in the reactor coolant loop model.Loads acting on the supports are obtained from the reactor coolant loop analysis. For each support analyzed, 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 interacti on equations for the specified operating condition. Maximum normal stre ss, shear stress, and comb ined load interaction equation values are printed. ASME Boiler and Pressure Vessel Code Section III, Subsection NF, stress and interaction equati ons are used and results are compared with limits for the operating condition specified.
The reactor vessel support structure is analyzed for all loading conditions using a finite element model. Vertical and horizontal fo rces delivered to th e support structures from the reactor vessel shoe are applied to the structure, and elem ent stresses and concrete forces are obtained.
3.9B.1.4.5 Equipment and Components The elastic analysis techniques de scribed in Section 3.7B.3.1.1 are u tilized in the qualification of Seismic Category I ASME code an d noncode equipment within BOP scope. Stress limits utilized for the faulted plant condition are as outlined in Section 3.9B.3.1. The design conditions and MPS3 UFSAR3.9B-10Rev. 30stress limits defined are applicable for an elastic system (and equipment) analysis. Inelastic analysis is not employed. If, and when, inelastic an alysis is contemplated, detailed design bases, demonstrating maintenance of function and/or structural integrity, will be established prior to implementation.
3.9B.2 DYNAMIC TESTING AND ANALYSIS 3.9B.2.1 Preoperational Vibration and Dynamic Effects Testing on Piping A preoperational vibrati on, thermal expansion (in discrete temperature step increments), and dynamic 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 fa ilure could reduce the functioning of any Seismic Category I plant featur e 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.
A list of the systems and the types of tests being conducted is contained in Table 3.9B-3. The different flow modes of operation and transients to which each system was subjected during the tests are contained in Chapter 14. The test titles, test prerequisites, test objectives, and summary of testing are also described in Chapter 14. For each system defined in items 1 through 4, all flow modes of operation that the systems are subjected to during the tests are visually observed where accessible. In addition, systems that were stress analyzed for fluid flow instabilities have instrumented measurements at selected locations for the specific flow modes analyzed. The measured results were compared to the analytically predicted values. Instrumented measurements are conducted (as needed) for all other systems and conditions. For ASME Code Class 1, 2, and 3 piping systems, design and supervision of the test s, definition of acceptanc e criteria, evaluations of test results, and the making of any changes in the piping system necessary to ensure that the piping is adequately designed and supported, were performed as required by Section III of the ASME code.
Observed vibrations which from visual examination appeared to be excessive in the opinion of experienced engineers who superv ised, conducted, and witnessed the various tests, resulted in either:
MPS3 UFSAR3.9B-11Rev. 301.An instrumented test program is c onducted and the system re-analyzed (or compared to existing analysis) to demonstrate that the observed levels do not cause 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.
In addition to the above, thermal expansion ef fects of piping are obser ved during the testing.
Locations for monitoring thermal expansion are chosen based on:1.Expected large movements2.Areas with tight clearances3.Snubber locations to verify correct and unrestricted motion 3.9B.2.2 Seismic Qualification Testing of Sa fety Related Mechanical Equipment The methods and procedures used in the de sign and qualification of Seismic Category I mechanical equipment within balance-of-plant (BOP) scope are outlined in Sections 3.7B.3.1.1, 3.9B.3, and 3.10B. Loading combinations include op erating as well as earthquake loading for qualification by testing and/
or analytical methods.Safety related mechanical equipment (Seismic Category I), not covered by the ASME Boiler and Pressure Vessel Code, are seismically qualifi ed in accordance with the procedures of Section 3.7B.3.1.1. Non-AS ME Code equipment typically in clude diesel generators, fans, coolers, and emergency ventilation equipment. Cranes are seismically qualified in accordance with criteria that preclude th e possibility of the crane being dislodged by a seismic disturbance.
Except as noted elsewhere in the FSAR, if c odes are used in the design of a component, the guidelines generally requi re the addition of operating loads to the operating base earthquake (OBE) (1/2 Safe Shutdown Earthquake (SSE)) load with no increase in code allowable stress. If no codes are used, the stress level under the combin ed loading is limited to 75 percent of the minimum yield strength of the material per th e ASTM specification. The general criteria for analysis of the SSE, pipe rupture (if applicable), and operating loads require that deformation of components be allowed only with no loss of safety function. Stresses under combined loadings are generally limited to the smaller of 100 percent of the minimum yield strength, or 70 percent of the minimum ultimate tensile strength, of the material (at temperature) per the ASTM, or equivalent specification for the material.
MPS3 UFSAR3.9B-12Rev. 30 3.9B.3 ASME CODE CLASS 1, 2, AND 3 COMP ONENTS, COMPONENT SUPPORTS, AND CORE SUPPORT STRUCTURES 3.9B.3.1 Loading Combinations, Design Transients, and Stress Limits 3.9B.3.1.1 ASME III Class 1 ComponentsComponents and systems which are Seismic Categor y I and which are within the jurisdiction of ASME Section III, Subsection NB (Class 1), have the following combinations and categorizations of load with relevant stress l imits in accordance with NB-3000:1.Design condition - Design pressure and temperature; design mechanical loads.2.Normal condition - Normal operating load s (deadweight, pressure, thermal, etc.).3.Upset condition - Upset plant condition lo ads plus the operating basis earthquake (OBE).4.Emergency condition - Emergency plant condition loads.5.Faulted condition - Dynamic system load s associated with the faulted plant condition plus the safe shutdown earthquake (SSE).6.Testing condition - Hydrostatic tests, pneum atic tests, leak tests, as applicable.
For the conditions specified, the allowable stress limits defined in Tables 3.9B-5 and 3.9B-10 are applicable to stress results obtained by elastic analysis techniques and are compared with Regulatory Guide 1.48 in Table 3.9B-6. The analysis techniques described in Section 3.7B.3.1 are utilized in implementing these criteria. Design transients incl uded in the above category are defined in Section 3.9B.1.
The general extent of compliance with Regul atory Guide 1.48 is di scussed in Section 1.8.
3.9B.3.1.2 ASME III Class 2 and 3 ComponentsTables 3.9B-7 and 3.9B-11 provide loading conditi ons and stress limits for ASME Code Class 2 and 3 components of Seismic Category I fluid syst ems which are constructe d in accordance with ASME III Subsections NC and ND. The conditions generally relate to ASME Section III, Code Class 1 requirements, and incl ude 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 spec ified design loads (as above) plus SSE loads, or pipe rupture loads (if applicable).
MPS3 UFSAR3.9B-13Rev. 30 The design loading combinations are analogous to either the Code Class 1 normal or upset conditions for Design Condition I, or to the faulted (or emergency, if applicable) condition for Design Condition II. The stress limi ts for these design conditions are presented in Tables 3.9B-7 and 3.9B-11. Since design temperature and pressu re exceed those asso ciated with upset, emergency, and faulted conditions, satisfac tion of primary stress limits is assured.
These requirements are intended to be consistent with the present code format and philosophy.
When implemented, they were a supplement to the requirements of ASME III, Subsection NC and ND.The stress limits and design conditions presented in Table 3.9B-7 are intended to ensure that no gross deformation of the component occurs. Table 3.9B-8 compares Table 3.9B-7 with Regulatory Guide 1.48. These limits are appli cable for an elastic system (and component) analysis. Inelastic deformation is not performe d on any ASME Code Class 2 and 3 Components. If, and when, inelastic analysis is contemplated, detailed design bases demonstrating maintenance of either function and/or structural integrit y will be proposed prior to implementation. The analysis techniques of Section 3.7B.3.1 are ut ilized in implementing these criteria.
The general extent of compliance with Regul atory Guide 1.48 is di scussed in Section 1.8.
3.9B.3.2 Pump and Valve Operability AssurancePumps and valves installed in Seismic Category I piping systems are designed in accordance with the requirements of ASME III, NB, NC, and ND.
Inactive pumps and valves are designed for the loading combinations of Sections, 3.9B.3.1 and 3.9B.3.2, and for the stress limits indicated in Table 3.9B-7.
Active components are those that must perfor m a mechanical motion during the course of accomplishing its safety function.
Inactive components are those for which mechanical movement does not occur in order for the component to accomplish it s intended safety function.
Operability of active pumps and valves is as sured by satisfying the re quirements of various programs. Safety related valves are qualified by prototype testing and analysis, and safety related active pumps by analysis with suitable stress limits and nozzle loads. The content of these programs is detailed below.
3.9B.3.2.1 Pump Operating Program Active pumps are qualified for ope rability by being subjec ted to rigid tests bot h prior to and after installation in the plant. The in-shop tests include:1.Hydrostatic tests to ASME Section III requirements.2.Seal leakage tests at the same pr essure used in the hydrostatic tests.
MPS3 UFSAR3.9B-14Rev. 303.Performance tests, while the pump is operated with flow, to determine total developed head, minimum and maximum he ad, net positive suction head (NPSH) requirements, and other pump/motor parameters.
Also monitored during these operation tests are b earing temperatures and vibration levels, which are shown to be below appropriate limits specified to the manufact urer for design of each active pump.After the pump is installed in the plant, it undergoes the cold hydro tests, hot functional tests, and the required periodic inservice inspection and ope ration as applicable. These tests demonstrate reliability of the pump for the design life of the plant.
In addition to these tests, the safety related ac tive pumps are qualified for operability during an SSE 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.
The pump manufacturer is required to show that the pump operates normally when subjected to the maximum applicable amplified seismic (floor) accelerations, attached piping nozzle loads, and dynamic system loads associated with the fa ulted operating condition. Analysis and/or testing procedures are utilized in accordance with those outlined in Section 3.7B.3.1.1. Natural frequency calculations are performed in order to dete rmine maximum seismic accelerations based on applicable amplified (f loor) response spectra.To avoid damage during the faulted condition, th e stresses caused by the combination of normal operating loads, SSE, and dynamic system loads are limited as indicated in Table 3.9B-7. The average membrane stress (P m) for the faulted condition loads is maintained at 1.2 S, or approximately 0.75 S y (S y = yield stress) and the maximum stress in local fibers (P m + bending stress P b) is limited to 1.8 S, or approximately 1.1 S
: y. In addition, the pump stresses caused by the maximum seismic nozzle loads are limited to the stresses outlined in Ta ble 3.9B-7. The maximum seismic nozzle loads are also considered in an analysis of the pump supports to assure that a system misalignment cannot occur. A static shaft deflection analysis of the rotor is performed with horizontal and vertical accelerations based on floor re sponse levels. The deflections determined from the static shaft analysis are compared to the allowable rotor clearances. The nature of seismic disturbances dictates that the maximum contact (if it occurs) is of short duration.Class 1 pumps are designed/analyzed according to the rules of ASME Section III, Subsection NB 3400.Performance of these analyses with the conservati ve loads stated and, with the restrictive stress limits of Table 3.9B.5 as allowables, assures th at critical parts of the pump are not damaged during the short duration of the faulted condition and that the reliability of the pump for post-faulted condition operation is not impaired by the seismic event.
MPS3 UFSAR3.9B-15Rev. 30 In addition to the post-faulted condition operati on, it is necessary to assure that the pump functions throughout the SSE. The pump/motor comb ination is designed to rotate at a constant speed under all conditions unless the rotor becomes completely seized (i.e., no rotation). Typically, the rotor can be seized 5 full seconds before a circuit breaker trips to prevent damage to the motor. However, the high rotary inertia in the operating pump rotor and the nature of the random, short duration loading characteristics of the seismic event prevent the rotor from becoming seized. In actuality, the seismic loadings cause only a slight increase, if any, in the torque (i.e., motor current) necessary to drive the pump at the constant design speed. Therefore, the pump does not shut down during the SSE and operates at the design despite the SSE loads.To complete the seismic qualific ation procedures, the pump motor is independently qualified for operation during the maximum seismic event. Any auxiliary equipment which is vital to the operation of the pump or pump motor, and whic h is not qualified for operation during the pump analysis or motor qualifications, is also separately qualified for operation at the accelerations occurring at its mounting. The pump motor and vi tal auxiliary equipment are qualified by meeting the requirements of IEEE Std 344-1975. If the test ing option is chosen, si nusoidal or sine-beat testing is justified by satisfying one or more of the following requirements to demonstrate that the multi-frequency response is negligible or the input is of sufficient magni tude to conservatively account for this effect:1.The equipment response is basically due to one mode.2.The sinusoidal or sine-beat response spec tra 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.
In general, the degree of coupling in the equipment determines if a single or multi-axis test is required. Multi-axis testing is requi red if there is considerable cr oss coupling. If coupling is very light, then single axis testing is justified. If th e degree of coupling can be determined, then single axis testing can be used with the input sufficient ly increased to include the effect of coupling on the response of the equipment.From previous arguments, the safety relate d pump/motor assemblies are not damaged and continue operating under SSE loading and, theref ore, perform their intended functions. These proposed requirements take into account the complex characteristics of the pump and are sufficient to demonstrate and assure the seismi c operability of the acti ve pumps. The functional ability of active pumps after a faulted condition is assured, since only normal operating loads and steady state nozzle loads exist. Since it is demonstrated that the pumps are not damaged during the faulted condition, the post-faulte d condition operating loads are id entical to the normal plant operating loads. This is assured by requiring that the imposed nozzle loads (steady state loads) for normal conditions and post-faul ted conditions are limited by the magnitudes of the normal condition nozzle loads. The post-faulted condition ability of the pumps to function under these applied loads is proven during the normal ope rating plant conditions for active pumps.
MPS3 UFSAR3.9B-16Rev. 30 3.9B.3.2.2 Valve Operability ProgramSafety related active valves must perform th eir mechanical motion during the course of performing their safety function. Assurance that th ese valves will operate during a seismic event must be supplied. Qualification te sts accompanied by analyses ar e conducted for all active valve assemblies.Valves without significant extended structure are proven seismically ad equate by analysis of piping seismic adequacy. For valves with operators having significant exte nded structures, and if these structures are essential to maintaining pressure integrity, analysis is performed based upon static forces resulting from equi valent earthquake accelerations acti ng at the centers of gravity of the extended masses. For active valves, this requirement for analysis is extended to the mechanical (nonpressur e boundary) components of valve top-works to ensure operability.
The safety related valves are subjected to a seri es of stringent tests prior to service and during plant 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 valv e 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.
Cold hydro qualification tests, hot functional qua lification tests, periodic inservice inspections, and periodic inservice operation are performed in situ to verify and assure the functional ability of the valve. These tests guarantee reliability of the valve for the desi gn life of the plant. The valves are designed using either stress analyses or the pressure containing minimum wall thickness requirements. On all active valves, an analysis of the extended structure is also performed for static equivalent seismic SSE lo ads supplied at the centers of gr avity of the extended structure. The maximum stress limits allowed in these analyses show structural integrity. The limits that are used for Class 2 and 3 active valves are shown in Table 3.9B-7. Class 1 valves are designed/
analyzed according to the rules of ASME Section III, Section NB-3500.In addition to these tests and analyses, representative valves of each design type are tested for verification of operability during a simulate d seismic event by de monstrating operational capabilities within the specified limits. The testing procedures are described below.
The valve is mounted in a manne r which conservatively represen ts a typical valve installation.
The valve includes the operator and all appurtenanc es normally attached to the valve in service.
MPS3 UFSAR3.9B-17Rev. 30 The operability of the valve during an SSE is de monstrated by satisfying the following criteria: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 load ed 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 th e 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 a ppurtenances 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.The accelerations used for static valve qualification are 3.0 g horizontal and 2.0 g vertical applied with the valve in its proper orientation. Where it is necessary to allow for random orientation of a valve, a 3.0 g horizontal and a 3.0 g vertical are applied. The pi ping designer maintains the motor operator accelerations to these levels with an adequate margin of safety.If the frequency of the valve, by test or analysis, is less than 33 Hz, the valve system is analyzed to determine the equivalent acceler ation applied during th e static test. The analysis provides the amplification of the input acceleration considering the natural frequency of the valve and the frequency content of the applicable plant floor response spectra. The adjusted acceleration is then used in the static analysis and valve operability is assured by the methods outlined in steps (2) to (4) above, using the modified acceleration input.
The above testing program applie s only to valves with overha nging structures (i.e., the motor operator). The testing is conducted on a representative number of valves. Valves from each of the primary safety related design types (e.g., motor-operated gate valve) are tested. Valve sizes which cover the range of sizes in service are qualified by the tests and the results are used to qualify all valves within the intermediate range of sizes. St ress and deformation analys is is used to support the extrapolation.Valves which are safety related, but can be classified as not having an overhanging structure, such as check valves and safety-relief valves, are considered separately.Check valves are characteristically simple in design and their operation is not affected by seismic accelerations or the maximum applied nozzle loads. The check valve design is compact and there are no extended structures or masses whose motion could cause distortions which could restrict operation of the valve. The nozzle loads due to maximum seismic excitation do not affect the MPS3 UFSAR3.9B-18Rev. 30functional ability of the valve, since the valve disc is designed to be isolated from the casing wall.
The clearance supplied by the design around the di sc prevents the disc from becoming bound or restricted due to any casing distortions caused by nozzle loads. Therefore, the design of these valves is such that once the structural integrity of the valve is assured using standard design or analysis methods, the ability of the valve to opera te is assured by the desi gn features. In addition to these 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.
Safety and relief valves are subj ected to tests and analyses simi lar to check valves; stress and deformation analyses for SSE lo ads, in-shop hydrostatic and seat leakage tests, and periodic in situ valve inspection. In addition, a static load equivalent to the SSE is applied to the top of bonnet and the pressure is increased until the valve mechanism is activated. Successful actuation within design requirements assures its overpressurization safety capabilities during a seismic event.
Using the methods described, all the safety rela ted valves in the system are qualified for operability during a seis mic event. These methods conservativ ely simulate the seismic event and ensure that the active valves perform their safety related function when necessary.Alternative valve operability testing, such as dynamic vibration testing, is allowed if it is shown to adequately assure the faulted condition functional ability of the valve system.
3.9B.3.3 Design and Installation Details for M ounting of Pressure Relief DevicesThe design criteria for all safety and relief valves are in accordance with the rules in Subarticles NB-3677 and NC-3677 of the ASME Boiler and Pressure Vessel Code, Section III (see Component Description for relief valves Letdow n Line Downstream of Low Pressure Letdown Valves, Sealwater Return Line, and Letdown Reheat Heat Exchanger section 9.3.4.2.5), and the rules of Code Case 1569, applicable to the classification of the piping component under investigation. For open relief systems, the design criteria and the analys es used to calculate maximum stresses and stress intensities are in accordance with Subarticles NB-3600 and NC-3600 of Section III. The maximum stresses are calculated based upon the full discharge loads, including the effects of the system dynamic respons e, and the system design internal pressure. Stresses are determined for all significant points in the piping sy stem including the safety valve inlet pipe nozzle and the nozzle to shell juncture.
MPS3 UFSAR3.9B-19Rev. 30 3.9B.3.3.1 Open Relief SystemThe total stead-state discharge thrust load for an open system discharge is expressed as the sum of the pressure and momentum forces as follows:F = 144PA + [
AV 2]/g(3.9B-1) where:F = Total reaction force (lb)
A = Exit flow area (sq ft)
P = Exit pressure (psig)
V = Exit fluid velocity (fps)
P = 32.2 fps 2 To ensure consideration of the effects of the s uddenly applied load at the junction of the valve nozzle and pipe run, a dynamic load factor is computed. The calculation of dynamic load factor is based on modeling the valve and nozzle as a si ngle degree of freedom dynamic system. The lumped mass of this system corresponds to the we ight of the valve and nozzle and is assumed to be at the valve center of gravity. The rotational de gree of freedom of this system is considered to be in the direction that causes maximum bending stress in the nozzl e at the junction of the nozzle and run-pipe. Rotational flexibility of the system is computed by a series combination of nozzle flexibility and local run-pipe flexibility (at the junction of th e nozzle and run-pipe).The rise time of the discharge force at the outlet of the safety valve elbow is assumed to be the minimum valve opening time, and the discharge force is assumed to rise linearly with time and remain thereafter constant value. The ratio of maximum dynamic rotations predicted by this single degree of freedom system to the corresponding static rotation caused by the steady state discharge force represents the dynamic load factor.To ensure the consideration of the effects of the suddenly applied loads on the pipe system, a dynamic time-history analysis is performed on the piping system.
The forcing function applied at the point of discharge is a linear force change from zero to the value of F that is determined in the above equation over a time period, t, that corresponds to the valve opening time which is provided by the valve manufacturer. After time t has been reached, the force remains at the value of F until the conclusion of the time-history integration. Th e lumped mass model that represents the piping system includes the safety/relief valves.Where more than one valve is mounted on a common header, two cases are computed. In the first, full discharge of all valves is assumed to occur simultaneously. In the second, the forcing functions are applied to a combinat ion of valves that yields the wo rst-load case. This worst-load case is first verified by trial thr ough a series of static load cases.
MPS3 UFSAR3.9B-20Rev. 30 3.9B.3.3.2 Closed Relief SystemFor relief valves discharging into a closed system, an analytical model of one-dimensional transient flow characteristics following the blow-off of the upstream safety/relief valve into the discharging piping system is established. The ti me-dependent pressure, temperature, density, velocity, and hence the momentum of the downstr eam pipe flow are then computed from this conservative hydrodynamic/thermodynamic flow model. The effects such as flow restrictions and frictional resistance are considered.
The unbalanced transient hydraulic forcing function acting on the piping system computed from the flow model is then used to determine the tr ansient dynamic responses of the piping structural model. Adapting the lumped-mass method incorporated with the modal analysis of piping system, the time-history modal responses are computed.
Computations of maximum stress intensities for ASME Code Class 1 piping or maximum stress le vels for ASME Code Class 2 and 3 piping are based on the dynamic analysis of the system.
3.9B.3.4 Component Supports Component supports which are Seismic Category I and which are within the jurisdiction of ASME Section III, Subsection NF, utilize applic able loading conditions outlined in Sections 3.9B.3.1.1 and 3.9B.3.1.2 for ASME Class 1, 2, and 3 components. Stress limits and loading combinations to be utilized for component support evaluation are as outlined in Tables 3.9B-9 and 3.9B-9A, respectively. The stress limits are used in conjunction with the applicable analytical procedures outlined in Section 3.7B.3 to assure integrity of supports and mounted components.
Component supports which do not come under the jurisdiction of ASME Section III, Subsection NF are summarized in Section 5.4.14.
The loads considered in the analysis of the supports are normal opera ting, seismic (OBE and SSE), 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 reac tor cavity and the SG and RCP cubicles.3.Fluid jet impinging on the supports or components, as applicable.Dynamic analyses were performed to determine the stresses in the component supports as well as the pipe rupture restraints. The results of th e analysis are shown in the following tables.
1.Table 3.9B-15 gives the location of the postula ted 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.
MPS3 UFSAR3.9B-21Rev. 30 4.Table 3.9B-19 gives the stresses in the component support of the pressurizer.
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.
MPS3 UFSAR3.9N-27Rev. 30 3.9N MECHANICAL SYSTEMS AND COMPONENTS 3.9N.1 SPECIAL TOPICS FOR MECHANICAL COMPONENTS 3.9N.1.1 Design Transients The following five operating conditions, as define d in Section III of the ASME B&PV Code, are considered in the design of the reactor coolan t system (RCS) component s 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 an ticipated to occur often enough that design should include a capability to with stand the conditions without operational impairment. The upset conditions include those transients which result from any single operator error or control malfuncti on, 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 10 6 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 speci fied by jurisdictional authorities.5.Testing Conditions MPS3 UFSAR3.9N-28Rev. 30Testing conditions are those pressure ove rload tests including hydrostatic tests and pneumatic tests. Other types of tests are classified under normal, upset, emergency or faulted conditions as appropriate.To provide the necessary high degree of integrity for the equipment in the RCS the transient conditions selected for equipment fatigue evalua tion are based upon a cons ervative estimate of the magnitude and frequency of the temperature a nd pressure transients resulting from various operating conditions in the plant. To a large extent, these transients ar e based upon engineering judgment and experience, and are c onsidered to be of such magnit ude and/or frequency as to be significant in the component de sign and fatigue evaluation. The transients selected may be regarded as a conservati ve representation of transients wh ich, used as a basis for component design and evaluation, provide conf idence that the component is a ppropriate for its application over the design life of the plant.
The design transients and the number of cycles of each that are used fo r fatigue evaluations are shown in Table 3.9N-1. In accordance with ASME III, emergency and faulted conditions are not included in fatigue evaluations.
Normal Conditions The following primary system transien ts are considered normal conditions:1.Heatup and cooldown at 100
&deg;F per hour2.Unit loading and unloading at 5 percent of full power/per minute3.Step load increase and decrease of 10 percent of full power4.Large step load decrease with steam dump5.Steady state fluctuationsa.Initialb.Random6.Feedwater cycling at hot shutdown7.Not Used.8.Unit loading and unloading between 0 and 15 percent of full power9.Boron concentration equalization10.Refueling MPS3 UFSAR3.9N-29Rev. 3011.Reduced temperature return to power12.Reactor coolant pumps startup and shutdown13.Turbine roll test14.Primary side leak test15.Secondary side leak test16.Tube leakage test17.Heaters out of service 1.Heatup and Cooldown at 100
&deg;F per Hour The design heatup and cooldown cas es are conservatively repres ented by continuous operations performed at a uniform temperature rate of 100
&deg;F per hour. (These operations can take place at lower rates approaching the minimum of 0
&deg;F per hour. The expected normal rates are 50
&deg;F per hour).For these cases, the heatup occurs from ambient (assumed to be 120
&deg;F) to the no-load temperature and pressure condition and the cool down represents the reverse situ ation. In actual practice, the rate of temperature change of 100
&deg;F per hour will not be attained because of other limitations such as:a.Material ductility considerations whic h 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 cy cles due to such factors as pressurizer steam bubble formation, rod withdrawal , sampling, water chemistry and gas adjustments.
The number of such complete heatup and cooldow n operations is specified as 200 each, for the 60 year plant design life.
2.Unit Loading and Unloading at 5 Percent of Full Power per Minute The unit loading and unloading ope rations are conservatively represented by continuous and uniform ramp power change of 5 percent per minut e between 15 percent load and full load. This load swing is the maximum possible consistent with operation under automatic reactor control.
The reactor temperature will vary with load as prescribed by the reactor control system. The number of loading and unloading operations is defined as 13,200 during the 60 year design life of the plant.
MPS3 UFSAR3.9N-30Rev. 30 3.Step Load Increase and Decrease of 10 Percent of Full Power The +/-10 percent step change in load demand result s from disturbances in the electrical network into which the plant output is tied. The reacto r control system is designed to restore plant equilibrium without reactor trip following a 10 percent step change in turbine load demand initiated from nuclear plant equilibrium condi tions in the range between 15 percent and 100 percent of full load, the power range for automatic reactor control. In effect, during load change conditions, the reactor control system attempts to match turbine and reactor outputs in such a manner that peak reactor coolant temperature is minimized and reactor coolant temperature is restored to its programmed setpoint at a sufficie ntly slow rate to prevent excessive pressurizer pressure decrease.
Following a step decrease in turbine load, the secondary side steam pr essure and temperature initially increase since the decrease in nuclear power lags behind the step decrease in turbine load. During the same increment of time, the RCS aver age temperature and pressurizer pressure also initially increase. Because of the power mism atch between the turbine and reactor and the increase in reactor coolant temperature, the control system automatically inserts the control rods to reduce core power. With the load decrease, the reactor coolant temperature will ultimately be reduced from its peak value to a value below its initial equilibrium value at the inception of the transient. The reactor coolant average temperatur e setpoint change is made as a function of turbine-generator load as determined by first stage turbine pressure measurement. The pressurizer pressure will also decrease from its peak pressure value and follow the reactor coolant decreasing temperature trend. At some point during the decreasing pr essure transient, the saturated water in the pressurizer begins to flash which reduces th e rate of pressure decrease. Subsequently the pressurizer heaters come on to restore the plant pressure to its normal value.
Following a step increase in turbine load, the re verse situation occurs, i.
e., the secondary side steam pressure and temperature initially decrease and the reactor coolant average temperature and pressure initially decrease. The control system automatically withdraws the control rods to increase core power. The decreasing pressure tran sient is reversed by actua tion of the pressurizer heaters and eventually the system pressure is re stored to its normal value. The reactor coolant average temperature will be raised to a value a bove its initial equilibriu m value at the beginning of the transient.
The number of each operation is specified at 2,000 times for the 60 year plant design life.
4.Large Step Load Decrease with Steam Dump This transient applies to a step decrease in turbine load from full power, of such magnitude that the resultant rapid increase in reactor coolant average temperat ure and secondary side steam pressure and temperature will automatically in itiate a secondary side steam dump that will prevent both reactor trip and lifting of steam generator and pressurizer safety valves. This plant is designed to accept a step decrease of 50 percent from full power, and the steam dump system provides the heat sink to accept the difference in allowable unloa ding rates between the turbine and the reactor coolant system.
MPS3 UFSAR3.9N-31Rev. 30 The number of occurrences of this transient is specified at 200 ti mes for the 60 year plant design life.5.Steady State Fluctuations It is assumed that reactor coolant pressure and temperature and pressure at any point in the system vary 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&deg;F and pressure by
+/-25 psi, once during each 2 mi nute period. The total number of occurrences is limited to 1.5 x 10
: 5. The fluctuations are assumed to occur consecutively, and not simultaneous ly with the random fluctuations.b.Random Fluctuations - Temperature is assumed to vary by
+/-0.5&deg;F and pressure by
+/-6 psi, once every 6 minutes. With a 6 minute period, th e total number of occurrences during the plant de sign life does not exceed 3.0 x 10 6.6.Feedwater Cycling at Hot Shutdown These transients can occur when the plant is at "no load" conditions, during which intermittent feeding of 32
&deg;F feedwater into the steam generators is assumed. Due to fluctuations arising from this mode of operation, the reactor coolant aver age temperature decreases to a lower value and then immediately begins to retu rn to normal no-load temperature. This transient is assumed to occur 2,000 times over the life of the plant.
7.Not Used 8.Unit Loading and Unloading Between 0 and 15 Percent of Full Power The unit loading and unloading cas es between the zero and 15 pe rcent power are represented by continuous and uniform ramp power changes, requiring 30 minutes for loading and 5 minutes for unloading. During loading, reactor coolant temperatur es are increased from the no load value to the normal load program temperatures at the 15 percent power level. The reverse temperature change occurs during unloading.
Prior to loading, it is assumed that the plant is at hot shutdown conditions, with 32
&deg;F feedwater cycling. During the 2-hour period following the beginning of loading, the feedwater temperature increases from 32
&deg;F to 300&deg;F due to steam dump and turbine startup heat input to the feedwater. Subsequent to unloading, feedwater heating is terminated, steam dump is reduced to residual heat removal requirements, and feedwater temperature decays from 300
&deg;F to 32&deg;F.The number of these loading and unloading transients is assumed to be 500 each during the 60 year plant design life.
MPS3 UFSAR3.9N-32Rev. 30 9.Boron Concentration Equalization Following any large change in boron concentration in the RCS, spray is initiated in order to equalize concentration between the loops and the pressurizer. This can be done by manually operating the pressurizer backup heaters, thus caus ing a pressure increase , which will initiate spray at a compensated pressurizer pressure of approximately 2,275 psia.
The proportional sprays return the pressure to 2,250 psia and maintain th is pressure by matching the heat input from the backup heater until the concentrat ion is equalized. For design purposes, it is assumed that this operation is performed once after each load change in the design load follow cycle. The total number of load change occurrences is 26,400 over the 60  year design life.
10.Refueling At the end of the plant cooldown, the temperatur e of the fluid in the RCS is less than 125
&deg;F. At this time, the vessel head is removed and the re fueling canal is filled.
This is done by pumping water from the refueling water storage tank, which is outside and conservatively assumed to be at 32&deg;F, into the loops by means of the low head safety injection pumps. The refueling water flows directly into the reactor vessel via the accumulator connections and cold legs.
This operation is assumed to occur 80 times over the life of the plant.11.Reduced Temperature Return to Power The reduced temperature return to power operati on is designed to improve the spinning reserve capabilities of the plant during load following ope rations without part le ngth rods. The transient will normally begin at the ebb (50 percent) of a load follow cycle and will proceed at a rapid positive rate (typically 5 percent per minute) until the ab ilities of the control rods and the coolant temperature reduction (negative moderator coefficient) to supply reactivity are exhausted. At that point, further power increases ar e limited to approximately one percent per minute by the ability of the boron system to dilute the reactor coolant. The reduction in primary coolant temperature is limited by the protection system to about 20
&deg;F below the programmed value.
The reduced temperature return to power operation is not intended fo r daily use. It is designed to supply additional plant capabilities when require d because of network fa ult or upset condition.
Hence, this mode of operation is e xpected to occur 2,000 times in 60 years.
12.Reactor Coolant Pump (RCP) Startup and Shutdown The reactor coolant pumps are st arted and stopped during routine operations such as RCS venting, plant heatup and cooldown, and in connection with recovery from certain transients such as loss of power. Other (undefined) circumstances may also require pump starting and stopping.
Of the spectrum of RCS pressure and temper ature conditions under whic h these operations may occur, three conditions have been selected for defining transients:a.Cold condition - 70
&deg;F and 400 psig MPS3 UFSAR3.9N-33Rev. 30b.Pump restart condition - 100
&deg;F and 400 psigc.Hot condition - 557
&deg;F and 2235 psig For RCP starting and stopping operations, it is a ssumed that variations in RCS primary side temperature and in pressurizer pressure and te mperature are negligible and that the steam generator secondary side is completely unaffecte
: d. The only significant va riables are the primary system flow and the pressure changes resulting from the pump operations. Occurrences for the pump starting/stopping conditions are given in Table 3.9N-1.
13.Turbine Roll Test This transient is imposed upon the plant during th e hot functional test period for turbine cycle checkout. Reactor coolant pump power is used to heat the reactor coolant to operating temperature (no-load conditions) a nd the steam generated is used to perform a turbine roll test. However, the plant cooldown during this test exceeds the 100
&deg;F per hour design rate.The number of such test cycles is specified at 20 times, to be performed at the beginning of plant operating life prior to reactor operation.
14.Primary Side Leakage Test Subsequent to each time the primary system has been opened, a leakage test will be performed.
During this test the primary side pressure may be raised to a maximum of 2500 psia, in accordance with the system temperature limitations imposed y the reactor vessel material ductility requirements, while the system is checked for leaks. Normally, to prevent the pressurizer safety valves from lifting during the leak test, the primary system will be pressurized to approximately 2,235 psig, as measured at the pressurizer.During this leakage test, the secondary side of the steam generator must be pressurized so that the pressure differential across the tube sheet does not exceed 1,600 psi. This is accomplished with the steam, feedwater, and blowdown lines closed off. For design purposes it is assumed that 200 cycles of this test will occur dur ing the 60 year life of the plant.
15.Secondary Side Leakage Test During the life of the plant, it may be necessary to check the secondary side of the steam generator (particularly, the manway closure) for leakage.
For the design purposes, it is assumed that the steam generator secondary side is pressurized to just below its design pressure, to prevent the safety valves from lifting. In order not to excee d a secondary side to primary side pressure differential of 670 psi, the primary side must al so be pressurized. The pr imary system must be above the minimum temperature impos ed by reactor vessel material ductility requirements. It is assumed that this test is performed 80 ti mes during the 60 year life of the plant.
16.Tube Leakage Test
 
MPS3 UFSAR3.9N-34Rev. 30During the life of the plant it may be necessary to check the steam generator for tube leakage and the tube to tube sheet leakage. This is done by visual inspection of the underside (channel head side) of the tube sheet for water leakage, with the secondary side pressurized. Tube leakage tests are performed during plant cold shutdowns.
For these tests the secondary side of the steam generator is pressurized with water, initially at a relatively low pressure, and the primary system remains depressuri zed. The underside of the tube sheet is examined visually for leaks. If any are observed, the secondary side is then depressurized and repairs made by tube plugging. The secondary side is then repr essurized (to a higher pressure) and the underside of the tube sheet is again checked for leaks. This process is repeated until all the leaks are repaired. The maximum (final) seconda ry side test pressure reached is 840 psig.
The total number of tube leakage te st cycles is defined as 800 duri ng the 60 year life of the plant.
Following is a breakdown of the anticipated number of occurrences at each secondary side test pressure: Both the primary and secondary sides of the stea m generators will be at ambient temperatures during these tests.
17.Heaters Out of Service These transients occur when one or more feedwate r heaters are taken out of service. During the period of time that the heaters are out of service, it is desirable to maintain the plant at full rated thermal load. To accomplish this, first the steam fl ow is reduced to the am ount that will maintain the plant at full rated thermal load when the heater(s) is taken out of service. It takes approximately 10 minutes for plant conditions to reach a new steady state. Then the heater(s) is taken out of service.The two cases considered here are one heater out of service and one bank of heaters out of service.
For design purposes, it is assumed that each of these transients occurs 120 times over the life of the plant.
Upset Conditions The following primary system transien ts are considered upset conditions:1.Loss of load (without immediate reactor trip)Test Pressure (psig)Number of Occurrences200400400200600120 840 80 MPS3 UFSAR3.9N-35Rev. 302.Loss of power3.Partial loss of flow4.Reactor trip from full power5.Inadvertent reactor coolant system depressurization6.Control rod drop7.Inadvertent safety injection actuation8.Operating basis earthquake9.Excessive feedwater flow10.RCS Cold Overpressurization 1.Loss of Load (without immediate reactor trip)
This transient applies to a step decrease in turb ine load from full power (turbine trip) without immediately initiating a reactor trip and represents the most severe pressure transient on the RCS under upset conditions. The reactor eventually trips as a conseque nce of a high pressurizer level trip initiated by the reactor pr otection system (RPS). Since redundant means of tripping the reactor are provided as part of the RPS, transients of this nature are not expected, but is included to ensure a conservative design.
The number of occurrences of this transient is specified at 80 ti mes for the 60 year plant design life.2.Loss of Power This transient applies to a blackout situation involving the loss of offsite electrical power to the station, assumed to be operating initially at 100-percent power, followed by reactor and turbine trips. Under these circumstances, the reactor coolant pumps are deenergized and, following coastdown of the reactor coolant pumps, natural circulation builds up in the system to some equilibrium value. This condi tion permits removal of core residual heat through the steam generators which at this time are receiving feedwater, assumed to be at 32
&deg;F, from the auxiliary feedwater system operating from diesel generator power. Steam is removed for reactor cooldown through atmospheric relief valv es provided for this purpose.
The number of occurrences of this transient is specified at 40 ti mes for the 60 year plant design life.3.Partial Loss of Flow
 
MPS3 UFSAR3.9N-36Rev. 30This transient applies to a partial loss of flow from full power, in which a reactor coolant pump is tripped out of service as the result of a loss of power to that pu mp. The consequences of such an event are a reactor trip, on low reactor coolant flow, followed by turbine trip and automatic opening of the steam dump system. Flow reversal occurs in the affected loop which causes reactor coolant at cold leg temperature to pass through the steam generator and be cooled still further.
This cooler water then flows through the hot le g piping and enters the reactor vessel outlet nozzles. The net result of the flow reversal is a sizable reduction in the hot leg coolant temperature of the affected loop.
The number of occurrences of this transient is specified at 80 ti mes for the 60 year plant design life.4.Reactor Trip From Full Power A reactor trip from full power may occur from a variety of causes resulting in temperature and pressure transients in the RCS and in the secondary side of the steam generator. This is the result of continued heat transfer from the reactor coolant in the steam generator. The transient continues until the reactor c oolant and steam generator secondary side temperatures are in equilibrium at zero power conditions. A continued supply of feedwater and controlled dumping of steam remove the core residual heat and prevent the steam ge nerator safety valves from lifting. The reactor coolant temperature and pressure undergo a rapid decrease from full power values as the RPS causes the control rods to move into the core. For design purposes, reactor trip is assumed to occur a total of 400 times over the life of the plant.The severity of the cooldown transients associat ed with reactor trips depends on the extent of secondary side cooling. Three cooldown cases are considered:a.Reactor trip with no inadve rtent cooldown - 230 occurrencesb.Reactor trip with cooldown but no safety injection - 160 occurrencesc.Reactor trip with cooldown actuat ing safety injection - 10 occurrences 5.Inadvertent Reactor Coolant System Depressurization Several events can be postulated as occurri ng during normal plant operation which will cause rapid 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 controll er causing two pressurizer spray valves to open.
MPS3 UFSAR3.9N-37Rev. 30d.Inadvertent opening of one pressurizer spray valve, due either to equipment malfunction or operator error.e.Inadvertent auxiliary spray. A "lockout" feature on auxiliary spray valve prevents inadvertent spray of unheated water in the pressurizer.Of these events, the pressurizer safety valve actuation causes the most severe transients, and is used as an "umbrella" case to conservatively represent the reactor coolant pressure and temperature variations ar ising from any of them.
When a pressurizer safety valve opens, and remain s open, the system rapidly depressurizes, the reactor trips, and the safety injection system is actuated. Also, the passive accumulators of the SIS are actuated when pressure decreases by a pproximately 1,600 psi. The depressurization and cooldown are eventually terminat ed by operator action. All of these effects are completed within approximately 18 minutes. It is conservatively assumed that none of the pressurizer heaters are energized.With pressure constant and sa fety injection in operation, boil off of hot leg liquid through the pressurizer and open safety valve will continue.
For design purposes this transient is assumed to occur 20 times during the 60 year design life of the plant.
6.Control Rod Drop This transient occurs if a bank of control rods drops into the fully inserted position due to a single component failure. The reactor is assumed to be tripped on OTT, OPT or low pressurizer pressure. It is assumed that this transien t occurs 80 times over the life of the plant.
7.Inadvertent Safety Injection Actuation For design purposes, no credit is taken for the P-19 cold leg injection permissive for the inadvertent safety injection actuation. A spurious safety injection signal re sults in an immediate reactor trip followed by actuation of the high head centrifugal charging pumps. Without crediting P-19, these pumps deliver borated water to the RCS cold legs. The initial portion of this transient is similar to the reactor trip from full pow er with no cooldown. Controlled steam dump and auxiliary feedwater flow after trip removes core residual heat. Reactor coolant temperature and pressure decrease as the cont rol rods move into the core.Later in the transient, the injected water causes the RCS pressure to increase to the pressurizer power operated relief valve (PORV) setpoint a nd the primary and secondary temperatures to decrease gradually. Operator action is credited to verify that the PORV block valves are open within 10 minutes of initiation of the event. Wate r relief through the safety valves will not occur provided at least one PORV actuates on demand. The transient c ontinues until the operator stops the charging pumps. It is assumed that the plant is then returned to no load conditions, with pressure and temperature changes controlled within normal limits.
MPS3 UFSAR3.9N-38Rev. 30 For design purposes this transient is assumed to occur 60 times during the 60 year design life of the plant.
8.Operating Basis Earthquake The operating basis earthquake is that earthquake which can reasonably be expected to occur during the plant life.
9.Excessive Feedwater Flow An excessive feedwater flow transient is conser vatively defined as an umbrella case to cover occurrence of several events of the same genera l nature. The postulated transient results from inadvertent opening of a feedwater control valve wh ile the plant is at the hot standby or no load condition, with the feedwater, condensate, and heater drain systems in operation.It is assumed that the stem of a feedwater cont rol valve fails and the valve immediately reaches the full open position. In the steam generator directly affect ed by the malfun ctioning valve
("failed loop"), the feedwater fl ow step increases from essentially zero flow to the value determined by the system resistance and the de veloped head of all operating feedwater pumps. Steam flow is assumed to remain at zero and the temperature of the feedwater entering the steam generator is conservatively assumed to be 32
&deg;F. Feedwater flow is isolated on a reactor coolant low T avg signal; a low pressurizer pressure signal actu ates the safety injection system. Auxiliary feedwater flow, initiated by the safety injection signal, is assumed, to c ontinue with all pumps discharging into the affected steam generator.
It is also assumed, for conservatism in the secondary side analysis, that auxiliary feedwater flows to the steam generators not affected by the malfunctioned valve, in the "unfai led loops". Plant conditions stabi lize at the values reached in 600 seconds, at which time auxiliar y feedwater flow is terminated.
The plant is then either taken to cold shutdown, or returned to the no load c ondition at a normal heatup rate with the auxiliary feedwater system under manual control.
For design purposes, this transien t is assumed to occur 30 times during the life of the plant.
10.RCS Cold Overpressurization RCS cold overpressurization can occur duri ng startup and shutdown conditions at low temperature, with or without the existence of a steam bubble in the pressurizer, and is especially severe when the reactor coolant system is in a water-solid configuration. The event is inadvertent, and usually generated by any one of a variety of malfunctions or operator errors. All events which have occurred to date may be categorized as belonging to either events resulting in the addition of mass (mass input transient), or events related to the addition of heat (heat input transient). All these possible transients are represented by composite "umbrella" design transients, referred to as RCS cold overpressurization.
For design purposes, this transient is assumed to occur 10 times during the 60 year design life of the plant.
MPS3 UFSAR3.9N-39Rev. 30Emergency Conditions The following primary system transients are considered emergency conditions:1.Small loss-of-coolant accident2.Small steam line break3.Complete loss of flow 1.Small loss of coolant accident For design transient purposes the small loss of coolant accident is defined as a break equivalent to the severance of a 1 inch inside diameter bran ch connection. (Liquid breaks are limited to less than 0.375 inch inside diameter and pressurizer steam space breaks are limited to 0.25 inch by the installation of an orifice. This ensures that the break can be handled by the normal makeup system and produce no significant fluid systems transients). Breaks which are much larger than one inch will cause accumulator injection soon after the accident and are regarded as faulted conditions.
For design purposes it is assumed that this transient occurs five times during the life of the plant.
It should be assumed that the safety injection system is actuated immediately after the break occurs and subsequently delivers wa ter at a minimum temperature of 32
&deg;F to the RCS.
2.Small Steam Line Break For design transient purposes, a small steam break is defined as a break equivalent in effect to a steam safety valve opening and remaining open. This transient is assumed to occur 5 times during the life of the plant. The follow ing 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.
3.Complete Loss of Flow This accident involves a complete loss of flow fr om full power resulting from simultaneous loss of power to all reactor coolant pumps. The conse quences of this incident are a reactor trip and turbine trip on undervoltage followed by automatic opening of the steam dump system. For design purposes this transient is assumed to occur five times during the plant lifetime.
MPS3 UFSAR3.9N-40Rev. 30 Faulted Conditions The following primary system transients are considered faulted conditions. Each of the following accidents should be eval uated for one occurrence:1.Reactor coolant pipe break (Large loss-of-coolant accident)2.Large steam line break3.Feedwater line break4.Reactor coolant pump locked rotor5.Control rod ejection6.Steam generator tube rupture7.Safe shutdown earthquake 1.Reactor Coolant Pipe Break (Large Loss-of-Coolant Accident)
Following postulated rupture of a reactor coolant pipe resulting in a large loss of coolant, the primary system pressure decreases causing the primary system temperature to decrease. Because of the rapid blowdown of coolant from the system and the comparatively large heat capacity of the metal sections of the components, it is likely that the metal will remain at or near the operating temperature by the end of the blow down. It is conservatively assumed that the safety injection system is actuated to introduce water at a minimum temperature of 32
&deg;F into the RCS. The safety injection signal will also result in reactor and turbine trips.
2.Large Steam Line Break The transient is based on the postulated complete severance of the largest steam line. The following 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.
3.Feedwater Line Break
 
MPS3 UFSAR3.9N-41Rev. 30 This accident involves the double-ended rupture of a main feedwater piping from full power, resulting in rapid blowdown of one steam generator and the termination of main feedwater flow to the others. The blowdown is completed in approximately 27 seconds. Conditions were conservatively chosen to give the most severe primary side and secondary side transients. All auxiliary feedwater flow exits at the break. The incident is terminated when the operator manually realigns the auxiliary feedwater sy stem to isolate the break and to deliver auxiliary feedwater to the intact steam generators.
4.Reactor Coolant Pump Locked Rotor This accident is based on the instantaneous seiz ure of a reactor coolant pump with the plant operating at full power. The locked rotor can oc cur in any loop. Reactor trip occurs almost immediately, as the result of low coolant flow in the affected loop.
5.Control Rod Ejection This accident is based on the single most reactiv e control rod being inst antaneously ejected from the core. This reactivity insertion in a particul ar region of the core causes a severe pressure increase in the RCS such that the pressurizer safety valves will lift and al so causes a more severe temperature transient in the loop associated with the affected region than in the other loops. For conservatism the analysis is based on the reacti vity insertion and does not include the mitigating effects (on the pressure transient) of coolan t blowdown through the hole in the vessel head vacated by the ejected rod.
6.Steam Generator Tube Rupture This accident postulates the double-ended rupture of a steam generator tube resulting in a decrease in pressurizer level and reactor coolan t pressure. Reactor trip will occur due to the resulting safety injection signal. In addition, safe ty injection actuation automatically isolates the feedwater lines, by tripping all feedwater pumps and closing the feedwater isolation valves. When this accident occurs, some of the reactor coolant blows down into the affected steam generator causing the shell side level to rise. The primary system pressure is reduced below the secondary safety valve setting. Subsequent recovery proce dures call for isolation of the steam line leading from the affected steam generator. The recovery actions may involve the use of unheated auxiliary spray into the pressurizer. Such an action will require override of the auxiliary spray valve "lockout."
7.Safe Shutdown Earthquake The safe shutdown earthquake is defined as the maximum vibratory ground motion which can reasonably be predicted from geologic and seismic evidence.Test Conditions The following primary system transien ts under test conditions are discussed:
MPS3 UFSAR3.9N-42Rev. 301.Primary Side Hydrostatic test2.Secondary Side Hydrostatic test 1.Primary Side Hydrostatic Test The pressure tests include both shop and field hydrostatic tests which occur as a result of component or system testing. This hydro test is performed at a water temperature which is compatible with reactor vessel material ductility requirements and a test pressure of 3,107 psig (1.25 times design pressure). In this test, the RC S is pressurized to 3,107 psig coincident with steam generator secondary side pr essure of 0 psig. The RCS is de signed for 10 cycles of these hydrostatic tests, which are performed prior to plant startup. The number of cycles is independent of other operating transients.
Additional hydrostatic tests may be performed to meet the inservice inspection requirements of ASME Section XI subarticle IS5-20. A total of f our such tests is expected. The increase in the fatigue usage factor caused by th ese tests is easily covered by th e conservative number (200) of primary side leakage tests that are considered for design.
2.Secondary Side Hydrostatic Test The secondary side of the steam generator is pressurized to 1.25 design pressure with a minimum water temperature of 120
&deg;F coincident with the primary side at 0 psig.
For design purposes it is assumed that the steam ge nerator will experience 10 cycles of this test.
These tests may be performed either prior to pl ant startup, or subseque ntly following shutdown for major repairs, or both.
3.9N.1.2 Computer Programs Used in Analyses The following computer programs have been us ed in the analysis of Seismic Category I components and equipment within the NSSS scope.1.ANSYS - General purpose finite element code used or problems of structural dynamic response in addition to static elas tic and inelastic analysis. The code is widely used in the public domain.
3.9N.1.3 Experimental Stress Analysis No experimental stress analys is methods are used for Cate gory I systems or components. However, for reactor internals, Westinghouse ma kes extensive use of m easured results from prototype plants and various scale mode l tests as discusse d in Sect ion 3.9N.2.
MPS3 UFSAR3.9N-43Rev. 30 3.9N.1.4 Considerations for the Evalua tion of the Faulted Condition 3.9N.1.4.1 Loading ConditionsThe structural evaluations performed on the reactor coolant system consider the loadings specified in Table 3.9N-2. These loads result from thermal expansion, pr essure, weight, operating basis earthquake (OBE), safe s hutdown earthquake (SSE), design basis loss-of-coolant accident, and plant operational thermal and pressure transients.
3.9N.1.4.2 Analysis of Primary Components Equipment which serves as part of the pressure boundary in the reactor coolant loop includes the steam generators, the reactor coolant pumps, the pressurizer, and the reactor vessel. This equipment is ANS Safety Class 1 and the pressure boundary meets the requirements of the ASME Boiler and Pressure Vessel Code , Section III, Subsection NB.
Loads are applied to the RCS components for all lo ading conditions on an "umbrella" load basis.
That is, on the basis of previous plant analyses , a set of loads are determined which should be larger than those seen in any single plant analys is. The umbrella loads represent a conservative means of allowing detailed component analysis pr ior to the completion of the system analysis. The results of the reactor coolant loop analysis are used to determine the actual loads acting on the equipment nozzles and the support/component interface locations. Upon completion of the system analysis, conformance is demonstrated between the actual plant loads and the loads used in the analyses of the components. Any deviations where the actual load is larger than the umbrella load are handled by indivi dualized analysis.Seismic analyses are performed individually for the reactor coolant pump, the pressurizer, and the steam generator. Detailed and complex dynamic models are used for the dynamic analyses. The response spectra corresponding to the building elevation at th e highest component/building attachment elevation are used for the component analysis. The reactor pressure vessel is qualified by static analysis based on loads de rived from dynamic system analysis.The pressure boundary portions of Class 1 valves in the RCS are designed and analyzed according to the requirements of NB-3500 of ASME III.Valves in sample lines connected to the RCS ar e not considered to be ANS Safety Class 1 nor ASME Class 1. This is because th e nozzles, where the lines connect to the primary system piping, are orificed to a 3/8-inch hole. This hole restri cts the flow such that lo ss through a severance of one of these lines can be made up by normal charging flow.
MPS3 UFSAR3.9N-44Rev. 30 3.9N.1.4.3 Dynamic Analysis of Reactor Pressure Vessel for Postulated Loss-of-Coolant Accident 3.9N.1.4.3.1 IntroductionThis section presents the method of computing the reactor pressure vessel response to a postulated loss-of-coolant accident (LOCA). The resulting stru ctural analysis considers loads on the reactor vessel resulting from the reactor coolant loop mechanical loads, internal hydraulic pressure transients, and reactor cavity pressurization (for postulate d breaks in the reactor coolant pipe at the vessel nozzles). The vessel is restrained by reactor vessel support pads and shoes beneath 4 of the reactor vessel nozzles, and th e reactor coolant loops with the primary supports of the steam generators and the reactor coolant pumps.
For the limiting break in the main loop piping, pi pe displacement restraints installed in the primary shield wall limit the brea k opening area of the vessel nozzl e pipe breaks to less than 144 square inches. This break area was determined to be an upper bound by using worst case vessel and pipe relative motions based on similar plant analyses. Detailed studies have shown that pipe breaks at the hot or cold leg reactor vessel nozzles, even with a limited break area, would give the highest reactor vessel support loads and the highe st vessel displacements, primarily due to the influence of reactor cavity pres surization. By considering these breaks, the most severe reactor vessel support loads are determin ed. In addition, two breaks outsid e the shield wall, for which there is no cavity pressurization, were also analyzed; specifically, the pump outlet nozzle pipe break and the steam generator inlet nozzle pipe break were considered.
Since the plant has been approved for the leak-before-break (LBB) analysis methodology, the LOCA analyses of the reactor vessel system are not required to include the dynamic effects of the main RCS pipe breaks, in accordance with GDC-4. The most limiting breaks to be considered are the branch line breaks. These break s consist of breaks in the accu mulator line in cold leg; the pressurizer surge line in the hot leg; and the residual heat removal (RHR) line in the hot leg. These three breaks were considered for the dynamic ef fects of the reactor vessel system. For the analyses of the branch line breaks, the reactor cavity pressuriza tion loads described in the next section are not utilized in the analysis.The results of the LOCA analysis of the reactor pressure vessel were determined using separate analyses. Westinghouse performed analyses to de termine the reactor vesse l displacement due to vessel internal hydraulic loads. RPV support and reactor coolant loop stiffnesses, supplied by SWEC, were included in the RPV structural model. The results of the Westinghouse analyses were dynamic displacement data for the RPV.SWEC determined the reactor vessel displace ments for the remaining LOCA loads. By combining these displacements with the displacements from the Westinghouse analyses, the total combined RPV displacements were determin ed for the postulated pipe ruptures.
MPS3 UFSAR3.9N-45Rev. 30 3.9N.1.4.3.2 Loading Conditions Following a postulated pipe rupture at the reactor vessel nozzle, the reactor vessel is excited by time-history forces. As previously mentioned, these forces are the combined effect of three phenomena:1.Reactor Coolant Loop Mechanical LoadsThe 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 rel eased 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 br oken pipe resulting in horizontal forces applied to the reactor vessel. Smaller vert ical forces arising from pressure on the bottom of the vessel and the vessel flanges are also a pplied to the reactor vessel.The internals reaction forces develop from asymmetric pressure distributions inside the reactor vessel.
For a vessel inlet nozzle brea k, the depressurization wave path is through the broken loop inlet no zzle and into the downcomer annulus between the core barrel and reactor ve ssel. 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 generato r inlet nozzle break, the wave passes through the RPV outlet nozzle and direct ly into the upper internals region, depressurizes the core, and enters the dow ncomer annulus from the bottom of the vessel. Thus, for an outlet nozzle break, the downcomer annulus is depressurized with much smaller difference s 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 reflect ion and translation through the reactor vessel fluid but the initial depressuriza tion wave has the greatest effect on the loads.3.Reactor Internal Hydraulic Forces The reactor internals hydrauli c 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-el astic coupling or fluid-structure interaction. The hydraulic analysis considers the fluid-structure interaction of the core barrel by account ing for the deflections of constraining boundaries which are repres ented by masses and springs. The dynamic response of the core barrel in its beam bendi ng mode responding to blowdown forces compensates for internal pressure variat ion by increasing the volume of the more MPS3 UFSAR3.9N-46Rev. 30 highly pressurized regions. The analytical methods used to develop the reactor internals hydraulics are described in WCAP-8708.
3.9N.1.4.3.3 Reactor Vessel and Internals Modeling The mathematical model of the RPV is a three-dimensional nonliner fini te element model which represents the dynamic characteristics of the reactor vessel and its internals in the six geometric degrees of freedom. The RPV th ree-dimensional nonlinear finite element model is shown in Figure 3.9N-1. The model consis ts of three concentric stru ctural submodels connected by nonlinear impact elements and stiffness matrices, which is conn ected to a submodel of the CRDMs and CRDM seismic platform, tie rods, and lifting legs. The first submodel, represents the reactor vessel shell and associated components. The reactor vessel is restrained by four reactor vessel supports (situated beneat h alternate nozzles) and by the at tached primary coolant piping.
Each reactor vessel support is modeled by a linear horizontal stiffness and a vertical impact element. The attached piping is represented by a stiffness matrix.
The second submodel, represents the reactor core barrel (RCB), neutron panels, lower support plate, tie plates, and secondary core support co mponents. This submodel is physically located inside the first, and is connected to it by a stiffness matrix at the internals support ledge. Core barrel to vessel shell impact is represented by nonlinear elements at the core barrel flange, core barrel nozzle, and lower radial support locations.
The third and innermost submodel, represents the upper support plate, guide tubes, support columns, upper and lower core plates, and fuel. Th e third submodel is connected to the first and second by stiffness matrices and nonlinear elements.
Fluid-structure or hydro-elastic in teraction is included in the re actor pressure vessel model for seismic evaluation. The horizontal hy dro-elastic interaction is signi ficant in the cylindrical fluid flow region between the core barrel and reactor ve ssel (the downcomer). Mass matrices with off-diagonal terms (horizontal degrees-of-freedom only) atta ch between nodes on th e core barrel and reactor vessel shell.The diagonal terms of the mass matrix are similar to the lumping of water mass to the vessel shell and core barrel. The off-diagonal terms reflect the fact that all the water does not participate when there is no relative motion of the vessel and co re barrel. It should be pointed out that the hydrodynamic mass matrix has no artificial virtual mass effect and is derived in a straight-forward, quantitative manner.
The matrices are a function of the properties of two cylinders with a fluid in the cylindrical annulus, specifically; inside and outside radius of the annulus, density of the fluid and length of the cylinders. Vertical segmenta tion of the RCB allows inclusi on of radii variations along the RCB height and approximates the effects of RCB beam deformati on. These mass matrices were inserted between selected nodes on the core barrel and reactor vessel shell. In the finite element approach, the structure is divided into a finite number of members or elements. Nodal disp lacements and impact forces are stored for post-processing.
MPS3 UFSAR3.9N-47Rev. 30The reactor vessel is restrained by the four attached reactor coolant loops with the steam generator and reactor coolant pump primary supports and the four reactor vessel s upports, situated beneath alternate reactor vessel nozzles. The RPV support model (supplied by SWEC) is shown in Figure 3.9N-3. All support spring elements are linear, doubl e-acting elements and represent the restraint provided by the attached piping and the vessel supports, includi ng the primary shield wall.
3.9N.1.4.3.4 Analytical MethodsThe time history effects of the internals hydraulic loads are applied to the appropriate nodes of the mathematical model of the reactor vessel and internals. The analysis is performed by numerically integrating the differential equations of motion to obtain the transient response. The output of the analysis is the time history displacements of the reactor vessel system. The output from the analysis is used for deta iled component evaluation.
3.9N.1.4.4 Stress Criteria for Reactor Coolant System Components All RCS components are designed and analyzed for the design, normal, upset, and emergency conditions to the rules and requirements of the ASME Code Section III. The analysis methods and associated stress or load allowable limits used in evaluation of faulted conditions are those that are defined in Appendix F of the ASME Code.Loading combinations and allowable stresses for RCS components are given in Tables 3.9N-2 and 3.9N-3, respectively. For faulted condition evaluations, the effects of the safe shutdown earthquake (SSE) and loss-of-coolant accident (LOCA) are combined using the square root of the sum of the square s (SRSS) method.
3.9N.2 DYNAMIC TESTING AND ANALYSIS 3.9N.2.1 Preoperational Vibration and Dynamic Effects Testing on Piping See Section 3.9B.2.1.
3.9N.2.2 Seismic Qualification Testing of Sa fety Related Mechanical Equipment See Section 3.10.
3.9N.2.3 Dynamic Response Analysis of Reactor Internals Under Operational Flow Transients and Steady State Conditions The vibration characteristics and behavior due to flow induced excitation are very complex and not readily ascertained by analytical means alone. Reactor components are excited by the flowing coolant which causes oscillatory pressures on the surfaces. The integration of these pressures over the applied area should provide th e forcing functions to be used in the dynamic analysis of the structures. In view of the comple xity of the geometries and the random character of the pressure oscillations, a closed form solu tion of the vibratory problem by integration of the differential equation of motion is not always practical and realistic. The determination of the forcing MPS3 UFSAR3.9N-48Rev. 30 functions as a direct correlat ion of pressure oscillations cannot be practi cally performed independent of the dynamic characteristics of the structure. The main objective is to establish the characteristics of the forcing functions that esse ntially determine the response of the structures.
By studying the dynamic properties of the structure from previous analytical and experimental work, the characteristics of the forcing function can be deduced. Th ese studies indicate that the most important forcing functions are flow turbulence, and pump-related excitation. The relevance of such excitations depends on many factors such as type and location of component and flow conditions. The effects of these forcing functions have been studied from test runs on models, prototype plants and in component tests, (WCAP-8303-P-A, WCAP-8517; Trojan Final Safety Analysis Report, Appendix A-12; WCAP 8780, WCAP-9945).
The Indian Point No. 2 plant (Docket No. 50.247) has been established as the prototype for a four-loop plant internals verifica tion program and was fully instru mented and tested during hot functional testing. In addition, the Trojan plant (Docket No. 50-344) has provided, and the Sequoyah No. 1 plant (Docket No. 50-327) has provi ded, prototype data applicable to Millstone 3, (WCAP-8517; Trojan Final Safety Anal ysis Report, Appendix A-12; WCAP-9945).Millstone 3 is similar to Indian Point No. 2; th e only significant differences are the modifications resulting from the use of 17 x 17 fuel, replacement of the annular thermal shield with neutron shielding panels, and the change to the UHI-style inverted top hat support structure configuration. These differences are addressed below.1.17 x 17 fuelThe only structural changes in the intern als 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 ri gid, 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 inte rnal vibration is expected from the vibration with the 15 x 15 fuel asse mblies vibration characteristics.2.Neutron shield panel lower internals The primary cause of core barrel excita tion 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 info rmation 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 stre sses approximately equal to MPS3 UFSAR3.9N-49Rev. 30 those of Indian Point No. 2 result on Mill stone 3 internals with the attendant large safety margins.3.UHI-style inverted top ha t upper support configuration The components of the upper in ternals are excited by turbul ent forces due to axial and crossflows in the upper plenum (W CAP-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.
The original test and analysis of the four-loop configuration is augmented by WCAP-8317-A; WCAP-8517; Trojan Final Safety Analys is Report, Appendix A-12, WCAP-8780; and WCAP-9945, supported by scale model and analytical work to cover the effects of successive hardware modifications.
3.9N.2.4 Preoperational Flow-Induced Vibration Testing of Reactor Internals Because the Millstone 3 reactor internals design configuration is well-c haracterized, as was discussed in Section 3.9N.2.3, it is not considered necessary to c onduct instrumented tests of the Millstone Unit 3 hardware. The recommenda tions of Regulatory Guide 1.20 are met by conducting the confirmatory preoperational testi ng examination for integr ity per Section D, of Regulatory Guide 1.20, Regulations for Reactor In ternals Similar to the Prototype Design. This examination will include some 30 points w ith special emphasis on the following areas:1.All major load-bearing elements of the re actor internals relied upon to retain the core structure in place2.The lateral, vertical and torsional restraints provided within the vessel3.Those locking and bolting devices whose failure could adversely affect the structural integrity of the internals4.Those other locations on the reactor in ternal components which are similar to those examined on the prototype designs5.The inside of the vessel will be inspected before and after th e hot functional test, with all the internals removed, to verify that no loose parts or foreign materials are in evidence.
MPS3 UFSAR3.9N-50Rev. 30 A particularly close inspection is made of th e following items or areas using a 5X or 10X magnifying glass or othe r appropriate inspection.1.Lower internalsa.Upper barrel to flange girth weldb.Upper barrel to lower barrel girth weldc.Upper core plate aligning pin. Exa mine bearing surfaces for any shadow marks, burnishing, buffing or scoring. Inspect welds for integrity.d.Irradiation specimen guide screw lo cking devices and dowel pins. Check for lockweld integrity.e.Baffle assembly locking devices. Check for lockweld integrity.f.Lower barrel to core support girth weldg.Neutron shield panel screw locki ng devices and dowel pin lockwelds.
Examine the interface surfaces for evidence of tightness. Check for lockweld integrityh.Radial support key weldsi.Insert screw locking devices. Examine soundness of lockwelds.j.Core support columns and instrumentat ion guide tubes. Check the joints for tightness and soundness of the locking devices.k.Secondary core support a ssembly weld integrityl.Lower radial support keys and inse rts. 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 ke y and keyway will show burnishing, buffing or shadow marks which indicat e pressure loading and relative motion between these parts. Minor sc oring of engaging surfaces is also possible and acceptable.m.Gaps and baffle joints. Check gaps between baffle-to-baffle joints.2.Upper internals MPS3 UFSAR3.9N-51Rev. 30a.Thermocouple conduits, clamps, and couplingsb.Guide tube, support column, orifice plate, and thermocouple assembly locking devicesc.Support column and thermocouple conduit assembly clamp weldsd.Upper core plate alignment inserts. Examine bearing surfaces for shadow marks, burnishing, buffing or scori ng. Check the locking devices for integrity of lockwelds.e.Thermocouple conduit fitting locktab and clamp weldsf.Guide tube enclosure and card welds Acceptance standards are the same as required in the shop by the original design drawings and specifications.During the hot functional test, the internals will be subjected to a total operating time greater than normal full-flow conditions (four pumps operating) of at least 240 hours. This provides a cyclic loading of approximately 10 7 cycles on the main structural elem ents of the internals. In addition there will be some operating time with only one, two and three pumps operating.
Pre- and post-hot functional inspect ion results serve to confirm predictions that the internals are adequately designed for flow induced vibrations. When no signs of abnormal wear, no harmful vibrations are detected and no a pparent structural changes take place, the four-loop core support structures are considered to be structurally adequate and sound for operations.
3.9N.2.5 Dynamic System Analysis of the Reac tor Internals Under Faulted Conditions The reactor internals analysis under faulted events consid ers the following conditions:*LOCA*SSE (Safe Shutdown Earthquake)
The criteria for acceptability in regard to mechan ical integrity analyses are that adequate core cooling and core shutdown must be assured. This implies that the deformation of the reactor internals must be sufficiently small so that ge ometry remains substantially intact. Consequently, the limitations established for the internals are c oncerned with the deflecti ons and stability of the parts in addition to stress criteria to assure integrity of the components.The MULTIFLEX digital com puter program (WCAP-8708 and WCAP-8709) which was developed for the purpose of calculating local fluid pressure, flow, and density transients that occur in pressurized water reactor coolant system s during a loss-of-coolant accident is applied to the subcooled, transition, and saturated two-phase blowdown regimes. This is in contrast to MPS3 UFSAR3.9N-52Rev. 30 programs such as WHAM (Fabic 1967) which are applicable only to the subcooled region and which, due to their method of solution, could not be extended into the region in which large changes in the sonic velocities and fluid densities take place. This blowdown code is based on the method of characteristics wherei n the resulting set of ordinary differential equations, obtained from the laws of conservation of mass, momentum, and energy, are solved numerically using a fixed mesh in both space and time.
Although spatially one-dimensional conservation laws are employed, the code can be applied to describe three-dimensional system geometries by use of the equi valent piping networks. Such piping networks may contain any num ber of pipes of channels of various diameters, dead ends, branches (with up to 6 pipes connected to each branch), contract ions, expansions, orifices, pumps and free surfaces (such as in the pressurizer).
System losses such as friction, contraction, expansion, etc are considered.The MULTIFLEX code evaluates the pressure a nd velocity transients at various locations throughout the system. These pressure and velocity tr ansients are stored as a permanent tape file and are made available to the programs LATFORC and FORCE2 which utilize a detailed geometric description in evaluating the loadings on the reactor internals. The LATFORC computer code calculates the lateral hydraulic load s on the reactor vessel wall and core barrel, while the FORCE2 code calculates the vertical hydraulic loads on the reactor vessel internals.Each reactor component for which LATFORC and FORCE2 calculations are performed is designated as an element and assigned an element number. Forces acting upon each of the elements are calculated summing the effects of:1.The pressure differential across the element2.Flow stagnation on, and unrecovere d orifice losses across the element3.Friction losses along the element Input to the codes, in addition to the blowdown pressure and velocity transients, includes the effective area of each element on which the force acts due to the pressure differential across the element, a coefficient to account for flow stagnation and unrecovere d orifice losses, and the total area of the element along which the shear forces act.The reactor internals analysis has been performe d using conservative as sumptions. Some of the more significant assumptions are:1.The mechanical and hydraulic analyses have considered the effect of hydroelasticity.2.The reactor internals are represented by concentric pi pes, 3-D beams and a multi-mass system connected with springs and dashpots simulating the elastic response and the viscous dampi ng of the components.
MPS3 UFSAR3.9N-53Rev. 30 The model described is considered to have a sufficient number of degrees of freedom to represent the most important modes of vibration in the horizontal and vertical directions.
The pressure waves generated within the reacto r are highly dependent on the location and nature of the postulated pipe failure. In general, the more rapid the severance of the pipe, the more severe the imposed loadings on the components. A one millisecond severance time is taken as the limiting case.
In the case of the hot leg break, the vertical hydraulic forces produce an initial upward lift of the core. A rarefaction wave propagates through the reactor hot leg nozzle into the interior of the upper core barrel. Since the wave has not reached the flow annulus on the outside of the barrel, the upper barrel is subjected to an impulsive compressive wave. Th us, dynamic instability (buckling) or large deflections of the upper core barrel, or bot h, are possible responses of the barrel during hot leg break results in transverse loading on the upper core components as the fluid exits the hot leg nozzle.In the case of the cold leg break, a rarefaction wave pr opagates along a reactor inlet pipe, arriving first at the core barrel at the inlet nozzle of the broken loop. The upper barrel is then subjected to a non-axisymmetric expansion radial impulse which changes as the rarefaction wave propagates both around the barrel and down th e outer flow annulus between ve ssel and barrel. After the cold leg break, the initial steady state hydraulic lift forces (upward) decreas e rapidly (within a few milliseconds) and then increase in the downwar d direction. These cause the reactor core and lower support structure to move initially downward.If a simultaneous seismic event with the intensity of the SSE is postulated with the loss-of-coolant accident, the imposed loading on the internals co mponent may be additive in certain cases and therefore the combined loading must be considered. In general, however, the loading imposed by the earthquake is small compared to the blowdown loading.The summary of the mechanical analysis follows:Transverse Excitati on Model for Blowdown Various reactor internal components are subject ed to transverse ex citation during blowdown. Specifically, the barrel, guide tubes, and upper s upport columns are analyzed to determine their response to this excitation.
Core Barrel - For the hydraulic an alysis of the pressure transi ents during hot leg blowdown, the maximum pressure drop across the barrel is a uniform radial compressive impulse.
The barrel is then analyzed for dynamic buckli ng using the following c onservative assumptions:1.The effect of the fluid environment is neglected.2.The shell is treated as simply supported.
MPS3 UFSAR3.9N-54Rev. 30 During cold leg blowdown, the barrel is subject ed to a non-axisymmetric expansion radial impulse which changes as the rarefaction wave propagates both around the barrel and down the outer flow annulus between vessel and barrel.
The 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 sh ield panels, is analyzed as a beam supported at the top and supported at bottom the lower radial support and the dynamic response is obtained.Guide Tubes - The guide tubes in closest proximity to the outlet nozzle of the ruptured loop are the most severely loaded during a blowdown. The transverse guide tube forces decrease with increasing distance from th e ruptured nozzle location.
All of the guide tubes are designed to maintain the function of the c ontrol rods for a break size of 144 in 2 and smaller. No credit for the function of the control rods is assumed for break size areas above 144 in
: 2. However, the design of the guide tube wi ll permit control rod operation in all but four control rod positions, which is sufficient to maintain the core in a subcritical configuration, for break sizes up to a double-ende d hot leg break. This double-e nded hot leg break imposes the limiting lateral guide tube loading.
Upper Support Columns - Upper support columns lo cated close to the broken nozzle during hot leg break will be subjected to transverse loads due to cross flow. The loads applied to the columns are computed with a method similar to the one used for the guide tubes, i.e., by taking into consideration the increase in flow across the column during the accident. The columns are studied as beams with variable section and the resulting stresses are obtained using the reduced section modulus and appropriate stress ri sers for the various sections.The effects of the gaps that could exist between vessel and barrel, between fuel assemblies, and between fuel assemblies and baffle plates are considered in the analysis. The stresses due to the safe shutdown earthquake (vertica l and horizontal components) were combined in the most unfavorable manner with the blowdow n stresses in order to obtain the largest principal stress and deflection.All reactor internals components were found to be within acceptable stress and deflection limits for both hot leg and cold leg loss of coolant acc idents occurring simultaneously with the safe shutdown earthquake.These results indicate that the ma ximum deflections and stress in the critical structures are below the established allowable limits. For the transver se excitation, it is shown that the upper barrel does not buckle during a hot leg break and that it has an allowable stress distribution during a cold leg break.
MPS3 UFSAR3.9N-55Rev. 30 Even though control rod insertion is not required for plant shutdown, this analysis shows that most of the guide tubes will deform within the limits established experimentally to assure control rod insertion. For the guide tubes deflected above the no loss of function limit, it must be assumed that the rods will not drop. However, the core will still shut down due to the negative reactivity insertion in the form of core voiding. Shutdown will be aided by the great majority of rods that do drop. Seismic deflections of the guide tubes are generally neglig ible by comparison with the no loss of function limit.
3.9N.2.6 Correlations of Reactor Internals Vibration Tests with the Analytical Results As stated in Section 3.9N.2.3, it is not considered necessary to c onduct instrumented tests of the reactor vessel internals. WCAP-8516-P, WCAP-8517 (1975) and the Trojan Final Safety Analysis Report, Appendix A12 describe pr edicted vibration behavior base d on studies performed prior to the plant tests. These studies, wh ich utilize analytical models, scale model test results, component tests, and results of previous plant tests, are used to characterize the forcing functions and establish component struct ural characteristics so that the fl ow induced vibratory behavior and response levels for reactor in ternals are estimated. These es timates are supported by values deduced from plant test data obtained from the Sequoyah and Trojan internals vibration measurement programs. Adequacy of the Millstone 3 internals will be verified by the use of Sequoyah and Trojan results supported by sc ale model tests and analytical work.
3.9N.3 ASME CODE CLASS 1, 2, AND 3 COMPONENTS, COMPONENT SUPPORTS, AND CORE SUPPORT STRUCTURESThe ASME Code Class components are constructed in accordance with the ASME Code, Section III.A detailed discussion of ASME Code Class 1 components is provided in Section 3.9N.1. For core support structures, design load ing conditions are given in Se ction 3.9N.2.3. Loading conditions are discussed in Section 4.2.
Method of analysis and testing for core suppor t structures are discus sed in Sections 3.9N.2.3, 3.9N.2.5; and 3.9N.2.6.
3.9N.3.1 Loading Combinations, Design Transients, and Stress Limits for Class 2 Components 3.9N.3.1.1 Design Loading Combinations The design loading combinations for ASME Code Class 2 and 3 components and supports are given in Table 3.9N-4.
3.9N.3.1.2 Design Stress Limits The design stress limits established for the components are sufficiently low to assure that violation of the pressure retaining boundary will not occur. These limits, for each of the loading combinations, are component oriented and are presented in Tables 3.9N-5 through 3.9N-10 for MPS3 UFSAR3.9N-56Rev. 30 vessels, pumps and valves. Active
* pumps and valves are furthe r discussed in Section 3.9N.3.2.
The design of component supports is discussed in Section 3.9N.3.4.
3.9N.3.2 Pumps and Valve Operability AssuranceEquipment for Millstone 3 is designed to comply with the intent of Regulatory Guide 1.48; i.e., it is designed/analyzed to ensure structural integrity and operability. However, the load combinations and stress limits that were used reflect NRC requirements when the components were purchased and subsequently designed. The codes and procedures which were available when the components were purchased ar e based on conservative design requirements. These codes and procedures have been used by th e nuclear industry for the design of components that are installed in plants that are presently operating. A list of active pumps and valves is presented in Tables 3.9N-11 and 3.9N-12, respectively.
3.9N.3.2.1 ASME Code Class PumpsActive pumps were designed in accordance with Section III of the ASME Code. The stress levels in the pumps do not exceed those provided in Ta bles 3.9N-7 and 3.9N-10. Forces resulting from seismic accelerations in the horizontal and vertical directions are included in the analysis of the pumps and their supports. To eliminate any amplification of the seismic floor accelerations in the pump support structure, the supports were designed to have natural frequencies in excess of 33 Hz.The pumps are subjected to a series of tests prior to installation and after installation in the plant. In-shop tests include hydrostatic tests to 150 percent of th e design pressure, seal leakage tests, net positive suction head (NPSH) test s to qualify the pumps for th e minimum available NPSH, and functional performance tests. For the NPSH a nd functional performance tests, the pumps are placed in a test loop and subjected to operating conditions. After installation, the pumps undergo cold hydrostatic tests, hot func tional tests to verify operation, and periodic inservice inspection and operation.
3.9N.3.2.2 Valve Operability Safety related valves are subjected to a series of tests prior to service and during the plant life.
Prior to installation, the following tests are pe rformed: shell hydrostati c test to ASME Code, Section III requirements, backseat and main s eat leakage tests, disc hydrostatic test, and operational tests to verify that the valve will ope n and close. Qualification of motor operators for environmental conditions is discussed in Sections 3.11 and 1.8 (in the discussion of Regulatory Guide 1.73, Qualification Tests of Electric Valve Op erators Installed Inside the Containment of Nuclear Power Plants). Cold hydrostatic tests, hot functional tests, periodic inservice inspections, and 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.
MPS3 UFSAR3.9N-57Rev. 30These tests guarantee reliability of the valve fo r the design life of the plant. The valves are constructed in accordance with the ASME Code, Section III. The stress limits used for active Class 2 and 3 valves are shown in Table 3.9N-8. Active valves are designed to have a first natural frequency which is equal to or greater than 33 he rtz and an analysis of the extended structure is performed for static equivalent safe shutdown earthquake loads applied at the center of gravity of the extended structure.
In addition to these tests and analyses, representative valves of each design type are tested during a simulated plant faulted condition event by demons trating operational capabilities within the specified limits. These tests veri fy the analysis methods describe d above. The testing procedures are described below.The valve is mounted in a manner which conservatively represents typical valve installations. The valve includes the operator, solenoid valves, and limit switches when such are normally attached to the valve in service. The opera bility of the valve dur ing a faulted condition is demonstrated by satisfying the following criteria:1.The actuator and yoke of the valve are st atically 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 di rection 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 deflect ed 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 swit ches, and solenoid valves necessary for operation are qualified in accordance with IEEE Standard 344-1975.
The piping designer must limit the inputs to the valve to those levels for which the valve is qualified.
The above operability program description applies to valves with extended structures. Valves which are safety related but can be classified as not having an ex tended structure, such as swing check valves and safety valves, are considered separately. Valves are qualified by analysis and testing verifies the method of analysis.Pressurizer safety valves are qualified for operability in the same manner as valves with extended structures as described above. Th e qualification methods include analysis of the bonnet for static equivalent safe shutdown earthquake loads, in shop hydrostatic and seat leakage tests, and periodic in situ valve inspection. Additionally, repres entative pressurizer safety valves are tested to verify 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.
MPS3 UFSAR3.9N-58Rev. 303.A static load representing the faulted condition seismic load is applied to the top of the valve bonnet in the direction of the weakest axis of the extended structure.4.The pressure is increased until the valve actuates.5.Actuation of the valve at its setpoint ensures its operability during the faulted condition.
Using these methods, all the safety related valves in the systems are qualified for operability during a faulted event. These methods outlined a bove conservatively simulate the seismic event and ensure that the active valves will perform their safety related function when necessary.
3.9N.3.2.3 Pump Motor and Valve Operator Qualification Active pump motors and active valve motor opera tors (and limit switches and solenoid valves) are seismically qualified in accordance with IEEE Standard 344-1975.
Qualification is accomplished by analysis, testi ng, or by a combination of analysis and testing.
When analysis is used, such methods can be justified by:1.Demonstrating that equipment being qualified is amenable to analysis, and2.That the analysis can be correlated wi th test or be performed using standard analysis techniques.
3.9N.3.3 Design and Installation Details for M ounting of Pressure-Relieving Devices Refer to Section 3.9B.3.3 3.9N.3.4 Component SupportsThe criteria for Westinghouse supplied supports for ASME Code Class 1 mechanical equipment is presented in Section 3.9N.1.The criteria for Westinghouse supplied supports for ASME Code Class 2 and 3 mechanical equipment is presented in Section 3.9N.3.4.1.
3.9N.3.4.1 Component Supports for Tanks and Heat Exchangers Component supports for Class 2 and 3 tanks and heat exchangers are of two types: linear and, for the most part, plate and shell type supports. The su pports for these tanks a nd heat exchangers are designed and analyzed to the rules and requirements of Subsection NF of Section III of the ASME Code. The design analyses and associated stress or load allowable limits for faulted conditions are those defined in Appendix F of the ASME Code.
The only exception to the above is the supports for the volume control tank which, because the procurement date of the tank predates Subsection NF, are designed to the requi rements of the AISC Code.
MPS3 UFSAR3.9N-59Rev. 30 3.9N.3.4.2 Component Supports for Pumps Component supports for Class 2 and 3 auxiliary pum ps are of two types: plate and shell and, for the most part, linear type s upports. These supports, w ith the exception of the supports for the charging and safety injection pumps, are de signed by the pump manufacturer to pressure boundary stress limits, but in no case is yield stress of the material exceeded. The supports for the charging and safety injection pumps meet the requirements of Subsection NF of Section III of the ASME Code.
3.9N.4 CONTROL ROD DRIVE SYSTEM (CRDS) 3.9N.4.1 Descriptive Information of CRDS Control Rod Drive Mechanism Control rod drive mechanisms are located on the dome of the reactor vessel head. They are coupled to rod cluster control assemblies (RCCAs) which have neutron absorber material over the entire length of the control rods and derive their name from this feature.
The control rod drive mechanism is shown in Figure 3.9N-4 and schematically in Figure 3.9N-5.The primary function of the control rod drive mechanism is to insert, withdraw or hold stationary, RCCAs within the core to control average core temperature and to shutdown the reactor.
The control rod drive mechanism is a magneti cally-operated jack. A magnetic jack is an arrangement of three electromagnets which are energized in a controlled sequence by a power cycler to insert or withdraw rod cluster control assemblies in the reactor core in discrete steps.
Rapid insertion of the rod cluster control assemblies occurs when electrical power is interrupted.
The control rod drive mechanism c onsists of four separate subasse mblies. They are the pressure vessel, 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 pr essure 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:
MPS3 UFSAR3.9N-60Rev. 301.The stationary gripper coil2.The movable gripper coil3.The lift coil The coil stack assembly is a separate uni t which is installed on the control rod drive mechanism by sliding it over the outsi de of the latch housing. It rests on the base of the latch housing wit hout 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 t ube, stationary pole pieces, movable pole pieces, and two sets of latches:1.The movable gripper latches2.The stationary gripper latchesThe latches engage grooves in the drive rod assembly. The movable gripper latches are moved up or down in 5/8-inch steps by th e 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 repositi oned 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 an d provides the means for coupling to the RCCA.The disconnect button, disconnect rod assembly, a nd locking button provide positive locking of the coupling to the RCCA and permits remote disconnection of the drive rod assembly.The control rod drive mechanism can be tripped during any part of the power cycler sequencing if electrical power to the coils is interrupted, thereby rel easing the drive rod assembly and inserting the RCCA.The control rod drive mechanism is threaded a nd seal welded on a head adaptor on top of the reactor vessel head. The drive rod assembly is coupled to the rod cluster control assembly directly below.
MPS3 UFSAR3.9N-61Rev. 30The mechanism is capable of ra ising or lowering a 360 pound load, (which includes the drive rod assembly weight) at a rate of 45 inches/minute. Withdrawal of the RCCA is accomplished by energizing the magnetic coils, and insertion is by gravity.The mechanism internals are designed to operate in 650
&deg;F reactor coolant. The pressure vessel assembly is designed to retain reactor coolant at 650
&deg;F and 2,500 psia. The three operating coils are designed to operate at an internal coil temperature of 392
&deg;F with forced air cooling required to maintain that temperature.
The control rod drive mechanism shown schematically in Figure 3.9N-5 withdraws and inserts an RCCA as shaped electrical pulses are received by the operating coils. An ON or OFF sequence, repeated by silicon controlled rectifiers in the power programmer, causes either withdrawal or insertion of the control rod. Position of the cont rol rod is measured by 42 discrete coils mounted on the rod position indicator coil stack assembly surrounding the rod travel housing. Each coil magnetically senses the entry and presence of the top of the ferromagnetic drive rod assembly as it moves through the coil center line.
During plant operation the stationary gripper coil of the control rod drive mechanism holds the rod cluster control assembly in a static position until a stepping sequence is initiated at which time the movable gripper coil and lift coil are energized sequentially.Drive Rod Assembly - RCCA Withdrawal The drive rod assembly along with a coupled RCCA is withdrawn by repetit ion of the following sequence of events (Figure 3.9N-5).1.The drive rod assembly is held in a st ationary 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 0.05 9-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 latche s and plunger to move downward 0.059 inch until the load of the drive rod assembly an d 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 MPS3 UFSAR3.9N-62Rev. 30The 5/8-inch gap between th e movable gripper pole and the lift pole closes and the drive rod assembly with attached RCCA raises one step length (5/8 inch).5.Stationary Gripper Coil - ON The plunger raises and closes the gap belo w the stationary gripper pole. The three links, pinned to the plunger, swing the stat ionary 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 st ationary 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 pol e and lift pole op ens. 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 thr ough 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 The sequence for RCCA insertion is similar to that for control r od withdrawal, except the timing of lift coil ON and OFF is changed to permit lowering the RCCA. The sequence begins with the SG energized in hold mode.1.Lift Coil - ON The 5/8-inch gap between the movable gr ipper 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 MPS3 UFSAR3.9N-63Rev. 30 The latch locking plunger raises and swi ngs the movable gripper latches into a drive rod assembly groove. A nominal 0.05 9-inch axial clearance exists between the latch tips and the drive rod assembly.3.Stationary Gripper Coil - OFFThe force of gravity, acting upon the dr ive rod assembly and attached RCCA, causes the stationary gripper latche s 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 sepa rates the movable gr ipper 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 belo w the stationary gripper pole. The three links, pinned to the plunger, swing the three stationary gripper la tches 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 st ationary 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 inse rtion rate of 45 inches per minute.Holding and Tripping of the Control Rods During most of the plant operating time, the control rod drive mechanisms hold the RCCA's withdrawn from the core in a static position. In the holding mode, only one coil, the stationary gripper coil, is energized on each mechanism. The drive rod assembly and attached RCCA's hang suspended from the three latches.
If power to the stationary gripper coil is cut of f, the stationary gripper return spring combined with the weight of the drive rod assembly and the RCCA is sufficie nt to move latches out of the MPS3 UFSAR3.9N-64Rev. 30 drive rod assembly groove. Follow ing a power interruption to the coil causing the magnetic field to collapse, the control rod falls by gravity into the core.
3.9N.4.2 Applicable CRDS Design Specifications For those components in the cont rol rod drive system comprising po rtions of the reactor coolant pressure boundary, conformance with the Gene ral Design Criteria a nd 10CFR50, Section 50.55a is discussed in Sections 3.1 and 5.2 conformance with regulatory guides pertaining in Section 4.5 and 5.2.3 and Section 1.8.
Design Bases Bases for temperature, stress and structural me mbers, and material compatibility are imposed on the design of the reactiv ity control components.Design Transient and Loading Combinations The CRDS is designed to withstand stresses origina ting from various operat ing design transients summarized in Table 3.9N-1. Structural eval uation performed on CRDM pressure retaining components consider the loading combinations specified in Table 3.9N-2.Allowable Stresses Allowable stresses for CRDM pressure retaining components are given in Table 3.9N-3.
Dynamic Analysis The cyclic stresses due to dynamic loads and defl ections are combined with the stresses imposed by loads from component weights, hydraulic forces and thermal grad ients for the determination of the total stresses of the CRDS.
Control Rod Drive Mechanisms The control rod drive mechanism (CRDM) pressure housings are Class 1 components designed to meet the stress requirements for normal operating co nditions of Section III of the ASME Boiler and Pressure Vessel Code. Both static and alternating stress intensities are considered. The stresses originating from the required design transients are included in the analysis.
A dynamic seismic analysis is required on the CRDM's when a seismic disturbance has been postulated to confirm the abilit y of the pressure housing to meet ASME Code, Section III allowable stresses and to confirm its ability to trip when subj ected to the seismic disturbance.Control Rod Drive Mechanism Operational Requirements The basic operational requirements for the CRDM's are:
MPS3 UFSAR3.9N-65Rev. 301.5/8-inch step2.144-inch travel (nominal)3.360 pound maximum load4.Step in or out of 45 in ches/minute (72 steps/minute)5.Electrical power interruption shall init iate release of drive rod assembly/RCCA6.Trip delay time of less than 150 millis econds 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
&deg;F to 550&deg;F7.60 year design life with normal refurbishment 3.9N.4.3 Design Loads, Stress Limits, and Allowable Deformations 3.9N.4.3.1 Pressure Vessel Assembly The pressure retaining components are analy zed for loads corresponding to normal, upset, emergency and faulted conditions. The analysis performed depends on the mode of operation under consideration.The scope of the analysis requires many differ ent techniques and methods, both static and dynamic.Some of the loads that are considered on each component where applicable are as follows:1.Control rod trip (equivalent static load)2.Differential Pressure3.Spring preloads4.Coolant flow forces (static)5.Temperature gradients6.Differences in thermal expansiona.Due to temperature differencesb.Due to expansion of different materials MPS3 UFSAR3.9N-66Rev. 307.Interference between components8.Vibration (mechanically or hydraulically induced)9.All operational transients listed in Table 3.9N-1 10.Pump overspeed11.Seismic loads (operational basis ea rthquake and safe shutdown earthquake)12.Blowdown forces (due to cold and hot leg break)
The main objective of the analysis is to satisfy allowable stress limits, to assure an adequate design margin, and to establish deformation li mits which are concerned primarily with the functioning of the components. The stress limits are established not only to assure that peak stresses will not reach unacceptable values, but also to limit the amplitude of the oscillatory stress component in consideration of fatigue characteristics of the materials. Standard method of strength of materials are used to establish the stresses and defl ections of these components. The dynamic behavior of the reactivity control compone nts has been studied using experimental test data and experience from operating reactors.
3.9N.4.3.2 Drive Rod Assembly All postulated failures of the dr ive rod assemblies either by frac ture or uncoupling lead to a reduction in reactivity. If the drive rod assembly fractures at any elevation, that portion remaining coupled falls with, and is guided by, the rod cluster control assembly. This always results in reactivity decrease for the control rods.
3.9N.4.3.3 Latch Assembly and Coil Stack Assembly Results of Dimensional and Tolerance Analysis With respect to the control rod drive mechanism system as a whole, critical clearances are present in the following areas:1.Latch assembly (diametral clearances)2.Latch arm-drive rod clearances3.Coil stack assembly-thermal clearances4.Coil fit in coil housingThe following defines clearances that are designed to provide reliable operation in the CRDM in these four critical areas. Thes e clearances have been proven by life tests and actual field performance at operating plants.
MPS3 UFSAR3.9N-67Rev. 30 Latch Assembly - Thermal Clearances The magnetic jack has several clearances where parts made of Type 410 st ainless steel fit over parts made from Type 304 stainless steel. Differential thermal e xpansion is therefore important.
Minimum clearances of these parts at 68
&deg;F is 0.011 inches. At a maximum design temperature of 650&deg;F minimum clearance is 0.0045 inches and at the maximum expected operating temperatures of 550&deg;F is 0.0057 inches.
Latch Arm - Drive Rod Clearances The CRDM incorporates a load transfer action. Th e movable or stationary gripper latch are not under load during engagement, as previously explained, due to load transfer action.
Figure 3.9N-6 shows latch clearance variation wi th the drive rod as a result of minimum and maximum temperatures. Figure 3.9N
-7 shows clearance variations over the design temperature range.Coil Stack Assembly - Thermal Clearances The assembly clearances of the coil stack assembly over the latch housing was selected so that the assembly could be removed under all anti cipated conditions of thermal expansion.
At 70&deg;F, the inside diameter of the coil stack is 7.428/7.438 inches. The outside diameter of the latch housing is 7.390/7.380 inches.
Thermal expansion of the mechanism due to operating temperat ure of the control rod drive mechanism results in minimum inside diameter of the coil stack being 7.440 inches at 222
&deg;F and the maximum latch housing diameter being 7.426 inches at 650
&deg;F.Under the extreme tolerance condi tions listed above it is necessary to allow time for a 70
&deg;F coil stack assembly to heat during a replacement operation.Four similar style coil stack assemblies we re removed from four hot control rod drive mechanisms mounted on 11.035-inch centers on a 550
&deg;F test loop, allowed to cool, and then replaced without incident as a test to prove the preceding.
Coil Fit in Core Housing CRDM coils and coil housing clearances are selected so that coil heat up results in a close to tight fit. This is done to facilitate thermal transfer and coil cooling in a hot CRDM.
MPS3 UFSAR3.9N-68Rev. 30 3.9N.4.4 CRDS Performance Assurance ProgramEvaluation of Material's Adequacy The ability of the pressure housing components to perform throughout the design lifetime as defined in the design specification is confir med by the stress analysis report required by the ASME Code, Section III.
Internal components subjected to wear have withstood a minimum of 2,500,000 steps without refurbishment as confirme d by life tests (Cooper 1974).To confirm the mechanical adequacy of the fuel assembly, the CRDM, and rod cluster control assembly (RCCA), functional test programs have been conducted on a full scale 12-foot control rod. The 12-foot prototype assembly was te sted under simulated c onditions of reactor temperature, pressure, and flow for approximately 1,000 hour
: s. The prototype mechanism accumulated about 2,500,000 steps a nd 600 trips. At the end of the test the CRDM was still operating satisfactorily. A correla tion was developed to predict the amplitude of flow-excited vibration of individual fuel rods and fuel assemblies. Inspection of the drive line components did not reveal significant fretting.These tests include verification that the trip time achieved by the CRDM's meet the design requirement of 2.2 seconds from start of RCCA motion to dashpot entry. This trip time requirement will be conf irmed for each CRDM prior to initi al reactor operation and at periodic intervals after initial reactor operation as required by the Technical Specifications.There are no significant differ ences between the prototype CRDMs and the production units.
Design materials, tolerances and fabricat ion techniques (Section 4.2.3.3.2) are the same.
These tests have been repor ted in WCAP-8446 and WCAP-8449.It is expected that all CRDM's will meet specified operating requirements for the duration of plant life with normal refurbishment.
If an RCCA cannot be moved by its mechanism, adjustments in the boron concentration ensure that adequate shutdown margin would be achieve d following a trip. Thus, inability to move one RCCA can be tolerated. More than one inoperabl e RCCA could be tolerated, but would impose additional demands on the plant operator. Therefor e, the number of inoperable RCCA's has been limited to one as discussed in the Technical Specifications (Chapter 16).In order to demonstrate proper operation of the CRDM and to ensure acceptable core power distributions during ope ration, RCCA partial-movement checks are performed (Technical Specifications). In addition, periodic drop tests of the RCCA ar e performed at each refueling shutdown to demonstrate continue d ability to meet trip time requirements, to ensure core subcriticality after reactor trip, and to limit potential reactivity insertions from a hypothetical RCCA ejection. During these tests the acceptable drop time of each assembly is not greater than MPS3 UFSAR3.9N-69Rev. 30 2.2 seconds, at full flow and operating temperatur e, from the beginning of motion to dashpot entry.Actual experience in operating Westinghouse plan ts indicates excellent performance of CRDM's.
All units are production tested prior to shipment to confirm ability of the CRDM to meet design specification-operation requirements.Each production CRDM undergoes a produc tion test as listed below:
3.9N.5 REACTOR VESSEL INTERNALS 3.9N.5.1 Design ArrangementsThe 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 (includi ng the entire core barrel and neutron shield pad assembly), the upper core support assembly and the inco re instrumentation support structure. The reactor internals support the core, maintain fuel alignment, limit fuel assembly movement, maintain alignment between fuel assemblies and CRDM's, dire ct 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 c oolant 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 co re plate. The lower core plate is sized to provide the Test Acceptance Criteria Cold (ambient) hydrostatic pressure test ASME Section III Confirm step length and load transfer (stationary gripper to movable gripper or movable gripper to stationary gripper)Step Length:
5/8 +/- 0.015 inches axial movementLoad Transfer:
0.059 inches nominal axial movement Cold (ambient) performance Test at design load -5 full travel
 
excursions Operating Speed:
45 inches/minuteTrip Delay:
Free fall of drive rod assembly to begin within 150 milliseconds as verified by normal gripper latch opening times
 
recorded during CRDM performance tests MPS3 UFSAR3.9N-70Rev. 30 desired inlet flow distribution to the core. After passing through the core, the coolant enters the region of the upper support structure a nd then flows radially to the core barrel 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 enteri ng flow is directed into the vessel head plenum and exits through th e 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 syst em 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 asse mbled core. The lower core support assembly and principally the core barrel serve to pr ovide 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 orie ntation for the fuel assemblies.
The lower core plate is a member through whic h the necessary flow distribution holes for each fuel assembly are located. Fuel assemb ly 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 dist ribution 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 st ainless 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 samp les 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 ir radiation specimen holders ar e given in WCAP-7870 (1972).Vertically downward loads from weight, fu el assembly preload, control rod dynamic loading, hydraulic loads, and earthquake acceler ation are carried by the lower core plate partially into the lower core plate support fl ange 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, a nd vibration are carried by the core barrel shell and distributed between th e lower radial support to the ve ssel wall and to the vessel flange. Transverse loads of the fuel assemblies are transmitted to th e core barrel shell by MPS3 UFSAR3.9N-71Rev. 30 direct connection of the lower core plate to the barrel wall and by upper core plate alignment pins which are welded into the core barrel.
The main radial support system of the lowe r end of the core barrel is accomplished by "key" and "keyway" joints to the reactor ve ssel wall. At equall y 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 AS ME Section III limits. In the event of an abnormal downward vertical displacement of the internals fo llowing 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 pa nel 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-indu ced 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 therma l shield results in some significant design advantages, specifically:
MPS3 UFSAR3.9N-72Rev. 301.The P in the downcomer region (annulus between core barrel and pressure vessel) is reduced by approximately 80 percent.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 freque ncy of the lower internals system. This is also an aid in reducing the tendency for indu ced vibrations. The adequacy of the design was confirmed by both analysis and test. Since the design satisfies Section 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 expe cted operating amplitude.To confirm that there were no deleterious ef fects 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 result s indicated extremely low levels of vibration.In summary, the neutron shie ld panel design of Millstone 3 has been shown to be structurally adequate by the following: 1.An in-depth analysis, considering lo adings for all plant operating conditions, indicates that the design sa tisfies all the cr iteria 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 th at the safety factors are approximately 3 on stress and 2 x 10 5 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 te sts with a thermal shield. The levels of vibration for the neutron shield pads were negligible.
MPS3 UFSAR3.9N-73Rev. 30 Upper Core Support Assembly 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 suppor t columns, and the guide tube assemblies.
The support columns establish the spacing be tween 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 plat es and serve the suppl ementary function of supporting thermocouples. The guide tube asse mblies, sheath and guide the control rod drive shafts and control rods. They are fast ened to the upper support and are restrained by pins in the upper core plate fo r 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 axia l direction. Lateral disp lacement 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 pl ate and engage the fuel assemblies as the upper assembly is lowered into place. Pr oper alignment of the lower core support assembly, the upper core support assembly, th e 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 barr el flange and the upper core support assembly and is co mpressed by installation of the reactor vessel head.Vertical loads from weight, earthquake accel eration, 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 vibrati ons 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 pene trating 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 he ad penetrations. The thermocouples are carried through these port columns and the uppe r support at positions above their readout locations. The thermocouple conduits are s upported from the columns of the upper core support system. The thermocouple conduits are stainless steel tubes.
MPS3 UFSAR3.9N-74Rev. 30 In addition to the upper incore instrume ntation, there are reactor vessel bottom instrumentation columns which carry the retractable, cold worked stainless steel flux 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 ar e 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 condui ts 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 clear ed 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.
3.9N.5.2 Design Loading Conditions Normal and Upset Conditions The normal and upset loading condi tions that provide the basis for the design of the reactor internals are:1.Fuel and reactor internals weight2.Fuel and core component spring for ces, including spring preloading forces3.Differential pressure and coolant flow forces4.Temperature gradients5.Vibratory loads including OBE seismic loads6.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 service9.Loss of load/pump overspeedEmergency Conditions
 
MPS3 UFSAR3.9N-75Rev. 30The emergency loading conditions that provide th e basis for design of the reactor internals are:1.Small LOCA2.Small steam break3.Complete loss of flow Faulted Conditions The faulted loading conditions that provide the ba sis for the design of the reactor internals are:1.Branch line breaks, as determined from leak-before-break (LBB) analysis2.SSE 3.9N.5.3 Design Loading CategoriesThe combination of design loadings fit into the normal, upset, emergency or faulted conditions as defined in the ASME Code, Section III.
Loads and deflections imposed on components due to shock and vibration are determined analytically and experimentally in both scaled models and operating reactors. The cyclic stresses due to these dynamic loads and deflections are combined with th e stresses imposed by loads from components weights, hydraulic forces and thermal gradients for the determination of the total stresses of the internals.The reactor internals are designed to withstan d stresses originating from various operating conditions as summarized in Table 3.9N-1.
The scope of the stress analysis problem is very large requiring many different techniques and methods, both static and dynamic. The analysis performed depends on the mode of operation under consideration.
Allowable Deflections For normal operating conditions, downward vertical deflection of the lower core support plate is negligible.
For the loss of coolant accident plus the sa fe shutdown earthquake condition, the deflection criteria of critical internal structures are the limiting values given in Table 3.9N-13. The corresponding no loss of function limits are included in Table 3.9N-13 for comparison purposes with the allowed criteria.The criteria for the core drop accident are based upon analyses which have to determine the total downward displacement of the internal structures following a hypothesized core drop resulting MPS3 UFSAR3.9N-76Rev. 30 from loss of the normal core barrel supports. Th e initial clearance between the secondary core support structures and the reactor vessel lower head in the hot condition is approximately one half inch. An additional displacement of approximately 3/4 inch woul d occur due to strain of the energy absorbing devices of the secondary core support; thus the total drop distance is about 1-1/4 inches which is insufficient to permit the tips of the rod cluster control assembly to come out of the guide thimble in the fuel assemblies.Specifically, the secondary core support is a device which will never be used, except during a hypothetical accident of the core support (core barrel , barrel flange, etc.).
There are 4 supports in each reactor. This device limits the fall of the core and absorbs much of the energy of the fall which otherwise would be imparted to the vessel. The energy of the fall is calculated assuming a complete and instantaneous failur e of the primary core support and is absorbed during the plastic deformation of the controlled volume of stainless steel, loaded in tension.The maximum deformation of this austenitic stainless steel piece is limited to approximately 15 percent, after which a positive stop is provided to ensure support.
3.9N.5.4 Design BasesThe design bases for the mechanical design of the reactor vessel internals components are as follows:1.The reactor internals in conjunction with the fuel assemblies shall direct reactor coolant through the core to achieve acceptab le flow distribution and to restrict bypass flow so that the heat transfer performance requireme nts 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 th e flange during reactor operation.2.In addition to neutron shie lding provided by the reactor coolant, a separate neutron pad assembly is provided to limit the exposure of the pr essure vessel in order to maintain the required ductility of the material for all modes of operation.3.Provisions shall be made for installing in core instrumentation useful for the plant operation and vessel material test speci mens required for a pressure vessel irradiation surv eillance program.4.The core internals are designed to wi thstand 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 tran sients arising from abnormal operating conditions.
MPS3 UFSAR3.9N-77Rev. 306.Following the design basis accident, the plant shall be capable of being shut down and cooled in an orderly fashion so that fuel cladding temperature is kept within specified limits. This implies that the deformation of certain critical reactor internals must be kept sufficient ly small to allow core cooling.
The functional limitations for the core structures during the design basis accident are shown in Table 3.9N-13. To ensure no column loading or r od cluster control guide tubes, the upper core plate deflection is limited to not ex ceed the value shown in Table 3.9N-13.
Details of the dynamic analyses, input forcing f unctions, and response loadings are presented in Section 3.9N.2.
The basis for the design stress and deflection criteria is identified below:Allowable Stresses For normal operating conditions the intent of Section III of the ASME Code is used as a basis for evaluating acceptability of calculated stresses. Both static and alternating stress intensities are considered.
It should be noted that the allowable stresses in Section III of the ASME Code are based on unirradiated material proper ties. In view of the fact that irradiation increases the strength of the Type 304 stainless steel used for the internals, although decreasing its elongation, it is considered that use of the allowable stresses in Section I II is appropriate and cons ervative for irradiated internal structures.The allowable stress limits during the design basis accident us ed for the core support structures are based on the 1974 Edition of the ASME Code for Core Support Structures, Subsection NG, and the Criteria fo r faulted conditions.
3.9N.6 INSERVICE TESTING OF PUMPS AND VALVES Refer to Section 3.9.7.
MPS3 UFSAR3.9-90Rev. 30
 
====3.9.7 INSERVICE====
TESTING OF PUMPS AND VALVES A test program has been developed to ensure that all safety related pumps a nd valves will be in a state of operational readin ess throughout plant life.
The ASME Code, Section XI, 1980 Edition through Winter 1980 Addendum, provided the basic rules used to identify applicable pumps an d valves and to develop test requirements.
3.9.7.1 Inservice Testing of Pumps Inservice testing is required for all Class 1, 2, and 3 pumps (both centr ifugal and displacement types) that are provided with an EMERGENCY POWER SOURCE. Drivers are excluded except when the pump and driver form an integral unit and the pump bearings are in the driver.
All tests and examination procedures required by the ASME code for Operation and Maintenance of Nuclear Power Plants including schedules, reference values, the location, and type of measurement for each of the requi red test quantities, records of the results, and all corrective action taken are defined in the Inservice Test Pump and Valve Program and performed by the owner for Millstone 3.
3.9.7.2 Inservice Testing of Valves This section describes the Class 1, 2, and 3 valves required to be exercised and tested to verify operational readiness. These valves (with thei r actuating and position in dicating devices) are required to perform a special function in bringing a reactor to cold shutdown condition or in mitigating the consequences of an accident.Valves used for operating convenience only such as manual vent, drain, instrument and test valves, and valves used for maintenance only do not require inservice testing. The following categories of valves are s ubject to inservice testing:1.Category A - Valves for which seat leak age is limited to a specific maximum amount in the closed position fo r fulfillment of their function.2.Category B - Valves for which seat leakage in the closed position is inconsequential for fulfil lment of their function.3.Category C - Valves whic h are self-actuating in response to some system characteristic, such as pressure (relief valves) or flow dire ction (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.Test and examination procedures required by Subsection IWV, including schedules and the limiting values of observed parameters are defined in the Inservice Test Pump and Valve Program and performed by the licensee for Millstone 3.
MPS3 UFSAR3.9-91Rev. 303.
 
==9.8 REFERENCES==
FOR SECTION 3.93.9-1Bohm, G.J. and LaFaille, J.P. 1971. Reac tor Internals Response Under a Blowdown Accident, Procedures, First Intl. Conf. on Structural Mech. in Reactor Technology. Berlin, September 20-24, 1971.3.9-2DeSalvo, G.J. and Swanson, J.A. 1972. ANSYS User's Manual. En gineering Analysis Systems Report, October 1, 1972.3.9-3Fabic, S. 1967. Computer Program WHAM fo r Calculation of Pressure, Velocity, and Force Transients in Liquid Filled Piping Networ ks. Kaiser Engineers Report No.
67-49-R.3.9-4Trojan Final Safety Analysis Re port, Appendix A-12 (Docket No. 50-344).3.9-5WCAP-7870, 1972, Kraus, S. "Neutron Shielding Pads."3.9-6WCAP-8252, Revision 1, 1977, "D ocumentation of Selected Westinghouse Structural Analysis Computer Codes."3.9-7WCAP-8303-P-A (Proprietary) and WCAP-8317-A (Non-Proprietary), 1975, "Prediction of the Flow-Induced Vibration of Reactor Internals by Scale Model Tests."3.9-8WCAP-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.3.9-9WCAP-8516-P (Proprietary) and WCAP-8517 (Non-Proprietary), 1975. Bloyd, C.N. and Singleton, N.R. UHI Plant Internals Vibrat ion Measurement Program and Pre and Post Hot Functional Examinations.3.9-10WCAP-8708-P-A (Proprieta ry) and WCAP-8709-A (Non-Proprietary), 1977. Takeuchi, K. "MULTIFLEX-A Fortran-IV Computer Pr ogram for Analyzing Thermal - Hydraulic - Structure System Dynamics."3.9-11WCAP-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."3.9-12WCAP-8929, 1977, "Benchmark Problem So lutions Employed for Verification WECAN Computer Program."3.9-13WCAP-9945, 1981. Altman, D.A. et al., "Verification of U pper Head Injection Reactor Vessel Internals by Preoperational Tests on the Sequoyah Unit 1 Power Plant."
MPS3 UFSAR3.9-92Rev. 30TABLE 3.9B-1 LIST OF INPUT DOCUMENTS DESCRIBING DESIGN TRANSIENT FOR FATIGUE ANALYSIS OF RCP AND ASSOCIATED CLASS 1 PIPING Piping System DescriptionMPS-3 Piping Drawing SeriesSWEC (MPS-3) Design Transient ReportSWEC (MPS-3) Piping Stress Analysis Problem Number Reactor coolant loops (including surge attachments to coolant
 
loops)EP 70TR 2658-27000 7001  7002  7003Residual heat removal pipingEP 71TR 2658-37000 7001 Chemical volume and control system EP 74 and part of EP107 & 108TR 2658-47419 7420  7422  7423  7425  7427 Low pressure safety injection systemEP 82TR 2658-57000 7001 7002 70013 Annulus piping (contains portions of high pressure safety injection, auxiliary
 
pressurizer spray and chemical volume and control)EP 107TR 2658-610700 10701  10702  10703  10704  10706  10707  10717  10729 Cubicle piping (includes letdown)EP 108TR 2658-710800 10802  10803 Pressurizer spray, safety and reliefEP 109TR 2658-810900 10901 MPS3 UFSAR3.9-93Rev. 30TABLE 3.9B-2 LIST OF MPS-3 DOCUM ENTS DESCRIBING COMPONENTS REQUIRING INELASTIC ANALYSISComponent DescriptionAssociated MPS-3 PPG System &
Component LocationDocument Providing Details of Analysis 3 inch Charging Nozzle on RC Loop RC Loop 1 &
3RCS-003-149-1 EP 74B (H-8)Teledyne Engineering Services Report TR 2658-213 inch Charging Nozzle on RC Loop RC Loop 4 &
3RCS-003-145-1 EP 74B (E-8)Teledyne Engineering Services Report TR 2658-21 3 inch Safety Injection Nozzle on RC Loop RC Loop 1 &
3RCS-003-121-1 EP 108A (I-4)Teledyne Engineering Services Report TR 2658-22 3 inch Safety Injection Nozzle on RC Loop RC Loop 2 &
3RCS-003-133-1 EP 108B (G-7)Teledyne Engineering Services Report TR 2658-22 3 inch Safety Injection Nozzle on RC Loop RC Loop 3 &
3RCS-003-139-1 EP 108C (C-7)Teledyne Engineering Services Report TR 2658-22 3 inch Safety Injection Nozzle on RC Loop RC Loop 4 &
3RCS-003-147-1 EP 108D (C-5)Teledyne Engineering Services Report TR 2658-2212 sets of Circumferential Butt Welds (as welded)Sets 1 through 3Low Pressure Safety Injection Piping to RC Loop 10 inch Schedule 140 Butt Welds SWEC Calculation 12179-NP(B)-315-XISets 4 & 5Boron Injection System 1.5 inch Schedule 160 Butt Welds SWEC Calculation 12179-NP(B)-317-XISets 6 through 10RCS Drains Loops 1, 2, 3 & 4 2 inch Schedule 160 Butt Welds SWEC Calculation 12179-NP(B)-299-XISet 11Pressurizer Spray Nozzle C-1 Weld 3RCS*TK1  3RCS-004-224-1 SWEC Calculation 12179-NP (F)-316-XI Set 12Bimetallic Welds between Steam Generator and RCS Loop SWEC Calculation 12179-NP(B)-4019-XI MPS3 UFSAR3.9-94Rev. 30TABLE 3.9B-3 PREOPERATIONAL TESTS (1)Types of Tests (2)System CodeSystem Title Reg. Guide 1.68, Rev. 2 Classification (3)Thermal ExpansionTransient VibrationsSteady State VibrationsPGSPrimary Grade WaterA, DNRNR (4)V (5)IASInstrument AirA, DNRNRNRGSNNitrogen SystemA, BNRNRV CCECharging Pump CoolingANRNRVSWPService WaterA, DNRNR (6)VWTCWater Treating -ChlorinationANRNRNRCHSChemical and Volume ControlA, B, A1V & I (7)VVCCPReactor Plant Component CoolingA, DNRNRVEGFEmergency Diesel FuelA, A1NRNRVCDSChilled WaterA, DNRNRVEGAAir Startup Emergency DieselAVNRVEGDEmergency Generator Exhaust and Combustion AirAVNRVEGSEmergency Diesel Jacket and Intercooler WaterANRNRVMSSMain SteamA, BV & IV & IVDTMTurbine Plant Miscellaneous DrainsAVNRVQSSQuench SprayA, DNRNR V (8)CCISafety Injection Pump CoolingA, A1NRNRV MPS3 UFSAR3.9-95Rev. 30HVCAir Conditioning -Control BuildingANRNRNRHVKChilled Water -Control BuildingANRNRVSIHSafety Injection -
High PressureA, BNRVVFWSFeedwaterA, BV & IV & IVFWAAuxiliary Feedwater and RecirculationA, B, A1NRNRVRHSResidual Heat RemovalA, B, DV & INRVSILSafety Injection -
Low PressureA, BV & INRVRCSReactor Coolant Main LoopsAV & IV & IVBDGSteam Generator BlowdownA, BV & IVVDASReactor Plant Aerated DrainsA, DVNRVSGFSteam Generator -
Chemical FeedANRNRVRSSContainment Recirculation SprayA, B, CNRNR V (8)SSRSampling System -
Reactor PlantAVNRNRHVUVentilation - Containment StructureANRNRNRSFCFuel Pool Cooling and PurificationA, DNRNRVTABLE 3.9B-3 PREOPERATIONAL TESTS (1)Types of Tests (2)System CodeSystem Title Reg. Guide 1.68, Rev. 2 Classification (3)Thermal ExpansionTransient VibrationsSteady State Vibrations MPS3 UFSAR3.9-96Rev. 30HCSHydrogen RecombinerA, A1NRNRVSSPSampling System -
Post AccidentAVNRNRGWSRadioactive Gaseous Waste (9)AVNRNRVRSReactor Plant Gaseous VentsA, DVNRNRLMSContainment Leakage MonitoringANRNRNRDGSReactor Plant Hydrogenated DrainsAVNRNRCMSContainment Atmosphere MonitoringANRNRNRCVSContainment VacuumA, B, DVNRVICIIncore Instrument LinesA, BNRNRNRRCSPressurizer Safety and Relief SystemA, BV & IV & IVRCSPressurizer Spray SystemA, BV & INRVFPWFire Protection WaterDNRNRNRDWS/ PBSDomestic Water/
Sanitary SystemDNRNRNRSASService Air Containment Service AirDNRNRNRSVVMain Steam Safety Valve Steam Vents
 
and DrainsBVNRNRTABLE 3.9B-3 PREOPERATIONAL TESTS (1)Types of Tests (2)System CodeSystem Title Reg. Guide 1.68, Rev. 2 Classification (3)Thermal ExpansionTransient VibrationsSteady State Vibrations MPS3 UFSAR3.9-97Rev. 30NOTES:1. The detailed test plans shall identify those portions of each system to be tested.2. Type of tests reflect the graded approach.3. NRC Regulatory Guide 1.68, Revision 2 Classifications:A= ASME III, High Energy Piping, Classes 1, 2, and 3.
A1= ASME III, Moderate Ener gy 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 seismi c category plant features outside Seismic Category I Structures, this classifi cation does not exist for Millstone 3.
D = Seismic Category I portions of Moderate Energy Piping located outside containment structures.4. NR = Testing not specifically required. Temperature < 200
&deg;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.5.V = Visual. When V appears in the table, hand held instruments may be required to perform observations.6.I = Instrumented measurements. When V & I appears in the table, only a portion of the system requires instrument observations.WSSRadioactive Solid WasteDNRNRNRVASNuclear Aerated VentsDNRNRNRLWSRadioactive Liquid WasteBVNRNRBRSBoron RecoveryB, DVNRVNSSNeutron Shield Tank Cooling SystemANRNRNRASSAuxiliary SteamBVNRVTABLE 3.9B-3 PREOPERATIONAL TESTS (1)Types of Tests (2)System CodeSystem Title Reg. Guide 1.68, Rev. 2 Classification (3)Thermal ExpansionTransient VibrationsSteady State Vibrations MPS3 UFSAR3.9-98Rev. 307.Fluid transient source was eliminated during initial system checkout (Phase II testing) by addition of an open-to-air vent. No retesti ng required during preoperational or initial startup testing program.8.Only that portion of system is olatable from the spray rings.9.Radioactive Gaseous Waste (GWS) System is non-seismic for design and analysis purposes.
MPS3 UFSAR3.9-99Rev. 30TABLE 3.9B-4 OMITTED MPS3 UFSARMPS3 UFSAR3.9-100Rev. 30NOTES:1. The nomenclature, conditions, and applicatio ns of the above allowables are in accordance with ASME Section III. Stress limit s above apply to design by elastic an alysis. Limit and plastic analysis is allowed in accordance with ASME Sect ion III criteria.
Special stress limits of NB 3227 as applicable. Stress limits of Subsection NF ar e used for the design of supports as applicabl e.TABLE 3.9B-5 STRESS LIMITS FOR ASME SECTION III CLASS 1 (NB) SEISMIC CATEGORY I COMPONENTS (ELASTIC ANALYSIS)Pressure Vessels, Pumps and Valve Bodies - Pressure Boundary - Designed by Analysis (1)Condition of DesignReference Paragraph ASME Section IIIPrimary Stress Limits  Secondary Stress Limits Peak Stress Limits P m P L P L + P b P e P L + P b + P e + QP L + P b + P e + Q + E Design (2)NB-3221S m 1.5 S m 1.5 S mNot RequiredNot RequiredNormal (3)NB-3222(4) (4) (4) 3 S m 3 S m S a Upset (3)NB-3223(4)(4)(4)3 S m 3 S m S a Emergency (3) (5)NB-3224Greater of 1.2S m or 1.0S y Greater of 1.8S m or 1.5S y Greater 1.8S m or 1.5S yNot RequiredNot Required Faulted (3) (4)NB-3225,Lesser of 2.4S m or 0.7S u Lesser of 3.6S m or 1.05S u Lesser of 3.6S m or 1.05S uNot RequiredNot Required NB-3221, App. F F 1323.1Testing (6)NB-32260.90 S y 1.35S yNot RequiredNot Required MPS3 UFSAR3.9-101Rev. 302. Use design loads.3. Use operating loads.4. Primary stresses often evaluated and combined with secondary effects.5. Use above limits for materials of Table 1-1.2 (A SME Section III). Use 0.7S u for materials of Table 1-1.1 (ASME Section III).6. Use test loads (pressure, temperature).
MPS3 UFSAR3.9-102Rev. 30(1) Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code including 1972 Winter Addenda thereto.
(2) The requirements of Case 1552 (Interpretations of ASME Boiler and Pressure Vessel Code) should be met for all sizes of Code Class 1 valves designed by analysis.
(3) In addition to compliance with design limits specified, assurance of operability under all design loading combinations should be provided by an appropriate combination of the following suggested measures:
In situ testing (e.g., preopera tional testing after the component is installed in the plant)(a)(a) Full-scale prototype testingTABLE 3.9B-6 COMPARISON OF CLASS 1 REQUIREMENTS REGULATORY GUIDE 1.48 VS. TABLES 3.9B-5 AND 3.9B-10 Component Normal or Upset + OBE EmergencyNormal + Faulted +
SSE Regulatory Position Comparison With Regulatory PositionVessels (1)NB-3223NB-3224NB-3225C.1Agree Pipe (1)NB-3654NB-3655NB-3656C.1Agree Nonactive Pumps, Valves (Design by Analysis)
(2)NB-3223NB-3224NB-3225C.2AgreeNonactive Valves (Design by Standard or Alternate Design Rules)1.1 P r1.2 P r1.5 P rC.3Agree Active Pumps, Valves (Design by Analysis)
(2)(3)NB-3222NB-3222NB-3222C.4Alternate acceptable basis, exception to footnotes (3) and (4)Active Valves (Design by Standard or Alternate Design Rules) (3)1.0 P r1.0 P r1.0 P rC.5Agree with pressure rating factors, exception to footnote (3)NOTES: (from Regulatory Guide 1.48, pages 1.48-7 and -8)
MPS3 UFSAR3.9-103Rev. 30(b)(b) Reduced-scale prototype testing(c)(c) Detailed stress and deformation an alyses (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 superpos ition of test results for other than the combined loading condition is proposed, the a pplicability of such a procedure should be demonstrated. The design limits for nonactive pumps and valves designed by analysis for applicable loading combinations if assu rance is provided by detailed stress and deformation analyses that operability is not impaired when designed to these limits. Similarly, the primary-pressure ratings P fo r nonactive valves designed by standard or alternative design rules may be used for the applicable loading combinations if appropriate testing demonstrates that operabi lity is not impaired when the valve is so rated.(4) Secondary affects (stresses and deformat ions) should be evaluated for the loading combinations designated by Regulatory Positions 4.a.(2) and 4.a.(3). Local affects (peak stresses) need not be considered for these loading combinations.
(5) Applies to all components (vessels, piping, pum ps, and valves) that are relied upon to cope with the effects of specified plant conditions.
(6) Identification of the specific transients or events to be co nsidered under each plant condition will be addressed in a future regulatory guide.(7) The provisions of NB-3411 a nd NB-3413 may be applied for all sizes of Code Class 1 pumps designed by analysis.(8) Table 1.3-0, "Permanent Strain Limiting Fact ors" of Appendix I of the ASME Boiler and Pressure Vessel Code, Section III, may be used as an aid in determini ng the relationship between design stress and deformation (see Note 2 to Table I-1.2 of Section II I of the ASME Code).
MPS3 UFSAR3.9-104Rev. 30NOTES:(1) S - Allowable stress values at design temperature from ASME Section III, Appendix I, as allowed by class Sm - Design stress intensity values at de sign temperature from ASME Section III, Appendix I, as allowed by class(2) Fatigue analysis may be required with operating conditions, reference paragraph NC-3219 and Appendix XIV of ASME Section III, Subsection NC.(3) When a complete analysis is performed in accordance with NC 3211.1(c), the faulted stress limits of Appendix F shall apply.TABLE 3.9B-7 STRESS LIMITS FOR AS ME SECTION III CLASS 2 AND 3 COMPONENTS (ELASTIC ANALYSIS)Design ConditionASME III Class CodePrimary Stress Limits (1)Membrane (P m)Membrane Plus Bending (P m + P b)Pressure Vessels I2(NC3300) or1.1 S1.65 SII3(ND3300)2.0 S2.40 S I (2)2(NC3200)1.1 S m 1.65 S m II (3)2.0 S m 2.40 S mPumps (4) (5), InactiveI2(NC3400) or1.1 S1.65 S II3(ND3400)2.0 S2.40 SPumps (4)(5), ActiveI2(NC3400) or1.0 S1.50 SII3(ND3400)1.2 S1.80 SValves (5)(6), Active and InactiveI2(NC3500) or1.1 S1.65 S II3(ND3500)2.0 S2.40 STanks (5) (Steel)I2(NC38-3900) or1.1 S1.65 SII3(ND38-3900)2.0 S2.40 S MPS3 UFSAR3.9-105Rev. 30(4) In accordance with NC-3400 and ND-3400, any design method which has been demonstrated to be satisfactory for the specified design conditions may be used.(5) Stress limits of ASME Section III, Subsection NF, are used for the design of supports as applicable.(6) The standard or alternative design ru les of NC-3500 and ND-3500 may be used in conjunction with the st ress limits specified.Valve nozzle (piping load) stress analysis is not required when both the following conditions are satisf ied by calculation:(a)Section modulus and area at the plan e 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 va lve body inlet and outlet nozzles; and, (b) Code allowable stress, S vlv , for valve body material, is equal to or greater than the code allowable stress, S pip , of connected piping material. If the valve body material allowabl e 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 requireme nts, 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 contai ned within the confines of the body and bonnet.
MPS3 UFSARMPS3 UFSAR3.9-106Rev. 30TABLE 3.9B-8 COMPARISON OF CLASSES 2 AND 3 REQUIREMENTS REGULATORY GUIDE 1.48 VS TABLE 3.9B-7Regulatory Guide 1.48 Table 3.9B-7Components Loading CombinationsDesign LimitsRegulatory Position Loading Combinations      Design Limits Comparison P m P m (or P L)+P b P m P m (or P 1)+P bPressure Vessels Classes 2 and 3Normal1.1 S1.65 SC.6aDesign+OBE1.1 S1.65 SAcceptable alternate, Code Case 1607, Regulatory Guide
 
1.84Upset+OBE1.1 S1.65 SC.6aEmergency1.1 S1.65 SC.6aDesign+SSE2.0 S2.40 S Faulted + SSE1.5 S2.25 SC.6b Piping Classes 2 and 3NormalNormalAcceptable alternate, Code Case 1606, Regulatory Guide 1.84Upset+OBENC-3611.1 (b)(4)(c) (b)(1)1.2 S hC.8.aUpset +OBENC-3611.1 (b)(4)(c) (b)(1)1.2 S hEmergencyEmergencyNC-3611.1 (b)(4)(c)
(b)(2)1.8 S hFaulted+SSENC-3611.1 (b)(4)(c) (b)(2)1.8 S hC.8.bFaulted+SSE2.4 S h
MPS3 UFSARMPS3 UFSAR3.9-107Rev. 30 Pumps Classes 2 and 3 Acceptable alternate, Code Case 1636, Regulatory Guide 1.84InactiveNormal1.1 S1.65 SDesign+OBE1.1 S1.65 SUpset+OBE1.1 S1.65 SC.9.a Emergency1.1 S1.65 SDesign+SSE2.0 S2.40 SFaulted+SSE1.2 S1.8 SC.9.bActiveNormalAcceptable alternate programUpset+OBES1.5 SC.10.aDesign+OBE1.0 S1.5 SEmergencyFaulted+SSEDesign+SSE1.2 S1.8 SValves Classes 2 and 3Acceptable alternate, Code Case 1635, Regulatory Guide
 
1.84InactiveNormalDesign+OBE1.1 S1.65 SUpset+OBE1.1 P r C.11.aEmergencyDesign+SSE2.0 S2.4 SFaulted+SSE1.2 P rC.11.bincluding P rActiveAll Conditions1.0 P rC.12 aSame as for inactiveAcceptable alternate programTABLE 3.9B-8 COMPARISON OF CLASSES 2 AND 3 REQUIREMENTS REGULATORY GUIDE 1.48 VS TABLE 3.9B-7Regulatory Guide 1.48 Table 3.9B-7Components Loading CombinationsDesign LimitsRegulatory Position Loading Combinations      Design Limits Comparison P m P m (or P L)+P b P m P m (or P 1)+P b MPS3 UFSAR3.9-108Rev. 30 NOTES from Regulatory Guide 1.48, pages 1.48-7 and -8 Rev. 0, May 1973)
: (1) For loadings designated in Regulatory Positi on 8.a(2), only Equation 9 of NC-3651 need be met.(2) In addition to compliance with the design limi ts specified, assurance of operability under all design loading combinations should be provided by any appropriate combination of the following suggested measures: (a) In situ testing (e.g., preope rational testing after the compon ent is installed in the plant)(b) Full-scale prototype testing (c) Reduced-scale prototype testing (d) Detailed stress and deformation anal yses (includes expe rimental stress and deformation analyses)In the performance of tests or analyses to demonstrate operability, the structural interaction of the entire assembly (e.g., valve-operator 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 proce dure should be demonstrated. The design limits for nonactive pumps ad valves may be used for the applicable loading combinations if appropriate analyses and/or testing confirms that operability is not impaired when designed to these limits.
MPS3 UFSARMPS3 UFSAR3.9-109Rev. 30TABLE 3.9B-9 STRESS LIMITS FOR ASME SECTION III CLASS 1, 2, AND 3 COMPONENT SUPPORTS *Condition of Design    Plate and Shell Linear Type Component Standard Type Class 1 (NF3220)Class 2 (NF3320)Class 3 (NF3400)Class 1 (NF3230)Class 2 (NF3330)Class 3 (NF3400)Class 1 (NF3240)Class 2 (NF3340)Class 3 (NF3400)DesignNF3221NF3321.1NF3321.1NF3231.1aNF3231.1aNF3231.1aNF3221NF3321.1NF3321.1 or or orNF3226App. XVIIApp. XVIIApp. XVINF3231.1aNF3231.1aNF3231.1aApp. XVIIApp. XVIIApp. XVIINormalNF3222NF3221.2NF3321.2NF3231.1aNE3231.1aNF3231.1aNF 3222NF3321.2NF3321.2 or or orNF3226NF3321.1NF3321.1App. XVIIApp. XVIIApp. XVIINF3231.1aNF3231.1aNF3231.1aApp. XVIIApp. XVIIApp. XVIIUpsetNF3223NF3321.2NF3321.2NF3231.1aNF3231.1aNF3231.1aNF3223NF3321.2NF3321.2 or or orNF3222NF3321.1NF3321.1App. XVIIApp. XVIIApp. XVIINF3231.1aNF3231.1aNF3231.1a NF3226App. XVIIApp. XVIIApp. XVIIEmergencyNF3224NF3321.3NF3321.3NF3231.1bNF3231.1bNF3231.1bNF3224NF3321.3NF3321.3 or or or NF3231.1bNF3231.1bNF3231.1bFaultedNF3225NF3321.4NF3321.4NF3231.1cNF3231.1cNF3231.1cNF3231.1cNF3321.4NF3321.4 App. F1320 App. F1370 App. F1370 App. F1370 App. F1370 or or MPS3 UFSARMPS3 UFSAR3.9-110Rev. 30NOTE:* The nomenclature, conditions, and applications of the above refe renced allowables are in accordance with ASME III requirement
: s. Bolts are qualified in accordance w ith Appendix XVII as directed by NF3280. Welded joints are designed in accordance with NF3290.
App. F1370or NF3225NF3231.1cNF3231.1c App. F1370 App. F1370TABLE 3.9B-9 STRESS LIMITS FOR ASME SECTION III CLASS 1, 2, AND 3 COMPONENT SUPPORTS *Condition of Design    Plate and Shell Linear Type Component Standard Type Class 1 (NF3220)Class 2 (NF3320)Class 3 (NF3400)Class 1 (NF3230)Class 2 (NF3330)Class 3 (NF3400)Class 1 (NF3240)Class 2 (NF3340)Class 3 (NF3400)
MPS3 UFSAR3.9-111Rev. 30 Loadings Applicable to Component Supports D - Sustained mechanical loads, includi ng deadweight of equipment and contentsT -Loads on supports due to thermal expansi on (constraint of free-end displacement) of componentsE -Inertia effects of the OBEE'-Inertia effects of the SSEAl-Loads resulting from primary loop pipe rupture asymmetric pressure effects (see FSAR Section 3.9B.1.4
)P-Piping associated loads as follows:d -sustained deadweight of piping contents and insulationr -loads induced on component supports due to thermal and pressure growth of piping for appropriate plant conditione -inertia effects of OBEe'-inertia effects of SSEa-loads induced in component supports due to response of ci vil structure for OBE (OBE anchor movement)a'-loads induced in component supports due to response of civil st ructure of SSE (SSE anchor movement)h-loads resulting from occasional loads other than seismic (water hammer, steam hammer, safety relief valve opening or clos ing, etc.) as appropriate for plant conditional-loads induced in component supports due to the effects of LOCATABLE 3.9B-9A  LOAD COMBINATIONS FO R ASME SECTION III CLASS 1, 2, AND 3 COMPONENT SUPPORTSPlant Operating ConditionLoad CombinationNormalD + T + P (d + r)UpsetD + T + E + P (d + r + e + a + h)Emergency D + T + E' + P (d + r + e' + a' + h)
FaultedD + T + E' + Al + P (d + r + e' + a' + h + al)
MPS3 UFSARMPS3 UFSAR3.9-112Rev. 30TABLE 3.9B-10  ASME III CLASS 1 STRESS AND FATIGUE ANALYSIS REQUIREMENTS PER NB3650 Normal and Upset Conditions Emergency Conditions Faulted Conditions Primary Stress Intensity (Equation 9) 2.25S m but not greater than 1.8S y3.0S mPrimary and Secondary Stress Range (Equation 10)
N/RN/RPeak Stress Range (Equation 11)
N/RN/R Thermal Expansions Range (Equation 12)
N/RN/R Primary and Secondary Membrane, and Bending Stress (Equation 13)
N/RN/RAlternating Stress (Equation 14)
N/RN/R Usage Factor N/RN/R NOTES: Nomenclature is as described in ASME Section III, NB-3600.
B 1 PD o 2t----------B 2 D o 2l------M 11.5S m+S n C 1 P o D o 2t---------------C 2 D o 2l-------M 1 121v-()---------------------ET 1 C 3 E aba T ab T b-3S m+++=S p K 1 C 1 P o D o 2t---------------K 2 C 2 D o 2l-------M 1 121v-()---------------------K 3 ET 1 K 3 C 3 E aba T ab T b-1 1v-------------ET 2++++=S e C 2 D o 2l-------M 1 3S m=C 1 P o D o 2t---------------C 2 D o 2l-------M 1+C 1 2 E aba T ab T b-3S m+S alt 1/2K e S p=UActualNo.CyclesAllowableNo.Cycles----------------------------------------------------------U1.0=
MPS3 UFSARMPS3 UFSAR3.9-113Rev. 30 B 1 = 0.5 may be used in lieu of 1.0 for branch connections, curved pipe/elbows, and tees.
C 1 and K 1 indices may be derived fr om NUREG CR-0778, June 1979.LOAD COMBINATIONS FOR ASME III CLASS 1 PIPING 1Plant Operating ConditionEquationsLoad (Moment Combination) 2Normal/Upset9P d + D + E+ H + V10P o + T + R + H + E + A + V + L11P o + T + R + H + E + A + V + L12T + R13P o + D + E+ H + V + L14P o + T + R+ H + E + A + V + L12aS 9P 1 + D + H + YEmergency9P 1 + D + Y' + E' + H + A1FaultedP t TestP t + PDTABLE 3.9B-10  ASME III CLASS 1 STRESS AND FATIGUE ANALYSIS REQUIREMENTS PER NB3650 MPS3 UFSAR3.9-114Rev. 30NOTES:1. This includes the follow ing 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 PipingPortions of Pressurizer Safety, Relief and Spray Piping.2See Table 3.9B-12 for De finition of Loadings.3Class 1 piping does not experience loading designated by "W."4. Occasional dynamic loads such as waterhammer, steamhammer, safety valve discharge, etc. are combined with seismic inertia by SRSS method.LOAD COMBINATIONS FOR AS ME III CLASS 1 PIPING (1)Plant Operating Condition EquationsAllowable StressNormal/Upset91.5 S m 10-3.0 S m 11-123.0 S m 133.0 S m 14-12a3.0S m Emergency92.25 S m or 1.8S y Faulted93.0 S m Test0.9S y 1.35S y MPS3 UFSAR3.9-115Rev. 30NOTE:*All nomenclature are as defined in ASME Section III NC, ND 3600.TABLE 3.9B-11 ASME III CLASS 2 AND 3* STRESS ANALYSIS REQUIREMENTS PER NC3650 AND ND3650 Normal and Upset Conditions Emergency ConditionsFaulted Conditions Sustained Loads (Equation 8)N/RN/ROccasional Loads (Equation 9) 1.8S h 2.4S hThermal Expansion Loads (Equation 10)N/RN/R Sustained and Thermal Expansion Loads (Expansion 11)N/RN/RTABLE 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 Plant Operating Condition NC 3600 EquationsLoad (Moment) Combinations **Allowable StressNormal/Upset8P d + DS h 9P p + D + E + H + V + W1.2 S h 10*T + R + AS A10aS3 S C11*P d + D + T + R + AS h + S AEmergency9P p + D + H + Y + W1.8 S hFaulted9P p + D + E' + H + Y' + W + A11.8 S h10*T + R' + A' + XS A PD o 4t----------
0.75i ()M A Z-------------
------------
-S h+PD o 4t----------
0.75i ()M A Z-------------
------------
-0.75i ()M B Z--------------
-----------
+1.2S h+i M c Z-------S A    F1.25S c0.25S h+()PD o 4t----------
0.75i ()M A Z-------------
------------
-i ()M C Z--------------
-+S a S+h+
MPS3 UFSAR3.9-116Rev. 30NOTES:*Either the requirements of Eq. (10) or Eq. (11) must be satisfied.**See Table 3.9B-12 for definition of loading.(1)Class 1 portions of the Safety Inject ion are analyzed in accordance with Table 3.9B.10
.(2)In break exclusion area, sum of stresses given by Equations 9 a nd 10 should not exceed 0.8 (S A + 1.2 S h). In crack exclusion area, sum of stresses given by Equations 9 and 10 should not exceed 0.4 (S A + 1.2S h)(3)Occasional dynamic loads such as waterhammer, steamhammer, safety valve discharge, etc. are combined with seismic inertia by SRSS method.11*P d + D + T + R' + A' + XS h + S ATest 8P t S c P t + D0.9 S yTABLE 3.9B-11 ASME III CLASS 2 AND 3* STRESS ANALYSIS REQUIREMENTS PER NC3650 AND ND3650 LOAD COMBINATIONS FOR ASME III CLAS S 2 AND 3 PIPING EXCEPT QUENCH, RECIRCULATION, AND SI Plant Operating Condition NC 3600 EquationsLoad (Moment) Combinations*Allowable StressNormal/Upset 8P d + DS h 9P p + D + E + H + V + W1.2 S h10(1)T + R + AS A 10aS3 S c11(1)P d + D + T + R + AS h + S AEmergency 9P p + D + H + Y + W1.8 S hFaulted 9P p + D + E' + H + W + A1 + Y'2.4 S hTest 8P t S cTABLE 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 Plant Operating Condition NC 3600 EquationsLoad (Moment) Combinations **Allowable Stress MPS3 UFSAR3.9-117Rev. 30NOTE:*See Table 3.9B-12 for definition of loading.(1)Either the requirements of Equation 10 or 11 must be satisfied.(2)In break exclusion area, sum of stresses given by Equations 9 a nd 10 should not exceed 0.8 (S A + 1.2 S h). In crack exclusion area, sum of stresses given by Equations 9 and 10 should not exceed 0.4 (S A + 1.2S h).(3)Occasional dynamic loads such as waterhammer, steamhammer, safety valve discharge, etc. are combined with seismic inertia by SRSS method.NOTES:*Equation numbers are from 1973 Ed ition ANSI B31.1, dated June 30, 1973.**Refer to Table 3.9B-12 for identification of loadings.
P t + D0.9 S yTABLE 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 Condition    104.8 Equations*Load (Moment) Combinations**Allowable StressDesign11P d + DS hNormal/Upset12P p + D + H + W1.2 S h13T + RS A14P d + D + T + RS h + S A10aS3 S cSeismic*Test11P t + DS c P t S cTABLE 3.9B-11 ASME III CLASS 2 AND 3* STRESS ANALYSIS REQUIREMENTS PER NC3650 AND ND3650 LOAD COMBINATIONS FOR ASME III CLAS S 2 AND 3 PIPING EXCEPT QUENCH, RECIRCULATION, AND SI Plant Operating Condition NC 3600 EquationsLoad (Moment) Combinations*Allowable Stress MPS3 UFSAR3.9-118Rev. 30***See NETM-42 for evaluation of ANSI B31.1 for seismic loading.
MPS3 UFSAR3.9-119Rev. 30TABLE 3.9B-12 LOADINGS APPLI CABLE TO PIPING SYSTEMSD - Sustained mechanical loads, including dead weight of piping, components, contents, and insulation.P - Internal pressure loads.T - Loads due to thermal expansion of the syst em in response to average fluid temperature.R - Loads induced in the piping due to the therma l growth of equipment and/or structures to which the piping is connected as a result of plant normal or upset plant conditions.R' - Loads induced in the piping due to therma l growth of equipment and/or structures to which the piping is connected as a re sult of plant faulted plant conditions.E - Inertia effects of the OBE.E' - Inertia effects of the SSE.A - Loads induced in the piping due to respons e of the connected equipment and/or civil structures to the OBE (commonly refe rred to as OBE anchor movements). A' - Loads induced in the piping due to res ponse of the connected equipment and/or civil structures to the SSE (commonly referred to as SSE movements).S - Loads induced due to building settlement effects.H - Loads resulting from occasional loads other than seismic. Examples of these loads would be: water hammer, steam hammer, opening a nd closing of safety relief valves, etc.L - Local stress effects in piping and/or pi ping components due to sudden changes in fluid temperature. These loads are commonly referred to as thermal transient effects.Y - 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 conditionY' - 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.V - Loads induced by operation of the system other than those loads described above (operational vibration).NOTE: These loads are evaluated duri ng preoperational and testing phaseB - Loads on restraints induced by blowdown a nd subsequent pipe re sponse of a ruptured system for emergency plant conditions.B' - Loads on restraints induced by blowdown and subsequent pipe re sponse of a ruptured system for faulted plant conditions.X - Loads induced in the piping due to pressure response (growth) of the containment during a faulted plant condition. P1 - Internal pressure loads for emerge ncy and faulted condition as applicable.
P d - Internal pressure due to design pressure.
P o - Internal pressure due to range of operating pressure.
P p - Internal pressure due to peak pressure.
P t - Internal pressure due to test pressure.
MPS3 UFSAR3.9-120Rev. 30A1 - Loads induced in piping due to inertia effects of LOCA or displacements due to LOCA or displacements due to pipe rupture.W - Loads imposed by wind, snow, or ice.
MPS3 UFSAR3.9-121Rev. 30TABLE 3.9B-13 ACTIVE PUMPS AND VALVES This table has been deleted. Refer to the Plant Design Data System (PDDS), Seismic Qualification Tracking System (SQT) List for details.
MPS3 UFSAR3.9-122Rev. 30TABLE 3.9B-14 OMITTED MPS3 UFSAR3.9-123Rev. 30TABLE 3.9B-15 POSTULATED PRIMARY LOOP BREAKSBreak No.Location of BreakType of Break1Reactor Vessel Outlet Nozzle1/4 Double-Ended Guillotine2Reactor Vessel Inlet Nozzle1/2 Double-Ended Guillotine3Steam Generator Inlet NozzleDouble-Ended Guillotine4Steam Generator Outlet NozzleDouble-Ended Guillotine 5Reactor Coolant Pump Inlet NozzleDouble-Ended Guillotine6Reactor Coolant Pump Outlet NozzleDouble-Ended Guillotine750-Degree Elbow on the IntradosSplit 8Flow Entrance to the 90-degree Elbow (Steam Generator Side) on the Crossover Leg Double-Ended Guillotine9Residual Heat Removal/Primary Loop ConnectionGuillotine (viewed from the residual heat removal line)10Safety Injection/
Primary Coolant Loop ConnectionGuillotine (viewed from the Safety Injection line)11Pressurizer Surge/Primary Coolant Loop ConnectionGuillotine (viewed from the pressurizer surge line)12Loop Closure Weld in Crossover LegDouble-Ended Guillotine MPS3 UFSAR3.9-124Rev. 30TABLE 3.9B-16 STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPORT SNUBBER EMBEDMENT LOADS (1) (KIPS)ComponentNo.
(2)Normal Operation (3) (4)Faulted Condition (4) (5)F x VM F x VMSteam Generator Upper Snubbers1 or 3668359+1153+59+1144-1443-56-10882 or 48257137+1553+59+1144-1443-56-1088Steam Generator Lower Snubbers11066117+1375+63+1226-988-73-14232749413+969+92+4523-409-78-360731186117+988+63+1226-1375-73-142341056117+988+63+1226-1375-73-1423 Reactor Coolant Pump Snubbers1836297+1168733333
-130721139413+841+89+4362-930-91-45103126636+970+75+3695-1391-86-4395 MPS3 UFSAR3.9-125Rev. 30NOTES:(1)Maximum for all loops. All loads are "
+/-" except where indicated. In those cases "+" means outward, "-" means inward on embedment.(2)Refer to Figure 3.9B-5(3)Includes seismic (OBE) and deadweight.(4)F x = axial, V = resultant shear, M = resultant moment(5)Includes seismic (SSE), deadweight, and pipe rupture MPS3 UFSAR3.9-126Rev. 30NOTES: (1) Refer to Table 3.9-2(2) Fx = axialV = resultant shear (horizontal)M = resultant momentTABLE 3.9B-17 STEAM GENE RATOR AND REACTOR COOLANT PUMP SUPPORT COLUMN EMBEDMENT LOADS (1)(KIPS)ComponentNo. (1)Normal Operation (1) (2)Faulted Condition (1) (2)F x VM F x VMSteam Generator Support Columns5+74NegligibleNegligible+1,197711200-501-13836+74NegligibleNegligible+1479721225-501-14117+74NegligibleNegligible+1354671146-501-12298+74NegligibleNegligible+1455711208-446-1282 Reactor Coolant Pump Columns4+96NegligibleNegligible+19961323039-363-15785+96NegligibleNegligible+11451874293-363-22286+96NegligibleNegligible+15791202767-363-1625 MPS3 UFSARMPS3 UFSAR3.9-127Rev. 30TABLE 3.9B-18 FACTORS OF SAFETY FOR PRIMARY MEMBERS OF S TEAM GENERATORS AND REACTOR COOLANT PUMP SUPPORTS Faulted Condition Support Section Members (1) Material Designation Material Stress (2) (ksi)Actual Stress (ksi)Allowable Stress (3) (ksi)Factor of SafetySteam GeneratorUpper Restraint RingA543 C1.286.9Stress Ratio = 0.741.36Splice PlateA543 C1.286.9Stress Ratio = 0.721.39Connecting Rod Assy.A487 - Gr. 10Q88.0Stress Ratio = 0.771.3CouplingA668-72 C1. L76.8Stress Ratio = 0.891.12 Cylinder Lug. Assy.A487 - Gr. 10Q97.0Stress Ratio = 0.25.0A668-72 Gr. M101.0Stress Ratio = 0.891.12Wall ClevisA487 - Gr. 10Q97.028.9109.03.8 Connecting Rod BoltA668 C1. M95.894.3122.01.36StudsSA-193 Gr. B-1694.022.7 55.92.5YokeA-543-72 C1. 286.935571.6Lower Restraint Ring ClevisA487 - Gr. 10Q83.5Stress Ratio = 0.531.88Snubber Lug Assy.A487 - Gr. 10Q97.0Stress Ratio = 0.2583.88Extension TubeA668 C1. L76.8Stress Ratio = 0.761.32Snubber ClevisA471 C1.5103.4Stress Ratio = 0.244.1Wall MountA487 - Gr. 10Q97.153.3 110.01.98 MPS3 UFSARMPS3 UFSAR3.9-128Rev. 30Vertical ColumnUpper Column ClevisA487 - Gr. 10Q83.5Stress Ratio = 0.521.9Upper Column LugA487 - Gr. 10Q88.0Stress Ratio = 0.782.28Column TubeA668-72 C1. L76.8Stress Ratio = 0.472.1 Floor ClevisA487 - Gr. 10Q97.0Stress Ratio = 0.42.5Lower Column LugA487 - Gr. 10Q97.0Stress Ratio = 0.422.38S/G Cap ScrewsSA-540 - B23 C1. 287.0Stress Ratio = 0.31.8Reactor Coolant PumpLateral RestraintWall MountA487 - Gr. 10Q97.155.3109.31.98Snubber ClevisA487 - Gr. 10Q97.0Stress Ratio = 0.61.7 Extension TubeA668-72 C1. L76.8Stress Ratio = 0.681.47Pump LinkA668-72 C1. M105.2Stress Ratio = 0.8171.22Pump LugA487 - Gr. 10Q88.0Stress Ratio = 0.61.7 Snubber Clevis PinA668-72 C1. M101.058.1110.01.89TABLE 3.9B-18 FACTORS OF SAFETY FOR PRIMARY MEMBERS OF S TEAM GENERATORS AND REACTOR COOLANT PUMP SUPPORTS Faulted Condition Support Section Members (1) Material Designation Material Stress (2) (ksi)Actual Stress (ksi)Allowable Stress (3) (ksi)Factor of Safety MPS3 UFSARMPS3 UFSAR3.9-129Rev. 30NOTES:(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 operati ng conditions, the corresponding design loads were signif icantly lower, thus making the faulted c ondition the critical design condition.Vertical ColumnUpper ClevisA487 - Gr. 10Q88.0Stress Ratio = 0.412.42Upper LugA487 - Gr. 10Q93.2Stress Ratio = 0.721.39Column TubeA-668-72 Gr. L76.8Stress Ratio = 0.731.37 Lower LugA487 - Gr. 10Q93.2Stress Ratio = 0.721.39Floor ClevisA487 - Gr. 10Q97.0Stress Ratio = 0.372.7Tie Down PinA471-70 C1.5102.6Stress Ratio = 0.7531.32PinA668-72 C1.M115.084.298.01.16TABLE 3.9B-18 FACTORS OF SAFETY FOR PRIMARY MEMBERS OF S TEAM GENERATORS AND REACTOR COOLANT PUMP SUPPORTS Faulted Condition Support Section Members (1) Material Designation Material Stress (2) (ksi)Actual Stress (ksi)Allowable Stress (3) (ksi)Factor of Safety MPS3 UFSARMPS3 UFSAR3.9-130Rev. 30NOTES:1Refer to Figure 5.4-13
.2Minimum specified yield or ultimate at temperature.TABLE 3.9B-19 MINIMUM DESIGN MARGINS FOR PRE SSURIZER SUPPORTFaulted Support Section Members (1) Material DesignationMaterial Stress (2) (ksi) Actual Stress (ksi)Allowable Stress (ksi)Factor of Safety Ring GirderSA-516 - Gr. 7036.3Stress Ratio = 0.4820.5 Column PipeSA-106-C39.2Stress Ratio = 0.841.2Upper Vertical ClevisA487 - Gr. 10Q100.Stress Ratio = 0.871.15Lower Vertical ClevisA487 - Gr. 10Q100.20.841.081.97 Upper Vertical LugA487 - Gr. 10Q100.19.839.52.02Horizontal Ring and Wall ClevisesA487 - Gr. 10Q100.11.627.52.36Lower Horizontal Wall LugsSA-533 - Gr. A C1. 149.65.969.81.06Spherical PlateSA-516 - Gr. 7036.8844.25.51Adjusting RodSA-540-B23 - C1.5103.625.349.71.98 Upper Lug Rest.A-543-72 C1. 292.55.277.31.4PinSA-540-B23 C1.2137.116.4143.71.23Embedment PlateSA-516 - Gr. 7036.816.555.23.34 Skirt Anchor BoltsSA-540-B23 C1.213784.8106.11.15 MPS3 UFSAR3.9-131Rev. 30NOTES:(1)A comparison of normal, upset, and faulted loads show that the faulted loads exceed the other loading conditions by a ratio greater than 2.(2)Refer to Figure 5.4-15(3)Minimum yield or ul timate at temperatureTABLE 3.9B-20 DESIGN MARGINS FOR THE PRIMARY MEMBERS OF THE PRESSURIZER SAFETY VALVE SUPPORT Faulted (1)Support Section (2)Material DesignationMaterial Stress (3) (ksi)Actual Stress (ksi)Allowable Stress (ksi)
Factor of SafetyValve Support Flange WeldSA-537-C1.252.225.48.1.9Radial ArmSA-537-C1.252.29.155.76.1Ring GirderSA-537-C1.252.240.55.71.39 ColumnSA-537-C1.247.6Stress Ratio = 0.731.37PinSA-540 B24 C1.2116.11241351.1Support BracketSA-537 C1.246.44.355.21.2 MPS3 UFSARMPS3 UFSAR3.9-132Rev. 30NOTES:(1)Refer to Figures 5.4-9 and 5.4-10.(2)Minimum yield or ul timate at temperatures.(3)Refer to Table 5.4-18
.TABLE 3.9B-21 DESIGN MARGINS FOR PRIMARY MEMBERS OF REACTOR PRESSURE VESSEL SUPPORTNormal OperatingFaulted Support Section Member (1) Material Designation Actual Stress (ksi)Allowable Stress (ksi)Factor of SafetyMaterial Stress(2) (ksi)Actual Stress (ksi)
Allowable Stress (3) (ksi)Factor of SafetyR.V. Vertical Restraint PadA668-72 Gr. Nsmall44.3large110.728.795.33.3R.V. Support GibkeysA668-72 Gr. N19.071.13.7118.566.3101.31.5R.V. Support GibgussetsSA-533-C1.26.039.966.666.639.859.91.56ShellSA-537-C1.210.926.72.4557.529.182.252.8Base FlangeSA-516-7011.1469.96.338.27.068.42.53 Anchor StudsSA-193-B714.962.54.210536.687.52.39R.V. BoltsSA-540-C1.1small70.2large127.749.998.31.97Leveling DeviceA668-72-N2.273.733.512387.51051.2 MPS3 UFSARMPS3 UFSAR3.9-133Rev. 30NOTE:(1)Minimum specified yield at temperatures.TABLE 3.9B-22 DESIGN MARG INS FOR PRIMARY LOOP BUMPER SUPPORT STRUCTURE Primary Loop Bumper SectionMaterial Designation Material Stress (1) (ksi)Max. Actual Stress (ksi) Allowable Stress (ksi)Factor of Safety Cross Over Leg BumperIntermediate StructureASTM 514 Steam Generator Side86.069.570.01.01Pump Side86.050.570.01.38SaddleASTM 514Steam Generator Side77.049.370.01.42Pump Side77.056.970.01.23Hot Leg BumperVertical I-BeamASTM 58843.8Stress ratio = 0.911.1SaddleASTM 51470.465.2 70.41.08 Cold Leg BumperKey RingASTM 51477.015.033.32.2Key SupportASTM 51477.045.870.01.53Key Support RingASTM 51477.055.070.01.27 Attachment BoltsASTM 193 GR B7125 (ult. stress)33.636.171.08 MPS3 UFSAR3.9-134Rev. 30TABLE 3.9N-1
 
==SUMMARY==
OF REACTOR COOLANT SYSTEM DESIGN TRANSIENTS Normal Conditions Occurrences 1.Heatup and cooldown at 100
&deg;F/hr (pressurizer cooldown 200
&deg;/hr)200 (each)2.Unit loading and unloading at 5% of full power/min13,200 (each)3.Step load increase and decrease of 10% of full power2,000 (each)4.Large step load decrease with steam dump 2005.Steady state fluctuations
: a. Initial fluctuations 1.5 x 10 5 b. Random fluctuations 3.0 x 10 66.Feedwater cycling at hot shutdown2,0007.Not Used 8.Unit loading and unloading between 0 and 15% of full power500 (each)9.Boron concentration equalization 26,40010.Refueling 8011.Reduce temperature return to power2,00012.Reactor coolant pumps startup/shutdown 3,80013.Turbine roll test2014.Primary side leak test200 15.Secondary side leak test8016.Tube leakage test 80017.Heaters out of servicea. One heater120b. One bank of heaters120 Upset Conditions Occurrences 1.Loss of load, without immediate reactor trip802.Loss of power (blackout with na tural circulation in the reactor coolant system) 403.Partial loss of flow (loss of one pump)804.Reactor trip from full power MPS3 UFSAR3.9-135Rev. 30NOTE:(1)In accordance with ASME III, emergency and faulted conditions are not included in fatigue evaluation.a. Without cooldown230b. With cooldown, without safety injection160c. With cooldown and safety injection105.Inadvertent reactor coolant depressurization 206.Not Used 7.Control rod drop808.Inadvertent emergency core cooling system actuation609.Operating basis earthquake (20 earthquakes of 20 cycles each)400 10.Excessive feedwater flow 3011.Reactor Coolant System (RCS) Cold Overpressurization10 Emergency Conditions (1)1.Small loss of coolant accident 52.Small steam break5 3.Complete loss of flow5 Faulted Conditions (1)1.Main reactor coolant pipe break (large loss of coolant accident)12.Large steam break13.Feedwater line break1 4.Reactor coolant pump locked rotor15.Control rod ejection16.Steam generator tube rupture1 7.Safe shutdown earthquake1Test Conditions Occurrences 1.Primary side hydrostatic test10 2.Secondary side hydrostatic test10TABLE 3.9N-1
 
==SUMMARY==
OF REACTOR COOLANT SYSTEM DESIGN TRANSIENTS MPS3 UFSAR3.9-136Rev. 30TABLE 3.9N-2 LOADING COMBINATIONS FOR REACTOR COOLANT SYSTEM COMPONENTS Condition ClassificationLoading CombinationDesignDesign pressure, Design temperature, Deadweight, Operating basis earthquakeNormalNormal conditions transients, DeadweightUpsetUpset condition tr ansients, Deadweight, Operating basis earthquakeEmergencyEmergency condition transients, Deadweight FaultedFaulted condition tran sients, Deadweight, Safe shutdown earthquake, Pipe rupture loads MPS3 UFSAR3.9-137Rev. 30TABLE 3.9N-3 ALLOWABLE STRESSES FOR REACTOR COOLANT SYSTEM COMPONENTS Operating Condition ClassificationVessels / Tanks PipingPumpsValvesNormalASME Section III NB-3000ASME Section III NB-3600ASME Section III NB-3400ASME Section III NB-3500UpsetASME Section III NB-3000ASME Section III NB-3600 ASME Section III NB-3400ASME Section III NB-3500EmergencyASME Section III NB-3000ASME Section III NB-3600ASME Section III NB-3400ASME Section III NB-3500FaultedASME Section III Appendix FASME Section III Appendix FASME Section III Appendix F Note 1 NOTE 1 TO TABLE 3.9N-3
:CLASS 1 VALVE FAULTED CONDITION CRITERIA Active Inactive a)Calculate Pm from para. a)Calculate Pm from para.NB3545.1 with InternalNB3545.1 with Internal
 
Pressure P s = 1.25PsPressure P s = 1.50Ps P m  1.5S m P m  2.4S m or 0.7S ub)Calculate S n from para.b)Calculate S n from para.NB3545.2 withNB3545.2 with C p = 1.5Cp = 1.5 P s = 1.25PsP s = 1.50Ps Q t2 = 0Q t 2 = 0 P ed = 1.3X value of PedP ed = 1.3X value of Pedfrom equations offrom equations of3545.2(b) (1)NB3545.2(b) (1)
MPS3 UFSAR3.9-138Rev. 30 S n  3S m S n  3S m P e , P m , P b , Q t , C p , S n , and S m as defined by Section III ASME Code.NOTE 1 TO TABLE 3.9N-3
:
MPS3 UFSAR3.9-139Rev. 30NOTES:(1)Temperature is used to determine allowable stress only.(2)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.TABLE 3.9N-4 DESIGN LOADING COMBIN ATIONS FOR ASME CODE CLASS 2 AND 3 COMPONENTS AND SUPPORTS Loading Combinations (1) (2)Design/Service Level Requirements1.Design pressure, Design temperature, Deadweight, Nozzle loads Design and Normal2.Normal condition pressure, Normal condition metal temperature, Deadweight, Nozzle loads Normal3.Upset condition pressure, Upset condition metal temperature, Deadweight, Nozzle loads, Operating basis earthquake Upset4.Emergency condition pressure, Emergency condition metal temperature, Deadweight, Nozzle loadsEmergency5.Faulted condition pressure, Faulted condition metal temperature, Deadweight, Nozzle loads, Safe shutdown earthquakeFaulted MPS3 UFSAR3.9-140Rev. 30NOTE:(1)Applies for tanks designed in accordance with ASME III, NC-3300.TABLE 3.9N-5 STRESS CRITERIA FOR SAFETY CODE CLASS 2 (1) AND CLASS 3 TANKSDesign/Service LevelStress Limits Design and Normalm  1.0 S (m or L) + b  1.5 SUpsetm  1.1 S (m or L) + b  1.65 SEmergencym  1.5 S (m or b) + b  1.80 S Faultedm  2.0 S (m or L) + b  2.4 S MPS3 UFSAR3.9-141Rev. 30NOTE:(1)Applies for tanks designed in accordance with ASME III, NC-3200.TABLE 3.9N-6 STRESS CRITERIA FOR ASME CODE CLASS 2 TANKS (1)Design/Service LevelStress LimitsDesign and NormalP m  1.0 S m P L  1.5 S m (P m or P L) + P b  1.5 S mUpsetP m  1.1 S m P L  1.65 S m (P m or P L) + P b  1.65 S mEmergencyP m  1.2 S m P L  1.8 S m (P m or P L) + P b 1.8 S mFaultedP m  2.0 S m P L  3.0 S m (P m or P L) + P b  3.0 S m MPS3 UFSAR3.9-142Rev. 30TABLE 3.9N-7 STRESS CRITERIA FOR AS ME CODE CLASS 2 AND CLASS 3 INACTIVE PUMPSDesign/Service Level Stress LimitsDesign/NormalThe pump shall confor m to the requirements of ASME Section III, NC-3400 (or ND-3400) m 1.0S and (m or L) + b <1.5SUpsetm 1.1S (m or L) + b  1.65SEmergency1.5S (m or L) + b  1.80S Faulted m  2.0S (m or m) + b 2.4S MPS3 UFSAR3.9-143Rev. 30NOTES:*Valve nozzle (piping load) stress analysis is not required when both the following conditions are satisfied: (1) the section modul us 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 e qual 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 allo wable stress. If unable to comply with this requirement, an analysis in accordance with the design procedure for Class I valves is an acc eptable alternate method.**The maximum pressure resulting from upset, emergency, or faulted conditions shall not exceed the tabulated factors listed under P max times the design pressure. If these pressure limits are met, the stress limits in this table are considered to be satisfied.TABLE 3.9N-8 STRESS CRITERIA FOR SAFETY RELATED ASME CODE CLASS 2 AND CLASS 3 ACTIVE AND INACTIVE VALVESConditionStress Limits*Pmax**Design and NormalValve bodies shal l conform to ASME Section IIIUpsetm 1.1S 1.1 (m or L) + b 1.65SEmergencym 1.5S 1.2 (m or L) + b 1.80S Faultedm 2.0S 1.5 (m or L) or b 2.4S MPS3 UFSAR3.9-144Rev. 30TABLE 3.9N-9 STRESS CRITERIA FOR ASME CODE CLASS 2 AND 3 PIPING See Section 3.9B, Table 3.9B-11 MPS3 UFSAR3.9-145Rev. 30TABLE 3.9N-10 DESIGN CRITERIA FOR ACTIVE PUMPSLoading CombinationDesign Criteria1. Design ASME Section III Subsection NC-3400 and ND-3400. Normalm 1.0S m + b 1.5S 2. Upsetm 1.0S m + b 1.5S3. Emergency m 1.2S m + b 1.65S 4. Faultedm 1.2S m + b 1.8S MPS3 UFSARMPS3 UFSAR3.9-146Rev. 30NOTES:(*)Only two charging pumps are required for proper safety inject ion operation, while the third pump is considered an installed spare.TABLE 3.9N-11 ACTIVE PUMPSPump Item No.SystemANS Safety Class Normal Mode Post LOCA ModeBasisCentrifugal charging Pump Number 1, 2 and 3APCHCVCS2ON/OFFONRequired for high head safety injection
(*); safe shutdown.
Residual removal Pump Number 1 and 2APRHRHRS2OFFONRequired for safety injection.
Safety injection Pump Number 1 and 2APSISIS2OFFONRequired for safety injection. Boric acid transfer Pumps 1 and 2APBACVCS3ON/OFFOFFRequired for reactor shutdown.
MPS3 UFSAR3.9-147Rev. 30TABLE 3.9N-12 ACTIVE VALVES This table has been deleted. Refer to the Plant Design Data System (PDDS) Seismic Qualification Tracking System (SQT).
MPS3 UFSAR3.9-148Rev. 30TABLE 3.9N-13 MAXIMUM DEFLECTIONS ALLOWED FOR REACTOR INTERNAL SUPPORT STRUCTURES ComponentAllowable Deflections (inches)No-Loss-of Function Deflections (inches)Upper barrel, Radial inward4.18.2Upper barrel, Radial outward 1.01.0 Upper package 0.100.15Rod cluster guide tubes 1.001.75 MPS3 UFSAR3.10-1Rev. 30 3.10 SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENTThis section presents information to demonstrate that instrumentation and electrical equipment classified as Seismic Category I is capable of performing designated safety related functions in the event of an earthquake. The information presented includes identification of the Category IE instrumentation and electrical e quipment that are within the scope of Westinghouse nuclear steam supply system (NSSS), and balance of plant (BOP) scope. The qualification criteria employed for each item of equipment, the designated safety related requirements, definition of the applicable seismic environment, and documentation of the qualification process employed to demonstrate the required seismic capability ar e described in this section.Sections whose identification numbers include the letter B contain material within the balance-of-plant (BOP) scope, while sections whose identification numbers include the letter N contain material within the nuclear steam supply system (NSSS) scope.
MPS3 UFSAR3.10B-2Rev. 30 3.10B.1 SEISMIC QUALIFICATION CRITERIA The methods of meeting the general requireme nts for seismic qualifi cation of Category I instrumentation and electrical equipment as described by Genera l Design Criteria (GDC) 1, 2, and 23 are described in Section 3.1. The general me thods of implementing the requirements of Appendix B to 10 CFR Part 50 are described in Chapter 17.
Seismic Category I instrumentation and electri cal equipment are designed to maintain the capability to:1.Initiate a protective action during a safe shutdown earthquake (SSE)2.Withstand seismic disturbances during post-accident operati on without loss of safety function Safety related instrumentation and electrical e quipment is seismically qualified in accordance with general instructions for earthquake require ments (Section 3.7B.3.1). In order to prevent any threat of impacting damage to Class IE equi pment during seismic events, nonsafety related instrumentation and electrical equipment locat ed adjacent to Class IE equipment is also seismically qualified. The earthquake requirement s and qualification methods conform to those outlined in IEEE Standard 3 44-1975, "IEEE Recommended Practices for Seismic Qualifications of Class IE Equipment for Nuclear Power Generating Stations," (Section 1.8, R.G. 1.100) and are in agreement with the recommendations of Branch Technical Po sition EICSB 10. Instrumentation and electrical equipment are test ed as individual components, e ither as part of a simulated structural section or as part of a completely assembled module or unit.
3.10B.2 METHODS AND PROCEDURES FOR QU ALIFYING ELECTRICAL EQUIPMENT AND INSTRUMENTATION Methods and procedures for qualifying electrical equipm ent and instrumentati on are contained in Section 3.7B.3.1 and 3.10B.4.
The response of racks, panels, cabinets, and cons oles is considered in assessing the seismic capability of instrumentation and electrical equipment. Mount ed equipment is qualified, as a minimum, to acceleration levels consistent wi th those transmitted by supporting structures. A design objective is to minimize amplification of floor acceleration by supporting members on mounted equipment.
Qualification of seismically related equipment is accomplished by the following methods:1.Test the equipment under simulated operating conditions2.Analyze the equipment for ve rification of proper function3.Combinations of test and analysis MPS3 UFSAR3.10B-3Rev. 30 The decision to qualify a component using one or a combination of the above methods is a judgement of the capability to perform the qualification based on the complexity of the equipment.
3.10B.3 METHODS AND PROCEDURES OF ANALYSIS OR TESTING OF SUPPORTS OF ELECTRICAL EQUIPMENT AND INSTRUMENTATION Supports for Category I electrical equipment, instrumentation, and control systems are seismically qualified by the analysis and testing procedures outlined in Section 3.7B.3.1. Supports are designed to withstand the combined effects of normal operating load s acting simultaneously with two orthogonal horizontal and the vertical components of earthquake loading without loss of functional capability or structural integrity as a pplicable. The stress levels due to the combined loading conditions do not exceed the maximum stress levels permitted under applicable codes. If there are no applicable codes, the stress le vel under combined loading for 1/2 SSE does not exceed 75 percent of the minimum yield strength of the material at service temperature per the ASTM specification. Under safe shutdown earthquake, the stress level does not exceed the smaller of 100 percent of the specified minimum yield strength or 70 percent of the specified ultimate strength of the material at temperature per ASTM specification.
A design objective is to provide supports for elect rical equipment, instrumentation, and control systems that are seismically rigid, i.e., with fundamental natural frequencies above the cutoff frequency, which separates the rela tively flat rigid range of a response curve from the resonant range of the relevant amplified response spectra curves. This assures that amplification of floor accelerations through suppor ting members to mounted equipment is minimized.
The dynamic analysis method is typically used to establish support spacing for cable trays and conduit. Additionally, restraints are used as ne cessary to limit the horizontal loads to allowable design values established on the basis of racew ay loading and unsupported span lengths. Design provisions for significant differential motions be tween buildings are made by breaks in raceways if these relative displacements would result in unacceptable equipmen t or support loadings.
In lieu of dynamic analysis, a provision is made fo r the use of static analysis in qualification of cable tray systems and conduit support systems. For use in the static analysis, the peak resonant "G" values from the ARS are amplified in accordance with Section 3.7. The damping values used for qualification of cable tray systems are 4 percent for 1/2 SSE and 8 percent for SSE. The damping values used for the qualification of conduit support systems are 4 percent for 1/2 SSE and 7 percent for SSE.
3.10B.4 REPLACEMENT ITEMS Methodologies for demonstrating equipment seismi c qualification have been established for the implementation of GDC-2. Qualification methods ha ve changed with time, resulting in various seismic FSAR commitments.IEEE 934-1987 specifically provides seismic guidance for replacement items and states that parts shall be selected and installed as to preserve the original seismic qualification of Class 1E MPS3 UFSAR3.10B-4Rev. 30equipment. Methods to be employed for seismic qualification of replacement electrical equipment are documented in IEEE 344-75 and IEEE 344-87 (Endorsed by Reg Guide 1.100, Rev. 2). Either of these documents may be utilized for plant design changes and modi fications. Replacement equipment originally qualified by Westinghouse to IEEE 344-71 meet the original criteria or either IEEE 344-75 or IEEE 344-87. The use of either method will ensure the original qualification basis remain unchanged. The use of seismic expe rience data may be employed for procurement of seismically in sensitive and rugged components.
In cases where a replacement item is not provi ded with documentation of seismic qualification which is identical to the original item, an equivalency evaluation must be performed. The equivalency evaluation must document the met hodology utilized for tec hnically evaluating the replacement item and provide justification that the item meets the seismic performance requirements necessary to maintain the seismic design basis.
3.10B.5 OPERATING LICENSE REVIEW The MP3 Plant Design Data System provide s the information to demonstrate proper implementation of the criteria in Section 3.10B
.1 and 3.10B.4 and demonstr ate adequate seismic qualification of Seismic Ca tegory I instrumentation and electrical equipment. Maintenance of this data is in accordance with the Standard Review Plan Section 3.10.
MPS3 UFSAR3.10N-5Rev. 30 3.10N SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT This section presents information to demonstrat e that instrumentation and electrical equipment classified as Seismic Category I is capable of performing designated safety related functions in the event of an earthquake. The information presented includes id entification of the Category IE instrumentation and electrical equipment that are within the scope of Westinghouse nuclear steam supply system (NSSS), and balance of plant (BOP) scope. The qualification criteria employed for each item of equipment, the designated safety re lated requirements, definition of the applicable seismic environment and documentation of the qualification process employed to demonstrate the required seismic capability ar e described in this section.Sections whose identification numbers include the letter B contain material within the BOP scope, while sections whose identificat ion numbers include the letter N contain material within the NSSS scope.
3.10N.1 SEISMIC QUALIFICATION CRITERIA 3.10N.1.1 Qualification Standards The methods of meeting the general requireme nts for seismic qualifi cation of Category I instrumentation and electrical equipment as described by Genera l Design Criteria (GDC) 1, 2, and 23 are described in Section 3.1. The general me thods of implementing the requirements of Appendix B to 10 CFR Part 50 are described in Chapter 17.The Commission's recommendati ons concerning the methods to be employed for seismic qualification of electri cal equipment are contained in Regulatory Guide 1.100, which endorses IEEE-344-1975. Westinghouse meets th is standard, as modified by Regulatory Guide 1.100, by either type of test, analysis, or an appropriate combination of these methods. Westinghouse meets this commitment employing the methodology described in the final staff approved version of WCAP-8587, Rev. 2 for all Seismic Category I instrumentation and el ectrical equipment.
3.10N.1.2 Performance Requirements for Seismic Qualification Equipment Qualification Data Packages (WCAP-8587) contains an equipment qualification data package (EQDP) for every item of instrumentation and electrical equipment classified as Seismic Category I within the Westinghouse NSSS scope of supply. Table 3.10N-1 identifies the Category I equipment supplied by Westinghouse for this appl ication and references the applicable EQDP contained in Supplement 1 of WCAP-8587. Each EQDP in Suppl ement 1 contains a section entitled "Performance Specificati on." This specification establishe s the safety related functional requirements of the equipment to be demonstrated during and after a seismic event. The required response spectrum (RSS) employed by Westinghouse for generic seismic qu alification is also identified in the specification, as applicable.
The spectra employed have been selected to envelope the plant specific sp ectra defined in Section 3.7.
MPS3 UFSAR3.10N-6Rev. 30 3.10N.1.3 Acceptance Criteria Seismic qualification must demonstrate that Category I instrument ation and electrical equipment is capable of performing designate d safety related functions during and after an earthquake of magnitude up to and including the safe shutdown earthquake (SSE). Any spurious actuation must not result in consequences adverse to safety. The qualification must also demonstrate the structural integrity of mechanical supports and structures at the operating basis earthquake (OBE) level. Some permanent mechanical deformation of supports and structures is acceptable at the SSE level provided that the ability to perform the designated safety related functions is not impaired.3.10N.2 METHODS AND PROCEDURES FOR QUAL IFYING ELECTRIC AL EQUIPMENT AND INSTRUMENTATIONIn accordance with IEEE 344-1975, seismic qualification of safety re lated electrical equipment is demonstrated by either type test ing, analysis, or a combination of these methods. The choice of qualification method employed by Westinghouse for a particular item of equipment is based upon many factors including practicability, complexity of equipment, economics, availability of previous seismic qualification to earlier standards, etc. The qualification method employed for a particular item of equi pment is identified in the individual equipment qualification data packages (EQDPs) of equipment qualifica tion data packages (WCAP-8587).
3.10N.2.1 Seismic Qualification by Type TestFrom 1969 to mid-1974 Westinghouse seismic test procedures employed si ngle axis sine-beat inputs in accordance with IEEE 344-1971 to seis mically qualify equipm ent. The input form selected by Westinghouse was chosen following an investigation of building responses to seismic events as reported in WCAP-7558 (1971). In addition, Westinghouse has conducted seismic retesting of certain items of equipment as pa rt of the Supplemental Qualification Program (NS-CE-692). This retesting was performed at the request of the NRC staff on agreed upon selected items of equipment employing multi-frequency, multi-axis test inputs (WCAP-8695, 1975) to demonstrate the conservatis m of the original sine-beat te st method with respect to the modified methods of testing for comp lex equipment recommended by IEEE 344-1975.
The original single axis sine-beat testi ng (WCAP-7821, 1971) and the additional retesting completed under the Supplemental Test Program ha s been the subject of generic review by the staff. For equipment which has been previously qualified by the single axis sine-beat method and included in the NRC seismic audit and, where required by the staff, the Supplemental Qualification Program, no additional qualification testing is required to demonstrate acceptability to IEEE 344-1975 since the Westinghouse aging evaluation program was implemented to determine aging effects on complex electronic e quipment located outside containment. This equipment is identified in WCAP-8587, Rev. 2 (1979), Table 7.1, and the test results in the applicable EQDPs of WCAP-8587. Subprogram C of the Westi nghouse aging evaluation program (Appendix B, WCAP-8587) has incorporated a repr esentative sample of components from these systems which use complex electronic equipment.
This program is completed and is reported in WCAP-8687, Supplement 2, Appendix A2. Subprogram C demonstrates that during the qualified MPS3 UFSAR3.10N-7Rev. 30 life there are no in-service aging mechanisms capable of reducing th e capability of these systems to perform during or after a seismic event.
For equipment tests after July 1974 (i.e., new de signs, equipment not previously qualified, or previously qualified equipment that does not meet 1, 2, and 3 above), seismic qualification by test is performed in accordance with IEEE 344-1975. Wh ere testing is utilized, multi-frequency, multi-axis inputs are developed by the genera l procedures outlined in WCAP-8695 (1975). The test results contained in the individual EQDPs of WCAP-8587 demonstrate that the measured test response spectrum envelopes the applicable required response spect rum (RSS) defined for generic testing as specified in Secti on 1 of the EQDP (WCAP-8587). Qual ification for plant specific use is established by verification that the generic RRS specified by Westinghouse envelopes the applicable plant specific respons e spectrum. Alternative test methods, such as single frequency, single axis inputs, are used in selected cases as permi tted by IEEE 344-1975 and Regulatory Guide 1.100.
3.10N.2.2 Seismic Qualification by Analysis The structural integrity of safety related motors (Table 3.10N-1 EQDP-AE-2 and AE-3) is demonstrated by a static seismic analysis in accordance with IEEE 344-1975, with justification. Should analysis fail to show the resonant frequency to be significantly greater than 33 Hz, a test is performed to establish the motor resonant frequency. Motor operability during a seismic event is demonstrated by calculating critical deflections, loads, and stresses under various combinations of seismic, gravitational, and operational loads. The worst case (maximum) values calculated are tabulated against the allowable values. On co mbining these stresses, the most unfavorable possibilities are considered in the following areas:1.Maximum rotor deflection2.Maximum shaft stresses3.Maximum bearing load and shaft slope at the bearings4.Maximum stresses in th e stator core welds5.Maximum stresses in the st ator core to frame welds6.Maximum stresses in the motor mounting bolts7.Maximum stress in the motor feetWhere minor differences exist between items of e quipment, analysis is employed to demonstrate that the test results obtained for one piece of equipment are equally applicable to a similar piece of equipment.
The analytical models employed and the results of the analysis are described in Section 4 of the Equipment Qualification Da ta Packages (WCAP-8587).
MPS3 UFSAR3.10N-8Rev. 30 3.10N.3 METHOD AND PROCEDURES FOR QUALIFYING SUPPORTS OF ELECTRICAL EQUIPMENT AND INSTRUMENTATION Where supports for the electrical equipment and instrumentation are within the Westinghouse NSSS scope of supply, the seismic qualification tests and/or analyses are conducted including the supplied supports. The EQDPs contained in WCAP-8587 identify the equipment mounting employed for qualification purpos es and establish interface requi rements for the equipment to ensure that subsequent inplant installation does not affect the qualification established by Westinghouse.
3.10N.4 OPERATING LICENSE REVIEW The results of tests and analyses that ensure th at the criteria establis hed in Section 3.10N.1 have been satisfied employing the qualification methods described in Section 3.10N.2 and 3.10N.3 are included in the individual EQDPs contained in WCAP-8587.
MPS3 UFSAR3.10-9Rev. 30 3.10.1 REPLACEMENT ITEMS Replacement items are procured and doc umented as shown in Section 3.10B.4.
3.
 
==10.2 REFERENCES==
FOR SECTION 3.103.10-1NS-CE-692. Letter of July 10, 1975 from C. Eicheldinger (Westinghouse) to D.B. Vasselo (NRC).3.10-2WCAP-7536-L (Proprietary) and WCAP-7821 (Non-proprietary), 1971 plus Supplements 1-6, "Seismic Testing of El ectrical and Control Equipment (High Seismic Plants),"
Povochnik, L.M. et al.3.10-3WCAP-7558, 1971. "Seismic Vibration Te sting with Sine Beats." Morrone, A.3.10-4WCAP-8587, Revision 2. 1979, "Methodology for Qualifying Westinghouse WRD Supplied NSSS Safety Related Electrical Equipment," Butterworth, G. and Miller, R.B.3.10-5WCAP-8587, 1978, "Equipment Qualificati on Data Packages," Supplement 1 1978.3.10-6WCAP-8634 (Proprietary) 1975 and WCAP-8695 (Non-proprietary), "General Method of Developing Multi-Frequency Biaxial Test Inputs for Bistables," Jarecki, S.J.3.10-7EPRI NP-7484: "Guideline for Seismic Tec hnical Evaluation of Replacement Items for Nuclear Power Plants."3.10-8EPRI TR-104871: "Generic Seismic Technica l Evaluations of Replacement Items for Nuclear Power Plants."
MPS3 UFSAR3.10-10Rev. 301. Equipment Qualification Data PackageTABLE 3.10N-1 SEISMIC CATEGORY I IN STRUMENTATION AND ELECTRICAL EQUIPMENT IN WESTINGHOUSE NSSS SCOPE OF SUPPLYEquipmentEQD (1)Safety-Related Valve Electric Motor OperatorsEQDP-HE-4Garrett Power-Operated Relief Valves (PORV)EQDP-HE-9Solenoid-Operated Isolation ValveEQDP-HE-10A Electronic Control ModuleEQDP-HE-10BModulating ValveEQDP-HE-10CLarge Pump Motors (Outside Containment)EQDP-AE-2 Canned Pump Motors (Outside Containment)EQDP-AE-3Pressure TransmittersEQDP-ESE-1BDifferential Pressure TransmittersEQDP-ESE-3B Resistance Temperature DetectorsEQDP-ESE-5 and 6Excore Neutron Detectors EQDP-ESE-8 and 9Nuclear Instrumentation System (NIS)EQDP-ESE-10
 
NIS Source Range PreamplifierEQDP-ESE-11 Operator Interface Modules (OIMs)EQDP-ESE-12AProcess Protection SetsEQDP-ESE-13
 
Indicators, Post-Accident MonitoringEQDP-ESE-14Recorders, Post-Accident MonitoringEQDP-ESE-15Solid-State Protection System and Safeguard Test Cabinet (2 Train)EQDP-ESE-16Reactor Trip SwitchgearEQDP-ESE-20Loop Stop Valve CabinetEQDP-ESE-23Reactor Coolant Pump Speed SensorEQDP-ESE-24 Differential Pressure Indicating Switch (Group B)EQDP-ESE-40Incore ThermocoupleEQDP-ESE-43AThermocouple Connectors and Thermocouple SpliceWCAP-10919 Incore Thermocouple Reference Junction Box SpliceEQDP-ESE-43CIncore Thermocouple Reference Junction BoxEQDP-ESE-44A MPS3 UFSAR3.11-1Rev. 303.11 ENVIRONMENTAL DESIGN OF MECH ANICAL AND ELECTRICAL EQUIPMENTElectrical equipment qualification is an integral part in the de sign, construction, and operation of Millstone Unit 3. The U.S. Nuclear Regulatory Commission's regulations in 10 CFR Part 50, "Domestic Licensing of Productio n and Utilization Facilities" require that categories of structures, systems, and components be designed to accommodate the effects of both normal and accident plant environmental conditions, and that control measures be employed to ensure the adequacy of design. Specific requirements pertaining to environmental qualification of certain categories of electrical equipment are embodied in Section 50.49 of 10 CFR Part 50, "Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants."The Millstone 3 Electrical Equipment Qualification (EEQ) Program complies with 10 CFR 50.49. The EEQ Program ensures the continued qualification of equipment that must function during and following design conditions postulated for design basis accidents and the post-accident duration.The constituent parts of the EEQ Program include the program basis, verification of equipment operability during and following exposure to plant environm ental conditions, and proper installation and maintenance of equipment in the plant. These elements are controlled through a set of administrative documents consisting of a program description, im plementing procedures, and reference documents. The procedures are retrievable through the Station's electronic procedure management system.Program output is generated by E quipment Qualification Records. These documents provide the auditable bases and evidence, which demonstrate that Millstone 3 is compliant with 10 CFR 50.49. EQRs utilize two main sources of design input - Test Report Assessments and an Environmental Specification.Test Report Assessments (TRA) are design calculations that evaluate and summarize the environmental qualification test report(s). The TRA documents the process of assessing a qualification test report, or analysis, as acceptable qualific ation documentation for use in the EEQ Program.The Environmental Specification provides the environmental conditions that are required to qualify electrical equipment in performing its function while exposed to normal and accident operating conditions; and provides the environmental design conditions for use as input in development of Equipment Qualification Records.Seismic qualification of safety-related mechanical and electrical equipment is presented in Sections 3.9 and 3.10, respectively.THE INFORMATION PROVIDED IN SECTIONS 3.11, 3.11B, AND 3.11N BELOW REPRESENTS HISTORICAL EEQ PROGRAM DETAILS, WHICH IS NOT SUBJECT TO FUTURE UPDATING.
MPS3 UFSAR HISTORICAL, NOT SUBJECT TO FUTURE UPDATING3.11-2Rev. 30This section presents information to demonstrate that the safety-related electrical equipment is capable of performing designated safety-related functions while exposed to applicable normal, abnormal, test, accident, and post-accident environmental conditions. The information presented includes the definition of the applicable environmental parameters, and description of the qualification process employed to demonstrate the required e nvironmental capability. The seismic qualification of safety-related mechanical and electrical equipment is presented in Sections 3.9and 3.10, respectively.Sections whose identification numbers include the letter B contain material within the balance-of-plant (BOP) scope, while sections whose identification numbers include the letter N contain material within the nuclear steam supply system (NSSS) scope.
MPS3 UFSAR HISTORICAL, NOT SUBJECT TO FUTURE UPDATING3.11B-3Rev. 303.11B.1 EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL CONDITIONS Safety-related equipment and components are qua lified to meet their performance requirements under normal, abnormal, and accident operating conditions. Evidence of qualification is presented in the form of an equipment qualification record (EQR), located in Specification (SP-M3-EE-0353), that reflects the equipment functional capability during the described conditions and required time frame.
The EQR also demonstrates th at the equipment remains in a safe mode after its safety functions are performed.
Since environmental conditions vary for different areas of the plant, there are several environmental zones. The environmental conditions to which the equipment is exposed depend on the environmental zone in which the equipment is located. The sa fety-related equipment located within each environmental zone can be determined from a sort of the information contained in the EEQ Master List (loc ated in SP-M3-EE-0353).
Engineering Specification SP-M3-EE-0333 provides a listing of the worst-case environmental conditions and profile figures for various areas in the plant and a listing of the environmental zones for various areas in the plant. These cond itions are determined by the criteria defined in Section 3.11B.1.1.3.11B.1.1 Environmental Conditions Environmental design conditions are specified in Engineer ing Specification SP-M3-EE-0333.
These environmental conditi ons are listed by a system of zones, each defining a specific area or areas in the plant. The environmental design c onditions in each zone are given for normal, abnormal, and accident conditions as applicable. A further description of each of these conditions is also included; for example, a one-time a ccident environment due to LOCA and MSLB.
Environmental parameters include temperature, pressure, relative humidity, chemicals, spray potential, submergence potential, accident duration, and gamma/beta (where applicable) radiation dose. Where applicable, these parameters ar e given in terms of a time-based profile.
The Equipment Qualification Program does not establish a fixed limitation on component service life to account for the aging effects of non-se ismic vibration. Rather than establish a fixed limitation, the program relies on two complimentary elements to address this aging mechanism. Initially, the seismic qualificat ion process provides a level of confidence that safety related components possess sufficient ruggedness to withstand a vibration environment. Thereafter, regular in-service surveillance and preventative maintenance diagnostics monitor vibration behavior thereby providing early warning agai nst premature service life expiration caused by non-seismic vibration environments.
Normal operating environmental conditions are de fined as conditions expe cted during routine plant operations.
Abnormal operating conditions are any deviations from normal conditions, but do not include accident conditions. These envir onmental conditions include transients cau sed by a fault in a MPS3 UFSAR HISTORICAL, NOT SUBJECT TO FUTURE UPDATING3.11B-4Rev. 30 system component requiring its isol ation from the system, transien ts caused by a loss of load or power, or any system upset not resulting in a forced outage.
Accident environmental conditions result from design basis acci dents (DBA), as described in Section 6.2 and Chapter 15. Safety-related equipmen t functions as required during all normal, abnormal, and accident events.3.11B.1.2 Equipment IdentificationAll safety-related equipment and components for each Class 1E specification, located throughout the plant, are listed in the EQRs which are in separate reports located in Specification (SP-M3-EE-0353). Each device specified is labeled, in the EQRs, with its plant identification number, location, and environmental zone.3.11B.2 QUALIFICATION TESTS AND ANALYSES The EQRs, listed in separate reports, detail the Millstone 3 compliance with 10 CFR 50.49 and NUREG-0588, Revision 1, as endorsed by Regulat ory Guide 1.89, Revision 1. The requirements of General Design Criterion 1 (GDC 1) of 10 CFR 50, Appendix A are achieved by incorporating performance, design, construction, and testing requirements into equipment specifications, and by the establishment of a system of reviews to ensu re conformance with the specified requirements. Appropriate auditable records are maintained in a permanent file. Chapter 17 provides a further definition of compliance to Crit erion III of Appendix B, 10 CFR 50.
Protection against earthquakes, as required by GDC 2, is provided by incorporating a description of seismically induced vibratio ns in equipment specifications and by requiring qualification in accordance with Section 3.10.
The environmental requirements of GDC 4 are addressed in Section 3.11B.1. The Class 1E equipment meets the requirements of GDC 4. The equipm ent is designed to operate satisfactorily or to fail in a safe mode. Since components ar e procured for their ability to withstand the environments resulting from both abnormal even ts and accidents, the Millstone 3 Class 1E equipment meets the re quirements of GDC 23.
GDC 50 requirements are achieved by analysis and testing of pr essure boundary components to ensure containment integrity. Inservice inspection is performed to demonstrate leaktight integrity of components, such as seals and seats.3.11B.2.1 Regulatory Guides The recommendations provided in the following list of regulator y guides have been included in appropriate equipment specifications. A detail ed discussion on compliance with the following regulatory guides is provided in Section 1.8:1.Regulatory Guide 1.30 - Quality Assuranc e Requirements for the Installation, Inspection, and Testing of Instrume ntation and Electric Equipment MPS3 UFSAR HISTORICAL, NOT SUBJECT TO FUTURE UPDATING3.11B-5Rev. 302.Regulatory Guide 1.40 - Qualification Test s of Continuous Duty Motors Installed Inside the Containment of Water-Cooled Nuclear Power Plants3.Regulatory Guide 1.63 - Electrical Pene tration Assemblies in Containment Structures for Water-Cooled Nuclear Power Plants4.Regulatory Guide 1.73 - Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plants5.Regulatory Guide 1.89 - Qualification of Class 1E Equipment for Nuclear Power Plants6.Regulatory Guide 1.131 - Qualification Tests of Electric Cables, Field Splices, and Connections for Light-Water-C ooled Nuclear Power Plants3.11B.2.2 Safety-Related (Class 1E) Equipm ent and Component QualificationsSafety-related equipment located in a mild environment is environmenta lly qualified based on the following:1.equipment is purchased based on th e normal and abnormal environmental conditions in which the equipment is require d to function;2.a periodic maintenance, inspection, and/
or replacement program based on sound engineering practice and r ecommendations 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; and4.An equipment surveillance program which includes periodic inspections, analysis of equipment and component failures, and a review of the results of preventive maintenance and period ic testing programs.A mild environment is an environment that would at no time be significantly more severe than the environment that would occur during normal pl ant operation or during anticipated operational occurrences. This type of environment is not in the scope of the EEQ program per 10 CFR 50.49(c)(3).
Environmental qualification of al l safety-related equipment meets the requirements of IEEE Std. 323-1974, Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations; the intent of NUREG-0588, Interim Staff Position on Environmental Qualification Electrical equipment within the scope of the EEQ Program, has a designated qualified life derived in accordance with IEEE Std. 323-1974 by type test, or combination of type test/analysis and/or documented operating experience.
MPS3 UFSAR HISTORICAL, NOT SUBJECT TO FUTURE UPDATING3.11B-6Rev. 30 Equipment specifications for com ponents within the EQ Program, de fine at least the worst case envelope service condition anticipated for partic ular types of Class 1E electrical equipment.Qualification by Type Testing or Combination of Type Test/Analysis and/or Documented Operating ExperienceTesting is the preferred method of qualification. Analysis has been used to verify or supplement test results.
Service conditions simulated during the test were reviewed to ensure that they enveloped the accident environments. The test duration and environmental paramete rs utilized in the test were reviewed to ensure that they equaled or exceeded specified values. When tests were found to be less severe than specified, a determination of the adequacy of the test is made on a case-by-case basis.The test specimen model, design, and construction material are reviewed against the equipment being qualified to verify a pplicability of test results.
The selected test sequence is re viewed and, when determined not to be in accordance with the guidelines of IEEE Std. 323-1974, is reevaluated for adequacy.Tests which are successful using components th at had not been pre-aged are considered acceptable, provided the components do not cont ain materials known to be susceptible to significant degradation due to aging effects.
Operational modes tested are revi ewed to ensure that they are representative of the actual application requirements as defi ned in the procurement documents. The length of time that each item of equipment is required to operate is reviewed against te st data to ensure that the equipment's designated life is greater than the length of time the equipment is required to operate.
Any failures identified during the environmental qualification effort are reviewed and evaluated relative to their effect on the ability of the component to perform its required function. If a component fails at any time dur ing the test, the applicability of the test with regard to demonstrating the ability of the co mponent to function for the entire period prior to the failure is considered on a case-by-case basis.
Where seals are included as part of the component, the test results and conclusions include the seals. Materials used for termin ating cables and simila r components are addressed separately in the EQR.Qualification by a combination of methods (test, evaluation, analysis) is id entified in the EQR. A determination of the adequacy of the qualification methods used is made on a case-by-case basis.
MPS3 UFSAR HISTORICAL, NOT SUBJECT TO FUTURE UPDATING3.11B-7Rev. 30 Replacement Program For equipment with a designated life less than design plant life, a maintenance surveillance replacement program based upon test da ta and analysis is utilized to lengthen the qualified life of all equipment to that of the plant design life. When new test data becomes available, it is used to modify the program as necessary.
Aging Considerations made for materi als which are susceptible to thermal, vibrat ion, electrical, mechanical, and/or radia tion aging are addressed in the appl icable Environmental Qualification documentation. The order of application of the simu lated aging conditions is reviewed to ensure that the most severe sequence is applied. Known synergistic effects, or those found during testing, are investigated on a case-by-case basis.
Equipment, within the EQ Program is assessed fo r operability at specif ied time intervals by systematic application of the plant periodic test program. Wh en this program detects any equipment degradation, the equi pment or component is analyze
: d. Any equipment or component found to have unsatisfactory aging characteristic s is reviewed to dete rmine necessary action.
Equipment or component operability is defined as the ca pability to perform its specified function.Margin Qualification type tests are revi ewed to verify that adequate margin exists between the most severe specified se rvice conditions of the plant and the c onditions used in type testing. This accounts for normal variations in commercial producti on of equipment and reasonable errors in defining satisfactory performance. Increased levels of testing, number of test cycles, and test duration are considered as methods of ensuring adequate margin.Submergence Equipment has been evaluated for submergence as a result of Design Ba sis Accidents (e.g. LOCA or HELB). Where submergence is possible, it is given as an environmental parameter in Appendix 3B. Equipment, within the EQ Program, located in areas with submergence potential, is either qualified to operate submerged, or it is el ectrically isolated, is admi nistratively controlled, or its location has been verified to be above the submergence level.
Documentation During the review of submitted qualification documentation, when it is determined that actual test data are not submitted (i.e., test summaries and/or certificates of conformance only are submitted), requests for act ual test data are initiated. If such data are not submitted because the data are considered proprietary by the manufacturer, an audit of these data are made. A certificate of conformance by itself is not acceptable unless it is accompanied by verification of test data and information on the qualification program.
MPS3 UFSAR HISTORICAL, NOT SUBJECT TO FUTURE UPDATING3.11B-8Rev. 30 A final determination on the accep tability of qualification docum entation is made based on the criteria that the documentation available is organized in an auditable form and sufficient to justify the conclusions reached.3.11B.3 QUALIFICATION TEST RESULTS This section provides a discussion of the test results associated with environmental qualification of equipment.3.11B.3.1 Nuclear Steam Supply System E quipment Qualification ProgramSection 3.11N describes the Nuclear Steam S upply System Equipment (NSSS) Qualification Program.3.11B.3.2 Qualification of Safety-Related Equipment (BOP) not covered by Section 3.11
.Qualification test results a nd supporting documentation for sa fety-related, non-NSSS equipment are summarized in the EQR (Section 1.7).3.11B.4 LOSS OF VENTILATION The following design features precl ude the possibility of a total system failure for ventilation systems serving areas where equipment require d to function during and following a DBA is located.1.All HVAC systems serving these equi pment 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 valv es and dampers are described in Section 9.4. Valves or dampers required for system ope ration after postulated accidents fail in the safe position.5.Equipment outside the containment bui lding required to operate following a LOCA or a high-energy pipe break is so lo cated 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 MPS3 UFSAR HISTORICAL, NOT SUBJECT TO FUTURE UPDATING3.11B-9Rev. 30 malfunction, the operator has the capability to switch manually to the HVAC standby equipment.
Based upon the above features and the detailed HVAC systems' evaluations in Section 9.4, only partial loss of the ventilation or air-conditioning system could occur in areas where equipment required to function during and/
or following a DBA is located. This loss would not adversely affect the availability of the safety-related equipment to func tion during and following a DBA. The effects of any partial loss of HVAC is re flected in the environments of Engineering Specification SP-M3-EE-0333.3.11B.5 CHEMICAL AND RADIATION ENVIRONMENTComponents of safety-related systems and their associated instrumentation and electrical equipment located inside the containment structure and elsewhere, which are required to function during and subsequent to an ac cident, are designed to operate under normal, abnormal, accident, and post-accident environmental conditions that may occur at their installation locations.3.11B.5.1 Radiation Environment Equipment is qualified for the radiation doses found in Engineering Specification SP-M3-EE-0333. These doses were calculated using source te rms given in Regulatory Guides 1.7 and 1.89, and NUREG 0588.
Assumptions include:1.For the design basis loss of coolant acci dent, which completely depressurizes the primary system, 100 percent of the core noble gases, 50 percent of the halogens, and 1 percent of the soli d 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 fissi on 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 insi de containment is based on the infinite medium dose. No beta dose ha s been considered for components that are sealed in shields which would preclude pene tration of the beta particles.4.LOCA dose values are derived from inte gration for a period of 1 year from the commencement of the accident. For HE LB and fuel handling accidents, integration is over 1 month.
MPS3 UFSAR HISTORICAL, NOT SUBJECT TO FUTURE UPDATING3.11B-10Rev. 305.Integrated doses for normal operation are the product of the dose rate for full power operations over a 40-year period, co rrected by a factor of 0.8 for plant availability.6.Equipment located within the containment building is qualified for radiation doses comprised of post-accident airborne nuc lide concentrations, airborne nuclide concentration due to primary coolant le akage inside containment during normal operations, and location dependent doses from sources cont aining 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 qua lification of equipment in the fuel building.3.11B.5.2 Chemical Environment A discussion of the LOCA and the source te rms is given in Section 15.6. The chemical composition and resulting pH of the spray water in the containment atmosphere, the liquids in the reactor core, and ESF sumps are identified in Section 6.1. The pH range (4.15 - approximately 11) of the containment sump wa ter following a LOCA is achieve d from the trisodium phosphate dodecahydrate baskets located in the contai nment sump structure. (See Section 6.2.2.)The concentration of chemicals in the containment sprays is equivalent to or more severe than that resulting from the most limiting mode of operation, including a sing le failure in the spray system.
MPS3 UFSAR HISTORICAL, NOT SUBJECT TO FUTURE UPDATING3.11N-11Rev. 303.11N ENVIRONMENTAL DESIGN OF MECH ANICAL AND ELECTRICAL EQUIPMENT This section presents information to demonstrat e that the safety-related electrical equipment of the engineered safety features and the reactor protection systems are capable of performing their designated safety-related functions while exposed to applicable nor mal, abnormal, test, accident, and post-accident environmental conditions. The info rmation presented includ es identification of the safety-related equipment that is within the scope of the Westinghouse Nuclear Steam Supply System (NSSS) and for each item of equipment, definition of the applicable environmental parameters, and description of the qualification process employe d to demonstrate the required environmental capability. The seismic qualificati on of safety-related mechanical and electrical equipment is presented in Sections 3.9 and 3.10, respectively.3.11N.1 EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL CONDITIONSA complete list of safety-related equipment within the Westinghouse NSSS scope of supply that is required to function during and subsequent to an accident is presented in Table 3.11N-1. The plant specific environmental parameters are discussed in Section 3.11B for normal operating and for accident conditions together with the time each item of equipment is required to perform post-accident, when applicable.3.11N.2 QUALIFICATION TESTS AND ANALYSIS3.11N.2.1 Environmental Qualification Criteria The methods of meeting the general requirement s for environmental design and qualification of safety-related equipment as described by GD C 1, 2, 4, and 23 are described in Section 3.1.
Additional specific information concerning th e implementation of GDC 23 is provided in Section 7.2.2.2. The general met hods of implementing the requirements of Appendix B to 10 CFR 50 are described in Chapter 17. Re gulatory Guides 1.40, 1.73, and 1.89 concerning environmental qualification are addressed in Section 1.8.Westinghouse meets the Institute of Electrical and Electronic Engineers (IEEE) Standard 323-1974, IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations, including IEEE Sta ndard 323a-1975, the Nuclear Po wer Engineering Committee (NPEC) Position Statement of July 24, 1975, by either type test, operating experience, analysis, or an appropriate combination of these methods. Westinghous e satisfies this commitment by employing the methodology described in the final staff approved version of WCAP-8587.3.11N.2.2 Performance Requirements for Environmental QualificationIn response to the NRC staff request for addi tional detailed informat ion on the qualification program, Westinghouse submitted Supplement 1 to WCAP-8587. The latest revision of this supplement, Supplement 1, WC AP-8587, contains an equipmen t qualification data package (EQDP) for every item of safety-related electrical equipment supplied by Westinghouse within the NSSS scope of supply. Table 3.11N-1 identifies the equipment supplied by Westinghouse for this application and identifies the applicable EQDP containe d in Supplement 1 of WCAP-8587.
MPS3 UFSAR HISTORICAL, NOT SUBJECT TO FUTURE UPDATING3.11N-12Rev. 30 Each EQDP in Supplement 1 contains a secti on entitled Performance Specification. This specification establishes the sa fety-related functiona l requirements of the equipment to be demonstrated under normal, abnormal, test, accident, and post-accident conditions. The environmental qualification para meters, e.g., temperature, humidity, pressure, radiation, etc, employed by Westinghouse for generic qualifica tion purposes are also identified in the specification, as applicable. The parameters employed have been reviewed to ensure that the plant specific conditions as described in Section 3.11B have been enveloped.3.11N.2.3 Methods and Procedures for Environmental QualificationThe basic methodology employed by Westinghouse for qualification of safety-related electrical equipment is described in WCAP-8587. Each EQDP (Supplement 1, WC AP-8587) contains a description of the qualification program plan for that piece of equipment. Qualification may be demonstrated by either type te st, operating experience, analysis , or a combination of these methods.3.11N.3 QUALIFICATION TEST RESULTS Qualification test results and supporting documentation for safety-related NSSS equipment are summarized in the EQDPs and their associated test reports. 3.11N.4 LOSS OF VENTILATION Loss of ventilation is discussed in Section 3.11B.4.3.11N.5 ESTIMATED CHEMICAL AND RADIATION ENVIRONMENTThe plant-specific estimates of the radiation dose incurred by equipment during normal operation are discussed in Section 3.11B. The estimated doses and chemical conditions following an accident are defined in Section 3.11B. The radiation and chemical environments for which the NSSS scope equipment is qualified are defined in the performance specification of the applicable EQDP contained in Supplement 1, WCAP-8587.
MPS3 UFSAR HISTORICAL, NOT SUBJECT TO FUTURE UPDATING3.11-13Rev. 303.
 
==11.6 REFERENCES==
FOR SECTION 3.113.11-1WCAP-8587, Revision 2, 1979, Butterworth, G. and Miller, R.B., "Methodology for Qualifying Westinghouse WRD Supplied NSSS Sa fety-Related Electrical Equipment."3.11-2WCAP-8587, 1978, "Equipment Qualificat ion Data Packages, Supplement 1."
MPS3 UFSAR3.11-14Rev. 30TABLE 3.11N-1 SAFETY-RELATED EQUIPMENT IN WESTINGHOUSE NSSS SCOPE OF SUPPLY (HISTORICAL, NOT SUBJECT TO FUTURE UPDATING)EquipmentEQD (1)Externally Mounted Limit SwitchesEQDP-H-3 and 6Safety-Related Valve Electric Motor OperatorsEQDP-HE-4Garrett Power-Operated Relief Valves (PORV)EQDP-HE-9Solenoid-Operated Isolation ValveEQDP-HE-10AElectronic Control ModuleEQDP-HE-10B Modulating ValveEQDP-HE-10CLarge Pump Motors (Outside Containment)EQDP-AE-2Canned Pump Motors (Outside Containment)EQDP-AE-3 Charging Pump MotorsEQDP-AE-5APressure TransmittersEQDP-ESE-1A and 1BDifferential Pressure TransmittersEQDP-ESE-3B and 4 Resistance Temperature Detectors (well mounted)EQDP-ESE-6WCAP-11587Excore Neutron DetectorsEQDP-ESE-8A
 
Nuclear Instrumentation System (NIS)EQDP-ESE-10Operator Interface Modules (OIMs)EQDP-ESE-12AProcess Protection SetsEQDP-ESE-13 Indicators, Post-Accident MonitoringEQDP-ESE-14Recorders, Post-Accident MonitoringEQDP-ESE-15Solid-State Protection System and Safeguard Test Cabinet (2 Train)EQDP-ESE-16Reactor Trip SwitchgearEQDP-ESE-20Pressure SensorEQDP-ESE-21Loop Stop Valve CabinetEQDP-ESE-23A Reactor Coolant Pump Speed SensorEQDP-ESE-24ADifferential Pressure Indicating Switch (Group B)EQDP-ESE-40AResistance Temperature Detectors (surface mounted)EQDP-ESE-42AIncore ThermocoupleEQDP-ESE-43A WC Series Connectors and Hardline Extension CableEQDP-ESE-43F MPS3 UFSAR3.11-15Rev. 30NOTE:1.Equipment Qualification Data PackagePotting Adaptor/Cable Splice AssemblyEQDP-ESE-43G Thermocouple Connectors and Thermocouple SpliceWCAP-10919Potting Adaptor/Splice AssemblyWCAP-10920 Incore Thermocouple Reference Junction BoxEQDP-ESE-44ATABLE 3.11N-1 SAFETY-RELATED EQUIPMENT IN WESTINGHOUSE NSSS SCOPE OF SUPPLY (HISTORICAL, NOT SUBJECT TO FUTURE UPDATING)EquipmentEQD (1)
MPS3 UFSAR Rev. 30 APPENDIX 3A - COMPUTER PROGRAMS FOR DYNAMIC AND STATIC ANALYSIS OF SEISMIC CATEGORY I STRUCTURES, EQUIPMENT, AND COMPONENTSPART 1 - STRUCTURESPART 2 - EQUIPMENT AND COMPONENTSPART 3 - PIPING SYSTEMS MPS3 UFSAR3A.1-1Rev. 30 3A.1 STRUCTURES The following are the major computer programs that are used in dynamic an d static analysis of Seismic Category 1 structures. (Note: Minor pr ograms such as post processors and simple programs are not listed.)1.STRUDL II - Structural Analysis Program2.SHELL 1 - Shell Analysis3.STRUDLSW - Structural Analysis Program4.ASAAS - Asymmetric Stress Analysis of Axisymmetric Solids5.TAC2D - Heat Transfer Program6.Time History Program - Dynamic Analysis7.PLAXLY - Finite Element Soil-Structure Analysis for Plain Strain Problems8.ANSYS9.MAT5 - Axisymmetric Mat Analysis10.MEMBRANE11.SBMMI12.GHOSH - WILSON3A.1.1STRUDL II The finite element method (Cheung and Ziankiew icz 1967) provides for the solution of a wide range of solid mechanics problems. Its implem entation within the context of the STRUDL analysis facilities expands these for the treatm ent of plane stress, pl ane strain, plate bending, shallow shell, and three-dimens ional stress analysis problems.STRUDL II (MIT 1968; 1971) has been designed as a modified subsystem of the Integrated Civil Engineering System (ICES), which was designed and formulated at the Massachusetts Institute of Technology, Department of Civil Engineering.
STRUDL II also provides a dynamic analysis capa bility for linear elastic structures undergoing small displacements. Eith er free or forced vibra tional response may be obtained and, in the later case, the forcing functions may be in the fo rm of time histories or response spectra.
MPS3 UFSAR3A.1-2Rev. 30 Seismic Category I structures are analyzed for seismic effect using the dynamic analysis capability of STRUDL II. The analysis yiel ds frequencies of vibrations, mode shapes, displacements, velocities, accelerations, and forces.STRUDL II is a recognized program in the public domain. Version 2- modification 2 (June 1972) of STRUDL is used. The software system is IBM-MVT-RELEASE 20.7. The hardware configuration is IBM-370 - Model 165. The Structural Design Language Engineering Users Manual. Vol 1, Frame Analysis. De partment of Civil Engineering.3A.1.2SHELL 13A.1.2.1General Description This program is based upon the general numerical procedure proposed by B. Budiansky and P.P.
Radkowski (1963) and Greenbaum (1963) to analyze a shell of revolution su bjected to arbitrary loadings.This is a finite difference stress analysis computer code. It can be used to determine the forces, moments, shears, displacements, rotations, and st ress in a thin shell of revolution subject to arbitrary loads expanded in a Four ier series of up to 150 terms. Si ngle layer shells with up to 30 simply connected branches may be analyzed. Poisson's ratio may change at discontinuity points, and Young's modulus and the thermal coefficient of expansion may be different at each point. The allowed types of loading include elastic restra ints, pressures in thr ee orthogonal directions, temperature changes which may have a gradient through the shell thickness, and simplified input for weight of the shell or earthquake forces.
The equilibrium equations for a thin shell are based on Sanders linear theory (Sanders 1959). Sanders' equations are expanded and modified slightly to handle a broader range of problems. All pertinent load, stress, and deformation variables are expanded into a Fourie r series. The individual Fourier components of stress a nd deflection are found separate ly by solution of the finite difference forms of the appropriate differential equations. The algorithm used to solve these equations is a minor modification of the Gaussian elimination method.3A.1.2.2Program Verification - Thin-Wall Cylinder A long thin-walled circular cylinde r is subjected to a constant internal pressure distribution. A solution of this problem ma y be obtained (Roark 1965).
The pertinent parameters of the cylinder are presented in Table 3A.1.2-1.
The following solution can be verified (Roark 1965). (3A.1.2-1)
R PR 2 Et----------=
MPS3 UFSAR3A.1-3Rev. 30 (3A.1.2-2)
The cylinder is idealized by 10 elements, as shown on Figure 3A.1.2-1. Computer results are presented in Table 3A.1.2-2 along with the re sults obtained from Equations 3A.1.2-1 and 3A.1.2-2. As can be seen, the com puter results compare very favorably. Therefore, this problem verifies the accuracy of SHELL 1.
where: s = hoop stress R = radial stress P = pressure R = radius t = thicknessE = Young's modulus3A.1.3STRUDL-SW3A.1.3.1General DescriptionThe STRUDL-SW computer code uses the stiffne ss analysis method to analyze a wide range of structural problems. It handles two and three-dimensiona l trusses and frames, having linear elastic members and statically applied loading.
STRUDL-SW has been documented by bench ma rking procedures against the GTSTRUDL computer code. GTSTRUDL is a rec ognized program in the public domain.3A.1.4ASAAS (Asymmetric Stress Anal ysis of Axisymmetric Solids)3A.1.4.1General Description This is a finite element comput er code (Crose 1971). It can be used to determine stresses and displacements in arbitrary ax isymmetric solids, including pr oblems involving asymmetric mechanical and thermal loads and asymmetric te mperature-dependent mech anical properties.
All dependent variables, including the mechani cal properties, are i nput by Fourier series expansions of the circumferentia l coordinate. The mechanical lo ads can be surface pressures, surface shears, and nodal point forces.The explicit stiffness relations for the axisymme tric solid ring elements of the triangular cross section are based on the classical theorem of potential energy and the assumption that, within any element, the displacement variation in the R-Z plane is linear. Al l dependent variables, including the material properties, are expa nded into the Fourier series. Th e harmonics are coupled and all PR t-------=
MPS3 UFSAR3A.1-4Rev. 30the equilibrium equations are solved simultaneously. The algorithm used to solve the equations is a block modified square root Cholesky method with iterati ve refinement (Crose 1971).3A.1.4.2Sample Program Verification - Harmonic Axisymmetric Plane Strain An infinitely long, solid, circul ar cylinder is subjected to cos  and cos 2 pressure distributions.
A closed-form solution of this pr oblem may be obtained (Love 1944).
The pertinent parameters of the cylinder are presented in Table 3A.1.4-1.
The following solution can be verified (Love 1944): (3A.1.4-1) (3A.1.4-2) (3A.1.4-3) (3A.1.4-4)
(3A.1.4-5)
The cylinder is idealized by 16 elements, as shown on Figure 3A.1.4-1. Computer results are depicted on Figure 3A.1.4-2, along wi th the exact results obtaine d from Equations 3A.1.4-4 and 3A.1.4-5. As can be seen, the comput er results are very close to the exact results. Therefore, this problem verifies the accuracy of ASAAS for mechanical loading problems where material properties are not variable.r P o r a--cos2cos+=P o 3 r a--cos 2r 2 a 2-a 2----------------
2cos+=rP o r a--r 2 a 2-a 2------------
-2sin-sin=U r U r1 U r2+=U r1 P o 14~-()1~+()r 2 2Ea----------------
---------------------------------cos=U r2 P o 1~+E------------r 2~r 3 3a 2-----------
-2cos=UU1 U2+=U1 P o 54~-()1~+()r 2sin 2Ea------------------
----------------------
--------------------
-=U2 P o 1~+E------------1 2~3-------r 3 a 2-----r-2 sin=
MPS3 UFSAR3A.1-5Rev. 303A.1.5TAC2D (A General Purpose Two-Dimensional Heat Transfer Computer Code)3A.1.5.1General DescriptionThis is a finite difference computer code (P etersen 1969) which can be used to determine steady-state and transient temper atures in two-dimensional probl ems. The configuration of the body to be analyzed is described in the rectangular, cylindrical, or circul ar (polar) coordinate system by orthogonal lines of constant coordinate called grid lines. These grid lines specify an array of nodal elements. Nodal points are defi ned as lying midway between the bounding grid lines of these elements. A finite difference equation is formulated for each nodal point in terms of its capacitance, heat generati on, and heat flow paths to nei ghboring nodal points. The equations for all the nodal points are assembled and solved us ing an implicit alternat ing gradient algorithm.3A.1.5.2Program Verification A sample problem is presented to compare the results from TAC2D with an analytical solution. The objective is to show that the TAC2D program yields the correct solution.The problem is to determine the transient temper ature distribution in a right circular cylinder which is initially at temperature T. At time, t = 0, the temperature at the surface is instantaneously changed to T and maintained at that value.Mathematically, the problem is defined by the following equations: (3A.1.5-0) 0  rR(3A.1.5-1)T(r,z,0) = T 1 (3A.1.5-2)
T(R,z,t) = T 2 (3A.1.5-3) (3A.1.5-4) where: 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.
1 r---d dr-----rdT dr---------d 2 T dZ 2---------+1---dT dt-------=T(r L Z---t),+/-, T 2=
MPS3 UFSAR3A.1-6Rev. 30Further, (3A.1.5-5) where: k = the thermal conductivitythe density,c = the specific heat capacity.For the specific problem analyzed, the fo llowing numerical values were used:
R = 12.0 inches L = 48.0 inchesK = 20.0 Btu/hr-ft-
&deg;Fc = 40.0 Btu/cu ft-
&deg;F T 1 = 0.0 &deg;F3A.1.5.2.1Analytical Solution It may be shown (Carlslaw and J aeger 1959) that the solution is:
(3A.1.5-6) (3A.1.5-7) (3A.1.5-8) where the
!m are the roots of:
(3A.1.5-9)The roots m of Equation 3A.1.5-9 and the Bessell functions J and (J i) are tabulated (Jahnke et al., 1945) and need not be computed.kc------=TT-T 2 T 1-------------------1fzt , () g rt , ()-=fzt , ()4"----1-()n2n1+()--------------------
e n0=#=-t$" 2L-------2n1+()2" z 2L-------2n1+()cosgrt , ()2 R----J orm ()m J i Rm ()--------------
--------------
-e-2$m t m1=#=J o R()0=
MPS3 UFSAR3A.1-7Rev. 30 From the definition of the problem there is symm etry about the geometric center of the cylinder and the origin of the coordinate system taken at that point, as is reflected in the boundary conditions, Equations 3A.1.5-3 and 3A.1.5-4.3A.1.5.2.2Numerical Solution With TAC2D A cross section of the problem model for TAC2D is shown on Figure 3A.1.5-1. The model extends only to the axial midplane of the cylinde r where an adiabatic boundary may be specified by virtue of the symmetr y condition described above. The soli d material is represented by one material block. The boundary conditions on the four external boundaries are described by Coolants 1 through 4 (specifically, Coolant Blocks 1 through 4). The material and coolant thermal parameters, as specified by the input functions, are given in Table 3A.1.5-1. All coolants have the standard specific heat of 1.0 Btu per pound-
&deg;F (Btu/lb-
&deg;F). Coolants 1 and 2, which represent the adiabatic external boundaries, have the standard heat transfer coefficient of 10 Btu/hr-sq ft-
&deg;F and the standard flow rate of 10 pounds per hour.3A.1.5.2.3Comparison of TAC2D Solution with the Analytical Solution A comparison of the output from the code with the series solution is shown on Figure 3A.1.5-2.
The temperature-versus-time function is plotted at three representative points within the cylinder. It can be seen that the results from TAC2D are almost identical to the series solution results. The maximum difference between the tw o sets of results is about 2
&deg;F out of a mean magnitude of 100&deg;F.3A.1.6Time History Program3A.1.6.1General DescriptionThe Time History Program computes time history response and am plified response spectra (ARS) at any mass location of a lumped mass system du e to a synthetic earthquake input. The responses are computed by integration of the modal equati ons of the system by exact methods (Nigam and Jennings 1968). The program's main application is the generation of ARS used in the design of Seismic Category I equipment and piping.3A.1.6.2Sample ProblemThe Time History Program's solution to a test pr oblem is substantially identical to the solution obtained using STRADYNE. STRADYNE is a r ecognized program in the public domain. The sample problem used consists of a structure subjected to an ea rthquake time history record. The structure is idealized by five lu mped masses interconnected by fi ve elastic beam elements, as shown on Figure 3A.1.6-1.
Peak acceleration and displacement as well as the horizontal ARS at the top mass point are compared in Tables 3A.1.6-1 and 3A.1.6-2, respectively.
MPS3 UFSAR3A.1-8Rev. 303A.1.7PLAXLY3A.1.7.1General DescriptionThe PLAXLY program provides a numerical soluti on for the dynamic analysis of plane systems under general dynamic loadings. This program works with a two-dime nsional plane-strain finite element idealization of the soil structure interaction problem.The original version of PLAXLY was developed at the University of California in Berkeley (Waas 1972). It was later modified and extended at Stone & Webster (SWE C) to incorporate transient seismic excitations, nonlinear soil behavior, and lumped mass representations of the structures.3A.1.7.2Sample ProblemThe PLAXLY program's solution to a test problem is substantially identical to the solution obtained by using the FLUSH program. FLUSH is a recognized program in the public domain.
The sample problem used consists of a st ructure represented by five lumped masses interconnected by four elastic beam elements. Th is structural model is connected to a finite element representation of the so il, as shown on Figure 3A.1.7-1.
The horizontal amplified res ponse spectra (ARS) at the top mass point are compared on Figure 3A.1.7-2.3A.1.8MAT5 (Circular Mat with Axisymmetric Loading)3A.1.8.1General Description This program is based upon the general numerical procedures proposed by Boris N. Zhemoshkin (1962) to analyze a circular plate on an elasti c foundation. It is used to determine moments, shears, vertical deflections, radi al displacements, tangential a nd radial inplane forces, plus rotations of the circular plate s ubjected to axisymmetric loadings.
The soil subgrade may be modeled as either a Winkler (1867) or a Boussinesqu (1885) type elastic foundation.3A.1.8.2Program VerificationThe results of this program have been reviewed in accordance with standard review procedures in effect at the time of use and, based on the user's and reviewer's knowle dge of plate and shell theory, have been found to be satisfactory.3A.1.9ANSYSThe ANSYS computer program is a large-scal e general purpose computer program for the solution of several classes of engineering anal ysis problems. ANSYS is capable of analyzing MPS3 UFSAR3A.1-9Rev. 30 structures with static and dynamic loadings, elas tic and plastic member pr operties, and small and large deflections.The matrix displacement method of analysis ba sed upon finite element id ealization is employed throughout the program.
ANSYS is a recognized program in the public domain.3A.1.10MEMBRANE (Membrane Stress Analysis)3A.1.10.1General Description This program computes membrane stresses and st rains in containment structures due to dead loads, internal pressure, and temperature gradients across the wall.
It analyzes cylinders, cones, and spherical domes which consist of a fully cracked reinforced concrete section with a steel liner.Stresses are computed by shell membrane theo ry (Billington 1965). The program automatically considers the effect of the uplift force acting on the roof of a cone or cylinder.3A.1.10.2Program Verification As an example, a cylindrical shell was anal yzed by the program and by a hand calculation. Table 3A.1.10-1 presents the pertinent parameters of the cylinder. The comparison of results is given in Table 3A.1.10-2.As can be seen, the computer program's results compare very favorably. This problem verifies the accuracy of MEMBRANE.3A.1.11SBMMI (Single Barr ier Mass Missi le Impact)3A.1.11.1General DescriptionThis program computes the elasto-plastic structural response of a barrier due to the following type of loads: (a) static loads; (b) suddenly applied constant dynamic loads which remain permanently on the structure; (c) suddenly a pplied constant dynamic loads repr esenting missile impact with a finite force and specific momentum; and (d) suddenly applied dynamic load of zero time duration and specific momentum representing missile impact. The barrier is modelle d as a single barrier mass and a non-linear spring, with the above loads applied. The equation of motion is integrated in time assuming constant acceleration in each time step (Billington 1965).3A.1.11.2Program VerificationAs an example, the program SBMMI was run for each load case separately. The computer's result were compared to those obtai ned from a hand calculation. The ha nd calculation was based on the elasto-plastic response charts found in Billington (1965). Table 3A.1.11-1 presents the pertinent model and load parameters. Th e hand and computer program results are compared in Table MPS3 UFSAR3A.1-10Rev. 303A.1.11-2. It can be seen that the computer program's result compare very favorably. This problem verifies the accuracy of the SBMMI computer program.3A.1.12GHOSH-WILSON3A.1.12.1General DescriptionDynamic Stress Analysis of Axisymmetric Stru ctures under Arbitrary Loadings, known as the GHOSH-WILSON computer code, is a finite-eleme nt based computer program developed by S. Ghosh and E. Wilson and modified by Stone & Webster as Code ST-200.
GHOSH-WILSON is capable of performing st atic and dynamic analysis of complex axisymmetric structures subjected to any arbitrary static (mechanical and temperature) and dynamic loading.
The method used to represent the three-dimensional continuum is either as an axisymmetric thin shell, a solid of revolution, or a combination of both. The arbitrary loading in the circumferential direction is represented by a fourier series, and the analysis is carried out for each term and summed up for the total response.Hamilton's variational principle is used to derive the equation of motion. This leads to a diagonal mass matrix and a stiffness matrix and load v ector which is consiste nt with the assumed displacement field. The equations of motion are solved numerically in the time-domain by direct integration using the Wilson method.
The input required by GHOSH-WILS ON is a description of geometry, materials, and boundary conditions. Loadings, damping factors, and time in tervals for integration should be provided for each fourier term. Additional in ertias can be added at join ts during a dynamic analysis.
GHOSH-WILSON provides time hist ory responses of the resultan t forces, moments, shears, displacements, rotations, acceler ations, and stresses at each node for the dynamic analysis.
Maximum responses can also be obtained for each Fourier term.3A.1.12.2Program VerificationStatic Case A cylinder is subjected to a constant internal pressure. The cylinder is modeled using the shell element, rectangular element, a nd the triangular element. The solutions for all three types of elements agree very well.
Dynamic Case A cylinder simply supported at both ends is subjected to a suddenl y applied load at midspan. The solution of the equations of motion is obtained by the direct integration method. The cylinder is modeled using rectangular el ements. The GHOSH-WILSON so lution (displacement under the MPS3 UFSAR3A.1-11Rev. 30applied load) is compared to the solution usi ng the ANSYS computer code. The results compare favorably.3A.1.13References for Appendix 3A.1 3A.1-1 Bathe, Klaus-Jurgen and Wilson, E. Numeri cal Methods in Finite Element Analysis. Prentice-Hall, Inc., Englewood Cliffs, NJ, 1976.
3A.1-2 FBillington, D.P. 1965. Introduction to St ructural Dynamics, First Edition, McGraw-Hill.
3A.1-3 Boussinesqu, J. 1885. Ap plication des Potentials a etude de l'equilibre et du Mouvement des Solides Elastiques. Paris, France.
3A.1-4 Budiansky B. and Radkowski, P.P. 1963. Numerical Analysis of Unsymmetrical
 
Bending of Shells of Revolu tion. AIAA Journal, 1, 1833.
3A.1-5 Carlslaw, H.S. and Jaeger, J.C. 1959. Conduction of Heat in Solids. Oxford at the Clarendon Press, p 227.
3A.1-6 Cheung, Y.K. and Zienkiewicz, O.C. 1967.
The Finite Element Method. McGraw-Hill Book Company, Inc., New York, N.Y.
3A.1-7 Crose, J.G. ASAAS Asymmetric Stress An alysis 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.
3A.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 Engineer ing Corporation Computer Code ST-200, September 1973.
3A.1-9 Greenbaum, G.A. 1963. Comments on Numerica l Analysis of Symmetrical Bending of Shells of Revolution. AIAA Journal 2, 1833.
3A.1-10 Janhke, E. and Emde, F. 1945. Table of Functi ons. Fourth Edition, Dover Publications, New York, N.Y.3A.1-11 Love, A.E.H. 1944. A Treatise on the Mathem atical Theory of Elasticity. Dover Publications, New York, NY.
3A.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.
MPS3 UFSAR3A.1-12Rev. 30 3A.1-13 Massachusetts Institute of Technology (MIT) 1971. ICES STRUDL II, The Structural Design Language Engineering Users Manual. Vol 2, Chapters III and IV, Additional Design and Analysis Facilities. De partment of Civil Engineering.
3A.1-14 Nigam, N.C. and Jennings, P.C. 1968. Digital Calculation of Response Spectrum from Strong-Motion Earthquake Record
: s. National Science Foundation.
3A.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-9262.3A.1-16 Roark, R.J. 1965. Formulas for Stress and Strain. Fourth Edition, McGraw-Hill Book Company, Inc., New York, N.Y. p 298.
3A.1-17 Sanders, J.L., Jr. 1959. An Improved First Approximation Theory for Thin Shells. NASA Technical Report R-24.
3A.1-18 Swanson Analysis Systems, Inc. 1979. Engineering Analysis System User's Manual (ANSYS). Volumes 1 and 2.
3A.1-19 Timashenko, S. and Goodier, J.N. Theory of Elasticity. McGraw-Hill Book Company, Inc., New York, N.Y., 1951, pp 58-60.
3A.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.
3A.1-21 Ulickii, I.I. 1972. Reinforced Concrete Structures (in Russian). Kiev, Russia.
3A.1-22 Waas, G. 1972. Linear Two-Dimensional An alysis of Soil D ynamics Problems in Semi-Infinite Layered Media. Ph.D.Thesis, University of California, Berkeley.
3A.1-23 Winkler, E. 1867. Die Lehre Vonder Elasticat and Festigkeit. Praha, Czechoslovakia.
3A.1-24 Zhemochkin, B. N. and Sinitsin, A. P. 1962. Practical Method in Analysis of Beams and Plates on Elastic Foundations (in Ru ssian). Gosstroiizdat, Russia.
MPS-3 FSARPage 1 of 1Rev. 30TABLE 3.A.1.2-1 THIN-WALL CYLINDER, PERTINENT PARAMETERSDimensions and PropertiesLoad ing and Boundary ConditionsR = 25 inchesAt z = 0 inch; F r = M = z = 0 l = 20 inchesAt z =
l = 20 inches; F r = M = F z = 0t = 0.5 inchP = 75 psi E = 28 x 10 6 psiPoisson's ratio = 0.3 M = Moment on free edge
 
F r = Radial forceF  = Force in z - direction z = Displacment in z - direction MPS-3 FSARPage 1 of 1Rev. 30TABLE 3.A.1.2-2 EXACT AND COMPUTER STRESSES FOR THIN-WALL CYLINDERVariableExactShell 1 R (inch)3.348 x 10
-3 3.348 x 10
-3R (psi)3,7503,750 MPS-3 FSARPage 1 of 1Rev. 30TABLE 3.A.1.4-1 INFINITELY LONG SOLID CYLINDER, PERTINENT PARAMETERSDimensions and PropertiesLoad ing and Boundary Conditions r o  = aP r  = P o (cos % + cos 2 %)l    = aro  = P o sin %E  = 10 x 10 6 psi U z  = 0Z  = 0.25At r = 0, U r = 0 a    = 1 inch P o  = 10,000 psi MPS-3 FSARPage 1 of 1Rev. 30TABLE 3.A.1.5-1 INPUT THERMAL PARAMETER FUNCTIONS FOR TAC2D SAMPLE PROBLEM Material Thermal Parameters Coolant Thermal Parameters SPEC1  (X)= 40.0 H3A (X) = 1.0 x 10 8 RCOH1 (X)= 20.0 FL03A (X) = 1.0 x 10 8 ACON1 (X)= 20.0TIN3A (X) = 1,460 H4A (X) = 1.0 x 10 8 FLO4A (X) = 1.0 x 10 8 TIN4A (X) = 1,460 MPS-3 FSARPage 1 of 1Rev. 30TABLE 3.A.1.6-1 PEAK ACCELERA TION AND DISPLACEMENTTime History Program  Stardyne Program Peak Acceleration0.922 g0.922 g Peak Displacement0.352 in0.352 in MPS-3 FSARPage 1 of 1Rev. 30TABLE 3.A.1.6-2 HORIZONTAL AMPL IFIED RESPONSE SPECTRA(Two Percent Oscillator Damping
)Period (Seconds)Time History Program (g)Stardyne Program (g)0.020.9290.9280.040.9990.9990.061.0401.039 0.081.3011.3010.101.2531.2520.121.6941.697 0.142.4192.4190.163.1583.1530.185.9625.961 0.2011.10411.1010.226.7776.8020.244.6945.025 0.263.4413.4510.282.5762.5760.302.4172.428 0.341.6131.6030.381.7201.7310.421.4931.491 0.461.4871.5070.501.2011.2010.700.6530.660 MPS-3 FSARPage 1 of 1Rev. 30TABLE 3.A.1.10-1 PERTINENT PARAMET ERS OF A CYLINDRICAL SHELLInput DataParameterRadius of outer layer of rebarsr o = 66.5 feet Radius of inner layer of rebarsr i = 63.67 feetRadius of linerr 1 = 63 feetRadius of reference surfacer = 65.25 feetHeight of cylinderh = 122 feetMeridional steel area/
unit length, outer layer A o = 4 in 2Meridional steel area/
unit length, inner layer A i = 4 in 2 Liner area/unit length A l = 4.5 in 2Circumferential stee l area, outer layer A o = 8 in 2 Circumferential steel area, inner layer A i = 8 in 2Internal pressurep = 9.72 ksiWall weight per unit surfaceq = 0.68 ksiTotal load at topw = 9726.5 kTemperature increment, outer rebarsT o = -12&deg;FTemperature increment, inner rebarsi = 27&deg;FTemperature increment, linerl = 230&deg;F MPS-3 FSARPage 1 of 1Rev. 30TABLE 3.A.1.10-2
 
==SUMMARY==
OF RESULTS Input Data Hand CalculationComputer Run% DifferenceLiner strain:Meridional0.0013080.0013080Circumferential0.0014120.0014120 Membrane stresses: (ksi)
Meridional rebars:Outer layer41.5841.590.02Inner layer33.97533.990.02 Liner-6.9857-6.970.2 Hoop rebars:Outer layer42.4742.470 Inner layer36.6535.650Liner-4.586-4.5860Membrane Forces: (k/ft)Meridional271.76271.900.05Circumferential612.32612.360 MPS-3 FSARPage 1 of 1Rev. 30NOTES:*Equivalent barrier weight = 16.66 k Barrier resistance function - From zero displa cement 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.TABLE 3.A.1.11-1 TEST PROBLEM DATALoad TypeLoad1. Static load
-15.5 k* 2. Suddenly applied constant load (remains on structure permanently)62.9 k*3. Missile Impact - finite force 264 k*specific momentum1.4 k*/sec
: 4. Suddenly applied dynamic load w/zero time duration (applied impulse)1.2 k*/sec MPS-3 FSARPage 1 of 1Rev. 30TABLE 3.A.1.11-2 A COMPARISON OF HAND AND COMPUTER PROGRAM RESULTSLoad NumberLoad CaseResults from Hand CalculationResults from Computer Run1Barrier deflection-0.000533 feet-0.0005 feet2Barrier deflection0.0054 feet0.0054 feetTime of maximum deflection0.01802 sec0.018378 sec3Barrier deflection0.0183 feet0.0187 feetTime of maximum deflection0.0180 sec0.018391 sec4Barrier deflection0.0171 feet0.0172 feetTime of maximum deflection0.01481 sec0.014419 sec MPS3 UFSAR3A.2-1Rev. 30 3A.2 EQUIPMENT AND COMPONENTS The following computer programs were used for the analysis of Seismic Category I equipment and components, as well as for pipe rupture design and analysis.1.ASAAS - Asymmetric Stress Analysis of Axisymmetric Solids2.LIMITA 25 - Nonlinear Static Analysis of Plane Frames3.MISSILE - Turbine Missile Probability Program4.SLOSH - Simplified Tank Sloshing Analysis5.LION - Temperature Distribution for Ar bitrary Shapes/Complicated Boundary Conditions6.LIMITA 3 - Nonlinear Dynamic Analysis of Frame Structures7.TAC2D - Two Dimensional Thermal Analysis8.SHELL 1 - Thin Shell of Revol ution Under Arbitrary Loading9.NOZZLE (ST-147) - Vessel Penetration Analysis10.DINASAW - Dynamic Inelastic Nonlinear Analysis by Stone & Webster Engineering Corporation11.LIMITA 2 D - Nonlinear Transient Dynamic Analysis12.STARDYNE - Linear and Nonlinear Elastic Structure Analysis13.ASYMPR - Asymmetric Pressure Force - Time History14.LIDOP - Local Inelastic Deformation of Piping15.DLF - Dynamic Load Factors16.STRUDL - Structural Analysis Progr am. STRUDL program descriptions and verification are presented in Sections 3A.3.8 and 3A.3.9 and are not duplicated here.
MPS3 UFSAR3A.2-2Rev. 30 3A.2.1 ASAAS (Asymmetric Stress Analysis of Axisymmetric Solids) 3A.2.1.1 General Description This is a finite element comput er code (Cross 1971). It can be used to determine stresses and displacements in arbitrary ax isymmetric solids, including pr oblems involving asymmetric mechanical and thermal loads and asymmetric te mperature-dependent mechanical properties. All dependent variables, including the mechanical prope rties, are input by Fourier series expansions of the circumferential coordinate. The mechanical loads can be surface pressures, surface shears, and nodal point forces.The explicit stiffness relations for the axisymme tric solid ring elements of the triangular cross section are based on the classical theorem of potential energy and the assumption that, within any element, the displacement variation in the R-Z plane is linear. Al l dependent variables, including the material properties, are expa nded into the Fourier series. Th e harmonics are coupled and all the equilibrium equations are solved simultaneously. The algorithm used to solve the equations is a block modified square root Cholesky method with iterati ve refinement (Cross 1971).
3A.2.1.2 Program Verification - Harmonic Assymetric - Plane Strain An infinitely long, solid, circular cylinder is s ubjected to cos and cos 2 pressure distributions. A closed-form solution of this pr oblem may be obtained (Love 1944).
The pertinent parameters of the cylinder are presented in Table 3A.2.1-1.
The following solution can be verified (Love 1944): = P o (r cos + cos2) (3A.2.1-1) (3A.2.1-2) (3A.2.1-3) (3A.2.1-4) (3A.2.1-5)
The cylinder is idealized by 16 elements, as shown on Figure 3A.1.4-1. Computer results are depicted on Figure 3A.1.4-2, along wi th the exact results obtaine d from Equations 3A.2.1-4 and P o3r cos 2r 2'2-'2------------------
-2 cos+=rP o r  sin r 2'2-r 2---------------
= 2sin =U r P o14Y-()1Y+()r 2 2E'----------------
--------------
--------------
-sin 1Y+E-------------(r 2Yr 3 3'2------------
-)cos+=UPo54Y-()1Y+()r 2 2E'----------------
--------------
--------------
-sin 1Y+E-------------
(1 2 3----)r 3'2-----Y- 2 sin+=
MPS3 UFSAR3A.2-3Rev. 30 3A.2.1-5. As can be seen, the comput er results are very close to the exact results. Therefore, this problem verifies the accuracy of ASAAS for mechanical loading problems where material properties are not variable.
3A.2.2 Limita 25 - 2D Nonlinear Transient Dynamic Analysis 3A.2.2.1 General DescriptionLimita 25 (ST-224) is a computer code written and fully documented by SWEC, which predicts two-dimensional structures. A plane frame is represented mathematic ally as a discrete system of beam members. Under loadings, the equilibriu m at each joint is ensured by the system equilibrium equa tion (Martin 1966):[K] {q} = {F}(3A.2.2-1) where:[K] = System stiffness matrix
{q} = Global displacement vector
{F} = External force vectorAn element of the stiffness matrix, Kij, situated in row i and column j, is the force in the i th degree of freedom required to produc e a unit displacement in the j th degree of freedom when all other degrees of freedom are restrained from moving (Martin 1966; Przemieniecki 1968).To account for nonlinear effects, such as plasticity and large deflec tions, Equation 3A.2.2-1 is solved by an incremental method at any partic ular load step, Equation (1) may be written:[K]i {q}i = {F}i (3A.2.2-2) where:{q}i = {q}i - {q}i=l, {F}i = {F}i - {F}i-l {q}0 = {q}0 , {F}0 = {F}0The stiffness matrix-[K], calculated based on the deformed structure at load step i (Martin, AFFDL IR66-80, 1966), is assumed constant through the load step. Displacements and member forces are given at load step i by: (3A.2.2-3) q{}q}0=i=
MPS3 UFSAR3A.2-4Rev. 30 where:{Q} = member force vector[k] = member stiffness matrix
{q} = member displacement vector The equilibrium equations, Equa tion (2), are solved by a sta ndard elimination technique.
Since no external loading is applied to a member between joints, the maximum value of the internal force acting on a member occurs at its ends. The transition from the elastic to the fully plastic state is disregarded, and the end sections are assumed to remain linearly elastic until a fully plastic state is reached. The yield surface is defined by scalar function, of the internal member forces, {Q}, having the form (Hodge 1959; Neal 1961; Stoc key et al., 1966): ({Q}i) = 1 is obtained by integrating stress across the member section with the stress fully developed over the section and satisfying the von Mises (or Tresca) yield criterion:2 + 2 T 2 = y 2 where: = Normal stressT = Shear stressy = Yield stress in simple tension 2 = 3 (von Mises) or 4 (Tresca)
Thus the function  depends on the shape of the cross se ction and the force components being considered. For a plane frame, the yielding norma lly occurs due to either a predominant bending moment or a predominant axial force. Therefore, two plastic models are used.Bending Yield Model:
Since a section is either elastic or full y plastic, there are four possible states: 1.Both ends A and B are elastic Q{}K[]q{}0=i=
MPS3 UFSAR3A.2-5Rev. 302.End A is plastic; end B is elastic3.End A is elastic; end B is plastic4.Both ends A and B are plasticA plastic hinge is introduced at any end section which is yielding. The force-displacement relation of the plastic hinge follows an ideal bilinear curve (Clough 1965; Guberson 1967). In situations where force reversal occurs, the elastic stiffn ess of the hinged member is restored, providing elastic unloading (isotropic strain hardening model).Axial Yield Model:
There are only two possible states:1.The entire member is elastic2.The entire member is plastic When the member yields, the member elastic Young's modulus is replace d by a plastic tangent modulus and the force-disp lacement relation follows a bilinear curve. If the member unloads, the elastic modulus is restored.
3A.2.2.2 Program Verification A center loaded beam, built in at one end a nd supported at the other end, is analyzed for comparison to data obtained analytically. The analytical so lution was obtained using limit analysis as explained in Hodge (1959). The displ acement at the point of loading calculated using Limita 25 and calculated analyt ically preshown in Figure 3A
.2.2-1. Additional problems where analyzed to ensure that all program options we re exercised and thus demonstrated the function and adequacy of the program.
3A.2.3 MISSILE 3A.2.3.1 General Description The Missile program calculates the impact probability (P
: 2) of postulated turbine missiles on specified targets. The solid angle method is used to calculate P 2: where:* = Solid angle subtended by the target P 2 1*m--------  *d+=
MPS3 UFSAR3A.2-6Rev. 30
*m = Total solid angle subtended by all possible missile trajectories The integral is evaluated by num erical integration, with consid eration of the missile ejection velocity and the relative positions of the turbine and targ et (Figure 3A.2.3-1).
3A.2.3.2 High Trajectory VerificationWestinghouse has derived a formula to predict th e probability of impact for high trajectory missiles. Some adjustments to the formula are n ecessary to enable direct comparison with the program results. The formula has been derived on the basis that the initial velocity is random and uniformly distributed between V 1 and V 2. The program uses a deterministic initial velocity. The formula may be specialized to this condition by setting V 1 equal to V 2 after applying L'Hopital's Rule. Also, the formula has been derived assuming a missile fragment occurs in the quadrant of the target, whereas the program assumes a miss ile fragment can occur in any of the four quadrants. These differing assumptions can be reconciled by using four fragments for program input.After making the above adjustments, the high trajectory formula becomes:
P = G 2/(2V 4)where:P = Impact probability per square foot of target G = Acceleration of gravity (ft/sec
: 2) = Deflection angle range (radians)
Missile (MA-057) is a computer code written and fully documented by SWEC for inhouse use.
3A.2.4 SLOSH 3A.2.4.1 General Description The purpose of this program is to compute the seismically-induced liquid pressures and the maximum vertical displacement of the liquid surface in a container under horizontal acceleration. The mathematical procedures and formulas used in developing the program were taken from AEC Report TID-7024. The program uses data fo r intensity of ground motion taken from average-acceleration-spectrum curves, as used in the analysis from the report.
The program is used for circular or rectangular, shallow or slender, ground-supported tanks and circular or rectangular, shallow (not slender) tower-supported tanks.
MPS3 UFSAR3A.2-7Rev. 30 3A.2.4.2 Program Verification A comparison of results of computer program SLOSH and those given in the AEC Report (US AEC 1963) shows that the program yields correct results (Table 3A.2.4-1).SLOSH (ME-111) is a computer code written and fully docum ented by SWEC for inhouse use.
3A.2.5 LION (ME-112)LION is a digital computer program which is used to solve three-dimensional transient and steady state temperature distribution problems. The program may also consider subcooled nucleate boiling and coolant heat transfer effects. The surface conditions may be forced convection, free convection, or radiation and heat may be externally or internally generate
: d. Input to the program consists of structural geometry, physical properties, boundary conditions, internal heat generation rates, coolant flow properties, and flow rates.
The program solves the transient heat conduction equations for a three-di mensional field using a first forward difference method. To ensure the temperature calculation stability, LION can determine the suitable time increment, if the specified input time increment is too large.
Since the original program (Bra y 1954) was developed, subsequent versions have evolved to solve larger and more complex problems (B ray 1954, 1959; Personal Communication; Briggs 1963; Lechliter 1963).
LION is a recognized program in the public domain and has been used extensively.
3A.2.6 LIMITA 3 3A.2.6.1 General DescriptionLIMITA 3 (ST-225) is a computer code written and fully documented by SWEC for inhouse use.
Its formulation is identical to that of Limita 2 (Paragraph 3A.2.11) in that the equations are applicable to a three-dimensional problem. For a space frame, yielding normally occurs due to either a predominant bending moment or a predominant torsion. Therefore, two plastic models are provided.1.Bending Yield ModelSince a beam section is either elastic or fu lly plastic, there are four possible states:a.Both ends A and B are elasticb.End A is plastic; end B is elasticc.End A is elastic; end B is plastic MPS3 UFSAR3A.2-8Rev. 30d.Both ends A and B are plastic.A plastic hinge is introduced at any end section which is yielding. The force-displacement relation of the plastic hinge follows an ideal bilin ear curve (Clough et al., 1965; Giberson 1967). In situations where force reversal occurs, the elastic stiffness of the hinged member is restored, providing elastic unloading (isotr opic strain hardening model).2.Torsional Yield ModelThere are only two possible states:a.The entire member is elasticb.The entire member is plastic When the member yields, the member elastic m odulus is replaced by a plastic tangent modulus and the force-displacement relation follows a bilinear curve. If the member unloads, the elastic modulus is restored.
3A.2.6.2 Program Verification 3A.2.6.2.1 Elastic ExampleAs a checkout of LIMITA 3, a space frame is considered. All members are W14x500. A step load of 30 kips is applied vertically at joint 6. This problem was analyzed by LIMITA 3 and ICES STRUDL II elastically. The results of displacements and moment Z at joint 6 were plotted against each other and found to be in excellent agreement.
3A.2.6.2.2 Plastic Example This example is provided to illustrate this ab ility of the program to determine the inelastic transient response of a three-dimensional struct ure. The structures considered consist of cantilevered steel tubes subjected to force transients causing bending and torsion in the structures.
The results obtained from an analysis using the LIMITA 3 code are compar ed with data obtained experimentally.The experiment was a drop test in which the cantilever tubes were loaded by weights at each tube end. The results tabulated we re the peak deflections and their corresponding times and the permanent deflections. These results are compared to those obtained using the LIMITA 3 code in Table 3A.2.6-1.Additional problems were analyzed (elastic and inelastic) to ensure that all program options were exercised and thus demonstrate the functions and ad equacy of the program.
MPS3 UFSAR3A.2-9Rev. 30 3A.2.7 TAC2D (A General Purpose, Two-Dimensional Heat Transfer Computer Code) 3A.2.7.1 General Description Refer to Appendix 3A.1, Section 3A.1.5.1.
3A.2.7.2 Program Verification Refer to Appendix 3A.1, Section 3A.1.5.2.
3A.2.7.2.1 Analytical Solution Refer to Appendix 3A.1, Section 3A.1.5.2.1 3A.2.7.2.2 Numerical Solution with TAC2D Refer to Appendix 3A.1, Section 3A.1.5.2.2.
3A.2.7.2.3 Comparison of TAC2D Solution with the Analytical Solution Refer to Appendix 3A.1, Section 3A.1.5.2.3.
3A.2.7.2.4 References for TAC2D Refer to Appendix 3A.1, Section 3A.1.12.
3A.2.8 SHELL 1 3A.2.8.1 General Description Refer to Appendix 3A.1, Section 3A.1.2.1.
3A.2.8.2 Program Verification - Thin-Wall Cylinder Refer to Appendix 3A.1, Section 3A.1.2.2.
3A.2.8.3 References for SHELL 1 Refer to Appendix 3A.1, Section 3A.1.12.
3A.2.9 Vessel Penetration Analysis 3A.2.9.1 General Description The vessel penetration analysis computer code (ST-147) is written and fully documented by SWEC for inhouse use. The code performs various analyses on tanks and pressure vessels. All of MPS3 UFSAR3A.2-10Rev. 30the analyses are concerned with local stresses at penetrations. Typical problems which can be handled include the following:1.Applied load stresses at vessel-nozzle junction for:a.Rigid attachment to cylinderb.Rigid attachment to spherec.Hollow attachment to sphere2.Pressure discontinuity analysis for thin shell interaction3.Allowable load functions on nozzles for each case Local stresses due to nozzle loads are found by the method prescribed by P.P. Bijlaard (Wichman et al., 1965). The method prescribed by Johns and Orange (Johns and Orange 1961) is used for pressure discontinuity stresses.
3A.2.9.2 Program Verification A sample problem of a thin-walle d cylindrical vessel is subjected to applied loads from a rigid cylindrical attachment. This problem may be solved using Johns and Orange's method (Johns and Orange 1961).
A summary of manual calculations were then compared with the computer summary. Additional problems were considered to ensure that all program options were exercised and thus demonstrate the function and adequacy of the program.
3A.2.10 DINASAW (Dynamic Inelastic Nonlinear Analysis by Stone and Webster) 3A.2.10.1 General DesignDINASAW may be used to predict the nonlinear, dynamic behavior of plane frames (pipes, rings, or beams) including large displacements, plasticity, and impacts.
Arbitrary force-time relations may be applied at any station. DINASAW can also be applied to pipe whip and pipe impact problems.The analysis, as derived (Wu and Witmer 1972; Collins and Witmer 1973), employs the spatial finite-element method in which the tangential a nd normal displacement fields are represented by cubic interpolations. By applying the principle of virtual work in conjunction with D'Alembert's principle, the equations of moti on may be derived in the form:
M[]q**{}+F{}P{}-H[]q{}-=
MPS3 UFSAR3A.2-11Rev. 30 where:{q} and  = the generalized displacements and generalized accelerations, respectively, for the complete assembled discretized struct ure 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 3A.2.10.2 Program VerificationTwo examples are discussed here. The first (Wu and Witmer 1972), involves a ring subjected to radial blast wave over a portion of its circumference. The resulti ng deformation severely distorts the ring, flattening it considerably. Still, the com puter code follows very closely not only to the displacement field, but also to the strain time history.The second case (Collins and Witm er 1973) involves the impact of a rotor segment onto a ring or shroud. Again, the program, in conjunction with the CIVM (Collision Imparted Velocity Method), follows experimental results very closely.3A.2.11 LIMITA 23A.2.11.1 General DescriptionLIMITA 2 (ST-223) predicts nonlinear, dynamic behavior of plane frames, including large displacements, plasticity, and imp act. A plane frame is simulated as a lumped parameter system, consisting of an assembly of discrete lumped masses connected by beam members. Under any loading, the equilibrium at the rth mass point is ensured by the equation of motion:
= f T (3A.2.11-1) where: = a series with one term for each of the i displacements C ri = the damping coefficient, which applies to the ith velocity in the rth equation of motion q**{}m r q**r C ri q i i K ri q i i++
MPS3 UFSAR3A.2-12Rev. 30 K ri = the member stiffness, which is defined as the force necessary to hold the structural member from moving the r th degree of freedom when th e ith degree of freedom is given a unit displacement when all other degrees of freedom are restrained from moving (Martin, 1966; Przemieniecki 1968).
f T = the external load factorTo take account of nonlinear effects, such as plasticity and large deflections, Equation 3A.2.2-1 is solved by an incremental method (Clough and Wi lson 1962). At any particular time, t, the displacement increment is obtained from (3A.2.11-2) where: C t ri = current damping coefficient K t ri = the member stiffness which are calculated based on the current de formed structure (Mar tin, 1966) and assumed constant through the time step, t.The displacement and member forces are thus given by:(3A.2.11-3)The second order differential system equations (Equation 3A.1.11-2) are solved by a linear acceleration implicity method (Hildebrand 1956).
Since no external loading is applied to a me mber between nodes, the maximum value of the internal force acting on a member occurs at its end sections. The tr ansition from the elastic to the fully plastic state is disregarded and the end sections are assumed to remain linearly elastic up to the full plastic yield surface. Th e yield surface is defined by a s calar function of the internal member forces, Q, of the form (Hodge 1959; Neal 1961; Stokey et al., 1966).
- (Q) = l m r q**r t C ri t q i ti K ri tq i t++f r t=( K ri Sq i s) iti=S----------------------
---------------------
-------------------------------
q r tq r s t S=Q r ( K i Sq i S) it S=
MPS3 UFSAR3A.2-13Rev. 30 Here the function is obtained by integrating the stress across the secti on with the stress fully developed over the section and satisfying the von Mises (or Tresca) yield criterion.2 + Y 2 T 2 = y 2 where: = normal stress T = shear stressy = yield stress in simple tension Y 2 = 3 (von Mises) or 4 (Tresca)
Thus, the function depends on the shape of the cross section and the force components being considered.
For a frame structure, the yielding normally occu rs due to either a predominant bending moment or to a predominant tension or compressi on. Thus, two plastic models are provided:1.Bending predominant membersSince a section is either elastic or full y plastic, there are four possible states:a.Both ends A and B are elasticb.End A is yielding and B is elasticc.End A is elastic and B is yieldingd.Both ends A and B are yieldingA plastic hinge is introduced at any end section which is yielding. The force-displacement relation of the plastic hinge follows an ideal bilin ear curve (Clough et al., 1965; Giberson 1967). In situations where force reversal occurs, the stiffness of the hing ed member is restored, providing unloading along the elastic line (iso tropic strain ha rdening model).2.Tension or compression of predominant membersThere are only two possible states:a.The entire member is elasticb.The entire member is plastic MPS3 UFSAR3A.2-14Rev. 30 When the member yields, the member is elastic but Young's Modulus is replaced by a plastic tangent modulus and the for ce-displacement curve follows a bilinear curve. If the member unloads, th e elastic modulus is restored.3A.2.11.2 Program Verification SWEC sponsored an experimental investigation performed by th e Massachusetts Institute of Technology (Wilson 1968). The problem consisted of the cantilevered pipe (Figure 3A.2.11-1) subjected to an impulsive load at its free end. The impulse is imparted by the detonation of a sheet of high explosive, separated from the pipe by a buffer material. A nearly uniform initial velocity is produced in the loaded region and is determined by high speed photography.This problem was analyzed by LIMITA 2. The results were compared with experimental data and output from another computer program, DINASAW.
Additional problems were considered to ensure th at all program options we re exercised and thus demonstrate the function and adequacy of the program.
3A.2.12 STARDYNEThe STARDYNE Structural Analysis System, wr itten by Mechanics Research, Inc., of Los Angeles, California, is a fully warranted and doc umented computer program available at Control Data Corporation. The latest version of this program became available August 1, 1973.The MRI STARDYNE Analysis System consists of a series of compatib le digital computer programs designed to analyze lin ear and nonlinear elastic stru ctural models. The system encompasses the full range of static and dynamic analyses.The static capability includes th e computation of structural deformations and member loads and stresses caused by an arbitrary set of thermal, m odal applied loads, and pr escribed displacements.
Utilizing the normal mode techni que, linear dynamic response analyses can be performed for a wide range of loading conditions , including transient, steady st ate harmonic, random, and shock spectra excitation types. Dynamic response results can be presente d as structural deformations and internal member loads.
The nonlinear dynamic analysis program is integrated in the rest of the STARDYNE system. The equations of motion for the linear portion of the structural model are generated and modified to account for the nonlinear spring
: s. The resulting nonlinear equa tions of motion are directly integrated using either the Newmark or Wilson im plicit integration operators. The user may enter sets of structural loadings which vary with time, and specify time points at which the program is to output th e structural response.
MPS3 UFSAR3A.2-15Rev. 30 3A.2.13 ASYMPR (ME-171) 3A.2.13.1 General Description This computer program is written to calculate the time history of th e resulting forces and moments at assigned nodes in the dynamic mode l of the RPV support syst em. These forces are produced due to the external asymmetric pressure in the reactor cavity, resulting from a LOCA near a hot leg or cold leg nozzle. The pressure forces may be acting at the reactor pressure vessel, the primary shield wall, or the neutron shield tank.
The program performs th e following calculation:
where: P i(t) =A pressure time histor y for pressure area No. i A i = 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 R j = Displacement vector from a force application point on an ar ea to the point of rotation M(t) = Resultant moment vector due to all pressuresTo calculate the force and moment time history at a node in a give n structural model, the surface area on which the pressure acts is divided into several regions, such that only a constant pressure acts on one region at any time.
The projection of a surface area, multiplied by the pressure acting perpendicular to it, gives the pressure force. This force is broken into thr ee global components by defini ng the direction cosines of the pressure force vector.
The centroid coordinates of the projected area are then calculated, and when these are subtracted from the coordinates of the node , the displacement components ar e obtained, which are used in calculating the three moments at the node.
The forces and moments for all projected areas, due to corres ponding pressure forces, are thus calculated and summed for each time point to get the force/moment time history for the node.
Pt ()Pjt ()A i xP i t ()==Mt ()Pj t () x R j              =
MPS3 UFSAR3A.2-16Rev. 30 3A.2.13.2 Program Verification (ME-171)A sample problem was performed by the use of the computer code ME-171. The results were then compared to results obt ained by hand calculations.
3A.2.14 LIDOP (ME-184) 3A.2.14.1 General Description LIDOP (ME-184) generates crush rigidities and deformation energies for pressurized or unpressurized piping in th e following geometries:1.Ring crush against flat rigid surface2.Indent of straight pi pe against rigid cylinder3.1.5D pipe elbow (extrados) against flat rigid surface4.Pipe bend (extrados) ag ainst flat rigid surface5.Indent of straight pipe against rectangular block Both dynamic and material properties are considered in generati on of the crush characteristics.
Unpressurized force-displacement and energy-displacement characteristics of pipe and elbows are generated from empirical equations which are based on experimental data. Pressurization effects, based on fluid displacement during deformat ion, are superimposed on the unpressurized characteristics. The overall dimensions of the contact area, where appl icable, are generated by empirically corrected geometric relationships. Dynamic effects of elbows are empirically determined from an experimental comparison of static and dynamic impact of spheres. Dynamic effects of all other geometries and elbows in ce rtain cases are based on the results of finite element computer simulations of rings impacting flat, rigid surfaces. The effects of material properties are determined from empirical relationships ba sed on computer predictions (EMTR-2-0, 1976; EMTG-33-0, 1977; EMTR-403-B, 1978; Sta ndard Review Plan 3.6.2, NUREG-75/087).
3A.2.14.2 Program VerificationSeveral sample problems were performed by the use of the code ME-184 to ensure that all options were exercised and thus demonstrate th e function and adequacy of the program.
MPS3 UFSAR3A.2-17Rev. 30 3A.2.15 Dynamic Load Factors (DLF ME-185) 3A.2.15.1 General Description The ME-185 program determines the dynamic load factor (DLF) for a si ngle degree-of-freedom harmonic oscillator subject to an arbitrary force history. At time zero, the oscillator is assumed to be in equilibrium and at rest. Its response to the force history, de fined by a series of force-time pairs, is then computed. If the force is not specified at time zero, it is automatically set at zero and ramps up linearly to the first specified force-time coordinate. If the force is specified at time zero, it is assumed to be suddenly applied.
Using the initial conditions at time zero, ME-185 solves the equa tion of motion and searches for maxima during the interval up to the next specified force-time pair. It then determines the boundary conditions (position and velo city) at the end of the interval and uses these as initial conditions for the solution during the next time increment. The process is repeated until the last force-time pair is reached. Assu ming this last force is applied as a continuing load, the steady state response is computed and maxima determine
: d. The greatest maxima is then divided by the greatest applied load to determine the maximum load factor. Since the solution method searches for the greatest absolute amplitude of the syst em response and applied force, the applied force may be positive or negative and may arbitrarily change signs during the specified force history.
The dynamic load factor depends on the natura l frequency of the si ngle degree of freedom oscillator. In order to provide the DLF at the frequency of the structural system be ing analyzed, as well as to show how the DLF changes to an error in the calculated frequency or in the duration of the applied force, DLFs are computed for a range of frequencies. The above calculation method is repeated for each of several discrete freque ncies in the range. The frequency range extends approximately one order of magnitude to either side of the frequency corresponding to the period (duration) of the applied force history.
For a force which varies linearly between two specified force-time pairs, the equation of motion is: where: M = mass of the oscillatork = stiffness of the oscillator F 0 = applied force at the start of the interval a 0 = rate of change in the applied force (F/ T)
The solution of this equation is:
Mx**kx+F o a o t+=
MPS3 UFSAR3A.2-18Rev. 30 x = C 1 sin(wt) + C 2 cos(wt) + C 3 + C 4 t where: w = , the natural frequency C 1 = (V o - a o/k)/w C 2 = X o - F o/k C 3 = F o/k C 4 = a o/k x 0 = initial position V 0 = initial velocity The above relations are simplified and the magnit ude of the spring force is made identical to displacement of K=1. This re lation is used in ME-185.
When transferring from one time interval to the next, the position and velocity must be determined before the new coefficients, C 1...,C 4 can be calculated. The position at the end of the previous time interval can be computed from the above equation for X and the velocity may be determined from:
V = C 1 w cos(wt) - C 2 w sin(wt) + C 4In any interval, the maxima or minima may be computed by substituting the times of zero velocity in the equation for X. These times may be computed from the relation:
0 = C 1 w cos(wt) - C 2 w sin(wt) + C 4 If C 2 = 0 then, wt = cos-1 (-C 4/(C 1 w))If C 4 = tan-1 (C 1/C 2)wt = tan-1 (C 1/C 2)If C 1 = 0 then, wt = sin-1 (C 4/C 2 w))K/M MPS3 UFSAR3A.2-19Rev. 30 otherwise:
3A.2.15.2 Program Verification Several problems were performed by the use of DLF (ME-184) to ensure that all options were used and thus demonstrate the func tion and adequacy of the program.
3A.2.16 References for Appendix 3A.23A.2-1Bray, A.P. TIGER-Temperatures from Internal Generation Rates. KAPL, Schenectady, NY, May 1954.3A.2-2Bray, A.P. and MacCracken, S.J. TIGER-II-Temperatures from Internal Generation Rates. KAPL-2044, May 29, 1959.3A.2-3Briggs, D.L. TIGER-Temp eratures from Internal Gene ration Rates. KAPL-M-EC-29, February 1963.3A.2-4Clough, 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 45.1-45.14.3A.2-5Clough, 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 68-69.3A.2-6Collins, 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 Astronauti cs, Massachusetts Institute of Technology, August 1973.3A.2-7Crose, 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, 1971.3A.2-8EMTG-33-0, Applicability of Pipe Crush Stiffness for Pipe Rupture Analysis, January 10, 1977.3A.2-9EMTR-2-0, Extrapolation and Application of Experimental Static Pipe Crush Data, April 20, 1976.wt 1-  cos C 1-C 4 wC 2 wC 1 2 C 2 2+()w 2 C 4 2-+/-C 1 2 C 2 2+()w 2------------------
------------------
-----------------
------------------
---------------
-=
MPS3 UFSAR3A.2-20Rev. 303A.2-10EMTR-403-B, Pipe Whip Analysis for Concrete Evalua tion, December 18, 1978.3A.2-11Giberson, M.F. The Response of Nonlinear, Multi-Story Structures Subjected to Earthquake Excitation. Earthquake Engineering Research Lab, California Institute of Technology, Pasadena, California, June 1967.3A.2-12Hildebrand, F.B. 1956. Introduction to Numerical Analysis. McGraw- Hill Book Company, Inc., New York, N.Y.3A.2-13Hodge, P.G. 1959. Plastics Analysis of Structures. McGraw-Hill Book Company, New York, N.Y.3A.2-14Johns, 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-103.3A.2-15Larson, L.D. Inelastic Response of Pressurized Tubes Under Dynamic Bending and Torsional Loads. PhD Thesis, Mechani cal Engineering Dept., Carnegie-Mellon University, 1973, University Microfilm Order No. 73-22872.3A.2-16Lechliter, G.L.; Liedel, A.L.; and Schmid, J.R. Mathematics Programs Available on Philco 2000 Computer, Part II, Curve Pl otting. KAPL-M-6416 (EC-40), October 1964.3A.2-17Love, A.E.H. 1944. A Treatise on the Ma thematical Theory of Elasticity. Dover Publications, New York, N.Y.3A.2-18Martin, H.C. 1966. Introduction to Matrix Methods of Structural Analysis. McGraw-Hill Book Company, Inc., New York, N.Y.3A.2-19Martin, 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, Oh io, October 26-28, 1965, AFFBL TR 66-80.3A.2-20Neal, B.G. The Effect of Shear and Norm al Forces on the Fully Plastic Moment of a Beam of Rectangular Cross Section. J ournal of Applied Mechanics, June, 1961, p 269-274.3A.2-21Personal Communication with Mrs. M. Helme concerning TIGER-IV, Thermal Design Memorandum No. 88 (Date).3A.2-22Przemieniecki, J.S. 1968. Theory of Matrix Structural Analys is, McGraw-Hill Book Company, Inc., New York, N.Y.3A.2-23Schmid, 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.
MPS3 UFSAR3A.2-21Rev. 303A.2-24Standard Review Plan 3.6.2, Determination of Break Locations and Dynamic Effects Associated with the Postulated R upture of Piping, USNRC, NUREG-75/087.3A.2-25Stokey, W.F.; Peterson, D.B.; and Wruder, R.A. 1966. Limit Load for Tubes Under Internal Pressure, Bending Moment, Axial Force and Torsi on. Nuclear Engineering and Design, 4, North-Holland Publishing Company, Amsterdam, p 193-261.3A.2-26U.S. Atomic Energy Commission. Nu clear Reactors and Ea rthquakes, Chapter 6, Dynamic Pressure on Fluid Containers. TI D-7024, Division of Reactor Development, Washington, D.C., August 1963.3A.2-27Wichman, 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-107.3A.2-28Wilson, 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 1968.3A.2-29Wu, 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 a nd Astronautics, Massachusetts Institute of Technology, January 1972.
MPS-3 FSARPage 1 of 1Rev. 30TABLE 3.A.2.1-1 INFINITELY LONG SOLID CYLINDER, PERTINENT PARAMETERSDimensions and PropertiesLoading and Boundary Conditions r 0 =ar = P (cos  + cos 2 )l = aq = P 0 sin E = 10 x 10 6 psi U z = 0 U r = 0 Y = 0.25  a = 1 inch At r = 0, P o = 10,000 psi MPS-3 FSARPage 1 of 2Rev. 30TABLE 3.A.2.4-1 COMPARISON OF RESULTS OF SLOSH VS AEC ANALYSIS*
Example 1 Page 188* SLOSH W 0, Eq. Impulsive Force (kips)298.5299.7 P 0, Impulsive Force (kips)105.4105.8M (EBP), Impulsive Moment (kip-ft)594595M (IBP), Impulsive Moment (kip-ft)1,1201,118 W 1, Convective Force (kips)133133 M 1 (EBP), Convective Moment (kip-ft)212214 M 1 (IBP), Convective Moment (kip-ft)252253 M max, Maximum Moment (kip-ft)1,3721,371 P max, Maximum Shear (kips)127.9128.5 Example 2 Page 192* SLOSH W 0, Eq. Impulsive Force (kips)458458 P 0, Impulsive Force (kips)277277M (EBP), Impulsive Moment (kip-ft)3,4603,470M (IBP), Impulsive Moment (kip-ft)4,0704,074 W, Convective Force (kips)139137 M 1 (EBP), Convective Moment (kip-ft)552547 M 1 (IBP), Convective Moment (kip-ft)560552 M max, Maximum Moment (kip-ft)4,6304,626 P max, Maximum Shear (kips)301301 Example 3 Page 197* SLOSH W 0, Eq. Impulsive Force (kips)298.5299.7 MPS-3 FSARPage 2 of 2Rev. 30 Source:*U.S. Atomic Energy Commission. Nuclear Reactors and Earthquakes, Chapter 6, Dynamic Pressure on Fluid Containers. TI D-7024, Division of Reactor Development, Washington, DC, August 1963.
W 1, Convective Force (kips)133133 Mode 1Frequency (cps)0.3330.331FB1, Seismic Force (kips)23.4623.80 FA1, Seismic Force (kips)1.261.28 Mode 2Frequency (cps)1.4861.482 FB2, Seismic Force (kips)-3.90-3.91FA2, Seismic Force (kips)174.39174.54 P max, Maximum Shear (kips)195.21195.72TABLE 3.A.2.4-1 COMPARISON OF RESULTS OF SLOSH VS AEC ANALYSIS*
MPS-3 FSARPage 1 of 1Rev. 30TABLE 3.A.2.6-1 COMPARISON OF EXPERI MENTAL DATA WITH ANALYTICAL DATA USING LIMITA 3 Experimental Value Limita 3 Computer ResultsPercent DifferencePeak Deflection Mode 60.2970.3104.2Peak Deflection Mode 9Not Determined0.870-Time at Peak Deflection in Mode 6~0.004~0.00424.8Permanent Deflection in Mode 60.144~0.1402.8Permanent Deflection in Mode 90.302~0.3102.6 MPS3 UFSAR3A.3-1Rev. 30 3A.3 PIPING SYSTEMSThe 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 No zzle/Run-Pipe Junction Due to Atmospheric Safety Valve Discharge
: 5. STEHAM: Steamhammer Transient Analysis
: 6. WATHAM: Waterhammer Transient Analysis
: 7. PSPECTRA: Peak Spreads and Enve lopes ARS Curves of Dynamic Events 8. STRUDL-SW: Structural Analysis Program
: 9. STRUDL-II (ICES): Structural Analysis Program
: 10. APEN: Anchor Penetration Analysis11. PITRIFE: Finite Elemen t 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 3A.3.1 NUPIPE II1.General Description MPS3 UFSAR3A.3-2Rev. 30 The NUPIPE II piping program performs a linear elastic analysis of three dimensional piping systems subjected to thermal, static, and dynamic loads. It utilizes the finite elem ent 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 anc hor displacements, may be combined to form additional analysis cases. The pi ping deadload analysis considers both distributed weight properti es of the piping and any a dded 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 comp utes values of translational lumped masses.Program input consists basically of program control, piping configuration description, and load specification in formation. The output for each loading condition analyzed consists of support r eactions, internal forces and moments, deflections and rotations, and member st resses. Output from seismic analysis includes system normal mode information.
Several reports may be generated based on report specification. These reports incl ude 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 combinati on of moment resultants, and provision to (automatically) generate results of comb ined loading cases. A program option is available to specify among Class 1 anal ysis 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 se ismic analysis. The results from both programs are presented in Ta bles 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 re sponse. 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
.
MPS3 UFSAR3A.3-3Rev. 30 The Class 1 piping stress conforms with the hand calculations. The model used is shown on Figure 3A.3-3. The results are tabulated in Tables 3A.3-9 and 3A.3-10.3A.3.2 PITRUST1.General Description PITRUST is a computer program which calcu lates the local stress intensity at the junction of two cylindrical vessels. The cal culated 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 consis ts basically of program control options, run-pipe dimensions, internal operating pressure, trunnion outside diameter, and loading specification. If the desi gn criteria for the stresses are exceeded, the program can incrementally increase the pad thickness an d 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 In stitute (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
.3A.3.3 PILUG1.General DescriptionPILUG 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 MPS3 UFSAR3A.3-4Rev. 30 calculated is in the run-pipe and incl udes pressure stresses. The program has specific application where a rectangular attachment is welded to a 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 Pl ant Components and ANSI-B31.1 Power Piping.PILUG input consists basically of program control, run-pipe dimensions, internal operating pressure, rectangular lug di mensions, 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
.3A.3.4 SAVAL1.General DescriptionSAVAL 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 openi ng time, pressure, and code allowable stresses.Computer output consists of resolved fo rces and moments at center line of run-pipe, dynamic loading factor , and calculated stresses.2.Program Verification MPS3 UFSAR3A.3-5Rev. 30Program SAVAL has been verified by compar ing its solution of a test problem to results obtained by hand calculations.
The test problem is illustrated on Figure 3A.3-4. A comparison of results is tabulated in Tables 3A.3-14 and 3A.3-15.3A.3.5 STEHAM1.General Description STEHAM is a computer program which is used to determine the steamhammer transients of piping system
: s. This program uses the me thod of characteristics with finite difference approximations both in sp ace and in time (Jons son et al.; Luk; Moody). It calculates the one-dimensi onal transient flow responses and the flow-induced forcing functions in a pi ping 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 co mponents are mathematically formulated as boundary conditions in th e 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 i nput: (1) the flow network representation of the piping system, (2) the initial fl ow conditions along the piping system, and (3) time-dependent flow char acteristics of piping compone nts. Output consists of time-histories of flow pressures, flow densities, flow velocities, inertia, and momentum functions.2.Program VerificationSTEHAM 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 m easurement, as published in Progelhof and Owczarek (1963) and ASME Paper No. 63-WA-10). A comparis on 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 +  u 2/g)A has also been checked by hand calculations as tabulated in Table 3A.3-16
.3A.3.6 WATHAM1.DescriptionWATHAM 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.
MPS3 UFSAR3A.3-6Rev. 30WATHAM is applicable to a waterhammer problem or, more generally, any unsteady, incompressible fluid flow. Thes e events may be caused by normal or 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 soluti on 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 fo rces 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 VerificationFor the verification of WATHAM, a pr oblem from Reference SWEC 1977 is employed.
Figure 3A.3-10 depicts the flow networ k with nine pipes, its geometrical properties and steady state fl ow conditions. The fl ow-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 Streeter's results (Streeter 1967). The small discrepancy is at tributed 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.
3A.3.7 PSPECTRA1.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 creat ed 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 MPS3 UFSAR3A.3-7Rev. 30application of PSPECTRA is the creation of disk or card data sets of OBE and SSE curves for input to NUPIPE.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 a nd 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 peak s actually span 0.001 second. The sides of parallel peaks are perfectly parallel to the original p eaks. 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 ea ch 0.001 second interval it either sums, finds the maximum value, or takes th e 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 "t rue" 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 sl ope. 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 pr oblem was run three times, producing maximum, sum, and SRSS envelopes (L uong 1977). The results were verified by comparing its solution of th e test problem to results obtained by hand calculations.
3A.3.8 STRUDL-SW1.General DescriptionThe STIFFNESS ANALYSIS Command of STRUDL-SW initiates the execution of a static, linear analysis of a structure considered as a lumped-parameter system.
MPS3 UFSAR3A.3-8Rev. 30 The analysis is performed by the stif fness (displacement) method, treating the displacements as unknowns. From the user
's structural and loading data, the program constructs the structural stiffness matrix and load vectors.
The computational procedure of the analys is 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 differ ent 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 e quations 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 th e 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 MPS3 UFSAR3A.3-9Rev. 30 during analysis. Requests for other types in itiate additional calculations when the output request is made.2.Program Verification STRUDL-SW has been verified by comp aring the results of the STRUDL-SW runs with either hand calcu lations or results obtained by running the same test problems under a public domain versi on of STRUDL (Luong 1980). In this particular case, the public domain ve rsion of STRUDL used was GTICES STRUDL (February 1980 VIM7).
3A.3.9 STRUDL-II (ICES)1.General Description The present capabilities of STRUDL provi de 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 desi gn capabilities which provide both a set of useful techniques for aiding the designer in hi s 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 st ructures 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 pa rticular type of element. The present version of ICES STRUDL provides a variet y of element types for the solution of plane stress/strain, plate bendi ng, and shell analysis problems.
ICES STRUDL-II also contains the fo llowing types of design procedures applicable to the design of framed st ructures and their components: a frame optimization procedure which employs a discrete variable optim ization procedure; a member selection capability with which various design procedures and codes can be used to design members from tables; de signs subjected to ar bitrary constraints; and reinforced concrete, beam, column, a nd slab adequacy checking and section proportioning, based on the ACI code.
MPS3 UFSAR3A.3-10Rev. 302.Program Verification Only partial verification has been pe rformed 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 me thod. A system test was run on both STRUDL-II (V10L06) and GTICES (VIM6) using all commands except for the CSM CALCULATIONS command, and th e 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.
3A.3.10 APEN1.General Description APEN is a computer program that an alyzes anchor penetrations. Anchor penetration consists of three phases, as follows:
PHASE 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 rea d, identified, and stored.
All earthquake type loads are stored as absolute values.
PHASE 2 Design The loads from both sides are combined an d 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 agains t 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.PHASE 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 comb ination. 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 th e analysis procedure are: a.Thickness of concrete wall must be 18 inches or greater.
MPS3 UFSAR3A.3-11Rev. 30b.Recommended only for piping with an operating temperature at 150&deg;F or lower 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' Co mputer Program," Calculation No.
570.470.1.13-NP(B)-010-Z14 (Luong 1977). The resu lts of the test problem from the APEN computer run were compared with that of the hand calculation.3A.3.11 PITRIFE1.General Description PITRIFE is a post processor program that ut ilizes the results of a finite element model of two intersecting cyli nders 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 fi nite element models to the stresses found in an actual pipe-trunnion problem subjecte d to actual loads. It acc omplishes 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 User's choice of one of the four available load combinati on methods. In the second stage, based upon the problem geometry and the type of we ld, the program generates a set of load independent stress coefficients by in terpolating between the non-dimensional stress coefficients generated from the fini te element models. In the third stage, it selects static and dynamic loads in accor dance 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 resu lts 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 progr am 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 MPS3 UFSAR3A.3-12Rev. 30comparison of these results is tabulated in Table 3A.3-19. The program was verified for other ratios of trunnion to pipe radius by demonstrating that the stress coefficients and maximum stress intensi ties 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. 3A.3.12 PITAB1.General DescriptionPITAB provides tabulations of vertical and other restraint data consisting of general support data and final design loads. The final de sign loads for each support are calculated using Millst one procedure. This proce dure specifies the method of combining individual support loading condi tions for various types of supports to arrive at the final design loads.
The program accepts titles, notes, proce dure selection, and support data for each support consisting of gene ral support information, indi vidual loading conditions for the support, and its thermal displace ments. 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 printe
: d. Each table contains a listing of the supports in ascending or der of support number with their general support information and final design loads.2.Program VerificationPITAB 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 we re compared with that of the hand calculation.
3A.3.13 CHPLOT1.General Description CHPLOT is a program which plots any number of data va lues (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 si zes; one with axes of 5 inch es (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 in che by 15 inche page. Pl ots are normally one data value versus time per graph, alt hough up to 14 data values (plots) can be plotted on one graph.
MPS3 UFSAR3A.3-13Rev. 30Each graph's abscissa will be labelled as TI ME (SEC) and the or dinate labels for each graph can be input. Or dinate axes can be select ively suppressed. Scaling is performed automatically to fit the size se lected. 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 labelli ng all graphs at the bottom of each graph.2.Program Verification All the output options such as plot si ze, scale groupings, curve labels, and data values versus time are verified by visu al inspection. It is found that CHPLOT performs all the options defi ned 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.
3A.3.14 LOADCOMB1.General DescriptionThe purpose of LOADCOMB computer program is for th e 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 a nd 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 differ ent coordinate systems, the local run-pipe coordinate system (identical to the NUPIPE, ME-110, local run-pipe coordinate system) and the STRU DL global coordinate system.
The user of LOADCOMB must input th e individual load components into LOADCOMB in the local run-pipe coordinate system. The local x axis for this system coincides with the longitudinal ax is 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 us ing the manual method found in the calculation titled, "Qualification Calculation to Verify Computer Program MPS3 UFSAR3A.3-14Rev. 30ME-163, 'LOADCOMB'," (Sexton 1977). The results of the test problem from the LOADCOMB computer run were compared with that of the hand calculation.
3A.3.15 BEARST1.General DescriptionBEARST 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 - (e arthquake loads); it then stores them along with the fixed - sign loads in an array of load s 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 ther mal 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 ge neral, stress intensities are calculated for each of the two elements taken from th e 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 ve rified by using th e manual method found in the calculation titled, "Verification of BEARST Co mputer Program," General -NP(B)-011-Z14 (Luong 1976). The results of the test problem from the BEARST computer run were compared wi th that of the hand calculation.
3A.3.16 ANCCOMB1.General Description MPS3 UFSAR3A.3-15Rev. 30 ANCCOMB is a computer pr ogram that combines the loads on both sides of the terminal anchor for each load type, in order to form a total anchor load for that load type and to select the smaller of the two ge neral stresses input for each side of the terminal anchor to obtain minimum normal stress (MNS). Terminal anchors are anchors common to two independently anal yzed piping systems, one on each side of the anchor.
The load combination method is base d 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 r oot of the sum of the squares of the two components from both sides of the anchor.
For example:
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, combin ed 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 possi ble 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 Mx OBEITotal Mx 2 OBEI 1 Mx 2 OBEI 2+=
MPS3 UFSAR3A.3-16Rev. 3001," 570.47.01-NP(B)-017 (Purohit 1979). The results of the test problem from the ANCCOMB computer run were compared with that of the hand calculation.
3A.3.17 BENDCORD1.General Description BENDCORD is a Fortran IV Program whic h 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 th e coordinates of the tangent point from the tangent intersection point of the tangent lines.2.Program VerificationThe 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 pipi ng 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 point s 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.
3A.3.18 NUPIPE-SWPC1.General Description MPS3 UFSAR3A.3-17Rev. 30The NUPIPE-SWPC program is used to perform detailed pipe stress analysis. This program is designed to perform analyses in accordance with the ASME B&PV Code, Section III Nuclear Power Plan t Components and the ANSI/ASME B31.1 Power Piping Code.2.Program Verification Using an approved Quality Assurance Progr am, 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.
3A.3.19 PC-PREPS1.General DescriptionPC-PREPS is a PC based computer program which performs a complete structural analysis, performing an AISC code check, weld qualificat ion and baseplate/anchor bolt qualifications.2.Program Verification Using an approved Quality Assurance Progr am, 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.3.20 PILUG-PC1.General Description PILUG-PC is a PC ba sed stress analysis program used to calculate stress intensity at the junction of a rectangular atta chment perpendicular to round pipe.2.Program Verification Using an approved Quality Assurance Progr am, 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.
MPS3 UFSAR3A.3-18Rev. 30 3A.3.21 References for Appendix 3A.3 3A.3-1 American Society of Mechanical Engineers. Pressure Vessel and Piping, 1972 Computer Programs Verification, Problem No. 5.
3A.3-2 Arthur D. Little Corporation. ADLPIPE: Static, Dynamic, Thermal Pipe Stress Analysis.
3A.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.
3A.3-4 EMTG-31-0, Load Conditions, Combinations, and Allowable Stresses, April 11, 1977.
3A.3-5 Fabic, S. 1967. Computer Program WHAM for calculation of Pressure, Velocity, and Force Transients in Liquid Filled Pipi ng Networks, Report No. 67-49-R, Kaiser Engineers, November 1967.
3A.3-6 Jonsson, V.K.; Matthews, L.; and Spalding, D.B. Numerical Solution Procedure for Calculating the Unsteady One-Dimensional Fl ow of Compressible Fluid, ASME Paper No. 73-FE-30.
3A.3-7 Luk, C.H. Effects of the Steam Chest on Steamhammer Analysis for Nuclear Piping Systems, ASME Paper No. 75-PVP-61.
3A.3-8 Luong, W.C. 1976. Verification of 'BEARST' Computer Program. Call. No. - General-NP(B)-011-Z14, December 15, 1976.
3A.3-9 Luong, W.C. 1977a. Verification of 'PSPECTRA' Computer Program. SWEC Computer Library, March 16, 1977a.
3A.3-10 Luong, W.C. 1977b. Verification of 'APEN' Computer Program. Call No. 570.470.1.13-NP(B)-010-Z14, October 28, 1977.3A.3-11 Luong, W.C. Call 570.470.1.13-NP(B)-012.
3A.3-12 Luong, W.C. 1980. Verification of 'STRUDL-SW' Computer Program. SWEC Computer Library, February 1980.
3A.3-13 Moody, F.J. Time - Dependent Pipe Forces Caused by Blowdown and Flow Stoppage, ASME Paper No. 73-FE-23.
3A.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.
MPS3 UFSAR3A.3-19Rev. 30 3A.3-15 Progelhof, R.C. and Owczarek, J.A. The Rapid Discharge of a Gas from a Cylindrical Vessel Through an Orifice, ASME Paper No. 63-WA-10.
3A.3-16 Purohit, S.N. 1979. Verification of
'ANCCOMB' Computer Program. Call. No.
570.47.01-NP(B)-017, November 6, 1979.
3A.3-17 Quan, T.F. Verification of 'PITAB' Computer Program. SWEC Computer Library.
3A.3-18 Sexton, R.W. 1977. Qualification Call. to Verify Computer Program ME-163, 'LOADCOMB'. SWEC Computer Library, November 23, 1977.
3A.3-19 Streeter, V.L. and Wylie, E.G. 1967. Hydraulic Transients, McGraw-Hill Book Company, New York.
3A.3-20 SWEC (Stone & Webster Engineering Co rporation) Computer Library 1977. Part Verification of STRUDL-II (ICES)
Computer Program. January 1977.
3A.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.
MPS3 UFSAR3A.3-20Rev. 30NOTE:*See Figure 3A.3-1 for General NUPIPE II Model.(All node points listed here are not shown.)TABLE 3A.3-1 COMPARISON OF SUPPO RT REACTION DUE TO THERMAL, ANCHOR MOVEMENT, AND EXTERNAL FORCE LOADINGNodeProgram
*Forces (lb)Moments (in-lb)FXFYFZMXMYMZ170NUPIPE-915475414492-5952-8234201241512ADLPIPE-917875404492-5529-8234201241512218NUPIPE16650ADLPIPE16622330NUPIPE34532-33620-31750-486338-1516811573673ADLPIPE34511-33608-31736-486386-1519359573438390NUPIPE8631ADLPIPE8678430NUPIPE170279812553-28147164346248852ADLPIPE174676812541-26917166180250956 MPS3 UFSAR3A.3-21Rev. 30TABLE 3A.3-2 COMPARISON OF DEFL ECTIONS AND ROTATIONS DUE TO THERMAL, ANCHOR MOVEMENT, AND EXTERNAL FORCE LOADING Node Program Deflection (in) Rotation (rad)DXDYDZRXRYRZ197NUPIPE0.348-0.1410.230-0.00260.0025-0.0084ADLPIPE0.348-0.1410.229-0.00260.0025-0.0084212NUPIPE1.1200.052-0.023-0.0092-0.0051-0.0115ADLPIPE1.1200.052-0.023-0.0092-0.0051-0.0115230NUPIPE1.276-0.028-0.548-0.0066-0.00440.0024ADLPIPE1.276-0.027-0.548-0.0066-0.00440.0024260NUPIPE0.512-0.001-0.520-0.0034-0.00050.0035ADLPIPE0.512-0.000-0.520-0.0035-0.00050.0035390NUPIPE0.066-0.0000.249-0.00100.0026-0.0020ADLPIPE0.067-0.0000.248-0.00100.0026-0.0020420NUPIPE-0.029-0.0790.011-0.0002-0.0002-0.0007ADLPIPE-0.029-0.0790.011-0.0002-0.0002-0.0007 MPS3 UFSAR3A.3-22Rev. 30TABLE 3A.3-3 COMPARISON OF STRESS DUE TO THERMAL, ANCHOR MOVEMENT, AND EXTERNAL FORCE LOADINGNodeNUPIPE (psi)ADLPIPE (psi)180189891901319917703177312142395823955 2361442714416265625462513051253912532 34411845118383706295629639534763473 43032823308 MPS3 UFSAR3A.3-23Rev. 30TABLE 3A.3-4 COMPARISON OF INTERNAL FORCES DUE TO DEADWEIGHT ANALYSISNodeProgramForces (lb)Moments (in-lb)FXFYFZMXMYMZ197NUPIPE295233714-35864521851979ADLPIPE290234115-35108523152081212NUPIPE29533061459390-539414010ADLPIPE29933101559735-550014542360NUPIPE3302781-2930930-22748-84971ADLPIPE3262783-3231920-23105-82784390NUPIPE3304933-29-255351701126476ADLPIPE3364707-32-256444916126716420NUPIPE330-492-29-89722707582202ADLPIPE336-497-32-91812772480676 MPS3 UFSAR3A.3-24Rev. 30TABLE 3A.3-5 COMPARISON OF DEFLEC TIONS AND ROTATION DUE TO DEADWEIGHT ANALYSISNodeProgramDeflection (in)Rotation (rad)DXDYDZRXRYRZ197NUPIPE0.007-0.014-0.0040.00010.00010.0002ADLPIPE0.007-0.014-0.0040.00010.00010.0002212NUPIPE-0.005-0.0130.0130.00060.00010.0004ADLPIPE-0.005-0.0130.0130.00060.00010.0004360NUPIPE-0.008-0.0680.0240.0004-0.0000-0.0004ADLPIPE-0.009-0.0690.0240.00040.0000-0.0004390NUPIPE-0.015-0.000-0.0030.0002-0.0002-0.0005ADLPIPE-0.015-0.000-0.0030.0002-0.0002-0.0005420NUPIPE-0.0010.002-0.001-0.0000-0.0001-0.0002ADLPIPE-0.0010.002-0.001-0.0000-0.0001-0.0002 MPS3 UFSAR3A.3-25Rev. 30TABLE 3A.3-6 COMPARISON OF STRESSE S DUE TO DEADWEIGHT ANALYSISNodeNUPIPE (psi)ADLPIPE (psi)18068569419944845821466767923624722449 265530524305515522344635631 37067967739557558043011011091 MPS3 UFSAR3A.3-26Rev. 30TABLE 3A.3-7 COMPARISON OF NATURAL FREQUENCIESMode1st2nd3rd4th5thNUPIPE7.1099.32812.29714.68118.043ADLPIPE7.1189.32912.49212.42717.714 MPS3 UFSAR3A.3-27Rev. 30TABLE 3A.3-8 COMPARISON OF NATURAL FREQUENCIESMode1st2ndNUPIPE2.40713.537Benchmark Pr.2.328813.0808 MPS3 UFSAR3A.3-28Rev. 30TABLE 3A.3-9 COMPARISON OF CLASS 1 PIPE STRESS ANALYSIS Point No. 20 Hand Calculation NUPIPE Min Wall Thickness0.032 in0.032 inPrimary Stress (Eq. 9)3,713 psi3,712 psiPrimary and Secondary Stress (Eq. 10)16,041 psi16,038 psiAlternating Stress (Eq. 11 & 14)13,468 psi13,465 psiUsage Factor0.06540.0631 Point No. 30 Min Wall Thickness0.047 in0.047 in Primary Stress (Eq. 9)8,748 psi8,741 psiPrimary and Secondary Stress (Eq. 10)117,655 psi117,546 psiExpansion Stress (Eq. 12)99,884 psi99,781 psiEq. 1318,252 psi18,246 psiAlternate Stress (Eq. 14)218,258 psi217,811 psiUsage FactorOut of Range MPS3 UFSAR3A.3-29Rev. 30TABLE 3A.3-10 INDIVIDUAL PAIR US AGE FACTOR FOR POINT NO. 30Hand CalculationNUPIPE P air (1, 5)0.1830.1803 P air (1, 8)1.6601.7361 P air (1, 9)0.00010.0001 P air (1, 10)Not in Range P air (5, 8)Not in Range P air (5, 9)0.2210.2646 P air (5, 10)0.7470.8051 P air (8, 9)0.8570.8832 P air (8, 10)5.55185.8608 P air (9, 10)0.00010.0001 MPS3 UFSAR3A.3-30Rev. 30TABLE 3A.3-11 COMPARISON OF PITRUST WITH FRANKLIN INSTITUTE PROGRAM CYLNOZ AND HAND CALCULATIONSource of StressFranklin Institute Corrected ValuesOutput from PITRUSTHand CalculationCircumferential p (Normal) (lb)395.00399.00399.99 p (Bending) (lb)1875.001883.001877.30M (Normal) (in-lb)35.8535.5736.06M (Bending) (in-lb)364.70366.60354.30 M (Normal) (in-lb)79.0579.6679.54M (Bending) (in-lb)90.5280.5779.42Axial p (Normal) (lb)813.00812.00814.80p (Bending) (lb)812.30827.00810.60M (Normal) (in-lb)91.79105.0095.45 M (Bending) (in-lb)158.80160.00158.80M (Normal) (in-lb)37.0637.0037.12M (Bending) (in-lb)117.90105.00103.85 Shear Stress6.636.636.63 by M (psi)Shear Stress106.10106.10106.10 by V (psi)Shear Stress106.10106.10106.10 by V (psi)
MPS3 UFSAR3A.3-31Rev. 30TABLE 3A.3-12 COMPARISON OF PITRUST WITH REFERENCE 8 RESULTSLocation and CausePITRUST ResultsExp. Results (Ref. 8)
Element "A"Longt. Moment(Fig. 16, Ref. 8)Circumf. Stress (psi)20,438.920,000Axial Stress (psi)26,292.625,000 Element "B"Circumf. Moment(Fig. 15, Ref. 8)Circumf. Stress (psi)22,016.224,000 Axial Stress (psi)13,105.813,000 M L M C 0&deg;270&deg;ELEMENT ELEMENT"A""B" MPS3 UFSAR3A.3-32Rev. 30 Stress in Circumferential Direction:TABLE 3A.3-13 COMPARISON OF PILUG COMPUTER PROGRAM OUTPUT WITH HAND CALCULATIONSTest 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; V c = 1,788 lb; V = 2478 lb; M c = 81,834 in-lb; M1 = 103,320 in-lb M t = 76,284 in-lbFigureStress From Hand CalculationComputer OutputRemarks3C0.5485387330Membrane stress due to P1C0.3262,1652,160Bending stress due to P 3A0.294671629Membrane stress due to Mc1A0.38818,97619,904Bending stress due to Mc3B0.4673,0142,961Membrane stress due to M1 1B0.4166,1435,969Bending stress due to M1Stress in Axial Direction:
4C0.4447683690Membrane stress due to P2C0.4632773792Bending stress due to P 4A0.2941,8971,864Membrane stress due to Mc2A0.5506,3575,942Bending stress due to Mc4B0.4672,3652,328Membrane stress due to M1 2B0.5824,989.74,842Bending stress due to M1 MPS3 UFSAR3A.3-33Rev. 30Shear Stress:1,304.81,304.8Shear stress due to Mt -366.99 -366.99Shear stress due to Vc127.15127.16Shear stress due to V1FigureStress From Hand CalculationComputer OutputRemarks MPS3 UFSAR3A.3-34Rev. 30NOTE:*Allowable Stress = 1.2 S h -1.2 (15,490) = 18,588 psiTABLE 3A.3-14
 
==SUMMARY==
OF COMPARISON OF SAVAL COMPUTER OUTPUT WITH HAND CALCULATION AS-DESIGNED CONDITIONVariableHand CalculationSAVALValve and Nozzle Weight (lb)714.09714.09Run Pipe Stiffness (in-lb/Rad)37,950,00038,083,056Nozzle Stiffness (in-lb/Rad)1,120,000,0001,120,619,008 Equivalent Stiffness (in-lb/Rad)36,700,00036,831,376Nat. Rotational Frequency (cps)22.022.19Time Ratio1.111.11 Dynamic Load Factor1.221.22 Circumferential Moment x DLF (in-lb)314,760314,760Net Vertical Force (lb)14,42314,436 Nozzle Stress (psi)6,600 6,597*PITRUST Stress Intensity (psi)62,782 62,789*Stress Intensification Factor5.65.608Equation (9) Stress (psi)36,651.9 36,681*
MPS3 UFSAR3A.3-35Rev. 30TABLE 3A.3-15
 
==SUMMARY==
OF COMPARISON OF SAVAL COMPUTER OUTPUT WITH HAND CALCULATION REINFORCED CONDITION (1 1/4" PAD)VariableHand CalculationSAVALValve and Nozzle Weight (lb)714.09714.09Run Pipe Stiffness (in-lb/Rad)181,500,000181,490,256Nozzle Stiffness (in-lb/Rad)1,120,000,0001,120,619,008 Equivalent Stiffness (in-lb/Rad)156,200,000156,193,808Nat. Rotational Frequency (cps)4545.7Time Ratio2.272.29 Dynamic Load Factor1.131.13Circumferential Moment x DLF (in-lb)291,540291,540Net Vertical Force (lb)14,43614,436 Nozzle Stress (psi)6.1446.144PITRUST Stress Intensity (psi)17.04617.046Stress Intensification Factor2.592.59 Equation (9) Stress (psi)16,18616,200 MPS3 UFSAR3A.3-36Rev. 30TABLE 3A.3-16 NODAL FORCE COMPARISON Diameter D = 0.25 ft Area A = D 2/4 = 0.0490874 ft 2 Nodal Force = (p + v 2 /g) A - P atm A p = pressure lb/ft 2 = density lb/ft 3v = velocity ft/sec g = gravitational constant 32.2 ft/sec 2 P atm = 14.7 x 144 lb/ft 2 At time t = 0.00650 sec Node No.Pressure (psia)Velocity (fps)Density (lb/ft 3)Force (lb) (STEHAM)Force (lb) (Hand Calculation)142.5230.00.23954186.57196.67542.7855.78430.24076198.43198.531044.23131.2190.24647209.00209.11 1547.00378.1720.25737230.62230.732050.214129.890.26979257.84257.972552.095159.430.27697274.93275.06 3052.209161.970.27742276.09276.233552.168162.210.27731275.83275.97 MPS3 UFSAR3A.3-37Rev. 30NOTE: The initial heads of all nodes are calculated by using the Darcy-Weisbach equation.TABLE 3A.3-17 INPUT DATA FOR WATHAM Pipe No.Total Length (ft)Inside Dia (ft)Friction Factor No. of Nodes Nodal Span (ft)Thickness (in)Velocity (fps)12,0003.00.0307333.330.308244.2441323,0002.50.0289375.000.442.92132 32,0002.00.0246400.000.500264.9847341,8001.50.0207300.000.111083.5933651,5001.50.0225375.000.2644.52142 61,6001.50.0256320.000.137962.2918372,2002.50.0408314.290.215343.6587881,5002.00.0306300.000.148113.83245 92,0003.00.0247333.330.308244.24413 MPS3 UFSAR3A.3-38Rev. 30 Nodal Force Calculation is ba sed on the following equation: F = A(pH + /g)V 2)where: F = nodal force (lb) = density (lb/ft 3)H = nodal head (ft) g = 32.2 ft/sec 2 V = nodal velocity (fps)
A = pipe area (ft 2)TABLE 3A.3-18 COMPARISON OF NODAL FORCE CALCULATION AT TIME = 2.34 SECONDSPipe No.Node No.Force, Kips (WATHAM)Force Kips (Hand Calculation)11276.34276.4812300.46300.62 13317.78317.9414329.59329.7615341.39341.56 16355.31355.4917369.52369.71 MPS3 UFSAR3A.3-39Rev. 30TABLE 3A.3-19 COMPARISON OF PITRIFE COMPUTER PROGRAM OUTPUT WITH STRUDL-II OUTPUTTest Problem:Size-on-Size 0.707 Size-on-SizeAverage Pipe Radius (in)3.003.00Average Trunnion Radius (in)3.002.12 Pipe Wall Thickness (in)0.300.30Trunnion Wall Thickness (in)0.300.21SIZE-ON-SIZE MAXIMUM STRESS INTENSITY--PSI ( = 30&deg;)Load PITRIFE Output STRUDL-II Output FX = 10,000 lbs5,7635,768FY = 10,000 lbs7,8447,846FZ = 10,000 lbs6,5075,506MX = 10,000 in-lb1,3291,329MY = 10,000 in-lb1,6881,687MZ = 10,000 in-lb4,0664,068 0.707 SIZE-ON-SIZE MAXIMUM STRESS INTENSITY--PSI ($ = 30&deg;)Load PITRIFE Output STRUDL-II Output FX = 10,000 lbs13,47113,458 FY = 10,000 lbs9,6169,611FZ = 10,000 lbs20,10520,030MX = 10,000 in-lb4,3714,368MY = 10,000 in-lb2,4672,467MZ = 10,000 in-lb6,1786,176 Y X Z MPS3 UFSAR3A.3-40Rev. 30TABLE 3A.3-20 COMPARISON OF PITRIFE COMPUTER PROGRAM OUTPUT WITH HAND CALCULATIONSTest Problem:Average Pipe Radius=1.5 inAverage Trunnion Radius=1.35 inPipe Wall Thickness=0.30 in Trunnion Wall Thickness=0.27 in LOADS FOR EACH LOAD TYPE COMBIN ED (DL, OBEI, THER, OCCU, ETC.)
FX = FY = FZ = 10,000 lbs MX = MY = MZ = 10,000 in-lbsMNS Stress = 200 psi Internal Pressure = 100 psi STRESS COEFFICIENTS--0.9 SI ZE-ON-SIZE--FX LOADING ($ = 30&deg;)Coefficient ByCoefficient From Stress Type Hand Calculation PITRIFE Longitudinal--Inside Fiber-1.2652-1.2652Circumferential--Inside Fiber-0.2764-0.2764Shear--Inside Fiber0.20410.2041
 
Longitudinal--Outside Fiber0.74540.7454 Circumferential--Outside Fiber1.35091.3509Shear--Outside Fiber0.20410.2041MAXIMUM STRESS INTENSITY--0.9 SIZE-ON-SIZE ($ = 30&deg;)               
 
Maximum Stress Intensity--psi Load Condition Hand Calculation PITRIFE P + DL + MNS 128,18128,182P + DL + SRSS (OBEI, OCCU) + MNS 273,22073,220P + DL + OBEA + THER + MNS 388,21688,216P + DL + OCCE + MNS 459,85359,853 P + DL + SRSS (SSEI, OCCF) + MNS 573,22073,220 MPS3 UFSAR3A.3-41Rev. 30TABLE 3A.3-21 BENDCORD PROGRAM--VERIFICATION PROBLEM Offsets Between Nodal Points (feet)ManualBENDCORD X 14.277214.27715 Z 14.277214.27713 X 23.730463.73042 Z 24.761584.76151 X 33.730433.73041 Z 34.761524.76151 X 43.129323.12929 Z 45.176525.17642 X 53.129323.12929 Z 35.176525.17642 X 62.482542.48254 Z 65.515985.51588 X 72.482542.48254 Z 75.515985.51587 X 81.799561.77956 Z 85.775005.77489 X 91.799561.79959 Z 95.775005.77490 MPS3 UFSAR3A.3-42Rev. 30TABLE 3A.3-22 PIPE 1, SEGMENT 1 FORCE VERSUS TIME FOR RUN NUMBER S2807011Time (sec)Seg. Force (lb)0.00.00.0040.00.0084415.83 0.0129626.32 (Maximum Force)0.0166712.860.0202681.74 0.024965.240.0282081.880.0324794.40 0.0366089.200.0405097.820.0443360.32 0.0482580.630.0523234.010.0564383.75 0.0604857.520.0644390.470.0683631.02 0.0723333.100.0763652.300.0804135.11 MPS3 UFSAR Rev. 30APPENDIX 3B - ENVIRONMENTAL DESIGN CONDITIONSFOR ENVIRONMENTAL DESIGN CONDITIONS SEE ENGINEERING SPECIFICATION SP-M3-EE-0333}}

Revision as of 07:24, 15 December 2018

Millstone Power Station Unit 3 Final Safety Analysis Report, Rev. 30, Chapter 3, Design of Structures, Components, Equipment, and Systems
ML17212A070
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