ML21179A124
| ML21179A124 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 06/15/2021 |
| From: | Southern Nuclear Operating Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML21179A130 | List:
|
| References | |
| ND-21-0486 | |
| Download: ML21179A124 (359) | |
Text
Chapter 3 UFSAR Table of Contents Chapter 1 Introduction and General Description of the Plant Chapter 2 Site Characteristics Chapter 3 Design of Structures, Components, Equipment and Systems Chapter 4 Reactor Chapter 5 Reactor Coolant System and Connected Systems Chapter 6 Engineered Safety Features Chapter 7 Instrumentation and Controls Chapter 8 Electric Power Chapter 9 Auxiliary Systems Chapter 10 Steam and Power Conversion Chapter 11 Radioactive Waste Management Chapter 12 Radiation Protection Chapter 13 Conduct of Operation Chapter 14 Initial Test Program Chapter 15 Accident Analyses Chapter 16 Technical Specifications Chapter 17 Quality Assurance Chapter 18 Human Factors Engineering Chapter 19 Probabilistic Risk Assessment UFSAR Formatting Legend Color Description Original Westinghouse AP1000 DCD Revision 19 content (part of plant-specific DCD)
Departures from AP1000 DCD Revision 19 content (part of plant-specific DCD)
Standard FSAR content Site-specific FSAR content Linked cross-references (chapters, appendices, sections, subsections, tables, figures, and references)
VEGP 3&4 - UFSAR TABLE OF CONTENTS Section Title Page CHAPTER 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS ..................................................................................................... 3.1-1 3.1 Conformance with Nuclear Regulatory Commission General Design Criteria ........................................................................................................... 3.1-1 3.1.1 Overall Requirements ................................................................... 3.1-1 3.1.2 Protection by Multiple Fission Product Barriers ............................ 3.1-4 3.1.3 Protection and Reactivity Control Systems ................................. 3.1-10 3.1.4 Fluid Systems ............................................................................. 3.1-14 3.1.5 Reactor Containment .................................................................. 3.1-23 3.1.6 Fuel and Reactivity Control ........................................................ 3.1-26 3.1.7 Combined License Information ................................................... 3.1-29 3.1.8 References ................................................................................. 3.1-29 3.2 Classification of Structures, Components, and Systems .............................. 3.2-1 3.2.1 Seismic Classification ................................................................... 3.2-1 3.2.1.1 Definitions .................................................................. 3.2-1 3.2.1.2 Classifications ............................................................ 3.2-3 3.2.1.3 Classification of Building Structures .......................... 3.2-3 3.2.2 AP1000 Classification System ...................................................... 3.2-3 3.2.2.1 Classification Definitions ............................................ 3.2-4 3.2.2.2 Application of Classification ....................................... 3.2-4 3.2.2.3 Equipment Class A .................................................... 3.2-5 3.2.2.4 Equipment Class B .................................................... 3.2-5 3.2.2.5 Equipment Class C .................................................... 3.2-6 3.2.2.6 Equipment Class D .................................................... 3.2-8 3.2.2.7 Other Equipment Classes ........................................ 3.2-10 3.2.2.8 Instrumentation and Control Line Interface Criteria .. 3.2-11 3.2.2.9 Electrical Classifications .......................................... 3.2-11 3.2.3 Inspection Requirements ............................................................ 3.2-12 3.2.4 Application of AP1000 Safety-Related Equipment and Seismic Classification System ................................................................. 3.2-12 3.2.5 Combined License Information ................................................... 3.2-17 3.2.6 References ................................................................................. 3.2-17 3.3 Wind and Tornado Loadings ......................................................................... 3.3-1 3.3.1 Wind Loadings .............................................................................. 3.3-1 3.3.1.1 Design Wind Velocity ................................................. 3.3-1 3.3.1.2 Determination of Applied Forces ............................... 3.3-1 3.3.2 Tornado Loadings ......................................................................... 3.3-1 3.3.2.1 Applicable Design Parameters .................................. 3.3-2 3.3.2.2 Determination of Forces on Structures ...................... 3.3-2 3.3.2.3 Effect of Failure of Structures or Components Not Designed for Tornado Loads ..................................... 3.3-3 3.3.2.4 Tornado Loads on the Passive Containment Cooling System Air Baffle .......................................... 3.3-3 3.3.3 Combined License Information ..................................................... 3.3-4 3.3.4 References ................................................................................... 3.3-4 3.4 Water Level (Flood) Design .......................................................................... 3.4-1 3.4.1 Flood Protection ........................................................................... 3.4-1 3-i Revision 10
VEGP 3&4 - UFSAR TABLE OF CONTENTS (CONTINUED)
Section Title Page 3.4.1.1 Flood Protection Measures for Seismic Category I Structures, Systems, and Components ..................... 3.4-1 3.4.1.2 Evaluation of Flooding Events ................................... 3.4-4 3.4.1.3 Permanent Dewatering System ............................... 3.4-22 3.4.2 Analytical and Test Procedures .................................................. 3.4-22 3.4.3 Combined License Information ................................................... 3.4-22 3.4.4 References ................................................................................. 3.4-22 3.5 Missile Protection .......................................................................................... 3.5-1 3.5.1 Missile Selection and Description ................................................. 3.5-3 3.5.1.1 Internally Generated Missiles (Outside Containment) 3.5-3 3.5.1.2 Internally Generated Missiles (Inside Containment) .. 3.5-7 3.5.1.3 Turbine Missiles ....................................................... 3.5-10 3.5.1.4 Missiles Generated by Natural Phenomena ............ 3.5-10 3.5.1.5 Missiles Generated by Events Near the Site ........... 3.5-11 3.5.1.6 Aircraft Hazards ....................................................... 3.5-11 3.5.2 Protection from Externally Generated Missiles ........................... 3.5-14 3.5.2.1 Protection from Externally Generated Missile through Wall 11...................................................................... 3.5-14 3.5.2.2 Protection of MSSV Functions from Externally Generated Missile....................................................................... 3.5-15 3.5.2.3 Protection from Externally Generated Missile through Wall I......................................................................... 3.5-15 3.5.2.4 Protection from Externally Generated Missile through Shield Building Air Inlets ........................................... 3.5-16 3.5.3 Barrier Design Procedures ......................................................... 3.5-16 3.5.3.1 Ductility Factors for Steel Structures ....................... 3.5-18 3.5.4 Combined License Information ................................................... 3.5-19 3.5.5 References ................................................................................. 3.5-19 3.6 Protection Against the Dynamic Effects Associated with the Postulated Rupture of Piping .......................................................................................... 3.6-1 3.6.1 Postulated Piping Failures in Fluid Systems Inside and Outside Containment .................................................................... 3.6-2 3.6.1.1 Design Basis .............................................................. 3.6-3 3.6.1.2 Description ................................................................. 3.6-5 3.6.1.3 Safety Evaluation ....................................................... 3.6-8 3.6.2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping ...................... 3.6-10 3.6.2.1 Criteria Used to Define High- and Moderate-Energy Break and Crack Locations and Configurations ...... 3.6-11 3.6.2.2 Analytical Methods to Define Jet Thrust Forcing Functions and Response Models ............................ 3.6-18 3.6.2.3 Dynamic Analysis Methods to Verify Integrity and Operability ................................................................ 3.6-19 3.6.2.4 Protective Assembly Design Criteria ....................... 3.6-23 3.6.2.5 Evaluation of Dynamic Effects of Pipe Ruptures ..... 3.6-24 3.6.2.6 Evaluation of Flooding Effects from Pipe Failures ... 3.6-26 3.6.2.7 Evaluation of Spray Effects from High- and Moderate-Energy Through-Wall Cracks .................. 3.6-26 3-ii Revision 10
VEGP 3&4 - UFSAR TABLE OF CONTENTS (CONTINUED)
Section Title Page 3.6.3 Leak-before-Break Evaluation Procedures ................................. 3.6-27 3.6.3.1 Application of Mechanistic Pipe Break Criteria ........ 3.6-28 3.6.3.2 Design Criteria for Leak-before-Break ..................... 3.6-29 3.6.3.3 Analysis Methods and Criteria ................................. 3.6-31 3.6.3.4 Documentation of Leak-before-Break Evaluations .. 3.6-32 3.6.4 Combined License Information ................................................... 3.6-32 3.6.4.1 Pipe Break Hazard Analysis .................................... 3.6-32 3.6.4.2 Leak-before-Break Evaluation of As-Designed Piping ....................................................................... 3.6-33 3.6.4.3 Leak-before-Break Evaluation of As-Built Piping ..... 3.6-33 3.6.4.4 Primary System Inspection Program for Leak-before-Break Piping ........................................ 3.6-33 3.6.5 References ................................................................................. 3.6-33 3.7 Seismic Design ............................................................................................. 3.7-1 3.7.1 Seismic Input ................................................................................ 3.7-1 3.7.1.1 Design Response Spectra ......................................... 3.7-1 3.7.1.2 Design Time History .................................................. 3.7-3 3.7.1.3 Critical Damping Values ............................................ 3.7-5 3.7.1.4 Supporting Media for Seismic Category I Structures.. 3.7-6 3.7.2 Seismic System Analysis .............................................................. 3.7-7 3.7.2.1 Seismic Analysis Methods ......................................... 3.7-8 3.7.2.2 Natural Frequencies and Response Loads ............. 3.7-10 3.7.2.3 Procedure Used for Modeling .................................. 3.7-10 3.7.2.4 Soil-Structure Interaction ......................................... 3.7-13 3.7.2.5 Development of Floor Response Spectra ................ 3.7-14 3.7.2.6 Three Components of Earthquake Motion ............... 3.7-14 3.7.2.7 Combination of Modal Responses ........................... 3.7-15 3.7.2.8 Interaction of Seismic Category II and Nonseismic Structures with Seismic Category I Structures, Systems, or Components ........................................ 3.7-15 3.7.2.9 Effects of Parameter Variations on Floor Response Spectra .................................................................... 3.7-19 3.7.2.10 Use of Constant Vertical Static Factors ................... 3.7-19 3.7.2.11 Method Used to Account for Torsional Effects ........ 3.7-19 3.7.2.12 Methods for Seismic Analysis of Dams ................... 3.7-20 3.7.2.13 Determination of Seismic Category I Structure Overturning Moments .............................................. 3.7-20 3.7.2.14 Analysis Procedure for Damping ............................. 3.7-20 3.7.3 Seismic Subsystem Analysis ...................................................... 3.7-20 3.7.3.1 Seismic Analysis Methods ....................................... 3.7-20 3.7.3.2 Determination of Number of Earthquake Cycles ..... 3.7-21 3.7.3.3 Procedure Used for Modeling .................................. 3.7-21 3.7.3.4 Basis for Selection of Frequencies .......................... 3.7-22 3.7.3.5 Equivalent Static Load Method of Analysis .............. 3.7-22 3.7.3.6 Three Components of Earthquake Motion ............... 3.7-23 3.7.3.7 Combination of Modal Responses ........................... 3.7-24 3.7.3.8 Analytical Procedure for Piping................................. 3.7-28 3.7.3.9 Combination of Support Responses ........................ 3.7-32 3-iii Revision 10
VEGP 3&4 - UFSAR TABLE OF CONTENTS (CONTINUED)
Section Title Page 3.7.3.10 Vertical Static Factors .............................................. 3.7-35 3.7.3.11 Torsional Effects of Eccentric Masses ..................... 3.7-35 3.7.3.12 Seismic Category I Buried Piping Systems and Tunnels .................................................................... 3.7-35 3.7.3.13 Interaction of Other Systems with Seismic Category I Systems ................................................................. 3.7-35 3.7.3.14 Seismic Analyses for Reactor Internals ................... 3.7-40 3.7.3.15 Analysis Procedure for Damping ............................. 3.7-41 3.7.3.16 Analysis of Seismic Category I Tanks ..................... 3.7-41 3.7.3.17 Time History Analysis of Piping Systems ................ 3.7-41 3.7.4 Seismic Instrumentation ............................................................. 3.7-42 3.7.4.1 Comparison with Regulatory Guide 1.12 ................. 3.7-42 3.7.4.2 Location and Description of Instrumentation ........... 3.7-43 3.7.4.3 Control Room Operator Notification ......................... 3.7-44 3.7.4.4 Comparison of Measured and Predicted Responses................................................................ 3.7-44 3.7.4.5 Tests and Inspections .............................................. 3.7-45 3.7.5 Combined License Information ................................................... 3.7-45 3.7.5.1 Seismic Analysis of Dams ....................................... 3.7-45 3.7.5.2 Post-Earthquake Procedures ................................... 3.7-45 3.7.5.3 Seismic Interaction Review ...................................... 3.7-45 3.7.5.4 Reconciliation of Seismic Analyses of Nuclear Island Structures ...................................................... 3.7-45 3.7.5.5 Free Field Acceleration Sensor ............................... 3.7-45 3.7.6 References ................................................................................. 3.7-45 3.8 Design of Category I Structures .................................................................... 3.8-1 3.8.1 Concrete Containment .................................................................. 3.8-1 3.8.2 Steel Containment ........................................................................ 3.8-1 3.8.2.1 Description of the Containment ................................. 3.8-1 3.8.2.2 Applicable Codes, Standards, and Specifications ..... 3.8-5 3.8.2.3 Loads and Load Combinations .................................. 3.8-5 3.8.2.4 Design and Analysis Procedures ............................... 3.8-7 3.8.2.5 Structural Criteria ..................................................... 3.8-15 3.8.2.6 Materials, Quality Control, and Special Construction Techniques .............................................................. 3.8-15 3.8.2.7 Testing and In-Service Inspection Requirements .... 3.8-16 3.8.3 Concrete and Steel Internal Structures of Steel Containment .... 3.8-16 3.8.3.1 Description of the Containment Internal Structures . 3.8-16 3.8.3.2 Applicable Codes, Standards, and Specifications ... 3.8-22 3.8.3.3 Loads and Load Combinations ................................ 3.8-23 3.8.3.4 Analysis Procedures ................................................ 3.8-25 3.8.3.5 Design Procedures and Acceptance Criteria ........... 3.8-31 3.8.3.6 Materials, Quality Control, and Special Construction Techniques .............................................................. 3.8-39 3.8.3.7 In-Service Testing and Inspection Requirements .... 3.8-40 3.8.3.8 Construction Inspection ........................................... 3.8-40 3.8.4 Other Category I Structures ........................................................ 3.8-40 3.8.4.1 Description of the Structures ................................... 3.8-41 3-iv Revision 10
VEGP 3&4 - UFSAR TABLE OF CONTENTS (CONTINUED)
Section Title Page 3.8.4.2 Applicable Codes, Standards, and Specifications ... 3.8-46 3.8.4.3 Loads and Load Combinations ................................ 3.8-47 3.8.4.4 Design and Analysis Procedures ............................. 3.8-50 3.8.4.5 Structural Criteria ..................................................... 3.8-55 3.8.4.6 Materials, Quality Control, and Special Construction Techniques .............................................................. 3.8-62 3.8.4.7 Testing and In-Service Inspection Requirements .... 3.8-65 3.8.4.8 Construction Inspection ........................................... 3.8-65 3.8.5 Foundations ................................................................................ 3.8-66 3.8.5.1 Description of the Foundations ................................ 3.8-66 3.8.5.2 Applicable Codes, Standards, and Specifications ... 3.8-69 3.8.5.3 Loads and Load Combinations ................................ 3.8-69 3.8.5.4 Design and Analysis Procedures ............................. 3.8-69 3.8.5.5 Structural Criteria ..................................................... 3.8-75 3.8.5.6 Materials, Quality Control, and Special Construction Techniques .............................................................. 3.8-79 3.8.5.7 In-Service Testing and Inspection Requirements .... 3.8-79 3.8.5.8 Construction Inspection ........................................... 3.8-79 3.8.6 Combined License Information ................................................... 3.8-79 3.8.6.1 Containment Vessel Design Adjacent to Large Penetrations ............................................................. 3.8-79 3.8.6.2 Passive Containment Cooling System Water Storage Tank Examination .................................................... 3.8-80 3.8.6.3 As-Built Summary Report ........................................ 3.8-80 3.8.6.4 In-Service Inspection of Containment Vessel .......... 3.8-80 3.8.6.5 Structures Inspection Program ................................ 3.8-80 3.8.6.6 Construction Procedures Program .......................... 3.8-80 3.8.7 References ................................................................................. 3.8-80 3.9 Mechanical Systems and Components ......................................................... 3.9-1 3.9.1 Special Topics for Mechanical Components ................................. 3.9-1 3.9.1.1 Design Transients ...................................................... 3.9-1 3.9.1.2 Computer Programs Used in Analyses .................... 3.9-23 3.9.1.3 Experimental Stress Analysis .................................. 3.9-23 3.9.1.4 Considerations for the Evaluation of the Faulted Conditions ................................................................ 3.9-23 3.9.1.5 Module Interaction, Coupling, and Other Issues ...... 3.9-23 3.9.2 Dynamic Testing and Analysis ................................................... 3.9-24 3.9.2.1 Piping Vibration, Thermal Expansion, and Dynamic Effects ...................................................................... 3.9-24 3.9.2.2 Seismic Qualification Testing of Safety-Related Mechanical Equipment ............................................ 3.9-26 3.9.2.3 Dynamic Response Analysis of Reactor Internals under Operational Flow Transients and Steady-State Conditions ................................................................ 3.9-28 3.9.2.4 Pre-operational Flow-Induced Vibration Testing of Reactor Internals ..................................................... 3.9-31 3.9.2.5 Dynamic System Analysis of the Reactor Internals Under Faulted Conditions ........................................ 3.9-32 3-v Revision 10
VEGP 3&4 - UFSAR TABLE OF CONTENTS (CONTINUED)
Section Title Page 3.9.2.6 Correlation of Reactor Internals Vibration Tests with the Analytical Results .............................................. 3.9-35 3.9.3 ASME Code Classes 1, 2, and 3 Components, Component Supports, and Core Support Structures ..................................... 3.9-36 3.9.3.1 Loading Combinations, Design Transients, and Stress Limits ............................................................ 3.9-36 3.9.3.2 Pump and Valve Operability Assurance .................. 3.9-50 3.9.3.3 Design and Installation Criteria of Class 1, 2, and 3 Pressure Relieving Devices .................................. 3.9-52 3.9.3.4 Component and Piping Supports ............................. 3.9-54 3.9.3.5 Instrumentation Line Supports ................................. 3.9-59 3.9.4 Control Rod Drive System (CRDS) ............................................. 3.9-59 3.9.4.1 Descriptive Information of CRDS ............................. 3.9-59 3.9.4.2 Applicable CRDS Design Specifications .................. 3.9-65 3.9.4.3 Design Loads, Stress Limits, and Allowable Deformations ........................................................... 3.9-67 3.9.4.4 Control Rod Drive Mechanism Performance Assurance Program ................................................. 3.9-68 3.9.5 Reactor Pressure Vessel Internals ............................................. 3.9-68 3.9.5.1 Design Arrangements .............................................. 3.9-68 3.9.5.2 Design Loading Conditions ...................................... 3.9-71 3.9.5.3 Design Bases ........................................................... 3.9-72 3.9.6 Inservice Testing of Pumps and Valves ..................................... 3.9-74 3.9.6.1 Inservice Testing of Pumps ..................................... 3.9-74 3.9.6.2 Inservice Testing of Valves ...................................... 3.9-74 3.9.6.3 Deviations from Code Requirements ....................... 3.9-75 3.9.6.4 Additional Testing .................................................... 3.9-75 3.9.7 Integrated Head Package ........................................................... 3.9-77 3.9.7.1 Design Bases ........................................................... 3.9-78 3.9.7.2 Design Description ................................................... 3.9-79 3.9.7.3 Design Evaluation .................................................... 3.9-80 3.9.7.4 Inspection and Testing Requirements ..................... 3.9-81 3.9.8 Combined License Information ................................................... 3.9-81 3.9.8.1 Reactor Internals Vibration Assessment and Predicted Response ................................................ 3.9-81 3.9.8.2 Design Specifications and Reports .......................... 3.9-81 3.9.8.3 Snubber Operability Testing .................................... 3.9-82 3.9.8.4 Valve Inservice Testing ............................................ 3.9-82 3.9.8.5 Surge Line Thermal Monitoring ............................... 3.9-82 3.9.8.6 Piping Benchmark Program ..................................... 3.9-82 3.9.8.7 As-Designed Piping Analysis ................................... 3.9-82 3.9.9 References ................................................................................. 3.9-82 3.10 Seismic and Dynamic Qualification of Seismic Category I Mechanical and Electrical Equipment ............................................................................ 3.10-1 3.10.1 Seismic and Dynamic Qualification Criteria ................................ 3.10-2 3.10.1.1 Qualification Standards ............................................ 3.10-2 3.10.1.2 Performance Requirements for Seismic Qualification ............................................................. 3.10-2 3-vi Revision 10
VEGP 3&4 - UFSAR TABLE OF CONTENTS (CONTINUED)
Section Title Page 3.10.1.3 Performance Criteria ................................................ 3.10-2 3.10.2 Methods and Procedures for Qualifying Electrical Equipment, Instrumentation, and Mechanical Components .......................... 3.10-3 3.10.2.1 Seismic Qualification of Instrumentation and Electrical Equipment ................................................ 3.10-4 3.10.2.2 Seismic and Operability Qualification of Active Mechanical Equipment ............................................ 3.10-4 3.10.2.3 Valve Operator Qualification .................................... 3.10-6 3.10.2.4 Seismic Qualification of Other Seismic Category I Mechanical Equipment ............................................ 3.10-6 3.10.3 Method and Procedures for Qualifying Supports of Electrical Equipment, Instrumentation, and Mechanical Components ....... 3.10-6 3.10.4 Documentation ........................................................................... 3.10-6 3.10.5 Standard Review Plan Evaluation .............................................. 3.10-7 3.10.6 Combined License Information Item on Experienced-Based Qualification ................................................................................ 3.10-7 3.10.7 References ................................................................................. 3.10-7 3.11 Environmental Qualification of Mechanical and Electrical Equipment ........ 3.11-1 3.11.1 Equipment Identification and Environmental Conditions ............ 3.11-1 3.11.1.1 Equipment Identification .......................................... 3.11-1 3.11.1.2 Definition of Environmental Conditions .................... 3.11-1 3.11.1.3 Equipment Operability Times ................................... 3.11-2 3.11.1.4 Standard Review Plan Evaluation ........................... 3.11-3 3.11.2 Qualification Tests and Analysis ................................................. 3.11-3 3.11.2.1 Environmental Qualification of Electrical Equipment 3.11-3 3.11.2.2 Environmental Qualification of Mechanical Equipment ................................................................ 3.11-3 3.11.3 Loss of Ventilation ...................................................................... 3.11-4 3.11.4 Estimated Radiation and Chemical Environment ....................... 3.11-4 3.11.5 Combined License Information Item for Equipment Qualification File .............................................................................................. 3.11-5 3.11.6 References ................................................................................. 3.11-6 APPENDIX 3A HVAC DUCTS AND DUCT SUPPORTS .....................................................3A-1 3A.1 Codes and Standards ....................................................................................3A-1 3A.2 Loads and Load Combinations ......................................................................3A-1 3A.2.1 Loads .............................................................................................3A-1 3A.2.1.1 Dead Load (D) ..........................................................3A-1 3A.2.1.2 Construction Live Load (L) .......................................3A-1 3A.2.1.3 Pressure (P) .............................................................3A-1 3A.2.1.4 Safe Shutdown Earthquake (Es) ..............................3A-2 3A.2.1.5 Wind Loads (W) ........................................................3A-2 3A.2.1.6 Tornado Loads (Wt) ..................................................3A-2 3A.2.1.7 External Pressure Differential Loads (PA) ................3A-2 3A.2.1.8 Thermal (TO/TA) .......................................................3A-2 3A.2.2 Load Combinations ........................................................................3A-2 3A.3 Analysis and Design .......................................................................................3A-2 3A.3.1 Response Due to Seismic Loads ..................................................3A-3 3A.3.2 Deflection Criteria ..........................................................................3A-3 3-vii Revision 10
VEGP 3&4 - UFSAR TABLE OF CONTENTS (CONTINUED)
Section Title Page 3A.3.3 Relative Movement ........................................................................3A-3 3A.3.4 Allowable Stresses ........................................................................3A-3 3A.3.5 Connections ...................................................................................3A-3 APPENDIX 3B LEAK-BEFORE-BREAK EVALUATION OF THE AP1000 PIPING .............3B-1 3B.1 Leak-before-Break Criteria for AP1000 Piping ...............................................3B-1 3B.2 Potential Failure Mechanisms for AP1000 Piping ..........................................3B-2 3B.2.1 Erosion-Corrosion Induced Wall Thinning .....................................3B-2 3B.2.2 Stress Corrosion Cracking .............................................................3B-3 3B.2.3 Water Hammer ..............................................................................3B-4 3B.2.4 Fatigue ...........................................................................................3B-5 3B.2.5 Thermal Aging ...............................................................................3B-6 3B.2.6 Thermal Stratification .....................................................................3B-6 3B.2.7 Other Mechanisms ........................................................................3B-7 3B.3 Leak-before-Break Bounding Analysis ...........................................................3B-8 3B.3.1 Procedure for Stainless Steel Piping .............................................3B-8 3B.3.1.1 Pipe Geometry, Material and Operating Conditions ................................................................3B-8 3B.3.1.2 Pipe Physical Properties ..........................................3B-9 3B.3.1.3 Low Normal Stress Case (Case 1) ...........................3B-9 3B.3.1.4 High Normal Stress Case (Case 2) ..........................3B-9 3B.3.1.5 Develop the Bounding Analysis Curve ...................3B-10 3B.3.2 Procedure for Non-stainless Steel Piping ....................................3B-10 3B.3.2.1 Pipe Geometry, Material and Operating Conditions ..............................................................3B-10 3B.3.2.2 Calculations Steps ..................................................3B-10 3B.3.2.3 Low Normal Stress Case (Case 1) .........................3B-11 3B.3.2.4 High Normal Stress Case (Case 2) ........................3B-11 3B.3.2.5 Develop the Bounding Analysis Curve ...................3B-11 3B.3.3 Evaluation of Piping System Using Bounding Analysis Curves ...3B-12 3B.3.3.1 Calculation of Stresses ...........................................3B-12 3B.3.3.2 Normal Loads .........................................................3B-13 3B.3.3.3 Maximum Loads .....................................................3B-13 3B.3.3.4 Bounding Analysis Curve Comparison - LBB Criteria ....................................................................3B-14 3B.3.4 Bounding Analysis Results ..........................................................3B-14 3B.4 Differences in Leak-before-Break Analysis for Stainless Steel and Ferritic Steel Pipe .........................................................................................3B-14 3B.5 Differences in Inspection Criteria for Class 1, 2, and 3 Systems .................3B-14 3B.6 Differences in Fabrication Requirements of ASME Class 1, Class 2, and Class 3 Piping .......................................................................................3B-14 3B.7 Sensitivity Study for the Constraint Effect on LBB .......................................3B-15 3B.8 References ...................................................................................................3B-15 APPENDIX 3C REACTOR COOLANT LOOP ANALYSIS METHODS ...............................3C-1 3C.1 Reactor Coolant Loop Model Description ......................................................3C-1 3C.1.1 Steam Generator Model ................................................................3C-1 3C.1.1.1 Steam Generator Mass and Geometrical Model ......3C-1 3C.1.1.2 Steam Generator Supports ......................................3C-1 3C.1.2 Reactor Coolant Pump Model .......................................................3C-1 3-viii Revision 10
VEGP 3&4 - UFSAR TABLE OF CONTENTS (CONTINUED)
Section Title Page 3C.1.2.1 Static Model ..............................................................3C-1 3C.1.2.2 Seismic Model ..........................................................3C-1 3C.1.2.3 Reactor Coolant Pump Supports ..............................3C-2 3C.1.3 Reactor Pressure Vessel Model ....................................................3C-2 3C.1.3.1 Mass and Geometrical Model ...................................3C-2 3C.1.3.2 Reactor Pressure Vessel Supports ..........................3C-2 3C.1.4 Containment Interior Building Structure Model ..............................3C-2 3C.1.5 Reactor Coolant Loop Piping Model ..............................................3C-2 3C.2 Design Requirements .....................................................................................3C-2 3C.3 Static Analyses ...............................................................................................3C-3 3C.3.1 Deadweight Analysis .....................................................................3C-3 3C.3.2 Internal Pressure Analysis .............................................................3C-3 3C.3.3 Thermal Expansion Analysis .........................................................3C-3 3C.4 Seismic Analyses ...........................................................................................3C-3 3C.5 Reactor Coolant Loop Piping Stresses ..........................................................3C-4 3C.6 Description of Computer Programs ................................................................3C-4 APPENDIX 3D METHODOLOGY FOR QUALIFYING AP1000 SAFETY-RELATED ELECTRICAL AND MECHANICAL EQUIPMENT .......................................3D-1 3D.1 Purpose ..........................................................................................................3D-2 3D.2 Scope .............................................................................................................3D-2 3D.3 Introduction ....................................................................................................3D-2 3D.4 Qualification Criteria .......................................................................................3D-2 3D.4.1 Qualification Guides ......................................................................3D-3 3D.4.1.1 IEEE Standards ........................................................3D-3 3D.4.1.2 NRC Regulatory Guides ...........................................3D-4 3D.4.2 Definitions ......................................................................................3D-6 3D.4.3 Mild Versus Harsh Environments ..................................................3D-6 3D.4.4 Test Sequence ..............................................................................3D-7 3D.4.5 Aging .............................................................................................3D-8 3D.4.5.1 Design Life ...............................................................3D-8 3D.4.5.2 Shelf Life ..................................................................3D-8 3D.4.5.3 Qualified Life ............................................................3D-8 3D.4.5.4 Qualified Life Reevaluation ......................................3D-9 3D.4.6 Operability Time ..........................................................................3D-10 3D.4.7 Performance Criterion .................................................................3D-10 3D.4.8 Margin ..........................................................................................3D-11 3D.4.8.1 Normal and Abnormal Extremes ............................3D-11 3D.4.8.2 Aging ......................................................................3D-11 3D.4.8.3 Radiation ................................................................3D-12 3D.4.8.4 Seismic Conditions .................................................3D-12 3D.4.8.5 High-Energy Line Break Conditions .......................3D-13 3D.4.9 Treatment of Failures ..................................................................3D-13 3D.4.10 Traceability ..................................................................................3D-13 3D.4.10.1 Auditable Link Document .......................................3D-13 3D.4.10.2 Similarity .................................................................3D-14 3D.5 Design Specifications ...................................................................................3D-14 3D.5.1 Normal Operating Conditions ......................................................3D-15 3D.5.1.1 Pressure, Temperature, Humidity ..........................3D-15 3-ix Revision 10
VEGP 3&4 - UFSAR TABLE OF CONTENTS (CONTINUED)
Section Title Page 3D.5.1.2 Radiation Dose .......................................................3D-15 3D.5.2 Abnormal Operating Conditions ..................................................3D-16 3D.5.2.1 Abnormal Environments Inside Containment .........3D-16 3D.5.2.2 Abnormal Environments Outside Containment ......3D-16 3D.5.3 Seismic Events ............................................................................3D-17 3D.5.4 Containment Test Environment ...................................................3D-17 3D.5.5 Design Basis Event Conditions ...................................................3D-17 3D.5.5.1 High-Energy Line Break Accidents Inside Containment ...........................................................3D-17 3D.5.5.2 High-Energy Line Break Accidents Outside Containment ...........................................................3D-20 3D.6 Qualification Methods ...................................................................................3D-20 3D.6.1 Type Test .....................................................................................3D-21 3D.6.2 Analysis .......................................................................................3D-21 3D.6.2.1 Similarity .................................................................3D-21 3D.6.2.2 Substitution .............................................................3D-22 3D.6.2.3 Analysis of Safety-Related Mechanical Equipment 3D-22 3D.6.3 Operating Experience ..................................................................3D-24 3D.6.4 On-Going Qualification ................................................................3D-25 3D.6.5 Combinations of Methods ............................................................3D-25 3D.6.5.1 Use of Existing Qualification Reports .....................3D-25 3D.7 Documentation .............................................................................................3D-26 3D.7.1 Equipment Qualification Data Package .......................................3D-26 3D.7.2 Specifications ..............................................................................3D-27 3D.7.2.1 Equipment Identification .........................................3D-27 3D.7.2.2 Installation Requirements .......................................3D-27 3D.7.2.3 Electrical Requirements .........................................3D-27 3D.7.2.4 Auxiliary Devices ....................................................3D-27 3D.7.2.5 Preventive Maintenance .........................................3D-27 3D.7.2.6 Safety Functions......................................................3D-28 3D.7.2.7 Performance Requirements ....................................3D-28 3D.7.2.8 Environmental Conditions .......................................3D-28 3D.7.3 Qualification Program ..................................................................3D-28 3D.7.4 Qualification by Test ....................................................................3D-28 3D.7.4.1 Specimen Description ............................................3D-29 3D.7.4.2 Number Tested .......................................................3D-29 3D.7.4.3 Mounting .................................................................3D-29 3D.7.4.4 Connections ...........................................................3D-29 3D.7.4.5 Test Sequence .......................................................3D-29 3D.7.4.6 Simulated Service Conditions ................................3D-30 3D.7.4.7 Measured Variables ...............................................3D-30 3D.7.4.8 Type Test Summary ...............................................3D-30 3D.7.5 Qualification by Analysis ..............................................................3D-31 3D.7.6 Qualification by Experience .........................................................3D-31 3D.7.7 Qualification Program Conclusions .............................................3D-31 3D.7.8 Combined License Information ....................................................3D-31 3D.8 References ...................................................................................................3D-31 Appendix 3D-Attachment A Sample Equipment Qualification Data Package (EQDP)............ 3D-58 3-x Revision 10
VEGP 3&4 - UFSAR TABLE OF CONTENTS (CONTINUED)
Section Title Page Appendix 3D-Attachment B Aging Evaluation Program .......................................................... 3D-75 B.1 Introduction ........................................................................................................ 3D-75 B.2 Objectives .......................................................................................................... 3D-75 B.3 Basic Approach .................................................................................................. 3D-75 B.4 Harsh Environment .......................................................................................3D-75 B.4.1 Scope................................................................................................ 3D-76 B.4.2 Aging Mechanisms............................................................................ 3D-76 B.4.3 Time .................................................................................................. 3D-76 B.4.4 Operational Stresses ........................................................................ 3D-76 B.4.5 External Stresses.............................................................................. 3D-77 B.4.6 Synergism ......................................................................................... 3D-78 B.4.7 Design Basis Event Testing .............................................................. 3D-78 B.4.8 Aging Sequence................................................................................ 3D-78 B.4.9 Performance Criterion....................................................................... 3D-78 B.4.10 Failure Treatment.............................................................................. 3D-78 B.5 Mild Environment ..........................................................................................3D-79 B.5.1 Scope................................................................................................ 3D-79 B.5.2 Performance Criteria......................................................................... 3D-79 B.5.3 Failure Treatment.............................................................................. 3D-80 Appendix 3D-Attachment C Effects of Gamma Radiation Doses Below 104 Rads on the Mechanical Properties of Materials............................................................ 3D-83 C.1 Introduction ........................................................................................................ 3D-83 C.2 Scope ................................................................................................................. 3D-83 C.3 Discussion.......................................................................................................... 3D-84 C.4 Conclusions........................................................................................................ 3D-84 C.5 References......................................................................................................... 3D-85 Appendix 3D-Attachment D Accelerated Thermal Aging Parameters..................................... 3D-89 D.1 Introduction ........................................................................................................ 3D-89 D.2 Arrhenius Model ................................................................................................. 3D-89 D.3 Activation Energy ............................................................................................... 3D-90 D.4 Thermal Aging (Normal/Abnormal Operating Conditions).................................. 3D-91 D.4.1 Normal Operation Temperature (T0)................................................. 3D-91 D.4.1.1 External Ambient Temperature (Ta) .............................. 3D-91 D.4.1.2 Temperature Rise in Enclosure (Tr) .............................. 3D-91 D.4.1.3 Self-Heating Effects (Tj) ................................................ 3D-92 D.4.2 Accelerated Aging Temperature (Ti) ................................................. 3D-92 D.4.3 Examples of Arrhenius Calculations ................................................. 3D-92 D.5 Post-Accident Thermal Aging............................................................................. 3D-93 D.5.1 Post-Accident Operating Temperatures............................................ 3D-93 D.5.2 Accelerated Thermal Aging Parameters for Post-Accident Conditions ......................................................................................... 3D-93 D.6 References......................................................................................................... 3D-93 Appendix 3D-Attachment E Seismic Qualification Techniques ............................................... 3D-99 E.1 Purpose.............................................................................................................. 3D-99 E.2 Definitions .......................................................................................................... 3D-99 E.2.1 1/2 Safe Shutdown Earthquake ........................................................ 3D-99 E.2.2 Seismic Category I Equipment.......................................................... 3D-99 E.2.3 Seismic Category II Equipment......................................................... 3D-99 3-xi Revision 10
VEGP 3&4 - UFSAR TABLE OF CONTENTS (CONTINUED)
Section Title Page E.2.4 Non-seismic Equipment .................................................................... 3D-99 E.2.5 Active Equipment .............................................................................. 3D-99 E.2.6 Passive Equipment ........................................................................... 3D-99 E.3 Qualification Methods......................................................................................... 3D-99 E.3.1 Use of Qualification by Testing ....................................................... 3D-100 E.3.2 Use of Qualification by Analysis...................................................... 3D-100 E.4 Requirements................................................................................................... 3D-100 E.4.1 Damping.......................................................................................... 3D-100 E.4.1.1 Testing......................................................................... 3D-100 E.4.1.2 Analysis ....................................................................... 3D-101 E.4.2 Interface Requirements................................................................... 3D-101 E.4.3 Mounting Simulation ....................................................................... 3D-101 E.4.4 1/2 Safe Shutdown Earthquake ...................................................... 3D-101 E.4.5 Safe Shutdown Earthquake ............................................................ 3D-101 E.4.6 Other Dynamic Loads ..................................................................... 3D-101 E.5 Qualification by Test......................................................................................... 3D-102 E.5.1 Qualification of Hard-Mounted Equipment ...................................... 3D-102 E.5.2 Qualification of Line-Mounted Equipment ....................................... 3D-103 E.5.2.1 Seismic Qualification Test Sequence.......................... 3D-103 E.5.2.2 Line Vibration Aging .................................................... 3D-104 E.5.2.3 Single Frequency Testing............................................ 3D-104 E.5.2.4 Seismic Aging.............................................................. 3D-104 E.5.2.5 Static Deflection Testing of Active Valves ................... 3D-104 E.5.3 Operational Conditions ................................................................... 3D-105 E.5.4 Resonant Search Testing ............................................................... 3D-105 E.6 Qualification by Analysis .................................................................................. 3D-105 E.6.1 Modeling ......................................................................................... 3D-105 E.6.2 Qualification by Static Analysis ....................................................... 3D-106 E.6.3 Qualification by Dynamic Analysis .................................................. 3D-106 E.6.3.1 Response Spectrum Analysis ..................................... 3D-106 E.6.3.2 Static Coefficient Method ............................................ 3D-107 E.6.3.3 Time History Analysis.................................................. 3D-107 E.7 Qualification by Test Experience...................................................................... 3D-107 E.8 Performance Criteria ........................................................................................ 3D-107 E.8.1 Equipment Qualification by Test ..................................................... 3D-107 E.8.2 Equipment Qualification by Analysis............................................... 3D-107 E.8.2.1 Structural Integrity ....................................................... 3D-107 E.8.2.2 Operability ................................................................... 3D-108 APPENDIX 3E HIGH-ENERGY PIPING IN THE NUCLEAR ISLAND .................................3E-1 APPENDIX 3F CABLE TRAYS AND CABLE TRAY SUPPORTS ....................................... 3F-1 3F.1 Codes and Standards ................................................................................... 3F-1 3F.2 Loads and Load Combinations ..................................................................... 3F-1 3F.2.1 Loads ............................................................................................ 3F-1 3F.2.1.1 Dead Load (D) ......................................................... 3F-1 3F.2.1.2 Construction Live Load (L) ...................................... 3F-1 3F.2.1.3 Safe Shutdown Earthquake (Es) ............................. 3F-1 3F.2.1.4 Thermal Load .......................................................... 3F-2 3F.2.2 Load Combinations ....................................................................... 3F-2 3-xii Revision 10
VEGP 3&4 - UFSAR TABLE OF CONTENTS (CONTINUED)
Section Title Page 3F.3 Analysis and Design ...................................................................................... 3F-2 3F.3.1 Damping ....................................................................................... 3F-2 3F.3.2 Seismic Analysis ........................................................................... 3F-2 3F.3.3 Allowable Stresses ....................................................................... 3F-3 3F.3.4 Connections .................................................................................. 3F-3 APPENDIX 3G NUCLEAR ISLAND SEISMIC ANALYSES ................................................. 3G-1 3G.1 Introduction ................................................................................................... 3G-1 3G.2 Nuclear Island Finite Element Models ........................................................... 3G-1 3G.2.1 Individual Building and Equipment Models ................................... 3G-2 3G.2.1.1 Coupled Auxiliary and Shield Building ..................... 3G-2 3G.2.1.2 Containment Internal Structures .............................. 3G-2 3G.2.1.3 Containment Vessel ................................................ 3G-2 3G.2.1.4 Polar Crane ............................................................. 3G-3 3G.2.1.5 Major Equipment and Structures Using Stick Models ..................................................................... 3G-4 3G.2.2 Nuclear Island Dynamic Models ................................................... 3G-4 3G.2.2.1 NI10 Model .............................................................. 3G-4 3G.2.2.2 NI20 Model .............................................................. 3G-5 3G.2.2.3 Nuclear Island Stick Model ...................................... 3G-5 3G.2.2.4 NI05 Model .............................................................. 3G-5 3G.2.2.5 Seismic Stability Model ........................................... 3G-5 3G.2.3 Static Models ................................................................................ 3G-6 3G.2.3.1 Quadrant Model of Shield Building Roof ................. 3G-6 3G.2.3.2 Containment Vessel 3D Finite Element Model ........ 3G-6 3G.2.3.3 Containment Vessel Axisymmetric Model ............... 3G-6 3G.3 2D SASSI Analyses ...................................................................................... 3G-6 3G.4 Nuclear Island Dynamic Analyses ................................................................. 3G-8 3G.4.1 ANSYS Fixed Base Analysis ........................................................ 3G-8 3G.4.2 3D SASSI Analyses ...................................................................... 3G-8 3G.4.3 Seismic Analysis .......................................................................... 3G-9 3G.4.3.1 Response Spectrum Analysis ................................ 3G-9 3G.4.3.2 Absolute Accelerations ........................................... 3G-9 3G.4.3.3 Seismic Response Spectra .................................... 3G-9 3G.4.3.4 Bearing Pressure Demand .................................... 3G-10 3G.5 References .................................................................................................. 3G-10 APPENDIX 3GG 3-D SSI ANALYSIS OF AP1000 AT VOGTLE SITE USING NI15 MODEL....................................................................................................3GG-1 APPENDIX 3H AUXILIARY AND SHIELD BUILDING CRITICAL SECTIONS .....................3H-1 3H.1 Introduction ....................................................................................................3H-1 3H.2 Description of Auxiliary and Shield Buildings .................................................3H-1 3H.2.1 Description of Auxiliary Building ....................................................3H-1 3H.2.2 Description of Shield Building ........................................................3H-2 3H.3 Design Criteria ...............................................................................................3H-4 3H.3.1 Governing Codes and Standards ..................................................3H-4 3H.3.2 Seismic Input .................................................................................3H-4 3H.3.3 Loads .............................................................................................3H-5 3H.3.4 Load Combinations and Acceptance Criteria ................................3H-8 3H.4 Seismic Analyses ..........................................................................................3H-8 3-xiii Revision 10
VEGP 3&4 - UFSAR TABLE OF CONTENTS (CONTINUED)
Section Title Page 3H.4.1 Live Load for Seismic Design ........................................................3H-8 3H.5 Structural Design of Critical Sections .............................................................3H-9 3H.5.1 Shear Walls .................................................................................3H-10 3H.5.1.1 Exterior Wall at Column Line 1 ...............................3H-11 3H.5.1.2 Wall at Column Line 7.3 .........................................3H-11 3H.5.1.3 Wall at Column Line L ............................................3H-12 3H.5.1.4 Wall at Column Line 11 ..........................................3H-12 3H.5.2 Composite Structures (Floors and Roof) .....................................3H-13 3H.5.2.1 Roof at Elevation 180-0, Area 6 (Critical Section is between Col. Lines N & K-2 and 3 & 4) ..............3H-14 3H.5.2.2 Floor at Elevation 135-3, Area 1 (Between Column Lines M and P) ..........................................3H-15 3H.5.3 Reinforced Concrete Slabs ..........................................................3H-15 3H.5.3.1 Operations Work Area (Tagging Room) Ceiling .....3H-16 3H.5.4 Concrete Finned Floors ...............................................................3H-16 3H.5.5 Structural Modules .......................................................................3H-18 3H.5.5.1 West Wall of Spent Fuel Pool .................................3H-18 3H.5.6 Shield Building Roof and Connections ........................................3H-19 3H.5.6.1 Air Inlets and Tension Ring ....................................3H-20 3H.5.6.2 Compression Ring and Interior Wall of Passive Containment Cooling Water Storage Tank .............3H-20 3H.5.6.3 Knuckle Region and Exterior Wall of Passive Containment Cooling System Tank ........................3H-20 3H.5.7 Shield Building Cylinder (SC) ......................................................3H-21 3H.5.7.1 Shield Building Cylindrical Wall ..............................3H-21 3H.5.7.2 Reinforced Concrete (RC)/Steel Concrete Composite (SC) Horizontal and Vertical Connections ...........................................................3H-22 3H.5.8 References ..................................................................................3H-22 APPENDIX 3I EVALUATION FOR HIGH FREQUENCY SEISMIC INPUT ......................... 3I-1 3I.1 Introduction ..................................................................................................... 3I-1 3I.2 High Frequency Seismic Input ........................................................................ 3I-1 3I.3 NI Models Used To Develop High Frequency Response ................................ 3I-1 3I.4 Evaluation Methodology .................................................................................. 3I-2 3I.5 General Selection Screening Criteria .............................................................. 3I-2 3I.6 Evaluation ....................................................................................................... 3I-3 3I.6.1 Building Structures ......................................................................... 3I-3 3I.6.2 Primary Coolant Loop ..................................................................... 3I-4 3I.6.3 Piping Systems ............................................................................... 3I-5 3I.6.4 Electrical and Electro-Mechanical Equipment ................................ 3I-5 3I.7 References ...................................................................................................... 3I-8 3-xiv Revision 10
VEGP 3&4 - UFSAR LIST OF TABLES Table Number Title Page 3.2-1 Comparison of Safety Classification Requirements ................................ 3.2-19 3.2-2 Seismic Classification of Building Structures .......................................... 3.2-20 3.2-3 AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment .................................................................. 3.2-21 3.2-201 Not Used ................................................................................................. 3.2-79 3.5-1 External Missile Protection Provided for Auxiliary Building Wall 11 Openings ................................................................................................. 3.5-20 3.5-2 External Missile Protection Provided for Annex Building Wall I Openings ................................................................................................. 3.5-22 3.5-3 External Missile Protection Provided for Shield Building Air Inlet Openings ................................................................................................. 3.5-23 3.5-201 Augusta APO Terminal Area Forecast Summary Report - Itinerant Operations ............................................................................................... 3.5-24 3.5-202 Deleted in Revision 2 .............................................................................. 3.5-25 3.6-1 High-Energy and Moderate-Energy Fluid Systems Considered for Protection of Essential Systems .............................................................. 3.6-35 3.6-2 Subcompartments and Postulated Pipe Ruptures .................................. 3.6-36 3.6-3 NI Rooms With Pipe Whip Restraints, Dual Acting Pipe Supports, Jet Shields and Corresponding Hazard Sources and Essential Targets ...... 3.6-43 3.7.1-1 Safe Shutdown Earthquake Damping Values ......................................... 3.7-49 3.7.1-2 Embedment Depth and Related Dimensions of Category I Structures ... 3.7-50 3.7.1-3 AP1000 Design Response Spectra Amplification Factors for Control Points ...................................................................................................... 3.7-51 3.7.1-4 Strain Compatible Soil Properties ........................................................... 3.7-52 3.7.2-1-3.7.2-16 Not Used ................................................................................................. 3.7-57 3.7.3-1 Seismic Category I Equipment Outside Containment by Room Number.................................................................................................... 3.7-58 3.7.3-2 Equipment Classified as Sensitive Targets for Seismically Analyzed Piping, HVAC Ducting, Cable Tray Systems ........................................... 3.7-61 3.8.2-1 Load Combinations and Service Limits for Containment Vessel............. 3.8-84 3.8.2-2 Containment Vessel Pressure Capabilities ............................................. 3.8-85 3.8.2-3 Analysis and Test Results of Fabricated Heads...................................... 3.8-86 3.8.2-4 Summary of Containment Vessel Models and Analysis Methods ........... 3.8-87 3.8.2-5 Maximum Absolute Nodal Acceleration (ZPA) Steel Containment Vessel...................................................................................................... 3.8-88 3.8.3-1 Shear and Flexural Stiffnesses of Structural Module Walls .................... 3.8-89 3.8.3-2 Summary of Containment Internal Structures Models and Analysis Methods................................................................................................... 3.8-90 3.8.3-3 Definition of Critical Locations and Thicknesses for Containment Internal Structures ................................................................................... 3.8-91 3.8.3-4 Design Summary of Southwest Wall of Refueling Cavity Design Loads, Load Combinations, and Comparison to Acceptance Criteria Mid-Span at Mid-Height ........................................................................................... 3.8-92 3.8.3-5 Design Summary of South Wall of West Steam Generator Compartment Design Loads, Load Combinations, and Comparison to Acceptance Criteria Mid-Span at Mid-Height .............................................................. 3.8-95 3-xv Revision 10
VEGP 3&4 - UFSAR LIST OF TABLES (CONTINUED)
Table Number Title Page 3.8.3-6 Design Summary of North-East Wall of IRWST Design Loads, Load Combinations, and Comparison to Acceptance Criteria Mid-Span at Mid-Height ............................................................................................... 3.8-98 3.8.3-7 Design Summary of Steel Wall of IRWST ............................................. 3.8-101 3.8.4-1 Load Combinations and Load Factors for Seismic Category I Steel Structures .............................................................................................. 3.8-102 3.8.4-2 Load Combinations and Load Factors for Seismic Category I Concrete Structures .............................................................................. 3.8-103 3.8.4-3 Acceptance Tests for Concrete Aggregates.......................................... 3.8-104 3.8.4-4 Criteria for Water Used in Production of Concrete ................................ 3.8-105 3.8.4-5 Not Used ............................................................................................... 3.8-106 3.8.4-6 Materials Used in Structural and Miscellaneous Steel .......................... 3.8-107 3.8.5-1 Minimum Required Factor of Safety for Overturning and Sliding of Structures .............................................................................................. 3.8-109 3.8.5-2 Factors of Safety for Flotation, Overturning and Sliding of Nuclear Island Structures ................................................................................... 3.8-110 3.8.5-3 Definition of Critical Locations, Thicknesses and Reinforcement for Nuclear Island Basemat (in2/ft) ............................................................. 3.8-111 3.8-201 Waterproof Membrane Inspections, Tests, Analyses, and Acceptance Criteria ................................................................................................... 3.8-112 3.9-1 Reactor Coolant System Design Transients ........................................... 3.9-86 3.9-2 Pump Starting/Stopping Conditions ........................................................ 3.9-88 3.9-3 Loadings for ASME Class 1, 2, 3, CS and Supports ............................... 3.9-89 3.9-4 Not Used ................................................................................................. 3.9-91 3.9-5 Minimum Design Loading Combinations For ASME Class 1, 2, 3 And CS Systems And Components ........................................................ 3.9-92 3.9-6 Additional Load Combinations and Stress Limits for ASME Class 1 Piping ...................................................................................................... 3.9-93 3.9-7 Additional Load Combinations and Stress Limits for ASME Class 2, 3 Piping ................................................................................................... 3.9-94 3.9-8 Minimum Design Loading Combinations for Supports for ASME Class 1, 2, 3 Piping and Components ..................................................... 3.9-95 3.9-9 Stress Criteria for ASME Code Section III Class 1 Components and Supports and Class CS Core Supports ............................................ 3.9-96 3.9-10 Stress Criteria for ASME Code Section III Class 2 and 3 Components and Supports...................................................................... 3.9-97 3.9-11 Piping Functional Capability - ASME Class 1, 2, and 3 .......................... 3.9-99 3.9-12 List of ASME Class 1, 2, and 3 Active Valves ....................................... 3.9-100 3.9-13 Control Rod Drive Mechanism Production Tests .................................. 3.9-107 3.9-14 Maximum Deflections Allowed for Reactor Internal Support Structures .............................................................................................. 3.9-108 3.9-15 Computer Programs for Seismic Category 1 Components ................... 3.9-109 3.9-16 Valve Inservice Test Requirements....................................................... 3.9-110 3.9-17 System Level Operability Test Requirements ....................................... 3.9-111 3.9-18 AP1000 Pressure Isolation Valves ........................................................ 3.9-112 3.9-19 Critical Piping Design Methods and Criteria (Piping Design Criteria).... 3.9-113 3.9-20 Piping Packages Chosen to Demonstrate Piping Design for Piping DAC Closure ......................................................................................... 3.9-115 3-xvi Revision 10
VEGP 3&4 - UFSAR LIST OF TABLES (CONTINUED)
Table Number Title Page 3.9-201 Safety Related Snubbers ...................................................................... 3.9-117 3.11-1 Environmentally Qualified Electrical and Mechanical Equipment............ 3.11-7 3B-1 AP1000 Leak-Before-Break Bounding Analysis Systems and Parameters ...............................................................................................3B-16 3D.4-1 Typical Mild Environment Parameter Limits .............................................3D-33 3D.4-2 Equipment Post-Accident Operability Times ............................................3D-34 3D.4-3 AP1000 EQ Program Margin Requirements ............................................3D-35 3D.5-1 Normal Operating Environments ..............................................................3D-36 3D.5-2 60-Year Normal Operating Doses ............................................................3D-39 3D.5-3 Abnormal Operating Environments Inside Containment ..........................3D-40 3D.5-4 Abnormal Operating Environments Outside Containment .......................3D-41 3D.5-5 Accident Environments.............................................................................3D-44 3D.6-1 Mechanical Equipment Components Requiring Environmental Qualification..............................................................................................3D-45 3D.B-1 Example Class 1E Equipment Scope and Category Allocation ...............3D-81 3D.B-2 Aging Mechanism Sequence ........................................................................ 3D-82 3D.C-1 Radiation-Induced Degradation of Material Mechanical Properties .............. 3D-86 3D.D-1 Activation Energies From Westinghouse Reports......................................... 3D-94 3G.1-1 Summary of Models and Analysis Methods ............................................ 3G-11 3G.1-2 Summary of Dynamic Analyses and Combination Techniques............... 3G-15 3G.2-1 Steel Containment Vessel Lumped-Mass Stick Model (Without Polar Crane) Modal Properties ......................................................................... 3G-16 3G.2-2 Comparison of Frequencies for Containment Vessel Seismic Model...... 3G-18 3G.3-1 AP1000 ZPA for 2D SASSI Cases .......................................................... 3G-19 3G.4-1 Key Nodes at Location ............................................................................ 3G-20 3G.4-2 Maximum Bearing Pressure from 2D Time History Analyses.................. 3G-21 3H.5-1 Nuclear Island: Design Temperatures for Thermal Gradient....................3H-23 3H.5-2 Exterior Wall at Column Line 1 Forces and Moments in Critical Locations ..................................................................................................3H-24 3H.5-3 Exterior Wall on Column Line 1 Details of Wall Reinforcement (in 2/ft).......................................................................................................3H-25 3H.5-4 Interior Wall at Column Line 7.3 Forces and Moments in Critical Locations ..................................................................................................3H-26 3H.5-5 Interior Wall on Column Line 7.3 Details of Wall Reinforcement..............3H-27 3H.5-6 Interior Wall at Column Line L Forces and Moments in Critical Locations ..................................................................................................3H-28 3H.5-7 Interior Wall on Column Line L Details of Wall Reinforcement.................3H-29 3H.5-8 Design Summary of Spent Fuel Pool Wall Design Loads, Load Combinations, and Comparisons to Acceptance Criteria - Element No. 20477.................................................................................................3H-30 3H.5-9 Shield Building Roof Reinforcement Summary ........................................3H-37 3H.5-10 Design Summary Of Roof At Elevation 180-0, Area 6............................3H-42 3H.5-11 Design Summary Of Floor At Elevation 135-3 Area 1 (Between Column Lines M And P) ...........................................................................3H-43 3H.5-12 Design Summary Of Floor At Elevation 135-3 (Operations Work Area (Previously Known As Tagging Room) Ceiling)) ............................3H-44 3H.5-13 Design Summary Of Floor At Elevation 135-3 Area 2 (Main Control Room Ceiling) .....................................................................................................3H-45 3-xvii Revision 10
VEGP 3&4 - UFSAR LIST OF TABLES (CONTINUED)
Table Number Title Page 3H.5-14 Design Summary of Enhanced Shield Building Cylindrical Wall Load Combinations, and Comparison to Acceptance Criteria Elevation 180 Feet Near Fuel Handling Building Roof ....................................................3H-46 3H.5-15 Shield Building Roof Reinforcement Ratio of Code Required Versus Provided ...................................................................................................3H-49 3I.6-1 Potential High Frequency Sensitive Equipment List.................................. 3I-10 3I.6-2 List of Potential High Frequency Sensitive AP1000 Safety-Related electrical and Electro-mechanical Equipment ........................................... 3I-11 3I.6-3 List Of AP1000 Safety-Related Electrical and Mechanical Equipment Not High Frequency Sensitive ................................................................... 3I-32 3-xviii Revision 10
VEGP 3&4 - UFSAR LIST OF FIGURES Figure Number Title Page 3.3-1 Velocity Pressure Variation with Radius from Center of Tornado ............. 3.3-6 3.4-1 Typical Details of Nuclear Island Waterproofing Below Grade................ 3.4-23 3.4-2 Typical Details of Nuclear Island Waterproofing Below Grade with Step Back ................................................................................................ 3.4-24 3.4-3 Not Used ................................................................................................. 3.4-25 3.4-4 Typical Details of Membrane Corner Detail at Basemat and Exterior Wall ......................................................................................................... 3.4-26 3.5-201 Airports Within 30 Miles of Vogtle Facility ............................................... 3.5-26 3.6-1 Typical U-Bar Restraint ........................................................................... 3.6-53 3.6-2 Typical Structural Components of Crushable Pipe Type Whip Restraints ................................................................................................ 3.6-54 3.6-3 Terminal Ends Definitions ....................................................................... 3.6-55 3.7.1-1 Horizontal Design Response Spectra Safe Shutdown Earthquake......... 3.7-62 3.7.1-2 Vertical Design Response Spectra Safe Shutdown Earthquake ............. 3.7-63 3.7.1-3 Design Horizontal Time History, "H1" Acceleration, Velocity &
Displacement Plots.................................................................................. 3.7-64 3.7.1-4 Design Horizontal Time History, "H2" Acceleration, Velocity &
Displacement Plots.................................................................................. 3.7-65 3.7.1-5 Design Vertical Time History Acceleration, Velocity & Displacement Plots ........................................................................................................ 3.7-66 3.7.1-6 Acceleration Response Spectra of Design Horizontal Time History, "H1" ......................................................................................................... 3.7-67 3.7.1-7 Acceleration Response Spectra of Design Horizontal Time History, "H2" ......................................................................................................... 3.7-68 3.7.1-8 Acceleration Response Spectra of Design Vertical Time History............ 3.7-69 3.7.1-9 Minimum Power Spectral Density Curve (Normalized to 0.3g) ............... 3.7-70 3.7.1-10 Power Spectral Density of Design Horizontal Time History, "H1" ........... 3.7-71 3.7.1-11 Power Spectral Density of Design Horizontal Time History, "H2" ........... 3.7-72 3.7.1-12 Power Spectral Density of Design Vertical Time History......................... 3.7-73 3.7.1-13 Not Used ................................................................................................. 3.7-74 3.7.1-14 Nuclear Island Structures Dimensions .................................................... 3.7-75 3.7.1-15 Strain Dependent Properties of Rock Material (Ref. 37) ......................... 3.7-76 3.7.1-16 Strain Dependent Properties of Soil Material (Ref. 38) ........................... 3.7-77 3.7.1-17 Generic Soil Profiles ............................................................................... 3.7-78 3.7.2-1-3.7.2-11 Not Used ................................................................................................. 3.7-79 3.7.2-12 (Sheet 1 of 12) Nuclear Island Key Structural Dimensions Plan at El. 66-6 .................................................................................................. 3.7-80 3.7.2-12 (Sheet 2 of 12) Nuclear Island Key Structural Dimensions Plan at El. 82-6 .............................................................................................. 3.7-81 3.7.2-12 (Sheet 3 of 12) Nuclear Island Key Structural Dimensions Plan at El. 100-0 & 107-2............................................................................. 3.7-82 3.7.2-12 (Sheet 4 of 12) Nuclear Island Key Structural Dimensions Plan at El. 117-6 ............................................................................................ 3.7-83 3.7.2-12 (Sheet 5 of 12) Nuclear Island Key Structural Dimensions Plan at El. 135-3 ............................................................................................ 3.7-84 3.7.2-12 (Sheet 6 of 12) Nuclear Island Key Structural Dimensions Plan at El. 153-0 & 160-6............................................................................. 3.7-85 3.7.2-12 (Sheet 7 of 12) Nuclear Island Key Structural Dimensions Plan at El. 160-6, 180-0, & 329-0 .............................................................. 3.7-86 3-xix Revision 10
VEGP 3&4 - UFSAR LIST OF FIGURES (CONTINUED)
Figure Number Title Page 3.7.2-12 (Sheet 8 of 12) Nuclear Island Key Structural Dimensions Section A - A ........................................................................................... 3.7-87 3.7.2-12 (Sheet 9 of 12) Nuclear Island Key Structural Dimensions Section B - B ........................................................................................... 3.7-88 3.7.2-12 (Sheet 10 of 12) Nuclear Island Key Structural Dimensions Sections C - C and H - H ......................................................................... 3.7-89 3.7.2-12 (Sheet 11 of 12) Nuclear Island Key Structural Dimensions Section G - G........................................................................................... 3.7-90 3.7.2-12 (Sheet 12 of 12) Nuclear Island Key Structural Dimensions Section J - J............................................................................................. 3.7-91 3.7.2-13 Not Used ................................................................................................. 3.7-92 3.7.2-14 Typical Design Floor Response Spectrum .............................................. 3.7-93 3.7.2-15-3.7.2-18 Not Used ................................................................................................ 3.7-94 3.7.2-19 (Sheet 1 of 10) Annex Building Key Structural Dimensions Plan at Elevation 100-0.................................................................................. 3.7-95 3.7.2-19 (Sheet 2 of 10) Annex Building Key Structural Dimensions Plan at Elevation 107-2 and 117-6 .............................................................. 3.7-96 3.7.2-19 (Sheet 3 of 10) Annex Building Key Structural Dimensions Plan at Elevation 135-3.................................................................................. 3.7-97 3.7.2-19 (Sheet 4 of 10) Annex Building Key Structural Dimensions Plan at Elevation 158-0 and 150-3 .............................................................. 3.7-98 3.7.2-19 (Sheet 5 of 10) Annex Building Key Structural Dimensions Roof Plan at Elevation 154-0 and 181-11 3/4 ...................................................... 3.7-99 3.7.2-19 (Sheet 6 of 10) Annex Building Key Structural Dimensions Section A - A ......................................................................................... 3.7-100 3.7.2-19 (Sheet 7 of 10) Annex Building Key Structural Dimensions Section B - B ......................................................................................... 3.7-101 3.7.2-19 (Sheet 8 of 10) Annex Building Key Structural Dimensions Section C - C ......................................................................................... 3.7-102 3.7.2-19 (Sheet 9 of 10) Annex Building Key Structural Dimensions Sections D - D, E - E, & F - F ................................................................ 3.7-103 3.7.2-19 (Sheet 10 of 10) Annex Building Key Structural Dimensions Sections G - G, H - H, & J - J ................................................................ 3.7-104 3.7.2-20 East-West 2D SASSI Model with Adjacent Buildings ............................ 3.7-105 3.7.2-21 2D North-South SASSI Model with Adjacent Buildings ......................... 3.7-106 3.7.2-22 3D SASSI Model with Adjacent Buildings ............................................. 3.7-107 3.7.3-1 Impact Evaluation Zone......................................................................... 3.7-108 3.7.3-2 Impact Evaluation Zone and Seismic Supported Piping ....................... 3.7-109 3.7-201 VEGP AP1000 Horizontal Spectra Comparison.................................... 3.7-110 3.7-202 VEGP AP1000 Vertical Spectra Comparison ........................................ 3.7-111 3.8.2-1 (Sheet 1 of 3) Containment Vessel General Outline ............................. 3.8-113 3.8.2-1 (Sheet 2 of 3) Containment Vessel General Outline ............................. 3.8-114 3.8.2-1 (Sheet 3 of 3) Containment Vessel General Outline ............................. 3.8-115 3.8.2-2 Equipment Hatches ............................................................................... 3.8-116 3.8.2-3 Personnel Airlock .................................................................................. 3.8-117 3.8.2-4 (Sheet 1 of 8) Containment Penetrations Main Steam .......................... 3.8-118 3.8.2-4 (Sheet 2 of 8) Containment Penetrations Startup Feedwater ............... 3.8-119 3.8.2-4 (Sheet 3 of 8) Containment Penetrations Normal RHR Piping.............. 3.8-120 3.8.2-4 (Sheet 4 of 8) Containment Penetrations .............................................. 3.8-121 3-xx Revision 10
VEGP 3&4 - UFSAR LIST OF FIGURES (CONTINUED)
Figure Number Title Page 3.8.2-4 (Sheet 5 of 8) Containment Penetrations Fuel Transfer Penetration .... 3.8-122 3.8.2-4 (Sheet 6 of 8) Containment Penetrations Typical Electrical Penetration ............................................................................................ 3.8-123 3.8.2-4 (Sheet 7 of 8) Containment Penetrations Steam Line and Feedwater Line Insert Plates................................................................................... 3.8-124 3.8.2-4 (Sheet 8 of 8) Containment Penetrations Main Feedwater ................... 3.8-125 3.8.2-5 (Sheet 1 of 5) Containment Vessel Response to Internal Pressure of 59 psig Displaced Shape Plot ........................................................... 3.8-126 3.8.2-5 (Sheet 2 of 5) Containment Vessel Response to Internal Pressure of 59 psig Membrane Stresses (ksi) ...................................................... 3.8-127 3.8.2-5 (Sheet 3 of 5) Containment Vessel Response to Internal Pressure of 59 psig Surface Meridional Stress (ksi) ............................................. 3.8-128 3.8.2-5 (Sheet 4 of 5) Containment Vessel Response to Internal Pressure of 59 psig Outside Surface Stresses (ksi) ............................................. 3.8-129 3.8.2-5 (Sheet 5 of 5) Containment Vessel Response to Internal Pressure of 59 psig Outer Stress Intensity (ksi) ................................................... 3.8-130 3.8.2-6 (Sheet 1 of 2) Containment Vessel Axisymmetric Model ...................... 3.8-131 3.8.2-6 (Sheet 2 of 2) Containment Vessel Axisymmetric Model ...................... 3.8-132 3.8.2-7 Finite Element Model for Local Buckling Analyses................................ 3.8-133 3.8.2-8 (Sheet 1 of 2) Location of Containment Seal ........................................ 3.8-134 3.8.2-8 (Sheet 2 of 2) Seal Sections and Details............................................... 3.8-135 3.8.3-1 (Sheet 1 of 7) Structural Modules in Containment Internal Structures .. 3.8-136 3.8.3-1 (Sheet 2 of 7) Structural Modules in Containment Internal Structures .. 3.8-137 3.8.3-1 (Sheet 3 of 7) Structural Modules in Containment Internal Structures .. 3.8-138 3.8.3-1 (Sheet 4 of 7) Structural Modules in Containment Internal Structures .. 3.8-139 3.8.3-1 (Sheet 5 of 7) Structural Modules in Containment Internal Structures .. 3.8-140 3.8.3-1 (Sheet 6 of 7) Structural Modules in Containment Internal Structures .. 3.8-141 3.8.3-1 (Sheet 7 of 7) Structural Modules in Containment Internal Structures .. 3.8-142 3.8.3-2 Typical Structural Wall Module .............................................................. 3.8-143 3.8.3-3 Structural Floor Module ......................................................................... 3.8-144 3.8.3-4 Reactor Vessel Supports....................................................................... 3.8-145 3.8.3-5 (Sheet 1 of 5) Steam Generator Supports............................................. 3.8-146 3.8.3-5 (Sheet 2 of 5) Steam Generator Supports............................................. 3.8-147 3.8.3-5 (Sheet 3 of 5) Steam Generator Supports............................................. 3.8-148 3.8.3-5 (Sheet 4 of 5) Steam Generator Supports............................................. 3.8-149 3.8.3-5 (Sheet 5 of 5) Steam Generator Supports............................................. 3.8-150 3.8.3-6 (Sheet 1 of 4) Pressurizer Support Columns......................................... 3.8-151 3.8.3-6 (Sheet 2 of 4) Pressurizer Lower Lateral Supports ............................... 3.8-152 3.8.3-6 (Sheet 3 of 4) Pressurizer Lower Supports ........................................... 3.8-153 3.8.3-6 (Sheet 4 of 4) Pressurizer Upper Supports .......................................... 3.8-154 3.8.3-7 IRWST Temperature Transient ............................................................. 3.8-155 3.8.3-8 (Sheet 1 of 3) Structural Modules - Typical Design Details .................. 3.8-156 3.8.3-8 (Sheet 2 of 3) Structural Modules - Typical Design Details .................. 3.8-157 3.8.3-8 (Sheet 3 of 3) Structural Modules - Typical Design Details .................. 3.8-158 3.8.3-9 Test Tank Finite Element Model............................................................ 3.8-159 3.8.3-10 (Sheet 1 of 2) IRWST Fluid Structure Finite Element Model CIS Structural Model .................................................................................... 3.8-160 3.8.3-10 (Sheet 2 of 2) IRWST Fluid Structure Finite Element Model IRWST Structural Model .................................................................................... 3.8-161 3-xxi Revision 10
VEGP 3&4 - UFSAR LIST OF FIGURES (CONTINUED)
Figure Number Title Page 3.8.3-11 IRWST Fluid Structure Finite Element Model Fluid Model .................... 3.8-162 3.8.3-12 IRWST Fluid Structure Finite Element Model Sparger Region Detail ... 3.8-163 3.8.3-13 Effective Sections for Floor Modules ..................................................... 3.8-164 3.8.3-14 (Sheet 1 of 5) CA-01 Module................................................................. 3.8-165 3.8.3-14 (Sheet 2 of 5) CA-02 Module................................................................. 3.8-166 3.8.3-14 (Sheet 3 of 5) CA-03 Module................................................................. 3.8-167 3.8.3-14 (Sheet 4 of 5) CA-04 Structural Module ................................................ 3.8-168 3.8.3-14 (Sheet 5 of 5) CA-05 Module................................................................. 3.8-169 3.8.3-15 (Sheet 1 of 2) Typical Submodule ......................................................... 3.8-170 3.8.3-15 (Sheet 2 of 2) Typical Submodule ......................................................... 3.8-171 3.8.3-16 Liner Modules ........................................................................................ 3.8-172 3.8.3-17 (Sheet 1 of 2) Structural Modules - Design Details Standard Floor Connection ............................................................................................ 3.8-173 3.8.3-17 (Sheet 2 of 2) Structural Modules - Design Details Heavily Loaded Floor Connection ............................................................................................ 3.8-174 3.8.3-18 Location of Structural Wall Modules ...................................................... 3.8-175 3.8.4-1 (Sheet 1 of 4) Containment Air Baffle General Arrangement ................ 3.8-176 3.8.4-1 (Sheet 2 of 4) Containment Air Baffle Panel Types............................... 3.8-177 3.8.4-1 (Sheet 3 of 4) Containment Air Baffle Typical Panel on Cylinder .......... 3.8-178 3.8.4-1 (Sheet 4 of 4) Containment Air Baffle Flexible Seal .............................. 3.8-179 3.8.4-2 Passive Containment Cooling Tank ...................................................... 3.8-180 3.8.4-3 Not Used ............................................................................................... 3.8-181 3.8.4-4 (Sheet 1 of 5) Structural Modules in Auxiliary Building ......................... 3.8-182 3.8.4-4 (Sheet 2 of 5) Structural Modules in Auxiliary Building ......................... 3.8-183 3.8.4-4 (Sheet 3 of 5) Structural Modules in Auxiliary Building ......................... 3.8-184 3.8.4-4 (Sheet 4 of 5) Structural Modules in Auxiliary Building ......................... 3.8-185 3.8.4-4 (Sheet 5 of 5) Structural Modules in Auxiliary Building ......................... 3.8-186 3.8.4-5 Shield Building Structure Key Areas ..................................................... 3.8-187 3.8.5-1 Foundation Plan .................................................................................... 3.8-188 3.8.5-2 Isometric View of Finite Element Model ................................................ 3.8-189 3.8.5-3 (Sheet 1 of 7) Radial Reinforcement, Top Side of DISH ....................... 3.8-190 3.8.5-3 (Sheet 2 of 7) Circumferential Reinforcement, Top Side of DISH ......... 3.8-191 3.8.5-3 (Sheet 3 of 7) Longitudinal Reinforcement Map, Top Side in NS Direction .......................................................................................... 3.8-192 3.8.5-3 (Sheet 4 of 7) Longitudinal Reinforcement Map, Top Side in EW Direction ......................................................................................... 3.8-193 3.8.5-3 (Sheet 5 of 7) Longitudinal Reinforcement, Bottom Side of DISH and 6 Basemat (NS) ............................................................................. 3.8-194 3.8.5-3 (Sheet 6 of 7) Longitudinal Reinforcement, Bottom Side of DISH and 6 Basemat (EW) ............................................................................ 3.8-195 3.8.5-3 (Sheet 7 of 7) Shear Reinforcement Map.............................................. 3.8-196 3.9-1-3.9-3 Not Used ............................................................................................... 3.9-118 3.9-4 Control Rod Drive Mechanism .............................................................. 3.9-119 3.9-5 Lower Reactor Internals ........................................................................ 3.9-120 3.9-6 Upper Core Support Structure............................................................... 3.9-121 3.9-7 Integrated Head Package...................................................................... 3.9-122 3.9-8 Reactor Internals Interface Arrangement .............................................. 3.9-123 3.9-9 Flow Skirt Schematic ............................................................................. 3.9-124 3B-1 Typical Bounding Analysis Curve (BAC) ..................................................3B-18 3-xxii Revision 10
VEGP 3&4 - UFSAR LIST OF FIGURES (CONTINUED)
Figure Number Title Page 3B-2 Bounding Analysis Curve for Primary Loop Hot Leg ................................3B-19 3B-3 Bounding Analysis Curve for Primary Loop Cold Leg ..............................3B-20 3B-4 Bounding Analysis Curve for 38 Main Steam Line..................................3B-21 3B-5 Bounding Analysis Curve for 20 Normal RHR ........................................3B-22 3B-6 (Sheet 1 of 2) Bounding Analysis Curve for 18 Surge Line.....................3B-23 3B-6 (Sheet 2 of 2) Bounding Analysis Curve for 18 Surge Line.....................3B-24 3B-7 Bounding Analysis Curve for 18 PRHR Supply/ADS 4 ...........................3B-25 3B-8 Bounding Analysis Curve for 14 PRHR Supply to Cold Trap, PRHR Supply/ADS4 .................................................................................3B-26 3B-9 Bounding Analysis Curve for 14 PRHR Supply after Cold Trap, Return - to Isolation Valve .......................................................................3B-27 3B-10 (Sheet 1 of 2) Bounding Analysis Curve for 14 ADS Stage 2, 3 .............3B-28 3B-10 (Sheet 2 of 2) Bounding Analysis Curve for 14 ADS Stage 2, 3 .............3B-29 3B-11 Bounding Analysis Curve for 14 PRHR Return - after Isolation Valve, 14 PRHR Return ..........................................................................3B-30 3B-12 Not Used ..................................................................................................3B-31 3B-13 Bounding Analysis Curve for 8 Accumulator to Isolation Valve ..............3B-32 3B-14 Bounding Analysis Curve for 8 CMT Cold Leg Balance Line and Vent, DVI Cold Trap to RV .......................................................................3B-33 3B-15 Bounding Analysis Curve for 8 CMT, DVI IRWST (Various Sections) ..................................................................................................3B-34 3B-16 Not Used ..................................................................................................3B-35 3B-17 Bounding Analysis Curve for Accumulator after Isolation Valve ..............3B-36 3B-18 Bounding Analysis Curve for RNS Discharge ..........................................3B-37 3B-19 (Sheet 1 of 2) Bounding Analysis Curve for ADS Header to RCS Safety Valve .............................................................................................3B-38 3B-19 (Sheet 2 of 2) Bounding Analysis Curve for ADS Header to RCS Safety Valve .............................................................................................3B-39 3B-20 Bounding Analysis Curve for 12 Normal RHR ........................................3B-40 3B-21 Bounding Analysis Curve for 10 Normal RHR ........................................3B-41 3B-22 Bounding Analysis Curve for 8 ADS Stage 2, 3 ......................................3B-42 3D.5-1 (Sheet 1 of 3) Typical Abnormal Environmental Test Profile: Main Control Room ...........................................................................................3D-46 3D.5-1 (Sheet 2 of 3) Typical Abnormal Environmental Test Profile: I&C and DC Equipment Rooms.......................................................................3D-47 3D.5-1 (Sheet 3 of 3) Typical Abnormal Environmental Test Profile: Voltage and Frequency Variations ........................................................................3D-48 3D.5-2 Gamma Dose and Dose Rate Inside Containment After a LOCA ............3D-49 3D.5-3 Beta Dose and Dose Rate Inside Containment After a LOCA .................3D-50 3D.5-4 Gamma Dose and Dose Rate Inside Containment After a Steam Line Break ................................................................................................3D-51 3D.5-5 Beta Dose and Dose Rate Inside Containment After a Steam Line Break ........................................................................................................3D-52 3D.5-6-3D.5-7 Not Used ..................................................................................................3D-53 3D.5-8 (Sheet 1 of 2) Typical Combined LOCA/SLB/FLB EQ Design Envelope for Inside Containment Temperature........................................3D-54 3D.5-8 (Sheet 2 of 2) Typical Combined LOCA/SLB/FLB EQ Design Envelope for Inside Containment Pressure ..............................................3D-55 3-xxiii Revision 10
VEGP 3&4 - UFSAR LIST OF FIGURES (CONTINUED)
Figure Number Title Page 3D.5-9 (Sheet 1 of 2) MSIV Compartment Response to MSLB (Short Term) ........................................................................................................3D-56 3D.5-9 (Sheet 2 of 2) MSIV Compartment Response to MSLB (Long Term) ........................................................................................................3D-57 3D.C-1 Histogram of Threshold Gamma Dose for Mechanical Damage to Elastomers, Plastics, and Encapsulation Compounds.................................. 3D-88 3D.D-1 Frequency Distribution of Activation Energies of Various Components/Materials (EPRI Data).............................................................. 3D-96 3D.D-2 Frequency Distribution of Activation Energies of Various Components/Materials (Westinghouse Data) ............................................... 3D-97 3D.D-3 Not Used ....................................................................................................... 3D-98 3E-1 (Sheet 1 of 2) High Energy Piping - Steam Generator System .................3E-2 3E-1 (Sheet 2 of 2) High Energy Piping - Steam Generator System .................3E-3 3E-2 High Energy Piping - Normal Residual Heat Removal System .................3E-4 3E-3 (Sheet 1 of 2) High Energy Piping - Reactor Coolant System ...................3E-5 3E-3 (Sheet 2 of 2) High Energy Piping - Reactor Coolant System ...................3E-6 3E-4 (Sheet 1 of 2) High Energy Piping - Passive Core Cooling System .......................................................................................................3E-7 3E-4 (Sheet 2 of 2) High Energy Piping - Passive Core Cooling System ..........3E-8 3E-5 (Sheet 1 of 2) High Energy Piping - Chemical and Volume Control System ...........................................................................................3E-9 3E-5 (Sheet 2 of 2) High Energy Piping - Chemical and Volume Control System .........................................................................................3E-10 3G.1-1 Nuclear Island Seismic Analysis Models ................................................. 3G-22 3G.2-1 3D Finite Element Model of Coupled Shield and Auxiliary Building ........ 3G-23 3G.2-2 3D Finite Element Model of Containment Internal Structures ................. 3G-24 3G.2-3 3D Finite Element Model of Containment Outer Basemat (Dish) ............ 3G-25 3G.2-4 Steel Containment Vessel and Polar Crane Models ............................... 3G-26 3G.2-5A Polar Crane Model Simplified Model ....................................................... 3G-27 3G.2-5B Polar Crane Model Detailed Model ......................................................... 3G-28 3G.2-6 Reactor Coolant Loop Lumped-Mass Stick Model .................................. 3G-29 3G.2-7 Pressurizer Model ................................................................................... 3G-30 3G.2-8 Core Makeup Tank Models ..................................................................... 3G-31 3G.2-9 AP1000 Nuclear Island Solid-Shell Model (NI10).................................... 3G-32 3G.2-10 Containment Internal Structure with the SCV, PC, Reactor Coolant Loop, and Pressurizer ............................................................................. 3G-33 3G.2-11 Soil Structure Interaction Model - NI20 Looking East ............................. 3G-34 3G.2-12 Coarse Model of Containment Internal Structures .................................. 3G-35 3G.2-13 Fine Mesh (NI05) Model of Auxiliary and Shield Building ....................... 3G-36 3G.2-14 NI05 Model of Containment Internal Structures ...................................... 3G-37 3G.2-15 3D NI05 Refined Mesh Model of Outer Containment Basemat (Dish) ....................................................................................................... 3G-38 3G.2-16 Quadrant Model of Shield Building Roof ................................................ 3G-39 3G.2-17 Detailed 3D Finite Element Model of Containment Vessel Including Large Penetrations .................................................................................. 3G-40 3G.2-18 Axisymmetric Model of Containment Vessel ........................................... 3G-41 3G.2-19 Schematic of Non-linear 2D East-West Nuclear Island Stick Model Used for Stability Evaluation that Addresses Sliding and Overturning .... 3G-42 3G.3-1 Generic Soil Profiles ................................................................................ 3G-43 3-xxiv Revision 10
VEGP 3&4 - UFSAR LIST OF FIGURES (CONTINUED)
Figure Number Title Page 3G.3-2 2D SASSI FRS - Node 41 X (ASB El. 99).............................................. 3G-44 3G.3-3 2D SASSI FRS - Node 41 Y (ASB El. 99).............................................. 3G-45 3G.3-4 2D SASSI FRS - Node 120 X (ASB El. 179.6)....................................... 3G-46 3G.3-5 2D SASSI FRS - Node 120 Y (ASB El. 179.6)....................................... 3G-47 3G.3-6 2D SASSI FRS - Node 310 X (ASB El. 333.2)....................................... 3G-48 3G.3-7 2D SASSI FRS - Node 310 Y (ASB El. 333.2)....................................... 3G-49 3G.3-8 2D SASSI FRS - Node 411 X (SCV El. 200.0) ...................................... 3G-50 3G.3-9 2D SASSI FRS - Node 411 Y (SCV El. 200.0) ...................................... 3G-51 3G.3-10 2D SASSI FRS - Node 535 X (CIS El. 134.3)........................................ 3G-52 3G.3-11 2D SASSI FRS - Node 535 Y (CIS El. 134.3)........................................ 3G-53 3G.4-1 Auxiliary Shield Building Rigid Nodes at El. 135 .................................. 3G-54 3G.4-2 Auxiliary Shield Building Flexible Nodes at El. 135 .............................. 3G-55 3G.4-3 Excavated Soil......................................................................................... 3G-56 3G.4-4 Additional Elements for Soil Pressure Calculations ................................ 3G-57 3G.4-5X X Direction FRS for Node 130401 (NI10) or 1761 (NI20) CIS at Reactor Vessel Support Elevation of 100 ............................................... 3G-58 3G.4-5Y Y Direction FRS for Node 130401 (NI10) or 1761 (NI20) CIS at Reactor Vessel Support Elevation of 100 ..................................................3G-59 3G.4-5Z Z Direction FRS for Node 130401 (NI10) or 1761 (NI20) CIS at Reactor Vessel Support Elevation of 100 .............................................. 3G-60 3G.4-6X X Direction FRS for Node 105772 (NI10) or 2199 (NI20) CIS at Operating Deck Elevation 134.25 .......................................................... 3G-61 3G.4-6Y Y Direction FRS for Node 105772 (NI10) or 2199 (NI20) CIS at Operating Deck Elevation 134.25 .......................................................... 3G-62 3G.4-6Z Z Direction FRS for Node 105772 (NI10) or 2199 (NI20) CIS at Operating Deck Elevation 134.25 .......................................................... 3G-63 3G.4-7X X Direction FRS for Node 4724 (NI10) or 2078 (NI20) ASB Control Room Side Elevation 116.50 .........................................................3G-64 3G.4-7Y Y Direction FRS for Node 4724 (NI10) or 2078 (NI20) ASB Control Room Side Elevation 116.50 .........................................................3G-65 3G.4-7Z Z Direction FRS for Node 4724 (NI10) or 2078 (NI20) ASB Control Room Side Elevation 116.50 .........................................................3G-66 3G.4-8X X Direction FRS for Node 5754 (NI10) or 2675 (NI20) ASB Fuel Building Roof Elevation 179.19............................................................... 3G-67 3G.4-8Y Y Direction FRS for Node 5754 (NI10) or 2675 (NI20) ASB Fuel Building Roof Elevation 179.19............................................................... 3G-68 3G.4-8Z Z Direction FRS for Node 5754 (NI10) or 2675 (NI20) ASB Fuel Building Roof Elevation 179.19 ..................................................................3G-69 3G.4-9X X Direction FRS for Node 8573 (NI10) or 3329 (NI20) ASB Shield Building Roof Elevation 327.41............................................................... 3G-70 3G.4-9Y Y Direction FRS for Node 8573 (NI10) or 3329 (NI20) ASB Shield Building Roof Elevation 327.41............................................................... 3G-71 3G.4-9Z Z Direction FRS for Node 8573 (NI10) or 3329 (NI20) ASB Shield Building Roof Elevation 327.41............................................................... 3G-72 3G.4-10X X Direction FRS for Node 130412 (NI10) or 2788 (NI20) SCV Near Polar Crane Elevation 224.00 ........................................................ 3G-73 3G.4-10Y Y Direction FRS for Node 130412 (NI10) or 2788 (NI20) SCV Near Polar Crane Elevation 224.00 ...........................................................3G-74 3-xxv Revision 10
VEGP 3&4 - UFSAR LIST OF FIGURES (CONTINUED)
Figure Number Title Page 3G.4-10Z Z Direction FRS for Node 130412 (NI10) or 2788 (NI20) SCV Near Polar Crane Elevation 224.00 ...........................................................3G-75 3H.2-1 General Layout of Auxiliary Building ........................................................3H-50 3H.5-1 (Sheet 1 of 3) Nuclear Island Critical Sections Plan at El. 135-3 ...........3H-51 3H.5-1 (Sheet 2 of 3) Nuclear Island Critical Sections Plan at El. 180-0 ...........3H-52 3H.5-1 (Sheet 3 of 3) Nuclear Island Critical Sections Section A-A .....................3H-53 3H.5-2 (Sheet 1 of 3) Wall on Column Line 1 ......................................................3H-54 3H.5-2 (Sheet 2 of 3) Wall on Column Line 7.3 ...................................................3H-55 3H.5-2 (Sheet 3 of 3) Wall on Column Line L ......................................................3H-56 3H.5-3 Typical Reinforcement in Wall on Column Line 1.....................................3H-57 3H.5-4 Typical Reinforcement in Wall 7.3 ............................................................3H-58 3H.5-5 (Sheet 1 of 3) Concrete Reinforcement in Wall 11 ...................................3H-59 3H.5-5 (Sheet 2 of 3) Concrete Reinforcement Layers in Wall 11 (Looking East) .........................................................................................................3H-60 3H.5-5 (Sheet 3 of 3) Wall 11 at Main Steamline Anchor Section A-A ................3H-61 3H.5-6 Auxiliary Building Typical Composite Floor ..............................................3H-62 3H.5-7 Typical Reinforcement and Connection to Shield Building.......................3H-63 3H.5-8 Auxiliary Building Operations Work Area (Tagging Room) Ceiling...........3H-64 3H.5-9 (Sheet 1 of 3) Auxiliary Building Finned Floor ..........................................3H-65 3H.5-9 (Sheet 2 of 3) Auxiliary Building Finned Floor ..........................................3H-66 3H.5-9 (Sheet 3 of 3) Auxiliary Building Finned Floor ..........................................3H-67 3H.5-10 Spent Fuel Pool Wall Divider Wall Element Locations .............................3H-68 3H.5-11 (Sheet 1 of 6) Design of Shield Building: Roof and Air Inlets ...................3H-69 3H.5-11 (Sheet 2 of 6) Design of Shield Building: Concrete Detail at Tension Ring ............................................................................................3H-70 3H.5-11 (Sheet 3 of 6) Design of Shield Building: Roof/Air Inlet Interface .............3H-71 3H.5-11 (Sheet 4 of 6) Design of Shield Building at Air Inlets................................3H-72 3H.5-11 (Sheet 5 of 6) Design of Shield Building: Tank/Roof Interface Reinforcement ..........................................................................................3H-73 3H.5-11 (Sheet 6 of 6) Design of Shield Building: Tank/Compression Ring Roof Interface Reinforcement ..................................................................3H-74 3H.5-12 Typical Reinforcement in Wall L ...............................................................3H-75 3H.5-13 Enhanced Shield Building Wall Panel Layout ..........................................3H-76 3H.5-14 Elevation View of Tension Ring and Air Inlets..........................................3H-77 3H.5-15 Shield Building Tension Ring ...................................................................3H-78 3H.5-16 (Sheet 1 of 2) Design of Shield Building: Surface Plates on Cylindrical Section - Developed View 90-270 Degrees ...........................3H-79 3H.5-16 (Sheet 2 of 2) Design of Shield Building: Surface Plates on Cylindrical Section - Developed View 270-90 Degrees ...........................3H-80 3I.1-1 Comparison of Horizontal AP1000 CSDRS and HRHF Envelope Response Spectra ..................................................................................... 3I-55 3I.1-2 Comparison of Vertical AP1000 CSDRS and HRHF Envelope Response Spectra ..................................................................................... 3I-56 3-xxvi Revision 10
VEGP 3&4 - UFSAR Chapter 3 Design of Structures, Components, Equipment and Systems 3.1 Conformance with Nuclear Regulatory Commission General Design Criteria This section discusses the extent to which the AP1000 design criteria for safety-related structures, systems, and components comply with 10 CFR 50, Appendix A. As presented in this section, each criterion is first quoted and then discussed. For some criteria, the AP1000 advanced passive design features are deemed to be significantly different in certain specific areas from those design features considered when the General Design Criteria were formulated. In those instances, the means by which the AP1000 design complies with the intent of the General Design Criterion is indicated. Where additional information is required for a complete discussion, the appropriate Design Control Document (DCD) sections are referenced.
3.1.1 Overall Requirements Criterion 1 - Quality Standards and Records Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety function to be performed.
Where generally recognized 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 safety functions. Appropriate records of the design, fabrication, erection, and testing of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit.
AP1000 Compliance The Quality Assurance Program for the AP1000 provides confidence that safety-related items and services are designed, procured, fabricated, inspected, and tested to quality standards commensurate with the safety-related functions to be performed. This program also applies to design services subcontracted to external organizations. The quality assurance program for erection of structures, systems, and components will be identified before the construction phase of the AP1000 project. The AP1000 quality assurance program is described in Chapter 17, including its compliance with ASME NQA-1.
Design, procurement, fabrication, inspection, and testing are performed according to recognized codes, standards, and design criteria that comply with the requirements of 10 CFR 50.55a. As necessary, supplemental standards, design criteria, and requirements are developed by the AP1000 designers. A portion of the chemical and volume control system that is defined as reactor coolant pressure boundary uses an alternate classification in conformance with the requirements of 10 CFR 50.55a(z). The alternate classification is discussed in Subsection 5.2.1.3.
Appropriate records documenting that design, procurement, fabrication, inspection, and testing comply with the applicable codes, standards, and design criteria are maintained according to appropriate, applicable laws and regulations, either by or under the control of the Combined License applicant.
In the passive AP1000 design, systems necessary to provide the reactor coolant pressure boundary, the capability to shut down the reactor and maintain it in a safe shutdown condition, and the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures 3.1-1 Revision 5
VEGP 3&4 - UFSAR comparable to the guideline exposures of 10 CFR 50.34 are classified as safety-related. Therefore, the AP1000 complies with the intent of Criterion 1.
The principal design criteria, design bases, codes, and standards applied to the facility are identified in Section 3.2. Additional details may be found in the pertinent sections dealing with safety-related structures, systems, and components.
Criterion 2 - Design Bases for Protection Against Natural Phenomena Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without the loss of the 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 conditions with the effects of the natural phenomena, and (3) the importance of the safety functions to be performed.
AP1000 Compliance The safety-related structures, systems, and components are designed to withstand the effects of natural phenomena without loss of the capability to perform their safety-related functions, or are designed such that their response or failure will be in a safe condition. Those structures, systems, and components vital to the shutdown capability of the reactor are designed to withstand the maximum probable natural phenomena at the intended site.
Accident analyses consider conservative site conditions that envelope expected sites. Appropriate combinations of structural loadings from normal, accident, and natural phenomena are considered in the plant design. The design of the plant in relationship to those natural phenomena is addressed.
Seismic and quality group classifications and other pertinent standards and information are given in the sections discussing individual structures, systems, and components as well as in Chapter 3. The nature and magnitude of the natural phenomena considered in the design of this plant are discussed in Chapter 2.
Criterion 3 - Fire Protection 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.
AP1000 Compliance The safety-related structures, systems, and components are designed to minimize the probability and effect of fires and explosions. Noncombustible and fire-resistant materials are used in the containment and main control room. Additionally noncombustible and fire-resistant materials are used on components of safety-related systems, and elsewhere in the plant where fire is a potential risk to safety-related systems.
3.1-2 Revision 5
VEGP 3&4 - UFSAR For example, electrical cables have a fire-retardant jacketing, and fire barriers are used at fire area boundaries. The AP1000 design approach includes designing the safety-related systems with redundant divisions, and locating these redundant divisions in separate safety-related areas.
Equipment and facilities for fire protection, including detecting, alarming, and extinguishing functions, are provided to help protect both plant equipment and personnel from fire, explosion, and the resultant release of toxic vapors. Fire protection is provided by deluge systems (water spray),
sprinklers, and portable extinguishers. Fire fighting systems are designed so that their rupture or inadvertent operation will not prevent safety-related systems from performing their design functions.
The following codes, guides, and standards are used as guidelines in the design of the fire protection system and equipment. The system and equipment conform to the applicable portions of the following documents:
National Fire Protection Association Codes and Standards BTP-CMEB 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants," July 1981 Subsection 9.5.1 describes the AP1000 fire protection system and equipment, including conformance with the applicable portions of these codes and standards and reference to specific fire protection codes and standards.
Criterion 4 - Environmental and Dynamic Effects Design Bases 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 and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.
AP1000 Compliance Safety-related structures, systems, and components are 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.
The AP1000 design has emphasized the minimization of missiles, pipe whip, and fluid discharge by a combination of separation of safe shutdown components and design to prevent the dynamic effects of postulated pipe ruptures based on the application of the leak-before-break approach. This analysis is discussed in Subsection 3.4.3.5 and Section 3.6.
The AP1000 structures, systems, and components are appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. Details of the design, environmental testing, and construction of these structures, systems, and components are given in the sections that discuss individual structures, systems, and components, as well as in Sections 3.5 and 3.6.
3.1-3 Revision 5
VEGP 3&4 - UFSAR Criterion 5 - Sharing of Structures, Systems, and Components 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 unit.
AP1000 Compliance The AP1000 is a single-unit plant. If more than one unit were built on the same site, none of the safety-related systems would be shared.
3.1.2 Protection by Multiple Fission Product Barriers Criterion 10 - Reactor Design 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 effects of anticipated operational occurrences.
AP1000 Compliance The reactor core and associated coolant, control, and protection systems are designed to the following criteria:
No fuel damage occurs during normal core operation and operational transients (Condition I) or during transient conditions arising from occurrences of moderate frequency (Condition II).
For normal operation, the plant is designed to accommodate a fuel defect level of up to 0.25 percent. Fuel damage, as used here, is defined as penetration of the fission product barrier, that is, the fuel rod cladding. The small number of clad defects that may occur are within the capability of the plant cleanup system and are consistent with the plant design bases. For additional information see Section 11.1.
The reactor can be returned to a safe shutdown state following a Condition III event, with only a small fraction of the fuel rods damaged, although sufficient fuel damage might occur to preclude the immediate resumption of operation.
The core remains intact with acceptable heat transfer geometry following transients arising from occurrences of limiting faults (Condition IV).
The reactor protection system is designed to actuate a reactor trip whenever necessary to prevent exceeding the fuel design limits. The core design, together with the process and decay heat removal systems, provide this capability under expected conditions of normal operation, with appropriate margins for uncertainties and anticipated transient situations. This includes the effects of the loss of reactor coolant flow, trip of the turbine generator, loss of normal feedwater, and loss of both normal and preferred power sources.
Chapter 4, Reactor, describes the mechanical components of the reactor and reactor core, including the fuel rods and fuel assemblies, the mechanical design, nuclear design, and the thermal hydraulic design. Chapter 7 provides details of the control and protection systems instrumentation design and logic. This information supports the accident analyses documented in Chapter 15. The acceptable fuel design limits are not exceeded for Condition I and II events. Acceptable core cooling is provided for Condition III and IV events.
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VEGP 3&4 - UFSAR Criterion 11 - Reactor Inherent Protection The reactor core and associated coolant systems shall be designed so that in the power-operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.
AP1000 Compliance When the reactor is critical, the negative fuel temperature reactivity effects (Doppler feedback) provide prompt reactivity feedback to compensate for a rapid, uncontrolled reactivity excursion. The negative Doppler coefficient of reactivity is provided by the use of a low-enrichment fuel design. This Doppler feedback is the primary reactivity feedback mechanism to provide the inherent core reactivity protection during rapid core reactivity excursions.
For slower reactivity transients that result in moderator temperature increases, the nonpositive moderator temperature coefficient of reactivity provides compensatory reactivity feedback to help control these slower transients. The overall core design establishes a nonpositive moderator temperature coefficient of reactivity.
Chapter 4 provides information pertaining to the core design.
Criterion 12 - Suppression of Reactor Power Oscillations 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 readily detected and suppressed.
AP1000 Compliance Power oscillations of the fundamental mode are inherently eliminated by negative Doppler and nonpositive moderator temperature coefficients of reactivity.
Oscillations, due to xenon spatial effects, in the radial and azimuthal overtone modes are heavily damped because of the inherent design and due to the negative Doppler and nonpositive moderator temperature coefficients of reactivity.
Oscillations due to xenon spatial effects may occur in the axial first overtone mode. Reactor trip functions are provided, using the measured axial power imbalance as an input, so that the fuel design limits are not exceeded during axial xenon oscillations.
If it is necessary to maintain axial imbalance within the limits (that is, imbalances that are alarmed to the operator and are within the imbalance trip setpoints), the operator can suppress axial xenon oscillations by control rod motions or temporary power reductions or both.
Oscillations due to spatial xenon effects, in axial modes higher than the first overtone, are heavily damped because of the inherent design and the negative Doppler coefficient of reactivity.
The stability of the core against xenon-induced power oscillations and the functional requirements of instrumentation for monitoring and measuring core power distribution are discussed in Chapter 4.
Details of the instrumentation design and logic are discussed in Chapter 7.
Criterion 13 - Instrumentation and Control Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its 3.1-5 Revision 5
VEGP 3&4 - UFSAR associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.
AP1000 Compliance Instrumentation and controls are provided to monitor and control neutron flux, control rod position, fluid temperatures, pressures, flows, and levels, as necessary, to maintain plant safety.
Instrumentation is provided in the reactor coolant system, steam and power conversion system, containment, engineered safety systems, radioactive waste management systems, and other auxiliary systems.
See Section 7.5 for a discussion of indications that are required for operator use under normal operating and accident conditions. Criteria regarding layout of the controls and displays are provided in Chapter 18.
The quantity and types of process instrumentation used provide safe and orderly operation of systems over the design range of plant operations, including accident conditions.
Criterion 14 - Reactor Coolant Pressure Boundary 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.
AP1000 Compliance The reactor coolant pressure boundary is designed to accommodate the system pressures and temperatures attained under the expected modes of plant operation, including anticipated transients, while maintaining stresses within applicable limits. Consideration is given to loadings under normal operating conditions and to abnormal loadings, such as seismic loadings. The piping is protected from overpressure by means of pressure-relieving devices, as required by ASME Code,Section III.
See Subsection 5.2.2 for additional information.
Reactor coolant pressure boundary materials and fabrication techniques are such that there is a low probability of gross rupture or significant leakage. The AP1000 reactor coolant system design incorporates revised pipe-break criteria (leak-before-break) to reduce or eliminate the need to consider the dynamic effects of pipe breaks. The configuration and materials of the reactor coolant system have been selected such that the pipe stresses meet the leak-before-break criteria. See Subsection 3.6.3 for additional information.
The AP1000 reactor core and reactor internals are designed to limit neutron fluence on the reactor vessel. See Section 5.4 and Chapter 4 for additional information.
The reactor vessel is manufactured from low-alloy carbon steel clad with 308L stainless steel weld overlay on wetted surfaces. The vessel shell is constructed of ring-rolled forgings that eliminate vertical weld seams. Chemical composition of the forging material is controlled to improve radiation resistance of the vessel. (See Criterion 31 for further discussion of the reactor coolant pressure boundary.)
Coolant chemistry is controlled to protect the materials of construction of the reactor coolant pressure boundary from corrosion. See Subsection 5.2.3 for additional information.
The reactor coolant pressure boundary welds are accessible for in-service inspections to assess structural and leaktight integrity. For the reactor vessel, a material surveillance program is provided.
Instrumentation is provided to detect significant leakage from the reactor coolant pressure boundary, with indication in the main control room. See Subsection 5.2.4 for additional information.
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VEGP 3&4 - UFSAR A portion of the chemical and volume control system that is defined as reactor coolant pressure boundary is nonsafety-related. This portion of the system is capable of being automatically isolated by safety-related valves that are designed and qualified for the design requirements.
Criterion 15 - Reactor Coolant System Design 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 normal operation, including anticipated operational occurrences.
AP1000 Compliance Steady-state and transient analyses are performed to demonstrate that reactor coolant system design conditions are not exceeded during normal operation. Protection and control setpoints are based on these analyses. See Chapter 15 for additional information.
The reactor coolant system stress analysis and the leak-before-break analyses are described in Appendices 3B and 3C. See Section 5.3 for additional information.
Two safety valves are provided for the reactor coolant system. These valves and their setpoints meet the ASME Code,Section III criteria for overpressure protection. See Subsection 5.2.2 for additional information.
Criterion 16 - Containment Design The reactor containment and associated 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.
AP1000 Compliance The containment is an integral part of the overall containment system, whose function is to contain the release of airborne radioactivity following postulated design basis accidents and to provide shielding for the reactor core and the reactor coolant system during normal operations. The containment consists of a steel containment vessel and is surrounded by a concrete shield building.
The containment vessel, which is a free-standing steel shell, is an integral part of the passive containment cooling system, whose function is to provide the safety-related ultimate heat sink for the removal of the reactor coolant system sensible heat, core decay heat, and stored energy. The containment vessel and the passive containment cooling system are designed to remove sufficient energy from the containment to prevent the containment from exceeding its design pressure following postulated design basis accidents.
The containment is designed to house the reactor coolant system and other related systems. The containment vessel functions as an essentially leaktight barrier. It is protected against postulated missiles from external sources as well as missiles produced by internal equipment failures.
Containment penetrations are isolated according to the provisions of GDCs 54, 55, 56, and 57.
Criterion 17 - Electrical Power Systems 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 that 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 3.1-7 Revision 5
VEGP 3&4 - UFSAR occurrences and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents.
The onsite electric power supplies, including the batteries, and the onsite electric distribution system 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 minimize, to the extent practical, the likelihood of their simultaneous failure under operating and postulated accident and environmental conditions. A switchyard common to both circuits is acceptable. Each of these circuits shall be designed to be available in sufficient time, following a loss of all onsite alternating current power supplies and 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 be 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 functions are maintained.
Provisions shall be included to minimize the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission network, or the loss of power from the onsite electric power supplies.
AP1000 Compliance The AP1000 plant design supports an exemption to the requirement of GDC 17 for two physically independent offsite circuits by providing safety-related passive systems for core cooling and containment integrity, and multiple nonsafety-related onsite and offsite electric power sources for other functions. See Section 6.3 for additional information on the systems for core cooling.
A reliable dc power source supplied by batteries provides power for the safety-related valves and instrumentation during transient and accident conditions.
The Class 1E dc and UPS system is the only safety-related power source required to monitor and actuate the safety-related passive systems. Otherwise, the plant is designed to maintain core cooling and containment integrity, independent of nonsafety-related ac power sources indefinitely. The only electric power source necessary to accomplish these safety-related functions is the Class 1E dc and UPS power system which includes the associated safety-related 120V ac distribution switchgear.
Although the AP1000 is designed with reliable nonsafety-related offsite and onsite ac power that are normally expected to be available for important plant functions, nonsafety-related ac power is not relied upon to maintain the core cooling or containment integrity.
The nonsafety-related ac power system is designed such that plant auxiliaries can be powered from the grid under all modes of operation. During loss of normal ac power and offsite power, the ac power is supplied by the onsite standby diesel-generators. Preassigned loads and equipment are automatically loaded on the diesel-generators in a predetermined sequence. Additional loads can be manually added as required. The onsite standby power system is not required for safe shutdown of the plant.
Criterion 18 - Inspection and Testing of Electric Power Systems 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 condition of their components. The 3.1-8 Revision 5
VEGP 3&4 - UFSAR 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 system, and the transfer of power among the nuclear power unit, the offsite power system, and the onsite power system.
AP1000 Compliance The AP1000 is designed so that only the Class 1E dc and UPS system is required in order to initiate and actuate the systems necessary for maintaining core cooling and containment integrity. The safety-related dc power system design complies with GDC 18. Compliance with GDC 18 is achieved by designing testability and inspection capability into the system. The associated testing requirements are contained in the Technical Specifications.
Criterion 19 - Control Room 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 condition 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.
AP1000 Compliance The AP1000 main control room provides the man-machine interfaces required to operate the plant safely and efficiently under normal conditions and to maintain it in a safe manner under accident conditions, including LOCAs. Simplified passive safety-related system designs are provided that do not rely upon operator action to maintain core cooling for design basis accidents. Operator action outside the main control room to mitigate the consequences of an accident is permitted.
The main control room is shielded by the containment and auxiliary building from direct gamma radiation and inhalation doses resulting from the postulated release of fission products inside containment. Refer to Chapter 15 for additional information on accident conditions. The main control room/control support area HVAC subsystem of the nuclear island nonradioactive ventilation system (VBS) allows access to and occupancy of the main control room under accident conditions as described in Subsection 9.4.1. Sufficient shielding and the main control room/control support area HVAC subsystem provide adequate protection so that personnel will not receive radiation exposure in excess of 5 rem whole-body or its equivalent to any part of the body for the duration of the accident.
If ac power is unavailable for more than 10 minutes or if main control room differential pressure is below the Low setpoint for more than 10 minutes or if High-2 particulate or High-2 iodine radioactivity is detected in the main control room supply air duct, which may lead to exceeding General Design Criteria 19 operator dose limits, the protection and safety monitoring system automatically isolates the main control room and operator habitability requirements are then met by the main control room emergency habitability system (VES). The main control room emergency habitability system also allows access to and occupancy of the main control room under accident conditions. The emergency main control room habitability system is designed to satisfy seismic Category I requirements as described in Section 3.2; the system design is described in Section 6.4.
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VEGP 3&4 - UFSAR In the event that the operators are forced to abandon the main control room, a workstation is provided with remote shutdown capability. A main control room evacuation is not assumed to occur simultaneously with design basis events. The remote shutdown workstation is described in Section 7.4.
3.1.3 Protection and Reactivity Control Systems Criterion 20 - Protection System Functions The protection system shall be designed (1) to initiate automatically 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.
AP1000 Compliance The protection system is a microprocessor-based system that trips the reactor and actuates engineered safety features when predetermined limits are exceeded or when manually initiated.
The reactor trip portion of the protection system includes four independent, redundant, physically separated, electrically-isolated divisions. The coincidence circuits guard against the loss of protection or the generation of false protection signals due to equipment failures through the use of a two-out-of-four logic and built-in operational bypasses.
Independent, redundant, physically separated, electrically-isolated engineered safety features trains are provided. Signal conditioning for the plant sensors is provided.
See Chapter 7 for additional information concerning the design of the protection system.
Criterion 21 - Protection System Reliability and Testability The protection system shall be designed for high functional reliability and in-service testability commensurate with the safety functions to be performed. Redundancy and independence designed into the protection system shall be sufficient to assure that (l) no single failure results in the loss of the protection function and (2) removal from service of any component or channel does not result in the loss of the required minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated. 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.
AP1000 Compliance The protection system is designed for functional reliability and in-service testability. The design employs redundant logic trains and measurement and equipment diversity.
The protection system equipment includes integral testing circuits. System equipment, from input to output, in the protection cabinets and the engineered safety features cabinets, is tested. Simulated inputs replace the field signals. Outputs are monitored for validity. Manual and automatic testing is used to test the final stages of the reactor trip circuits and the reactor trip switchgear. Testing of cabinets and communications links verifies the functional operation of the equipment and the hardware. See Chapter 7 for further information concerning the test capabilities of the protection system.
Criterion 22 - Protection System Independence 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 3.1-10 Revision 5
VEGP 3&4 - UFSAR do not result in the 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.
AP1000 Compliance Design of the protection systems includes consideration of natural phenomena, normal maintenance, testing, and accident conditions so that the protection functions are available.
Protection system components are designed, arranged, and qualified for operation in the environment accompanying any emergency situation in which the components are required to function.
Functional diversity has been designed into the system. The extent of this functional diversity is demonstrated for a variety of postulated accidents. Diverse protection functions automatically serve to mitigate the consequences of an event. Chapter 15 identifies the primary and diverse protective functions for each of the analyzed events.
Sufficient redundancy and independence are designed into the protection systems so that no single failure or removal from service of any component or channel of a system results in loss of the protection function. Functional diversity and location diversity are designed into the system.
Automatic reactor trip is initiated by neutron flux measurements, reactor coolant system overtemperature delta-T, reactor coolant system overpower delta-T, pressurizer pressure and level measurements, reactor coolant flow, reactor coolant pump speed, reactor coolant pump bearing water temperature, either of the PRHR heat exchanger discharge valves is not closed, ADS actuation, CMT actuation and steam generator water level measurements. Trips may also be initiated manually or by a safeguards actuation signal.
For additional information pertaining to the reactor trip logic, see Section 7.2.
High-quality components, conservative design and quality control, inspection, calibration, and tests are used to guard against common-mode failure. Qualification testing and analysis are performed on the safety-related systems to demonstrate functional operation at normal and post-accident conditions of temperature, humidity, pressure, and radiation for specified periods, as required.
Typical protection system equipment is subjected to type tests under simulated seismic conditions, using conservatively large accelerations and applicable frequencies.
See Section 7.1 for additional information concerning the equipment design of the protection and safety monitoring system.
See Sections 3.10 and 3.11 for information pertaining to environmental and seismic qualification of the protection system equipment.
The AP1000 includes a nonsafety-related diverse actuation system. The diverse actuation system provides specific automatic functions including control rod insertion, turbine trip, passive residual heat removal heat exchanger actuation, core makeup tank actuation, isolation of critical containment lines, and passive containment cooling system actuation. This system is diverse and independent from the reactor protection system from sensors up to the actuation devices.
See Section 7.7 for additional information concerning the diverse actuation system.
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VEGP 3&4 - UFSAR Criterion 23 - Protection System Failure Modes The protection system shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air) or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced.
AP1000 Compliance The protection system is designed considering the most probable failure modes of the components under various perturbations of the environment and energy sources. Reactor trip channels are designed on the deenergize-to-trip principle so that a single event (that is, loss of power) that could affect many functions at the same time causes the channels to actuate to their tripped conditions.
Criterion 24 - Separation of Protection and Control Systems The protection system shall be separated from the 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.
AP1000 Compliance The protection system is separate and distinct from the control systems. Control systems are, in some cases, dependent on the protection system for control signals that are derived from protection system measurements, where applicable. These signals are transferred to the control system by isolation devices classified as protection components.
The adequacy of the system isolation is verified by testing under conditions of postulated credible faults. Due to their inherent electrical isolation characteristics, fiber optic cables are exempt from electrical isolation qualification testing. The failure of a single control system component or channel, or the failure or removal from service of a single protection system component or channel common to the control and protection system, leaves intact a system that satisfies the requirements of the protection system. The removal of a protection division from service is allowed during testing of the division.
Criterion 25 - Protection System Requirements for Reactivity Control Malfunctions The protection system shall be designed to assure that specified 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 the control rods.
AP1000 Compliance The protection system is designed to limit reactivity transients so that the fuel design limits are not exceeded. Reactor shutdown by control rod insertion is independent of the normal control functions since the trip breakers interrupt power to the rod mechanisms regardless of existing control signals.
Thus, in the postulated accidental withdrawal of a control rod or control rod bank (assumed to be initiated by a control malfunction), neutron 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.
The AP1000 is designed to automatically terminate a boron dilution during manual or automatic operation at power, and also during startup and shutdown conditions. See Chapter 7 for a discussion of the signals used in the logic to terminate a boron dilution. Subsection 9.3.6.4.5 discusses the 3.1-12 Revision 5
VEGP 3&4 - UFSAR chemical and volume control system design features for addressing boron dilution. The Chapter 15 safety analyses demonstrate that fuel design limits are not exceeded.
Criterion 26 - Reactivity Control System Redundancy and Capability 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 reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure that the acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.
AP1000 Compliance Two reactivity control systems are provided. These are rod cluster control assemblies and gray rod assemblies, and chemical shim (boric acid). The rod cluster control and gray rod assemblies are inserted into the core by the force of gravity.
During operation, the shutdown rod banks are fully withdrawn. The control rod system automatically maintains a programmed average reactor temperature compensating for reactivity effects associated with scheduled and transient load changes. See Section 4.3 for additional information.
The shutdown and control rod banks are designed to provide reactivity margin to shut down the reactor during normal operating conditions and during anticipated operational occurrences, without exceeding specified fuel design limits. The safety analyses assume the most restrictive time in the core operating cycle and that the most reactive control rod cluster assembly is in the fully withdrawn position. See Chapter 15 for summaries of the analyses, assumptions, and results.
The safety-related passive systems provide the required boration to establish and maintain safe shutdown condition for the reactor core. See Section 6.3 for additional information.
Criterion 27 - Combined Reactivity Control Systems Capability 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.
AP1000 Compliance The plant is provided with the means of making and holding the core subcritical under any anticipated conditions and with appropriate margin for contingencies. Combined use of the control rod 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 rod assembly is assumed to be stuck in the fully withdrawn position for this determination.
Criterion 28 - Reactivity Limits 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 (l) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures, or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include 3.1-13 Revision 5
VEGP 3&4 - UFSAR 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.
AP1000 Compliance The maximum reactivity worth of the control rods and the maximum rates of reactivity increase employing control rods and boron removal are limited by design and operating procedures.
The appropriate reactivity addition rate for the withdrawal of control rods and the dilution rate of the boric acid in the reactor coolant system are specified in the precautions, limitations, and setpoint document and the control system setpoint study. Technical specifications explicitly specify control rod bank alignment and insertion limits in addition to shutdown margin reactivity requirements.
The control rod reactivity addition rate is determined by the allowable rod control system withdrawal speed, in conjunction with the control rod worth, which varies throughout the operating cycle. The capability to change boron concentration is determined by the various plant systems that provide makeup to the reactor coolant system. The reactivity insertion rates, rod withdrawal limits, and boron dilution limits are discussed in Chapter 4.
Core cooling capability following events such as rod ejection and steam line breaks is provided by keeping the reactor coolant pressure boundary stresses within faulted condition limits, as specified by applicable ASME codes. Structural deformations are also checked and limited to values that do not jeopardize the operation of needed safety-related features.
Criterion 29 - Protection Against Anticipated Operational Occurrences 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.
AP1000 Compliance The protection and reactivity control systems have an extremely high probability of performing their required safety-related functions in the event of anticipated operational occurrences. High-quality equipment, diversity, and redundancy, support this probability. Loss of power to the protection system results in a reactor trip. Defense in depth is designed into AP1000 to reduce challenges to the protection and reactivity control systems.
3.1.4 Fluid Systems Criterion 30 - Quality of Reactor Coolant Pressure Boundary 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.
AP1000 Compliance Reactor coolant pressure boundary components are designed, fabricated, inspected, and tested in conformance with the ASME Code,Section III. A portion of the chemical and volume control system that is defined as reactor coolant pressure boundary uses an alternate classification in conformance with the requirements of 10 CFR 50.55a(z). The alternate classification is discussed in Section 5.2.
Leakage detection monitoring is accomplished using instrumentation and other components of several systems. See Subsection 5.2.5 for additional information. Reactor coolant pressure boundary leakage is classified as either identified or unidentified leakage.
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VEGP 3&4 - UFSAR Auxiliary systems connected to the reactor coolant pressure boundary incorporate design and administrative provisions that limit leakage. Leakage is detected by increasing auxiliary system level, temperature, flow, or pressure, by lifting of relief valves, or by increasing values of monitored radiation in the auxiliary system.
Leakage from the reactor coolant pressure boundary and other components not otherwise identified inside the containment will condense and flow by gravity via the floor drains and other drains to the containment sump. Leakage is indicated by an increase in the sump level.
Reactor coolant system inventory monitoring provides an indication of system leakage. The reactor coolant system inventory balance is a quantitative inventory or mass balance calculation.
Leakage from the reactor coolant pressure boundary will result in an increase in the radioactivity levels inside containment. The containment atmosphere is monitored for airborne gaseous radioactivity and F18 particulate. From the concentration of F18 particulate and the power level, reactor coolant pressure boundary leakage can be estimated.
Criterion 31 - Fracture Prevention of Reactor Coolant Pressure Boundary The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (l) the boundary behaves 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 boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (l) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state, and transient stresses, and (4) size of flaws.
AP1000 Compliance Control is maintained over material selection and fabrication for the reactor coolant pressure boundary components so that the boundary behaves in a nonbrittle manner. The portion of the chemical and volume control system that uses an alternate classification is not required to meet the requirements to prevent brittle failure. The reactor coolant pressure boundary materials exposed to the coolant are corrosion-resistant stainless steel or nickel-chromium-iron alloy. The nil-ductility transition reference temperature of the reactor vessel structural steel is established by Charpy V-notch and drop weight tests in accordance with 10 CFR 50, Appendix G (Reference 1). See Section 5.3 for additional information.
The following requirements are imposed in addition to those specified by the ASME Code,Section III.
A 100 percent volumetric ultrasonic shear wave test of reactor vessel plate and a post-hydrotest ultrasonic map of welds in the pressure vessel are required. Cladding bond ultrasonic inspection to more restrictive requirements than those specified in the ASME Code,Section III is also required in order to preclude interpretation problems during in-service inspection.
In the surveillance programs, the evaluation of the radiation damage is based on pre-irradiation testing of Charpy V-notch and tensile specimens and post-irradiation testing of Charpy V-notch, tensile, and l/2T compact tension specimens. These programs are directed toward evaluation of the effect of radiation on the fracture toughness of reactor vessel steels based on the reference transition temperature approach and the fracture mechanics approach, and are in accordance with ASTM, E-185 (Reference 2).
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VEGP 3&4 - UFSAR Reactor vessel core region material chemistry (copper, phosphorous, and vanadium) is controlled to reduce sensitivity 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 governed by ASME Code,Section III requirements.
Allowable pressure-temperature relationships for plant heatup and cooldown rates are calculated using methods derived from the ASME Code,Section III, Appendix G. The approach specifies that the allowable stress intensity factors for vessel-operating conditions do not exceed the reference stress intensity factor for the metal temperature. Operating specifications include conservative margins for predicted changes in the material reference temperatures due to irradiation.
Criterion 32 - Inspection of Reactor Coolant Pressure Boundary Components which are part of the reactor coolant pressure boundary shall be designed to permit (l) 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.
AP1000 Compliance The design of the reactor coolant pressure boundary provides accessibility to the internal surfaces of the reactor vessel and most external zones of the vessel, including the nozzle-to-reactor coolant piping welds, the top and bottom heads, and external surfaces of the reactor coolant piping, except for the area of pipe within the primary shield concrete. The inspection capability complements the leakage detection systems in assessing the integrity of the pressure boundary components. The reactor coolant pressure boundary will be periodically inspected under the provisions of the ASME Code,Section XI. Section 5.1 provides the reactor coolant system primary loop drawings. The portion of the chemical and volume control system that uses an alternate classification is constructed to requirements that do not require in-service inspection.
Monitoring of changes in the fracture toughness properties of the reactor vessel core region plates, forgings, weldments, and associated heat-treated zones is performed according to 10 CFR 50, Appendix H. Additionally, 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 conventional tensile and impact tests and fracture mechanics specimens. The observed shifts in the nil-ductility transition reference temperature of the core region materials with irradiation is used to confirm the allowable limits calculated for operational transients.
The design of the reactor coolant pressure boundary piping provides for accessibility of welds requiring in-service inspection under the provisions of the ASME Code,Section XI. Removable insulation is provided at welds requiring in-service inspection. See Section 5.3 and Subsection 5.2.4 for additional information.
Criterion 33 - Reactor Coolant Makeup 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 3.1-16 Revision 5
VEGP 3&4 - UFSAR using the piping, pumps, and valves used to maintain coolant inventory during normal reactor operation.
AP1000 Compliance Changes in the reactor coolant volume will be accommodated by the pressurizer level program for normal power changes, including the transition from hot standby to full-power operation and returning to hot standby. In addition, the pressurizer has sufficient volume to accommodate minor reactor coolant system leakage.
Safety-related passive reactor coolant system makeup is provided to accommodate small leaks when the normal makeup system is unavailable and to accommodate larger leaks resulting from loss of coolant accidents. Safety-related reactor coolant makeup and safety injection are provided by two core makeup tanks, two accumulators, and an in-containment refueling water storage tank. Long-term cooling is provided by containment gravity recirculation of reactor coolant within containment.
See Section 6.3 for additional information. The safety-related reactor coolant makeup relies on the Class 1E and UPS system. Neither onsite or offsite ac power is required.
In addition, the nonsafety-related chemical and volume control system automatically provides inventory control to accommodate minor leakage from the reactor coolant system, expansion during heatup from cold shutdown, and contraction during cooldown. This inventory control is provided by letdown and makeup connections to the chemical and volume control system purification loop.
Redundant pumps with connections to redundant nonsafety-related onsite ac power are provided when offsite power is not available and these pumps can be supplied from offsite power when onsite power is not available. See Section 5.2 for additional information.
Criterion 34 - Residual Heat Removal 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 reactor 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.
AP1000 Compliance The AP1000 design satisfies the intent of GDC 34 by reducing the risk associated with loss of the decay heat removal function through a combination of safety-related passive systems, together with nonsafety-related active systems. Specific decay heat removal systems include the following:
A safety-related passive residual heat removal heat exchanger that uses natural circulation flow and that does not require electrical power for operation Automatic, safety-related feed and bleed using the core makeup tanks, accumulators, and the in-containment refueling water storage tank for injection and the automatic depressurization system valves for reactor coolant system venting The nonsafety-related main feedwater system with motor-driven pumps supplied by the main generator or by offsite power 3.1-17 Revision 5
VEGP 3&4 - UFSAR The nonsafety-related startup feedwater system with motor-driven pumps supplied by offsite or onsite power, including automatic sequencing on the nonsafety-related diesel generators The nonsafety-related normal residual heat removal system with motor-driven pumps supplied by offsite or onsite power, including nonsafety-related diesel generators, for use at low reactor coolant system pressures A safety-related emergency feedwater system is not required for the AP1000 design. An active safety-related residual heat removal system is not required for the AP1000.
The AP1000 passive core cooling system, in conjunction with the passive containment cooling system, provides a reliable capability for removing decay heat from the reactor core and maintains sufficient water inventory to provide adequate core cooling for an extended period of time. The system does not depend upon pumped injection or recirculation, and actuates automatically, requiring no operator actions.
The containment arrangement addresses the Regulatory Guide 1.82 issues. Functional performance of the system addresses the guidelines of Regulatory Guide 1.139, except that cooldown rate is somewhat more limited when using the passive residual heat removal equipment. See Subsection 1.9.1 for additional information.
The passive core cooling system provides both gravity injection and gravity recirculation, automatically shifting injection modes when the proper containment flood-up conditions are achieved.
The AP1000 design provides a passive decay heat removal system that functions independent of nonsafety-related ac power supplies and can accommodate single active failures. (The Class 1E dc and UPS system supplies power to the safety-related monitoring and control instrumentation.) The passive core cooling system complies with General Design Criterion 34 by providing the capability to remove decay heat without relying on nonsafety-related ac power.
Criterion 35 - Emergency Core Cooling 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 (l) 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.
AP1000 Compliance The AP1000 design provides for safety-related passive reactor coolant makeup. Core makeup tanks accommodate small leaks when the normal makeup system is unavailable and provide safety injection for small-break loss of coolant accidents. Accumulators provide the high makeup flow required for a large loss of coolant accident and initiate injection when the reactor coolant system pressure is below the static accumulator pressure during a small-break loss of coolant accident.
The in-containment refueling water storage tank, and after containment flood-up, containment recirculation capability provide the long-term source of gravity injection to the core after the reactor coolant system is depressurized. The automatic depressurization system valves provide the vent path to transfer the core decay heat to the containment and then to the ultimate heat sink.
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VEGP 3&4 - UFSAR The AP1000 design provides a passive core cooling system that functions independent of ac power supplies, assuming single active failures. The passive core cooling system does not need the nonsafety-related diesel-generators for electrical power to either actuate or operate the various system components. Therefore, the passive core cooling system complies with the intent of GDC 35 by providing the capability for core cooling without relying on nonsafety-related ac power sources.
Criterion 36 - Inspection of Emergency Core Cooling System 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.
AP1000 Compliance The AP1000 design includes a passive core cooling system that provides emergency core decay heat removal, emergency reactor coolant system makeup and boration, safety injection, and containment sump pH control. The system piping and components are designed to permit access for periodic inspection and testing of equipment, according to the ASME Code and technical specification requirements, to provide confidence in the integrity and capability of the system.
The core makeup tanks, accumulators, and passive residual heat removal heat exchanger have manways which permit access for inspection and required maintenance. The in-containment refueling water storage tank design provides access for both the tank itself and for the passive residual heat removal heat exchanger, spargers, and other components located inside the tank.
In addition, the system piping provides accessibility for inspection and maintenance to the extent practical. See Section 6.3 for additional information.
Criterion 37 - Testing of Emergency Core Cooling System The emergency core cooling system shall be designed to permit appropriate periodic pressure and functional testing to assure (l) the structural and leak-tight 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.
AP1000 Compliance The AP1000 passive core cooling system is designed to permit the periodic inspection and testing of the appropriate system components. The testing capabilities of the system including in-service testing and inspection to confirm the structural and leaktight integrity of various components, technical specification operability and performance of the active system components, and additional in-service testing to confirm the overall operability of the system.
The stage 1, 2, and 3 automatic depressurization system valves have provisions for shutdown in-service testing and at-power operability testing.
Planned shutdown testing includes operability testing of the component and system performance, including operation of applicable portions of the protection and safety monitoring system and the use of the appropriate power sources for the system.
The AP1000 design has significantly reduced the support systems required for system operation. In-service testing of the required support systems is also planned.
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VEGP 3&4 - UFSAR Criterion 38 - Containment Heat Removal System 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 temperature following any loss of coolant accident and maintain them at acceptably low levels.
Suitable redundancy in components and features and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.
AP1000 Compliance The AP1000 design uses passive systems for post-loss of coolant accident core and containment heat removal and for the prevention of overpressurization failure of the containment building. Heat is transferred from the containment atmosphere to the steel containment shell by natural convection and condensation. Heat removal from the exterior of the containment shell is enhanced by a directed-flow natural convection design and a passive, external cooling water distribution system.
The AP1000 passive containment cooling system is designed with sufficient capacity to prevent the containment from exceeding its design pressure with no operator action or outside assistance for a minimum of 3 days. After 3 days, limited operator action is required.
The AP1000 passive containment cooling system consists of a steel containment shell and associated water supplies, piping, valves, and air baffle. The passive containment cooling system is a passive system that uses gravity and natural circulation as driving forces. The design of the AP1000 passive containment cooling system does not require the use of any pumps, and it functions independent of nonsafety-related ac power sources for 3 days. Therefore, the passive containment cooling system can function during loss of offsite or onsite power. GDC 38 is satisfied by using appropriate redundancy and by the design of the passive containment cooling system and its reliance on natural forces.
Criterion 39 - Inspection of Containment Heat Removal System The containment heat removal system shall be designed to permit appropriate periodic inspection of important components, such as the torus, sumps, spray nozzles and piping, to assure the integrity and capability of the system.
AP1000 Compliance The AP1000 design uses safety-related passive means for containment heat removal. The design of the system allows for inspection of piping, valves, the containment shell and air baffle, and other components to provide confidence in the integrity and capability of the system.
The periodic inspections specified in the ASME Code and technical specifications provide confidence that the capability of these heat removal systems is retained through plant life.
Criterion 40 - Testing of Containment Heat Removal System The containment heat removal system shall be designed to permit appropriate periodic pressure and functional testing to assure (l) 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 the full operational sequence that brings the system into operation, including operation of applicable 3.1-20 Revision 5
VEGP 3&4 - UFSAR portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.
AP1000 Compliance The AP1000 design includes a passive containment cooling system that provides containment heat removal to limit the peak containment pressure following design basis events. The system piping and components are designed to permit access for periodic inspection and testing of equipment, according to the ASME Code and technical specification requirements, to provide confidence in the integrity and capability of the system.
The passive containment cooling water storage tank design allows access for both the tank and for the various components located inside the tank.
In addition, the system piping provides accessibility for inspection and maintenance to the extent practical. See Section 6.2 for additional information.
Criterion 41 - Containment Atmosphere Cleanup Systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be provided, as necessary, to reduce, consistent with the functioning of other associated systems, the concentration and quantity of fission products released to the environment following postulated accidents and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained.
Each system shall have suitable redundancy in components and features and suitable interconnections, leak detection, isolation, and containment capabilities to assure that for onsite electric power system operation (assuming offsite 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.
AP1000 Compliance Fission product control for the AP1000 plant is provided via natural removal processes within containment and by limiting containment leakage. The passive removal processes such as deposition and sedimentation are evaluated based on a physically-based source term with large scale core damage. See Section 6.5 for additional details. The containment and penetration design includes features specifically designed to minimize overall containment leakage. See Subsection 6.2.3 for additional details.
The generation of hydrogen in the containment under post-accident conditions has been evaluated, and the containment hydrogen control system has been designed such that the following criteria are satisfied:
In compliance with Section 50.44 of 10 CFR 50, means are provided to measure and control post-loss of coolant accident hydrogen concentrations.
The combustible concentrations of hydrogen do not accumulate in the areas where unintended combustion or detonation could cause loss of containment integrity or loss of appropriate mitigating features.
Internal passive autocatalytic recombiners are provided for hydrogen control following a design basis loss of coolant accident.
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VEGP 3&4 - UFSAR Hydrogen igniters are provided to limit local and global hydrogen concentrations to below 10 percent following a degraded core event with the reaction of 100 percent of the zircaloy cladding.
The concentration of uniformly distributed hydrogen produced by the equivalent of a 75 percent active fuel-clad metal water reaction does not exceed 13 percent by volume during and following a degraded core event. (The AP1000 containment volume is large enough to provide passive protection for the hydrogen produced by 75 percent zircaloy cladding reaction following a severe accident.)
The nonsafety-related ventilation system, normally used during refueling, is designed with the capability for a controlled purge of the containment atmosphere to assist in post-accident cleanup, but is not required for hydrogen control.
Criterion 42 - Inspection of Containment Atmosphere Cleanup System 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.
AP1000 Compliance The containment atmosphere cleanup systems are designed and located so that they can be inspected periodically, as appropriate.
Criterion 43 - Testing of Containment Atmosphere Cleanup Systems The containment atmosphere cleanup systems shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leak-tight integrity of its components, (2) the operability and performance of the active components 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.
AP1000 Compliance The appropriate portions of the containment atmosphere cleanup system are designed to permit periodic pressure and functionality testing.
As described in GDC 41, the containment atmosphere cleanup system has no safety-related post-accident cleanup functions. Dose mitigation is passively provided by the containment isolation and integrity, natural removal processes, and limited containment leakage. Periodic containment integrity is verified in accordance with 10 CFR 50 Appendix J testing as described in Subsection 6.2.3.
Criterion 44 - Cooling Water 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 under 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.
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VEGP 3&4 - UFSAR AP1000 Compliance The passive containment cooling system is the ultimate heat sink for the AP1000 and does not rely upon offsite or onsite ac power sources. Heat transfer by convection from the containment shell to the atmosphere meets the intent of GDC 44. Additional information is provided in the responses for GDC 34 and GDC 38.
Criterion 45 - Inspection of Cooling Water System 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.
AP1000 Compliance Refer to the discussion provided for GDC 39.
Criterion 46 - Testing of Cooling Water System The cooling water system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leak-tight integrity of its components, (2) the operability and the 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 for reactor shutdown and for loss of coolant accidents, including operation of applicable portions of the protection system and the transfer between normal and emergency power sources.
AP1000 Compliance Refer to the discussion provided for GDC 40.
3.1.5 Reactor Containment Criterion 50 - Containment Design Basis The reactor containment structure, including access opening, penetrations, and the containment heat removal system, shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any 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 energy from metal-water and other chemical reactions that may result from degraded 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.
AP1000 Compliance The design of the containment structure is based on the containment design basis accidents, which include the rupture of a reactor coolant pipe or the rupture of a main steam or feedwater line. The maximum pressure and temperature reached, a description of the calculational model, and input parameters for a containment design basis accident are presented in Section 6.2. The containment design provides margin to the design basis limits.
Criterion 51 - Fracture Prevention of Containment Pressure Boundary 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 3.1-23 Revision 5
VEGP 3&4 - UFSAR 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 stresses, and (3) size of flaws.
AP1000 Compliance Principal load-carrying components of ferritic materials of the reactor containment boundary exposed to the external environment are selected so that they behave in a nonbrittle manner and so that the probability of fracture propagation is minimized. See Subsection 3.8.2 for additional information.
Criterion 52 - Capability for Containment Leakage Rate Testing 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.
AP1000 Compliance The containment system is designed and constructed and the necessary equipment is provided to permit periodic integrated leakage rate tests according to the requirements of 10 CFR 50, Appendix J.
Criterion 53 - Provisions for Containment Testing and Inspection 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 leak-tightness of penetrations which have resilient seals and expansion bellows.
AP1000 Compliance Provisions exist for conducting individual leakage rate tests on containment penetrations.
Penetrations are visually inspected and pressure-tested for leak tightness at periodic intervals. Other inspections are performed as required by 10 CFR 50, Appendix J.
Criterion 54 - Piping Systems Penetrating Containment Piping systems penetrating the 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.
AP1000 Compliance Piping systems penetrating the primary reactor containment are provided with containment isolation valves. Penetrations that close for containment isolation have redundant valving. Automatic isolation valves with air-, solenoid-, or motor-operators, which do not restrict normal plant operation, are periodically tested to verify operability.
The AP1000 containment isolation design satisfies the current NRC requirements including the post-TMI requirements, as discussed in Subsection 1.9.3. In general, this means that two barriers are provided, one inside containment and the other outside containment. Usually these barriers are valves, but in some cases they are closed piping systems not connected to the reactor coolant system or to the containment atmosphere.
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VEGP 3&4 - UFSAR The AP1000 design incorporates a reduction in the number of existing penetrations. Most penetrations are normally closed. Those few that are normally open and are required to close use remotely operated valves for isolation that close automatically. See Subsection 6.2.3 for additional information.
Nonessential systems that may be normally open, such as the mini-purge system, are provided with automatic containment isolation valves that close automatically on a containment isolation signal.
The containment isolation signal is actuated by the protection and safety monitoring system. See Section 7.3 for additional information.
Piping and electrical containment penetrations are equipped with test connections and test vents or have other provisions to allow periodic leak rate testing so that leakage is within the acceptable limits established in technical specifications consistent with 10 CFR 50, Appendix J.
Criterion 55 - Reactor Coolant Pressure Boundary Penetrating Containment Each line that is part of the reactor coolant pressure boundary and that penetrates primary reactor containment 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; or
- 2. One automatic isolation valve inside and one locked closed isolation valve outside containment; or
- 3. One locked closed isolation valve inside and one automatic isolation valve outside the containment. A simple check valve may not be used as the automatic isolation valve outside containment; or
- 4. One automatic isolation valve inside and one automatic isolation valve outside containment.
A simple check valve may not be used as the automatic isolation valve outside containment.
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.
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 for in-service inspection, protection against more severe natural phenomena, and additional isolation valves and containment, shall include consideration of the population density, and use characteristics, and physical characteristics of the site environs.
AP1000 Compliance Lines that penetrate containment that are connected to the reactor coolant pressure boundary are provided with containment isolation valves in accordance with one of the acceptable arrangements as described in GDC 55. Additional information is found in Subsection 6.2.3.
Criterion 56 - Primary Containment Isolation Each line that connects directly to the containment atmosphere and penetrates the primary reactor containment shall be provided with containment isolation valves as follows, unless it can be 3.1-25 Revision 5
VEGP 3&4 - UFSAR 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 the containment; or
- 2. One automatic isolation valve inside and one locked closed isolation valve outside the containment; or
- 3. One locked closed isolation valve inside and one automatic isolation valve outside the containment. A simple check valve may not be used as the automatic isolation valve outside containment; or
- 4. One automatic isolation valve inside and one automatic isolation valve outside the containment. A simple check valve may not be used as the automatic isolation valve outside the containment.
Isolation valves outside the 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.
AP1000 Compliance Lines connecting directly with the containment atmosphere and penetrating the reactor containment are normally provided with two isolation valves in series, one inside and one outside the containment, in accordance with one of the acceptable arrangements as described in GDC 56. Isolation of instrument lines for containment pressure measurement is demonstrated on a different basis and does not require isolation valves. Additional information is found in Subsection 6.2.3.
Criterion 57 - Closed System Isolation Valves Each line that penetrates the 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, locked closed, or capable of remote manual operation. This valve shall be outside the containment and located as close to the containment as practical. A simple check valve may not be used as the automatic isolation valve.
AP1000 Compliance Lines that penetrate the containment and are neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere are considered closed systems within the containment and are equipped with at least one containment isolation valve of one of the following types:
An automatic isolation valve (a simple check valve is not used as this automatic valve)
A locked-closed valve This valve is located outside the containment and as close to the containment wall as practical.
3.1.6 Fuel and Reactivity Control Criterion 60 - Control of Releases of Radioactive Materials to the Environment 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 3.1-26 Revision 5
VEGP 3&4 - UFSAR normal reactor operation, including anticipated operational occurrences. Sufficient holdup capacity shall be provided for the 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.
AP1000 Compliance Means are provided to control the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation, including anticipated operational occurrences.
The radioactive waste management systems are designed to minimize the potential for an inadvertent release of radioactivity from the facility and to provide confidence that the discharge of radioactive wastes is maintained below regulatory limits of 10 CFR 50, Appendix I, during normal operation. The gaseous radwaste and liquid radwaste processing systems include continuous radiation monitoring of their discharge paths. High radiation automatically closes a discharge isolation valve. The liquid radwaste system also has provisions to prevent inadvertent siphoning of its monitor tank contents which could cause an uncontrolled discharge. The radioactive waste management systems, the design bases, and the estimated amounts of radioactive effluent releases to the environment are described in Chapter 11.
Criterion 61 - Fuel Storage and Handling and Radioactivity Control The fuel storage and handling, radioactive waste, and other 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 capability 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 inventory under accident conditions.
AP1000 Compliance The spent fuel pool cooling system, and the fuel handling and refueling system are designed to provide cooling and shielding for the fuel assemblies stored in the spent fuel pool and to provide purification of the water in the pit. The system design provides adequate safety under normal and postulated accident conditions.
The spent fuel pool cooling system normal system operation is described in Subsection 9.1.3.
Sampling of the spent fuel pool water for gross activity, tritium, and particulate matter is conducted periodically. The concentration of tritium in the spent fuel pool water is maintained at less than 0.5 microcuries per gram to provide confidence that the airborne concentration of tritium in the fuel handling area is within the limits specified in 10 CFR 20, Appendix B. See Subsection 12.2.2 for additional information.
The spent fuel pool is designed so that a water level is maintained above the spent fuel assemblies for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a loss of the spent fuel pool cooling system, without ac power. See Subsection 9.1.2 for additional information.
The spent fuel pool cooling system maintains the water in the in-containment refueling water storage tank consistent with activity requirements of the water in the refueling cavity during a refueling. Two spent fuel pool cooling filters are provided, one downstream of each demineralizer in the purification branch line of each mechanical train. The filters are sized to collect particulates and suspended solids passed by the demineralizer.
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VEGP 3&4 - UFSAR The AP1000 spent fuel pool cooling system is not required to operate to mitigate design basis events. In the event the spent fuel pool cooling system is unavailable, the spent fuel pool cooling is provided by the heat capacity of the water in the pool and in the passive sources of makeup water.
Normal HVAC to the spent fuel pool area is provided by a subsystem of the radiologically controlled area ventilation system described in Subsection 9.4.3. No credit is taken for this system in evaluation of fuel handling accidents discussed in Subsection 15.7.4.
Connections to the spent fuel pool are provided at an elevation that prevents inadvertent draining of the water in the pool to an unacceptable level.
The design of spent fuel storage pool and the spent fuel pool cooling system satisfies GDC 61. See Subsection 9.1.3 for additional information.
Criterion 62 - Prevention of Criticality in Fuel Storage and Handling Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.
AP1000 Compliance The restraints, interlocks, and physical arrangement provided for the safe handling and storage of new and spent fuel are discussed in Section 9.1. The spent fuel assemblies are stored in the spent fuel pool until fission product activity is low enough to permit shipment.
Criterion 63 - Monitoring Fuel and Waste Storage Appropriate systems shall be provided in the fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in the loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions.
AP1000 Compliance Instrumentation is provided to monitor spent fuel storage pool temperature and water level. Indication and alarms are provided in the main control room. Area radiation monitoring is provided in the fuel storage area for personnel protection and general surveillance. The area monitor alarms locally and in the main control room.
If radiation levels in the ventilation effluent reach a predetermined point, an alarm is actuated in the main control room, and the ventilation discharge path is automatically transferred through filter absorber units that provide filtration before discharge from the plant vent.
Criterion 64 - Monitoring Radioactivity Releases Means shall be provided for monitoring 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.
AP1000 Compliance The containment atmosphere is monitored during normal and transient operations by the containment gaseous radiation monitors. Under accident conditions, samples of the containment atmosphere taken via the sampling system provide data on airborne radioactive concentrations within the containment.
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VEGP 3&4 - UFSAR No reactor coolant fluids are required to be recirculated outside of containment following an accident.
Radioactivity levels contained in the facility effluent and discharge paths and in the plant environs are monitored during normal and accident conditions by the plant radiation monitoring systems. High radiation in a discharge path causes automatic closure of the discharge isolation valve.
Area radiation monitors (ARMs) are provided to supplement the personnel and area radiation survey provisions of the AP1000 health physics program described in Section 12.5 and to comply with the personnel radiation protection guidelines of 10 CFR 20, 10 CFR 50, and Regulatory Guides 1.97, 8.2, 8.8, and 8.12. In addition to the installed detectors, periodic plant environmental surveillance is established.
Measurement capability and reporting of effluents are based on the guidelines of Regulatory Guides 1.4 and 1.21, as discussed in Subsection 1.9.1. Additional information is contained in Chapters 11 and 12.
3.1.7 Combined License Information This section contained no requirement for additional information.
3.1.8 References
- 1. 10 CFR 50, Appendix G, "Fracture Toughness Requirements."
- 2. American Society of Testing Materials E-185, Standard Recommended Practice for Surveillance Test for Nuclear Reactor Vessels, and the requirements for 10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements."
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VEGP 3&4 - UFSAR 3.2 Classification of Structures, Components, and Systems Structures, systems, and components in the AP1000 are classified according to nuclear safety classification, quality groups, seismic category, and codes and standards. This section provides the methodology used for safety-related and seismic classification of AP1000 structures, systems, and components. The seismic classification is described in Subsection 3.2.1. Subsection 3.2.2 describes the classification including nuclear safety-related classification and the corresponding codes and standards. Additionally, Subsection 3.2.2 describes nonsafety-related equipment classifications.
3.2.1 Seismic Classification General Design Criterion 2 requires that nuclear power plant Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena, such as earthquakes, tornados, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. 10 CFR 100, Appendix A sets forth the criteria to which the plant design bases demonstrate the capability to function during and after vibratory ground motion associated with the safe shutdown earthquake conditions.
The seismic classification methodology used in AP1000 complies with the preceding criteria, as well as with recommendations stated within Regulatory Guide 1.29. Conformance with the recommendations of Regulatory Guide 1.29 is discussed in Subsection 1.9.1. The methodology classifies structures, systems, and components into three categories: seismic Category I (C-I),
seismic Category II (C-II) and non-seismic (NS).
Seismic Category I applies to both functionality and integrity, and seismic Category II applies only to integrity. Non-seismic items located in the proximity of safety-related items, the failure of which during a safe shutdown earthquake could result in loss of function of safety-related items, are designated as seismic Category II.
There are no safety-related structures, systems, or components outside the scope of the DCD, except for engineered fill which is classified as a Seismic Category I, safety-related structure. See Table 3.2-2. Refer to Subsection 2.5.4 for a discussion of safety-related backfill.
The nonsafety-related structures, systems, and components outside the scope of the DCD are classified as non-seismic (NS).
3.2.1.1 Definitions 3.2.1.1.1 Seismic Category I (C-I)
Seismic Category I applies to, in general, safety-related structures, systems, and components.
Seismic Category I also applies to those structures, systems, and components required to support or protect safety-related structures, systems, and components. The exceptions to this general rule are a limited number of structures, such as those required for tornado missile protection, which do not have a safety-related function to perform during or following a seismic event. (See Subsection 3.2.2.5.)
Safety-related items are those necessary to provide for the following:
The integrity of the reactor coolant pressure boundary The capability to shut down the reactor and maintain it in a safe shutdown condition Capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of 10 CFR 50.34.
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VEGP 3&4 - UFSAR Seismic Category I structures, systems, and components are designed to withstand the appropriate seismic loads, as discussed in Section 3.7, and other applicable loads without loss of function.
Seismic Category I structures are protected from interaction with adjacent non-seismic structures as described in Subsection 3.7.2.8.
Systems and components identified as safety-related systems and components in Table 3.2-3, and electrical and instrumentation components identified in Table 3.11-1, are the systems and components necessary for continued operation that must remain functional without undue risk of the health and safety of the public during and following an operating basis earthquake. Systems and components identified as Equipment Class A, B, and C in Table 3.2-3, and electrical and instrumentation components identified in Table 3.11-1, are the systems and components that per the criteria of 10 CFR Part 50, Appendix S, must be demonstrated, prior to resuming operations, to have no functional damage following a seismic ground motion exceeding the operating basis earthquake ground motion. See Section 3.7 for information on the operating basis earthquake.
Seismic Category I structures, systems, and components meet the quality assurance requirements of 10 CFR 50, Appendix B. The criteria used for the design of seismic Category I structures, systems, and components are discussed in Section 3.7.
3.2.1.1.2 Seismic Category II (C-II)
Seismic Category II applies to plant structures, systems, and components which perform no safety-related function, and the continued function of which is not required. Seismic Category II applies to structures, systems, and components designed to prevent their collapse under the safe shutdown earthquake. Structures, systems and components are classified as seismic Category II to preclude their structural failure during a safe shutdown earthquake or interaction with seismic Category I items which could degrade the functioning of a safety-related structure, system, or component to an unacceptable level, or could result in incapacitating injury to occupants of the main control room. The turbine building first bay building structure, including Wall 11.2, is a seismic Category II structure as identified in Table 3.2-2. The seismic Category II systems, structures, and components which provide tornado missile protection for openings and penetrations are described in Subsection 3.5.2. Wall 11.2 protects Wall 11 from the dynamic effects of pipe failure events in the nonseismic portion of the turbine building.
Seismic Category II structures, systems, and components are designed so that the safe shutdown earthquake does not cause unacceptable structural failure of or interaction with seismic Category I items. Seismic Category II fluid systems require an appropriate level of pressure boundary integrity if located near sensitive equipment.
The criteria used for the design of seismic Category II structures, systems, and components are discussed in Section 3.7. As identified in Subsection 3.7.2, seismic Category II building structures are designed for the safe shutdown earthquake using the same methods and design stress limits as are used for seismic Category I structures. Seismic Category II building structures are also designed to withstand the design basis tornado loads, including missiles, in accordance with the loading combinations identified in Table 3.8.4-2.
Pertinent portions of 10 CFR 50, Appendix B apply to the analysis and design of seismic Category II structures, systems, and components. The quality assurance requirements for the analysis and design of seismic Category II structures, systems, and components are performed in accordance with the Westinghouse AP1000 quality plan as described in Section 17.3 and are sufficient to provide that these components will meet the requirement to not cause unacceptable structural failure of or interaction with seismic Category I items. See Section 17.7 for the Combined License applicant quality assurance program requirement. These quality requirements are applicable to the seismic Category II turbine building first bay building structure.
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VEGP 3&4 - UFSAR 3.2.1.1.3 Non-Seismic Non-seismic (NS) structures, systems, and components are those that are not classified seismic Category I or Category II.
The criteria used for the design of non-seismic structures, components and systems are discussed in Section 3.7.
The non-seismic lines and associated equipment are routed, to the extent practicable, outside of safety-related buildings and rooms to avoid adverse system interactions. In cases where these lines are routed in safety-related areas, the non-seismic item is evaluated for the safe shutdown earthquake and is upgraded to seismic Category II if a credible failure could cause an unacceptable interaction.
Although the seismic category for an item located in the proximity of safety-related structures, systems, and components may be upgraded to seismic Category II, its pre-assigned equipment class remains unchanged.
3.2.1.2 Classifications Table 3.2-1 illustrates the general relationship between safety-related equipment classes and seismic categories. In most cases, except as noted in Subsection 3.2.2.5, safety-related items are also seismic Category I items. When portions of systems are identified as seismic Category I, the boundaries of seismic Category I portions of the system are shown on the piping and instrumentation diagram (P&ID) of that system. See Subsection 1.7.2 for a list of the piping and instrumentation diagrams.
3.2.1.3 Classification of Building Structures Building structures are assigned a seismic category as indicated in Table 3.2-2. Codes and standards used in the design and construction of building structures are given in Section 3.8. The building structures are not assigned a safety classification in Subsection 3.2.2 with the exception of the containment vessel.
3.2.2 AP1000 Classification System The assignment of safety-related classification and use of codes and standards conforms to the requirements of 10 CFR 50.55a for the development of a Quality Group classification and the use of codes and standards. The description of the equipment classification which follows identifies the classifications requiring the full 10 CFR 50, Appendix B quality assurance program as described in Chapter 17 and the Quality Group associated with each classification.
The classification system provides a means of identifying the extent to which structures, systems, and components are related to safety-related and seismic requirements. The classification system provides an easily recognizable means of identifying the extent to which structures, systems, and components are related to ANS nuclear safety classification, NRC quality groups, ASME Code,Section III classification, seismic category and other applicable industry standards, as shown in Table 3.2-3.
There are no safety-related structures, systems, or components outside the scope of the DCD, except for engineered fill which is classified as a Seismic Category I, safety-related structure. See Table 3.2-2. Refer to Subsection 2.5.4 for a discussion of safety-related backfill.
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VEGP 3&4 - UFSAR 3.2.2.1 Classification Definitions The definitions used in the classification of structures, systems and components are provided in the following. Unless otherwise noted these definitions apply throughout the Design Control Document.
These definitions are consistent with the ANS Definitions for Light Water Reactor Standards (ANS-58.14-1993).
Safety-related is a classification applied to items relied upon to remain functional during or following a design basis event to provide a safety-related function. Safety-related also applies to documentation and services affecting a safety-related item.
Safety-related function is a function that is relied upon during or following a design basis event to provide for the following:
The integrity of the reactor coolant pressure boundary The capability to shut down the reactor and maintain it in a safe shutdown condition The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of 10 CFR 50.34.
Design basis event is an event that is a condition of normal operation (including anticipated operational occurrences), a design basis accident, an external event, or natural phenomena for which the plant must be designed so that the safety-related functions are achievable.
Design basis accidents and transients are those design basis events that are accidents and transients and are postulated in the safety analyses. The design basis accidents and transients are used in the design of the plant to establish acceptable performance requirements for structures, systems, and components.
3.2.2.2 Application of Classification The AP1000 requires adaptation of safety classification documents and standards because of the way that the AP1000 accomplishes safety-related functions.
In addition to 10 CFR 50.55a, the AP1000 classification has been developed considering requirements and guidelines in the following:
ANSI N18.2a (Reference 1) - safety classification ANS 51.1 (Reference 2) - safety classification Regulatory Guide 1.26 - Quality Groups Regulatory Guide 1.97 - instrumentation requirements 10 CFR 21.
Conformance with the guidelines of Regulatory Guides 1.26 and 1.97 is discussed in Subsection 1.9.1.
The general guidelines for safety classification in the ANSI and ANS standards are useful in the development of the AP1000 classification. The specific classifications for various structures, systems, and components included in Regulatory Guide 1.26 and ANSI N18.2a and ANS 51.1 are based on a nuclear power plant with active safety systems and are not necessarily appropriate for the passive safety systems of the AP1000.
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VEGP 3&4 - UFSAR For the purposes of equipment classification, structures, systems, and components are classified as Class A, B, C, D, E, F, G, L, P, R, or W. For mechanical equipment Classes A, B, and C are equivalent to ANS Safety Class 1, 2, and 3. For electrical equipment Class C is equivalent to Class 1E.
Structures, systems, and components classified equipment class A, B, or C or seismic Category I are basic components as defined in 10 CFR Part 21.
Equipment Class D is a nonsafety-related class. Classes E, F, G, L, P, R, and W are nonsafety-related classes associated with different industry codes and standards.
Components are classified down to the replacement part level according to the definitions and criteria of the classification system. A single item or portion thereof, which provides two or more functions of different classes, is classified according to the most stringent function. Different portions of the same structure, system, or component may perform different functions and be assigned to different equipment classes if the structure, system, or component contains a suitable interface boundary.
The definitions and criteria for the AP1000 equipment classes follow.
3.2.2.3 Equipment Class A Class A is a safety-related class equivalent to ANS Safety Class 1. It applies to the reactor coolant system pressure boundary, including the required isolation valves and mechanical supports. This class has the highest integrity, and the lowest probability of leakage.
10 CFR 21 applies to Class A structures, systems, and components. Class A structures, systems, and components are seismic Category I and use codes and standards consistent with the guidelines for NRC Quality Group A. 10 CFR 50, Appendix B applies. ASME Code,Section III, Class 1 applies to pressure retaining components.
3.2.2.4 Equipment Class B Class B is a safety-related class equivalent to ANS Safety Class 2. It limits the leakage of radioactive material from the containment following a design basis accident. This class is designed to accomplish the following:
It provides fission product barrier or primary containment radioactive material holdup or isolation.
It provides the containment boundary including penetrations and isolation valves. This also includes piping that functions as the containment boundary. For example, the steam and feedwater system inside containment and the secondary shell of the steam generator are Class B by this criterion.
It circulates a non-containment/non-reactor coolant fluid to provide a post-accident safety-related function into and out of the containment. These lines have a Class B pressure boundary inside the containment. The outside containment lines in this circulation loop can be Class C or a nonsafety-related class if suitable containment isolation valves are provided.
It introduces emergency negative reactivity to make the reactor subcritical (for example, control rods).
This class also applies to structures, systems, and components where leakage could cause a loss of adequate core cooling. In isolating leaks, credit can be taken for automatic safety-related isolation and for appropriate operator action. As a minimum, operator action needs redundant safety-related indication and alarm followed by 30 minutes for operator action.
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VEGP 3&4 - UFSAR 10 CFR 21 applies to Class B structures, systems, and components. Class B structures, systems, and components are seismic Category I and use codes and standards consistent with the guidelines for NRC Quality Group B. 10 CFR 50, Appendix B applies. ASME Code,Section III, Class 2 or Class MC applies to pressure retaining components. ASME Code,Section III, Subsection NE applies to the containment vessel and guard pipes.
3.2.2.5 Equipment Class C Class C is a safety-related class equivalent to ANS Safety Class 3. It applies to other safety-related functions required to mitigate design basis accidents and other design basis events. Minor leakage will not prevent Class C structures, systems, and components from meeting the safety-related function, either from the regard of radiation dose or system functioning.
This class also applies to equipment that, upon rupturing, would cause dose limits for unrestricted areas, as specified in 10 CFR 20, to be exceeded or would cause a loss of core cooling.
10 CFR 21 applies to Class C structures, systems, and components. Class C structures, systems, and components use codes and standards consistent with the guidelines for NRC Quality Group C.
Class C structures, systems, and components are seismic Category I except those noted below which are not required to provide a safety-related function following a seismic event. 10 CFR 50, Appendix B applies. ASME Code,Section III, Class 3 applies to pressure retaining components with some exceptions as identified below. In addition to these requirements, for systems that provide emergency core cooling functions, full radiography in accordance with the requirements of ASME Code,Section III, ND-5222 of a random sample of welds will be conducted on the piping butt welds during construction. For Class C air and gas storage tanks fabricated without welding, ASME Code,Section VIII, Appendix 22 may be used in lieu of Section III, Class 3. For Class C fuel handling system weir gate pressure boundary components containing only compressed air, Manufacturers Standard may be used in lieu of Section III, Class 3. 10 CFR 50, Appendix B requirements and 10 CFR 21 apply to the manufacture of safety-related air and gas storage tanks. For core support structures ASME Code,Section III, Subsection NG applies. For electrical systems, appropriate IEEE standards, including IEEE standard 323-74 (Reference 3) and IEEE standard 344-87 (Reference 4),
apply.
Class C applies to structures, systems, and components not included in Class A or Class B that are designed and relied upon to accomplish one or more of the following safety-related functions:
Provide safety injection or maintain sufficient reactor coolant inventory to allow for core cooling Provide core cooling Provide containment cooling Provide for removal of radiation from the containment atmosphere as necessary to meet the offsite dose limits Limit the buildup of radioactive material in the atmosphere of rooms and areas outside containment as necessary to meet the offsite dose limits Introduce negative reactivity control measures to achieve or maintain safe shutdown conditions (for example, boron addition)
Maintain geometry of structures inside the reactor vessel so that the control rods can be inserted (when required) and the fuel remains in a coolable geometry 3.2-6 Revision 10
VEGP 3&4 - UFSAR Provide load-bearing structures and supports for Class A, B, and C structures, systems, and components. This applies to structures and supports that are not part of the pressure boundary.
Provide structures to protect Class A, B, and C structures, systems, and components from events such as internal/external missiles, seismic, flooding, and the dynamic and environmental effects of pipe failures. Structures protecting equipment from nonseismic events are not required to be seismic Category I. Exceptions to this equipment Class C criterion are the equipment Class D barriers and seismic Category II structures described in Subsection 3.5.2 that provide tornado missile protection. Turbine building first bay building structure Wall 11.2 provides protection of Wall 11 from HELB loads resulting from postulated ruptures in main steam and main feedwater piping north of the turbine building first bay.
Provide permanent radiation shielding to allow operator access to the main control room and to limit the exposure to Class A, B and C structures, systems, and components Provide safety support functions to Class A, B and C structures, systems, and components, such as, heat removal, room cooling, and electrical power Provide instrumentation and controls for automatic or manual actuation of Class A, B, and C structures, systems, and components necessary to perform the safety-related functions of the Class A, B, or C structure, system or component. This includes the processing of signals and interlock functions required for proper safety performance of these structures, systems, and components.
Maintain spent fuel integrity, the failure of which could result in fuel damage such that significant quantities of radioactive material could be released from the fuel and results in offsite doses greater than normal limits (for example, spent fuel pool, fuel transfer tube isolation valve)
Maintain spent fuel sub-critical Monitor variables to indicate status of Class A, B or C structures, systems, and components required for post-accident mitigation Provide for functions defined in Class B where structures, systems, and components, or portions thereof are not within the scope of the ASME Code,Section III, Class 2.
Provide provisions for connecting temporary equipment to extend the use of safety related systems. See Subsection 1.9.5 for a discussion of actions required for an extended loss of onsite and offsite ac power sources.
The components and portions of systems that provide emergency core cooling functions and are required to have radiography of a random sample of welds during construction include the following:
Accumulators Injection piping from the accumulators to the reactor coolant system isolation check valves in the direct vessel injection line Piping from the in-containment refueling water storage tank (IRWST) and recirculation screens to the reactor coolant system isolation check valves in the direct vessel injection line 3.2-7 Revision 10
VEGP 3&4 - UFSAR Piping from the Stage 1, 2, and 3 automatic depressurization system valves to the IRWST including the spargers.
The IRWST is formed from portions of structural modules that are elements of the containment internal structures. The inspection requirements for the welds in these structural modules are provided in Subsection 3.8.3.6.2.
3.2.2.6 Equipment Class D Class D is nonsafety-related with some additional requirements on procurement, inspection or monitoring.
For Class D structures, systems, and components containing radioactivity, it is demonstrated by conservative analysis that the potential for failure due to a design basis event does not result in exceeding the normal offsite doses per 10 CFR 20. This criterion is in conformance with the definition of Class D in Regulatory Guide 1.26.
A structure, system or component is classified as Class D when it directly acts to prevent unnecessary actuation of the passive safety systems. Structures, systems and components which support those which directly act to prevent the actuation of passive safety systems are also Class D.
The inclusion of these nonsafety-related structures, systems, and components in Class D recognizes that these systems provide an important first level of defense that helps to reduce the calculated probabilistic risk assessment core melt frequency. These structures, systems, and components are normally used to support plant cooldown and depressurization and to maintain shutdown conditions during maintenance and refueling outages.
For Class D structures, systems, and components considered to be risk significant as defined in the reliability assurance program (see Section 16.2). Provisions are made to check for operability, including appropriate testing and inspection, and to repair out-of-service structures, systems, and components. These provisions are documented and administered in the plant reliability assurance plan and operating and maintenance procedures.
A portion of chemical and volume control system is defined as the reactor coolant pressure boundary and is Class D. This portion of the chemical and volume control system is seismically analyzed. See Subsection 5.2.1.1 for the seismic analysis requirements.
Some Class D structures, systems, and components are assumed to function in a severe containment environment. The design requirements for these components include operation in such an environment. An evaluation is done to confirm that the structure, system, or component can be expected to function in such an environment.
Standard industrial quality assurance standards are applied to Class D structures, systems, and components to provide appropriate integrity and function although 10 CFR 50, Appendix B and 10 CFR 21 do not apply. 10 CFR 50, Appendix B and 10 CFR 21 do apply to Class D structures, systems, and components that are seismic Category I. Pertinent portions of 10 CFR 50, Appendix B are applied to seismic Category II applications as described in Subsection 3.2.1.1.2. These industrial quality assurance standards are consistent with the guidelines for NRC Quality Group D. The industry standards used for Class D structures, systems and components are widely used industry standards. Typical industrial standards used for Class D systems and components are provided as follows:
Pressure vessels - ASME Code,Section VIII 3.2-8 Revision 10
VEGP 3&4 - UFSAR Piping - ASME B31.1. Power Piping, (Reference 5). NDE of welds during fabrication and installation may use methods and acceptance criteria of ASME B31.1, 2007 Edition. Post Weld Heat Treatment of welds may use methods and acceptance criteria of ASME B31.1, 2014 Edition.
Pumps - API 610 (Reference 6), or Hydraulic Institute Standards Valves - ANSI B16.34 (Reference 8)
Atmospheric storage tanks - API-650 (Reference 9), AWWA D 100 (Reference 10), or ANSI B96.1 (Reference 11) 0 - 15 psig Storage Tanks - API-620 (Reference 12)
AC motor and generators - NEMA MG1 (Reference 13)
Circuit breakers, switchgear, relays, substations and fuses - IEEE C37 (Reference 14).
The NS buildings (except for the NS portions of the turbine and annex buildings outlined in Table 3.2-2) containing Class D structures, systems, and components, as well as the anchorage of the structures, systems, and components to the building, are designed to the seismic requirements of the Uniform Building Code (Reference 15). The NS portions of the turbine building and annex building are designed to the requirements of the International Building Code, IBC-06 (Reference 19).
The systems and components are not designed for seismic loads. However, when Class D structures, systems, and components are located near a Class A, B, or C structure, system, or component, the requirements for seismic Category II may apply. The tornado missile barriers in the turbine building first bay building structure identified in Table 3.5-1 are seismic Category II, equipment Class D.
For Class D structures, systems, and components required to be monitored for maintenance effectiveness by 10 CFR 50.65, the availability parameters and criteria are included in the maintenance monitoring plan for evaluating the effectiveness of the maintenance program.
As examples, Class D applies to structures, systems, and components not included in Class A, B or C that provide the following functions:
Provide core or containment cooling which prevents challenges to the passive core cooling system and the passive containment cooling system Process, extract, encase, store or reuse radioactive fluid or waste Verify that plant operating conditions are within technical specification limits Provide permanent shielding for post accident access to Class A, B or C structures, systems, and components or of offsite personnel Handle spent fuel, the failure of which could result in fuel damage such that limited quantities of radioactive material could be released from the fuel (for example, fuel handling machine, spent fuel handling tool, new and spent fuel racks)
Protect Class B or C structures, systems, and components necessary to attain or maintain safe shutdown following fire 3.2-9 Revision 10
VEGP 3&4 - UFSAR Indicate the status of protection system bypasses that are not automatically removed as a part of the protection system operation Aid in determining the cause or consequences of an event for post-accident investigation Prevent interaction that could result in preventing Class A, B or C structures, systems, and components from performing required safety-related functions Limit the buildup of hydrogen in the containment atmosphere to acceptable values 3.2.2.7 Other Equipment Classes Equipment classes E, F, G, L, P, R, and W are nonsafety-related. They apply to structures, systems, and components not covered in the above classes. They have no safety-related function to perform.
They do not contain sufficient radioactive material that a release could exceed applicable limits.
Structures, systems, and components that do not normally contain radioactive fluids, gases, or solids but have the potential to become radioactively contaminated are classified as one of these nonsafety-related classes if all of the following criteria are satisfied:
The system is only potentially radioactive and does not normally contain radioactive material, and The system has shown in plant operations that the operation with the system containing radioactive material meets or can meet unrestricted area release limits, and An evaluation of the system confirms that the system contains features and components that keep the consequences of a system failure as low as reasonably achievable, and The system has no other regulatory guidance requiring its inclusion in Classes A, B, C or D.
This review of the system functions, features, and components includes the following as a minimum:
Features and components that control and limit the radioactive contamination in the system Features that facilitate an expeditious cleanup should the system become contaminated Features and components that limit and control the radiological consequences of a potential system failure Monitor radioactive effluent to confirm that release rates or total releases are within limits established for normal operations and transient operation The means by which the system prevents propagation to an event of greater consequence.
There are no special quality assurance requirements for Class E, F, G, L, P, R, and W structures, systems, and components. Unless specifically specified, 10 CFR Part 21 and Part 50, Appendix B do not apply. The systems and components are normally not designed for seismic loading. However, there may be special cases where some seismic design is required. See Subsection 3.2.1 for more details.
Structures, systems, and components are designed in accordance with an industry standard at the discretion of the designer. The following provides examples of industry standards which may be used for these classes:
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VEGP 3&4 - UFSAR Class E - This class is used for nonsafety-related structures, systems, and components constructed to industry standards which are not noted in the following classes. Class Y is a subset of Class E and is not utilized in the UFSAR.
Class F and G - These classes are used for Fire Protection Systems. They comply with National Fire Protection Association Codes which invoke ANSI B31.1 (Reference 5), AWWA (American Water Works Association), API (American Petroleum Institute), Underwriters Laboratories (UL), and other codes, depending on service. See Subsection 9.5.1 for quality assurance requirements for fire protection structures, systems, and components. Portions of fire protection systems that protect safety-related SSCs are designated as AP1000 equipment Class F, which meets the requirements of ASME B31.1 and requires seismic analysis. NDE of welds during fabrication and installation may use methods and acceptance criteria of ASME B31.1, 2007 Edition. Post Weld Heat Treatment of welds may use methods and acceptance criteria of ASME B31.1, 2014 Edition.
Class L - This class is used in heating, ventilation and air-conditioning systems. It complies with SMACNA - 1995 (Reference 16). Components may also be procured to AMCA and ASHRAE standards.
Class P - This class is used for plumbing equipment. It complies with the National Plumbing Code (Reference 17).
Class R - This class is for air cleaning units and components that may be required to contain, clean, or exclude radioactively contaminated air. It complies with ASME N509 (Reference 18) and ASME AG-1 (Reference 20) as noted in Table 3.2-3. When used with 10 CFR Part 50 Appendix B quality assurance and 10 CFR Part 21, it is equivalent to Class C. If Class R components are nonsafety-related, the quality assurance requirements invoked in ASME N509 and ASME AG-1 do not apply. When used with 10 CFR Part 50 Appendix B quality assurance, it is equivalent to Class C.
Class W - This class complies with American Water Works Association guidelines with no specific quality assurance requirements.
3.2.2.8 Instrumentation and Control Line Interface Criteria Class C instrumentation, as defined in Subsection 3.2.2.5 have a safety-related equipment class pressure boundary including the sensing line, valves and instrument sensor. The pressure boundary is the same safety-related equipment class as the systems or components it is connected to. Sensing lines connected to the reactor coolant system pressure boundary are Class B if a suitable flow restrictor is provided.
Nonsafety-related instrumentation that monitors safety-related fluid systems is Class D, as defined in Subsection 3.2.2.6. The instrument sensing line is safety-related, seismic Category I from the connected fluid system to the instrument manifold. The instrument, manifold, and impulse line (interconnecting tubing between the manifold and instrument) are Class D, seismic Category II, as defined in Subsection 3.2.1.1.2. The Class D quality requirements include a pressure test at 1.5 times design pressure and a requirement for certified material test reports.
The parts of the sensor, outside the pressure boundary, are designated Class C (1E) if they provide a safety-related function per Subsection 3.2.2.1. They are Class D if the instrument supports Class D functions per Subsection 3.2.2.6. Otherwise the parts are Class E.
3.2.2.9 Electrical Classifications Safety-related electrical equipment is equipment Class C, as outlined in Subsection 3.2.2.5, and is constructed to IEEE standards for Class 1E. The nonsafety-related electrical equipment and 3.2-11 Revision 10
VEGP 3&4 - UFSAR instrumentation is constructed to standards including non-Class 1E IEEE standards and National Electrical Manufacturers Association (NEMA) standards. Safety-related electrical equipment and instrumentation is identified in Section 3.11.
3.2.3 Inspection Requirements Safety-related structures, systems, and components built to the requirements of the ASME Code,Section III, are required by 10 CFR 50.55a to have in-service inspections. The requirements of the in-service inspection program for ASME Code,Section III structures, systems, and components are found in Section XI of the ASME Code.
The following ASME standards apply to safety-related structures, systems, and components:
Pumps (Class A, B, C) - ASME OM Code, Subsection ISTB Valves (Class A, B, C) - ASME OM Code, Subsection ISTC Equipment supports (Class A, B, C) - ASME Code,Section XI, Subsection IWF Metal containments and vessels - ASME Code,Section XI, Subsection IWE Other Class A components such as pipes and tanks - ASME Code,Section XI, Subsection IWB Other Class B components such as pipes and tanks - ASME Code,Section XI, Subsection IWC Other Class C components such as pipes and tanks - ASME Code,Section XI, Subsection IWD.
The inspection requirements, if applicable, for Class D structures, systems, and components are established by the designer for each structure, system, and component. These inspection requirements are developed so that the reliability of the structures, systems, and components is not degraded. The inspection requirements are included in the administratively controlled inspection or maintenance plans.
3.2.4 Application of AP1000 Safety-Related Equipment and Seismic Classification System The application of the AP1000 equipment and seismic classification system to AP1000 systems and components is shown in Table 3.2-3. Table 3.2-3 lists safety-related and seismic Category I mechanical and fluid system component and associated equipment class and seismic category as well as other related information. The table also provides information on the systems that contain Class D components. Additional information on the Class D functions of the various systems can be found in the description in the Design Certification Document (DCD) for the systems. Mechanical and fluid systems that contain no safety-related or Class D systems are included in the table and general information provided on the system. Supports for piping and components have the same classification as the component or piping supported. Supports for AP1000 equipment Class A, B, and C mechanical components and piping are constructed to ASME Code,Section III, Subsection NF requirements. The principal construction code for supports for nonsafety-related components and piping is the same as that for the supported component or piping.
Following the name of each system is the building location of the system components. Some of the systems supply all or most of the buildings. This is indicated by identifying the location as various.
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VEGP 3&4 - UFSAR Where a system includes piping or ducts that only passed through a building without including any components that building is generally not included in the list.
The following list includes the systems in Table 3.2-3. The three letters in the beginning of each line is the acronym for the system. The systems included in Table 3.2-3 are listed alphabetically by three letter acronym. Those systems marked with an asterisk
- are electrical or instrumentation systems and are not included in Table 3.2-3. The components in the incore instrumentation system that have a pressure boundary function are included in the table. See Section 3.11 for identification of safety-related electrical and instrumentation equipment.
NSSS/Steam Generator Controls and Auxiliaries BDS Steam Generator Blowdown System CNS Containment System CVS Chemical and Volume Control System PCS Passive Containment Cooling System PXS Passive Core Cooling System RCS Reactor Coolant System RNS Normal Residual Heat Removal System RXS Reactor System SGS Steam Generator System Nuclear Control and Monitoring
- DAS Diverse Actuation System IIS Incore Instrumentation System
- OCS Operation and Control Centers
- PMS Protection and Safety Monitoring System PSS Primary Sampling System
- RMS Radiation Monitoring System
- SJS Seismic Monitoring System
- SMS Special Monitoring System Main Power Cycle and Auxiliaries CDS Condensate System 3.2-13 Revision 10
VEGP 3&4 - UFSAR CFS Turbine Island Chemical Feed System CPS Condensate Polishing System DTS Demineralized Water Treatment System DWS Demineralized Water Transfer and Storage System FWS Main and Startup Feedwater System GSS Gland Seal System HDS Heater Drain System MSS Main Steam System MTS Main Turbine System RWS Raw Water System TDS Turbine Island Vents, Drains and Relief System Class 1E and Emergency Power Systems
- IDS Class 1E dc and UPS System Cooling and Circulating Water CCS Component Cooling Water System CES Condenser Tube Cleaning System CWS Circulating Water System SFS Spent Fuel Pool Cooling System SWS Service Water System TCS Turbine Building Closed Cooling Water System Auxiliary Steam ASS Auxiliary Steam Supply System Generation and Transmission
- ZAS Main Generation System 3.2-14 Revision 10
- ZBS Transmission Switchyard and Offsite Power System
- ZVS Excitation and Voltage Regulation System Radwaste WGS Gaseous Radwaste System WLS Liquid Radwaste System WRS Radioactive Waste Drain System WSS Solid Radwaste System HVAC VAS Radiologically Controlled Area Ventilation System VBS Nuclear Island Nonradioactive Ventilation System VCS Containment Recirculation Cooling System VES Main Control Room Emergency Habitability System VFS Containment Air Filtration System VHS Health Physics and Hot Machine Shop HVAC System VLS Containment Hydrogen Control System VRS Radwaste Building HVAC System VTS Turbine Building Ventilation System VUS Containment Leak Rate Test System VWS Central Chilled Water System VXS Annex/Auxiliary Nonradioactive Ventilation System VYS Hot Water Heating System VZS Diesel Generator Building Ventilation System Turbine-Generator Controls and Auxiliary CMS Condenser Air Removal System HCS Generator Hydrogen and CO2 Systems 3.2-15 Revision 10
VEGP 3&4 - UFSAR HSS Hydrogen Seal Oil System LOS Main Turbine and Generator Lube Oil System
- TOS Main Turbine Control and Diagnostics System Material Handling FHS Fuel Handling and Refueling System MHS Mechanical Handling System Piping Services CAS Compressed and Instrument Air Systems DOS Standby Diesel Fuel Oil System FPS Fire Protection System PGS Plant Gas Systems PWS Potable Water System Non-Class 1E Power Systems
- ECS Main AC Power System
- EDS Non-Class 1E dc and UPS System ZOS Onsite Standby Power System
- ZRS Offsite Retail Power System Miscellaneous Electrical Systems
- EFS Communication Systems
- EGS Grounding and Lightning Protection System
- EHS Special Process Heat Tracing System
- ELS Plant Lighting System
- EQS Cathodic Protection System Non-Nuclear Controls and Monitoring 3.2-16 Revision 10
- DDS Data Display and Processing System
- MES Meteorological and Environmental Monitoring System
- PLS Plant Control System
- SES Plant Security System SSS Secondary Sampling System
- TVS Closed Circuit TV System Non-Radioactive Drains DRS Storm Drain System RDS Gravity and Roof Drain Collection System SDS Sanitary Drainage System WWS Waste Water System Those systems marked with an asterisk (*) are electrical or instrumentation systems and are not included in Table 3.2-3.
3.2.5 Combined License Information This section contained no requirement for additional information.
3.2.6 References
- 1. ANSI N18.2a - 1975, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants.
- 2. ANS/ANSI 51.1 - 1983, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants.
- 3. IEEE 323-74, IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations.
- 4. IEEE 344-1987, IEEE Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations.
- 5. ASME B31.1-1989 Edition, Power Piping, including 1989 Addenda.
- 6. API 610-81, Centrifugal Pumps for General Refinery Services.
- 7. Not used.
- 8. ANSI B 16.34 - 1981, Valves - Flanged and Buttwelding End.
- 9. API-650-80, Welded Steel Tanks for Oil Storage, Revision 1, February 1984.
3.2-17 Revision 10
- 10. AWWA D100-84, Welded Steel Tanks for Water Storage.
- 11. ANSI B96.1-81, Welded Aluminum-Alloy Storage Tanks.
- 12. API-620-82, Recommended Rules for Design and Construction of Large, Welded, Low-Pressure Storage Tanks, Revision 1, April 1985.
- 13. NEMA MG-1-98, Motors and Generators, Revision 1, January 1998, National Electric Manufacturers Association.
- 14. IEEE C37, IEEE standards on circuit breakers, switch gear, relays, substations, fuses, etc.
- 15. Uniform Building Code (1997), International Conference of Building Officials.
- 16. SMACNA - 1995, HVAC Duct Construction Standards - Metal and Flexible, Sheet Metal and Air-Conditioning Contractors National Association.
- 17. The BOCA Basic/National Plumbing Code 1984: Model Plumbing Regulations for the Protection of Public Health, Safety and Welfare: Sixth Edition, Building Officials and Code Administrators International.
- 18. ASME N509-1989, Nuclear Power Plant Air Cleaning Units and Components.
- 19. International Building Code, 2006.
- 20. ASME AG-1-1997, Code on Nuclear Air and Gas Treatment.
3.2-18 Revision 10
VEGP 3&4 - UFSAR Table 3.2-1 Comparison of Safety Classification Requirements AP1000 ANS Equipment RG 1.29 Seismic ASME Code, RG 1.26 NRC 10 CFR 50 Inspection &
Code Letter Safety Class Design Reqmnts Sec. III Class IEEE Quality Group Appendix B Testing Required (1) (2) (3) (4) Requirements (5) (6) Requirements Test & Maint.
A SC-1 I 1 NA GROUP A YES YES(7) (8)
B SC-2 I 2 NA GROUP B YES YES(7) (8)
C SC-3 I(14) 3 1E GROUP C YES YES(7) (8)
D NNS(2) NA(9)(14) NA(10) (10) GROUP D NO(10) YES(11) (11)
OTHER NNS(2) NA(13) NA NA NA NA(12) NA NA NA - Not Applicable OTHER includes Classes E, F, L, P, R, and W.
Notes:
- 1. A single letter equipment classification identifies the safety class, quality group, and other classifications for AP1000. See the Subsection 3.2.2 for definition.
- 2. AP1000 safety classification is an adaptation of that defined in ANSI 51.1. The NNS defined in the ANSI 51.1 standard is divided into several AP1000 equipment classifications namely, Classes D E, F, L, P, R, and W.
- 3. See Subsection 3.2.1 for definition of seismic categories.
- 4. ASME Boiler and Pressure Vessel Code,Section III defines various classes of structures, systems, and components for nuclear power plants. It defines criteria and requirements based on the classification. It is not applicable for nonsafety-related components.
- 5. The quality group classification corresponds to those provided in Regulatory Guide 1.26.
- 6. Yes means quality assurance program is required according to 10 CFR 50 Appendix B.
No means quality assurance program is not required according to 10 CFR 50 Appendix B.
- 7. Class A, B, and C, structures, systems, and components built to ASME Code,Section III are inspected to ASME Code,Section XI requirements. See the text for additional specification of requirements.
- 8. Class A, B, and C structures, systems, and components that are required to function to mitigate design base accidents have some testing requirements included in the plant technical specifications. In addition to the requirements in the technical specifications, testing and maintenance requirements are included in an administratively controlled reliability assurance plan.
- 9. See Subsection 3.2.1 for cases when seismic Category II requirements are applicable for Class D structures, systems, and components.
- 10. See the text for a discussion of the industry standards used in the construction of Class D structures, systems and components.
- 11. Class D structures, systems, and components have selected reliability assurance programs and procedures to provide availability when needed. These programs are administratively controlled programs and are not included in the technical specifications.
- 12. Normal industrial procedures are followed in procuring, designing, fabricating, and testing these nonsafety-related structures, systems, and components.
- 13. Some Class E, F, G, L, P, R, and W structures, systems, and components may be classified as seismic Category II. See Subsection 3.7.3.
- 14. See Subsection 3.5.2 for seismic Category II tornado missile barriers that provide tornado missile protection for the openings and penetrations in auxiliary building walls.
3.2-19 Revision 10
VEGP 3&4 - UFSAR Table 3.2-2 Seismic Classification of Building Structures Structure Category(1)
Nuclear Island C-I Basemat Containment Interior Shield Building Auxiliary Building Containment Air Baffle Containment Vessel C-I Plant Vent and Stair Structure C-II Turbine Building - First bay adjacent to Nuclear Island outlined by Columns I.1 to R, 11.05 to 11.2, C-II and 11.02 to 11.2 Turbine Building - All portions of Turbine Building except first bay adjacent to Nuclear Island as NS(2) outlined by Columns H.05 to R and 12.1 to 20 Turbine Building NS(2)
Annex Building Area Outlined by Columns A - D and 8 - 13 NS(2)
Area Outlined by Columns A - G and 13 - 16 Annex Building Area Outlined by Columns E - I.1 and 2 - 13 C-II Radwaste Building NS(2)
Diesel-Generator Building NS(3)
Circulating Water Pumphouse and Towers NS Safety-Related Backfill C-I C-I - Seismic Category I C-II - Seismic Category II NS - Non-seismic Note:
- 1. Within the broad definition of seismic Category I and II structures, these buildings contain members and structural subsystems the failure of which would not impair the capability for safe shutdown. Examples of such systems would be elevators, stairwells not required for access in the event of a postulated earthquake, and nonstructural partitions in nonsafety-related areas. These substructures are classified as non-seismic.
- 2. The NS designation for the turbine building, the radwaste building, and a portion of the annex building indicates that the buildings are not seismic Category I or seismic Category II. The seismic requirements for these buildings are outlined in Subsection 3.7.2.8.
- 3. The NS designation for the diesel-generator building indicates that the building is not seismic Category I or seismic Category II.
The seismic requirements for buildings containing Class D equipment, including the diesel generator building, are outlined in Subsection 3.2.2.6.
3.2-20 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 1 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Auxiliary Steam Supply System (ASS) Location: Turbine Building System components are Class E Steam Generator Blowdown System (BDS) Location: Turbine Building BDS-PL-V009 Discharge to WLS Vent D NS ANSI B16.34 BDS-PL-V037 Discharge to WLS Valve D NS ANSI B16.34 BDS-PL-V046 WWS Contaminated Flow D NS ANSI B16.34 Connection Balance of system components are Class E Compressed and Instrument Air System (CAS) Location: Various CAS-PL-V014 Instrument Air Supply Outside B I ASME III-2 Containment Isolation CAS-PL-V015 Instrument Air Supply Inside B I ASME III-2 Containment Isolation CAS-PL-V027 Containment Penetration Test B I ASME III-2 Connection Isolation CAS-PL-V204 Service Air Supply Outside B I ASME III-2 Containment Isolation CAS-PL-V205 Service Air Supply Inside B I ASME III-2 Containment Isolation CAS-PL-V219 Containment Penetration Test B I ASME III-2 Connection Isolation CAS-PY-C02 Containment Instrument Air B I ASME III, MC Inlet Penetration CAS-PY-C03 Containment Service Air B I ASME III, MC Inlet Penetration Balance of system components are Class E Component Cooling Water System (CCS) Location: Auxiliary Building and Turbine Building n/a Heat Exchangers, CCS and D NS ASME VIII SWS Side n/a Pumps D NS Hydraulic Institute Stds.
n/a Tanks D NS ASME VIII n/a Valves Providing CCS D NS ANSI B16.34 AP1000 Equipment Class D Function CCS-PL-V200 CCS Containment Isolation B I ASME III-2 Valve - Inlet Line ORC CCS-PL-V201 CCS Containment Isolation B I ASME III-2 Valve - Inlet Line IRC CCS-PL-V207 CCS Containment Isolation B I ASME III-2 Valve - Outlet Line IRC CCS-PL-V208 CCS Containment Isolation B I ASME III-2 Valve - Outlet Line ORC CCS-PL-V209 Containment Isolation Valve B I ASME III-2 Test Connection - Outlet Line CCS-PL-V214 CCS Supply Containment C I ASME III-3 Isolation - IRC 3.2-21 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 2 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Component Cooling Water System (Continued)
CCS-PL-V215 CCS Supply Containment C I ASME III-3 Isolation Valve Test Connection - IRC CCS-PL-V216 Containment Leak Test Outlet C I ASME III-3 Line - IRC CCS-PL-V217 Containment Isolation Valve C I ASME III-3 V207 Body Test Connection Valve CCS-PL-V270 CCS IRC Relief Valve C I ASME III-3 CCS-PL-V271 CCS IRC Relief Valve C I ASME III-3 CCS-PL-V220 CCS Containment Isolation B I ASME III-2 Relief Valve CCS-PL-V257 Containment Isolation Valve B I ASME III-2 Test Connection - Inlet Line CCS-PY-C01 Containment Supply Header B I ASME III, MC Penetration CCS-PY-C02 Containment Return Header B I ASME III, MC Penetration Balance of system components are Class E Condensate System (CDS) Location: Turbine Building System components are Class E Condenser Tube Cleaning System (CES) Location: Turbine Building System components are Class E Turbine Island Chemical Feed System (CFS) Location: Turbine Building and SWS Chemical Treatment Building/Area System components are Class E Condenser Air Removal System (CMS) Location: Turbine Building n/a Condenser Vacuum Breakers E NS ANSI B16.34 Balance of system components are Class D Containment System (CNS) Location: Containment CNS-MV-01 Containment Vessel B I ASME III, MC CNS-MY-Y01 Equipment Hatch B I ASME III, MC CNS-MY-Y02 Maintenance Hatch B I ASME III, MC CNS-MY-Y03 Upper Personnel Hatch - B I ASME III, MC 135-3 CNS-MY-Y04 Lower Personnel Hatch - B I ASME III, MC 107-2 n/a Spare Containment B I ASME III, MC Penetrations Condensate Polishing System (CPS) Location: Turbine Building System components are Class E Location: Containment, Auxiliary Building, and Chemical and Volume Control System (CVS) Annex Building n/a Heat Exchangers, CVS and D NS ASME VIII/ TEMA CCS Side 3.2-22 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 3 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Chemical and Volume Control System (Continued) n/a Pumps D NS Hydraulic Institute Stds.
n/a Tanks (Except CVS-MT-03 D NS API 650 and CVS-MT-05)
CVS-MT-03 CVS Chemical Mixing Tank E NS ASME VIII CVS-MT-05 CVS Air Intrusion Prevention D NS ASME VIII Tank n/a Demineralizers D NS ASME VIII n/a Filters D NS ASME VIII n/a Valves Providing CVS D NS ANSI B16.34 AP1000 Equipment Class D Function CVS-PL-V001 RCS Purification Stop A I ASME III-1 CVS-PL-V002 RCS Purification Stop A I ASME III-1 CVS-PL-V003 RCS Purification Stop C I ASME III-3 CVS-PL-V040 Resin Flush IRC Isolation B I ASME III-2 CVS-PL-V041 Resin Flush ORC Isolation B I ASME III-2 CVS-PL-V042 Flush Line Containment B I ASME III-2 Isolation Relief CVS-PL-V045 Letdown Containment B I ASME III-2 Isolation IRC CVS-PL-V046 Letdown Pressure Instrument B I ASME III-2 Root CVS-PL-V047 Letdown Containment B I ASME III-2 Isolation ORC CVS-PL-V058 Letdown Line Containment B I ASME III-2 Isolation Thermal Relief CVS-PL-V065 Zinc Addition - IRC Shutoff C I ASME III-3 CVS-PL-V067 Makeup Return Line Bypass A I ASME III-1 Check Valve CVS-PL-V080 RCS Purification Return Line C I ASME III-3 Check Valve CVS-PL-V081 RCS Purification Return Line A I ASME III-1 Stop Valve CVS-PL-V082 RCS Purification Return Line A I ASME III-1 Check Valve CVS-PL-V084 Auxiliary Pressurizer Spray A I ASME III-1 Line Isolation CVS-PL-V085 Auxiliary Pressurizer Spray A I ASME III-1 Line CVS-PL-V090 Makeup Line Containment B I ASME III-2 Isolation CVS-PL-V091 Makeup Line Containment B I ASME III-2 Isolation CVS-PL-V092 Zinc Injection Containment B I ASME III-2 Isolation ORC 3.2-23 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 4 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Chemical and Volume Control System (Continued)
CVS-PL-V094 Zinc Injection Containment B I ASME III-2 Isolation IRC CVS-PL-V095 Zinc Add Containment C I ASME III-3 Isolation Test Connection CVS-PL-V096 Zinc Injection Containment B I ASME III-2 Isolation Test Connection CVS-PL-V098 Zinc Addition Line B I ASME III-2 Containment Isolation Thermal Relief Valve CVS-PL-V100 Makeup Line Containment B I ASME III-2 Isolation Relief CVS-PL-V136A Demineralized Water System C I ASME III-3 Isolation CVS-PL-V136B Demineralized Water System C I ASME III-3 Isolation CVS-PL-V215 Hydrogen Injection - IRC C I ASME III-3 Shutoff CVS-PL-V216 Hydrogen Injection C I ASME III-3 Containment Isolation Test Connection CVS-PL-V217 Hydrogen Injection B I ASME III-2 Containment Isolation Check IRC CVS-PL-V218 Hydrogen Injection B I ASME III-2 Containment Isolation Test Connection CVS-PL-V219 Hydrogen Injection B I ASME III-2 Containment Isolation ORC CVS-PY-C01 Demineralizer Resin Flush B I ASME III, MC Line Containment Penetration CVS-PY-C02 Letdown Line Containment B I ASME III, MC Penetration CVS-PY-C03 Makeup Line Containment B I ASME III, MC Penetration CVS-PY-C04 Zinc Add Line Containment B I ASME III, MC Penetration CVS-PY-C05 Hydrogen Add Line B I ASME III, MC Containment Penetration Balance of system components are Class D or E Circulating Water System (CWS) Location: Turbine Building and pump intake structure System components are Class E 3.2-24 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 5 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Standby Diesel Fuel Oil System (DOS) Location: Diesel Generator Building and yard n/a Fuel Oil Transfer Package D NS Manufacturer Std.
n/a Fuel Oil Storage Tanks D NS API 650 n/a Fuel Oil Day Tanks D NS ASME VIII n/a Valves Providing DOS D NS ANSI B16.34 AP1000 Equipment Class D Function n/a Ancillary Diesel Generator D II Non-Stamped ASME Located in Annex Fuel Tank VIII Building Balance of system components are Class E Storm Drain System (DRS) Location: Various System components are Class E Demineralized Water Treatment System (DTS) Location: Turbine Building System components are Class E Demineralized Water Transfer and Storage System (DWS) Location: Various n/a Condensate Storage Tanks D NS API 650 n/a Valves Providing DWS D NS ANSI B16.34 AP1000 Equipment Class D Function DWS-PL-V241 DWS Containment C I ASME III-3 Penetration Thermal Relief Valve DWS-PL-V244 Demineralized Water Supply B I ASME III-2 Containment Isolation -
Outside DWS-PL-V245 Demineralized Water Supply B I ASME III-2 Containment Isolation - Inside DWS-PL-V248 Containment Penetration Test B I ASME III-2 Connection Isolation DWS-PY-C01 Containment Demineralized B I ASME III, MC Water Supply Penetration Balance of system components are Class E Electrical Distribution System (ECS) Location: Annex Building n/a Ancillary Diesel Generator D NS Manufacturer Anchorage is SCII Engines Standard n/a Ancillary Diesel Generator D NS Manufacturer Std.
Radiators n/a Ancillary Diesel Generator D NS Manufacturer Std.
Silencers n/a Valve providing fuel to ECS D NS ANSI B16.34 Ancillary Diesel Generators Balance of system components are Class E 3.2-25 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 6 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Fuel Handling and Refueling System (FHS) Location: Containment and Auxiliary Building FHS-FH-01 Refueling Machine D II AISC FHS-FH-02 Fuel Handling Machine D II AISC FHS-FH-04 New Fuel Elevator D II AISC FHS-FH-05 Fuel Transfer System D II AISC FHS-FH-52 Spent Fuel Assembly D II AISC Handling Tool FHS-FS-01 New Fuel Storage Rack D I Manufacturer Std.
FHS-FS-02 Spent Fuel Storage Rack D I Manufacturer Std.
FHS-FT-01 Fuel Transfer Tube B I ASME III Class MC Jurisdictional Boundary CV Only, Remaining Portion Optional FHS-MT-01 Spent Fuel Pool C I ACI 349 ACI 349 Evaluation of Structural Boundary Only FHS-MT-02 Fuel Transfer Canal C I ACI 349 ACI 349 Evaluation of Structural Boundary Only FHS-MT-03 Refueling Cavity C I ACI 349 ACI 349 Evaluation of Structural Boundary Only FHS-MT-05 Spent Fuel Cask Loading Pit C I ACI 349 ACI 349 Evaluation of Structural Boundary Only FHS-MT-06 Spent Fuel Cask Washdown C I ACI 349 ACI 349 Evaluation of Pit Structural Boundary Only FHS-MY-Y01 Spent Fuel Transfer Gate C I ANSI/AISC N690 for Manufacturers Std for structural steel non-structural components FHS-MY-Y02 Spent Fuel Cask Loading Pit C I ANSI/AISC N690 for Manufacturers Std for Gate structural steel non-structural components FHS-MY-Y03 Permanent Cavity Seal Ring C I ANSI/AISC N690 FHS-PL-V001 Fuel transfer tube Isolation C I ASME-III-3 Valve FHS-PY-B01 Fuel Transfer Tube B I ASME III Hatch/Blind Flange Class MC Balance of system components are Class E Fire Protection System (FPS) Location: Various FPS-PL-V050 Fire Water Containment B I ASME III-2 Supply Isolation FPS-PL-V051 Fire Water Containment Test B I ASME III-2 Connection Isolation FPS-PL-V052 Fire Water Containment B I ASME III-2 Supply Isolation - Inside FPS-PY-C01 Fire Protection Containment B I ASME III, MC Penetration 3.2-26 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 7 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Fire Protection System (Continued)
FPS-PL-V441 Auxiliary Connection to CCS D NS ASME B31.1 Isolation FPS-PL-V702 FPS Containment Penetration C I ASME III-3 Thermal Relief Valve Containment Includes all FPS components F NS ASME B31.1 Seismic Analysis standpipe and fire Inside Reactor Containment Consistent with ASME hose connections with the exception of those Section III Class 3 used for containment Systems isolation, cable tray suppression, and containment spray Various Auxiliary Building F NS ASME B31.1 Seismic Analysis (Non-Radiologically Consistent with ASME Controlled) Standpipe and Section III Class 3 Non-1E Equipment Systems Penetration Room Preaction Sprinkler System piping Balance of system components are Class E, F & G Main and Startup Feedwater System (FWS) Location: Turbine Building n/a Startup Feedwater Pumps D NS Hydraulic Institute Standards n/a Valves Providing SFW D NS ANSI B16.34 AP1000 Equipment Class D Function Balance of system components are Class E Gland Seal System (GSS) Location: Turbine Building System components are Class D Generator Hydrogen and CO2 Systems (HCS) Location: Turbine Building System components are Class E Heater Drain System (HDS) Location: Turbine Building System components are Class E Hydrogen Seal Oil System (HSS) Location: Turbine Building System components are Class E Incore Instrumentation System (IIS) Location: Containment n/a In-core Instrument Thimble D II Manufacturer Std.
Assembly Tubes n/a Thimble assemblies B I Manufacturer Std.
Main Turbine and Generator Lube Oil System (LOS) Location: Turbine Building System components are Class E 3.2-27 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 8 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Mechanical Handling System (MHS) Location: Various MHS-MH-01 Containment Polar Crane D I NUREG-0554 supplemented by ASME NOG-1 MHS-MH-02 Cask Handling Crane D I NUREG-0554 supplemented by ASME NOG-1 MHS-MH-05 Equipment Hatch Hoist C I NUREG-0554 supplemented by ASME NOG-1 MHS-MH-06 Maintenance Hatch Hoist C I NUREG-0554 supplemented by ASME NOG-1 Balance of system components are Class E Main Steam System (MSS) Location: Turbine Building System components are Class E Main Turbine System (MTS) Location: Turbine Building System components are Class E Location: Containment Shield Building and Auxiliary Passive Containment Cooling System (PCS) Building PCS-JE-FE001 PCCWST Flow Through C I ASME III-3 L001A Element PCS-JE-FE002 PCCWST Flow Through C I ASME III-3 L001B Element PCS-JE-FE003 PCCWST Flow Through C I ASME III-3 L001C Element PCS-JE-FE004 PCCWST Flow Through C I ASME III-3 L001D Element PCS-MT-01 Passive Containment Cooling C I ACI 349 See subsection Water Storage Tank 6.2.2.2.3 for additional design requirements PCS-MT-02 PCS Chemical Addition Tank D II ASME VIII PCS-MT-03 Water Distribution Bucket C I Manufacturer Std. See subsection 6.2.2.2.3 for additional design requirements PCS-MT-04 Water Collection Troughs C I Manufacturer Std. See subsection 6.2.2.2.3 for additional design requirements PCS-MT-05 Passive Containment Cooling D II API 650 Ancillary Water Storage Tank PCS-MT-06 PCCWST Leak Chase D II ASME VIII Collection Pot PCS-MY-Y01 PCCWST Screen 1 C I Manufacturer Std.
PCS-MY-Y02 PCCWST Screen 2 C I Manufacturer Std.
PCS-MY-Y03 PCCWST Screen 3 C I Manufacturer Std.
PCS-MY-Y04 PCCWST Screen 4 C I Manufacturer Std.
PCS-MY-Y05 PCCWST Screen 5 C I Manufacturer Std.
PCS-PL-V001A PCCWST Isolation C I ASME III-3 3.2-28 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 9 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Passive Containment Cooling System (Continued)
PCS-PL-V001B PCCWST Isolation C I ASME III-3 PCS-PL-V001C PCCWST Isolation C I ASME III-3 PCS-MP-01A PCS Recirculation Pump D NS Hydraulic Institute Equipment Anchorage Standards is Seismic Category II PCS-MP-01B PCS Recirculation Pump D NS Hydraulic Institute Equipment Anchorage Standards is Seismic Category II PCS-PL-V002A PCCWST Series Isolation C I ASME III-3 PCS-PL-V002B PCCWST Series Isolation C I ASME III-3 PCS-PL-V002C PCCWST Series Isolation C I ASME III-3 PCS-PL-V004 Recirculation Bypass Isolation D NS ANSI B16.34 Equipment Anchorage Valve is Seismic Category II PCS-PL-V005 PCCWST Supply to FPS C I ASME III-3 Isolation PCS-PL-V009 Spent Fuel Pool Emergency C I ASME III-3 Makeup Isolation Valve PCS-PL-V010A Flow Transmitter FT001 Root C I ASME III-3 Valve PCS-PL-V010B Flow Transmitter FT001 Root C I ASME III-3 Valve PCS-PL-V011A Flow Transmitter FT002 Root C I ASME III-3 Valve PCS-PL-V011B Flow Transmitter FT002 Root C I ASME III-3 Valve PCS-PL-V012A Flow Transmitter FT003 Root C I ASME III-3 Valve PCS-PL-V012B Flow Transmitter FT003 Root C I ASME III-3 Valve PCS-PL-V013A Flow Transmitter FT004 Root C I ASME III-3 Valve PCS-PL-V013B Flow Transmitter FT004 Root C I ASME III-3 Valve PCS-PL-V014 Chemical Addition Tank D NS ANSI B16.34 Equipment Anchorage Isolation Valve is Seismic Category II PCS-PL-V015 Water Bucket Makeup Line C I ASME III-3 Drain Valve PCS-PL-V016 PCCWST Drain Isolation C I ASME III-3 Valve PCS-PL-V017 Chemical Addition Tank Vent D NS ANSI B16.34 Equipment Anchorage Isolation Valve is Seismic Category II PCS-PL-V018 Recirculation Pump Throttle D NS ANSI B16.34 Equipment Anchorage Valve is Seismic Category II PCS-PL-V019 Chemical Addition Tank Fill D NS ANSI B16.34 Equipment Anchorage Isolation Valve is Seismic Category II PCS-PL-V020 Water Bucket Makeup Line C I ASME III-3 Isolation Valve PCS-PL-V021 PCCWST TO Recirculation D NS ANSI B16.34 Equipment Anchorage Pump Suction Isolation Valve is Seismic Category II 3.2-29 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 10 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Passive Containment Cooling System (Continued)
PCS-PL-V022 Chemical Addition Tank Drain D NS ANSI B16.34 Equipment Anchorage Isolation Valve is Seismic Category II PCS-PL-V023 PCS Recirculation Return C I ASME III-3 Isolation PCS-PL-V025 Pressure Transmitter PT 031 D NS ANSI B16.34 Equipment Anchorage Root Isolation Valve is Seismic Category II PCS-PL-V026 Makeup to Distribution Bucket C I ASME III-3 Isolation Valve PCS-PL-V029 PCCWST Isolation Valve C I ASME III-3 Leakage Detection Drain PCS-PL-V030 PCCWST Isolation Valve C I ASME III-3 Leakage Detection Crossconnect Valve PCS-PL-V031A Level Transmitter LT 016 & C I ASME III-3 010 Root Isolation Valve PCS-PL-V031B Level Transmitter LT 015 & C I ASME III-3 011 Root Isolation Valve PCS-PL-V033 Recirculation Pump Suction C I ASME III-3 from Long Term Makeup Isolation Valve PCS-PL-V035A Recirculation Pump Suction D NS ANSI B16.34 Equipment Anchorage Isolation Valve is Seismic Category II PCS-PL-V035B Recirculation Pump Suction D NS ANSI B16.34 Equipment Anchorage Isolation Valve is Seismic Category II PCS-PL-V036A/B Recirculation Pump Discharge D NS ANSI B16.34 Equipment Anchorage Check Valve is Seismic Category II PCS-PL-V037 PCCAWST Discharge D NS ANSI B16.34 Equipment Anchorage Isolation Valve is Seismic Category II PCS-PL-V038 PCCAWST Drain Isolation D NS ANSI B16.34 Equipment Anchorage Valve is Seismic Category II PCS-PL-V039 PCCWST Long-Term Makeup C I ASME III-3 Check Valve PCS-PL-V040 Recirculation Pump Suction D NS ANSI B16.34 Equipment Anchorage from PCCAWST Isolation is Seismic Category II Valve PCS-PL-V041 PCCAWST Recirculation D NS ANSI B16.34 Equipment Anchorage Return Line Isolation Valve is Seismic Category II PCS-PL-V042 PCCWST Long-Term Makeup C I ASME III-3 Isolation Drain Valve PCS-PL-V043 PCCAWST Recirculation D NS ANSI B16.34 Equipment Anchorage Return Line Drain Isolation is Seismic Category II Valve PCS-PL-V044 PCCWST Long-Term Makeup C I ASME III-3 Isolation Valve PCS-PL-V045 Emergency Makeup to the C I ASME III-3 Spent Fuel Pool Isolation Valve 3.2-30 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 11 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Passive Containment Cooling System (Continued)
PCS-PL-V046 PCCWST Recirculation C I ASME III-3 Return Isolation Valve PCS-PL-V047A/B PCS Recirculation Pump D NS ANSI B16.34 Equipment Anchorage Discharge Isolation Valve is Seismic Category II PCS-PL-V048 Recirculation Pump Fire D NS ANSI B16.34 Seismically Analyzed Suction Isolation Valve for Operability PCS-PL-V049 Emergency Makeup to the C I ASME III-3 Spent Fuel Pool Drain Isolation Valve PCS-PL-V050 Spent Fuel Pool Long Term C I ASME III-3 Makeup Isolation Valve PCS-PL-V051 Spent Fuel Pool Emergency C I ASME III-3 Makeup Lower Isolation PCS-PL-V052 Spent Fuel Pool Emergency C I ASME III-3 Makeup Isolation Valve PCS-PL-V053 PCS Recirculation Heater D NS ASME VIII Pressure Relief Valve PCS-PL-V060A Shutoff Valve for Leakage C I ASME III-3 Sensor PCS-PL-V060B Shutoff Valve for Leakage C I ASME III-3 Sensor PCS-PL-V100 Temporary Containment D NS ANSI B16.34 Equipment Anchorage Washdown Isolation Valve is Seismic Category II PCS-PL-V301 PCCWST to Recirculation D NS ANSI B16.34 Equipment Anchorage Pump Suction Drain Isolation is Seismic Category II Valve PCS-PL-V303 Recirculation Header C I ASME III-3 Discharge to SFS Pool Vent Isolation Valve PCS-PL-V304 Recirculation Header C I ASME III-3 Discharge to SFS Pool Drain Isolation Valve PCS-PL-V305 PCCWST Recirculation C I ASME III-3 Return Drain Isolation Valve PCS-PY-B01 Spent Fuel Pool Emergency C I ASME III-3 Makeup Isolation PCS-PY-C01 Containment Pressure B I ASME, MC Instrument Line Penetration PCS-PY-C02 Containment Pressure B I ASME, MC Instrument Line Penetration PCS-PY-C03 Containment Pressure B I ASME, MC Instrument Line Penetration PCS-PY-C04 Containment Pressure B I ASME, MC Instrument Line Penetration Balance of system components are Class E or F Plant Gas Systems (PGS) Location: Various System components are Class E 3.2-31 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 12 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Primary Sampling System (PSS) Location: Containment and Auxiliary Building n/a Grab Sample Unit D NS Manufacturer Std.
n/a Sample Cooler, PSS and CCS D NS ASME VIII/ TEMA Side n/a Valves Providing PSS AP1000 D NS ANSI B16.34 Equipment Class D Function PSS-PL-V001A Hot Leg Sample Isolation B I ASME III-2 PSS-PL-V001B Hot Leg Sample Isolation B I ASME III-2 PSS-PL-V003 Pressurizer Sample Isolation B I ASME III-2 PSS-PL-V004A PXS Accumulator Sample C I ASME III-3 Isolation PSS-PL-V004B PXS Accumulator Sample C I ASME III-3 Isolation PSS-PL-V005A PXS CMT A Sample B I ASME III-2 Isolation PSS-PL-V005B PXS CMT B Sample B I ASME III-2 Isolation PSS-PL-V005C PXS CMT A Sample B I ASME III-2 Isolation PSS-PL-V005D PXS CMT B Sample B I ASME III-2 Isolation PSS-PL-V008 Containment Air Sample B I ASME III-2 Containment Isolation IRC PSS-PL-V010A Liquid Sample Line B I ASME III-2 Containment Isolation IRC PSS-PL-V010B Liquid Sample Line B I ASME III-2 Containment Isolation IRC PSS-PL-V011A Liquid Sample Line B I ASME III-2 Containment Isolation ORC PSS-PL-V011B Liquid Sample Line B I ASME III-2 Containment Isolation ORC PSS-PL-V012A Liquid Sample Isolation Valve C I ASME III-3 PSS-PL-V012B Liquid Sample Check Valve C I ASME III-3 PSS-PL-V013 RCS Pressurizer Sample B I ASME III-2 Isolation Valve PSS-PL-V014A RCS Hot Leg 1 Sample B I ASME III-2 Isolation Valve PSS-PL-V014B RCS Hot Leg 2 Sample B I ASME III-2 Isolation Valve PSS-PL-V015A PXS Accumulator Sample C I ASME III-3 Isolation Valve PSS-PL-V015B PXS Accumulator Sample C I ASME III-3 Isolation Valve PSS-PL-V016A PXS CMT A Sample Isolation B I ASME III-2 Valve PSS-PL-V016B PXS CMT B Sample Isolation B I ASME III-2 Valve 3.2-32 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 13 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Primary Sampling System (Continued)
PSS-PL-V016C PXS CMT A Sample Isolation B I ASME III-2 Valve PSS-PL-V016D PXS CMT B Sample Isolation B I ASME III-2 Valve PSS-PL-V023 Sample Return Line B I ASME III-2 Containment Isolation ORC PSS-PL-V024 Sample Return Containment B I ASME III-2 Isolation IRC PSS-PL-V046 Air Sample Line Containment B I ASME III-2 Isolation ORC PSS-PL-V076A Containment Testing C I ASME III-3 Boundary Isolation Valve PSS-PL-V076B Containment Testing C I ASME III-3 Boundary Isolation Valve PSS-PL-V082 Containment Isolation Test C I ASME III-3 Connection Isolation Valve PSS-PL-V083 Containment Isolation Test C I ASME III-3 Connection Isolation Valve PSS-PL-V085 Containment Isolation Test B I ASME III-2 Connection Isolation Valve PSS-PL-V086 Containment Isolation Test C I ASME III-3 Connection Isolation Valve PSS-PY-C01 Common Primary Sample B I ASME III, MC Line Penetration PSS-PY-C02 Containment Atmosphere B I ASME III, MC Sample Line Penetration PSS-PY-C03 Containment Atmosphere B I ASME III, MC Sample Line Penetration PSS-PY-C04 RCS Hot Leg Sample Line B I ASME III, MC Penetration PSS-PY-Y01 Delay Coil 1 for RCS Hot C I ASME III-3 Leg 1 PSS-PY-Y02 Delay Coil 2 for RCS Hot C I ASME III-3 Leg 2 PSS-MY-Y05 Delay Coil Assembly C I ASME III-3 Balance of system components are Class E Potable Water System (PWS) Location: Various PWS-PL-V418 PWS MCR Isolation Valve C I ASME III-3 PWS-PL-V420 PWS MCR Isolation Valve C I ASME III-3 PWS-PL-V498 PWS MCR Vacuum Relief C I ASME III-3 Balance of system components are Class E, P, and W 3.2-33 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 14 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Passive Core Cooling System (PXS) Location: Containment PXS-ME-01 Passive Residual Heat A I ASME III-1 Removal Heat Exchanger PXS-MT-01A Accumulator Tank A C I ASME III-3 PXS-MT-01B Accumulator Tank B C I ASME III-3 PXS-MT-02A Core Makeup Tank A A I ASME III-1 PXS-MT-02B Core Makeup Tank B A I ASME III-1 PXS-MT-03 In-Containment Refueling C I ACI 349/ANSI/AISC ACI 349 is used for Water Storage Tank N690 Evaluation of Structural Boundary PXS-MT-04 IRWST Gutter C I Manufacturer Std.
PXS-MW-01A Reactor Coolant C I ASME III-3 Depressurization Sparger A PXS-MW-01B Reactor Coolant C I ASME III-3 Depressurization Sparger B PXS-MY-Y01A IRWST Screen A C I Manufacturer Std. Structural frame and attachment use ASME III, Subsection NF criteria. Screen modules use manufacturer std.
PXS-MY-Y01B IRWST Screen B C I Manufacturer Std. Structural frame and attachment use ASME III, Subsection NF criteria. Screen modules use manufacturer std.
PXS-MY-Y01C IRWST Screen C C I Manufacturer Std. Structural frame and attachment use ASME III, Subsection NF criteria. Screen modules use manufacturer std.
PXS-MY-Y02A Containment Recirculation C I Manufacturer Std. Structural frame and Screen A attachment use ASME III, Subsection NF criteria. Screen modules use manufacturer std.
PXS-MY-Y02B Containment Recirculation C I Manufacturer Std. Structural frame and Screen B attachment use ASME III, Subsection NF criteria. Screen modules use manufacturer std.
PXS-MY-Y03A pH Adjustment Basket 3A C I Manufacturer Std.
PXS-MY-Y03B pH Adjustment Basket 3B C I Manufacturer Std.
PXS-MY-Y04A pH Adjustment Basket 4A C I Manufacturer Std.
PXS-MY-Y04B pH Adjustment Basket 4B C I Manufacturer Std.
PXS-MY-Y11A CMT A Upper Level Standpipe A I ASME III-1 3.2-34 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 15 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Passive Core Cooling System (Continued)
PXS-MY-Y11B CMT B Upper Level Standpipe A I ASME III-1 PXS-MY-Y12A CMT A Upper Level Standpipe A I ASME III-1 PXS-MY-Y12B CMT B Upper Level Standpipe A I ASME III-1 PXS-MY-Y13A CMT A Lower Level Standpipe A I ASME III-1 PXS-MY-Y13B CMT B Lower Level Standpipe A I ASME III-1 PXS-MY-Y14A CMT A Lower Level Standpipe A I ASME III-1 PXS-MY-Y14B CMT B Lower Level Standpipe A I ASME III-1 PXS-MY-Y21 IRWST Hood Vent Cover C I Manufacturer Std.
PXS-MY-Y22 IRWST Hood Vent Cover C I Manufacturer Std.
PXS-MY-Y23 IRWST Hood Vent Cover C I Manufacturer Std.
PXS-MY-Y24 IRWST Hood Vent Cover C I Manufacturer Std.
PXS-MY-Y25 IRWST Hood Vent Cover C I Manufacturer Std.
PXS-MY-Y26 IRWST Hood Vent Cover C I Manufacturer Std.
PXS-MY-Y27 IRWST Hood Vent Cover C I Manufacturer Std.
PXS-MY-Y28 IRWST Hood Vent Cover C I Manufacturer Std.
PXS-MY-Y29 IRWST Hood Vent Cover C I Manufacturer Std.
PXS-MY-Y30 IRWST Hood Vent Cover C I Manufacturer Std.
PXS-MY-Y31 IRWST Hood Vent Cover C I Manufacturer Std.
PXS-MY-Y32 IRWST Hood Vent Cover C I Manufacturer Std.
PXS-MY-Y33 IRWST Hood Vent Cover C I Manufacturer Std.
PXS-MY-Y41 IRWST Hood Vent Cover C I Manufacturer Std.
PXS-MY-Y47 IRWST Hood Vent Cover C I Manufacturer Std.
PXS-MY-Y48 IRWST Hood Vent Cover C I Manufacturer Std.
PXS-MY-Y61 IRWST SG Wall Vent Cover C I Manufacturer Std.
PXS-MY-Y62 IRWST SG Wall Vent Cover C I Manufacturer Std.
PXS-MY-Y63 IRWST SG Wall Vent Cover C I Manufacturer Std.
PXS-MY-Y64 IRWST SG Wall Vent Cover C I Manufacturer Std.
PXS-MY-Y71 IRWST Overflow Weir Cover C I Manufacturer Std.
PXS-MY-Y72 IRWST Overflow Weir Cover C I Manufacturer Std.
PXS-MY-Y73 IRWST Overflow Weir Cover C I Manufacturer Std.
PXS-MY-Y74 IRWST Overflow Weir Cover C I Manufacturer Std.
PXS-MY-Y75 IRWST Overflow Weir Cover C I Manufacturer Std.
PXS-MY-Y76 IRWST Overflow Weir Cover C I Manufacturer Std.
PXS-MY-Y81 Downspout Screen 1A C I Manufacturer Std.
PXS-MY-Y82 Downspout Screen 1B C I Manufacturer Std.
PXS-MY-Y83 Downspout Screen 1C C I Manufacturer Std.
PXS-MY-Y84 Downspout Screen 1D C I Manufacturer Std.
PXS-MY-Y85 Downspout Screen 2A C I Manufacturer Std.
PXS-MY-Y86 Downspout Screen 2B C I Manufacturer Std.
PXS-MY-Y87 Downspout Screen 2C C I Manufacturer Std.
PXS-MY-Y88 Downspout Screen 2D C I Manufacturer Std.
PXS-PL-V002A CMT A CL Inlet Isolation A I ASME III-1 PXS-PL-V002B CMT B CL Inlet Isolation A I ASME III-1 PXS-PL-V010A CMT A Upper Sample B I ASME III-2 3.2-35 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 16 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Passive Core Cooling System (Continued)
PXS-PL-V010B CMT B Upper Sample B I ASME III-2 PXS-PL-V011A CMT A Lower Sample B I ASME III-2 PXS-PL-V011B CMT B Lower Sample B I ASME III-2 PXS-PL-V012A CMT A Drain A I ASME III-1 PXS-PL-V012B CMT B Drain A I ASME III-1 PXS-PL-V013A CMT A Discharge Manual A I ASME III-1 Isolation PXS-PL-V013B CMT B Discharge Manual A I ASME III-1 Isolation PXS-PL-V014A CMT A Discharge Isolation A I ASME III-1 PXS-PL-V014B CMT B Discharge Isolation A I ASME III-1 PXS-PL-V015A CMT A Discharge Isolation A I ASME III-1 PXS-PL-V015B CMT B Discharge Isolation A I ASME III-1 PXS-PL-V016A CMT A Discharge Check A I ASME III-1 PXS-PL-V016B CMT B Discharge Check A I ASME III-1 PXS-PL-V017A CMT A Discharge Check A I ASME III-1 PXS-PL-V017B CMT B Discharge Check A I ASME III-1 PXS-PL-V019A RNS to CMT Injection Line A B I ASME III-2 Drain PXS-PL-V019B RNS to CMT Injection Line B B I ASME III-2 Drain PXS-PL-V020A IRWST Injection Line A Drain B I ASME III-2 PXS-PL-V020B IRWST Injection Line B Drain B I ASME III-2 PXS-PL-V021A Accumulator A Nitrogen Vent C I ASME III-3 PXS-PL-V021B Accumulator B Nitrogen Vent C I ASME III-3 PXS-PL-V022A Accumulator A Pressure C I ASME III-3 Relief PXS-PL-V022B Accumulator B Pressure C I ASME III-3 Relief PXS-PL-V023A Accumulator A Pressure C I ASME III-3 Transmitter B Isolation PXS-PL-V023B Accumulator B Pressure C I ASME III-3 Transmitter B Isolation PXS-PL-V024A Accumulator A Pressure C I ASME III-3 Transmitter A Isolation PXS-PL-V024B Accumulator B Pressure C I ASME III-3 Transmitter A Isolation PXS-PL-V025A Accumulator A Sample C I ASME III-3 PXS-PL-V025B Accumulator B Sample C I ASME III-3 PXS-PL-V026A Accumulator A Drain C I ASME III-3 PXS-PL-V026B Accumulator B Drain C I ASME III-3 PXS-PL-V027A Accumulator A Discharge C I ASME III-3 Isolation PXS-PL-V027B Accumulator B Discharge C I ASME III-3 Isolation 3.2-36 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 17 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Passive Core Cooling System (Continued)
PXS-PL-V028A Accumulator A Discharge A I ASME III-1 Check PXS-PL-V028B Accumulator B Discharge A I ASME III-1 Check PXS-PL-V029A Accumulator A Discharge A I ASME III-1 Check PXS-PL-V029B Accumulator B Discharge A I ASME III-1 Check PXS-PL-V030A CMT A Highpoint Vent B I ASME III-2 PXS-PL-V030B CMT B Highpoint Vent B I ASME III-2 PXS-PL-V031A CMT A Highpoint Vent B I ASME III-2 PXS-PL-V031B CMT B Highpoint Vent B I ASME III-2 PXS-PL-V033A Accumulator A Check Valve B I ASME III-2 Drain PXS-PL-V033B Accumulator B Check Valve B I ASME III-2 Drain PXS-PL-V042 Nitrogen Supply Containment B I ASME III-2 Isolation ORC PXS-PL-V043 Nitrogen Supply Containment B I ASME III-2 Isolation IRC Check Valve PXS-PL-V052 Accumulator Nitrogen B I ASME III-2 Containment Penetration TC PXS-PL-V080A CMT A WR Level Isolation B I ASME III-2 PXS-PL-V080B CMT B WR Level Isolation B I ASME III-2 PXS-PL-V081A CMT A WR Level Isolation B I ASME III-2 PXS-PL-V081B CMT B WR Level Isolation B I ASME III-2 PXS-PL-V082A CMT A Upper Level A A I ASME III-1 Isolation 1 PXS-PL-V082B CMT B Upper Level A A I ASME III-1 Isolation 1 PXS-PL-V083A CMT A Upper Level A A I ASME III-1 Isolation 2 PXS-PL-V083B CMT B Upper Level A A I ASME III-1 Isolation 2 PXS-PL-V084A CMT A Upper Level A Vent B I ASME III-2 PXS-PL-V084B CMT B Upper Level A Vent B I ASME III-2 PXS-PL-V085A CMT A Upper Level A Drain B I ASME III-2 PXS-PL-V085B CMT B Upper Level A Drain B I ASME III-2 PXS-PL-V086A CMT A Upper Level B A I ASME III-1 Isolation 1 PXS-PL-V086B CMT B Upper Level B A I ASME III-1 Isolation 1 PXS-PL-V087A CMT A Upper Level B A I ASME III-1 Isolation 2 PXS-PL-V087B CMT B Upper Level B A I ASME III-1 Isolation 2 PXS-PL-V088A CMT A Upper Level B Vent B I ASME III-2 3.2-37 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 18 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Passive Core Cooling System (Continued)
PXS-PL-V088B CMT B Upper Level B Vent B I ASME III-2 PXS-PL-V089A CMT A Upper Level B Drain B I ASME III-2 PXS-PL-V089B CMT B Upper Level B Drain B I ASME III-2 PXS-PL-V092A CMT A Lower Level A A I ASME III-1 Isolation 1 PXS-PL-V092B CMT B Lower Level A A I ASME III-1 Isolation 1 PXS-PL-V093A CMT A Lower Level A A I ASME III-1 Isolation 2 PXS-PL-V093B CMT B Lower Level A A I ASME III-1 Isolation 2 PXS-PL-V094A CMT A Lower Level A Vent B I ASME III-2 PXS-PL-V094B CMT B Lower Level A Vent B I ASME III-2 PXS-PL-V095A CMT A Lower Level A Drain B I ASME III-2 PXS-PL-V095B CMT B Lower Level A Drain B I ASME III-2 PXS-PL-V096A CMT A Lower Level B A I ASME III-1 Isolation 1 PXS-PL-V096B CMT B Lower Level B A I ASME III-1 Isolation 1 PXS-PL-V097A CMT A Lower Level B A I ASME III-1 Isolation 2 PXS-PL-V097B CMT B Lower Level B A I ASME III-1 Isolation 2 PXS-PL-V098A CMT A Lower Level B Vent B I ASME III-2 PXS-PL-V098B CMT B Lower Level B Vent B I ASME III-2 PXS-PL-V099A CMT A Lower Level B Drain B I ASME III-2 PXS-PL-V099B CMT B Lower Level B Drain B I ASME III-2 PXS-PL-V101 PRHR HX Inlet Isolation A I ASME III-1 PXS-PL-V102A PRHR HX Inlet Head Vent B I ASME III-2 PXS-PL-V102B PRHR HX Inlet Head Drain B I ASME III-2 PXS-PL-V103A PRHR HX Outlet Head Vent B I ASME III-2 PXS-PL-V103B PRHR HX Outlet Head Drain B I ASME III-2 PXS-PL-V104A PRHR HX Flow Transmitter A B I ASME III-2 Isolation PXS-PL-V104B PRHR HX Flow Transmitter B B I ASME III-2 Isolation PXS-PL-V105A PRHR HX Flow Transmitter A B I ASME III-2 Isolation PXS-PL-V105B PRHR HX Flow Transmitter B B I ASME III-2 Isolation PXS-PL-V106 Containment Recirculation A C I ASME III-3 Highpoint Vent PXS-PL-V107 Containment Recirculation A C I ASME III-3 Highpoint Vent PXS-PL-V108A PRHR HX Control A I ASME III-1 PXS-PL-V108B PRHR HX Control A I ASME III-1 3.2-38 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 19 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Passive Core Cooling System (Continued)
PXS-PL-V109 PRHR HX/RCS Return A I ASME III-1 Isolation PXS-PL-V111A PRHR HX Highpoint Vent B I ASME III-2 PXS-PL-V111B PRHR HX Highpoint Vent B I ASME III-2 PXS-PL-V113 PRHR HX Pressure B I ASME III-2 Transmitter Isolation PXS-PL-V115A Containment Recirculation A C I ASME III-3 Drain PXS-PL-V115B Containment Recirculation B C I ASME III-3 Drain PXS-PL-V116A Containment Recirculation A C I ASME III-3 Drain PXS-PL-V116B Containment Recirculation B C I ASME III-3 Drain PXS-PL-V117A Containment Recirculation A C I ASME III-3 Isolation PXS-PL-V117B Containment Recirculation B C I ASME III-3 Isolation PXS-PL-V118A Containment Recirculation A C I ASME III-3 Isolation PXS-PL-V118B Containment Recirculation B C I ASME III-3 Isolation PXS-PL-V119A Containment Recirculation A C I ASME III-3 Check PXS-PL-V119B Containment Recirculation B C I ASME III-3 Check PXS-PL-V120A Containment Recirculation A C I ASME III-3 Isolation PXS-PL-V120B Containment Recirculation B C I ASME III-3 Isolation PXS-PL-V121A IRWST Line A Isolation C I ASME III-3 PXS-PL-V121B IRWST Line B Isolation C I ASME III-3 PXS-PL-V122A IRWST Injection A Check A I ASME III-1 PXS-PL-V122B IRWST Injection B Check A I ASME III-1 PXS-PL-V123A IRWST Injection A Isolation A I ASME III-1 PXS-PL-V123B IRWST Injection B Isolation A I ASME III-1 PXS-PL-V124A IRWST Injection A Check A I ASME III-1 PXS-PL-V124B IRWST Injection B Check A I ASME III-1 PXS-PL-V125A IRWST Injection A Isolation A I ASME III-1 PXS-PL-V125B IRWST Injection B Isolation A I ASME III-1 PXS-PL-V126A IRWST Injection Check Test C I ASME III-3 PXS-PL-V126B IRWST Injection Check Test C I ASME III-3 PXS-PL-V127 IRWST Injection Line A Drain C I ASME III-3 PXS-PL-V128A IRWST Injection Check Test A I ASME III-1 PXS-PL-V128B IRWST Injection Check Test A I ASME III-1 PXS-PL-V129A IRWST Injection Check Test A I ASME III-1 3.2-39 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 20 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Passive Core Cooling System (Continued)
PXS-PL-V129B IRWST Injection Check Test A I ASME III-1 PXS-PL-V130A IRWST Gutter Bypass A C I ASME III-3 Isolation PXS-PL-V130B IRWST Gutter Bypass B C I ASME III-3 Isolation PXS-PL-V131A IRWST Injection Line A Drain B I ASME III-2 PXS-PL-V131B IRWST Injection Line B Drain B I ASME III-2 PXS-PL-V132A IRWST Injection Line A Drain B I ASME III-2 PXS-PL-V132B IRWST Injection Line B Drain B I ASME III-2 PXS-PL-V133A IRWST Injection Line A B I ASME III-2 Highpoint Vent PXS-PL-V133B IRWST Injection Line B B I ASME III-2 Highpoint Vent PXS-PL-V134A IRWST Injection Line A B I ASME III-2 Highpoint Vent PXS-PL-V134B IRWST Injection Line B B I ASME III-2 Highpoint Vent PXS-PL-V135A IRWST Injection Line A B I ASME III-2 Highpoint Vent Isolation PXS-PL-V135B IRWST Injection Line B B I ASME III-2 Highpoint Vent Isolation PXS-PL-V149 RNS Suction Pump Line Drain C I ASME III-3 PXS-PL-V151B IRWST Wide Range Level C I ASME III-3 Transmitter B Isolation PXS-PL-V151C IRWST Wide Range Level C I ASME III-3 Transmitter C Isolation PXS-PL-V151D IRWST Wide Range Level C I ASME III-3 Transmitter D Isolation PXS-PL-V161A IRWST Lower Narrow Range C I ASME III-3 Level Transmitter A Isolation PXS-PL-V161B IRWST Lower Narrow Range C I ASME III-3 Level Transmitter B Isolation PXS-PL-V161C IRWST Lower Narrow Range C I ASME III-3 Level Transmitter C Isolation PXS-PL-V161D IRWST Lower Narrow Range C I ASME III-3 Level Transmitter D Isolation PXS-PL-V162A IRWST Lower Narrow Range C I ASME III-3 Level Transmitter A Reference Leg Isolation PXS-PL-V162B IRWST Lower Narrow Range C I ASME III-3 Level Transmitter B Reference Leg Isolation PXS-PL-V162C IRWST Lower Narrow Range C I ASME III-3 Level Transmitter C Reference Leg Isolation 3.2-40 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 21 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Passive Core Cooling System (Continued)
PXS-PL-V162D IRWST Lower Narrow Range C I ASME III-3 Level Transmitter D Reference Leg Isolation PXS-PL-V170A PRHR Flow Transmitter A B I ASME III-2 Vent PXS-PL-V170B PRHR Flow Transmitter B B I ASME III-2 Vent PXS-PL-V171A PRHR Flow Transmitter A B I ASME III-2 Vent PXS-PL-V171B PRHR Flow Transmitter B B I ASME III-2 Vent PXS-PL-V201A Accumulator A Leak Test B I ASME III-2 PXS-PL-V201B Accumulator B Leak Test B I ASME III-2 PXS-PL-V202A Accumulator A Leak Test C I ASME III-3 PXS-PL-V202B Accumulator B Leak Test C I ASME III-3 PXS-PL-V205A RNS Discharge Leak Test B I ASME III-2 PXS-PL-V205B RNS Discharge Leak Test B I ASME III-2 PXS-PL-V206 RNS Discharge Leak Test C I ASME III-3 PXS-PL-V207A RNS Suction Leak Test B I ASME III-2 PXS-PL-V207B RNS Suction Leak Test B I ASME III-2 PXS-PL-V208A RNS Suction Leak Test B I ASME III-2 PXS-PL-V217 PXS Leak Test Line Isolation D NS ASME B31.1 PXS-PL-V221 Test Header to IRWST D NS ASME B31.1 PXS-PL-V230A CMT A Fill Isolation B I ASME III-2 PXS-PL-V230B CMT B Fill Isolation B I ASME III-2 PXS-PL-V231A CMT A Fill Check B I ASME III-2 PXS-PL-V231B CMT B Fill Check B I ASME III-2 PXS-PL-V232A Accumulator A Fill/Drain C I ASME III-3 Isolation PXS-PL-V232B Accumulator B Fill/Drain C I ASME III-3 Isolation PXS-PL-V250A CMT A Check Valve Test A I ASME III-1 Valve PXS-PL-V250B CMT B Check Valve Test A I ASME III-1 Valve PXS-PL-V251A CMT A Check Valve Test A I ASME III-1 Valve PXS-PL-V251B CMT B Check Valve Test A I ASME III-1 Valve PXS-PL-V252A CMT A Check Valve Test A I ASME III-1 Valve PXS-PL-V252B CMT B Check Valve Test A I ASME III-1 Valve PXS-PY-C01 Nitrogen Makeup B I ASME III, MC Containment Penetration PXS-PY-E01 DVI-B Inline Expansion Joint C I ASME III-3 3.2-41 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 22 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Passive Core Cooling System (Continued)
PXS-PY-R01A Core Makeup Tank A Orifice A I ASME III-1 PXS-PY-R01B Core Makeup Tank B Orifice A I ASME III-1 PXS-PY-R02A Accumulator Tank A Orifice C I ASME III-3 PXS-PY-R02B Accumulator Tank B Orifice C I ASME III-3 Balance of system components are Class E Reactor Coolant System (RCS) Location: Containment RCS-MB-01 Steam Generator 1 A I ASME III-1 n/a SG 1 Shell B I ASME III-1 n/a SG 1 Channel Head Divider B I ASME III-1 Plate n/a SG 1 Tube Bundle Support C I Manufacturer Std Assembly n/a SG 1 Steam Flow Limiting B I ASME III-1 See Subsections Venturi 5.4.4.2 and 5.4.4.3 for additional requirements.
n/a SG 1 Feedwater Distribution C I Manufacturer Std Ring Supports RCS-MB-02 Steam Generator 2 A I ASME III-1 n/a SG 2 Shell B I ASME III-1 n/a SG 2 Channel Head Divider B I ASME III-1 Plate n/a SG 2 Tube Bundle Support C I Manufacturer Std Assembly n/a SG 2 Steam Flow Limiting B I ASME III-1 See Subsections Venturi 5.4.4.2 and 5.4.4.3 for additional requirements.
n/a SG 2 Feedwater Distribution C I Manufacturer Std Ring Supports RCS-MP-01A/B SG 1A(B) Reactor Coolant A I ASME III-1 Pump Motor - Class D Pump n/a Rotor Shaft C I Manufacturer Std n/a Impeller C I Manufacturer Std n/a Flywheel C I Manufacturer Std n/a RCP Heat Exchanger (Tube A I ASME III-1 Shellside - Class D, Side) ASME VIII, Div. 1 n/a Pump Motor Cooling Water to A I ASME III-1 HX Inlet Connector n/a Pump Motor Cooling Water A I ASME III-1 from HX Outlet Connector RCS-MP-02A/B SG 2A(B) Reactor Coolant A I ASME III-1 Pump Motor - Class D Pump n/a Rotor Shaft C I Manufacturer Std n/a Impeller C I Manufacturer Std n/a Flywheel C I Manufacturer Std 3.2-42 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 23 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Reactor Coolant System (Continued) n/a RCP Heat Exchanger (Tube A I ASME III-1 Shellside - Class D, Side) ASME VIII, Div. 1 n/a Pump Motor Cooling Water to A I ASME III-1 HX Inlet Connector n/a Pump Motor Cooling Water A I ASME III-1 from HX Outlet Connector RCS-MV-01 Reactor Vessel A I ASME III-1 The reactor vessel has two tag numbers; RCS-MV-01 for the reactor coolant system (RCS) and RXS-MV-01 for the reactor system (RXS)
RCS-MV-02 Pressurizer A I ASME III-1 RCS-PL-V001A First Stage ADS A I ASME III-1 RCS-PL-V001B First Stage ADS A I ASME III-1 RCS-PL-V002A Second Stage ADS A I ASME III-1 RCS-PL-V002B Second Stage ADS A I ASME III-1 RCS-PL-V003A Third Stage ADS A I ASME III-1 RCS-PL-V003B Third Stage ADS A I ASME III-1 RCS-PL-V004A Fourth Stage ADS A I ASME III-1 RCS-PL-V004B Fourth Stage ADS A I ASME III-1 RCS-PL-V004C Fourth Stage ADS A I ASME III-1 RCS-PL-V004D Fourth Stage ADS A I ASME III-1 RCS-PL-V005A Pressurizer Safety Valve A I ASME III-1 RCS-PL-V005B Pressurizer Safety Valve A I ASME III-1 RCS-PL-V007A ADS Test Valve B I ASME III-2 RCS-PL-V007B ADS Test Valve B I ASME III-2 RCS-PL-V007C ADS Test Valve B I ASME III-2 RCS-PL-V008 ADS Valve Leakage Check C I ASME III-3 Valve RCS-PL-V010A ADS Discharge Header A C I ASME III-3 Vacuum Relief RCS-PL-V010B ADS Discharge Header B C I ASME III-3 Vacuum Relief RCS-PL-V011A First Stage ADS Isolation A I ASME III-1 RCS-PL-V011B First Stage ADS Isolation A I ASME III-1 RCS-PL-V012A Second Stage ADS Isolation A I ASME III-1 RCS-PL-V012B Second Stage ADS Isolation A I ASME III-1 RCS-PL-V013A Third Stage ADS Isolation A I ASME III-1 RCS-PL-V013B Third Stage ADS Isolation A I ASME III-1 RCS-PL-V014A Fourth Stage ADS Isolation A I ASME III-1 RCS-PL-V014B Fourth Stage ADS Isolation A I ASME III-1 RCS-PL-V014C Fourth Stage ADS Isolation A I ASME III-1 RCS-PL-V014D Fourth Stage ADS Isolation A I ASME III-1 3.2-43 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 24 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Reactor Coolant System (Continued)
RCS-PL-V095 Hot Leg 2 Level Instrument B I ASME III-2 Root RCS-PL-V096 Hot Leg 2 Level Instrument B I ASME III-2 Root RCS-PL-V097 Hot Leg 1 Level Instrument B I ASME III-2 Root RCS-PL-V098 Hot Leg 1 Level Instrument B I ASME III-2 Root RCS-PL-V101A Hot Leg 1 Flow Instrument B I ASME III-2 Root RCS-PL-V101B Hot Leg 1 Flow Instrument B I ASME III-2 Root RCS-PL-V101C Hot Leg 1 Flow Instrument B I ASME III-2 Root RCS-PL-V101D Hot Leg 1 Flow Instrument B I ASME III-2 Root RCS-PL-V101E Hot Leg 1 Flow Instrument B I ASME III-2 Root RCS-PL-V101F Hot Leg 1 Flow Instrument B I ASME III-2 Root RCS-PL-V102A Hot Leg 2 Flow Instrument B I ASME III-2 Root RCS-PL-V102B Hot Leg 2 Flow Instrument B I ASME III-2 Root RCS-PL-V102C Hot Leg 2 Flow Instrument B I ASME III-2 Root RCS-PL-V102D Hot Leg 2 Flow Instrument B I ASME III-2 Root RCS-PL-V102E Hot Leg 2 Flow Instrument B I ASME III-2 Root RCS-PL-V102F Hot Leg 2 Flow Instrument B I ASME III-2 Root RCS-PL-V103 PRHR HX Outlet Line Drain B I ASME III-2 RCS-PL-V108A Hot Leg 1 Sample Isolation B I ASME III-2 RCS-PL-V108B Hot Leg 2 Sample Isolation B I ASME III-2 RCS-PL-V110A Pressurizer Spray Valve A I ASME III-1 RCS-PL-V110B Pressurizer Spray Valve A I ASME III-1 RCS-PL-V111A Pressurizer Spray Block A I ASME III-1 Valve RCS-PL-V111B Pressurizer Spray Block A I ASME III-1 Valve RCS-PL-V120 Reactor Vessel Flange D NS ASME B31.1 Leakoff RCS-PL-V121 Reactor Vessel Flange D NS ASME B31.1 Leakoff RCS-PL-V122A Reactor Vessel Flange D NS ASME B31.1 Leakoff 3.2-44 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 25 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Reactor Coolant System (Continued)
RCS-PL-V122B Reactor Vessel Flange D NS ASME B31.1 Leakoff RCS-PL-V150A Reactor Vessel Head Vent A I ASME III-1 RCS-PL-V150B Reactor Vessel Head Vent A I ASME III-1 RCS-PL-V150C Reactor Vessel Head Vent A I ASME III-1 RCS-PL-V150D Reactor Vessel Head Vent A I ASME III-1 RCS-PL-V171A Cold Leg 1A Bend Instrument B I ASME III-2 Root RCS-PL-V171B Cold Leg 1A Bend Instrument B I ASME III-2 Root RCS-PL-V172A Cold Leg 1B Bend Instrument B I ASME III-2 Root RCS-PL-V172B Cold Leg 1B Bend Instrument B I ASME III-2 Root RCS-PL-V173A Cold Leg 2A Bend Instrument B I ASME III-2 Root RCS-PL-V173B Cold Leg 2A Bend Instrument B I ASME III-2 Root RCS-PL-V174A Cold Leg 2B Bend Instrument B I ASME III-2 Root RCS-PL-V174B Cold Leg 2B Bend Instrument B I ASME III-2 Root RCS-PL-V204 Pressurizer Manual Vent A I ASME III-1 RCS-PL-V205 Pressurizer Manual Vent A I ASME III-1 RCS-PL-V210A Pressurizer Spray Bypass B I ASME III-2 RCS-PL-V210B Pressurizer Spray Bypass B I ASME III-2 RCS-PL-V015 Pressurizer Vent to RCDT D NS ANSI B16.34 Test Valve RCS-PL-V225A Pressurizer Level Steam B I ASME III-2 Space Instrument Root RCS-PL-V225B Pressurizer Level Steam B I ASME III-2 Space Instrument Root RCS-PL-V225C Pressurizer Level Steam B I ASME III-2 Space Instrument Root RCS-PL-V225D Pressurizer Level Steam B I ASME III-2 Space Instrument Root RCS-PL-V226A Pressurizer Level Liquid B I ASME III-2 Space Instrument Root RCS-PL-V226B Pressurizer Level Liquid B I ASME III-2 Space Instrument Root RCS-PL-V226C Pressurizer Level Liquid B I ASME III-2 Space Instrument Root RCS-PL-V226D Pressurizer Level Liquid B I ASME III-2 Space Instrument Root RCS-PL-V227A Pressurizer Level Reference B I ASME III-2 Leg Tubing Instrumentation Root 3.2-45 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 26 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Reactor Coolant System (Continued)
RCS-PL-V227B Pressurizer Level Reference B I ASME III-2 Leg Tubing Instrumentation Root RCS-PL-V227C Pressurizer Level Reference B I ASME III-2 Leg Tubing Instrumentation Root RCS-PL-V227D Pressurizer Level Reference B I ASME III-2 Leg Tubing Instrumentation Root RCS-PL-V228 Wide Range Pressurizer Level B I ASME III-2 Steam Space Instrument Root RCS-PL-V229 Wide Range Pressurizer Level B I ASME III-2 Liquid Space Instrument Root RCS-PL-V232 Manual Head Vent C I ASME III-3 RCS-PL-V233 Head Vent Isolation C I ASME III-3 RCS-PL-V241 ADS Valve Discharge Header C I ASME III-3 Drain Isolation RCS-PL-V242 ADS Valve Discharge Header D NS ANSI B16.34 Drain Check RCS-PL-V250 ADS Discharge Line Isolation C I ASME III-3 RCS-PL-V260A RCP 1A Vent A I ASME III-1 RCS-PL-V260B RCP 1B Vent A I ASME III-1 RCS-PL-V260C RCP 2A Vent A I ASME III-1 RCS-PL-V260D RCP 2B Vent A I ASME III-1 RCS-PL-V261A RCP 1A Drain A I ASME III-1 RCS-PL-V261B RCP 1B Drain A I ASME III-1 RCS-PL-V261C RCP 2A Drain A I ASME III-1 RCS-PL-V261D RCP 2B Drain A I ASME III-1 RCS-PY-B01 ADS Hydrostatic Test C I ASME III-3 Spectacle Flange RCS-PY-B02 ADS Hydrostatic Test C I ASME III-3 Spectacle Flange RCS-PY-B03 RCS Vacuum Pump Suction C I ASME III-3 Spectacle Blind RCS-PY-B04 RCS Vacuum Ejector C I ASME III-3 Package Suction Spectacle Blind RCS-PY-K03 Safety Valve Discharge C I ASME III-3 Chamber Rupture Disk RCS-PY-K04 Safety Valve Discharge C I ASME III-3 Chamber Rupture Disk RCS-PY-R01A Reactor Vessel Head Vent C I ASME III-3 Flow Orifice A RCS-PY-R01B Reactor Vessel Head Vent C I ASME III-3 Flow Orifice B RCS-PY-Y01A Pressurizer Level Reference B I ASME III-2 Leg L225A Flex Hose 3.2-46 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 27 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Reactor Coolant System (Continued)
RCS-PY-Y01B Pressurizer Level Reference B I ASME III-2 Leg L225B Flex Hose RCS-PY-Y01C Pressurizer Level Reference B I ASME III-2 Leg L225C Flex Hose RCS-PY-Y01D Pressurizer Level Reference B I ASME III-2 Leg L225D Flex Hose Gravity and Roof Drain Collection System (RDS) Location: Various System components are Class E Normal Residual Heat Removal System (RNS) Location: Containment and Auxiliary Building RNS-JE-FE001A RNS Train A Discharge Flow C I ASME III-3 Element RNS-JE-FE001B RNS Train B Discharge Flow C I ASME III-3 Element RNS-ME-01A Normal Residual Heat C I ASME III-3 Shellside - Class D Removal Heat Exchanger A ASME VIII, Div. 1 (Tube Side)
RNS-ME-01B Normal Residual Heat C I ASME III-3 Shellside - Class D Removal Heat Exchanger B ASME VIII, Div. 1 (Tube Side)
RNS-MP-01A Residual Heat Removal C I ASME III-3 Pump Motor - Class D Pump A RNS-MP-01B Residual Heat Removal C I ASME III-3 Pump Motor - Class D Pump B RNS-PL-V001A RNS HL Suction Isolation - A I ASME III-1 Inner RNS-PL-V001B RNS HL Suction Isolation - A I ASME III-1 Inner RNS-PL-V002A RNS HL Suction and A I ASME III-1 Containment Isolation - Outer RNS-PL-V002B RNS HL Suction and A I ASME III-1 Containment Isolation - Outer RNS-PL-V003A RCS Pressure Boundary B I ASME III-2 Valve Thermal Relief RNS-PL-V003B RCS Pressure Boundary B I ASME III-2 Valve Thermal Relief RNS-PL-V004A RCS Pressure Boundary B I ASME III-2 Valve Thermal Relief Isolation RNS-PL-V004B RCS Pressure Boundary B I ASME III-2 Valve Thermal Relief Isolation RNS-PL-V005A RNS Pump A Suction C I ASME III-3 Isolation RNS-PL-V005B RNS Pump B Suction C I ASME III-3 Isolation RNS-PL-V006A RNS HX A Outlet Flow Control C I ASME III-3 RNS-PL-V006B RNS HX B Outlet Flow Control C I ASME III-3 RNS-PL-V007A RNS Pump A Discharge C I ASME III-3 Isolation 3.2-47 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 28 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Normal Residual Heat Removal System (Continued)
RNS-PL-V007B RNS Pump B Discharge C I ASME III-3 Isolation RNS-PL-V008A RNS HX A Bypass Flow C I ASME III-3 Control RNS-PL-V008B RNS HX B Bypass Flow C I ASME III-3 Control RNS-PL-V010 RNS Discharge Containment C I ASME III-3 Isolation Valve Test RNS-PL-V011 RNS Discharge Containment B I ASME III-2 Isolation Valve - ORC RNS-PL-V012 RNS Discharge Containment B I ASME III-2 Isolation Valve Test Connection ORC RNS-PL-V013 RNS Discharge Containment B I ASME III-2 Isolation - IRC RNS-PL-V014 RNS Discharge Containment C I ASME III-3 Isolation Valve Test Connection RNS-PL-V015A RNS Discharge RCS A I ASME III-1 Pressure Boundary RNS-PL-V015B RNS Discharge RCS A I ASME III-1 Pressure Boundary RNS-PL-V016 RNS Discharge Containment C I ASME III-3 Penetration Isolation Valves Test RNS-PL-V017A RNS Discharge RCS A I ASME III-1 Pressure Boundary RNS-PL-V017B RNS Discharge RCS A I ASME III-1 Pressure Boundary RNS-PL-V020 RNS HL Suction Pressure B I ASME III-2 Relief RNS-PL-V021 RNS HL Suction Pressure B I ASME III-2 Relief RNS-PL-V022 RNS Suction Header B I ASME III-2 Containment Isolation - ORC RNS-PL-V023 RNS Suction from IRWST - B I ASME III-2 Containment Isolation RNS-PL-V024 RNS Discharge to IRWST C I ASME III-3 Isolation RNS-PL-V025 RNS Suction from IRWST - C I ASME III-3 Bonnet Relief Isolation RNS-PL-V026 RNS Suction from IRWST - C I ASME III-3 Containment Isolation Test RNS-PL-V029 RNS Discharge to CVS C I ASME III-3 RNS-PL-V030A RNS HX A Shell Drain D NS ASME B31.1 RNS-PL-V030B RNS HX B Shell Drain D NS ASME B31.1 RNS-PL-V031A RNS Train A Discharge Flow C I ASME III-3 Instrument Isolation 3.2-48 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 29 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Normal Residual Heat Removal System (Continued)
RNS-PL-V031B RNS Train B Discharge Flow C I ASME III-3 Instrument Isolation RNS-PL-V032A RNS Train A Discharge Flow C I ASME III-3 Instrument Isolation RNS-PL-V032B RNS Train B Discharge Flow C I ASME III-3 Instrument Isolation RNS-PL-V033A RNS Pump A Suction C I ASME III-3 Pressure Instrument Isolation RNS-PL-V033B RNS Pump B Suction C I ASME III-3 Pressure Instrument Isolation RNS-PL-V034A RNS Pump A Discharge C I ASME III-3 Pressure Instrument Isolation RNS-PL-V034B RNS Pump B Discharge C I ASME III-3 Pressure Instrument Isolation RNS-PL-V035A RNS HX A Shell Vent D NS ANSI B16.34 RNS-PL-V035B RNS HX B Shell Vent D NS ANSI B16.34 RNS-PL-V036A RNS Pump A Suction Piping C I ASME III-3 Drain. Isolation RNS-PL-V036B RNS Pump B Suction Piping C I ASME III-3 Drain. Isolation RNS-PL-V045 RNS Pump Discharge Relief C I ASME III-3 RNS-PL-V048A RNS Pump Seal Cooler A C I ASME III-3 Vent Isolation RNS-PL-V048B RNS Pump Seal Cooler B C I ASME III-3 Vent Isolation RNS-PL-V049A RNS Pump Seal Cooler A C I ASME III-3 Drain Isolation RNS-PL-V049B RNS Pump Seal Cooler B C I ASME III-3 Drain Isolation RNS-PL-V050 RNS Pump A Casing Drain. C I ASME III-3 Isolation RNS-PL-V051 RNS Pump B Casing Drain. C I ASME III-3 Isolation RNS-PL-V052 RNS Pump Suction From C I ASME III-3 Spent Fuel Pool Isolation RNS-PL-V053 RNS Pump Discharge to C I ASME III-3 Spent Fuel Pool Isolation RNS-PL-V055 RNS Pump Suction to Cask C I ASME III-3 Loading Pit Isolation RNS-PL-V056 RNS Pump Suction to Cask C I ASME III-3 Loading Pit Isolation RNS-PL-V057A RNS Pump A Miniflow C I ASME III-3 Isolation RNS-PL-V057B RNS Pump B Miniflow C I ASME III-3 Isolation 3.2-49 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 30 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Normal Residual Heat Removal System (Continued)
RNS-PL-V059 RNS Pump Suction C I ASME III-3 Containment Isolation Test Connection RNS-PL-V061 RNS Return from CVS - B I ASME III-2 Containment Isolation RNS-PL-V065 RNS Discharge Drain Valve C I ASME III-3 RNS-PL-V066A RNS Discharge to DVI Line A C I ASME III-3 Drain RNS-PL-V066B RNS Discharge to DVI Line B C I ASME III-3 Drain RNS-PL-V067A RNS Discharge to DVI Line A B I ASME III-2 Drain RNS-PL-V067B RNS Discharge to DVI Line B B I ASME III-2 Drain RNS-PL-V068 RNS Discharge to IRWST C I ASME III-3 Drain RNS-PL-V069A RNS Pump A Miniflow Vent C I ASME III-3 RNS-PL-V069B RNS Pump B Miniflow Vent C I ASME III-3 RNS-PL-V071A RNS HX A Channel Head C I ASME III-3 Drain Isolation RNS-PL-V071B RNS HX B Channel Head C I ASME III-3 Drain Isolation RNS-PL-V072A RNS HX A Channel Head C I ASME III-3 Drain Isolation RNS-PL-V072B RNS HX B Channel Head C I ASME III-3 Drain Isolation RNS-PL-V073A RNS HX A Channel Head C I ASME III-3 Drain Isolation RNS-PL-V073B RNS HX B Channel Head C I ASME III-3 Drain Isolation RNS-PL-V074A RNS HX A Channel Head C I ASME III-3 Drain Isolation RNS-PL-V074B RNS HX B Channel Head C I ASME III-3 Drain Isolation RNS-PL-V075A RNS HX A Channel Head C I ASME III-3 Drain Isolation RNS-PL-V075B RNS HX B Channel Head C I ASME III-3 Drain Isolation RNS-PL-V080 IRWST Suction Line to RNS B I ASME III-2 Pump Vent RNS-PL-V081 RNS Cask Loading Pit C I ASME III-3 Suction Line Vent RNS-PL-V082 RNS Discharge Drain C I ASME III-3 RNS-PY-C01 Normal Residual Heat B I ASME III, MC Removal Suction Line Penetration 3.2-50 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 31 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Normal Residual Heat Removal System (Continued)
RNS-PY-C02 Normal Residual Heat B I ASME III, MC Removal Discharge Line Penetration RNS-PY-R01A Residual Heat Removal C I ASME III-3 Discharge to DVI Line Orifice A
RNS-PY-R01B Residual Heat Removal C I ASME III-3 Discharge to DVI Line Orifice B
RNS-PY-R02A Residual Heat Removal Pump C I ASME III-3 Miniflow Orifice A RNS-PY-R02B Residual Heat Removal Pump C I ASME III-3 Miniflow Orifice B RNS-PY-R03 Residual Heat Removal to C I ASME III-3 IRWST Line Orifice RNS-PY-R04 Residual Heat Removal to C I ASME III-3 Spent Fuel Pool Line Orifice Balance of system components are Class E Raw Water System (RWS) Location: Yard, Turbine Building System components are Class E Reactor System (RXS) Location: Containment n/a Fuel Assemblies C I Manufacturer Std.
RXS-FR-B06 Control Rod Cluster B6 B I Manufacturer Std.
RXS-FR-B10 Control Rod Cluster B10 B I Manufacturer Std.
RXS-FR-C05 Control Rod Cluster C5 B I Manufacturer Std.
RXS-FR-C07 Control Rod Cluster C7 B I Manufacturer Std.
RXS-FR-C09 Control Rod Cluster C9 B I Manufacturer Std.
RXS-FR-C11 Control Rod Cluster C11 B I Manufacturer Std.
RXS-FR-D06 Control Rod Cluster D6 B I Manufacturer Std.
RXS-FR-D08 Control Rod Cluster D8 B I Manufacturer Std.
RXS-FR-D10 Control Rod Cluster D10 B I Manufacturer Std.
RXS-FR-E03 Control Rod Cluster E3 B I Manufacturer Std.
RXS-FR-E05 Control Rod Cluster E5 B I Manufacturer Std.
RXS-FR-E07 Control Rod Cluster E7 B I Manufacturer Std.
RXS-FR-E09 Control Rod Cluster E9 B I Manufacturer Std.
RXS-FR-E11 Control Rod Cluster E11 B I Manufacturer Std.
RXS-FR-E13 Control Rod Cluster E13 B I Manufacturer Std.
RXS-FR-F02 Control Rod Cluster F2 B I Manufacturer Std.
RXS-FR-F04 Control Rod Cluster F4 B I Manufacturer Std.
RXS-FR-F12 Control Rod Cluster F12 B I Manufacturer Std.
RXS-FR-F14 Control Rod Cluster F14 B I Manufacturer Std RXS-FR-G03 Control Rod Cluster G3 B I Manufacturer Std.
RXS-FR-G05 Control Rod Cluster G5 B I Manufacturer Std.
RXS-FR-G07 Control Rod Cluster G7 B I Manufacturer Std.
RXS-FR-G09 Control Rod Cluster G9 B I Manufacturer Std.
3.2-51 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 32 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Reactor System (Continued)
RXS-FR-G11 Control Rod Cluster G11 B I Manufacturer Std.
RXS-FR-G13 Control Rod Cluster G13 B I Manufacturer Std.
RXS-FR-H04 Control Rod Cluster H4 B I Manufacturer Std.
RXS-FR-H08 Control Rod Cluster H8 B I Manufacturer Std.
RXS-FR-H12 Control Rod Cluster H12 B I Manufacturer Std.
RXS-FR-J03 Control Rod Cluster J3 B I Manufacturer Std.
RXS-FR-J05 Control Rod Cluster J5 B I Manufacturer Std.
RXS-FR-J07 Control Rod Cluster J7 B I Manufacturer Std.
RXS-FR-J09 Control Rod Cluster J9 B I Manufacturer Std.
RXS-FR-J11 Control Rod Cluster J11 B I Manufacturer Std.
RXS-FR-J13 Control Rod Cluster J13 B I Manufacturer Std.
RXS-FR-K02 Control Rod Cluster K2 B I Manufacturer Std.
RXS-FR-K04 Control Rod Cluster K4 B I Manufacturer Std.
RXS-FR-K12 Control Rod Cluster K12 B I Manufacturer Std.
RXS-FR-K14 Control Rod Cluster K14 B I Manufacturer Std.
RXS-FR-L03 Control Rod Cluster L3 B I Manufacturer Std.
RXS-FR-L05 Control Rod Cluster L5 B I Manufacturer Std.
RXS-FR-L07 Control Rod Cluster L7 B I Manufacturer Std.
RXS-FR-L09 Control Rod Cluster L9 B I Manufacturer Std.
RXS-FR-L11 Control Rod Cluster L11 B I Manufacturer Std.
RXS-FR-L13 Control Rod Cluster L13 B I Manufacturer Std.
RXS-FR-M06 Control Rod Cluster M6 B I Manufacturer Std.
RXS-FR-M08 Control Rod Cluster M8 B I Manufacturer Std.
RXS-FR-M10 Control Rod Cluster M10 B I Manufacturer Std.
RXS-FR-N05 Control Rod Cluster N05 B I Manufacturer Std.
RXS-FR-N07 Control Rod Cluster N07 B I Manufacturer Std.
RXS-FR-N09 Control Rod Cluster N09 B I Manufacturer Std.
RXS-FR-N11 Control Rod Cluster N11 B I Manufacturer Std.
RXS-FR-P06 Control Rod Cluster P06 B I Manufacturer Std.
RXS-FR-P10 Control Rod Cluster P10 B I Manufacturer Std.
RXS-MI-01 Reactor Upper Internals C I ASME III, CS RXS-MI-02 Reactor Lower Internals C I ASME III, CS RXS-MI-10 Non-Threaded Fasteners D II Manufacturer Std.
RXS-MI-11 Threaded Structural C I ASME III, CS Fasteners RXS-MI-20 Lower Core Support Plate C I ASME III, CS RXS-MI-22 Vortex Suppression Plate D II Manufacturer Std.
RXS-MI-23 Core Shroud Assembly D II Manufacturer Std.
RXS-MI-24 Radial Supports [4] C I ASME III, CS RXS-MI-25 Core Barrel C I ASME III, CS RXS-MI-26 Core Barrel Nozzle C I Manufacturer Std.
RXS-MI-27 Head and Vessel Pins D II Manufacturer Std.
RXS-MI-28 Lower Support Plate Fuel C I ASME III, CS Alignment Pins 3.2-52 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 33 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Reactor System (Continued)
RXS-MI-29 Core Barrel Hold Down Spring C I Manufacturer Std.
RXS-MI-50 Upper Support C I ASME III, CS RXS-MI-51 Upper Core Plate C I ASME III, CS RXS-MI-52 Support Columns [42] C I ASME III, CS RXS-MI-53 Guide Tube Assemblies [69] C I Manufacturer Std.
RXS-MI-54 Upper Core Plate Fuel C I ASME III, CS Alignment Pins RXS-MI-55 Upper Core Plate Inserts C I ASME III, CS RXS-MI-56 Safety Injection Deflector D II Manufacturer Std.
RXS-MI-57 Irradiation Specimen Guide D II Manufacturer Std.
Tubes RXS-MI-58 Head Cooling Nozzles D II Manufacturer Std.
n/a Neutron Pad D II Manufacturer Std.
n/a Instrument Grid Assembly D II Manufacturer Std.
n/a DVI Flow Diverter D II Manufacturer Std.
RXS-MI-80 Reactor Vessel Flow Skirt D II Manufacturer Std.
RXS-MN-01 Reactor Vessel Cavity D II Manufacturer Std.
Reflective Insulation RXS-MV-01 Reactor Vessel A I ASME III-1 The reactor vessel has two tag numbers; RCS-MV-01 for the reactor coolant system (RCS) and RXS-MV-01 for the reactor system (RXS)
RXS-MV-10 Reactor Integrated Head C I ANSI/AISC N690 Package RXS-MV-10A Integrated Head Package C I ASME-NF Shroud RXS-MV-10B Integrated Head Package C I ASME-NF Seismic Support System 3.2-53 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 34 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Reactor System (Continued)
RXS-MV-11B06/ CRDM Latch Assemblies C I Manufacturer Std.
B08/B10/C05/C07/
C09/C11/D04/
D06/D08/D10/
D12/E03/E05/E07/
E09/E11/E13/F02/
F04/F06/F08/F10/
F12/F14/G03/G05/
G07/G09/G11/
G13/
H02/H04/H06/H08/
H10/H12/H14/J03/
J05/J07/J09/J11/
J13/K02/K04/K06/
K08/K10/K12/K14/
L03/L05/L07/L09/
L11/L13/M04/M06/
M08/M10/M12/
N05/N07/N09/N11/
P06/P08/P10 RXS-MV-11B06/ CRDM Drive Rod Assemblies D NS Manufacturer Std.
B08/B10/C05/C07/
C09/C11/D04/
D06/D08/D10/
D12/E03/E05/E07/
E09/E11/E13/F02/
F04/F06/F08/F10/
F12/F14/G03/G05/
G07/G09/G11/
G13/
H02/H04/H06/H08/
H10/H12/H14/J03/
J05/J07/J09/J11/
J13/K02/K04/K06/
K08/K10/K12/K14/
L03/L05/L07/L09/
L11/L13/M04/M06/
M08/M10/M12/
N05/N07/N09/N11/
P06/P08/P10 3.2-54 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 35 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Reactor System (Continued)
RXS-MV-11B06/ CRDM Coil Stack Assemblies D NS Manufacturer Std.
B08/B10/C05/C07/
C09/C11/D04/D06/
D08/D10/D12/E03/
E05/E07/E09/E11/
E13/F02/F04/F06/
F08/F10/F12/F14/
G03/G05/G07/
G09/
G11/G13/H02/H04/
H06/H08/H10/H12/
H14/J03/J05/J07/
J09/J11/J13/K02/
K04/K06/K08/K10/
K12/K14/L03/L05/
L07/L09/L11/L13/
M04/M06/M08/
M10/M12/N05/
N07/N09/N11/P06/
P08/P10 RXS-MV-11B06L/ CRDM Latch Housings A I ASME III-1 B08L/B10L/C05L/
C07L/C09L/C11L/
D04L/D06L/D08L/
D10L/D12L/E03L/
E05L/E07L/E09L/
E11L/E13L/F02L/
F04L/F06L/F08L/
F10L/F12L/F14L/
G03L/G05L/G07L/
G09L/G11L/G13L/
H02L/H04L/H06L/
H08L/H10L/H12L/
H14L/J03L/J05L/
J07L/J09L/J11L/
J13L/K02L/K04L/
K06L/K08L/K10L/
K12L/K14L/L03L/
L05L/L07L/L09L/
L11L/L13L/M04L/
M06L/M08L/
M10L/M12L/N05L/
N07L/N09/N11L/
P06L/P08L/P10L 3.2-55 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 36 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Reactor System (Continued)
RXS-MV-11B06R/ CRDM Rod Travel Housings A I ASME III-1 B08R/B10R/C05R/
C07R/C09R/C11R/
D04R/D06R/D08R/
D10R/D12R/E03R/
E05R/E07R/E09R/
E11R/E13R/F02R/
F04R/F06R/F08R/
F10R/F12R/F14R/
G03R/G05R/
G07R/
G09R/G11R/
G13R/
H02R/H04R/H06R/
H08R/H10R/H12R/
H14R/J03R/J05R/
J07R/J09R/J11R/
J13R/K02R/K04R/
K06R/K08R/K10R/
K12R/K14R/L03R/
L05R/L07R/L09R/
L11R/L13R/M04R/
M06R/M08R/
M10R/M12R/
N05R/N07R/N11R/
P06R/P08R/P10R RXS-MY-Y01 IHP Lower Shroud Assembly C I ASME-NF RXS-MY-Y51 Integrated Head Package C II NUREG-0612, ANSI Lifting Rig N14.6 RXS-MY-Y11 Incore Instrumentation A I ASME III-1 QuickLoc Assembly 1 RXS-MY-Y12 Incore Instrumentation A I ASME III-1 QuickLoc Assembly 2 RXS-MY-Y13 Incore Instrumentation A I ASME III-1 QuickLoc Assembly 3 RXS-MY-Y14 Incore Instrumentation A I ASME III-1 QuickLoc Assembly 4 RXS-MY-Y15 Incore Instrumentation A I ASME III-1 QuickLoc Assembly 5 RXS-MY-Y16 Incore Instrumentation A I ASME III-1 QuickLoc Assembly 6 RXS-MY-Y17 Incore Instrumentation A I ASME III-1 QuickLoc Assembly 7 RXS-MY-Y18 Incore Instrumentation A I ASME III-1 QuickLoc Assembly 8 Balance of system components are Class E 3.2-56 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 37 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Sanitary Drainage System (SDS) Location: Various SDS-PL-V001 SDS MCR Vent Isolation C I ASME III-3 Valve SDS-PL-V002 SDS MCR Vent Isolation C I ASME III-3 Valve Balance of system components are Class D, P, and W Spent Fuel Pool Cooling System (SFS) Location: Auxiliary Building, Containment n/a Heat Exchangers, SFS and D NS ASME VIII CCS Side n/a Pumps D NS Hydraulic Institute Std.
n/a Demineralizers D NS ASME VIII n/a Filters D NS ASME VIII n/a Valves Providing SFS AP1000 D NS ANSI B16.34 Equipment Class D Function SFS-PL-V024A Spent Fuel Pool Level C I ASME III-3 Instrument Isolation SFS-PL-V024B Spent Fuel Pool Level C I ASME III-3 Instrument Isolation SFS-PL-V024C Spent Fuel Pool Level C I ASME III-3 Instrument Isolation SFS-PL-V028 Cask Washdown Pit Level C I ASME III-3 Instrument Isolation SFS-PL-V031 SFS Refueling Cavity Drain to C I ASME III-3 SGS Compartment Isolation SFS-PL-V032 SFS Refueling Cavity Suction C I ASME III-3 Isolation SFS-PL-V033 SFS Refueling Cavity Drain to C I ASME III-3 Containment Sump Isolation SFS-PL-V034 SFS Suction Line B I ASME III-2 Containment Isolation SFS-PL-V035 SFS Suction Line B I ASME III-2 Containment Isolation SFS-PL-V037 SFS Discharge Line B I ASME III-2 Containment Isolation SFS-PL-V038 SFS Discharge Line B I ASME III-2 Containment Isolation SFS-PL-V039 SFS Suction Line from IRWST C I ASME III-3 Isolation SFS-PL-V040 SFS Fuel Transfer Canal C I ASME III-3 Drain Isolation SFS-PL-V041 SFS Cask Loading Pit Drain C I ASME III-3 Isolation SFS-PL-V042 Cask Loading Pit to Pump C I ASME III-3 Suction Isolation SFS-PL-V043 Cask Loading Pit Level C I ASME III-3 Transmitter Root Isolation Valve 3.2-57 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 38 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Spent Fuel Pool Cooling System (Continued)
SFS-PL-V045 SFS Discharge to Cask C I ASME III-3 Loading Pit Isolation SFS-PL-V047 SFS Demineralized Water D NS ANSI B16.34 Makeup to SFP Reverse Flow Prevent SFS-PL-V048 SFS Containment Penetration B I ASME III-2 Test Connection SFS-PL-V049 SFS Cask Loading Pit Drain to C I ASME III-3 WLS Isolation SFS-PL-V056 SFS Containment Penetration B I ASME III-2 Test Connection Isolation SFS-PL-V058 SFS Containment Isolation C I ASME III-3 Valve V034 Test SFS-PL-V066 Spent Fuel Pool to Cask C I ASME III-3 Washdown Pit Isolation SFS-PL-V067 SFS Containment Isolation B I ASME III-2 Relief Valve SFS-PL-V068 Cask Washdown Pit Drain C I ASME III-3 Isolation SFS-PL-V071 Refueling Cavity Overflow to C I ASME III-3 SG Compartment SFS-PL-V072 Refueling Cavity Overflow to C I ASME III-3 SG Compartment SFS-PL-V075 SFS Containment Floodup C I ASME III-3 Isolation Valve SFS-PL-V117 Refueling Cavity Drain Line C I ASME III-3 Test Connection SFS-PY-C01 Spent Fuel Cooling Pump B I ASME III, MC Discharge to IRWST SFS-PY-C02 Spent Fuel Cooling Pump B I ASME III, MC Suction from IRWST SFS-PY-R04 SFS Cask Loading and C I ASME III-3 Washdown Orifice Balance of system components are Class D Steam Generator System (SGS) Location: Containment and Auxiliary Building SGS-MY-Y01A Steam Generator A PORV D NS Manufacturer Std.
Silencer SGS-MY-Y01B Steam Generator B PORV D NS Manufacturer Std.
Silencer SGS-PL-V001A LT001, LT011 Root Isolation B I ASME III-2 Valve SGS-PL-V001B LT005, LT013 Root Isolation B I ASME III-2 Valve SGS-PL-V002A LT001 Root Isolation Valve B I ASME III-2 SGS-PL-V002B LT005 Root Isolation Valve B I ASME III-2 SGS-PL-V003A LT002, LT012 Root Isolation B I ASME III-2 Valve 3.2-58 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 39 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Steam Generator System (Continued)
SGS-PL-V003B LT006, LT014 Root Isolation B I ASME III-2 Valve SGS-PL-V004A LT002 Root Isolation Valve B I ASME III-2 SGS-PL-V004B LT006 Root Isolation Valve B I ASME III-2 SGS-PL-V005A LT003, LT015, LT044 Root B I ASME III-2 Isolation Valve SGS-PL-V005B LT007, LT017, LT046 Root B I ASME III-2 Isolation Valve SGS-PL-V006A LT003 Root Isolation Valve B I ASME III-2 SGS-PL-V006B LT007 Root Isolation Valve B I ASME III-2 SGS-PL-V007A LT004, LT016, LT045 Root B I ASME III-2 Isolation Valve SGS-PL-V007B LT008, LT018, LT047 Root B I ASME III-2 Isolation Valve SGS-PL-V008A LT004 Root Isolation Valve B I ASME III-2 SGS-PL-V008B LT008 Root Isolation Valve B I ASME III-2 SGS-PL-V010A LT011 Root Isolation Valve B I ASME III-2 SGS-PL-V010B LT013 Root Isolation Valve B I ASME III-2 SGS-PL-V011A LT011 Root Isolation Valve B I ASME III-2 SGS-PL-V011B LT013 Root Isolation Valve B I ASME III-2 SGS-PL-V012A LT012 Root Isolation Valve B I ASME III-2 SGS-PL-V012B LT014 Root Isolation Valve B I ASME III-2 SGS-PL-V013A LT012 Root Isolation Valve B I ASME III-2 SGS-PL-V013B LT014 Root Isolation Valve B I ASME III-2 SGS-PL-V014A PORV Discharge Condensate D NS ASME B31.1 Drain Isolation SGS-PL-V014B PORV Discharge Condensate D NS ASME B31.1 Drain Isolation SGS-PL-V015A FT021 Root Isolation Valve B I ASME III-2 SGS-PL-V015B FT023 Root Isolation Valve B I ASME III-2 SGS-PL-V016A FT020 Root Isolation Valve B I ASME III-2 SGS-PL-V016B FT022 Root Isolation Valve B I ASME III-2 SGS-PL-V017A FT021 Root Isolation Valve B I ASME III-2 SGS-PL-V017B FT023 Root Isolation Valve B I ASME III-2 SGS-PL-V018A FT020 Root Isolation Valve B I ASME III-2 SGS-PL-V018B FT022 Root Isolation Valve B I ASME III-2 SGS-PL-V019A Main Steam Line Vent B I ASME III-2 Isolation SGS-PL-V019B Main Steam Line Vent B I ASME III-2 Isolation SGS-PL-V020A FT024 Root Isolation Valve B I ASME III-2 SGS-PL-V020B FT025 Root Isolation Valve B I ASME III-2 SGS-PL-V021A FT024 Root Isolation Valve B I ASME III-2 SGS-PL-V021B FT025 Root Isolation Valve B I ASME III-2 SGS-PL-V022A PT030 Root Isolation Valve B I ASME III-2 3.2-59 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 40 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Steam Generator System (Continued)
SGS-PL-V022B PT034 Root Isolation Valve B I ASME III-2 SGS-PL-V023A PT031 Root Isolation Valve B I ASME III-2 SGS-PL-V023B PT035 Root Isolation Valve B I ASME III-2 SGS-PL-V024A PT032 Root Isolation Valve B I ASME III-2 SGS-PL-V024B PT036 Root Isolation Valve B I ASME III-2 SGS-PL-V025A PT033 Root Isolation Valve B I ASME III-2 SGS-PL-V025B PT037 Root Isolation Valve B I ASME III-2 SGS-PL-V027A PORV Block Valve SG 01 B I ASME III-2 SGS-PL-V027B PORV Block Valve SG 02 B I ASME III-2 SGS-PL-V030A Main Steam Safety Valve B I ASME III-2 SG 01 SGS-PL-V030B Main Steam Safety Valve B I ASME III-2 SG 02 SGS-PL-V031A Main Steam Safety Valve B I ASME III-2 SG 01 SGS-PL-V031B Main Steam Safety Valve B I ASME III-2 SG 02 SGS-PL-V032A Main Steam Safety Valve B I ASME III-2 SG 01 SGS-PL-V032B Main Steam Safety Valve B I ASME III-2 SG 02 SGS-PL-V033A Main Steam Safety Valve B I ASME III-2 SG 01 SGS-PL-V033B Main Steam Safety Valve B I ASME III-2 SG 02 SGS-PL-V034A Main Steam Safety Valve B I ASME III-2 SG 01 SGS-PL-V034B Main Steam Safety Valve B I ASME III-2 SG 02 SGS-PL-V035A Main Steam Safety Valve B I ASME III-2 SG 01 SGS-PL-V035B Main Steam Safety Valve B I ASME III-2 SG 02 SGS-PL-V036A Steam Line Condensate Drain B I ASME III-2 Isolation SGS-PL-V036B Steam Line Condensate Drain B I ASME III-2 Isolation SGS-PL-V038A Steam Line #1 Nitrogen B I ASME III-2 Supply Isolation SGS-PL-V038B Steam Line #2 Nitrogen B I ASME III-2 Supply Isolation SGS-PL-V040A Main Steam Line Isolation B I ASME III-2 SGS-PL-V040B Main Steam Line Isolation B I ASME III-2 SGS-PL-V042A MSIV Bypass Control Isolation B I ASME III-2 SGS-PL-V042B MSIV Bypass Control Isolation B I ASME III-2 SGS-PL-V043A MSIV Bypass Control Isolation C I ASME III-3 3.2-60 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 41 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Steam Generator System (Continued)
SGS-PL-V043B MSIV Bypass Control Isolation C I ASME III-3 SGS-PL-V045A SG 1 Condensate Pipe Drain B I ASME III-2 Valve SGS-PL-V045B SG 2 Condensate Pipe Drain B I ASME III-2 Valve SGS-PL-V046A LT015 Root Isolation Valve B I ASME III-2 SGS-PL-V046B LT017 Root Isolation Valve B I ASME III-2 SGS-PL-V047A LT015, LT044 Root Isolation B I ASME III-2 Valve SGS-PL-V047B LT017, LT046 Root Isolation B I ASME III-2 Valve SGS-PL-V048A LT016 Root Isolation Valve B I ASME III-2 SGS-PL-V048B LT018 Root Isolation Valve B I ASME III-2 SGS-PL-V049A LT016, LT045 Root Isolation B I ASME III-2 Valve SGS-PL-V049B LT018, LT047 Root Isolation B I ASME III-2 Valve SGS-PL-V050A LT044 Root Isolation Valve B I ASME III-2 SGS-PL-V050B LT046 Root Isolation Valve B I ASME III-2 SGS-PL-V051A LT044 Root Isolation Valve B I ASME III-2 SGS-PL-V051B LT046 Root Isolation Valve B I ASME III-2 SGS-PL-V052A LT045 Root Isolation Valve B I ASME III-2 SGS-PL-V052B LT047 Root Isolation Valve B I ASME III-2 SGS-PL-V053A LT045 Root Isolation Valve B I ASME III-2 SGS-PL-V053B LT047 Root Isolation Valve B I ASME III-2 SGS-PL-V056A PT062 Root Isolation Valve C I ASME III-3 SGS-PL-V056B PT063 Root Isolation Valve C I ASME III-3 SGS-PL-V057A Main Feedwater Isolation B I ASME III-2 SGS-PL-V057B Main Feedwater Isolation B I ASME III-2 SGS-PL-V058A Main Feedwater Check B I ASME III-2 SGS-PL-V058B Main Feedwater Check B I ASME III-2 SGS-PL-V062A FT055A Root Isolation Valve C I ASME III-3 SGS-PL-V062B FT056A Root Isolation Valve C I ASME III-3 SGS-PL-V063A FT055A Root Isolation Valve C I ASME III-3 SGS-PL-V063B FT056A Root Isolation Valve C I ASME III-3 SGS-PL-V064A FT055B Root Isolation Valve C I ASME III-3 SGS-PL-V064B FT056B Root Isolation Valve C I ASME III-3 SGS-PL-V065A FT055B Root Isolation Valve C I ASME III-3 SGS-PL-V065B FT056B Root Isolation Valve C I ASME III-3 SGS-PL-V066A FT055 Root Isolation Valve C I ASME III-3 SGS-PL-V066B FT056 Root Isolation Valve C I ASME III-3 SGS-PL-V067A Startup Feedwater Isolation B I ASME III-2 SGS-PL-V067B Startup Feedwater Isolation B I ASME III-2 SGS-PL-V068A FT055 Root Isolation Valve C I ASME III-3 SGS-PL-V068B FT056 Root Isolation Valve C I ASME III-3 3.2-61 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 42 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Steam Generator System (Continued)
SGS-PL-V074A SG Blowdown Isolation B I ASME III-2 SGS-PL-V074B SG Blowdown Isolation B I ASME III-2 SGS-PL-V075A SG Series Blowdown Isolation C I ASME III-3 SGS-PL-V075B SG Series Blowdown Isolation C I ASME III-3 SGS-PL-V076A Blowdown Vent Line Isolation C I ASME III-3 SGS-PL-V076B Blowdown Vent Line Isolation C I ASME III-3 SGS-PL-V084A SG 1 Nitrogen Sparging B I ASME III-2 Isolation SGS-PL-V084B SG 2 Nitrogen Sparging B I ASME III-2 Isolation SGS-PL-V086A Steam Line Condensate Drain C I ASME III-3 Control SGS-PL-V086B Steam Line Condensate Drain C I ASME III-3 Control SGS-PL-V093A Orifice Isolation Valve C I ASME III-3 SGS-PL-V093B Orifice Isolation Valve C I ASME III-3 SGS-PL-V094A Orifice Cleanout Line Isolation C I ASME III-3 Valve SGS-PL-V094B Orifice Cleanout Line Isolation C I ASME III-3 Valve SGS-PL-V095A Orifice Isolation Valve C I ASME III-3 SGS-PL-V095B Orifice Isolation Valve C I ASME III-3 SGS-PL-V096A Steam Line Condensate Drain B I ASME III-2 Level Isolation Valve SGS-PL-V096B Steam Line Condensate Drain B I ASME III-2 Level Isolation Valve SGS-PL-V097A Steam Line Condensate Drain B I ASME III-2 Level Isolation Valve SGS-PL-V097B Steam Line Condensate Drain B I ASME III-2 Level Isolation Valve SGS-PL-V100A Startup Feedwater Drain C I ASME III-3 Isolation Valve SGS-PL-V100B Startup Feedwater Drain C I ASME III-3 Isolation Valve SGS-PL-V101A Main Feedwater Drain B I ASME III-2 Isolation Valve SGS-PL-V101B Main Feedwater Drain B I ASME III-2 Isolation Valve SGS-PL-V102A Startup Feedwater Vent C I ASME III-3 Isolation Valve SGS-PL-V102B Startup Feedwater Vent C I ASME III-3 Isolation Valve SGS-PL-V103A Main Feedwater Vent Isolation B I ASME III-2 Valve SGS-PL-V103B Main Feedwater Vent Isolation B I ASME III-2 Valve 3.2-62 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 43 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Steam Generator System (Continued)
SGS-PL-V104A Main Feedwater Drain C I ASME III-3 Isolation Valve SGS-PL-V104B Main Feedwater Drain C I ASME III-3 Isolation Valve SGS-PL-V233A Power Operated Relief Valve C I ASME III-3 SGS-PL-V233B Power Operated Relief Valve C I ASME III-3 SGS-PL-V240A MSIV Bypass Isolation B I ASME III-2 SGS-PL-V240B MSIV Bypass Isolation B I ASME III-2 SGS-PL-V250A Main Feedwater Control C I ASME III-3 SGS-PL-V250B Main Feedwater Control C I ASME III-3 SGS-PL-V255A Startup Feedwater Control C I ASME III-3 SGS-PL-V255B Startup Feedwater Control C I ASME III-3 SGS-PL-V256A Startup Feedwater Check C I ASME III-3 Valve SGS-PL-V256B Startup Feedwater Check C I ASME III-3 Valve SGS-PL-V257A Main Feedwater Thermal C I ASME III-3 Relief Valve SGS-PL-V257B Main Feedwater Thermal C I ASME III-3 Relief Valve SGS-PL-V258A Startup Feedwater Thermal C I ASME III-3 Relief Valve SGS-PL-V258B Startup Feedwater Thermal C I ASME III-3 Relief Valve SGS-PL-V300 SG1 NR LT001 Upper Isol B I ASME III-2 SGS-PL-V301 SG1 NR LT001 Lower Isol B I ASME III-2 SGS-PL-V302 SG1 NR LT002 Upper Isol B I ASME III-2 SGS-PL-V303 SG1 NR LT002 Lower Isol B I ASME III-2 SGS-PL-V304 SG1 NR LT003 Upper Isol B I ASME III-2 SGS-PL-V305 SG1 NR LT003 Lower Isol B I ASME III-2 SGS-PL-V306 SG1 NR LT004 Upper Isol B I ASME III-2 SGS-PL-V307 SG1 NR LT004 Lower Isol B I ASME III-2 SGS-PL-V308 SG1 NR LT011 Lower Isol B I ASME III-2 SGS-PL-V309 SG1 NR LT012 Lower Isol B I ASME III-2 SGS-PL-V310 SG1 NR LT015 Lower Isol B I ASME III-2 SGS-PL-V311 SG1 NR LT016 Lower Isol B I ASME III-2 SGS-PL-V312 SG1 Main Steam FT020 Isol B I ASME III-2 SGS-PL-V313 SG1 Main Steam FT021 Isol B I ASME III-2 SGS-PL-V314 SG1 Main Steam FT024 Isol B I ASME III-2 SGS-PL-V320 SG2 NR LT005 Upper Isol B I ASME III-2 SGS-PL-V321 SG2 NR LT005 Lower Isol B I ASME III-2 SGS-PL-V322 SG2 NR LT006 Upper Isol B I ASME III-2 SGS-PL-V323 SG2 NR LT006 Lower Isol B I ASME III-2 SGS-PL-V324 SG2 NR LT007 Upper Isol B I ASME III-2 SGS-PL-V325 SG2 NR LT007 Lower Isol B I ASME III-2 3.2-63 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 44 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Steam Generator System (Continued)
SGS-PL-V326 SG2 NR LT008 Upper Isol B I ASME III-2 SGS-PL-V327 SG2 NR LT008 Lower Isol B I ASME III-2 SGS-PL-V328 SG2 NR LT013 Lower Isol B I ASME III-2 SGS-PL-V329 SG2 NR LT014 Lower Isol B I ASME III-2 SGS-PL-V330 SG2 NR LT017 Lower Isol B I ASME III-2 SGS-PL-V331 SG2 NR LT018 Lower Isol B I ASME III-2 SGS-PL-V332 SG2 Main Steam FT022 Isol B I ASME III-2 SGS-PL-V333 SG2 Main Steam FT023 Isol B I ASME III-2 SGS-PL-V334 SG2 Main Steam FT025 Isol B I ASME III-2 SGS-PY-C01A Main Steam Line A B I ASME III, MC Penetration SGS-PY-C01B Main Steam Line B B I ASME III, MC Penetration SGS-PY-C02A Main Feedwater Line A B I ASME III, MC Penetration SGS-PY-C02B Main Feedwater Line B B I ASME III, MC Penetration SGS-PY-C03A Steam Generator A Blow- B I ASME III, MC down Line Penetration SGS-PY-C03B Steam Generator B Blow- B I ASME III, MC down Line Penetration SGS-PY-C05A Startup Feedwater Line A B I ASME III, MC Penetration SGS-PY-C05B Startup Feedwater Line B B I ASME III, MC Penetration SGS-PY-R02A SG 1 Main Steam Line Drain C I ASME III-3 Plug-Resistant Orifice SGS-PY-R02B SG 2 Main Steam Line Drain C I ASME III-3 Plug-Resistant Orifice SGS-PY-Y01A Flexible Hose from SG 1 to B I ASME III-2 Upper Tap LT001/LT011 SGS-PY-Y01B Flexible Hose from SG 2 to B I ASME III-2 Upper Tap LT005/LT013 SGS-PY-Y02A Flexible Hose from SG 1 to B I ASME III-2 Upper Tap LT001 SGS-PY-Y02B Flexible Hose from SG 2 to B I ASME III-2 Upper Tap LT005 SGS-PY-Y03A Flexible Hose from SG 1 to B I ASME III-2 Upper Tap LT002/LT015 SGS-PY-Y03B Flexible Hose from SG 2 to B I ASME III-2 Upper Tap LT006/LT014 SGS-PY-Y04A Flexible Hose from SG 1 to B I ASME III-2 Upper Tap LT002 SGS-PY-Y04B Flexible Hose from SG 2 to B I ASME III-2 Upper Tap LT006 3.2-64 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 45 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Steam Generator System (Continued)
SGS-PY-Y05A Flexible Hose from SG 1 to B I ASME III-2 Upper Tap LT003/LT012/
LT044 SGS-PY-Y05B Flexible Hose from SG 2 to B I ASME III-2 Upper Tap LT007/LT017/
LT046 SGS-PY-Y06A Flexible Hose from SG 1 to B I ASME III-2 Upper Tap LT003 SGS-PY-Y06B Flexible Hose from SG 2 to B I ASME III-2 Upper Tap LT007 SGS-PY-Y07A Flexible Hose from SG 1 to B I ASME III-2 Upper Tap LT004/LT016/
LT045 SGS-PY-Y07B Flexible Hose from SG 2 to B I ASME III-2 Upper Tap LT008/LT018/
LT047 SGS-PY-Y08A Flexible Hose from SG 1 to B I ASME III-2 Upper Tap LT004 SGS-PY-Y08B Flexible Hose from SG 2 to B I ASME III-2 Upper Tap LT008 SGS-PY-Y09A Flexible Hose from SG 1 to B I ASME III-2 Upper Tap LT011 SGS-PY-Y09B Flexible Hose from SG 2 to B I ASME III-2 Upper Tap LT013 SGS-PY-Y10A Flexible Hose from SG 1 to B I ASME III-2 Upper Tap LT012 SGS-PY-Y10B Flexible Hose from SG 2 to B I ASME III-2 Upper Tap LT014 SGS-PY-Y11A Flexible Hose from SG 1 to B I ASME III-2 Upper Tap LT015/LT044 SGS-PY-Y11B Flexible Hose from SG 2 to B I ASME III-2 Upper Tap LT017/LT046 SGS-PY-Y12A Flexible Hose from SG 1 to B I ASME III-2 Upper Tap LT016/LT045 SGS-PY-Y12B Flexible Hose from SG 2 to B I ASME III-2 Upper Tap LT018/LT047 SGS-PY-Y13A Flexible Hose from SG 1 to B I ASME III-2 Lower Tap FT024 SGS-PY-Y13B Flexible Hose from SG 2 to B I ASME III-2 Lower Tap FT025 SGS-PY-Y14A Flexible Hose from SG 1 to B I ASME III-2 Lower Tap FT021 SGS-PY-Y14B Flexible Hose from SG 2 to B I ASME III-2 Lower Tap FT023 SGS-PY-Y15A Flexible Hose from SG 1 to B I ASME III-2 Lower Tap FT020 SGS-PY-Y15B Flexible Hose from SG 2 to B I ASME III-2 Lower Tap FT022 3.2-65 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 46 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Steam Generator System (Continued)
SGS-PY-Y16A Flexible Hose from Main B I ASME III-2 Steam Line to Upper Tap FT020 SGS-PY-Y16B Flexible Hose from Main B I ASME III-2 Steam Line to Upper Tap FT022 SGS-PY-Y17A Flexible Hose from Main B I ASME III-2 Steam Line to Upper Tap FT021 SGS-PY-Y17B Flexible Hose from Main B I ASME III-2 Steam Line to Upper Tap FT023 SGS-PY-Y18A Flexible Hose from Main B I ASME III-2 Steam Line to Upper Tap FT024 SGS-PY-Y18B Flexible Hose from Main B I ASME III-2 Steam Line to Upper Tap FT025 SGS-PY-Y19A Flexible Hose from Main B I ASME III-2 Steam Line to PT030 SGS-PY-Y19B Flexible Hose from Main B I ASME III-2 Steam Line to PT034 SGS-PY-Y20A Flexible Hose from Main B I ASME III-2 Steam Line to PT032 SGS-PY-Y20B Flexible Hose from Main B I ASME III-2 Steam Line to PT036 Secondary Sampling System (SSS) Location: Turbine Building System components are Class E Service Water System (SWS) Location: Turbine Building and Yard n/a Service Water Cooling Tower D NS Manufacturer Std.
Fans n/a Service Water Cooling Tower D NS Manufacturer Std.
n/a Service Water Pumps D NS Hydraulic Institute Std.
n/a Valves Providing SWS D NS ANSI B16.34 AP1000 Equipment Class D Function Turbine Building Closed Cooling Water System (TCS) Location: Turbine Building System components are Class E Turbine Island Vents, Drains and Relief System (TDS) Location: Turbine Building n/a Piping and components that D NS ASME B31.1 provide the path from the GSS and CMS to atmosphere and rad monitor Balance of system components are Class E Main Turbine Control and Diagnostic System (TOS) Location: Turbine Building System components are Class E 3.2-66 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 47 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Radiologically Controlled Area Ventilation System (VAS) Location: Auxiliary Building and Annex Building n/a CVS Pump Room Coolers Note 2 NS Manufacturer Std.
n/a RNS Pump Room Coolers Note 2 II Manufacturer Std.
n/a Valves Providing VAS AP1000 D NS ANSI B16.34 Equipment Class D Function n/a Shutoff, Isolation, and L NS ANSI/AMCA-500-D Balancing Dampers n/a Shutoff and Isolation Dampers R NS ASME AG-1 n/a Fire Dampers L NS UL-555 n/a Air Handling Units L NS Manufacturer Std.
n/a Filters L NS UL 900 n/a Ductwork L NS SMACNA n/a Ductwork in Auxiliary Building L II SMACNA Ductwork in radiation except ductwork attached to chemistry laboratory is mechanical modules nonseismic.
n/a Ductwork R II ASME AG-1 n/a Fans L NS AMCA Balance of system components are Class E or Class L Nuclear Island Nonradioactive Ventilation System (VBS) Location: Shield Building, Auxiliary Building, and Annex Building n/a Battery Rooms Exhaust Fans Note 2 II AMCA n/a PCS Room Heaters Note 3 II Manufacturer Std.
n/a Fire Dampers Note 2, L NS UL-555 n/a Fire Dampers Note 2, NS ASME AG-1 and R UL-555 n/a Combination Fire/Smoke Note 2, L NS UL-555 and UL-555S Dampers n/a Combination Fire/Smoke Note 2, NS ASME AG-1, Dampers R UL-555, and UL-555S n/a Dampers Providing AP1000 Note 2 NS ANSI/AMCA-500-D Equipment Class D Function n/a Shutoff, Balancing, and R NS ASME AG-1 Isolation Dampers n/a Shutoff and Balancing L NS ANSI/AMCA-500-D Dampers VBS-MP-01A Sample Pump A C I Manufacturer Std.
VBS-MP-01B Sample Pump B C I Manufacturer Std.
n/a MCR/CSA Supplemental Air Note 2 II ASME AG-1, ASME Filtration Units N509, Note 4 VBS-PL-V186 MCR Isolation Valve C I ASME III-3 VBS-PL-V187 MCR Isolation Valve C I ASME III-3 VBS-PL-V188 MCR Isolation Valve C I ASME III-3 VBS-PL-V189 MCR Isolation Valve C I ASME III-3 VBS-PL-V190 MCR Isolation Valve C I ASME III-3 VBS-PL-V191 MCR Isolation Valve C I ASME III-3 3.2-67 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 48 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Nuclear Island Nonradioactive Ventilation System (Continued) n/a Valves Providing VBS AP1000 D NS ANSI B16.34 Equipment Class D Function n/a Other Air Handling Units Note 2 NS Manufacturer Std.
n/a Filters Note 2 NS UL 900 n/a Other Fans Note 2, L NS AMCA VBS-MA-10A Ancillary Fan D II ANSI/AMCA 210, Equipment Anchorage 300 is Seismic Category II VBS-MA-10B Ancillary Fan D II ANSI/AMCA 210, Equipment Anchorage 300 is Seismic Category II VBS-MA-11 Ancillary Fan D II ANSI/AMCA 210, Equipment Anchorage 300 is Seismic Category II VBS-MA-12 Ancillary Fan D II ANSI/AMCA 210, Equipment Anchorage 300 is Seismic Category II n/a MCR Pressure Boundary C I ANSI/AISC N690 CA52 Penetration Cast Sleeve n/a Ductwork in Auxiliary Building Note 2, L II SMACNA n/a Ductwork in Annex Building Note 2, L NS SMACNA n/a Ductwork in Auxiliary Building Note 2, II ASME AG-1 R
n/a Ductwork in Annex Building Note 2, NS ASME AG-1 R
VBS-MY-Y100 Silencer D II Manufacturer Std. Equipment Anchorage is Seismic Category II VBS-MY-Y101 Silencer D II Manufacturer Std. Equipment Anchorage is Seismic Category II VBS-MY-Y102 Static Air Mixer Note 1 I Manufacturer Std. Inside Seismic Category II Supported Duct; Commercially dedicated to ASME AG-1 VBS-PL-V184A VBS-PDT-032A Isolation C I ASME III-3 Valve VBS-PL-V184B VBS-PDT-032B Isolation C I ASME III-3 Valve VBS-PL-V184C VBS-PDT-032C Isolation C I ASME III-3 Valve Balance of system components are Class E or Class L Containment Recirculation Cooling System (VCS) Location: Containment n/a Dampers L II ANSI/AMCA-500-D n/a Fan Coil Units L II Manufacturer Std.
n/a Ductwork L II SMACNA n/a Fans L II AMCA n/a Fire Damper L II UL-555 Balance of system components are Class L 3.2-68 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 49 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Main Control Room Emergency Habitability System (VES) Location: Auxiliary Building VES-MT-01 Emergency Air Storage C I ASME VIII, Appendix Tank 01 22 VES-MT-02 Emergency Air Storage C I ASME VIII, Appendix Tank 02 22 VES-MT-03 Emergency Air Storage C I ASME VIII, Appendix Tank 03 22 VES-MT-04 Emergency Air Storage C I ASME VIII, Appendix Tank 04 22 VES-MT-05 Emergency Air Storage C I ASME VIII, Appendix Tank 05 22 VES-MT-06 Emergency Air Storage C I ASME VIII, Appendix Tank 06 22 VES-MT-07 Emergency Air Storage C I ASME VIII, Appendix Tank 07 22 VES-MT-08 Emergency Air Storage C I ASME VIII, Appendix Tank 08 22 VES-MT-09 Emergency Air Storage C I ASME VIII, Appendix Tank 09 22 VES-MT-10 Emergency Air Storage C I ASME VIII, Appendix Tank 10 22 VES-MT-11 Emergency Air Storage C I ASME VIII, Appendix Tank 11 22 VES-MT-12 Emergency Air Storage C I ASME VIII, Appendix Tank 12 22 VES-MT-13 Emergency Air Storage C I ASME VIII, Appendix Tank 13 22 VES-MT-14 Emergency Air Storage C I ASME VIII, Appendix Tank 14 22 VES-MT-15 Emergency Air Storage C I ASME VIII, Appendix Tank 15 22 VES-MT-16 Emergency Air Storage C I ASME VIII, Appendix Tank 16 22 VES-MT-17 Emergency Air Storage C I ASME VIII, Appendix Tank 17 22 VES-MT-18 Emergency Air Storage C I ASME VIII, Appendix Tank 18 22 VES-MT-19 Emergency Air Storage C I ASME VIII, Appendix Tank 19 22 VES-MT-20 Emergency Air Storage C I ASME VIII, Appendix Tank 20 22 VES-MT-21 Emergency Air Storage C I ASME VIII, Appendix Tank 21 22 VES-MT-22 Emergency Air Storage C I ASME VIII, Appendix Tank 22 22 VES-MT-23 Emergency Air Storage C I ASME VIII, Appendix Tank 23 22 VES-MT-24 Emergency Air Storage C I ASME VIII, Appendix Tank 24 22 3.2-69 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 50 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Main Control Room Emergency Habitability System (Continued)
VES-MT-25 Emergency Air Storage C I ASME VIII, Appendix Tank 25 22 VES-MT-26 Emergency Air Storage C I ASME VIII, Appendix Tank 26 22 VES-MT-27 Emergency Air Storage C I ASME VIII, Appendix Tank 27 22 VES-MT-28 Emergency Air Storage C I ASME VIII, Appendix Tank 28 22 VES-MT-29 Emergency Air Storage C I ASME VIII, Appendix Tank 29 22 VES-MT-30 Emergency Air Storage C I ASME VIII, Appendix Tank 30 22 VES-MT-31 Emergency Air Storage C I ASME VIII, Appendix Tank 31 22 VES-MT-32 Emergency Air Storage C I ASME VIII, Appendix Tank 32 22 VES-PL-V001 Air Delivery Alternate Isolation C I ASME III-3 Valve VES-PL-V002A Pressure Regulating Valve A C I ASME III-3 VES-PL-V002B Pressure Regulating Valve B C I ASME III-3 VES-PL-V005A Air Delivery Main Isolation C I ASME III-3 Valve A VES-PL-V005B Air Delivery Main Isolation C I ASME III-3 Valve B VES-PL-V006A Air Delivery Line Pressure C I ASME III-3 Instrument Isolation Valve A VES-PL-V006B Air Delivery Line Pressure C I ASME III-3 Instrument Isolation Valve B VES-PL-V010A Air Delivery Line Maintenance C I ASME III-3 Isolation Valve A VES-PL-V010B Air Delivery Line Maintenance C I ASME III-3 Isolation Valve B VES-PL-V011A Air Delivery Line Maintenance C I ASME III-3 Isolation Valve A VES-PL-V011B Air Delivery Line Maintenance C I ASME III-3 Isolation Valve B VES-PL-V016 Temporary Instrument C I ASME III-3 Isolation Valve A VES-PL-V018 Temporary Instrument C I ASME III-3 Isolation Valve A VES-PL-V019 Temporary Instrument C I ASME III-3 Isolation Valve B VES-PL-V020 Temporary Instrument C I ASME III-3 Isolation Valve B VES-PL-V022A Pressure Relief Isolation C I ASME III-3 Valve A 3.2-70 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 51 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Main Control Room Emergency Habitability System (Continued)
VES-PL-V022B Pressure Relief Isolation C I ASME III-3 Valve B VES-PL-V024A Air Bank 1 Isolation Valve A C I ASME III-3 VES-PL-V024B Air Bank 2 Isolation Valve B C I ASME III-3 VES-PL-V024C Air Bank 3 Isolation Valve C C I ASME III-3 VES-PL-V024D Air Bank 4 Isolation Valve D C I ASME III-3 VES-PL-V025A Air Bank 1 Isolation Valve A C I ASME III-3 VES-PL-V025B Air Bank 2 Isolation Valve B C I ASME III-3 VES-PL-V025C Air Bank 3 Isolation Valve C C I ASME III-3 VES-PL-V025D Air Bank 4 Isolation Valve D C I ASME III-3 VES-PL-V026A Air Bank 1 Fill/Vent Isolation C I ASME III-3 Valve A VES-PL-V026B Air Bank 2 Fill/Vent Isolation C I ASME III-3 Valve B VES-PL-V026C Air Bank 3 Fill/Vent Isolation C I ASME III-3 Valve C VES-PL-V026D Air Bank 4 Fill/Vent Isolation C I ASME III-3 Valve D VES-PL-V040A Air Tank Safety Relief Valve A C I ASME III-3 VES-PL-V040B Air Tank Safety Relief Valve B C I ASME III-3 VES-PL-V040C Air Tank Safety Relief Valve C C I ASME III-3 VES-PL-V040D Air Tank Safety Relief Valve D C I ASME III-3 VES-PL-V043A Differential Pressure C I ASME III-3 Instrument Line Isolation Valve A VES-PL-V043B Differential Pressure C I ASME III-3 Instrument Line Isolation Valve B VES-PL-V044 Main Air Flowpath C I ASME III-3 Isolation Valve VES-PL-V045 Eductor Flow Path Isolation C I ASME III-3 Valve VES-PL-V046 Eductor Bypass Isolation C I ASME III-3 Valve VES-PY-N01 MCR Air Filtration Line C I ASME III-3 Eductor VES-PY-N02 Eductor Bypass Isolation C I ASME III-3 Discharge Silencer VES-PY-R02 Eductor Bypass Flow Orifice C I ASME III-3 VES-MY-F01 MCR Air Filtration Line Note 1 I ASME AG-1 Section Charcoal Filter FD VES-MY-F02 MCR Air Filtration Line HEPA Note 1 I ASME AG-1 Section Filter FC VES-MY-F03 MCR Air Filtration Line Note 1 1 ASME AG-1 Postfilter VES-MD-D001A Relief Damper A Note 1 I ASME N509/N510 VES-MD-D001B Relief Damper B Note 1 I ASME N509/N510 3.2-71 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 52 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Main Control Room Emergency Habitability System (Continued)
VES-MD-D002 MCR Air Filtration Line Supply Note 1 I ASME AG-1 Section Damper DA VES-MD-D003 MCR Air Filtration Line Supply Note 1 I ASME AG-1 Section Damper DA VES-MY-G001 Ducting Grill Note 1 I ASME N509/N510 VES-MY-R001 Ducting Register Note 1 I ASME N509/N510 VES-MY-R002 Ducting Register Note 1 I ASME N509/N510 VES-MY-R003 Ducting Register Note 1 I ASME N509/N510 VES-MY-Y01 MCR Air Filtration Line Note 1 I ASME AG-1 Section Silencer SA VES-MY-Y02 MCR Air Filtration Line Note 1 I ASME AG-1 Section Silencer SA Containment Air Filtration System (VFS) Location: Containment, Auxiliary Building, and Annex Building VFS-PY-C01 Containment Supply Duct B I ASME III, MC Penetration VFS-PY-C02 Containment Exhaust Duct B I ASME III, MC Penetration VFS-MY-Y01 Containment Air Supply C I ASME Sec. III Class Debris Screen 3 VFS-MY-Y02 Containment Air Exhaust C I ASME Sec. III Class Debris Screen 3 VFS-PL-V003 Containment Purge Supply B I ASME III-2 Containment Isolation Valve VFS-PL-V004 Containment Purge Supply B I ASME III-2 Containment Isolation Valve VFS-PL-V006 Containment Isolation Test C I ASME III-3 Connection VFS-PL-V007 RCS Ejector Discharge C I ASME III-3 Isolation VFS-PL-V008 Containment Isolation Test B I ASME III-2 Connection VFS-PL-V009 Containment Purge Discharge B I ASME III-2 Containment Isolation Valve VFS-PL-V010 Containment Purge Discharge B I ASME III-2 Containment Isolation Valve VFS-PL-V012 Containment Isolation Test B I ASME III-2 Connection VFS-PL-V015 Containment Isolation Test B I ASME III-2 Connection VFS-PL-V101 Containment Air Supply Line C I ASME III-3 Test Connection VFS-PL-V202 Containment Atmosphere to C I ASME III-3 Filtration Units Isolation VFS-PL-V587 Filtration Units to Containment C I ASME III-3 Atmosphere Manual Isolation 3.2-72 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 53 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Containment Air Filtration System (Continued)
VFS-PL-V800A Vacuum Relief Containment B I ASME III-2 Isolation A - ORC VFS-PL-V800B Vacuum Relief Containment B I ASME III-2 Isolation B - ORC VFS-PL-V803A Vacuum Relief Containment B I ASME III-2 Isolation Check Valve A - IRC VFS-PL-V803B Vacuum Relief Containment B I ASME III-2 Isolation Check Valve B - IRC n/a Valves Providing VFS AP1000 D NS ANSI B16.34 Equipment Class D Function n/a Shutoff, Balancing, Isolation, R NS ASME AG-1 and Pressure Differential Control Dampers n/a Shutoff and Balancing L NS ANSI/AMCA-500-D Dampers n/a Fire Dampers L NS UL-555 n/a Fire Damper R NS ASME AG-1 and UL-555 n/a Supply Air Handling Units L NS Manufacturer Std.
n/a Air Exhaust Filtration Units R NS ASME AG-1, ASME N509, Note 4 n/a Other Fans L NS AMCA n/a Ductwork in Auxiliary Building L II SMACNA n/a Ductwork in Annex Building L NS SMACNA n/a Ductwork in Containment R II ASME AG-1 n/a Ductwork in Annex Building R NS ASME AG-1, Note 4 Balance of system components are Class E, L and R Health Physics and Hot Machine Shop HVAC System (VHS) Location: Annex Building n/a Shutoff, Isolation, Balancing, L NS ANSI/AMCA-500-D and Backdraft Dampers n/a Shutoff Dampers R NS ASME AG-1 n/a Fire Dampers L NS UL-555 n/a Air Handling Units w/ Filters L NS Manufacturer Std.
n/a Ductwork L NS SMACNA n/a Ductwork R NS ASME AG-1 n/a Fans L NS AMCA Balance of system components are Class E or Class L Containment Hydrogen Control System (VLS) Location: Containment n/a Hydrogen Igniters D NS Manufacturer Std. Provides Hydrogen Control Following Severe Accidents VLS-MY-E01A Catalytic Hydrogen D NS Manufacturer Std.
Recombiner A VLS-MY-E01B Catalytic Hydrogen D NS Manufacturer Std.
Recombiner B Balance of system components are Class E or Class L 3.2-73 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 54 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Radwaste Building Ventilation System (VRS) Location: Annex Building and Radwaste Building n/a Shutoff, Isolation, Balancing, L NS ANSI/AMCA-500-D and Backdraft Dampers n/a Shutoff Dampers R NS ASME AG-1 n/a Fire Damper R NS ASME AG-1 &
UL-555 n/a Air Handling Units L NS Manufacturer Std.
n/a Filters L NS UL 900 n/a Ductwork L NS SMACNA n/a Ductwork R NS ASME AG-1 n/a Fans L NS AMCA Balance of system components are Class E or Class L Turbine Building Ventilation System (VTS) Location: Turbine Building n/a Shutoff, Isolation, Balancing, L NS ANSI/AMCA-500-D and Backdraft Dampers n/a Fire Dampers L NS UL-555 n/a Air Handling Units w/ Filters L NS Manufacturer Std.,
UL-900 n/a Ductwork L NS SMACNA n/a Fans L II AMCA n/a Fire Damper L II UL-555 Balance of system components are Class E or Class L Containment Leak Rate Test System (VUS) Location: Containment and Auxiliary Building VUS-PL-V013 Main Equipment Hatch Test B I ASME III-2 Connection VUS-PL-V014 Maintenance Equipment B I ASME III-2 Hatch Test Connection VUS-PL-V015 Main Equipment Hatch Test B I ASME III-2 Connection VUS-PL-V016 Maintenance Equipment B I ASME III-2 Hatch Test Connection VUS-PL-V017 Personnel Hatch Test B I ASME III-2 Connection VUS-PL-V018 Personnel Hatch Test B I ASME III-2 Connection VUS-PL-V019 Personnel Hatch Test B I ASME III-2 Connection VUS-PL-V020 Personnel Hatch Test B I ASME III-2 Connection VUS-PL-V021 Personnel Hatch Test B I ASME III-2 Connection VUS-PL-V022 Personnel Hatch Test B I ASME III-2 Connection VUS-PL-V023 Fuel Transfer Tube Test B I ASME III-2 Connection 3.2-74 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 55 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Containment Leak Rate Test System (Continued)
VUS-PL-V140 Spare Penetration Test B I ASME III-2 Connection VUS-PL-V141 Spare Penetration Test B I ASME III-2 Connection VUS-PL-V142 Spare Penetration Test B I ASME III-2 Connection Balance of system components are Class E Central Chilled Water System (VWS) Location: Various VWS-MS-02 Air-Cooled Chiller 2 D NS ARI/ASME VIII VWS-MS-03 Air-Cooled Chiller 3 D NS ARI/ASME VIII n/a Pumps D NS Manufacturer Std.
n/a Tanks D NS ASME VIII n/a Valves Providing VWS D NS ANSI B16.34 AP1000 Equipment Class D Function VWS-PY-C01 Containment Chilled Water B I ASME III, MC Return Penetration VWS-PY-C02 Containment Chilled Water B I ASME III, MC Supply Penetration VWS-PL-V053 VWS Containment C I ASME III-3 Penetration Thermal Relief Valve VWS-PL-V057 VWS Containment C I ASME III-3 Penetration Thermal Relief Valve VWS-PL-V058 Fan Coolers Supply B I ASME III-2 Containment Isolation VWS-PL-V062 Fan Coolers Supply B I ASME III-2 Containment Isolation Check Valve VWS-PL-V080 VWS Containment Isolation B I ASME III-2 Relief Valve VWS-PL-V082 Fan Coolers Return B I ASME III-2 Containment Isolation VWS-PL-V086 Fan Coolers Return B I ASME III-2 Containment Isolation VWS-PL-V424 Containment Penetration Test B I ASME III-2 Connection VWS-PL-V425 Containment Penetration Test B I ASME III-2 Connection Balance of system components are Class E 3.2-75 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 56 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Annex/Auxiliary Nonradioactive Ventilation System (VXS) Location: Auxiliary Building and Annex Building n/a Air Handling Unit Fans Note 2 NS AMCA Providing AP1000 Equipment Class D Function n/a Dampers Providing VXS Note 2 NS ANSI/AMCA-500-D AP1000 Equipment Class D Function n/a Exhaust Fan Providing Note 2 NS AMCA Ancillary Diesel Room Ventilation n/a Fire Dampers L NS UL-555 n/a Fire Dampers R NS ASME AG-1 and UL-555 n/a Combination Fire/Smoke L NS UL-555 and UL-555S Dampers n/a Combination Fire/Smoke R NS ASME AG-1, Dampers UL-555, and UL-555S n/a Air Handling Units L NS Manufacturer Std.
n/a Filters L NS UL 900 n/a Ductwork L NS SMACNA n/a Ductwork in Auxiliary Building L II SMACNA except ductwork attached to mechanical modules n/a Ductwork R NS ASME AG-1 n/a Fans L NS AMCA Balance of system components are Class E or Class L Hot Water Heating System (VYS) Location: Various System components are Class E Diesel Generator Building Ventilation System (VZS) Location: Diesel Generator Building n/a Unit Heaters Providing Note 3 NS UL-2021; NFPA 70 AP1000 Equipment Class D Function n/a Fans Providing AP1000 Note 2 NS AMCA Equipment Class D Function n/a Dampers Providing VZS Note 2 NS ANSI/AMCA 500-D AP1000 Equipment Class D Function n/a Fire Dampers L NS UL-555 n/a Air Handling Units L NS Manufacturer Std.
n/a Filters L NS UL 900 n/a Ductwork L NS SMACNA n/a Fans L II AMCA n/a Fire Damper L II UL-555 Balance of system components are Class E or Class L 3.2-76 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 57 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Gaseous Radwaste System (WGS) Location: Auxiliary Building n/a Gas Cooler D NS Manufacturer Std.
n/a Sample Pumps D NS Manufacturer Std.
n/a Guard and Delay Beds D NS ASME VIII Design for 1/2 SSE n/a Moisture Separator D NS ASME VIII n/a Valves Providing WGS D NS ANSI B16.34 AP1000 Equipment Class D Function Liquid Radwaste System (WLS) Location: Containment, Auxiliary, and Radwaste Buildings n/a Heat Exchangers, WLS and D NS ASME VIII/ TEMA CCS Side n/a Pumps D NS Manufacturer Std.
n/a Tanks (except WLS-MT-13A, D NS ASME III without WLS-MT-13B, WLS-MT-17, Code Stamp WLS-MT-23A, WLS-MT-23B WLS-MT-13A Waste Holdup Tank A D NS ASME VIII-1 Chemical Addition Pot WLS-MT-13B Waste Holdup Tank B D NS ASME-VIII-1 Chemical Addition Pot WLS-MT-17 Chemical Waste Tank D NS ASME VIII-1 Chemical Addition Pot WLS-MT-23A WLS Leak Chase Collection D NS ASME VIII-1 Pot A WLS-MT-23B WLS Leak Chase Collection D NS ASME VIII-1 Pot B n/a Degasifier D NS ASME VIII n/a Ion Exchangers D NS ASME VIII n/a Filters D NS ASME VIII n/a Valves Providing WLS D NS ANSI B16.34 AP1000 Equipment Class D Function (local drain valves in Radwaste Building) n/a Floor Drain Hubs D NS Manufacturer Std.
WLS-MT-02 Containment Sump D NS ACI 349 ACI 349 Evaluation of Structural Boundary Only WLS-PL-V055 Sump Discharge Containment B I ASME III-2 Isolation IRC WLS-PL-V057 Sump Discharge Containment B I ASME III-2 Isolation ORC WLS-PL-V058 WLS Containment Isolation B I ASME III-2 Relief Valve WLS-PL-V067 RCDT Gas Outlet B I ASME III-2 Containment Isolation IRC WLS-PL-V068 RCDT Gas Outlet B I ASME III-2 Containment Isolation ORC WLS-PL-V071A CVS Compartment to Sump C I ASME III-3 WLS-PL-V071B PXS A Compartment to Sump C I ASME III-3 3.2-77 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 58 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Liquid Radwaste System (Continued)
WLS-PL-V071C PXS B Compartment to Sump C I ASME III-3 WLS-PL-V072A CVS Compartment to Sump C I ASME III-3 WLS-PL-V072B PXS A Compartment to Sump C I ASME III-3 WLS-PL-V072C PXS B Compartment to Sump C I ASME III-3 WLS-PY-C02 Reactor Coolant Drain Tank B I ASME III, MC WLS Connection Penetration WLS-PY-C03 Containment Sump Pumps B I ASME III, MC Combined Discharge Penetration WLS-MY-Y34 Containment Sump Level D I Manufacturer Std.
Instrument Stilling Well WLS-MY-Y35 Containment Sump Level D I Manufacturer Std.
Instrument Stilling Well WLS-MY-Y36 Containment Sump Level D I Manufacturer Std.
Instrument Stilling Well Balance of system components are Class E Radioactive Waste Drain System (WRS) Location: Auxiliary Building n/a Pumps D NS Manufacturer Std.
n/a Valves Providing WRS D NS ANSI B16.34 AP1000 Equipment Class D Function n/a Floor Drain Hubs D NS Manufacturer Std.
WRS-MT-02A WRS Leak Chase Collection D NS ASME VIII-1 Pot A WRS-MT-02B WRS Leak Chase Collection D NS ASME VIII-1 Pot B Solid Radwaste System (WSS) Location: Auxiliary Building n/a Pumps D NS Manufacturer Std.
n/a Tanks D NS ASME VIII n/a Filters D NS ASME VIII n/a Valves Providing WSS D NS ANSI 16.34 AP1000 Equipment Class D Function Balance of system components are Class E Waste Water System (WWS) Location: Various WWS-PL-V506 MCR WWS Isolation Valve C I ASME III-3 Balance of system components are Class E 3.2-78 Revision 10
VEGP 3&4 - UFSAR Table 3.2-3 (Sheet 59 of 59)
AP1000 Classification of Mechanical and Fluid Systems, Components, and Equipment AP1000 Seismic Principal Con-Tag Number Description Class Category struction Code Comments Onsite Standby Power System (ZOS) Location: Diesel Generator Building n/a Diesel Generator Engines D NS Manufacturer Std.
n/a Diesel Generator Starting D NS CAGI Units n/a Diesel Generator Radiators D NS API 661 n/a Diesel Generator Silencers D NS Manufacturer Std.
n/a Valves Providing ZOS Diesel D NS ANSI 16.34 Generator Engines AP1000 Equipment Class D Function Balance of system components are Class E Notes:
- 1. Component performs a safety-related function equivalent to AP1000 equipment Class C. The component is constructed using the standards for Class R and a quality assurance program in conformance with 10 CFR Part 50 Appendix B.
- 2. Component performs an AP1000 equipment Class D function and is constructed using the standards for Class L or Class R.
- 3. Component performs an AP1000 Equipment Class D function and is constructed using the standards for Class E.
- 5. Not used.
- 6. The classification of NS for structures, systems and components is upgraded to seismic Category II if the criterion in Subsection 3.2.1.1.3 is met.
- 7. For specific SMACNA and AMCA standards, see the descriptions in Section 9.4.
Table 3.2-201 Not Used 3.2-79 Revision 10
VEGP 3&4 - UFSAR 3.3 Wind and Tornado Loadings 3.3.1 Wind Loadings The wind loadings for seismic Category I structures are in accordance with American Society of Civil Engineers, "Minimum Design Loads for Buildings and Other Structures," ASCE 7-98 (Reference 1).
3.3.1.1 Design Wind Velocity The design wind is specified as a basic wind speed of 145 mph with an annual probability of occurrence of 0.02 based on the most severe location identified in Reference 1 This wind speed is the 3 second gust speed at 33 feet above the ground in open terrain (Reference 1, exposure C). The basic wind speed of 145 mph is the 3 second gust speed that has become the basis of wind design codes since 1995. It corresponds to the 110 mph fastest mile wind used as the basis for the AP600 design in accordance with the 1988 edition of Reference 1.
Higher winds with a probability of occurrence of 0.01 are used in the design of seismic Category I structures by using an importance factor of 1.15. This is obtained by classifying the AP1000 seismic Category I structures as essential facilities and using the design provisions for Category IV of Reference 1.
Velocity pressure exposure coefficients and gust response factors are calculated according to Reference 1 for exposure C, which is applicable to shorelines in hurricane prone areas in the 1998 edition of Reference 1. The topographic factor is taken as unity.
The design wind loads calculated as described above exceed those required at other locations in the United States, where the more severe Exposure Category D is specified in Reference 1. Exposure Category D is applicable for sites near the open inland waterways, the Great Lakes, and the coastal areas of California, Oregon, Washington, and Alaska. For such locations, the basic wind speed is less than 130 mph.
The wind velocity characteristics for the Vogtle Electric Generating Plant, Units 3 and 4 (VEGP), are given in Subsection 2.3.1.3.1. These values are bounded by the design wind velocity values given in Subsection 3.3.1.1 for the AP1000 plant.
3.3.1.2 Determination of Applied Forces The procedures used in transforming the wind velocity into an effective pressure to be applied to structures and parts and portions of structures follow the guidelines of Reference 1.
Effective pressures applied to interior and exterior surfaces of the buildings and corresponding shape coefficients are calculated according to Reference 1 for exposure C. Shape coefficients, defining the variation around the circumference of the shield building, are calculated using ASCE Paper No. 3269 (Reference 2). These shape coefficients are consistent with those observed in the model tests described in Reference 6.
3.3.2 Tornado Loadings Seismic Category I structures are designed to resist tornado wind loads without exceeding the allowable stresses defined in Subsection 3.8.4. These tornado loads exceed the loads for hurricanes with a probability of occurrence comparable to that of the tornado. In addition, seismic Category I structures are designed to remain functional when subjected to tornado-generated missiles as discussed in Subsection 3.5.1.4. Seismic Category I structures are permitted to sustain local missile damage such as partial penetration and local cracking or permanent deformation or both, provided 3.3-1 Revision 5
VEGP 3&4 - UFSAR that structural integrity is maintained and seismic Category I systems, components, and equipment required to function during or after passage of a tornado are not subject to damage by secondary missiles, such as from concrete spalling. See Subsection 3.5.2.
3.3.2.1 Applicable Design Parameters The design parameters applicable to the design basis tornado are as follows:
Maximum wind speed - 300 mph Maximum rotational speed - 240 mph Maximum translational speed - 60 mph Radius of maximum rotational wind from center of tornado - 150 ft Atmospheric pressure drop - 2.0 psi Rate of pressure change - 1.2 psi/sec It is estimated that the probability of wind speeds greater than the design basis tornado is between 10-6 and 10-7 per year for an AP1000 at a "worst location" anywhere within the contiguous United States.
The tornado characteristics for the VEGP are given in Subsection 2.3.1.3.2. These values are bounded by the tornado design parameters given in Subsection 3.3.2.1 for the AP1000 plant.
3.3.2.2 Determination of Forces on Structures The procedures described in Subsection 3.3.1.2 are used to transform the tornado wind loading and differential pressure loading into effective loads on structures, with a wind velocity of 300 mph (translational plus rotational velocities). The dynamic wind pressure is applied to the structures in the same manner as the wind loads described in Subsection 3.3.1.2, except that the importance factor, gust factor, and the variation of wind speed with height do not apply. Loading combinations and load factors used are as follows:
Wt = Ww Wt = Wp Wt = Wm Wt = Ww + 0.5 Wp Wt = Ww + Wm Wt = Ww + 0.5 Wp + Wm where:
Wt = total tornado load Ww = total wind load Wp = total differential pressure load Wm = total missile load 3.3-2 Revision 5
VEGP 3&4 - UFSAR The maximum pressure drop of 2.0 psi, applicable to a nonvented structure, is used for Wp for all structures except the upper portion of the shield building. The portion of the shield building surrounding the upper annulus is designed as fully vented (zero differential pressure) due to the large area of the air inlets and discharge stack. Figure 3.3-1 shows the velocity pressure variation with the radius from the center of the tornado. When the tornado loading includes the missile load, the structure locally may go into the plastic range because of missile impact. Subsection 3.5.3 discusses the procedure for analyzing local missile effects.
3.3.2.3 Effect of Failure of Structures or Components Not Designed for Tornado Loads The failure of structures not designed for tornado loadings does not affect the capability of seismic Category I structures or safety-related systems performance. This is accomplished by one of the following:
Designing the adjacent structure to seismic Category I structure tornado loading Investigating the effect of adjacent structure failure on seismic Category I structures to determine that no impairment of function results Designing a structural barrier to protect seismic Category I structures from adjacent structural failure.
The structures adjacent to the nuclear island are the annex building, the radwaste building, and the turbine building.
The portions of the annex and turbine buildings adjacent to the nuclear island are classified as seismic Category II and are designed to seismic Category I structure tornado loading. The acceptance criteria are based on ACI 349 for concrete structures, on ANSI/AISC N690 for steel structures, including the supplemental requirements described in Subsections 3.8.4.4.1 and 3.8.4.5, and on AWS D1.1-2000 for weld design, qualification, fabrication, and inspection as described in Subsections 3.8.3.2 and 3.8.4.2. The structures are constructed to the same requirements as nonseismic structures, ACI 318 for concrete structures, and AISC Specification for Structural Steel Buildings, Allowable Stress Design and Plastic Design June 1989 for steel structures. Siding is permitted to blow off during the tornado.
The radwaste building is a small steel-frame building. If it were to collapse in the tornado, it would not impair the integrity of the reinforced concrete nuclear island.
The main area of the turbine building is classified as nonseismic and is designed to seismic Category I structure tornado loading. The acceptance criteria for tornado loading are based on ACI 318 for concrete structures using a load factor of 1.0 and on 1.7 times the AISC 360 allowables for steel structures. Siding is permitted to blow off during the tornado.
Consideration of the effects of wind and tornado due to failures in an adjacent AP1000 plant and VEGP Units 1 and 2 are bounded by the evaluation of the buildings and structures in a single unit.
3.3.2.4 Tornado Loads on the Passive Containment Cooling System Air Baffle The containment air baffle is located within the annulus between the containment vessel and the shield building. It interfaces with the passive containment cooling system and separates downward flowing air entering at the air intake openings at the top of the cylindrical portion of the shield building from upward flowing air that cools the containment vessel and flows out of the discharge diffuser.
3.3-3 Revision 5
VEGP 3&4 - UFSAR Loads due to the atmospheric pressure drop (Wp) are calculated assuming the tornado is centered over the containment. Differential pressure between the air intakes and the discharge is calculated based on the radius of the shield building and the parameters of the tornado defined in Subsection 3.3.2.1. The differential pressure is used with the pressure loss coefficients in the air flow path to determine pressures throughout the flow path.
The development of loads on the air baffle due to the design wind and tornado (Ww) are described in the test reports (References 3, 4, and 5). Models of the AP600 were tested in a wind tunnel and subjected to representative wind profiles. Pressures were measured on each side of the baffle, and the differential pressures were normalized to the input wind velocity. The pressure coefficients are applied to the effective dynamic pressure for the design wind and the tornado to obtain the wind loads across the baffle. The tornado wind is specified to be constant with height. The tornado loads calculated for the AP600 are applicable to the AP1000. The AP1000 configuration is similar to the AP600. The height of the shield building roof increases by 20 6"; the exterior diameter of the passive containment cooling storage tank increases from 80 0" to 89 0". The pressure coefficients measured in the AP600 tests are not significantly affected by these changes in geometry.
Wind conditions result in a pressure reduction in the annulus between the shield building and the containment vessel as well as above the containment dome. This reduced pressure is equivalent to an increase in containment internal pressure and is within the normal operating range for containment pressure (-0.2 to 1.0 psig).
Wind conditions result in a small wind load across the containment vessel. This is maximum opposite the air intakes where positive pressures occur on the windward side and negative pressures occur on the leeward side. Lateral loads on the containment vessel are developed in Reference 5.
3.3.3 Combined License Information The site interface criteria for wind and tornado are addressed in APP-GW-GLR-020 (Reference 7).
The VEGP site satisfies the site interface criteria for wind and tornado (see Subsections 3.3.1.1, 3.3.2.1 and 3.3.2.3) and will not have a tornado-initiated failure of structures and components within the applicants scope that compromises the safety of AP1000 safety-related structures and components (see also Subsection 3.5.4).
Subsection 1.2.2 discusses differences between the plant specific site plan and the AP1000 typical site plan.
There are no other structures adjacent to the nuclear island other than as described and evaluated in the DCD.
Missiles caused by external events separate from the tornado are addressed in Subsections 3.5.1.3, 3.5.1.5, and 3.5.1.6.
3.3.4 References
- 1. American Society of Civil Engineers, "Minimum Design Loads for Buildings and Other Structures," ASCE 7-98.
- 2. ASCE Paper No. 3269, "Wind Forces on Structures," Transactions of the American Society of Civil Engineers, Vol. 126, Part II (1961).
- 3. WCAP-13323-P and WCAP-13324-NP, "Phase II Wind Tunnel Testing for the Westinghouse AP600 Reactor," August 1992.
3.3-4 Revision 5
- 4. WCAP-14068-P, "Phase IVA Wind Tunnel Testing for the Westinghouse AP600 Reactor,"
May, 1994.
- 5. WCAP-14169-P, "Phase IVA Wind Tunnel Testing for the Westinghouse AP600 Reactor, Supplemental Report," September, 1994.
- 6. WCAP-13294-P and WCAP-13295-NP, "Phase I Wind Tunnel Testing for the Westinghouse AP600 Reactor," April 1992.
- 7. APP-GW-GLR-020, "Wind and Tornado Site Interface Criteria," Westinghouse Electric Company LLC.
- 8. American Society of Civil Engineers, "Minimum Design Loads for Buildings and Other Structures," ASCE 7-05.
- 9. AISC 360, "Specification for Structural Steel Buildings," March 9, 2005.
3.3-5 Revision 5
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U5P Figure 3.3-1 Velocity Pressure Variation with Radius from Center of Tornado 3.3-6 Revision 5
VEGP 3&4 - UFSAR 3.4 Water Level (Flood) Design External flooding of a nuclear power plant from natural causes can be attributed to probable maximum flood, site and adjacent area probable maximum precipitation runoff, seiche, and ground water. Criteria for the design basis flood are in accordance with the provisions of Regulatory Guide 1.59, Design Basis Floods for Nuclear Power Plants, and Regulatory Guide 1.102, Flood Protection for Nuclear Power Plants. Conformance with the Regulatory Guides is described in Section 1.9.
External events are described in Section 2.4. Chapter 2 provides interface data for AP1000 which has an interface flood level at plant grade.
Internal plant flooding can be attributed to piping ruptures, tank failures, or the actuation of fire suppression systems.
3.4.1 Flood Protection 3.4.1.1 Flood Protection Measures for Seismic Category I Structures, Systems, and Components The seismic Category I structures, systems, and components identified in Section 3.2 are designed to withstand the effects of flooding due to natural phenomena or postulated component failures. A description of the structures is provided in Subsections 3.8.2, 3.8.3, and 3.8.4. None of the nonsafety-related structures, systems and components were found to be important based on flooding considerations. As a result, nonsafety-related structures, systems and components are not important in mitigation of flood events and are not required to be protected from either internal or external flooding.
3.4.1.1.1 Protection from External Flooding The probable maximum flood for the AP1000 has been established at less than plant elevation 100 as discussed previously in Section 2.4. The probable maximum flood results from site specific events, such as river flooding, upstream dam failure, or other natural causes.
Flooding does not occur from the probable maximum precipitation. The roofs do not have internal roof drains. The annex, radwaste, and diesel generator buildings have parapets with large openings to drain to scuppers/drains to preclude accumulation of water on the roofs. The roofs are sloped such that rainfall is directed towards gutters located along the edges of the roofs. Therefore, ponding of water on the roofs is precluded. Water from roof drains and/or scuppers, as well as runoff from the plant site and adjacent areas, is conveyed to catch basins, underground pipes, or directly to open ditches by sloping the tributary surface area. The site is graded to offer protection to the seismic Category I structures.
The high ground water table interface is at two feet below the grade elevation, as discussed in Section 2.4.
The components that may be potential sources for external flooding are nonsafety-related, nonseismic tanks as shown in Figure 1.2-2:
Fire water tanks as described in Subsection 9.5.1. These two tanks have maximum usable volumes of approximately 504,000 gallons each, and are located at the north end of the turbine building. Water will drain from the tanks away from the nuclear island and adjacent buildings due to the required site grading.
3.4-1 Revision 8
VEGP 3&4 - UFSAR Condensate storage tank as described in Subsection 9.2.4. This tank has a volume of approximately 656,000 gallons, and is located at the west side of the turbine building. Water will drain from the tank away from the turbine and auxiliary buildings due to site grading.
Demineralized water tank as described in Subsection 9.2.4. This tank has a volume of 126,000 gallons and is located adjacent to the annex building at elevation 107-2. Water will drain from the tank away from the annex building to elevation 100-0. Nearby doors lead to areas in the annex building which do not contain safety-related components or systems.
Boric acid storage tank as described in Subsection 9.3.6. This tank has a volume of 79,315 gallons and is located adjacent to the demineralized water storage tank.
Diesel fuel oil tanks as described in Subsection 9.5.4. These two tanks have volumes of 94,000 gallons each. They are located remote from safety-related structures and are provided with dikes to retain leaks and spills.
Passive containment cooling ancillary water storage tank, which is seismic category II, as described in Subsection 6.2.2.2.3. This tank has a volume of 1,080,027 gallons and is located at the west side of the auxiliary building. Water will drain from the tank away from the auxiliary building due to site grading.
In addition, failure of the cooling tower or the service water or circulating water piping under the yard could result in a potential flood source. However, these potential sources are located far from safety-related structures and the consequences of a failure in the yard would be enveloped by the analysis described in Subsection 10.4.5.
For the AP1000, the 100-0 building floor elevations are slightly above the grade elevation. In addition, the slope of the yard grade directs water away from the buildings. Because the probable maximum flood for AP1000 is less than grade elevation, the exterior doors are not required to be watertight for protection from external flooding.
Process piping penetrations through the exterior walls of the nuclear island below grade are embedded in the wall or are welded to a steel sleeve embedded in the wall. There are no access openings or tunnels penetrating the exterior walls of the nuclear island below grade.
The reinforced concrete seismic Category I structures, incorporating the waterproofing and sealing features described above and in Subsection 3.4.1.1.1.1, provide hardened protection for safety-related structures, systems, and components as defined in Regulatory Guide 1.59.
3.4.1.1.1.1 Waterproofing A waterproof membrane or waterproofing system for the seismic Category I structures below grade will be installed as an architectural aide to limit the infiltration of subsurface water. The COL applicant will use a waterproofing system for foundation mat (mudmat) and the below grade exterior walls exposed to flood and groundwater that will demonstrate a friction coefficient 0.55 with all horizontal concrete surfaces. This friction coefficient is maintained for the life expectancy of the plant and will not introduce a horizontal slip plane increasing the potential for movement during an earthquake (see Subsection 3.8.5.5.3). Typical waterproofing approaches are described as follows:
HDPE Double-Sided Textured Waterproof Membrane Figures 3.4-1and 3.4-2 show the typical application of this waterproofing approach for a mechanically stabilized earth (MSE) wall and for a step-back configuration.
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VEGP 3&4 - UFSAR HDPE Single-Sided Self-Adhering Sheet Waterproofing Membrane The HDPE single-sided adhesive sheet membrane is interchangeable with the HDPE double-sided textured membrane for use in the mudmat, and it may be used in certain circumstances to waterproof the walls.
Self-adhesive, Rubberized Asphalt/Polyethylene Waterproofing Membrane The self-adhesive rubberized membrane is for application to waterproof the walls only.
Sprayed-on Waterproofing Membrane This method may be used either for soil sites, in conjunction with an MSE wall, or for rock sites, where an open excavation may be used. The membrane consists of 100-percent solids materials based on polymer-modified asphalt or polyurea. This system may include a polyester reinforcement fabric having properties necessary to meet the performance requirements for the system. Figure 3.4-4 shows the typical installation using MSE walls with the sprayed-on, liquid-applied waterproofing membrane placed on the MSE wall panels and between the two layers of the mudmat. Where the vertical face of excavation is used as a form for the exterior walls, the waterproof membrane is installed on the vertical face of the excavation prior to placement of concrete in the exterior walls. An alternate waterproofing system for the seismic Category I structures below grade is as presented in Subsection 3.8.5.1.1.
HDPE Waterproof Membrane System The HDPE waterproof membrane system uses an embedded HDPE Liner, with additional rubberized waterproof sealant materials forming a transition to the waterproof membrane on the MSE walls. It may be used for vertical waterproofing applications on the NI wall.
The waterproof function of the membrane is not safety-related; however, the membrane between the mudmats must provide adequate shear strength to transfer horizontal shear forces due to seismic (SSE) loading. This function is seismic Category I. The waterproof membrane will have physical properties, including surface and texture, to achieve the required coefficient of friction. Primer or geotextile may be added as required.
3.4.1.1.2 Protection from Internal Flooding The nuclear island general arrangement drawings provided in Section 1.2 are a useful reference for the internal flooding discussion.
The AP1000 arrangement provides physical separation of redundant safety-related components and systems from each other and from nonsafety-related components. As a result, component failures resulting from internal flooding do not prevent safe shutdown of the plant or prevent mitigation of the flooding event. Protection mechanisms are generally described in Subsection 3.6.1.3.2. The protection mechanisms related to minimizing the consequences of internal flooding include the following:
Structural enclosures Structural barriers Curbs, weirs and elevated thresholds Watertight hatches Leak detection systems Drain systems Flood relief louvers 3.4-3 Revision 8
VEGP 3&4 - UFSAR Pressure relief panels Flappers The AP1000 minimizes the number of penetrations through enclosure or barrier walls below the flood level. Those few penetrations through flood protection walls that are below the maximum flood level are watertight. Any process piping penetrating below the maximum flood level either is embedded in the wall or floor or is welded to a steel sleeve embedded in the wall or floor. There are no watertight doors in the AP1000 used for internal flood protection because, as described in Subsection 3.4.1.2.2, they are not needed to protect safe shutdown components from the effects of internal flooding.
Watertight hatches are provided to close openings not required for hydrogen venting. The walls, floors, hatches, and penetrations are designed to withstand the maximum anticipated hydrostatic loads associated with a pipe failure as described in Section 3.6. The two watertight doors on the waste holdup tank compartments limit the consequence of a failure on spent fuel pool water level.
3.4.1.2 Evaluation of Flooding Events 3.4.1.2.1 External Flooding Base mat and exterior walls of seismic Category I structures are designed to resist upward and lateral pressures caused by the probable maximum flood and high ground water level. The vertical hydrostatic pressure acting uniformly at the bottom of the base mat is the product of the height to the high water level and the unit weight of water assumed as 62.4 lb/ft3. The horizontal hydrostatic pressure acting on the exterior walls varies with height, with the maximum at the bottom of the wall and zero at the maximum water level. Minimum factors of safety for overturning, sliding, and flotation are described in Subsection 3.8.5. There are no dynamic water forces associated with the probable maximum flood or high ground water level because they are below the finished grade. Dynamic forces associated with the probable maximum precipitation are not factors in the analysis or design since the finished grade is adequately sloped.
There are no safety-related hydraulic structures for AP1000.
3.4.1.2.2 Internal Flooding This section describes the consequences of compartment flooding for various postulated component failures. The equipment required to achieve and maintain safe shutdown depends on the initiating event. The safety-related systems and components available for safe shutdown are described in Section 7.4. This equipment is located in the auxiliary building and inside containment. Except for floor drains, no credit is taken in this evaluation for the availability of nonsafety-related systems or components.
Each area of the plant containing safety-related systems or equipment is reviewed to determine the postulated fluid system failures which would result in the most adverse internal flooding conditions.
For the internal flooding analysis, the failure of safety-related systems, structures or components is acceptable provided they have no safe shutdown function or the safe shutdown function is otherwise accomplished. The internal flooding analysis shows that systems, structures, and components are not prevented from performing their required safe shutdown functions due to the effects of the postulated failure. In addition, the analysis identifies the protection features that mitigate the consequences of flooding in an area that contains safety-related equipment.
The flooding sources considered in the analysis consist of the following:
High-energy piping (breaks and cracks)
Through-wall cracks in seismically-supported moderate energy piping Breaks and through-wall cracks in non-seismically-supported moderate energy piping 3.4-4 Revision 8
VEGP 3&4 - UFSAR Pump mechanical seal failures Storage tank ruptures Actuation of fire suppression systems Flow from upper elevations, lower elevations and adjacent areas The analysis is performed based on the criteria and assumptions provided in Section 3.6 and ANS-56.11 (Reference 1). Section 3.6 provides the criteria used to define break and crack locations and configurations for high and moderate-energy piping failures. Additional design criteria pertaining to the internal flooding analysis are provided in this section.
The analysis consists of the following steps:
Identification of the flood sources Identification of essential equipment in area Determination of flowrates and flood levels Evaluation of effects on essential equipment As stated in Section 3.6, high-energy ASME Code Class 1, 2, and 3 piping of 6 inch nominal diameter or larger inside the containment is evaluated for mechanistic pipe break (leak-before-break) for AP1000. Those high-energy piping systems that do not satisfy the mechanistic pipe break requirements inside containment and high-energy lines outside containment are evaluated for non-mechanistic breaks and cracks, as above.
Fluid flowrates from high- and moderate-energy piping ruptures are determined based on the criteria provided in Section 3.6 and ANSI 56.11 (Reference 1). Fluid flowrates through stairwells, floor openings, and floor sleeves are determined in accordance with the formulas given in Reference 1.
No breaks or cracks are assumed for piping with nominal diameters of 1 inch or less. For each storage tank rupture, it is assumed that the entire tank inventory is drained.
The analysis of potential flooding events is performed on a floor-by-floor and room-by-room basis depending upon the relative location of safety-related equipment. No credit is taken for operation of sump pumps to mitigate the consequences of flooding.
3.4.1.2.2.1 Containment Flooding Events General The safe shutdown systems and components located inside the containment are associated with the passive core cooling system (PXS), the automatic depressurization system (ADS), and containment isolation.
The evaluation of containment flooding events addresses the impact of flooding on the safe shutdown systems and components. The AP1000 passive core cooling system, the internal containment compartments, and the equipment locations are designed for internal flooding to maintain post accident long-term cooling flow to the reactor core from the flooded volumes.
In the unlikely event of a loss-of-coolant accident (LOCA), the combined water inventory from available sources within the containment is sufficient to flood the reactor and steam generator compartments to a level above the reactor coolant system piping to provide water flow back into the reactor coolant system via the break location or via the passive core cooling system containment recirculation subsystem (see Section 6.3) flow path.
The potential for flooding safe shutdown components inside containment that would be required to perform safe shutdown functions is limited to two equipment compartments. These compartments 3.4-5 Revision 8
VEGP 3&4 - UFSAR are located in the southeast and northeast quadrants of the containment below the floor at elevation 107-2. For flood evaluation, these compartments extend up to the top of the curbs through the openings in the floor. These two compartments contain passive core cooling system components that provide two redundant means for delivering borated water to the reactor coolant system when required for safe shutdown.
The two passive core cooling system compartments primarily contain passive core cooling system components. The southeast compartment is referred to as the PXS-A compartment. The northeast compartment is interconnected with the RNS valve room by a passageway. These two compartments are treated, in this discussion, as one floodable volume referred to as the PXS-B compartment. The principal passive core cooling system component in each passive core cooling system compartment is an accumulator. A passive core cooling system core makeup tank is located above each passive core cooling system compartment. Each passive core cooling system compartment also contains isolation valves for the accumulator, the core makeup tank, the in-containment refueling water storage tank, and the passive core cooling system containment recirculation subsystem line.
There are eight automatically actuated containment isolation valves inside containment subject to flooding. These normally closed containment isolation valves are not required to operate during a safe shutdown operation and they would not fail open as a result of the compartment flooding. Also, there is a redundant, normally closed, containment isolation valve located outside containment in series with each of these valves.
The PXS-A compartment contains one normally closed spent fuel pool cooling system containment isolation valve and one normal residual heat removal system containment isolation valve. The PXS-B compartment contains four normally closed normal residual heat removal system containment isolation valves. The maintenance floor contains two normally closed liquid radwaste system containment isolation valves located partially below the maximum flood level.
Except for the valves mentioned above, the rest of the automatically actuated containment isolation valves are located above the maximum flood level; therefore, these components would not be adversely affected by postulated flooding.
Flooding can be postulated from a failure of several systems located inside the containment. The worst case flooding scenario is a LOCA. The maximum flood level for a LOCA is based on the combined inventory of the reactor coolant system, the two accumulators, the two core makeup tanks, and the in-containment refueling water storage tank flooding the containment. The maximum inventory also considers makeup from the cask loading pit and boric acid tank.
Curbs are provided around openings through the maintenance floor at elevation 107-2 or watertight hatches are provided to close openings not required for hydrogen venting to control flooding.
Overflow into the refueling canal occurs through a pipe centered at elevation 110-0. Curbs around openings into the chemical and volume control system compartment extend up to elevation 110-0.
Curbs around openings into the PXS-A compartment extend up to elevation 110-2. Curbs around openings into the PXS-B compartment extend up to elevation 110-1. With these curb elevations, water flooding the maintenance floor is directed first into the refueling canal, then into the CVS compartment, then into the PXS-B compartment, and finally into the PXS-A compartment.
The evaluation of containment flooding from postulated component failures includes the compartments that are located below the maximum flood level. There are seven subcompartments that contain components below the floor at elevation 107-2. The active safe shutdown components inside containment which are located below the maximum flood level are located in only two of the seven floodable compartments.
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VEGP 3&4 - UFSAR The seven compartments partially or completely below the maximum flood level include the reactor vessel cavity, the two steam generator compartments, the vertical access tunnel, the two passive core cooling system compartments, and the chemical and volume control system compartment. The safe shutdown components are located in the two passive core cooling system compartments.
The reactor vessel cavity and the two steam generator compartments are interconnected by a large vertical access tunnel. These four compartments are treated, in this discussion, as one large floodable volume and they are referred to as the reactor coolant system compartment. Flooding of this compartment above elevation 107-2 also includes the maintenance floor outside the curbs around the other three compartments.
The PXS-A compartment (Room 11206), PXS-B compartment (Room 11207 and its connected RNS Valve Room 11208), and the chemical and volume control system compartment (Room 11209) are physically separated and isolated from each other by structural walls and curbs such that flooding in any one of these compartments or in the reactor coolant system compartment cannot cause flooding in any of the other compartments. Located in the 107'-2 floor are hatches which allow personnel access into PXS-A, PXS-B, and the CVS compartment for maintenance. These hatches are leak-tight and designed to prevent flooding through the 107'-2 floor into the compartments below.
The access hatch to the PXS-B compartment is located near the containment wall and is normally closed to address severe accident considerations. The access hatch to the PXS-B compartment is accessible from Room 11300 on elevation 107-2.
The access hatches and curb on the 107'-2 maintenance floor are important design features which support containment flood-up during a LOCA. The access hatches are designed to be opened or removed during plant shutdown to support maintenance work in the PXS-A, PXS-B, and CVS compartments. The design of the curbs may also allow them to be removed to support component inspections or other plant shutdown maintenance activities. Removal of these components to conduct shutdown maintenance activities cannot be started until after the plant is in MODE 4. In addition to the delay in hatch or curb removal, only one of the three compartments (PXS-A, PXS-B, and CVS compartments) can have the access hatches or curbs not fully intact at one time. These limitations preserve an adequate flood-up recirculation level to be able to support core cooling in the unlikely event of a loss of shutdown cooling or loss or RCS inventory.
The fire protection system and the demineralized water transfer and storage system are open-cycle systems that enter the containment. During plant operation, the containment piping for these systems is isolated by containment isolation valves and is not a potential flooding source. These systems are not open systems as defined in Bulletin 80-24 (one that has an essentially unlimited source).
Reactor Coolant System Compartment The reactor coolant system compartment, represented by the reactor vessel cavity, the two steam generator compartments, and the large vertical access tunnel, is the largest of the separate floodable compartments. With the exception of the pressurizer which is at a higher elevation, the principal components of the reactor coolant system are contained in this compartment.
The reactor vessel cavity and the adjoining equipment room are at the lowest level in the containment. The floor level of these rooms is at elevation 71-6. The floor level of the two steam generator compartments is at elevation 83-0. A portion of each compartment has low point areas at elevation 80-0.
The containment sump pumps are located in the equipment room at elevation 71-6. The arrangement for the floor drains from the two passive core cooling system compartments and the chemical and volume control system compartment provide a drain path for each compartment to the lowest level of containment (elevation 71-6) where the containment sump is located. Therefore, the 3.4-7 Revision 8
VEGP 3&4 - UFSAR source of the flooding in the reactor coolant system compartment is not limited to the components or systems contained within this compartment.
Any leakage that occurs within the containment drains by gravity to the elevation 71-6 equipment room. Reverse flow into the two passive core cooling system compartments and the chemical and volume control system compartment is prevented by redundant backflow preventers in each of the three compartment drain lines.
Flooding in any compartment of the containment is detected by the containment sump level monitoring system and the containment flood-up level instrumentation.
The containment sump level monitoring system consists of three seismically qualified level sensors for the containment sump. These sensors transmit the sump level indication through a qualified isolation device to the main control room and the plant instrumentation system.
The plant instrumentation system monitors the rate of the sump level rise, calculates the leakage collection rate, and initiates the appropriate alarms in the main control room. A description of this leak detection system is provided in Subsection 5.2.5.3.1.
Another indication of flooding in this compartment is provided by the containment flood-up level instrumentation consisting of three redundant Class 1E level sensor racks. Multiple discrete level signals are provided from the bottom of the reactor vessel cavity to the top of the vertical access tunnel. These level sensors transmit the containment sump water level indication to the main control room.
In the event that the source of the containment flooding can not be terminated, the water level in the reactor vessel cavity and the steam generator compartment continues to increase until the water source has been depleted or the leak has been isolated. The maximum level that could occur in the compartment from all of the water which is available in containment is elevation 110-2.
Essential, safety-related components in the RCS compartment subjected to submergence are either protected or qualified for submergence. The flooding of this compartment has no impact on safe shutdown capability.
Passive Core Cooling System Compartments The PXS-A and PXS-B compartments, located in the southeast and northeast quadrants of the containment, primarily contain components associated with the passive core cooling system. The safe shutdown related components of the passive core cooling system located in these two compartments are redundant and essentially identical. One set of the redundant equipment is located in each of the two separate compartments.
The redundant passive core cooling system components located in these two compartments provide coolant to the reactor vessel from the two core makeup tanks, the two accumulators, and the in-containment refueling water storage tank via two independent and redundant direct vessel injection lines.
Each passive core cooling system compartment contains a parallel set of normally closed, air operated, core makeup tank isolation valves that receive actuation signals to open during a safe shutdown operation. These valves are approximately 10 feet above the floor level of the passive core cooling system compartments and 26 feet above the floor of the reactor vessel cavity.
Each passive core cooling system compartment also contains one normally open accumulator isolation valve and one normally open in-containment refueling water storage tank isolation valve.
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VEGP 3&4 - UFSAR These valves do not have to be repositioned during a safe shutdown operation and a coincident flooding event.
In addition, each passive core cooling system compartment contains four passive core cooling system containment recirculation subsystem isolation valves. A normally closed, squib valve is located in each of two parallel flow paths. One of the lines includes a check valve in series with the squib valve. The other line includes a normally open, motor-operated valve in series with the squib valve. The squib valve and motor-operated valves are opened on a Low-3 in-containment refueling water storage tank level signal to provide a redundant flow path from the flooded reactor/steam generator compartments to the reactor vessel. One set of these redundant containment recirculation subsystem isolation valves is required to open to provide a redundant recirculation flow path to the reactor vessel. In the unlikely event that one of the two passive core cooling system compartments were to be flooded, the set of recirculation valves in the other, unflooded, compartment could be opened. Note that these squib valves are qualified to operate after being flooded.
The design bases for this system are described in Section 6.3. The passive core cooling system is designed to perform its safety functions in the unlikely event of the most limiting single failure occurring coincident with any design basis event. For example, a direct vessel injection line could break in one of the two passive core cooling system compartments, thus preventing the core makeup tank and the accumulator located in the compartment from delivering borated water to the reactor vessel. A coincident single failure in the other passive core cooling system compartment would prevent only one of the two parallel injection paths from opening. This series of events would not prevent the passive core cooling system from performing its safety function.
The maximum flooding rate to either of these passive core cooling system compartments would occur on a postulated LOCA of one of the eight inch direct vessel injection lines at a location inside one of the two compartments. This postulated rupture would result in direct blowdown from the reactor coolant system to the compartment as well as blowdown of the associated core makeup tank and accumulator. The resulting flooding in one of two passive core cooling system compartments would not prevent the passive core cooling system from performing its safe shutdown function.
Another postulated LOCA, that would cause rapid flooding in the PXS-B compartment, is a rupture of the 12 inch normal residual heat removal system line. This line is routed from one of two reactor coolant system hot legs to a containment penetration in the PXS-B compartment.
The evaluation of containment flooding events is also concerned with non-LOCA flooding events.
The maximum flooding rate to either of the passive core cooling system compartments, for a non-LOCA event, would be based on a postulated rupture of one of the two in-containment refueling water storage tank lines or a postulated rupture of one of the two accumulator injection lines.
A 10-inch line is routed from the in-containment refueling water storage tank to the PXS-A compartment and a 10-inch line is routed to the PXS-B compartment. The driving head from a full in-containment refueling water storage tank to either of these two compartments is approximately 35 feet. A rupture in one of these lines would result in flooding of the associated passive core cooling system compartment and the reactor coolant system compartment via the normal drain path or by overflowing the passive core cooling system compartment.
The 8-inch accumulator injection lines are routed from the accumulators to the 8-inch direct vessel injection lines. A rupture of either of these two injection lines at a point upstream of the two series reactor coolant system pressure boundary check valves would result in the blowdown of the accumulator to the associated compartment. The water level attained in this case would be limited to the water volume of the accumulator. The water level would not reach the level of the core makeup tank isolation valves.
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VEGP 3&4 - UFSAR The total flood-up of either the PXS-A or PXS-B compartments from any source of water is acceptable and does not prevent the passive core cooling system from performing its required safe shutdown function.
The PXS-A and the PXS-B compartments are physically separated and isolated from each other by a structural wall so that flooding in one compartment can not cause flooding in the other compartment.
They are located below the maintenance floor level which is at elevation 107-2. A curb or watertight hatch is provided at openings that penetrate through the maintenance floor into these compartments from the elevation 107-2 floor level.
There are several HVAC ducts, cable trays, and pipes that penetrate the maintenance floor into the passive core cooling system compartments. These penetrations are properly protected to prevent leakage into the passive core cooling system compartments.
The floor drains for these two compartments are located at elevation 84-6. Reverse flow through the floor drains is blocked by redundant, safety-related backflow preventers in the drain lines.
When the flooding rate exceeds the ability of the floor drain lines to drain the water from the compartment, or in the event that the floor drain line is blocked, the water level in that compartment increases to the entrance curb elevation.
Should the flooding continue, the water overflows from that compartment to the maintenance floor at elevation 107-2. The water overflowing to this level would immediately drain to the reactor coolant system compartment via the vertical access tunnel. There is no curb at the entrance to the vertical access tunnel; therefore, water on the maintenance floor (elevation 107-2) flows freely into the reactor coolant system compartment. For LOCA events, flooding via this path continues to a level above the reactor coolant system cold legs.
If the leakage rate into PXS-A or PXS-B were not excessive, the compartment drain lines would prevent significant flood-up in that compartment. Consequently, the flooding of the components could be prevented for postulated flooding events of limited duration and flowrates less than the drain line capacity.
The flowrate from the compartments is a function of the water height in the PXS compartments and the water height in the reactor coolant system compartment. The differential head between the two water levels establishes the flowrate from the compartment.
The draining of these compartments initiates flooding of the reactor vessel cavity and the adjoining cavity equipment room. If the operator does not terminate the leak, action is taken to shut down the reactor.
If the flooding rate is not greater than the compartment drain line capacity, the large volume of the reactor vessel cavity and the adjoining equipment room provides additional time for the operator to identify the source of leakage before any significant flooding occurs in the compartments containing the passive core cooling system equipment.
Should the flooding continue, the water level eventually reaches the steam generator compartment floor at elevation 83-0. The large floor area of the two steam generator compartments and the vertical access tunnel provides additional volume for flood-up and reduces the rate of level increase.
The containment isolation valves in these two passive core cooling system compartments are located above elevation 95-0, but below the maximum flood-up level. The PXS-A compartment contains one normally closed, motor operated, spent fuel pool cooling system containment isolation valve. The PXS-B compartment contains four normally closed, motor operated, normal residual heat removal 3.4-10 Revision 8
VEGP 3&4 - UFSAR system containment isolation valves. These containment isolation valves are not required to operate for safe shutdown and they do not fail open as a result of compartment flooding. Also, there are redundant outside containment isolation valves for each line that penetrates the containment boundary.
Chemical and Volume Control System Compartment The majority of the components associated with the chemical and volume control system are located inside the containment in a separate compartment in the north quadrant of the containment below elevation 107-2.
There are several HVAC ducts, cable trays, and pipes that penetrate the maintenance floor into the chemical and volume control system compartment. These penetrations are properly protected to prevent leakage around the ducts into the chemical and volume control system compartment. The entrance curb elevation for the chemical and volume control system compartment is lower than the PXS-A and B compartment curbs to preferentially flood the chemical and volume control system compartment.
A single floor drain line is routed from this compartment to the containment sump at elevation 71-6.
Reverse flow from the containment sump to this compartment is prevented by redundant, safety-related backflow preventers in the drain lines.
In the event that the single drain line were to be blocked, the water level in the chemical and volume control system compartment would flood to the level of the entrance curb elevation and would over flow to the maintenance floor at elevation 107-2. The water overflowing to this level would drain to the reactor coolant system compartment via the vertical access tunnel. There is no adverse effect on safe shutdown of the plant from flooding of the chemical and volume control system compartment.
The fire protection system and the demineralized water transfer and storage system are open-cycle systems that enter the containment. During plant operation, the containment piping for these systems is isolated by containment isolation valves and is not a potential flooding source. These systems are not open systems as defined in Bulletin 80-24 (one that has an essentially unlimited source).
3.4.1.2.2.2 Auxiliary Building Flooding Events General The AP1000 auxiliary building contains radiologically controlled areas and nonradiologically controlled areas which are physically separated by 2 and 3 foot structural walls and floor slabs.
These structural barriers are designed to prevent flooding across the boundary between these areas by locating penetrations for piping and HVAC duct above maximum flood levels, or by sealing these penetrations. Process piping penetrations between the radiologically controlled areas and nonradiologically controlled areas are embedded in the wall or are welded to a steel sleeve in the wall. Electrical penetrations between the radiologically controlled areas and nonradiologically controlled areas are located above the maximum flood level. Electrical penetrations subject to the effects of the local build up of water on floors above the maximum flood level are also sealed.
For example, flooding in the auxiliary building at elevation 66-6 of the radiologically controlled area would not cause flooding in the nonradiologically controlled areas since the two areas are completely separated by a three foot thick structural wall. In the non-radiologically controlled area (non-RCA) of the auxiliary building, the four Class 1E electrical divisions are separated by 3-hour fire barriers.
Portions of these fire barriers also serve as flood barriers.
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VEGP 3&4 - UFSAR Nonradiologically Controlled Areas The safe shutdown systems and components that are located in the nonradiologically controlled area are associated with the protection and safety monitoring and Class 1E dc system, and containment isolation. The safe shutdown components associated with the protection and safety monitoring system are the instrumentation and control (I&C) cabinets located in the nonradioactive controlled area on level 3 (elevation 100-0). The safe shutdown components associated with the Class 1E dc system are the Class 1E batteries on level 1 (elevation 66-6) and level 2 (elevation 82-6) and dc electrical equipment also on level 2.
The nonradiologically controlled areas of the auxiliary building are designed to provide maximum separation between the mechanical and electrical equipment areas. This separation prevents the propagation of leaks from the piping areas and the mechanical equipment areas to the Class 1E electrical and Class 1E I&C equipment rooms.
The major piping compartments in the nonradiologically controlled area are the main steam isolation valve compartments on levels 4 and 5 (elevations 117-6 and 135-3, respectively) and the valve/piping penetration compartment on level 3 (elevation 100-0). The mechanical equipment rooms in the nonradiologically controlled area are the HVAC compartments on levels 4 and 5.
Drain lines are provided in each of the piping and mechanical equipment compartments which drain to the turbine building drain tank. Leakage from postulated pipe ruptures in these compartments will drain to the turbine building.
Radiologically Controlled Areas The safe shutdown components located in radiologically controlled areas (RCA) are primarily containment isolation valves which are located near the containment vessel and above elevation 82-6. The containment isolation valves below the maximum flood level are either air-operated and fail closed, or remain closed during a safe shutdown operation.
The evaluation of potential flooding within the radiologically and nonradiologically controlled areas of the auxiliary building is performed on a floor-by-floor basis as described below.
Auxiliary Building Level 1 (Elevation 66-6)
Nonradiologically Controlled Area Level 1 of the nonradiologically controlled area has five individual rooms that contain Class 1E batteries: four divisional (A, B, C, and D) Class 1E battery rooms and one Class 1E spare battery room. The doors are not water tight.
The primary line of defense for level 1 is to exclude fluid systems and their associated piping from this area. The only fluid systems in level 1 are the potable water and fire protection systems.
Potable or demineralized water brought into the area in portable containers is used for battery washdown. Potable water is used for the emergency eye wash/shower facilities. The maximum nominal diameter of potable water piping in this area is 1 inch; therefore, it is excluded from consideration as a source of flooding.
The seismically qualified fire protection system piping routed through levels 1, 2, 3, and 4 is the only piping in level 1 that is greater than 1 inch in diameter.
Fire fighting activities or fire protection system piping through-wall cracks in levels 1, 2, 3, or 4 would contribute to flooding in level 1. The drain lines, stairwells, and the elevator shaft direct the 3.4-12 Revision 8
VEGP 3&4 - UFSAR water from fire fighting activities down to the auxiliary building nonradiologically controlled area sump located on level 1.
Fire fighting in these five battery rooms is accomplished by manual means from two fire hose stations located adjacent to the two stairwells. The maximum flowrate to this area from the two hose stations is assumed to be 250 gpm.
A limited supply of water is initially provided to the fire protection system standpipe fire hose stations (see Subsection 9.5.1) from the passive containment cooling water storage tank. A nominal volume of 18,000 gallons is provided for the fire protection system. A volume of approximately 26,200 gallons is conservatively assumed; this is the volume in the tank between the elevations of the fire protection system inlet and the tank overflow, and associated system piping. In the event of a through-wall crack in the fire protection system piping, the maximum water depth in the battery rooms would be less than 7.5 inches, assuming that the water could propagate into all rooms on this level. This maximum water depth is substantially below the terminal height on the first row of battery jars, which are located 7.5 inches above the floor. The terminal height is approximately 30 inches above the floor.
Since a limited supply of fire water is provided, inadvertent initiation of the fire protection system can not exceed the flooding levels described above. Operator action to stop inadvertent water flow from the fire protection system is expected to limit flooding to only a small fraction of this water supply.
Structural walls, drain line routing, and raised platforms prevent leakage that may occur in piping or mechanical areas on levels 4 and 5 from propagating to the electrical areas on levels 1, 2, 3, or 4.
Dual sump pumps and water level sensors are also provided in the sump on level 1. The level sensors transmit water level indication to the main control room and the plant control system.
Level alarms alert the operator to take corrective action.
The sump pumps are sized to remove approximately 250 gpm (with two pumps operating) based on a maximum flow from two fire hose stations of 250 gpm. The discharge of these pumps is directed to the turbine building drain tank of the waste water system (WWS) located on elevation 89-0 of the turbine building as described in Subsection 9.2.9. The discharge line into the tank is provided with a standpipe to prevent siphoning back to the auxiliary building nonradiologically controlled area sump. These sump pumps and level sensors are not required to maintain safe shutdown capability.
Radiologically Controlled Area There are no safe shutdown components located on level 1 of the radiologically controlled area.
The radiologically controlled area of the auxiliary building is subject to flooding from a variety of potential sources including, but not limited to, the component cooling water, central chilled water, hot water, spent fuel pool cooling, normal residual heat removal system, fire protection system, and chemical and volume control system, as well as various tanks. Most of the piping associated with these systems is above level 1; however, the flow from any postulated rupture in the radiologically controlled area will eventually flood level 1. The principal flow paths to level 1 are the vertical pipe chase and the floor gratings provided in the elevator lobbies on levels 2 and 3.
Other flow paths include the floor drain system, the stairwell, and the elevator shaft. The western wall of the stairwell contains a safety-related, seismically qualified, and fire-rated flood relief louver that provides a flow path into the adjacent corridor in the event of a fire protection system line rupture.
3.4-13 Revision 8
VEGP 3&4 - UFSAR The auxiliary building radiologically controlled area sump is located on level 1 with dual sump pumps and water level sensor provided in the sump. The level sensor transmits water level indication to the main control room and the plant control system. High level alarms alert the operator to take corrective action. Two seismic Category I flood level sensors, which alert in the main control room, are also located in this area.
The sump pumps are sized to remove approximately 250 gpm (with two pumps operating) based on a maximum flow from two fire hose stations of 250 gpm. The discharge of these pumps is directed to the waste holdup tanks of the liquid radioactive waste system as described in Section 11.2. These sump pumps and sump level sensors are not required to maintain safe shutdown capability.
For the component cooling water and central chilled water systems, the maximum flooding volume is bounded by the system volume plus a reasonable period of makeup. This includes any discharges from the pressure relief valves on the cooling water lines from the RNS heat exchangers. Flow from these two valves is also directed to the auxiliary building sump through the radioactive waste drain system. For the spent fuel pool cooling system, the maximum flooding volume is limited to the volume of water above the non-seismic spent fuel pool cooling suction line, plus a reasonable period of makeup. This flooding volume is approximately equal to that of the component cooling water and chilled water systems above.
The normal residual heat removal system is primarily operated when the plant is being shutdown.
Since it is not normally operating, it is evaluated as a moderate-energy system. Flooding is determined based on the maximum flowrate from a through-wall crack in a 10 inch normal residual heat removal system suction line.
Flooding due to a break in the high-energy chemical and volume control system makeup pump discharge line bounds the normal residual heat removal system through-wall crack.
For a flooding event due to a moderate-energy line break of the fire protection system main header loop running through the auxiliary building, the maximum connected fire protection system inventory is assumed to be drained.
A flooding event due to a moderate-energy line break of the fire protection system main header loop running through the auxiliary building bounds the above conditions. Due to a circumferential break in this non-seismic moderate-energy piping, flooding flow output from both break ends originating from higher elevations eventually flows down to elevation 66'-6. Based on notification of flooding conditions received in the main control room, credited operator action is taken after pipe ruptures occur to limit the flood to the design basis level. Due to limited volume of the fire water storage tanks, maximum height is limited to 19 feet in non-tank rooms. To provide indication that supports initiation of credited operator action, two redundant, seismic Category I flood level sensors are used to alert the operators in the main control room of a floodup within the radiologically controlled area of the auxiliary building.
Flow from the postulated bounding break spreads throughout the level 1 rooms and corridor via flow under doors and interconnecting floor drains if the auxiliary building radiologically controlled area sump pumps are inoperable. The maximum flood level in the area, for any of the cases above, is 19 ft. This flooding will extend up to level 2 to elevation 85'-6 after credited operator action is taken to isolate fire protection system makeup at 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> past rupture initiation. This flooding has no impact on safe shutdown since there are no flooded components on level 1 or 2 that cannot perform the functions required for safe shutdown. The maximum flood height in waste holdup tank rooms (12166 and 12167) is 23 feet due to the water-tight doors and the locked-closed drain.
3.4-14 Revision 8
VEGP 3&4 - UFSAR Flow from a tank rupture in one of the tank rooms will initially flood the tank room, and begin to flow to the auxiliary building radiologically controlled area sump via floor drains. If the sump pumps are inoperable, the tank volume floods the balance of level 1 via the interconnecting floor drains. The maximum flood level for this event is less than for the piping failures discussed above.
Auxiliary Building Level 2 (Elevation 82-6)
Nonradiologically Controlled Area Level 2 of the nonradiologically controlled area has two Class 1E battery rooms, four divisional Class 1E dc electrical equipment rooms, and one Class 1E reactor coolant pump trip switchgear room. The doors to these rooms are not water tight.
Level 2 contains an arrangement of fire protection and potable water piping similar to level 1.
The potential for flooding on this level is limited to fire fighting activities and through-wall cracks in fire protection system piping. Fire fighting in these rooms is accomplished by manual means from two fire hose stations located adjacent to the two stairwells. The maximum flowrate to this area from the two hose stations is assumed to be 250 gpm.
The drains, elevator shafts, and stairwells drain water spilled on this level to level 1. Therefore, no significant accumulation of water occurs on level 2.
Radiologically Controlled Area The radiologically controlled area on level 2 contains a few containment isolation valves and an effluent monitor tank at elevation 92-6". The horizontal pipe chase at elevation 92-6 contains two normally closed normal residual heat removal system isolation valves. One spent fuel pool cooling system containment isolation valve is located, above 92-6, in the adjacent vertical pipe chase. The area on the north side of the lower annulus contains two chemical and volume control system and two liquid radwaste automatically operated containment isolation valves above elevation 82-6. These valves are required to close or remain closed during a safe shutdown operation.
Two chemical and volume control system valves used to isolate the chemical and volume control system makeup pump suction from the demineralized water storage tank are located in the demineralizer/filter access area at 82-6. These safety-related valves close or remain closed to prevent boron dilution events. They are not required for safe shutdown.
Potential sources of flooding for this area include, but are not limited to, the chemical and volume control system, a liquid radioactive waste system effluent monitor tank, and the fire protection system, including an automatic suppression system in the CVS makeup pump room. Flow from a component rupture or from fire fighting activities on level 2 drains to level 1 as described below.
To protect the above valves from flooding, the makeup pump compartment at elevation 82-6 drains, via the floor grating located in the corridor adjacent to the stairwell, directly to elevation 66-6. Flooding in the lower annulus drains directly to elevation 66-6 via the floor grating, drains, and various openings to the tank rooms. The stairwell and elevator shaft on the east wall are additional flow paths to level 1. The horizontal pipe chase at elevation 92-6 drains under the door to elevation 66-6 via the vertical pipe chase in the spent fuel pool cooling system penetration room. As a result of these drain paths, there is less than a 12-inch accumulation of water in the horizontal pipe chase from any postulated pipe ruptures. The containment isolation valves in these areas are capable of moving to their safety position during the event.
3.4-15 Revision 8
VEGP 3&4 - UFSAR Due to the moderate-energy line break of the fire protection system and the subsequent drainage of the entire connected fire protection system inventory to level 1, the bounding flood height originating from elevation 66'-6 is 19 feet. That flood height extends up to level 2 and terminates at elevation 85'-6. Therefore, the bounding flood height for the rooms on auxiliary building level 2 is 3 feet for a room with a floor elevation of 82'-6. This flooding has no impact on safe shutdown since there are no components on level 2 that cannot perform the functions required for safe shutdown if flooded.
Auxiliary Building Level 3 (Elevation 100-0)
Nonradiologically Controlled Area Level 3 of the nonradiologically controlled area includes the remote shutdown room, one reactor coolant pump trip switchgear room, four divisional Class 1E I&C rooms, one equipment room, and the valve/piping penetration room. The division A, B, C and D I&C rooms and the electrical room also include containment electrical penetrations. The doors are not water tight.
The level 3 Class 1E and non-Class 1E electrical areas contain only fire protection system piping.
Fire hose stations are provided near each of the two stairwells and normally dry fire protection piping, supplied from the passive containment system tank, serves the preaction sprinkler system in the non-1E equipment/penetration room. The potential for flooding in the electrical areas on this level is limited to fire fighting activities. The maximum flowrate to this area from either automatic or manual fire fighting activities is assumed to be 250 gpm. The floor drains, stairwells, and elevator shaft drain water spilled on this level down to level 1. Therefore, no significant accumulation of water occurs in this area.
The valve/piping penetration room on level 3 is physically separated from the electrical rooms.
The valve/piping penetration room contains automatically actuated containment isolation valves for the steam generator blowdown system and the hydrogen line in the chemical and volume control system. Access to this room is from the turbine building. The access door and drain lines provided in this room drain from the auxiliary building to the turbine building. Maximum postulated flood level for this room is less than 36 inches. The containment isolation valves in the area are located above this maximum flood level.
Radiologically Controlled Area Level 3 of the radiologically controlled area includes two containment pressure transmitters, three spent fuel pool level transmitters, four spent fuel pool cooling system isolation valves, and two passive containment cooling system safety-related valves, which are required for safe shutdown.
Potential sources of flooding for this area include, but are not limited to, the normal residual heat removal system, the component cooling water system, an effluent monitor tank and the fire protection system, including an automatic suppression system in the rail car bay. The worst case flooding scenario on this level is the moderate-energy line break in the fire protection system located in the rail car bay area. This scenario results in significant flooding in the rail car bay, residual heat removal system heat exchanger room, and resin transfer pump room due to the high flow rate in the fire protection system and limited flow paths out of the room. The maximum flood level in the rail car bay, resin transfer pump room, and residual heat removal system heat exchanger room is less than 4 feet. Fluid from this break will flow from the rail car bay through a safety-related flood relief flapper into the normal residual heat removal heat exchanger room, and through floor penetrations to the lower levels.
The containment pressure transmitters and one of the spent fuel pool level transmitters are located in the middle annulus. The safety-related passive containment cooling valves are located in the maintenance floor staging area. A moderate-energy line break in the fire protection system 3.4-16 Revision 8
VEGP 3&4 - UFSAR located in the maintenance floor staging area causes significant flooding in the middle annulus.
The maximum flood level in the middle annulus and the staging area is below elevation 107'-9.
The spent fuel pool level transmitter will be submerged; however, the transmitter is qualified for submergence.
The other safe shutdown components are located above the maximum flood level. The other two spent fuel pool level transmitters and the spent fuel pool cooling system isolation valves are located in the waste monitor tank room B. The maximum flood level in this room is due to a waste monitor tank rupture and will be less than 109 inches. The spent fuel pool level transmitters in this area are located below the maximum flood level; however, they are qualified for submergence.
For each scenario, the plant is maintained in a safe condition and the safety functions maintain their integrity without the need for operator action to mitigate the flooding. Flow from a component rupture or from fire fighting activities in other rooms on level 3 drains directly to level 1.
Auxiliary Building Level 4 (Elevation 117-6)
Nonradiologically Controlled Area Level 4 of the nonradiologically controlled area includes the main control room, one divisional Class 1E penetration room, one non-Class 1E electrical penetration room, two main steam isolation valve compartments, and one mechanical equipment room.
The doors to these rooms are not water tight. There are no doors from the main steam isolation valve compartments to the Class 1E electrical areas. The main steam isolation valve compartments are only accessible from the turbine building at elevation 135-3. The mechanical equipment room is only accessible from the turbine building at elevation 117-6.
The potential for flooding Class 1E electrical areas on this level is limited to fire fighting activities.
The Class 1E electrical penetration room and main control room are accessible from a hose station near the east stairwell. While the main control room kitchen and restroom are provided with potable water, the lines are 1 inch and smaller, and are not evaluated for pipe ruptures. The main control room contains a water heater tank and expansion tank Section in the mens toilet room to provide hot water for hand washing and dishwashing. To mitigate the effects of a postulated tank rupture, the water heater tank and expansion tank are enclosed in a seismic Category I structure and the room design is configured to contain the water and direct it to the floor drain, where it is gravity-drained to the sanitary drainage system (SDS).
Fire fighting in the control room is done manually using portable extinguishers or a fire hose from a hose station in the east corridor. In the event that a hose is brought into the main control room through the east corridor access doors, water accumulation is limited by flow through the access doors which are open. The threshold of the east corridor access door is at the elevation of the floor slab. Once in the corridor this flow drains, via floor drains, the stairwell and elevator shaft, to level 1. An emergency egress door and stairwell is located on the west end of the main control room, which leads down to the remote shutdown room. The threshold of the emergency egress door is flush with the raised portion of flooring in the main control room, which is approximately 14 inches above the east corridor entrance. Water being discharged in this area will flow through the porous raised flooring and flow back out the east access doors. The main control room has a normally closed floor drain which can be manually opened to drain water to the auxiliary building non-RCA sump at level 1. The drain paths prevent significant flooding of the adjacent rooms.
In the event of fire fighting activity in the non-Class 1E electrical penetration rooms, the accumulation of water is prevented by floor drains and flows through the stairwell and elevator shaft to level 1.
3.4-17 Revision 8
VEGP 3&4 - UFSAR The mechanical equipment room contains containment isolation valves for the chilled water, compressed air, component cooling water, and passive core cooling (nitrogen) systems. Flooding in the mechanical equipment room due to fire fighting or piping ruptures is directed to the turbine building through the access door at elevation 117-6 or through floor drains to the turbine building. The maximum flood level for this room is 4 inches. The containment isolation valves in this area are located above this maximum flood level.
The main steam isolation valve compartments contain the main steam and main feedwater piping and their isolation valves. In the event of a pipe break or leak in the area, floor drains to the turbine building are provided. Structural walls and floors are designed to prevent flow of water to levels 1, 2, or 3. For larger flows, wall openings and pressure relief panels, located at floor elevation, open to drain the rooms to the turbine building. The maximum flood level for these rooms is less than 36 inches. The isolation valves in this area are located above this maximum flood level.
Radiologically Controlled Area In the radiologically controlled area, there are six containment isolation valves on level 4. Five of these are located in the vertical pipe chase. These are for the primary sampling system, spent fuel pool cooling system and containment air filtration system. The primary source of flooding in the vertical pipe chase is from a moderate-energy line break in the fire protection system located in the personnel access area on level 5. Flow from this break will be directed through the open pipe penetration on level 5, into the vertical pipe chase, through the grating down to level 3 where water will flow under the door to the waste monitor tank room and through floor drains and an open sleeve penetration to the auxiliary building RCA sump which limits the flood level to less than 12 inches in the vertical pipe chase and 109 inches in the waste monitor tank room. The containment isolation valves are located above maximum flood level. The other containment isolation valve for the containment air filtration system is located in a separate compartment adjacent to column line 5. The principal source of flooding for this area is from a moderate-energy line break in the fire protection system located in the personnel access on level 5. Flow from this source will be directed through the open pipe penetration on level 5, into the containment air filtration system penetration room, under the door and through floor drains to the auxiliary building radiologically controlled area sump which limits the flood level to less than 30 inches in the containment air filtration system penetration room. The containment isolation valve body will be partially submerged; however, the limit switch and solenoid are above the maximum flood level. No other safe shutdown equipment is located in this area.
Auxiliary Building Level 5 (Elevation 135-3)
Nonradiologically Controlled Area Level 5 of the nonradiologically controlled area contains two mechanical HVAC equipment rooms and the upper portion of the two main steam isolation valve compartments. The only safety-related equipment on level 5 that could be affected by flooding is the potable water system main control room isolation valve located in the main mechanical HVAC equipment room, but it is positioned above the maximum flood height for this room.
The evaluation of the main steam isolation valve compartments is addressed in the discussion of level 4.
Water from fire fighting, postulated pipe, or potable water storage tank ruptures in the main mechanical HVAC equipment rooms drains to the turbine building via floor drains. Therefore, no significant accumulation of water occurs in this room. Floor penetrations are sealed and a 6 inch platform is provided at the elevator and stairwell such that flooding in these rooms does not propagate to levels below.
3.4-18 Revision 8
VEGP 3&4 - UFSAR The mechanical room between the main steam isolation valve compartments at level 5 is accessed from the turbine building on the same level. This room is drained to the turbine building.
In the event of fire fighting or postulated pipe ruptures, the accumulation of water is prevented by directing the flow to the drains into the turbine building. Floor penetrations are sealed such that flooding in this area does not propagate to other areas of the auxiliary building.
Radiologically Controlled Area Level 5 of the radiologically controlled area contains the fuel handling area operating deck, HVAC equipment and access rooms, the main equipment hatch staging area, and the component cooling water system valve room. Safety-related equipment on level 5 includes the compressed air tanks for the main control room emergency habitability system and associated valves located in the main equipment hatch staging area, containment leak rate test system valves in the personnel access area, and spent fuel equipment located in the fuel transfer canal and spent fuel pool.
Over-filling of the spent fuel pool would flood the fuel handling area operating deck. The flooding flowrate is limited by the makeup capacity from the demineralized water or chemical and volume control systems. Accumulation of water in this area is prevented by floor drains. Spent fuel pool cooling is not adversely affected by this event. There is no safe shutdown equipment in this area.
The component cooling water system valves with the regulatory treatment of nonsafety-related system important missions located in the component cooling water system valve room which support the normal residual heat removal system, are located well above the maximum flood level for this area and are expected to remain functional in a flooding event.
The shield building stairwell serves as a pipe chase for passive containment cooling system supply and return lines, drains for the passive containment cooling system valve room and passive containment cooling system air outlet shield plug, and a fire water line. Leakage from a crack in one of these lines flows down the stairwell to level 5, under the stairwell door to the HVAC equipment room, and then to the auxiliary building radiologically controlled area sump via floor drains or to the annex building. There is no significant accumulation in the stairwell or the equipment and access rooms. There is no safe shutdown equipment in this area. The passive containment cooling system supply and return line connections to the passive containment cooling system storage tank are above the maximum water level, thus a leak in these lines would not adversely affect the safe shutdown capability of the passive containment cooling system.
Water from fire fighting in the main equipment hatch staging area drains to the auxiliary building radiologically controlled area sump via floor drains. Therefore, no significant accumulation of water occurs in this area.
Auxiliary Building Upper Annulus (Elevation 132-3)
This area serves as the air flow path for the passive containment cooling system. It is bounded by the seismic Category I shield building on the outside and the seismic Category I containment vessel on the inside. The floor has a curb on the outside with a flexible seal connected to the shield building.
The curb and seal block communication with the middle annulus area, below. The outside wall of the annulus is provided with redundant safety-related drains to the yard drainage system.
The worst case flooding scenario is postulated as blockage of the nonsafety-related floor drains concurrent with inadvertent opening of a passive containment cooling system cooling water isolation valve. The maximum flood level is determined by the flow gradient to the operating drains. Maximum level will be approximately 2 feet. This level does not affect the capability of passive containment cooling system air cooling. No other safe shutdown equipment is affected. Passive containment cooling system operation or leakage is detected by sensors on the passive containment cooling 3.4-19 Revision 8
VEGP 3&4 - UFSAR system discharge line. During non-accident conditions the annulus is accessible to manually clear any drain blockage.
PCS Valve Room (Elevation 284-10)
This room contains three redundant safety-related valve trains for the passive containment cooling system water cooling subsystem. One train must open to provide the required containment cooling.
The only source of flooding for this room is a through-wall crack in the passive containment cooling system piping. The bounding through-wall crack is postulated in the 4 inch recirculation piping. This leak is not isolable from the 756,700-gallon passive containment cooling system water storage tank above the valve room.
The PCS valve room contains floor drains with sufficient capacity to prevent water accumulation in the room following a postulated through-wall crack. Water from the through-wall crack flows into the floor drains out to the auxiliary building roof, then to the storm drain system. The passive containment cooling system isolation valves are located above the maximum flood level in the valve room, so they remain operable.
The PCCWST level instruments are relied upon to monitor PCCWST water level for operation of the PCS and to identify leakage from the PCS. The leakage does not adversely affect containment or any other essential system.
3.4.1.2.2.3 Adjacent Structures Flooding Events Turbine Building The turbine building is subject to flooding from a variety of potential sources including the circulating water, service water, condensate/feedwater, component cooling water, turbine building cooling water, demineralized water and fire protection systems as well as the deaerator storage tank. Flow from any postulated ruptures above elevation 100-0 flows down to elevation 100-0 via floor grating and stairwells. Thus, there will be a negligible contribution from these sources to flooding of the auxiliary building compartments at elevations 135-3 and 117-6 via flow under doors. Auxiliary building flooding is bounded by the effects of postulated breaks in the compartments.
The bounding flooding source for the turbine building is a break in the circulating water piping which would result in flooding of the elevation 100-0 floor. Flow from this break runs out of the building to the yard through a relief panel in the turbine building west wall and limits the maximum flood level to less than 6 inches. The only area of the auxiliary building which interfaces with the turbine building at elevation 100-0 is the valve/piping penetration room. This room could be flooded via flow under the door or backflow through the drains, however the flood level would be less than postulated for a break in the valve/piping penetration room itself.
The waste water system (WWS) sump pumps located in the nonradiologically controlled area of the auxiliary building discharge to the turbine building drain tank. Backflow from the drain tank is prevented as described in Subsection 3.4.1.2.2.2 There is no safety-related equipment in the turbine building. The component cooling water and service water components on elevation 100-0, which provide the regulatory treatment of nonsafety-related systems important to support the normal residual heat removal system, are expected to remain functional following a flooding event in the turbine building since the pump motors and valve operators are above the expected flood level.
3.4-20 Revision 8
VEGP 3&4 - UFSAR Annex Building Nonradiologically Controlled Areas The primary sources of flooding in the nonradiologically controlled areas of the annex building are the component cooling water, chilled water and fire protection systems. Water from postulated breaks above elevation 100-0 flows primarily through floor drains to the annex building sump that discharges to the turbine building drain tank. Alternate paths include flows to the turbine building via flow under access doors at elevations 135-3 and 117-6 and flows down to elevation 100-0 via stairwells and elevator shaft. Water accumulation at elevation 100-0 is minimized by floor drains to the annex building sump and by flow under the access doors leading directly to the yard area.
There is no safety-related equipment in the nonradiologically controlled area portion of the annex building. The main ac power system components with regulatory treatment of nonsafety-related systems important missions are located on elevation 117-6 in the electrical switchgear rooms, which are separated from potential flood sources. Water from manual fire fighting operations is collected by floor drains discharging to the annex building sump or down a hatch or stairwell to elevation 100-0. The non-Class 1E dc and UPS system (EDS) equipment with regulatory treatment of nonsafety-related systems important missions is located on elevation 100-0 in separate battery rooms. Water in one of these rooms due to manual fire fighting in the room is collected by floor drains to the annex building sump and by flow under the access doors leading directly to the yard area. This is not expected to affect functionality of equipment in the adjacent rooms.
Radiologically Controlled Areas There is no safety-related equipment in the radiologically controlled area portion of the annex building. The primary sources of flooding in the radiologically controlled areas of the annex building are the component cooling water, chilled water and fire protection systems, including an automatic suppression system that protects the containment access corridor. Water from postulated breaks above elevation 100-0 drains through floor drains to the radioactive waste drain system sump in the radiologically controlled area of the auxiliary building or drains to elevation 100-0 via stairwells and equipment handling hatches or under access doors to the radiologically controlled area portion of the auxiliary building. Accumulated water at elevation 100-0 is minimized by floor drains discharging to the radioactive waste drain system sump or chemical waste tank in the auxiliary building. The contribution of water to the flooding of the radiologically controlled area portion of the auxiliary building is bounded by flooding events which could occur in the auxiliary building.
Radwaste Building The potential sources of flooding in the radwaste building are the chilled water, hot water, and fire protection systems or from failure of one of the three waste monitor tanks. Flow from postulated breaks is directed to floor drains via a curb/sloped floor around the perimeter to drain to the radioactive waste drain system sump in the radiologically controlled area of auxiliary building. The contribution of water to flooding of the auxiliary building is bounded by flooding events which could occur in the auxiliary building. There are no safety-related systems or components or equipment with regulatory treatment of nonsafety-related systems important missions in the radwaste building.
Diesel Generator Building The potential source of flooding in the diesel generator building is the fire protection system. There is no safety-related equipment in the diesel generator building. The diesel generator system which has regulatory treatment of nonsafety-related systems important mission has each diesel and associated auxiliaries in a separate compartment. Flooding due to a break in a fire water header is directed to 3.4-21 Revision 8
VEGP 3&4 - UFSAR the respective diesel generator building sump and subsequently pumped to the yard oil separator or is drained by gravity to the yard area under the access doors. The equipment in the adjacent diesel generator compartment should remain functional following the event.
3.4.1.3 Permanent Dewatering System The need for a permanent dewatering system is site specific and is defined as discussed in Subsection 3.4.3.
No permanent dewatering system is required because site groundwater levels are two feet or more below site grade level as described in Subsection 2.4.12.
3.4.2 Analytical and Test Procedures The AP1000 is designed so that the maximum water levels considered due to natural phenomena or internal flooding do not jeopardize the safety of the plant or the ability to achieve and maintain safe shutdown conditions. The analytical approach in the consideration of external and internal flooding events is described in Subsection 3.4.1.2.
3.4.3 Combined License Information The site-specific water levels given in Subsection 3.4.1.3 and Section 2.4 satisfy the interface requirements identified in Section 2.4.
3.4.4 References
- 1. ANSI/ANS-56.11-1988, "Design Criteria for Protection against the Effects of Compartment Flooding in Light Water Reactor Plants."
3.4-22 Revision 8
VEGP 3&4 - UFSAR EL 100' -0" GRADE LEVEL NOTES:
- 1. HDPE DOUBLE-SIDED TEXTURED WATERPROOF MEMBRANE ON TOP OF FIRST LAYER OF MUDMAT AND ON OUTSIDE VERTICAL FACE OF AUXILIARY BUILDING WALL UP TO EL. 100' -0" GRADE LEVEL NOTE 1 AUXILIARY BUILDING (WITH PROTECTIVE SHIELD ON VERTICAL FACE)
- 2. MSE WALL TO BE DESIGNED WITH GEOREINFORCED MATERIALS AND 18" THICK LAYERS OF COMPACTED FREE DRAINING GRANULAR SOIL BASEMAT 6
SECOND LAYER OF MUDMAT NOTE 2 EL 66' -6" 1 6" EL 60' -6" EL 59' -6" 5' -0" 6" 6" MINIMUM THICKNESS OF FIRST LAYER OF MUDMAT NOTE 1 1' -0" 6' -0" Figure 3.4-1 Typical Details of Nuclear Island Waterproofing Below Grade 3.4-23 Revision 8
- 1. HDPE DOUBLE-SIDED TEXTURED WATERPROOF MEMBRANE ON TOP OF EL 100' -0" GRADE LEVEL FIRST LAYER OF MUDMAT AND ON OUTSIDE VERTICAL FACE OF AUXILIARY BUILDING WALL UP TO EL 100' -0" GRADE LEVEL (WITH PROTECTIVE SHIELD ON VERTICAL FACE)
AUXILIARY BUILDING NOTE 1 BASEMAT SECOND LAYER OF MUDMAT EL 66' -6" 6"
EL 60' -6" EL 59' -6" 6"
6" MINIMUM THICKNESS OF FIRST LAYER OF MUDMAT NOTE 1 1' -0" 6' -0" Figure 3.4-2 Typical Details of Nuclear Island Waterproofing Below Grade with Step Back 3.4-24 Revision 8
VEGP 3&4 - UFSAR Figure 3.4-3 Not Used 3.4-25 Revision 8
VEGP 3&4 - UFSAR 2' -6" 2' -0" 6"
FINISH YARD GRADE STRUCTURE WALL EXTERIOR STRUCTURE WALL SOIL STABILIZING TIE BACKS (TYP)
AS REQUIRED BASEMAT FOUNDATION SPRAY-ON SPRAY-ON WATERPROOFING WATERPROOFING MEMBRANE MSE WALL PANELS AS REQUIRED MEMBRANE SOIL STABILIZING TIE BACK MSE WALL PANEL 1' -6" (TYP)
MUDMAT LAYERS CANT (CONT) 6" 1' -0" MSE WALL PANELS (TYP)
PANEL END 6" 6"
LEVELING PAD (CONT) 1' -0" COMPACTED BACKFILL DETAIL 1 DETAIL 1 ELEVATION Figure 3.4-4 Typical Details of Membrane Corner Detail at Basemat and Exterior Wall 3.4-26 Revision 8
VEGP 3&4 - UFSAR 3.5 Missile Protection General Design Criterion 4 of Appendix A to 10 CFR 50 requires that structures systems and components important to safety be protected from the effects of missiles. The AP1000 criteria for protection from postulated missiles provide the capability to safely shut down the reactor and maintain it in a safe shutdown condition. The AP1000 criteria also protect the integrity of the reactor coolant system pressure boundary and maintain offsite radiological dose/concentration levels within the limits defined in 10 CFR 50.34.
Missiles may be generated by pressurized components, rotating machinery, and explosions within the plant and by tornadoes or transportation accidents external to the plant. Potential missile hazards are eliminated to the extent practical by minimizing the potential sources of missiles through proper selection of equipment, and by arrangement of structures and equipment in a manner to minimize the potential for damage from missiles. Potential missiles due to failures of nonseismic items are addressed in Subsection 3.7.3.13. Heavy load-drop evaluations are described in Subsection 9.1.5.
The following are definitions for missile protection terminology:
Internally Generated Missile - A mass that may be accelerated by energy sources continuously present on site.
Single Active Failure - Malfunction or loss of a component of electrical or fluid systems. The failure of an active component of a fluid system is considered to be a loss of component function as a result of mechanical, hydraulic, pneumatic, or electrical malfunction, but not the loss of component structural integrity.
High-Energy System - Fluid systems that, during normal plant conditions, are operated or maintained pressurized with a maximum operating temperature greater than 200°F and/or a maximum operating pressure greater than 275 psig, as discussed in Subsection 3.6.1.
The following criteria are applied in the identification of missiles and the protection requirements that must be satisfied:
A missile must not damage structures, systems, or components to the extent that could prevent achieving or maintaining safe shutdown of the plant or result in a significant release of radioactivity.
A single active component failure is assumed in systems used to mitigate the consequences of the postulated missile and achieve a safe shutdown condition. The single active component failure is assumed to occur in addition to the postulated missile and any direct consequences of the missile. When the postulated missile is generated in one of two or more redundant trains of a dual-purpose safety-related fluid system, which is designed to seismic Category I standards and is capable of being powered from both onsite and offsite sources, a single active component failure need not be assumed in the remaining train(s), or associated supporting trains.
Walls, partitions, and other items that enclose safety-related systems, or separate redundant trains of safety related equipment, must be constructed so that a postulated missile cannot damage components required to achieve safe shutdown nor damage components required to prevent a release of radioactivity producing offsite doses in excess of 10 CFR 50.34 limits.
A postulated missile from the reactor coolant system must not cause loss of integrity of the primary containment, main steam, feedwater, or other loop of the reactor coolant system.
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VEGP 3&4 - UFSAR A postulated missile from any system other than the reactor coolant system must not cause loss of integrity of the containment or the reactor coolant system pressure boundary.
Other plant accidents or severe natural phenomena are not assumed to occur in conjunction with a postulated missile (except for tornado).
Offsite power is assumed to be unavailable if a trip of the turbine-generator or reactor protection system is a direct consequence of the postulated missile.
Safe shutdown is accomplished using only safety-related systems with a coincident single active failure, although nonsafety-related systems not affected by the missile are available to support safe shutdown.
Missiles are postulated to occur where the single failure of a retention mechanism can result in a missile, unless the missile is not considered credible as discussed later. Missiles created by the independent failures of two retention mechanisms are not postulated.
The energy of postulated missiles produced by rotating components is based on a 120 percent overspeed condition, unless such an overspeed condition is not possible (such as a synchronous motor).
Equipment required for safe shutdown is located in plant areas separate from potential missile sources wherever practical.
Spatial separation may be used to demonstrate protection from missile hazards when it is shown that the range and trajectory of the generated missile is less than the distance to or is directed away from the potential target.
The AP1000 passive design minimizes the number of safety-related structures, systems, and components required for safe shutdown. Systems required for safe shutdown are identified in Chapter 7. Safety class structures, systems and components, their location, seismic category, and quality group classifications are given in Section 3.2. General arrangement drawings showing locations of the structures, systems, and components are given in Section 1.2 The areas required for safe shutdown, and the major systems and components housed therein that are required to be protected from internally and externally generated missiles for safe shutdown, are summarized below:
The containment vessel, including the reactor coolant loop, and passive core cooling system inside containment The shield building, including the passive containment cooling system Containment penetration areas, including containment isolation valves and Class IE cables The control complex including the main control room, reactor protection system, batteries, and dc switchgear The spent fuel pool The AP1000 relies on safety-related systems and equipment to establish and maintain safe shutdown conditions. There are no nonsafety-related systems or components that require protection from missiles.
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VEGP 3&4 - UFSAR Evaluations are performed to demonstrate that the criteria are satisfied in the event a credible missile is produced coincident with a single active component failure. These evaluations include the following:
For those potential missiles considered to be credible, a realistic assessment is made of the postulated missile size and energy, and its potential trajectories.
Potentially impacted components associated with systems required to achieve and maintain safe shutdown are identified.
Loss of these potentially impacted components coincident with an assumed single active component failure is evaluated to determine if sufficient redundancy remains to achieve and maintain a safe shutdown condition. If these criteria are satisfied, no further protection is required for the identified missile. If these conditions are not satisfied, additional protective features are incorporated (for example, plant layout is modified, or barriers are added).
3.5.1 Missile Selection and Description 3.5.1.1 Internally Generated Missiles (Outside Containment) 3.5.1.1.1 Criteria for Missile Prevention Equipment for the AP1000 is selected to minimize the potential for missiles to be generated. Missiles are postulated as described in Subsection 3.5.1.1.2. The following items are the major equipment selection considerations with regards to missile prevention:
Safety-related rotating equipment is designed so that the surrounding housings would contain fragments in the event of failure of the rotating parts.
Valves that have only a threaded connection between the body and the bonnet are not used in high-energy systems. ASME Code,Section III valves with removable bonnets should be of the pressure-seal type or have bolted bonnets.
Valve stems of valves located in high-energy systems have at least two retention features. In addition to the stem threads, acceptable features include back seats on the stem or a power actuator, such as an air or motor operator.
Thermowells and other instrument wells, vents, drains, test connections, and other fittings located in high-energy systems are attached to the piping or pressurized equipment by welding. The completed joint should have a greater design strength than the parent metal.
Threaded connections in high-energy systems are avoided.
High-pressure gas cylinders permanently installed in safety-related areas are constructed to the criteria of ASME Code,Section III or Section VIII. Portable and temporary cylinders and cylinders periodically replaced in safety-related areas are constructed and handled in accordance with applicable Department of Transportation requirements for seamless steel cylinders.
3.5.1.1.2 Missile Selection 3.5.1.1.2.1 Missiles not Considered Credible This subsection describes internally generated missiles (outside of containment) not considered credible. Missiles not considered credible include the following:
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VEGP 3&4 - UFSAR Catastrophic failure of safety-related rotating equipment (such as pumps, fans, and compressors) leading to the generation of missiles is not considered credible. These components are designed to preclude having sufficient energy to move the masses of their rotating parts through the housings in which they are contained. In addition, material characteristics, inspections, quality control during fabrication and erection, and prudent operation as applied to the particular component reduce the likelihood of missile generation.
Catastrophic failure of nonsafety-related rotating equipment is not considered credible in situations where measures similar to those just described for safety-related rotating equipment are applied to them. Protection from nonsafety-related equipment will normally be provided by separation. In special situations, equipment features may be used to prevent missile formation.
Provisions to preclude generation of missiles due to failure of the turbine generator are discussed in Subsection 3.5.1.3.
Missiles originating in non-high-energy fluid systems are not considered credible because these systems have insufficient stored energy.
The valve bonnets of pressure-seal, bonnet-type valves, constructed in accordance with ASME Code,Section III, are not considered credible missiles. The valve bonnets are prevented from becoming missiles by the retaining ring, which would have to fail in shear, and by the yoke capturing the bonnet or reducing bonnet energy. Because of the conservative design of the retaining ring of these valves, bonnet ejection is unlikely.
The valves of the bolted bonnet design, constructed in accordance with ASME Code,Section III, are not considered credible missiles. These bolted bonnets are prevented from becoming missiles by limiting stresses in the bonnet-to-body bolting material according to ASME Code,Section III requirements, and by designing flanges in accordance with applicable code requirements. Even if bolt failure would occur, the likelihood of all bolts experiencing simultaneous complete severance failure is not credible. The widespread use of valves with bolted bonnets, and the low historical incidence of complete severance failure of the bonnet, confirm that bolted valve bonnets are not credible missiles. Safety-relief valves in high energy systems use the bolted bonnet design.
Valve stems are not considered as credible missiles if at least one feature (in addition to the stem threads) is included in their design to prevent ejection. Valve stems with back seats are prevented from becoming missiles by this feature. In addition, the valve stems of valves with power actuators, such as air- or motor-operated valves, are effectively restrained by the valve actuator. Valve stems of rotary motion valves, such as plug valves, ball valves (except single-seat ball valves) and butterfly valves, as well as diaphragm and bellows type valves are not considered as credible missiles. Because these valves do not have a large reservoir of pressurized fluid acting on the valve stem, there is little stored energy available to produce a missile.
Nuts, bolts, nut and bolt combinations, and nut and stud combinations have only a small amount of stored energy and thus are not considered as credible missiles.
Thermowells and similar fittings attached to piping or pressurized equipment by welding are not considered as credible missiles where the completed joint has a greater design strength than the parent metal. Such a design makes missile formation not credible. Threaded connections are not used to connect instrumentation to high-energy systems or components.
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VEGP 3&4 - UFSAR Instrumentation such as pressure, level, and flow transmitters and associated piping and tubing are not considered as credible missiles. The quantity of high energy fluid in these instruments is limited and will not result in the generation of missiles. The connecting piping and tubing is made up using welded joints or compression fittings for the tubing. Tubing is small diameter and has only a small amount of stored energy.
ASME Code,Section III vessel ruptures and ruptures of gas storage vessels constructed without welding using ASME Code,Section VIII criteria are not considered credible due to the conservative design, material characteristics, inspections, quality control during fabrication and erection, and prudent operation.
Rotating components that operate less than 2 percent of the time are not considered credible sources of missiles. Components that are excluded by this criterion include motors on valve operators and pumps in systems that operate infrequently, such as the chemical and volume control makeup pumps. This exclusion is similar to the exclusion mentioned in Subsection 3.6.1.1, that is, of lines from the high-energy category of lines that have limited operating time in high energy conditions.
Valves, rotating equipment, vessels, and small fittings not otherwise considered to be credible missiles due to design features or other considerations are not considered to be a potential source of missiles when struck by a falling object.
3.5.1.1.2.2 Explosions Missiles can potentially be generated by a hydrogen explosion. Missiles that could prevent achieving or maintaining a safe shutdown or result in significant release of radioactivity are precluded by design of the plant systems that use or generate hydrogen.
The battery compartments are ventilated by a system that is designed to preclude the possibility of hydrogen accumulation. Therefore, a hydrogen explosion in a battery compartment is not postulated.
The hydrogen injection package (CVS-MS-02) in the turbine building is supplied from the Plant Gas System (PGS) high pressure hydrogen switchover station bottles located in the yard, adjacent to the turbine building. The hydrogen is not located in a compartment that contains safety-related systems or components. The hydrogen is supplied to the chemical and volume control system inside the containment. There are two parallel flow paths: one for continuous addition of hydrogen during power operation, and one for batch additions of hydrogen during startup and restart. Each line has a control valve and flow transmitter to provide operators with system parameters and control capabilities. The hydrogen injection line break analysis has determined that a worst case failure of the Chemical and Volume Control System (CVS) hydrogen injection line outside the containment building would not lead to an explosive atmosphere in the compartment where the failure occurred.
Mixing within a compartment is achieved by normal convection caused by thermal forces from hot surfaces and air movement due to operation of HVAC systems. The hydrogen supply line is not routed through compartments that do not have air movement due to HVAC systems.
The bulk gas plant storage area for the plant gas system (PGS) stores hydrogen for use in generator cooling. This storage area is located sufficiently far from the nuclear island that an explosion would not result in missiles more energetic than the tornado missiles for which the 3.5-5 Revision 9
VEGP 3&4 - UFSAR nuclear island is designed. The hydrogen is piped to the generator in the turbine building. The turbine building includes sufficient ventilation to prevent an explosive concentration of hydrogen in the event of a leak.
A detonation of a flammable vapor cloud (delayed ignition) due to the accidental release of hydrogen from the PGS bulk gas storage area or from the high-pressure hydrogen bottles area would not result in missiles more energetic than the tornado missiles for which the nuclear island is designed.
3.5.1.1.2.3 Missiles to be Considered The following missiles are considered:
Nonsafety-related rotating equipment, not excluded above, Pressurized components, not excluded above, located in high-energy systems High pressure gas storage cylinders that may experience a failure of the outlet pipe or valve if accidentally impacted.
3.5.1.1.2.4 Credible Sources of Internally Generated Missiles (Outside Containment)
The consideration of missile sources outside containment that can adversely affect safety-related structures, systems or components is limited to a few rotating components inside the auxiliary building and a few pressurized components in the chemical and volume control system. The safety-related systems and components needed as described in Section 7.4 to bring the plant to a safe shutdown are located inside the containment shield building and auxiliary building, both of which have thick structural concrete exterior walls that provide protection from missiles generated in other portions of the plant. Safety-related systems and components located in the auxiliary building, including the main control room, are protected from missiles generated in other portions of the auxiliary building by the structural concrete interior walls and floors. Protection against potential missiles from the turbine-generator is discussed in Subsection 3.5.1.3.
Rotating components located inside the auxiliary building that are either safety-related or are constructed as canned motor pumps would contain fragments from a postulated fracture of the rotating elements. These are excluded from evaluation as missile sources. Rotating components used less than 2 percent of the time are also excluded from evaluation as missile sources. This exclusion of equipment that is used for a limited time is similar to the approach used for the definition of high-energy systems. Nonsafety-related rotating equipment in compartments surrounded by structural concrete walls with no safety-related systems or components inside the compartment is not considered a missile source. Rotating equipment with a housing or an enclosure that contains the fragments of a postulated impeller failure is not considered a credible source of missiles. For one or more of these reasons the nonsafety-related rotating equipment inside the auxiliary building is not considered to be a credible missile source. Nonsafety-related rotating equipment in compartments with safety-related systems or components that do not provide other separation features have design requirements for a housing or an enclosure to retain fragments from postulated failures of rotating elements.
The high-energy system inside the auxiliary building that includes pressurized components in the high-energy portions that are constructed to standards other than the ASME Code criteria outlined in Subsection 3.5.1.1.1 is the chemical and volume control system. The high-energy portion of this system inside the auxiliary building that is not constructed to ASME Code criteria outlined in Subsection 3.5.1.1.1 is from the makeup pumps to the containment and system isolation valves. The nonsafety-related, high-energy portion of this system is not required to be protected from missiles.
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VEGP 3&4 - UFSAR The nonsafety-related, high-energy portion of the chemical and volume control system is not to be considered a missile source. It includes the design features that are outlined above to exclude components from consideration as missile sources. These considerations include features such as a pump housing or enclosure that contains fragments of a postulated impeller fracture, valve design requirements, vessel design requirements, or enclosure requirements. See Table 3.6-1 for a list of the high-energy systems.
Falling objects (i.e. gravitational missiles) heavy enough to generate a secondary missile are postulated as a result of movement of a heavy load or from a nonseismically designed structure, system, or component during a seismic event. Movements of heavy loads are controlled to protect safety-related structures, systems, and components, see Subsection 9.1.5. Safety-related structures, systems, or components are protected from nonseismically designed structures, systems, or components or the interaction is evaluated. See Subsection 3.7.3.13 for additional discussion on the interaction of other systems with Seismic Category I systems. Valves, rotating equipment, vessels, and small fittings not otherwise considered to be credible missiles due to design features or other considerations are not considered to be a potential source of missiles when struck by a falling object.
The outlet pipes and valves for the air storage bottles for the main control room are constructed to the ASME Code,Section III, requirements and are designed for seismic loads. The attached pipes and valves are not credible missile sources due to an accidental impact. The air storage bottles are located within a structural steel frame and are in an area with no activity directly above. For the reasons noted above, secondary missiles are not considered credible missiles.
3.5.1.2 Internally Generated Missiles (Inside Containment)
Selection of equipment for the AP1000 considers provisions to minimize the potential for missiles to be generated. The considerations previously discussed in Subsection 3.5.1.1 are also applicable to equipment inside the containment.
3.5.1.2.1 Missile Selection 3.5.1.2.1.1 Missiles not Considered Credible Potential missiles are not considered credible when sufficient energy is not available to produce the missile, or by design the probability of creating a missile is negligible. The following are not considered credible sources of internally generated missiles:
Reactor coolant pump design requirements are established so that any failure of the rotating parts would be retained within the casing at specified overspeed conditions. This is discussed in Subsection 5.4.1.3.6.
Catastrophic failure of rotating equipment such as pumps, fans, and compressors leading to the generation of missiles is not considered credible as described previously in Subsection 3.5.1.1.2.
Failure of the reactor vessel, steam generators, pressurizer, core makeup tanks, accumulators, reactor coolant pump castings, passive residual heat exchangers, and piping leading to the generation of missiles is not considered credible. This is due to the material characteristics, preservice and inservice inspections, quality control during fabrication, erection and operation, conservative design, and prudent operation as applied to the particular component.
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VEGP 3&4 - UFSAR Gross failure of a control rod drive mechanism housing, sufficient to create a missile from a piece of the housing or to allow a control rod to be ejected rapidly from the core, is not considered credible. This is because of the same reasons listed above for the reactor vessel and other components and is based on the following:
- The control rod drive mechanisms are shop hydrotested to 125 percent of system design pressure.
- The housings are hydrotested to 125 percent of system design pressure after they are installed on the reactor vessel head. They are included as part of the hydrotest of the completed reactor coolant system.
- The housings are made of Type 304 or 316 stainless steel, which exhibits excellent notch toughness.
- The allowable stress levels in the mechanism are not exceeded due to system thermal transients at power or by thermal movement of the coolant loops.
- The welds in the pressure boundary of the control rod drive mechanism meet the same design, procedure, examination, and inspection requirements as the welds on other ASME Code,Section III, Class 1 components.
- A nonmechanistic control rod ejection is considered in the safety analyses in Chapter 15 and the design transients in Subsection 3.9.1.1. The integrated head package and control rod drive mechanisms are not designed for the dynamic effects of a missile generated by a rupture of the control rod housing.
Valves, valve stems, nuts and bolts, and thermowells in high-energy fluid systems and missiles originating in non-high-energy fluid systems are not considered credible missiles as discussed previously in Subsection 3.5.1.1.2.1.
3.5.1.2.1.2 Explosions Missiles can potentially be generated by a hydrogen explosion. Missiles that could prevent achieving or maintaining a safe shutdown or result in significant release of radioactivity are precluded by design of the plant systems that use or generate hydrogen.
The hydrogen injection package (CVS-MS-02) in the turbine building is supplied from the Plant Gas System (PGS) high pressure hydrogen switchover station bottles located in the yard, adjacent to the turbine building. The hydrogen is supplied to the chemical and volume control system inside the containment. Hydrogen is injected into the Reactor Coolant System (RCS) at a steady continuous flow rate for normal operation and at higher flows for batch operation when it is desirable to change the hydrogen concentration in the coolant quickly. In case of a Chemical and Volume Control System (CVS) hydrogen injection line failure inside containment, hydrogen would rapidly dilute due to the mixing with air. The hydrogen injection line break analysis conclusion has determined that a failure of the CVS hydrogen injection line inside the containment would not lead to an explosive atmosphere in the compartment where the failure occurred. Mixing within a compartment is achieved by normal convection caused by thermal forces from hot surfaces and air movement due to operation of HVAC systems.
3.5.1.2.1.3 Missiles to be Considered The following missiles are considered:
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VEGP 3&4 - UFSAR Nonsafety related rotating equipment, not excluded above, Pressurized components, not excluded above, located in high-energy systems 3.5.1.2.1.4 Evaluation of Internally Generated Missiles (Inside Containment)
The consideration of credible missile sources inside containment that can adversely affect safety-related structures, systems, or components is limited to a few rotating components. The safety-related systems and components needed to bring the plant to a safe shutdown are inside the containment shield building and auxiliary building both of which have thick structural concrete exterior walls that provide protection from missiles generated in other portions of the plant.
Rotating components inside containment that are either safety-related or are constructed as sealless pumps would contain fragments from a postulated fracture of the rotating elements and are excluded from evaluation as missile sources. Rotating components in use less than 2 percent of the time are also excluded from evaluation as missile sources. This exclusion of equipment that is used for a limited time is similar to the approach used for the definition of high-energy systems. This includes the reactor coolant drain pumps, the containment sump pumps and motors for valve operators, and mechanical handling equipment. Non-safety-related rotating equipment in compartments surrounded by structural concrete walls with no safety-related systems or components inside the compartment is not considered a missile source. Rotating equipment with a housing or an enclosure that contains the fragments of a postulated impeller failure is not considered a credible source of missiles. For one or more of these reasons the nonsafety-related rotating equipment inside containment is considered not to be a credible missile source. Non-safety-related rotating equipment in compartments with safety-related systems or components that do not provide other separation features has design requirements for a housing or an enclosure to retain fragments from postulated failures of rotating elements.
The high-energy portions of high-energy systems inside the containment shield building except for a portion of the chemical and volume control system are constructed to the requirements of the ASME Code,Section III. The nonsafety-related, high-energy portion of the chemical and volume control system between the inside containment isolation valves and the outermost reactor coolant system isolation valves is not required to be protected from missiles and is not to be considered a missile source. It includes design features outlined above to exclude components from consideration as missile sources. In addition most of the nonsafety-related portion of the chemical and volume control system is contained in a compartment located away from safety-related equipment. See Table 3.6-1 for a list of the high-energy systems.
Falling objects heavy enough to generate a secondary missile are postulated as a result of movement of a heavy load or from a nonseismically designed structure, system, or component during a seismic event. Movements of heavy loads are controlled to protect safety-related structures, systems, and components (see Subsection 9.1.5). Design and operational procedures of the polar crane inside containment precludes dropping a heavy load. Additionally, movements of heavy loads inside containment occur during shutdown periods when most of the high-energy systems are depressurized. Valves, rotating equipment, vessels, and small fittings not otherwise considered to be credible missiles due to design features or other considerations are not considered to be a potential source of missiles when struck by a falling object. Secondary missiles are not considered credible.
Striking a component with a falling object will not generate a secondary missile if design of the component precludes generation of missiles due to pressurization of the component. Safety-related structures, systems, or components are protected from nonseismically designed structures, systems, or components or the interaction is evaluated. Nonsafety-related equipment that could fall and damage safety-related equipment during an earthquake is classified as seismic Category II and is designed and supported to preclude such failure. See Subsection 3.7.3.13 for additional discussion on the interaction of other systems with Seismic Category I systems. There are no high-pressure gas 3.5-9 Revision 9
VEGP 3&4 - UFSAR storage cylinders inside the containment shield building. For the reasons noted above, secondary missiles are not considered credible missiles.
3.5.1.3 Turbine Missiles The turbine generator is located north of the nuclear island with its shaft oriented north-south. In this orientation, the potential for damage from turbine missiles is negligible. Safety-related structures, systems and components are located outside the high-velocity, low-trajectory missile strike zone, as defined by Regulatory Guide 1.115. Thus, postulated low-trajectory missiles cannot directly strike safety-related areas.
The turbine and rotor design is described in Section 10.2. Protection is provided by the orientation of the turbine-generator and by the use of robust turbine rotors as described in Section 10.2. The rotor design, manufacturing, and material specification and the inspections recommended for the AP1000 provide an acceptably very low probability (see Subsection 10.2.2) of missile generation. Turbine rotor integrity is discussed in Subsection 10.2.3. This discussion includes fatigue and fracture analysis, material selection, and the maintenance program requirements.
The potential for a high-trajectory missile to impact safety-related areas of the AP1000 is less than 10-7. Based on this very low probability, the potential damage from a high-trajectory missile is not evaluated. The probability of an impact in the safety-related areas is the product of the probability of missile generation from the turbine; the probability, assuming a turbine failure, that a high-trajectory missile would land within a few hundred feet from the turbine (10-7 per square foot); and the area of the safety-related area. In the AP1000, the safety-related area is contained within the containment shield building and the auxiliary building.
The potential for a turbine missile from another AP1000 plant in close proximity has been considered.
As noted in Subsection 10.2.2, the probability of generation of a turbine missile (or P1 as identified in SRP 3.5.1.3) is less than 1 x 10-5 per year. This missile generation probability (P1) combined with an unfavorable orientation P2 x P3 conservative product value of 10-2 (from SRP 3.5.1.3) results in a probability of unacceptable damage from turbine missiles (or P4 value) of less than 10-7 per year per plant which meets the SRP 3.5.1.3 acceptance criterion and the guidance of Regulatory Guide 1.115.
Thus, neither the orientation of the side-by-side AP1000 turbines nor the separation distance is pertinent to meeting the turbine missile generation acceptance criterion. In addition, the shield building and auxiliary building walls, roofs, and floors, provide further conservative, inherent protection of the safety-related SSCs from a turbine missile.
The orientation of the Units 1 and 2 turbines has been evaluated and Vogtle Units 3 and 4 are located outside of the low trajectory strike zones as described in Regulatory Guide 1.115. Therefore, there is no potential for a turbine missile from Units 1 and 2 to impact Units 3 and 4.
The turbine system maintenance and inspection program is discussed in Subsection 10.2.3.6.
3.5.1.4 Missiles Generated by Natural Phenomena Tornado missiles are defined in accordance with Standard Review Plan, Subsection 3.5.1.4. The velocities are adjusted to the maximum wind velocity defined in Section 3.3. The following missiles are postulated:
A massive high-kinetic-energy missile, which deforms on impact. It is assumed to be a 4000-pound automobile impacting the structure at normal incidence with a horizontal velocity of 105 mph or a vertical velocity of 74 mph. This missile is considered at all plant elevations up to 30 feet above grade. In addition, to consider automobiles parked within half a mile of the plant at higher elevations than the plant grade elevation, the evaluation of the automobile 3.5-10 Revision 9
VEGP 3&4 - UFSAR missile is considered at all plant elevations up to the junction of the outer wall of the passive containment cooling water storage tank with the roof of the shield building. This elevation is approximately 193 feet above grade. This evaluation bounds sites with automobiles parked within half a mile of the shield building and auxiliary building at elevations up to the equivalent of 163 feet above grade.
A rigid missile of a size sufficient to test penetration resistance. It is assumed to be a 275 pound, eight inch armor-piercing artillery shell impacting the structure at normal incidence with a horizontal velocity of 105 mph or a vertical velocity of 74 mph.
A small rigid missile of a size sufficient to just pass through any openings in protective barriers. It is assumed to be a one inch diameter solid steel sphere assumed to impinge upon barrier openings in the most damaging direction at a velocity of 105 mph.
In addition to the missile spectrum specified above, the impact of tornado-driven sheet metal siding on the shield building is evaluated. The evaluation considers siding representative of the siding used on the turbine building, radwaste building, diesel generator building, and portions of the annex building. The evaluation considers a flat steel sheet, which bounds the corrugated siding design used on the buildings adjacent to the nuclear island.
3.5.1.5 Missiles Generated by Events Near the Site As described previously in Section 2.2, the site interface is established to address site specific missiles as discussed in Subsection 3.5.4. The AP1000 missile interface criteria are based on the tornado missiles described in Subsection 3.5.1.4. Additional analyses are required to evaluate other site specific missiles.
The primary access point, administrative building, communications support center, warehouse and shops, engineering and administrative building, maintenance support building and miscellaneous structures are common structures that are located at a nuclear power plant. They are of similar design and construction to those that are typical at nuclear power plants. Therefore, any missiles resulting from a tornado-initiated failure are not more energetic than tornado missiles postulated for design of the AP1000. Additionally, there are no other structures adjacent to the nuclear island other than the turbine building, annex building, radwaste building and passive containment cooling ancillary water storage tank.
In accordance with Subsection 2.2.3, the effects of explosions have been evaluated and it has been determined that the overpressure criteria of Regulatory Guide 1.91 is not exceeded. Consistent with Regulatory Guide 1.91, the effects of blast-generated missiles will be less than those associated with the blast overpressure levels considered; therefore, no further evaluation of blast-generated missiles is required.
3.5.1.6 Aircraft Hazards As described previously in Section 2.2, the site interface is established to address aircraft hazards as discussed in Subsection 3.5.4. The AP1000 missile interface criteria are based on the tornado missiles described in Subsection 3.5.1.4. Additional analyses are required to evaluate other site specific missiles. Aircraft crash probability, and the effects of this hazard on the plant, is determined as described in Section 2.2.
Airports and airways in the VEGP site vicinity are discussed in Subsection 2.2.2.6. Aircraft hazards related to these airports and airways (shown in Figure 3.5-201) have been evaluated in accordance with Regulatory Standard 002, Processing Applications for Early Site Permits, May 2004 (RS-002),
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VEGP 3&4 - UFSAR and NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Draft Revision 3, 1996 (NUREG-0800), Subsection 3.5.1.6.
3.5.1.6.1 Airports RS-002 acceptance criteria provide a distance threshold for evaluating aircraft hazards due to nearby airports.
All airports in the VEGP site vicinity are greater than 10 mi from the site. The hazard probability for these airports is considered acceptable if the projected annual number of operations is less than 1,000 D2, where D is the site-to-airport distance.
Bush Field is the closest (17 mi) and largest commercial airport in the VEGP site vicinity. The Federal Aviation Administration (FAA) (Reference 201) has projected the number of aircraft that will be in operation at Bush Field for every year up to 2025 for each of the following four types of aircraft:
general aviation, air taxi and commuter, commercial air carrier, and military. The projected flight data (which include landings and takeoffs) are provided in Table 3.5-201. As noted in the table, the total number of projected aircraft operations is substantially less than 1,000 D2 (289,000).
The other airports in the vicinity are much smaller than Bush Field. Since they are all at least 10 mi from the VEGP site, their aircraft hazard threshold is greater than 100,000 operations, which significantly exceeds their annual traffic.
As discussed in Subsection 2.2.2.6.1, a small unimproved grass airstrip is located immediately north of the VEGP site (north of Hancock Landing Road and west of the Savannah River). This privately owned and operated airstrip has a 1,650-foot turf runway oriented 80° East - 260° West. The airstrip is for personal use and the associated traffic consists only of small single-engine aircraft. In addition, there is a small helicopter landing pad on the VEGP site. This facility exists for corporate use and for use in case of emergency. The traffic associated with either of these facilities may be characterized as sporadic. Due to the small amount and the nature of the traffic, these facilities do not present a safety hazard to the VEGP site.
3.5.1.6.2 Airway V185 The VEGP site is approximately 1.5 mi east of the centerline of Federal Airway V185, which runs between Augusta and Savannah. A more detailed review of aircraft hazards was performed because the VEGP site is within the 2 statute mile limit. This review is summarized below.
Airways are typically used by commercial flights and by general aviation for inclement weather and nighttime operations. In general, military aircraft do not use the federal airways. To be allowed to fly in a federal airway, an aircraft needs to have the proper communication equipment and the pilot needs to have specific qualifications. In addition, most general aviation flights do not use a federal airway in favorable weather conditions. When these factors are considered, along with the fact that there are no regularly scheduled direct commercial flights between Augusta and Savannah, it is expected that the total number of aircraft using Airway V185 is relatively small.
Although the FAA does not maintain records of air traffic in Airway V185, informal communications with air traffic control personnel at the Augusta airport revealed that the southeast quadrant of the air space around the airport (of which Airway V185 is a part) has the least air traffic compared to the other quadrants and that the total traffic in Airway V185 is a fraction of the total operations into and out of the Augusta airport.
Because of the unavailability of traffic data for Airway V185, the following evaluation calculates the maximum number of airway flights per year above which the acceptance guideline probability of 10-7 3.5-12 Revision 9
VEGP 3&4 - UFSAR per year contained in RS-002 and NUREG-0800 is exceeded. Regulation 14 CFR 71 provides the criteria for determining the width of the airway. It is 4 nautical miles on either side of the centerline, for a total width of 8 nautical miles (9.2 mi).
PFA = CxNxA/W where:
PFA = probability per year of an aircraft crashing into a VEGP Units 3 and 4 safety-related structure, 1 x 10-7 C = in-flight crash rate per mile for aircraft using airway = 4 x 10-10 (RS-002)
N = number of flights per year along the airway A = effective area of plant or site area in square miles, see below W = airway width, 9.2 mi By rearranging this equation, the maximum number of flights corresponding to the acceptance guideline probability of 10-7 may be calculated.
NUREG-0800 and RS-002 also provide alternate guidance on the acceptable method for calculating area A. RS-002 specifies the use of the site area because, for ESP Applications where the type of power plant has not been selected, the plant cross-sectional area cannot be defined. However, because the Westinghouse AP1000 design has been selected, the effective area of the plant was used in this analysis.
The effective plant area (A) depends on the length, width, and height of the facility, as well as the aircrafts wingspan, skid distance, and impact angle (Reference 203).
The safety-related structures of the AP1000 design include only the containment and the auxiliary building; the remainder of the structures is not safety related. The AP1000 containment height is about 234 ft above grade, and the diameter is about 146 ft (Reference 204).
For traffic in Airway V185, the fractions of the types of aircraft using the airway were assumed to be the same as the fractions of the types of aircraft using Bush Field. Representative values for wingspan, skid distance, and impact angle for each aircraft type follow those suggested in (Reference 203). For military aviation, large aircraft are conservatively used in the estimates. The effective areas for general aviation, air taxi and commuter, commercial air carrier, and military aircraft are 0.025, 0.061, 0.073, and 0.086 sq mi, respectively. Using these effective areas and the fractions of aircraft types (52.9, 29.3, 12.8, and 5 percent for general aviation, air taxi and commuter, commercial air carrier, and military aircraft, respectively), the average of the weighted effective plant area, 0.045 mi2, is determined for the calculation.
Among the representative wingspans, the large military aircraft has the longest wingspan of 223 ft (Reference 203). The physical separation of the new reactor buildings is about 650 ft. Since this distance is longer than the largest representative wingspan (223 ft), the estimate of the effective area involves only one unit. In addition, Subsection 3.5.1.6 of NUREG-0800 also suggests the use of an effective area of one unit of the plant.
To reach the permissible crash probability of 1 x 10-7, the total number of flights traveling along Airway V185 would need to be about 51,100 per year. This value is higher than the total of all projected itinerant flights for 2025 at Bush Field (see Table 3.5-201).
3.5-13 Revision 9
VEGP 3&4 - UFSAR Although the flight data associated with Airway V185 are not available from the FAA, the number of flights in this airway is expected to be only a fraction of the total Bush Field flights. Therefore, the presence of Airway V185 is not a safety concern for the VEGP site.
3.5.2 Protection from Externally Generated Missiles Systems required for safe shutdown are protected from the effects of missiles. These systems are identified in Section 7.4. Protection from external missiles, including those generated by natural phenomena, is provided by the external walls and roof of the Seismic Category I nuclear island structures. The external walls and roofs are reinforced concrete. The structural design requirements for the shield building and auxiliary building are outlined in Subsection 3.8.4. Openings through these walls are evaluated on a case-by-case basis to provide confidence that a missile passing through the opening would not prevent safe shutdown and would not result in an offsite release exceeding the limits defined in 10 CFR 50.34.
Where necessary, adjacent structures and/or missile barriers are used to protect openings in the nuclear island building structures. Building structures credited in this evaluation to protect openings in seismic Category I nuclear island building structures are seismic Category I or seismic Category II (see Subsection 3.2.1.1.2). As identified in Subsection 3.7.2, seismic Category II building structures are designed for the safe shutdown earthquake using the same methods and design stress limits as are used for seismic Category I structures. Seismic Category II building structures are also designed to withstand the design basis tornado loads, including missiles, in accordance with the loading combinations identified in Table 3.8.4-2. The evaluation of site-specific hazards for external events that may produce missiles more energetic than tornado missiles is discussed in Subsection 2.2.1.
Evaluation of turbine missiles is provided in Subsection 3.5.1.3. Evaluation of tornado missiles is provided in Subsection 3.5.1.4. Conformance with regulatory guide recommendations is provided in Appendix 1A.
3.5.2.1 Protection from Externally Generated Missile through Wall 11 Evaluation of the openings in the north wall of the auxiliary building (Wall 11) considered the protection provided by the seismic Category II turbine building first bay building structure and associated missile barriers, consistent with the provisions in Subsections 3.2.1.1.1, 3.2.1.1.2, 3.2.2.5, and 3.2.2.6. Since the turbine building first bay is surrounded by a large amount of structures, equipment and components in the turbine building and adjacent annex building, large deformable missiles (represented by automobile) passing through turbine building and annex building into turbine building first bay is a highly tortuous path. The structures, equipment and components in the turbine building and part of the annex building are credited to stop or break apart an automobile missile and protect Wall 11 from an automobile missile.
In accordance with the missile identification and protection criteria provided in Section 3.5, a realistic assessment of potential missile paths was conducted. Where the line of sight of a missile passing through a turbine building first bay building structure opening could potentially result in missile impact upon a Wall 11 opening, the missile was considered credible, and additional protection was provided or additional evaluation performed. As a result of this analysis, two missile barriers designed to stop the 8-inch artillery shell and 1-inch sphere missiles are provided within the interior of the turbine building first bay building structure, at elevation 117'-6, northeast of MSIV Compartment B, and at elevation 100', north of Valve/Piping Penetration Room. Where the 1-inch diameter solid steel sphere or 8-inch artillery shell missiles were determined to have a line of sight to a Wall 11 penetration, analysis was performed to confirm that there was no line of sight through the penetration, the contents of the penetration would prevent the missile from passing through Wall 11, or equipment required to achieve safe shutdown could not be adversely affected.
3.5-14 Revision 9
VEGP 3&4 - UFSAR The steel tornado missile barriers located within the turbine building first bay are designed and analyzed in accordance with the barrier design procedures contained in Subsection 3.5.3 and applicable ANSI/AISC N690 requirements.
The steel tornado missile barriers located within the turbine building first bay, identified in Table 3.5-1, are designed and analyzed in accordance with the barrier design procedures and ductility requirements contained in Subsection 3.5.3 and applicable ANSI/AISC N690 requirements. The missile barriers are permanently anchored to the turbine building first bay or auxiliary building as applicable. Where anchored to concrete, the anchors conform with the anchorage requirements contained in Subsection 3.8.4.5.1. The configuration of the barriers is based on the lines of sight to be eliminated and the missile sizes pertinent to those lines of sight. The configuration also provides for functional requirements such as airflow, venting, and personnel access. Steel barrier designs are provided in both solid and grating-type configurations to address the required missile protection and functional requirements on a case-by-case basis. The missile protection provided for Wall 11 openings and penetrations is tabulated in Table 3.5-1. Protection of Wall 11 openings from external missiles is provided by the seismic Category II turbine building first bay building structure, the seismic Category II missile barriers located within the turbine building first bay building structure, the seismic Category I Wall 11 doors, and the seismic Category I Wall 11 penetrations and spare penetration covers within missile lines of sight.
3.5.2.2 Protection of MSSV Functions from Externally Generated Missile The main steam safety valve (MSSV) vent stacks are evaluated for tornado missiles in the horizontal and vertical directions. Based on elevation of the vent stacks, site topography, and arrangement of surrounding buildings, large debris (represented by the automobile) does not have a line of sight to the MSSV vent stacks. For horizontal missile impact, it is analyzed and confirmed that the steam can be relieved through Wall 11 doors and vents into turbine building first bay without impacting the pressure limit in the MSIV rooms and Valve/Piping Penetration Room if a vent stack is crimped. For the vertical missile impact, it is evaluated and confirmed that 1) ASME limits are not exceeded under impact of small debris; 2) The main steam pressure boundary remains intact under the impact of the 8-inch artillery shell and 1-inch sphere missiles; 3) the passive safety features are not impacted.
Therefore, the tornado missile impact on the MSSV vent stacks does not affect safe shutdown.
3.5.2.3 Protection from Externally Generated Missile through Wall I Evaluation of the openings in the east wall of the auxiliary building (Wall I) considered the protection provided by the seismic Category II annex building structures and associated missile barriers. Since the auxiliary building is surrounded by a large amount of structures, equipment, and components in the annex building, large deformable missiles (represented by automobile) passing through the annex building into auxiliary building is a highly tortuous path. The structures, equipment and components in the annex building are credited to stop or break apart an automobile missile and protect Wall I from an automobile missile.
In accordance with the missile identification and protection criteria provided in Section 3.5, a realistic assessment of potential missile paths was conducted. Where the line of sight of a missile passing through the annex could potentially result in missile impact upon a Wall I opening or penetration, the missile was considered credible, and additional protection was provided or additional evaluation performed. The rooms with line of sight are 12351, 12421, 12501, and 12541. The missile protection provided for Wall I openings and penetrations is tabulated in Table 3.5-2.
3.5-15 Revision 9
VEGP 3&4 - UFSAR 3.5.2.4 Protection from Externally Generated Missile through Shield Building Air Inlets Evaluation of the openings in the shield building air inlets considered the protection provided by the seismic Category I missile barriers and air baffles, and seismic Category II missile barriers, consistent with the provisions in Subsections 3.2.1.1.1, 3.2.1.1.2, 3.2.2.5, and 3.2.2.6.
In accordance with the missile identification and protection criteria provided in Section 3.5, a realistic assessment of potential missile paths was conducted. Where the line of sight of a missile passing through the shield building air inlets could potentially result in missile impact upon a safe shutdown required SSC, the missile was considered credible, and additional protection was provided or additional evaluation performed. As a result of this analysis, potential missile paths through some of the upper and lower air inlet openings to some PCS piping and Class 1E cabling/conduit are identified. Four missile barriers are provided on the outside of the exterior circular walkway of the shield building to prevent the 1-inch diameter solid steel sphere or 8-inch artillery shell missiles from entering the upper air inlet openings. The external shield building walkway prevents the 1-inch diameter solid steel sphere and 8-inch artillery shell missiles from entering the lower air inlet openings. The automobile missile is too large to enter the air inlets and therefore can be stopped by the shield building wall. The missile protection provided for the shield building air inlet openings is tabulated in Table 3.5-3.
3.5.3 Barrier Design Procedures Missile barriers and protective structures are designed to withstand and absorb missile impact loads to prevent damage to safety-related components.
Formulae used for missile penetration calculations into steel or concrete barriers are the Modified National Defense Research Committee (NDRC) formula for concrete and either the Ballistic Research Laboratory (BRL) or Stanford formulae for steel.
Concrete (Modified NDRC Formula) 0.5 V
1.8 x = 4 KNWd x 1000 d for 2.0 d
1.8 V x x = KNW +d for > 2.0 1000 d d where x = penetration depth, inches W = missile weight, lbs d = missile diameter, inches N = missile shape factor = 1.0 V = impact velocity, feet/sec 180 K = experimentally obtained material coefficient for penetration = fc fc = concrete compressive strength 3.5-16 Revision 9
VEGP 3&4 - UFSAR Scabbing thickness, t s , and perforation thickness, tp is given by:
ts x x
= 2.12 + 1.36 for 0.65 11.75 d d d 2
ts x x x
= 7.91 - 5.06 for 0.65 d d d d tp x x
= 1.32 + 1.24 for 1.35 13.5 d d d tp x x
= 3.19 ( ) - 0.718 ( )2 d d d Steel (Stanford Formula)
E S W
= 16,000 T 2 + 1,500 T D 46,500 Ws Where:
E = critical kinetic energy required for perforation, foot pounds D = effective missile diameter, inches S = ultimate tensile strength of the target (steel plate), pounds per square inch T = target plate thickness, inches W = length of a square side between rigid supports, inches Ws = length of a standard window, 4 inches The ultimate tensile strength is directly reduced by the amount of bilateral tension stress already in the target. The equation is good within the following ranges:
0.1 < T/D < 0.8, 0.002 < T/L < 0.05, 10 < L/D < 50, 5 < W/D <8, 8 < W/T < 100, 70 < V < 400 3.5-17 Revision 9
L = missile length, inches V = impact velocity, feet/second Steel ( BRL Formula )
(N
WS
'
Where:
tp = steel plate thickness for threshold of perforation, inches D = equivalent missile diameter, inches Ek = missile kinetic energy, foot pounds
= M V2/2 M = mass of the missile, lb-sec2/ft.
In using the Modified NDRC, BRL and Stanford formulae for missile penetration, it is assumed that the missile impacts normal to the plane of the wall on a minimum impact area and, in the case of reinforced concrete, does not strike the reinforcing. Due to the conservative nature of these assumptions, the minimum thickness required for missile shields is taken as the thickness just perforated.
Structural members designed to resist missile impact are designed for flexural, shear, and buckling effects using the equivalent static load obtained from the evaluation of structural response. Stress and strain limits for the equivalent static load comply with applicable codes and Regulatory Guide 1.142, and the limits on ductility of steel structures as given in Subsection 3.5.3.1. The consequences of scabbing are evaluated if the thickness is less than the minimum thickness to preclude scabbing.
The thicknesses of the exterior walls above grade and of the roof of the nuclear island are 24 inches and 15 inches, respectively. The roof is constructed using left-in-place metal deck. These thicknesses exceed the minimum thicknesses for Region II tornado missiles specified in Standard Review Plan 3.5.3.
3.5.3.1 Ductility Factors for Steel Structures Ductility factors for the design of steel structures are as follows:
For tension due to flexure, < 10.0 For columns with slenderness ratio (L/r) equal to or less than 20, < 1.3 For columns with slenderness ratio greater than 20, < 1.0 Where: L = effective length of the member r = the least radius of gyration For members subjected to tension, < .5*(eu/ey)
Where: eu = ultimate strain ey = yield strain 3.5-18 Revision 9
VEGP 3&4 - UFSAR 3.5.4 Combined License Information The evaluation for those external events that produce missiles that are more energetic than the tornado missiles postulated for design of the AP1000 is addressed in APP-GW-GLR-020 (Reference 1).
In addition, the VEGP site satisfies the site interface criteria for wind and tornado (see Subsections 3.3.1.1, 3.3.2.1 and 3.3.2.3) and will not have a tornado-initiated failure of structures and components within the applicants scope that compromises the safety of AP1000 safety-related structures and components (see also Subsection 3.3.3).
Subsection 1.2.2 discusses differences between the plant specific site plan and the AP1000 typical site plan.
There are no other structures adjacent to the nuclear island other than as described and evaluated in this document.
Missiles caused by external events separate from the tornado are addressed in Subsections 3.5.1.3, 3.5.1.5, and 3.5.1.6.
3.5.5 References
- 1. APP-GW-GLR-020, Wind and Tornado Site Interface Criteria, Westinghouse Electric Company LLC.
201. (APO 2006) APO Terminal Area Forecast Summary Report, Federal Aviation Administration, http://www.apo.data.faa.gov/wtaf/, issued February 2006, accessed 5/2/
2006.
202. (Atlanta 2005) Atlanta Sectional Aeronautical Chart, 74th Edition, U.S. Department of Transportation, Federal Aviation Administration, March 17, 2005.
203. (DOE 1996) Accident Analysis for Aircraft Crash into Hazardous Facilities, DOE Standard, DOE-STD-3014-96, US Department of Transportation, October 1996.
204. (Westinghouse 2001) Nuclear Island General Arrangement, AP1000 Advanced Passive Light Water Reactor, Rev. 0, Section B-B, DCD Number APP 1000 P2 902, Westinghouse Electric Company, 08/06/2001.
3.5-19 Revision 9
VEGP 3&4 - UFSAR Table 3.5-1 (Sheet 1 of 2)
External Missile Protection Provided for Auxiliary Building Wall 11 Openings Wall 11 Opening Protected Room Elevation Missile(1) Protection(2)(3)
Room 12306 doorway Valve/Piping 100'-0 Sphere First bay building structure Penetration Room First bay interior missile barrier Artillery Shell First bay building structure first bay interior missile barrier Automobile First bay building structure Structure, equipment, and components in turbine building and annex building Room 12404 vent Lower MSIV 117'-6 Sphere First bay building structure Compartment B First bay interior missile barrier Artillery Shell First bay building structure First bay interior missile barrier Automobile First bay building structure Structure, equipment, and components in turbine building and annex building Room 12405 doorway Lower VBS B&D 117'-6 Sphere Room 12405 Wall 11 door Equipment Room Artillery Shell Room 12405 Wall 11 door Automobile First bay building structure Structure, equipment, and components in turbine building and annex building Room 12406 vent Lower MSIV 117'-6 Sphere First bay building structure Compartment A Artillery Shell First bay building structure Automobile First bay building structure Structure, equipment, and components in turbine building and annex building Room 12504 doorway Upper MSIV 135'-3 Sphere First bay building structure Compartment B Artillery Shell First bay building structure Automobile First bay building structure Structure, equipment, and components in turbine building and annex building Room 12505 doorway Upper VBS B&D 135'-3 Sphere Room 12505 Wall 11 door Equipment Room Artillery Shell Room 12505 Wall 11 door Automobile First bay building structure Structure, equipment, and components in turbine building and annex building Room 12506 doorway Upper MSIV 135'-3 Sphere First bay building structure Compartment A Artillery Shell First bay building structure Automobile First bay building structure Structure, equipment, and components in turbine building and annex building 3.5-20 Revision 9
VEGP 3&4 - UFSAR Table 3.5-1 (Sheet 2 of 2)
External Missile Protection Provided for Auxiliary Building Wall 11 Openings Wall 11 Opening Protected Room Elevation Missile(1) Protection(2)(3)
Wall 11 penetrations Various Various Sphere First bay building structure Pipe sleeves Penetration contents Spare penetration covers Artillery Shell First bay building structure Pipe sleeves Penetration contents Spare penetration covers Automobile First bay building structure Structure, equipment, and components in turbine building and annex building Notes:
- 1. Tornado missiles are defined in Subsection 3.5.1.4.
- 2. Turbine building first bay is a seismic Category II building structure. Wall 11 doors, Wall 11 penetration sleeves, and Wall 11 spare penetration covers that provide tornado missile protection are equipment Class C. Barriers in the turbine building first bay building structure that provide tornado missile protection are equipment Class D.
- 3. The materials of construction for the missile barriers installed within the turbine building first bay and on Wall 11 shall be steel as specified below, or steel with equal or better material properties:
- First bay interior missile barrier at elevation 117'-6
- barrier plate - ASTM A572, Gr. 50
- barrier support frame - ASTM A500, Gr. B
- anchors - ASTM A325 or ASTM A490, and ASTM F1554, Gr. 105 as required
- First bay interior missile barrier at elevation 100'-0
- barrier plate - ASTM A572, Gr. 50
- barrier support frame - ASTM A500, Gr. B
- anchors - manufacturer standard
- Wall 11 missile doors
- door plate - ASTM A572, Gr. 50
- embed plates - ASTM A572, Gr. 50
- anchors - ASTM A 1064
- Wall 11 spare penetration covers
- ASTM A240
- Pipe sleeves
- material shall be as specified for their penetration function 3.5-21 Revision 9
VEGP 3&4 - UFSAR Table 3.5-2 External Missile Protection Provided for Annex Building Wall I Openings(3)
Wall I Opening Protected Room Elevation Missile(1) Protection(2)
Room 12351 Maintenance Floor 107'-2 Sphere Room 12351 Wall I door doorway Staging Area Artillery Shell Room 12351 Wall I door Automobile Room 12351 Wall I door Room 12421 Non 1E Equipment/ 117'-6 Sphere Pipe sleeves electrical Penetration Room Penetration contents penetrations Spare penetration covers Artillery Shell Pipe sleeves Penetration contents Spare penetration covers Automobile Annex Building Wall along column line E Room 12501 VBS MCR/A&C 135'-3 Sphere Room 12501 Wall I door doorway Equipment Room Artillery Shell Security grating Electrical control panels Room 12501 Wall I door Automobile Structures, equipment, and components in annex building Room 12541 Operating Deck 135'-3 Sphere Room 12556 Wall I door doorway Staging Area Artillery Shell Room 12556 Wall I door Automobile Room 12556 Wall I door Notes:
- 1. Tornado missiles are defined in Subsection 3.5.1.4.
- 2. Annex Building Areas 1, 2 and 3 are a Seismic Category II building structure. Wall I doors, Wall I penetration sleeves and Wall I spare penetration covers that provide tornado missile protection are equipment Class C. Barriers in the Annex Building Areas 1, 2 and 3 building structures that provide tornado missile protection are equipment Class D.
- 3. Other Wall I door and penetrations are protected from the tornado missiles by the Seismic Category II Annex Building Structure.
3.5-22 Revision 9
VEGP 3&4 - UFSAR Table 3.5-3 External Missile Protection Provided for Shield Building Air Inlet Openings Opening Protected Room Elevation Missile Protection Shield Building Upper Annulus 251'-8 Sphere External Shield Building Walkway Lower Air Inlet Artillery Shell External Shield Building Walkway Openings Automobile Shield Building Wall around Air Inlets Shield Building Upper Annulus 257'-9 Sphere Upper Air Baffle (SB02)
Upper Air Inlet External Shield Building Walkway Openings Missile Barriers Artillery Shell Upper Air Baffle (SB02)
External Shield Building Walkway Missile Barriers Automobile Shield Building Wall around Air Inlets 3.5-23 Revision 9
VEGP 3&4 - UFSAR Table 3.5-201 Augusta APO Terminal Area Forecast Summary Report - Itinerant Operations General Air Taxi & Commercial Air Year Aviation Commuter Carrier Military Total 1990 22,023 14,941 6,495 4,522 47,981 1991 19,175 9,462 6,576 3,242 38,455 1992 17,872 9,393 7,196 3,221 37,682 1993 16,902 8,821 6,455 4,068 36,246 1994 16,896 5,961 6,473 3,727 33,057 1995 16,597 8,876 5,024 3,511 34,008 1996 17,016 9,325 4,225 2,780 33,346 1997 18,995 8,304 4,599 2,561 34,459 1998 19,611 7,518 5,028 2,271 34,428 1999 22,653 6,954 5,183 2,841 37,631 2000 21,975 6,663 4,969 3,354 36,961 2001 19,961 7,378 4,929 2,954 35,222 2002 20,085 7,164 4,286 3,082 34,617 2003 17,622 9,058 4,393 2,843 33,916 2004 18,658 9,441 4,934 2,528 35,561 2005 13,307 8,226 4,585 1,799 27,917 2006 13,618 8,328 4,585 1,799 28,330 2007 13,937 8,432 4,585 1,799 28,753 2008 14,263 8,537 4,585 1,799 29,184 2009 14,597 8,644 4,585 1,799 29,625 2010 14,939 8,751 4,585 1,799 30,074 2011 15,288 8,860 4,585 1,799 30,532 2012 15,646 8,971 4,585 1,799 31,001 2013 16,012 9,083 4,585 1,799 31,479 2014 16,387 9,196 4,585 1,799 31,967 2015 16,611 9,310 4,585 1,799 32,305 2016 16,837 9,426 4,585 1,799 32,647 2017 17,067 9,544 4,585 1,799 32,995 2018 17,300 9,663 4,585 1,799 33,347 2019 17,536 9,783 4,585 1,799 33,703 2020 17,776 9,905 4,585 1,799 34,065 2021 18,018 10,028 4,585 1,799 34,430 2022 18,264 10,153 4,585 1,799 34,801 2023 18,514 10,280 4,585 1,799 35,178 2024 18,766 10,408 4,585 1,799 35,558 2025 19,023 10,538 4,585 1,799 35,945 Source: Reference 201 3.5-24 Revision 9
VEGP 3&4 - UFSAR Table 3.5-202 Deleted in Revision 2 3.5-25 Revision 9
VEGP 3&4 - UFSAR McCormick Co.
Edgefield Aiken Site location Co.
Co.
30-mile radius !
Columbia Co.
Orangeburg Co.
Harman Airport Daniel V4 Airport with Control Tower 17 Augusta V1 Class B Airspace Low Ft. Regional 8 5 Barnwell Co. Altitude Federal Airways Gordon SO Richmond Co. U TH VR CA 9 7 -1 004 Rea RO Savannah 05 9
Military Training Route R-3 Patch GE LI O NA River RG IA Site BULLDOG Military Operations D MOA NATIONAL Par Pond SECURITY Area Jefferson Barnwell AREA Co.
Barrow Restricted Area Rhodes Air Ranch !
( Vogtle Site National Security Area Sa v an 85 na hR V1 .
Burke Co.
BULLDOG Burke Co.
Unnamed N A & B MOA BULLDOG B MOA Allendale 0 5 10 Wade Co. Miles Millhaven Millen VR 97 Aiport locations derived from
-1059 FAA Sectional Aeronautical Chart, Atlanta, 1:500,000 Allendale Co.
V70 Hampton Co.
Emanuel Co.
Landings East Jenkins Sylvania Co. Screven Co.
Source: Reference 202 Figure 3.5-201 Airports Within 30 Miles of Vogtle Facility 3.5-26 Revision 9
VEGP 3&4 - UFSAR 3.6 Protection Against the Dynamic Effects Associated with the Postulated Rupture of Piping The effects of a postulated pipe rupture in the AP1000 are of several types. This section considers the effects that are localized to the area of the break and are a result of the dynamic effects of the pipe rupture including jet impingement, pipe whip, subcompartment pressurization, and fluid system decompression. This section describes the evaluation of the potential for and effects of these dynamic effects. It describes measures taken to protect systems and equipment from dynamic effects of pipe rupture when necessary. This section also considers the effects of spray wetting and flooding from pipe ruptures and cracks.
Chapters 6 and 15 discuss the response of the system to changes in flow and pressure and loss of coolant and the response of the containment to the pressure and temperature changes. Pressure due to a break in a high energy line in the auxiliary building is vented into an adjacent building or to the atmosphere. The design transients listed in Subsection 3.9.1 are used in evaluating the components of the reactor coolant system for effects due to internal pressure and temperature changes from postulated accidents. Section 3.11 discusses the qualification of the equipment required to function in the adverse environmental conditions including temperature, humidity, pressure, radiation, spray wetting, and chemical consequences.
Pipe failure protection is provided according to the requirements of 10 CFR 50, Appendix A, General Design Criterion 4. In the event of a high- or moderate-energy pipe failure within the plant, adequate protection is provided so that essential structures, systems, or components are not impacted by the adverse effects of postulated piping failure. Essential systems and components are those required to shut down the reactor and mitigate the consequences of the postulated piping failure. Nonsafety-related systems are not required to be protected from the dynamic and environmental effects associated with the postulated rupture of piping except as described in Subsection 3.6.1.1, item Q.
The criteria used to evaluate pipe failure protection are generally consistent with NRC guidelines including those in the Standard Review Plan Sections 3.6.1 and 3.6.2, NUREG-1061, Volume 3 (Reference 11) and applicable Branch Technical Positions.
Subsection 3.6.1 provides the design bases and criteria for the analysis required to demonstrate that essential systems are protected. The high- and moderate-energy systems representing the potential source of adverse effects are listed. Additionally, the criteria for separation and the effects of adverse consequences are defined.
Subsection 3.6.2 defines the criteria for postulated break location and configuration. High-energy pipes are evaluated for the effects of circumferential and longitudinal pipe breaks and through-wall cracks. Moderate-energy pipes are evaluated for the effects of through-wall cracks. Analysis methods and criteria for evaluating pipe whip and evaluating the consequences of jet impingement, motions of the pipe, and system depressurization on integrity and operability are provided. The evaluation of containment penetrations, pipe whip restraints, guard pipes, and other protective devices is also described. The criteria for excluding breaks in high-energy piping adjacent to containment penetrations are also provided.
Evaluation of the dynamic effects of postulated breaks in the reactor coolant loop, main steam lines inside containment, and other primary piping inside containment equal to or greater than the 6-inch nominal pipe size (NPS) is eliminated for AP1000 based on mechanistic pipe break (leak-before-break) considerations. Those sections of high-energy piping that qualify for mechanistic pipe break are evaluated for only the effects of leakage cracks.
Subsection 3.6.3 describes the application of leak-before-break criteria to permit the elimination of pipe rupture dynamic effects considerations. Design guidelines aid in the design of piping systems 3.6-1 Revision 9
VEGP 3&4 - UFSAR that satisfy the requirements for mechanistic pipe break. Dynamic effects of postulated breaks are evaluated for those analyzable sections of high-energy piping systems that do not use the mechanistic pipe break methods.
The safety analyses in Chapter 15 and the requirements for emergency core cooling discussed in Section 6.3 and the environmental qualification of equipment discussed in Section 3.11 of this report are not changed by the use of mechanistic pipe break considerations for pipe rupture dynamic effects evaluations. Chapter 6 describes the containment subcompartment pressurization analyses including mechanistic pipe break considerations.
3.6.1 Postulated Piping Failures in Fluid Systems Inside and Outside Containment A number of systems and components are necessary to shut the plant down in the event of a pipe rupture. These systems, termed essential systems, are protected from the postulated pipe ruptures.
The essential systems for various pipe ruptures are the reactor coolant system, the steam generator system, the passive core cooling system, and the passive containment cooling system. In addition to these fluid systems, the protection and safety monitoring system and the Class 1E dc and UPS system are essential. The main control room and main control room habitability system are also protected as essential systems. In addition, containment penetrations and isolation valves (including those for nonessential systems) are essential.
Most of the equipment required for plant safety or safety-related shutdown is located inside containment. The piping inside containment also represents the most significant piping relative to plant safety and, therefore, is subject to the most stringent design and analysis requirements.
Essential equipment in the vicinity of piping that does not satisfy leak-before-break criteria is protected as required by the use of protective structures, pipe restraints, and separation. The need for protection of essential structures, systems and components is determined by evaluation of the dynamic effects. The design bases and criteria for the evaluation follow.
Evaluations are made based upon circumferential or longitudinal pipe breaks, through-wall cracks, or leakage cracks as determined by the appropriate criteria. At locations determined to be subject to a circumferential or longitudinal pipe break, dynamic effects such as jet impingement and pipe whip are evaluated.
At locations subject to through-wall cracks, only environmental effects (temperature, humidity, pressure, radiation, spray wetting and chemical consequences) and flooding are evaluated.
The pressurization loads on structures and components are evaluated for postulated circumferential breaks and longitudinal breaks in piping that does not meet leak-before-break requirements and for postulated leakage cracks in piping that meets the leak-before-break requirements. See Subsection 3.8.3.4 and Subsection 3.8.4.3.1.4 for a discussion of pressurization loads on structures.
The in-containment refueling water storage tank is evaluated for pressurization as described in Subsection 3.6.1.2.1.
Pressurization loads for pipe failures in the main steam and feedwater break exclusion zones for high-energy lines in the vicinity of containment penetrations are evaluated for a 1.0 square foot break. Structures in the steam generator blowdown break exclusion zone are evaluated for subcompartment pressurization effects due to worst case circumferential pipe rupture in the 4-inch steam generator blowdown piping. Pipe whip and jet impingement are not evaluated for structures in the break exclusion zones per NRC Branch Technical Position MEB 3-1, section B.1.b, except that the east wall and the floor at elevation 117-6 of the east main steam subcompartment is designed 3.6-2 Revision 9
VEGP 3&4 - UFSAR for pipe whip and jet impingement loads for worst case breaks in either the main steam line or the main feedwater line. See Subsection 3.6.1.2.2.
3.6.1.1 Design Basis The following design bases relate to the evaluation of the effects of the pipe failures at locations determined in Subsection 3.6.2.
A. The selection of the failure type is based on whether the system is high or moderate-energy during normal operating conditions of the system. High-energy piping includes those systems or portions of systems in which the maximum normal operating temperature exceeds 200°F or the maximum normal operating pressure exceeds 275 psig. Piping systems or portions of systems pressurized above atmospheric pressure during normal plant conditions and not identified as high-energy are considered moderate-energy. Piping systems that exceed 200°F or 275 psig for two percent or less of the time during which the system is in operation or that experience high-energy pressures or temperatures for less than one percent of the plant operation time are considered moderate-energy.
B. The following assumptions are used to determine the thermodynamic state in the piping system for the calculation of fluid reaction forces:
- 1. For those portions of piping systems normally pressurized during operation at power, the thermodynamic state in the pipe and associated reservoirs is that of normal full-power operation.
- 2. For those portions of piping systems pressurized only during other normal plant conditions (for example, startup, hot standby, reactor cooldown), the thermodynamic state and associated operating condition are determined as the mode giving the most severe fluid reaction forces. Moderate-energy systems that are occasionally at higher temperature or pressure (see design basis A.) are not evaluated for pipe failures at the high-energy conditions.
- 3. High-stress pipe rupture locations are based on calculated stresses due to Level A and Level B loading. Seismic loads are not included.
C. Circumferential and longitudinal breaks in high-energy pipes, except in pipes satisfying leak-before-break requirements, are evaluated for effects including subcompartment pressurization, pipe whip, jet impingement, jet reaction thrust, internal fluid decompression loads, spray wetting, and flooding.
D. High-energy and moderate-energy pipe through-wall cracks are evaluated for environmental and flooding effects. Moderate-energy systems that are occasionally at higher temperature or pressure (see design basis A.) are evaluated for environmental effects at only the moderate-energy conditions. Dynamic effects are not evaluated for these cracks.
E. Through-wall cracks are not postulated in the break exclusion zones. The effects of flooding, spray wetting, and subcompartment pressurization are evaluated for a postulated 1.0 square foot break for the main steam and feedwater lines.
F. Where postulated, each longitudinal or circumferential break in high-energy fluid system piping, leakage crack in high-energy piping with mechanistic pipe break, or through-wall crack in high-energy or moderate-energy fluid system piping is considered separately as a single initial event occurring during normal plant conditions.
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VEGP 3&4 - UFSAR For systems not seismically analyzed for a safe shutdown earthquake, the safe shutdown earthquake is assumed to cause a pressure boundary failure, as described in Subsections 3.6.2.1.1.3 and 3.6.2.1.2.2.
G. AC power is not required for the actuation of the passive safety systems. The only electrical system required to function is the Class 1E dc and UPS system.
H. A single active component failure is assumed in systems used to mitigate the consequences of the postulated piping failure or to safely shut down the reactor. The single active component failure is assumed to occur in addition to the postulated piping failure and any direct consequences of the piping failure, such as unit trip and loss of offsite power.
I. The function of the containment to act as the ultimate heat sink is maintained for any postulated pipe rupture.
J. Safety-related systems and components are used to mitigate the effects of postulated pipe ruptures. However, the seismic Category II wall separating the first bay of the turbine building from the main area of the turbine building (Wall 11.2) is credited with protecting the north wall of the auxiliary building (Wall 11) from high energy line break (HELB) loads resulting from postulated ruptures in the main steam and main feedwater piping north of the turbine building first bay. In addition, the turbine control and stop, moisture separator reheater 2nd stage steam isolation, and turbine bypass (steam dump) valves (which are not safety-related) are credited in single failure analyses to mitigate postulated steam line ruptures.
K. A whipping pipe is considered capable of rupturing impacted pipes of smaller nominal pipe diameter, irrespective of pipe-wall thickness. This is based on the assumption that only piping is determined to do the impacting. A whipping pipe is considered capable of developing a through-wall crack in a pipe of equal or larger nominal pipe size with equal or thinner wall thickness, assuming that only piping is determined to do the impacting. The preceding criterion is not used where the potential exists for valves or other components in the whipping pipe to impact the targets, since these are treated on a case-by-case basis.
L. Pipe whip is assumed to occur in the plane defined by the piping geometry and to cause movement in the direction of the jet reaction.
If unrestrained, a whipping pipe having a constant energy source sufficient to form a plastic hinge is considered to form a plastic hinge and rotate about the nearest rigid pipe whip restraint, anchor, or wall penetration capable of resisting the pipe whip loads or the calculated dynamic plastic hinge location.
If the direction of the initial pipe movement caused by the thrust force is such that the whipping pipe impacts a flat surface normal to its direction of travel, it is assumed that the pipe comes to rest against that surface, with no pipe whip in other directions.
Pipe whip restraints are provided wherever postulated pipe breaks could impair the capability of any essential system or component to perform its intended safety functions.
M. The calculation of thrust and jet impingement forces considers any line restrictions (that is, flow limiter) between the pressure source and break location and the absence of energy reservoirs, as applicable.
N. Breaks are not postulated to occur in pump and valve bodies since the wall thickness exceeds that of connecting pipe.
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VEGP 3&4 - UFSAR O. Components impacted by jets from breaks in piping containing high-pressure (870 to 2466 psia) steam or subcooled liquid (subcooled no more than 126°F) that would flash at the break, such as piping connected to the steam generators or reactor coolant loops, are evaluated as follows:
- 1. Impacted components within 10 piping inside diameters of the broken pipe are assumed to fail. Specific jet loads are calculated and evaluated only when failure of the component, when combined with a single active failure, could adversely affect safe shutdown or accident mitigation capability. These jet loads are calculated according to Subsection 3.6.2.3.1.
- 2. Components beyond 10 inside diameters of the broken pipe are considered to be undamaged by the jet and are not analyzed. The basis for these criteria is contained in NUREG/CR-2913 (Reference 1).
P. Pipe breaks are not postulated to occur in systems for which postulated leakage cracks have been shown to be stable for worst case loadings. (See Subsection 3.6.3.) Leak detection systems are provided that are capable of detecting the leakage from a postulated leakage crack.
For these systems, leakage cracks are postulated and evaluated for subcompartment pressure loads on structures and components. When the mechanistic pipe break approach is used, subcompartment pressure loads on structures and essential components are based on the small leakage crack determined from the mechanistic pipe break approach. Where the subcompartment includes lines not qualified for mechanistic pipe break, subcompartment pressurization is evaluated for a break in the line with the largest effect.
The leakage crack effects of jet impingement, pipe whip, and internal fluid system loads are considered negligible and are not evaluated. The leakage crack effects of flooding and environmental effects are less limiting than the corresponding effects for postulated high-energy through-wall cracks. These through-wall cracks are not eliminated by mechanistic pipe break.
Q. Nonessential systems, structures, and components are not required to meet the criteria outlined in this section. However, while none of the nonessential systems are needed during or following a pipe break event, pipe whip protection is evaluated in cases where a high-energy nonessential system failure could initiate a failure in an essential system or component or where a high-energy nonessential system failure could initiate a failure in another nonessential system whose failure could affect an essential system.
R. The escape of steam, water, combustible or corrosive fluids, gases, and heat in the event of a pipe rupture will not preclude:
Subsequent access to any areas, as required, to recover from the postulated pipe rupture Habitability of the control room Capability of essential instrumentation, electric power supplies, components, and controls to perform safety functions to the extent necessary to meet the criteria outlined in this section 3.6.1.2 Description Essential systems are evaluated to demonstrate conformance with the design bases and to determine their susceptibility to the failure effects. Table 3.6-1 identifies systems which contain high 3.6-5 Revision 9
VEGP 3&4 - UFSAR and moderate-energy lines. The systems listed include all high- and moderate-energy systems inside containment plus the high- and moderate-energy systems in the auxiliary building near containment penetrations (including access hatches), the main control room, the Class 1E dc and UPS system or the portions of the passive containment cooling system located in the auxiliary building. The table does not list systems that operate at or close to atmospheric pressure including air handling and gravity drains. High energy system piping in the turbine building adjacent to the auxiliary building is evaluated for potential effects on the main control room. These systems are included on Table 3.6-1.
The definition of high and moderate-energy systems is provided in paragraph A of Subsection 3.6.1.1.
The postulated break, through-wall crack, and leakage crack locations are determined according to Subsections 3.6.2 and 3.6.3.
Equipment is considered to be separated from the dynamic effects of pipe rupture when the equipment is located in a different subcompartment. For the case of pipe whip, equipment may be considered separated for dynamic effects based on the distance from the pipe and the length of pipe that is moving. For the case of jet impingement in a line with saturated or subcooled fluid, equipment more than ten inside pipe diameters from the break location and the tip of the pipe whip trajectory, including the resting location of the broken pipe, is considered separated for dynamic effects.
Equipment located in the same subcompartment as a break, through-wall crack, or leakage crack is subject to potential environmental and flooding effects. Equipment may also be subject to environmental and flooding effects of steam and water vented into a subcompartment from an adjoining subcompartment.
3.6.1.2.1 Pressurization Response Pressurization response analyses are performed for subcompartments containing high-energy piping for which break locations are defined by Subsections 3.6.2.1.1.1, 3.6.2.1.1.2, and 3.6.2.1.1.3 or postulated leakage flaws are defined based on Subsection 3.6.3.3. Table 3.6-2 identifies those pipe breaks considered for the evaluation of the effects of pressurization loads on subcompartments. The pipe breaks inside containment that are postulated in piping that is not evaluated to the leak-before-break requirements of Subsection 3.6.3 are summarized in Table 3.6-2. The subcompartments are identified using the room numbers and room names given on Figures 1.2-4 through 1.2-10 as supplemented by Table 3.6-2. The subcompartments inside containment are designed to accommodate the pressurization loads from these breaks. In order to account for high stress break locations and the additional pressure boundary leakages from manways and flanges, pressurization loads on compartments inside containment enclosing high-energy piping are designed as described in Subsection 3.8.3.4.
There is no high-energy piping that can pressurize the annulus between the containment vessel and the shield building. Guard pipes are provided for the main steam, feedwater, startup feedwater and steam generator blowdown containment penetrations passing through the annulus as shown on Figure 3.8.2-4. The chemical and volume control system makeup piping is classified as high energy due to its design pressure, but does not cause pressurization because it is at ambient temperature.
The pressurization loads for the in-containment refueling water storage tank are based on the pressure and hydrodynamic loads due to the maximum discharge through the first, second, and third stages of the automatic depressurization system valves.
The pressurization loads for the reactor vessel annulus for the evaluation of asymmetric compartment pressurization are negligible based on a 5-gallon per minute leakage crack in the primary loop piping. The internal reactor pressure vessel asymmetric pressurization loads are based 3.6-6 Revision 9
VEGP 3&4 - UFSAR on a break in the largest pipe connected to the reactor coolant system that does not qualify for the application of mechanistic pipe break.
There are limited areas in the auxiliary building where the potential for pressurization loads from high-energy lines are considered. The pressurization loads for the steam tunnels are addressed in the discussion of loads due to a break in the break exclusion zone of the main steam and feedwater lines. The pressurization loads for the Elevation 100 containment penetration room containing the steam generator blowdown break exclusion zone are based on a circumferential rupture of the 4-inch steam generator blowdown piping. The areas through which the chemical and volume control system make-up line run, including the annulus between the containment and the containment shield building, are not subject to pressurization since the temperature of these lines is less than 212°F.
For a discussion of the criteria and analysis methods for subcompartment pressurization analysis, see Subsection 6.2.1.2. The analytical methods for transient mass distribution, used for pressure response analysis, are the same as described in WCAP-8077 (Reference 2).
3.6.1.2.2 Main Control Room Habitability The high-energy lines in closest proximity to the main control room are the main steam line and main feedwater line. The portions of these lines near the main control room are in the main steam line isolation valve compartment and are part of the break exclusion areas.
The main control room is separated from the isolation valve compartment by two structural walls. The areas between the two walls is used for nonessential office and administrative space associated with the control room. The walls separating the main control room from the main steam isolation valve compartment are thick, reinforced-concrete walls.
Consistent with the criteria for evaluation of leaks in the break exclusion area, the subcompartment, including the walls, is evaluated for the effects of flooding, spray wetting and subcompartment pressurization from a 1-square-foot break from either main steam or feedwater line within the respective break exclusion areas. The wall between the main steam line isolation valve compartment and the main control room, and the floor slab between the main steam line isolation valve compartment and the safety related electrical equipment room are also evaluated for pipe whip and jet impingement loads for worse case breaks in either the main steam line or the main feedwater line.
The subcompartment pressure loads from the 1-square-foot break are not combined with the pipe whip and jet impingement loads for the worse case breaks.
The effects upon the habitability of the main control room resulting from postulated pipe breaks and cracks in the auxiliary building are evaluated. In addition to pipe ruptures and cracks in lines in the auxiliary building, the main control room is evaluated for the dynamic effects and environmental effects of a postulated circumferential or longitudinal break of either the main steam line or main feedwater line in the turbine building.
Further description of the control room habitability systems, including options for remote shutdown, is provided in Section 6.4. The remote shutdown workstation is not subject to adverse effects of high-energy pipe rupture.
Main Steam Isolation Valve Compartment De-Pressurization Flow Paths The MSIV Compartment B (Figures 1.2-7, 1.2-8, 1.2-10, and 1.2-11), which is adjacent to the main control room, houses the main feedwater, startup feedwater, and main steam lines, as they pass to and from primary containment through the shield building, and to and from the turbine building, the MSIVs, the main steam safety valves (MSSVs), the power-operated relief valves (PORVs), the PORV block valves, and other essential components. As a result of a main steam or main feedwater 3.6-7 Revision 9
VEGP 3&4 - UFSAR high-energy line pipe break in MSIV Compartment B adjacent to the Main Control Room or MSIV Compartment A, the MSIV compartments de-pressurize by venting through the following pathways:
MSIV Compartment A
- drains/vents to Turbine Building first bay
- floor penetrations to Room 12306, which is the Valve/Piping Penetration Room located below, through Room 12306 door to Turbine Building first bay
- door to Turbine Building first bay
- roof vent to atmosphere MSIV Compartment B
- drains/vents to Turbine Building first bay
- door to Turbine Building first bay
- roof vent to atmosphere These pressure relief flow paths limit the design pressure loading while maintaining the environmental and structural integrity of the MSIV compartments, adjacent compartments, the Main Control Room, and equipment housed therein.
3.6.1.3 Safety Evaluation 3.6.1.3.1 General An analysis of postulated pipe failures is performed to determine the impact of such failures on those safety-related systems or components that provide protective actions and are required to mitigate the consequences of the failure. Through such protective measures, as separation, barriers, and pipe whip restraints, the effects of breaks, through-wall cracks, and leakage cracks are prevented from damaging essential items to an extent that would impair their essential function or necessary component operability.
Typical measures used for protecting the essential systems, components, and equipment are outlined in the next subsection and are discussed in Subsection 3.6.2. The capability of specific safety-related systems to withstand a single active failure concurrent with the postulated event is discussed, as applicable. When the results of the pipe failure effects analysis show that the effects of a postulated pipe failure are isolated, physically remote, or restrained by protective measures from essential systems or components, no further dynamic analysis is performed.
3.6.1.3.2 Protection Mechanisms The plant arrangement is based on maximizing the physical separation of redundant or diverse safety-related components and systems from each other and from nonsafety-related items.
Therefore, in the event a pipe failure occurs, there is a minimal effect on other essential systems or components required for safe shutdown of the plant or to mitigate the consequences of the failure.
The effects associated with a particular pipe failure are mechanistically consistent with the failure.
Thus, pipe dimensions, piping layouts, material properties, and equipment arrangements are considered in defining the specific measures for protection against the consequences of postulated failures.
Protection against the dynamic effects of pipe failures is provided by physical separation of systems and components, barriers, equipment shields, and pipe whip restraints. The precise method chosen depends largely upon considerations such as accessibility and maintenance. The preferred method of providing protection is by separation. When separation is not practical pipe whip restraints are used. Barriers or shields are used when neither separation nor pipe whip restraints are practical. This protection is not required when piping satisfies leak-before-break criteria.
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VEGP 3&4 - UFSAR Separation The plant arrangement provides separation, to the extent practicable, between redundant safety systems (including their appurtenances) to prevent loss of safety function as a result of events for which the system is required to be functional. Separation between redundant safety systems, with their related appurtenances, therefore, is the basic protective measure incorporated in the design to protect against the dynamic effects of postulated pipe failures.
In general, separation is achieved by:
Safety-related systems located remotely from high-energy piping, where practicable Redundant safety systems located in separate compartments, where practicable Specific components enclosed to retain the redundancy required for those systems that must function to mitigate specific piping failures Drainage systems provided for flooding control Pressure relief panels, safety-related, seismically qualified, and fire-rated barrier louvers and flappers to prevent flooding of essential areas Where physical separation is not possible, the pipe rupture hazard analysis includes an evaluation to determine the systems and components that require a structure for separation from the effects of a break in a high energy line. For these structures specifically included to separate breaks from essential systems or components, the evaluation considers that the break may be at the closest point in the line to the separating structure; not only at the break locations identified in Subsection 3.6.2.1.1. High energy lines qualified as leak-before-break lines and the lines in containment penetration break exclusion areas are not included as possible break locations in this evaluation. For a discussion of the information included in the pipe rupture hazard analysis see Subsection 3.6.2.5.
Barriers and Shields Protection requirements are met through the protection afforded by walls, floors, columns, abutments, and foundations. Where adequate protection does not already exist as a result of separation, a separating structure such as additional barriers, deflectors, or shields is provided to meet the functional protection requirements.
Inside the containment, the secondary shield wall serves as a barrier between the reactor coolant loops and the containment. In addition, the refueling cavity walls, operating floor, and secondary shield walls minimize the possibility of an accident that may occur in any one reactor coolant loop affecting the other loop or the containment. Those portions of the steam and feedwater lines located within the containment are routed in such a manner that possible interaction between these lines and the reactor coolant piping is minimized. The direct vessel injection valves for train A and train B are separated by the secondary shield wall.
Outside the containment, the wall separating the first bay of the turbine building from the main area of the turbine building (Wall 11.2) is credited with protecting the north wall of the auxiliary building (Wall
- 11) from HELB loads resulting from postulated ruptures in main steam and main feedwater piping north of the turbine building first bay. The wall is designed and analyzed for the loadings associated with the effects of postulated pipe ruptures in high energy lines north of the turbine building first bay in conjunction with the required load combinations identified in Subsection 3.8.4, and determined to meet the acceptance criteria contained in ACI 349-01 and ANSI/AISC N690-1994.
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VEGP 3&4 - UFSAR Barriers and shields that are identified as required by the pipe rupture hazard analysis are designed for loads from a break in the line at the closest location to the structure. This criterion is in conformance with the guidance of Branch Technical Position MEB 3-1. Rev. 2. Subsection 3.6.2.4 further discusses barriers and shields.
Piping Restraint Protection Measures for protection against pipe whip are provided where the unrestrained pipe movement of either end of the ruptured pipe could cause damage at an unacceptable level to any structure, system, or components required to meet the criteria outlined in this subsection.
Subsection 3.6.2.3 gives the design criteria for and description of pipe whip restraints.
3.6.1.3.3 Specific Protection Considerations The analysis of the consequences of pipe breaks, through-wall cracks, and leakage cracks uses the following criteria.
High-energy containment penetrations are subject to special protection mechanisms.
Restraints are provided to maintain the operability of the isolation valves and the integrity of the penetration due to a break in the safety-related and nonsafety piping beyond the restraint if required. These restraints are located as close as practicable to the containment isolation valves associated with these penetrations.
Instrumentation required to function following a pipe rupture is protected.
High-energy fluid system pipe whip restraints and protective measures are designed so that a postulated break in one pipe cannot lead to a rupture of other nearby essential pipes or components, if the secondary rupture results in consequences that are unacceptable for the initial postulated break.
For those cases in which the rupture of the main steam or feedwater piping inside containment is the postulated initiating event, the turbine control, turbine stop, moisture separator reheater 2nd stage steam isolation, and turbine bypass valves, and to a limited extent, the control systems for the turbine stop and feedwater control valves, are credited in single failure analysis to mitigate the event. This equipment is not protected from pipe ruptures in the turbine building because the postulated pipe rupture for which it provides protection is inside containment. The assumed single active failure for this analysis is the function of the safety-related valve that would normally isolate the piping. This isolation function is addressed in more detail in Chapter 10.
3.6.2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping This subsection describes the design bases for locating postulated breaks and cracks in high- and moderate-energy piping systems inside and outside the containment; the procedures used to define the jet thrust reaction at the break location; the procedures used to define the jet impingement loading on adjacent essential structures, systems, or components; pipe whip restraint design; and the protective assembly design. Pipe breaks in several high-energy systems, including the reactor coolant loop and surge line, are replaced by small leakage cracks when the leak-before-break criteria are applied. (See Subsection 3.6.3.) Jet impingement and pipe whip effects are not evaluated for these small leakage cracks.
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VEGP 3&4 - UFSAR 3.6.2.1 Criteria Used to Define High- and Moderate-Energy Break and Crack Locations and Configurations The NRC Branch Technical Position MEB 3-1 is used as the basis of the criteria for the postulation of high-energy pipe breaks and through-wall cracks, except for piping that satisfies the requirements for mechanistic pipe break, as described in Subsection 3.6.3.
A postulated high-energy pipe break is defined as a sudden, gross failure of the pressure boundary of a pipe either in the form of a complete circumferential severance (that is, a guillotine break) or as a sudden longitudinal, uncontrolled crack. For high-energy and moderate-energy fluid systems, pipe failures are also defined by postulation of controlled through-wall cracks in piping. For those piping lines that satisfy leak-before-break requirements, the guillotine breaks and sudden longitudinal cracks are replaced by postulated controlled leakage cracks.
Subsection 3.6.1 describes the evaluation and criteria for the effects of these breaks and cracks on the safety-related equipment.
3.6.2.1.1 High-Energy Break Locations The locations for postulated breaks in high-energy piping are dependent on the classification, quality group, and design standards used for the piping system. The break locations for high-energy piping are described in the following subsections. These locations are based on the design configuration and include changes due to the as-built piping configuration. As a result of piping reanalysis due to differences between the design configuration and the as-built configuration, the high stress and usage factor location may be shifted. The intermediate break (if any) locations need not be changed unless one of the following conditions exists:
A. The dynamic effects from new (as-built) intermediate break locations are not mitigated by the original pipe whip restraints and jet shields.
B. There is a significant change in pipe design parameters such as pipe size, wall thickness or pressure rating.
Breaks are not postulated in piping in the vicinity of containment penetrations. The portion of the piping that does not have postulated breaks is the break exclusion area. Subsection 3.6.2.1.1.4 identifies the requirements for the piping in the containment penetration break exclusion area.
Breaks are not postulated for those sections of pipe, including the reactor coolant loop and pressurizer surge line, that meet the requirements for leak-before-break as described in Subsection 3.6.3.
The leak-before-break methodology is applied to the candidate high-energy lines in the nuclear island identified in Appendix 3E. This appendix also identifies other high-energy lines in the nuclear island with diameters larger than 1 inch and the break exclusion areas inside and outside containment. The evaluation criteria for lines that do not satisfy the leak-before-break criteria are described in Subsection 3.6.2.
3.6.2.1.1.1 ASME Code,Section III, Division 1 - Class 1 Piping
[Pipe breaks are postulated to occur at the following locations in piping designed and constructed to the requirements for Class 1 piping in the ASME Code,Section III, Division 1.
At terminal ends of the piping, including:
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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- The extremity of piping connected to structures, components, or anchors that act as essentially rigid restraints to piping translation and rotational motion due to static or dynamic loading.
- Branch intersection points are considered a terminal end for the branch line unless the following are met: The branch and the main piping systems are modeled in the same static, dynamic and thermal analyses, and the branch and main run are of comparable size and fixity (that is, the nominal size of the branch is at least one-half of that of the main run).
- In piping runs that are maintained pressurized during normal plant conditions for only a portion of the run, the terminal end, for purposes of defining break locations, is the piping connection to the first normally closed valve.
At intermediate locations where the following conditions are satisfied:
- Intermediate locations where the maximum stress range as calculated by Equation (10) of Paragraph NB-3653 of the ASME Code,Section III exceeds 2.4 Sm (where Sm is the design stress intensity), and either Equation (12) or Equation (13) of Paragraph NB-3653.6, exceed 2.4 Sm.
- Intermediate locations where the cumulative usage factor as determined by the ASME Code exceeds 0.1.
- Efforts will be made to avoid intermediate break locations through appropriate piping layout and pipe support design.
The loading conditions considered for the stress range and usage factors calculated to determine break locations are those defined for Level A and B Service conditions for the piping system with the exception that seismic loads do not need to be considered for the postulation of intermediate break locations.
For those sections of pipe that satisfy the requirements for leak-before-break, leakage cracks are postulated for evaluation of subcompartment pressurization.]*
3.6.2.1.1.2 ASME Code,Section III - Class 2 and Class 3 Piping Systems
[For those piping system lines designed and analyzed to the requirements of the ASME Code,Section III, Class 2 and 3, except for those sections that satisfy the mechanistic pipe break criteria (subsection 3.6.3), the following criteria apply.
Pipe breaks are postulated to occur at terminal ends, using the same definition for terminal ends as for Class 1 pipe.
Pipe breaks are postulated at intermediate locations between terminal ends where the maximum stress value, as calculated by the sum of Equations (9) and (10) in Subarticle NC-3600 (Class 2) and ND-3600 (Class 3) of the ASME Code,Section III, considering Level A and B Service conditions. That is, breaks are postulated at locations for sustained loads, occasional loads, and thermal expansion exceeding 0.8 (1.8 Sh + SA) or 0.8 (1.5 Sy + SA), where Sh, SA, and Sy are the allowable stress at maximum hot temperature stress, allowable stress range for thermal expansion, and yield strength, respectively, for Class 2 and 3 piping, as defined in Subarticle NC-3600 and Subarticle ND-3600 of the ASME Code,Section III. Efforts will be made to avoid intermediate break locations through appropriate piping layout and pipe support design.
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR For those ASME Code,Section III, Class 2 and 3 systems that satisfy the leak-before-break criteria, postulated leakage crack locations are defined in the same way as for the Class 1 systems.]*
3.6.2.1.1.3 Piping Not Designed to ASME Code
[Breaks in piping systems designed to requirements other than the ASME Code, such as ASME B31.1 (Reference 3), are postulated at the following locations:
If the piping is analyzed and supported to withstand safe shutdown earthquake loadings, pipe ruptures are postulated to occur at the following locations:
- At terminal ends, using the same definition for terminal ends as for Class 1 pipe
- At intermediate locations where the stresses, as calculated by the sum of Equations (9) and (10) in Subarticle NC3600 of the ASME Code,Section III, considering normal and upset plant conditions, exceeds 0.8 (1.8 Sh + SA) or 0.8 (1.5 Sy + SA)
- Efforts will be made to avoid intermediate break locations through appropriate piping layout and pipe support design.]*
In the absence of stress analysis, breaks in non-nuclear piping are postulated at the following locations in each run or branch run:
- Terminal ends
- Intermediate fittings; (short- and long-radius elbows, crosses, flanges, nonstandard fittings, tees, reducers, welded attachments, and valves) 3.6.2.1.1.4 High-Energy Piping in Containment Penetration Areas The AP1000 does not have any ASME Code,Section III Class 1 pipe in containment penetration areas. Breaks are not postulated in the portions of ASME Code,Section III, Class 2 or Class 3 piping, defined below as break exclusion piping, provided subject piping meets the following provisions:
Stresses do not exceed those specified in Subsection 3.6.2.1.1.2.
The maximum stress in this piping as calculated by Equation (9), of paragraph NC-3653 of ASME Code Section III, when subjected to the combined loadings of internal pressure, deadweight, and postulated pipe rupture outside the break exclusion zone, does not exceed 2.25 Sh or 1.8 Sy.
The number of circumferential piping welds is minimized by using pipe bends in place of welding elbows when practicable. There are no longitudinal piping welds in the break exclusion zone. Where guard pipes are used, there are no circumferential or longitudinal welds in the piping enclosed within the guard pipe. Details of the arrangement are shown in Figure 3.8.2-4.
When required for isolation valve operability, structural integrity, or containment integrity, anchors or five-way restraints capable of resisting torsional and bending moments produced by a postulated pipe break, either upstream or downstream of the piping and valves which form the containment isolation boundary, are located reasonably close to the isolation valves or penetration.
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.6-13 Revision 9
VEGP 3&4 - UFSAR The anchors or five-way restraints do not prevent the access required to conduct in-service inspection examinations specified in Section XI of the ASME Code. In-service examinations completed during each inspection interval provide 100-percent volumetric examination (according to IWA-2400, ASME Code,Section XI) of circumferential pipe welds within the boundary of these portions of piping during each inspection interval. This volumetric inspection applies to circumferential piping welds that are equal to or greater than a 3-inch nominal diameter and piping branch connection welds that are equal to or greater than 4-inch nominal diameter.
Welded attachments to these portions of piping for pipe supports or other purposes are avoided. Where welded attachments are necessary, detailed stress analyses are performed to demonstrate compliance with the limits of Subsection 3.6.2.1.1 and applicable requirements of Section XI of the ASME Code.
The requirements of ASME Code,Section III, Subarticle NE-1120, are satisfied for the containment penetration.
Class 3 pipe satisfies the fabrication and inspection requirements for Section III, Class 2 pipe.
For evaluation of spray wetting, flooding, and subcompartment pressurization effects, longitudinal cracks (with crack flow areas of 1 square foot) are postulated in the main steam and main feedwater piping. The dynamic effects of pipe whip and jet impingement are not evaluated for these cracks. Locations having the greatest effect on essential equipment are chosen.
Guard pipe assemblies for high-energy piping in the containment annulus region between the containment shell and shield building that are part of the containment boundary are designed according to the rules of Class MC, subsection NE, of the ASME Code. The following requirements also apply. The design pressure and temperature are equal to or greater than the maximum operating pressure and temperature of the enclosed process pipe under normal plant conditions. Level C service limits of the ASME Code,Section III, Paragraph NE-3221, are not exceeded by the loadings associated with containment design pressure and temperature in combination with a safe shutdown earthquake. The guard pipe assemblies are subjected to a pressure test performed at the maximum operating pressure of the enclosed process pipe.
Areas of system piping where no breaks, except as noted in Subsections 3.6.1.2.1 and 3.6.1.2.2, are postulated are as follows:
The main steam piping from the containment penetration flued head inboard weld to the auxiliary building anchor downstream of the main steam isolation valves, including the main steam safety valves and the connecting branch piping The main feedwater piping from the auxiliary building side of the containment penetration flued head to the auxiliary building anchor upstream of the isolation valve The startup feedwater piping from the auxiliary building side of the containment penetration flued head to the auxiliary building anchor upstream of the isolation valve The steam generator blowdown piping from the auxiliary building side of the containment penetration flued head to the auxiliary building anchor downstream of the isolation valve The chemical and volume control system makeup piping from the containment penetration flued head to the outboard isolation valve 3.6-14 Revision 9
VEGP 3&4 - UFSAR The chemical and volume control system makeup piping from the containment penetration flued head to the inboard isolation valve For main steam line breaks in the Turbine Building, Wall 11 is capable of resisting torsional and bending moments produced by the postulated pipe breaks. When Wall 11.2 protects Wall 11 against the whipping effects of main steam line breaks, its piping sleeve helps in minimizing axial movement of Wall 11. The axial movement is due to pipe deflection at Wall 11.2, resulting from a main steam piping break in the Turbine Building north of Wall 11.2, and deflection of the main steam piping at Wall 11.2. Wall 11.2, along with its piping sleeve, acts as a barrier which helps in minimizing the lateral displacement of the main steam piping at Wall 11.2. This, in turn, reduces the axial piping displacements at Wall 11 due to pipe deflection at Wall 11.2, resulting in stresses in the break exclusion zone portion of the main steam piping that meet the criteria of Subsection 3.6.2.1.1.4.
Those portions of the containment penetration flued heads identified above that have the same nominal dimensions as the connected pipe are also considered as part of the break exclusion zone piping. The auxiliary building anchors also have flued head designs and the same requirement applies to these.
The main steam and main feedwater containment penetration flued heads are attached to expansion bellows, which are attached to the containment vessel via insert plates (Subsection 3.8.2.1.5, Figure 3.8.2-4, Sheet 1 and Sheet 8, respectively). The function of the expansion bellows is to minimize any piping loads applied to the containment vessel. The containment is not a terminal end for these piping analyses; the terminal ends are the main steam and main feedwater piping anchors in the auxiliary building exterior wall and their respective steam generator nozzles inside containment. The portion of the main steam piping that is inside containment is evaluated to meet the leak-before-break mechanistic pipe break criteria in accordance with Subsection 3.6.3; the portion of the main feedwater piping that is inside containment is analyzed to meet the high-energy pipe break criteria in accordance with Subsection 3.6.2.
All other fluid system containment penetrations are for moderate-energy systems or for pipe of 1-inch nominal diameter or smaller. See Subsection 6.2.3 for a discussion of containment penetrations.
3.6.2.1.2 Types of Breaks/Cracks Postulated 3.6.2.1.2.1 Break in Piping - High-Energy The following types of breaks are postulated to occur in ASME Code Class 1, 2, and 3 and non-ASME Code,Section III high-energy piping at the locations determined according to Subsection 3.6.2.1.1, except when the leak-before-break criteria are satisfied.
In piping with a nominal diameter of greater than or equal to 4 inches, both circumferential and longitudinal breaks are postulated at each selected break location unless eliminated by comparison of longitudinal and axial stresses with the maximum stress as follows:
- If the maximum stress range exceeds the limits specified in Subsections 3.6.2.1.1.1, 3.6.2.1.1.2, and 3.6.2.1.1.3, but the circumferential stress range is at least 1.5 times the axial stress range, only a longitudinal break is postulated.
- If the maximum stress range exceeds the limits specified in Subsections 3.6.2.1.1.1, 3.6.2.1.1.2, and 3.6.2.1.1.3, but the axial stress is at least 1.5 times the circumferential stress range, only a circumferential break is postulated.
- Longitudinal breaks, however, are not postulated at terminal ends.
3.6-15 Revision 9
VEGP 3&4 - UFSAR In piping with a nominal diameter of greater than 1 inch but less than 4 inches, only circumferential breaks are postulated at each selected break location.
No breaks are postulated for piping with a nominal diameter of 1 inch or less.
3.6.2.1.2.2 Through-Wall Cracks in High- or Moderate-Energy Piping Through-wall cracks are postulated in high-energy or moderate-energy piping, including branch runs larger than 1-inch nominal diameter as defined in the following paragraphs:
A. Through-wall cracks are not postulated in the break exclusion areas of high-energy pipe defined in Subsection 3.6.2.1.1.4 and in those portions of moderate-energy piping between containment isolation valves, provided the containment penetration meets the requirements of ASME Code,Section III, Sub-article NE-1120, and the piping is designed so that the maximum stress range based on the sum of equations (9) and (10) in Subarticle NC3600 of the ASME Code,Section III, considering Level A and B Service conditions, does not exceed either 0.4 (1.8 Sh + SA) or 0.4 (1.5 Sy + SA).
B. Through-wall cracks are not postulated in high- or moderate-energy fluid system piping located in an area where a break in the high-energy fluid system is postulated, provided that such cracks do not result in environmental conditions more limiting than the high-energy pipe break.
C. Subject to Paragraphs A and D, through-wall cracks are postulated in:
ASME Code,Section III, Division 1 - Class 1 piping where the maximum stress range as calculated by Equation (10) of Paragraph NB-3653 of the ASME Code,Section III exceeds 1.2 Sm. Cracks are also postulated at locations where the cumulative usage factor exceeds 0.1.
ASME Code,Section III, Division 1 - Class 2 or 3 piping at locations where the maximum stress range, as calculated by the sum of Equations (9) and (10) in Subarticle NC-3600 (Class 2) and ND-3600 (Class 3) of the ASME Code,Section III, considering Level A and B Service conditions, is greater than 0.4 (1.8 Sh + SA) or 0.4 (1.5 Sy + SA).
Seismically analyzed ASME B31.1 piping at locations defined in the same way as ASME Code,Section III, Class 3 piping.
Nonseismically analyzed ASME B31.1 piping at the following locations:
- Terminal ends
- Intermediate fittings; (short- and long-radius elbows, crosses, flanges, nonstandard fittings, tees, reducers, welded attachments, and valves)
D. Individual through-wall cracks are not postulated at specific locations determined by stress analyses when a review of the piping layout and plant arrangement drawings shows that the effects of through-wall leakage cracks at any location in the piping designed to seismic or nonseismic standards are isolated or are physically remote from structures, systems, and components required for safe shutdown.
E. Through-wall cracks are postulated to be in those circumferential locations that result in the most severe environmental consequences.
3.6-16 Revision 9
VEGP 3&4 - UFSAR 3.6.2.1.2.3 Leakage Cracks in High-Energy Piping with Leak-before-Break In those sections of piping that satisfy the requirements for leak-before-break, leakage cracks are postulated for evaluation of subcompartment pressurization. The size of the crack is such that the expected leakage is 10 times the minimum leak detection capability for that location. See Subsection 3.6.3 for a discussion of crack size and leakage detection.
3.6.2.1.3 Break and Crack Configuration 3.6.2.1.3.1 High-Energy Break Configuration Following a circumferential break, the two ends of the broken pipe are assumed to move clear of each other unless physically limited by piping restraints, structural members, or piping stiffness. The effective cross-sectional (inside diameter) flow area of the pipe is used in the jet discharge evaluation. Movement is assumed to be in the direction of the jet reaction initially with the total path controlled by the piping geometry.
The orientation of a longitudinal break, except when otherwise justified by a detailed stress analysis, is assumed to be at opposing points on a line perpendicular to the plane of a fitting for a non-axisymmetric fitting. The flow area of such a break is equal to the cross-sectional flow area of the pipe. The geometry of the longitudinal break may be assumed elliptical (2D along pipe axis and D/2 along pipe transverse) or circular. Both circumferential and longitudinal breaks are postulated to occur, but not concurrently, in high-energy piping systems at the locations specified in Subsection 3.6.2.1.2.1, except as follows:
Where the postulated break location is at a tee or elbow, the locations and types of breaks are determined as follows:
- Without the benefit of a detailed stress analysis, such as a finite element analysis, circumferential breaks are postulated to occur individually at each pipe-to-fitting weld.
Longitudinal breaks are postulated to occur individually (except in piping with a nominal diameter less than 4-inches) on each side of the fitting at its center and oriented perpendicular to the plane of the fitting, or
- Alternatively, if a detailed stress analysis or test is performed, the results may be used to predict the most probable rupture location(s) and type of break.
Where the postulated break location is at a branch/run connection, a circumferential break is postulated at the branch pipe-to-branch fitting weld unless otherwise justified by detailed analysis.
Where the postulated break location is at a welded attachment (lugs, stanchions), a circumferentially oriented break is postulated at the centerline of the welded attachment unless otherwise justified by a detailed analysis. The break area is equal to the pipe surface area that is bounded by the welded attachment.
Where the postulated break location is at a reducer, circumferential breaks are postulated at each pipe-to-fitting weld. Longitudinal breaks are oriented to produce out-of-plane bending of the piping configuration on both sides of the reducer at each pipe-to-fitting weld.
3.6.2.1.3.2 High-Energy and Moderate-Energy Through-Wall Crack Configuration High-and moderate-energy through-wall crack openings are assumed to be a circular orifice with cross-sectional flow area equal to that of a rectangle one-half the pipe inside diameter in length and 3.6-17 Revision 9
VEGP 3&4 - UFSAR one-half pipe wall thickness in width. The flow from a through-wall crack is assumed to result in an environment that wets unprotected components within the compartment with consequent flooding in the compartment and communicating compartments, unless analysis shows otherwise. Flooding effects are determined on the basis of a conservatively estimated time period required to take corrective actions.
3.6.2.2 Analytical Methods to Define Jet Thrust Forcing Functions and Response Models To determine the forcing function, the fluid conditions at the upstream source and at the break exit dictate the analytical approach and approximations that are used.
Analytical methods for calculation of jet thrust for the preceding situations are discussed in ANS-58.2-1988 (Reference 4) and Moody, F. J. (Reference 5). The discussion of the jet thrust forcing functions on the reactor coolant loop follows.
Since a rupture of the large-diameter reactor coolant loop piping does not have to be considered, based on satisfying mechanistic pipe break criteria, the jet thrust and reactive loads considered in the analysis are those associated with breaks in branch line sections that do not satisfy the mechanistic pipe break criteria.
To determine the thrust and reactive force loads to be applied to the reactor coolant loop during the postulated pipe rupture, it is necessary to have a detailed description of the hydraulic transient.
Hydraulic forcing functions are calculated for the reactor coolant loops as a result of a postulated loss of coolant accident. These forces result from the transient flow and pressure histories in the reactor coolant system (RCS).
The calculation is performed in two steps. The first step is to calculate the transient pressure, mass flowrates, and thermodynamic properties as a function of time. The second step uses the results obtained from the hydraulic analysis, along with input of areas and direction coordinates, and calculates the time-history of forces at appropriate locations in the reactor coolant loops.
The hydraulic model represents the behavior of the coolant fluid within the entire reactor coolant system. Key parameters calculated by the hydraulic model are pressure, mass flowrate, and density.
These are supplied to the thrust calculation, together with plant layout information, to determine the time-dependent loads exerted by the fluid on the loops. In evaluating the hydraulic forcing functions during a postulated loss of coolant accident, the pressure and momentum flux terms are dominant.
The inertia and gravitational terms are taken into account in the evaluation of the local fluid conditions in the hydraulic model.
The blowdown hydraulic analysis provides the basic information concerning the dynamic behavior of the reactor core environment for the loop forces. This requires the ability to predict the flow, quality, and pressure of the fluid throughout the reactor system. [MULTIFLEX (Reference 6) or an equivalent computer code is used to provide this information.]*
MULTIFLEX calculates the hydraulic transients within the entire primary coolant system. This hydraulic program considers a coupled, fluid-structure interaction by accounting for the deflection of the core support barrel. The depressurization of the system is calculated using the method of characteristics applicable to transient flow of a homogenous fluid in thermal equilibrium.
The ability to treat multiple flow branches and a large number of mesh points gives MULTIFLEX the flexibility to represent the various flow passages within the primary reactor coolant system. The system geometry is represented by a network of one-dimensional flow passages.
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.6-18 Revision 9
[The THRUST computer program or equivalent is used to compute the transient (blowdown) hydraulic loads resulting from a loss of coolant accident.]*
The blowdown hydraulic loads on primary loop components are computed from the equation:
§ m2 *º F = 144 A << (P - 14.7) + ¨¨ ¸>>
2¸
<<¬ © 144 U g (A m ) ¹ >>1/4 where:
F = Force (lbf)
A = Aperture area (ft2)
P = System pressure (psia)
= Mass flowrate (lbm/s)
= Density (lbm/ft3) g = Gravitational constant = 32.174 ft-lbm/lbf - s2 Am = Mass flow area (ft2)
In the model to compute forcing functions, the reactor coolant loop system is represented by a model similar to that employed in the blowdown analysis. The entire loop layout is represented in a global coordinate system. Each node is described by blowdown hydraulic information and the orientation of the streamline of the force nodes in the system, which includes flow areas and projection coefficients along the three axes of the global coordinate system.
Each node is modeled as a separate control volume with one or two flow apertures associated with it.
Two apertures are used to simulate a change in flow direction and area.
Each force is divided into its x, y and z components using the projection coefficients. The force components are then summed over the total number of apertures in any one node to give a total x force, a total y force, and a total z force. These thrust forces serve as input to the piping/
restraint dynamic analysis.
[The THRUST code calculates forces the same way as the STHRUST code described in WCAP-8252 (Reference 7).]*
3.6.2.3 Dynamic Analysis Methods to Verify Integrity and Operability This subsection describes the pipe rupture design criteria for auxiliary piping systems.
Subsection 3.6.2.2 describes the analysis methods for thrust loadings. To mitigate each postulated pipe rupture, auxiliary piping systems required to maintain pressure boundary integrity or to provide for fluid flow are identified. The loadings on these systems may consist of jet impingement loads, transient motions at terminal end connections, or internal system depressurization loadings.
The application of leak-before-break analysis eliminates evaluation of postulated pipe ruptures in the primary coolant loop piping and selected piping systems of 6-inch nominal size or larger. The piping
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.6-19 Revision 9
VEGP 3&4 - UFSAR system mechanical components and supports are designed for the effects of the remaining postulated pipe ruptures and leaks.
To confirm the continued integrity of the essential components and the engineered safety systems, consideration is given to the consequential effects of the pipe break to the extent that:
The minimum performance capabilities of the engineered safety systems are not reduced below that required to protect against the postulated break.
The containment leaktightness is not decreased below the design value if the break leads to a loss of coolant accident.
Propagation of damage is limited in type or degree or both to the extent that:
- A pipe break that is not a loss of coolant accident, steam line break, or main feedwater break will not cause a loss of coolant accident or steam line or feedwater line break.
- A break in the nonsafety portion of the chemical and volume control system purification loop will not cause a break in the safety-related portion of the system. In addition, the ability to isolate the reactor coolant system flow will not be adversely affected.
- A reactor coolant system pipe break will not cause a steam or feedwater system pipe break, and vice versa.
3.6.2.3.1 Jet Impingement Analytical methods for the calculation of jet impingement forces are based on Moody, F. J.
(Reference 5), NUREG/CR-2913 (Reference 1), and Section 7.3 of ANS-58.2-1988 (Reference 4).
For piping systems this loading is a suddenly applied load that can have significant energy content.
These loads are generally treated as statically applied constant loads.
Two separate structural evaluations are performed. For the short-term response, snubber supports are considered to be active and a dynamic load factor of 2 is used. For the longer-term response, snubber supports are considered inactive, and no dynamic load factor is used.
If simplified static analysis is performed instead of a dynamic analysis, the preceding jet load (FT) is multiplied by a dynamic load factor. For an equivalent static analysis of the target structure, the jet impingement force is multiplied by a dynamic load factor of 1.2 to 2.0, depending upon the time variance of the jet load and the elastic/plastic behavior of the target. This factor assumes that the target can be represented as essentially a one-degree-of-freedom system.
3.6.2.3.2 Transient Motions at Terminal Ends This loading is displacement limited and has a short duration of about 0.5 seconds. An example is the motions of the primary loop piping at the terminal end connection of the Class 1 pressurizer surge line piping due to a postulated pipe rupture in a Class 2 pipe connected to the steam generator.
When there are active in-line components in the piping system that must function to mitigate the postulated pipe rupture, dynamic structural analyses are performed for the terminal end motions. The calculated accelerations are evaluated to confirm the operability of the active in-line components. For piping systems with no active in-line components, static structural analyses with no dynamic amplification are performed for the terminal end motions.
These analyses may consider nonlinear geometric and material characteristics of the piping system.
3.6-20 Revision 9
VEGP 3&4 - UFSAR 3.6.2.3.3 Internal System Depressurization This loading has a short duration of approximately 0.5 seconds and arises from rapidly traveling pressure waves in piping systems connected to the broken piping system. Two types of configurations are possible: systems without check valves and systems with check valves. In systems with check valves, the valve closure can increase the duration and magnitude of these loads.
An example of the former is the pressure waves in the Class 1 letdown line of the chemical and volume control system piping due to a postulated pipe rupture in a Class 1 pipe connected to the primary loop piping. An example of the latter is the closure of the feedwater check valve due to a postulated pipe rupture upstream of the valve.
For piping systems without closing check valves, there is little energy in the high-frequency depressurization loadings. These loadings are therefore not considered in the piping and support analysis.
For piping system with closing check valves, the magnitude of the loadings depends on the valve closure time, with shorter closing times generally causing higher loadings. For this loading the potential system failure mechanisms evaluated are: 1) excessive pipe and valve hoop stress;
- 2) tensile loads on the valve pressure boundary bolting; and 3) excessive distortion of the valve disc or seat.
The maximum internal pressure and the kinetic energy of the valve disc at the time of closure are used to verify the pressure boundary integrity of the piping and valve based on the preceding failure mechanisms. RELAP5 is used to calculate the pressure and kinetic energy. The supports on these systems are designed in such a way that support failure will occur prior to local pipe pressure boundary failure at the support connection.
3.6.2.3.4 Pipe Whip Restraints To satisfy varying requirements of available space, permissible pipe deflection, and equipment operability, the restraints are generally designed as a combination of an energy-absorbing element and a restraint structure suitable for the geometry required to pass the restraint load from the whipping pipe to the main building structure. The restraint structure is typically a structural steel frame or truss, and the energy-absorbing element is usually either stainless steel U-bars or energy-absorbing material.
Another option is for the restraints to be designed as entirely elastic structures and simply transfer the pipe break loads to the main building structure. In addition, pipe supports (including snubbers, struts, or rigid frames) can be used as pipe whip restraints when qualified for the pipe break loads.
3.6.2.3.4.1 Location of Pipe Whip Restraints For purposes of determining pipe hinge length and thus locating the pipe whip restraints, the plastic moment of the pipe is determined in the following manner:
Mp = 1.1 zpSy where:
zp = Plastic section modulus of pipe Sy = Yield stress at pipe operating temperature 3.6-21 Revision 9
VEGP 3&4 - UFSAR 1.1 = 10-percent factor to account for strain hardening.
Pipe whip restraints are located as close to the axis of the reaction thrust force as practicable. Pipe whip restraints are generally located so that a plastic hinge does not form in the pipe. If, because of physical limitations, pipe whip restraints are located so that a plastic hinge can form, the consequences of the whipping pipe and the jet impingement effect are further investigated. Lateral guides are provided where necessary to predict and control pipe motion.
Generally, pipe whip restraints are designed and located with sufficient clearances between the pipe and the restraint in such a way that they do not interact and cause additional piping stresses. A design hot position gap is provided that allows maximum predicted thermal, seismic, and seismic anchor movement displacements to occur without interaction.
Exception to this general criterion may occur when a pipe support and restraint are incorporated into the same structural steel frame, or when a zero design gap is required. In these cases, the pipe whip restraint is included in the piping analysis and designed to the requirements of pipe support structures for all loads except pipe break and designed to the requirements of pipe whip restraints when pipe break loads are included.
In general, the pipe whip restraints do not prevent the access required to conduct in-service inspection examination of piping welds. When the location of the restraint makes the piping welds inaccessible for in-service inspection, a portion of the restraint is designed to be removable to provide accessibility.
3.6.2.3.4.2 Analysis and Design of Pipe Whip Restraints The criteria for analysis and design of pipe whip restraints for postulated pipe break effects are provided in the following. These criteria are consistent with the guidelines in ANS-58.2-1988 (Reference 4).
Pipe whip restraints are designed based on energy absorption principles by considering the elastic-plastic, strain-hardening behavior of the materials used.
Non-energy absorbing portions of the pipe whip restraints are designed to the requirements of ANSI/AISC N690 Code supplemented by the requirements given in Subsection 3.8.4.5.
American Welding Society (AWS), Structural Welding Code-Steel, AWS D1.1-2000 provides an acceptable alternative for ANSI/AISC N690 weld requirements as described in Subsections 3.8.3.2 and 3.8.4.2.
Standard pipe support components, such as struts and snubbers, used as pipe whip restraints are designed to ASME Section III, Subsection NF, Service Level D allowables.
Loads on pipe whip restraints are developed either from an energy balance method (as described in ANS-58.2) with a 1.1 rebound factor or from dynamic time-history analyses.
Except in cases where calculations are performed to verify that a plastic hinge is formed, the energy absorbed by the ruptured pipe is conservatively assumed to be zero. That is, the thrust force developed goes directly into moving the broken pipe and is not reduced by the force required to bend the pipe.
Other structural members of the pipe whip restraints are designed for elastic response. A dynamic increase factor is used for those members that are designed to remain elastic.
3.6-22 Revision 9
VEGP 3&4 - UFSAR The criteria for allowable strain in a pipe whip restraint are dependent on the type of restraint.
The following discussions address the types of restraints used and the allowable strain for each. Note - = allowable strain used in design.
Energy Absorbing Materials Stainless Steel U-Bar - This type of restraint consists of one or more U-shaped, upset-threaded rods of stainless steel looped around the pipe but not in contact with the pipe. This allows unimpeded pipe motion during seismic and thermal movement of the pipe. At rupture, the pipe moves against the U-bars, which absorb the kinetic energy of pipe motion by yielding plastically. Figure 3.6-1 shows a typical example of a U-bar restraint.
= 0.5u where:
u = ultimate uniform strain of stainless steel (strain at ultimate stress)
Crushable Pipe - The crushable pipe type of restraint consists of a short section of standard pipe (length at least three times diameter of crushable pipe) that is situated approximately perpendicular to the process pipe and whose longitudinal axis is approximately perpendicular to the plane of pipe movement. The crushable pipe is attached to a supporting structure. The crushable pipe is the part of the assembly that absorbs the kinetic energy that the broken pipe has accumulated after rupture, and the crushable pipe typically undergoes considerable plastic deformation during the pipe whip event.
A design hot position gap is provided between the process pipe and the energy-absorbing crushable pipe to allow unimpeded pipe motion during seismic and thermal pipe movements. Figure 3.6-2 shows typical examples of a crushable pipe type whip restraint. The allowable capacity of the crushable pipe is limited to 80 percent of its rated energy dissipating capacity as determined by dynamic testing, at loading rates within +/- 50 percent of the specified design loading rate. An additional conservative measure is used to maintain ductile yield of the energy absorbing pipe by limiting flattening of the crushed pipe to the limits provided in ASTM A530. The rated energy dissipating capacity is not greater than the area under the load-deflection curve as illustrated in Figure 3.6.2-1 of NUREG-0800, Standard Review Plan, Section 3.6.2, Revision 2.
3.6.2.4 Protective Assembly Design Criteria In addition to pipe whip restraints, other protective devices are designed to protect against the effects of postulated pipe ruptures. Barriers and shields are designed to protect against jet impingement.
Guard pipes in the break exclusion zones provide additional confidence that pipes will not leak into the annulus between the containment vessel and the shield building.
3.6.2.4.1 Jet Impingement Barriers and Shields Barriers and shields, constructed of either steel or concrete, are provided to protect essential equipment, including instrumentation, from the effects of jet impingement resulting from postulated pipe breaks. Barriers differ from shields in that they may also accept the impact of whipping pipes.
Barriers and shields include walls, floors, and structures specifically designed to provide protection from postulated pipe breaks. Barrier and shield design is based on elastic methods and the elastic-plastic methods for dynamic analysis included in Biggs, J. M. (Reference 9). Design criteria and loading combinations are according to Subsections 3.8.3 and 3.8.4.
3.6.2.4.2 Auxiliary Guardpipes The use of guard pipes has been minimized by plant arrangement and routing of high-energy piping.
Guard pipes in the containment annulus areas of the break exclusion zones are designed as 3.6-23 Revision 9
VEGP 3&4 - UFSAR described in Subsection 3.6.2.1.1.4. Other guard pipes are designed and constructed to the same ASME rules as the enclosed process pipe.
3.6.2.5 Evaluation of Dynamic Effects of Pipe Ruptures The preceding information provides the criteria and methods for the evaluation of the dynamic effects of pipe ruptures. The pipe rupture hazard analysis report (also referred to as the pipe break evaluation report) includes the following:
Prepare a stress summary Identify pipe break locations in high energy piping Identify through-wall crack locations in high and moderate energy piping Identify and locate essential structures, systems, and components Evaluate consequences of pipe whip and jet impingement For rooms with both high energy breaks and essential items, confirm that there is no adverse interaction between the essential items and the whipping pipe or jet.
The plant layout is modified as required to provide separation to protect essential systems.
Evaluate consequences of flooding, environment, and compartment pressurization Evaluate compartment pressurization in the break exclusion zones in the vicinity of containment penetrations due to 1.0 square foot breaks in the main steam and feedwater lines.
Design and locate protective hardware Prepare isometric piping sketches that identify the break locations, the basis for these locations and the protective hardware which mitigates the consequences of these breaks.
Reconciliation of as-built condition Pipe breaks that are larger than 1-inch nominal diameter are evaluated for pipe whip and jet impingement. Lines that are located in a break exclusion zone or are qualified to leak-before-break are not evaluated for pipe whip and jet impingement effects on systems and components, except for the portions of the lines in the MSIV compartment adjacent to the main control room as noted in Subsection 3.6.1.2.2.
Where these systems are qualified for mechanistic pipe break and pipe rupture loads prior to fabrication, the qualification is based on design information, not on as-built information. As-built information and the final configuration of valves and other equipment is used to verify the design analysis.
High Energy Break Locations High energy break locations evaluated are on the nuclear island and in the turbine building for evaluation of the wall loadings in the south end of the turbine building adjacent to the main control room.
For ASME Class 1 piping terminal end locations are determined from the piping isometric drawings.
Intermediate break locations depend on the ASME Code stress report fatigue analysis results. These 3.6-24 Revision 9
VEGP 3&4 - UFSAR results were not available at design certification. For the design of the AP1000, breaks are postulated at locations typically associated with a high cumulative fatigue usage factor. These locations are at valves, tees, and branch connections which have significant structural discontinuities. These locations are part of the as-built reconciliation as discussed in Subsection 3.6.4.1. The following ASME Class 1 lines are evaluated to terminal end and intermediate high energy break locations if applicable.
Line Diameter (inches)
Pressurizer Spray 4 Automatic Depressurization Stage 1 4 Chemical and Volume Control Letdown 3 Chemical and Volume Control Makeup 3 Reactor Coolant Pump Piping to/from Heat Exchangers (2/pump) 3 Pressurizer Auxiliary Spray 2 For ASME Class 2 and 3 piping, terminal end break locations are determined from the piping isometric drawings. The intermediate break locations depend on the stress level. The AP1000 ASME Class 2 and 3 lines do not have intermediate breaks based on the low stress. The following ASME Class 2 and 3 lines have terminal end high energy break locations.
Line Diameter (inches)
Main Feedwater 16, 20 Startup Feedwater 6 Steam Generator Blowdown 4 For ASME B31.1 piping, terminal end break locations are determined from the piping isometric drawings. The intermediate break locations in seismically analyzed pipe depend on the stress level. The AP1000 ASME seismically analyzed ASME B31.1 piping does not have intermediate breaks based on the low stress. For nonseismically analyzed high-energy ASME B31.1, intermediate breaks locations are postulated at each fitting.
Rooms subject to pressurization due to high energy pipe break are listed in Table 3.6-2 with the break location.
Essential Systems and Components In rooms that contain high energy pipe breaks, the systems and components that are needed to mitigate the postulated break and achieve a safe plant shutdown are identified. Rooms that contain both high energy pipe break locations and essential systems or components that must be protected are listed in Table 3.6-3. No high energy pipe break protection is required in other areas of the plant.
Essential Target Evaluation To complete the essential target evaluation jet parameters, volumetric area of affected compartments, plant layout, and separating structures are considered. Parameters that determine the shape of the jet and the magnitude of the jet and thrust loads include pressure, temperature, and friction losses between the break and the reservoir. The volumetric area affected is determined by considering jet shape and loads at the postulated location of the breaks. Where an initial evaluation of essential targets indicated adverse effects, layout may be changed to relocate the target or 3.6-25 Revision 9
VEGP 3&4 - UFSAR postulated break. If necessary, the location of whip restraints and jet shields is established to protect essential systems and components. Essential equipment protected by pipe whip restraints or jet shields is listed in Table 3.6-3. The criteria for the break location postulated for evaluation of separating structures is outlined in Subsection 3.6.1.3.2.
Verification of the Pipe Break Hazard Analysis A pipe rupture hazard analysis is prepared based on the as-designed piping stress analyses and pipe whip restraint design information. The as-designed piping analysis is based on piping routings, layouts, and isometrics. Intermediate break locations are identified using the as-designed piping stress analysis, including the fatigue analysis required for ASME Code Class 1 piping. As-designed piping stress analysis information is used to confirm the location and configuration of pipe whip restraints and jet impingement shields. The information included in Tables 3.6-2 and 3.6-3 is updated and validated as part of the as-designed pipe rupture hazard analysis. Large leakage cracks in moderate energy pipes are evaluated for adverse effects as part of the pipe break hazard evaluation.
The ASME Code,Section III, requires that each plant have a Design Report for the piping system that includes as-built information. Included in the Design Reports are the loads and loading combinations used in the analysis. Where mechanistic pipe break requirements are used to eliminate the evaluation of dynamic effects of pipe rupture in ASME Code,Section III, Class 1, 2, and 3 piping system, the basis for the exclusion is documented in the Design Report.
The final piping stress analyses, pipe whip restraint design, and as-built reconciliation of the pipe break hazard analysis is discussed in Subsection 3.6.4.1. The final piping stress analysis includes design properties and characteristics of procured components selected to be included in the piping system that are not available for the as-designed evaluation. The as-built reconciliation is required prior to fuel loading and includes evaluation of the ASME Code fatigue analysis, pipe break dynamic loads, reconciliation to the certified design floor response spectra, confirmation of the reactor coolant loop time history seismic analyses, changes in support locations, preoperational testing, and construction deviations.
3.6.2.6 Evaluation of Flooding Effects from Pipe Failures The effect of flooding due to high and moderate energy pipe failures on essential systems and components is described in Section 3.4.
3.6.2.7 Evaluation of Spray Effects from High- and Moderate-Energy Through-Wall Cracks Essential systems and components are evaluated for the potential effects of spray from high- and moderate-energy through-wall cracks. Spray effects are assumed to be limited to the compartment where the pipe failure occurs. The spray is assumed to wet unprotected components in the compartment. It is further assumed the spray does not damage non-electrical passive components, including piping, ducts, valve bodies, or mechanical components of valve operators. Spray may cause failure of electrical components not designed to withstand wetting. Components protected by NEMA 4 or NEMA 12 enclosures are not affected by spray effects.
The safe shutdown components inside containment are subject to wetting from design basis events inside containment. These conditions bound the effects of spray from moderate energy cracks.
Sensitive components are qualified for this environment as described in Section 3.11.
The doors to the auxiliary Class 1E battery rooms are normally closed, so spray cannot affect the batteries if fire fighting activities or a pipe crack were to occur in the corridor. If fire fighting activities were to occur in a particular room, all of the equipment is assumed inoperable due to the fire, therefore, no further spray effects need be considered. The containment isolation valves subject to 3.6-26 Revision 9
VEGP 3&4 - UFSAR spray and the safe shutdown components in the main steam tunnels are evaluated for the consequence of spray; and, if essential functions are compromised, are provided with spray protection. The sensitive components of the main control room emergency habitability system are protected from spray effects.
3.6.3 Leak-before-Break Evaluation Procedures This subsection describes the design basis for mechanistic pipe break (leak-before-break) evaluation of high-energy piping systems.
Mechanistic pipe break evaluations demonstrate that for piping lines meeting the criteria, sudden catastrophic failure of the pipe is not credible. It is demonstrated that piping that satisfies the criteria leaks at a detectable rate from postulated flaws prior to growth of the flaw to a size that would fail because of applied loads resulting from normal conditions, anticipated transients, and a postulated safe shutdown earthquake.
The use of mechanistic pipe break criteria represents a higher level of confidence of the integrity of piping systems based on additional criteria compared to the existing high level of integrity provided by the requirements of the ASME Code. Evaluations of the mechanistic pipe break criteria are commonly called leak-before-break evaluations.
The use of mechanistic pipe break criteria permits the elimination of the evaluation of dynamic effects of sudden circumferential and longitudinal pipe breaks in the design basis analysis of structures, systems, and components. General Design Criterion 4 of Appendix A, 10 CFR Part 50 allows the use of analyses to eliminate from the design basis the dynamic effects of pipe ruptures.
Without the application of mechanistic pipe break criteria, the dynamic effects are evaluated for pipe ruptures postulated at locations defined in Subsection 3.6.2. Dynamic effects include jet impingement, pipe whip, jet reaction forces on other portions of the piping and components, subcompartment pressurization including reactor cavity asymmetric pressurization transients, pump overspeed and traveling pressure waves from the depressurization of the system.
Incorporating leak-before-break criteria and guidelines into the design process maximizes the benefits of applying mechanistic pipe break. Eliminating the dynamic effects permits minimizing the size and number of protective structures and eliminates the use of pipe whip restraints. This permits design optimization and avoids obstruction of pipe welds for in-service inspection by protective structures and restraints.
High-energy ASME Code Section III piping that is evaluated to the leak-before-break criteria is identified in Appendix 3E. This applies to the main steam piping as follows. The main steam piping from the steam generator outlet nozzle to the anchor downstream of the isolation valve is analyzed for applicable loadings including the safe shutdown earthquake. This anchor is at the exterior wall of the auxiliary building. The portion of this piping from the containment penetration flued head inboard weld to the above anchor satisfies the break exclusion zone requirements described in Subsection 3.6.2. The portion of this piping from the steam generator outlet nozzle to flued head inboard weld, including the welds, is evaluated to the leak-before-break criteria. The portion of the auxiliary building flued head (anchor in the wall) that has the same nominal dimensions as the main steam pipe is also classified as a break exclusion zone. High-energy piping that does not satisfy the leak-before-break criteria is designed to the requirements discussed in Subsections 3.6.1 and 3.6.2.
The piping to which mechanistic pipe break is applied is analyzed to demonstrate that the piping has leak-before-break characteristics. The leak-before-break analysis is either a fracture-mechanics based stability analysis or a plastic-instability limit load analysis as appropriate. The analysis combines normal and abnormal (including seismic) loads to determine a critical crack size for a 3.6-27 Revision 9
VEGP 3&4 - UFSAR postulated through-wall crack. The critical crack size is compared to the size of a leakage crack for which, with appropriate margin, detection is certain. When the critical crack size is sufficiently larger than the leakage crack size the leak-before-break requirements are satisfied.
Mechanistic pipe break is not used for purposes of specifying non-structural design criteria for emergency core cooling, containment systems, or other non-structural engineered safety features, or for the evaluation of environmental effects including spray wetting, humidity, temperature, pressure, radiation, and adverse reactions with chemicals in the coolant. This includes piping for which leak-before-break is demonstrated.
A bounding analysis is performed for each piping system. The bounding analysis is applied as discussed in Subsection 3.6.4.2 to verify that the as-built piping satisfies the requirements for leak-before-break.
3.6.3.1 Application of Mechanistic Pipe Break Criteria Piping systems to which mechanistic pipe break are applied are high integrity systems with well understood loading combinations and conditions. The piping systems to which it is applied satisfy the requirements of the ASME Code,Section III. ASME Code requirements also apply to the pre-service and in-service inspection which confirm continued integrity.
The mechanistic pipe break approach is applicable to high-energy piping provided plant design, operating experience, tests, or analyses have indicated low probability of failure from effects of intergranular stress corrosion cracking, water hammer, steam hammer, fatigue (thermal or mechanical), or erosion.
The plant design and operating features permit the application of the mechanistic pipe break approach. The piping to which the leak-before-break criteria is applied is evaluated for fatigue due to cyclic loads as required by the appropriate requirements of the ASME Code.
The piping in the AP1000 does not operate at temperatures for which creep or creep fatigue must be considered.
The reactor coolant loop piping, branch lines, and other lines in contact with reactor coolant are fabricated of austenitic stainless steel, which is very resistant to erosion and corrosion in typical reactor coolant chemistries and flowrates. Intergranular stress corrosion cracking has not been associated with reactor coolant piping in pressurized water reactors.
The design of the reactor coolant loop is not conducive to the generation of water hammer loads. The reactor coolant loop does not have any valves that could result in a water hammer due to rapid valve closure. The steam bubble in the pressurizer is not subject to the introduction of a large volume of cold water sufficient to result in a bubble collapse water hammer.
The design and component selection of reactor coolant branch lines and other lines evaluated for mechanistic pipe break follow design guidelines intended to minimize the potential for water hammer.
Comparison of the AP1000 piping to the screening criteria in Subsection 5.29 of NUREG/CR-6519 (Reference 13) demonstrates that there is not a significant potential for water hammer in the leak-before-break piping.
Thermal stratification of water in stagnant or slowly flowing lines can result in thermal fatigue in a pipe. The piping and system design requirements for AP1000 address the potential for thermal stratification. For additional information of thermal stratification, see Subsections 3.9.3, 5.4.3, and 5.4.5.
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VEGP 3&4 - UFSAR The composition of the main steam lines has been selected to minimize the potential for erosion and corrosion. The main steam lines are fabricated from SA335 Grade P11 Alloy steel, which is composed of sufficient levels of chromium to preclude erosion and corrosion mechanisms. The main steam lines are not subject to water hammer or thermal stratification by the nature of the fluid transported.
The steam line is protected from being filled with water due to steam generator overfill by implementation of operating instructions or isolation requirements included in the protection system logic or both. See Section 7.3 for information on the protection system design to prevent overfill.
In addition to requirements on the design, fabrication, and inspection of the piping systems, the application of mechanistic pipe break requires a qualified leak detection capability. Leak detection systems inside containment meet the guidelines of Regulatory Guide 1.45. See Subsection 5.2.5 for a discussion of the leak detection system for the reactor coolant system and connected piping.
3.6.3.2 Design Criteria for Leak-before-Break The methods and criteria to evaluate leak-before-break in the AP1000 are consistent with the guidance in NUREG-1061 (Reference 11) and Draft Standard Review Plan 3.6.3 (Reference 12)
The application of the mechanistic pipe break in AP1000 requires that the following design requirements are met.
Pre-service inspection of welds is required.
For ASME Code Class 1, Class 2, and Class 3 systems for which leak-before break is demonstrated, the ASME Code,Section III and Section XI preservice and inservice inspection requirements will provide for the integrity of each system. The weld and welder qualification, and weld inspection requirements for ASME Code,Section III, Class 3 leak-before-break lines are equivalent to the requirements for Class 2. The inservice inspection requirement for each Class 3 leak-before-break line includes a volumetric inspection equivalent to the requirements for Class 2 for the weld at or closest to the high stress location.
Inservice inspection and testing of snubbers (if used) are performed to provide for a low snubber failure rate.
For the maximum stress due to steady-state vibration refer to Subsection 3.9.2.
The leak-before-break bounding analysis curves are developed for each applicable piping system. The bounding analysis methods are described in Appendix 3B. These curves give the design guidance to satisfy the stress limits and leak-before-break acceptance criteria.
The highest stressed point (critical location) determined from the piping stress analysis is compared to the bounding analysis curve and has to fall on or under the curve. The points on or under the bounding analysis curve satisfy the requirements for leak-before-break.
The analyzed normal stress and maximum stress are not required to construct the bounding analysis curve. The analyzed stresses are calculated by the equation; Fx M
= +
A Z 3.6-29 Revision 9
is the stress Fx is the axial force M is the applied moment A is the piping cross-sectional area Z is the piping section modulus.
The normal stress is calculated by the algebraic summation of load combination method and the maximum stress is calculated by the absolute summation of load combination method.
The corrosion-resistant piping materials, including base metal and welds, have an appropriate toughness. The piping materials containing primary coolant are wrought stainless steel. The welds in stainless steel pipe are made using the gas tungsten arc (GTAW) process. These materials are very resistant to crack extension. The tensile properties for the leak-before-break evaluation are those found in the Section II Appendices of the ASME Code. During the design stage, the material properties used are based on the ASME Code minimum values. During the as-built reconciliation stage, certified material test report values are reviewed to verify that ASME Code requirements are satisfied.
For those lines fabricated using non-stainless ferritic materials, the materials used and the associated welds have adequate toughness to demonstrate that leak-before-break criteria are satisfied. The welds are made using the gas tungsten arc (GTAW) process. The tensile properties for the leak-before-break evaluation are obtained from actual material tests. During the design stage, the material properties are based on test results. During the as-built reconciliation stage, certified material test report values are reviewed to verify that the toughness and strength requirements of the ASME Code,Section III are satisfied.
Potential degradation by erosion, erosion/corrosion and erosion cavitation is examined to provide low probability of pipe failure.
Wall thicknesses in elbows and other fittings are evaluated to confirm that ASME Code,Section III piping requirements are met as a minimum.
The as-built condition of the piping and support system is evaluated based on the guidelines in EPRI NP-5630 (Reference 10) and reconciled to the analysis of the leak-before-break criteria based on the design information. The locations and characteristics of the supports, including any gaps between the supports and piping, or other configurations that result in a nonlinear response are included in the as-built evaluation.
Adjacent structures and components are designed for the safe shutdown earthquake event to provide low probability of indirect pipe failure.
The piping supports are anchored to reinforced concrete structures, to concrete-filled steel plate structures, or to steel structures anchored to these types of structures. Piping is not supported by masonry block walls.
3.6-30 Revision 9
VEGP 3&4 - UFSAR 3.6.3.3 Analysis Methods and Criteria The methods used to develop the bounding analysis curves are described in Appendix 3B.
Development of the bounding analysis curves provides an evaluation method that is consistent with NRC requirements and guidance. The calculation method and computer codes used for AP1000 are benchmarked to test data and have been previously accepted by the NRC for leak-before-break evaluations in operating nuclear power plants.
Analyzable sections run from one terminal end or anchor to another terminal end or anchor. A terminal end is typically a connection to a larger pipe or a component. For the structural analysis, a normally closed valve between pressurized and unpressurized portions of a line is not considered a terminal end. Figure 3.6-3 is a schematic of a portion of a piping system that illustrates the meaning of analyzable segments. In the figure the analyzable portion of the pipe runs from point A to point D.
The leak-before-break evaluation is based on a fracture mechanics stability analysis comparing the selected leakage crack to the critical crack size. The following discussion outlines the analysis method.
The development of leak-before-break bounding analysis curves assume that circumferentially oriented postulated cracks are limiting. Stability is established by analyzing through-wall flaws.
Leakage Flaw Through-wall flaws in candidate leak-before break piping systems are postulated. [The size of the postulated flaws are large enough so that the leakage is detectable with adequate margin, using 10 times the minimum installed leak detection capability when the pipes are subjected to normal operational loads combining by algebraic sum method.]* That is, the size of the leakage flaw postulated would be expected to have a leak rate 10 times the size of the rated leak rate detection capability.
As noted in Subsection 5.2.5, the rated capability of the leak detection systems for the primary coolant inside containment is 0.5 gpm. The methods used to detect leakage are described in Subsection 5.2.5.3. The methods used for primary coolant are the containment sump level, inventory balance, and containment atmosphere radiation. The method used to detect leakage from the main steam line inside containment is the containment sump level. Containment water level sensors provide a diverse backup for main steam line leakage by detecting a 0.5 gpm leak within 17 days.
Containment air cooler condensate flow, and containment atmosphere pressure, temperature, and humidity also provide an indication of possible leakage.
Stability and Critical Flaw Sizes The local and global failure mechanisms are evaluated, as appropriate, to provide margin on flaw size and load. The local mode of failure addresses crack tip behavior: blunting, initiation, extension, and instability. The local failure mechanism is evaluated for ferritic steel piping systems using the J-integral method. The global mode of failure addresses the behavior of the net section: initial yielding, strain hardening, and plastic hinge formation. The global failure mechanism (limit load method) is evaluated for stainless steel piping with no cast material and GTAW welding. From these evaluations a critical crack size is determined. That is, a crack larger than the critical crack size would have unstable growth characteristics.
Acceptance Standards
[The results of the preceding evaluations are compared to show that the critical flaw size, which is shown to be stable when the maximum loads are combined based on individual absolute values, is at least twice the size (to satisfy margin of 2 on flaw size) of the leakage flaw size. To satisfy a margin
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.6-31 Revision 9
VEGP 3&4 - UFSAR on load of 1.0, the maximum loads are combined using absolute summation of individual values.]*
The maximum loads are described in Appendix 3B Subsection 3B.3.3.
Bounding Analyses Evaluations are provided for each different combination of material type, pipe size, pressure, and temperature. These evaluations are used to develop a set of curves of maximum faulted stress versus the corresponding normal stress that satisfy the criteria for leak-before-break. These curves are used in the design of the piping systems and will be used to verify that the as-built piping satisfies the requirements for leak-before-break as discussed in Subsection 3.6.4.2.
3.6.3.4 Documentation of Leak-before-Break Evaluations The leak-before-break evaluation is used to support the elimination of dynamic effects of pipe breaks from the loading conditions for the piping analysis. An evaluation of leak-before-break using the as-built configuration of the piping system and supports is required as part of the Design Report (also referred to as LBB evaluation report where applicable) of the as-built configuration required to meet ASME Code requirements and LBB criteria. Appendix 3B contains a discussion of the bounding analysis methods for the leak-before-break evaluation.
The analysis methods, criteria, and loads used for evaluation of stress in piping systems are outlined in Subsections 3.7.3 and 3.9.3.
3.6.4 Combined License Information 3.6.4.1 Pipe Break Hazard Analysis The as-built reconciliation of the pipe break hazards analysis and as-built pipe rupture hazard analysis are addressed in APP-GW-GLR-021 (Reference 14), and the applicable changes are incorporated into the UFSAR.
The as-designed pipe rupture hazards evaluation is made available for NRC review. The completed as-designed pipe rupture hazards evaluation will be in accordance with the criteria outlined in Subsections 3.6.1.3.2 and 3.6.2.5. Systems, structures, and components identified to be essential targets protected by associated mitigation features (Reference is Table 3.6-3) will be confirmed as part of the evaluation, and updated information will be provided as appropriate.
A pipe rupture hazard analysis is part of the piping design. The evaluation will be performed for high and moderate energy piping to confirm the protection of systems, structures, and components which are required to be functional during and following a design basis event. The locations of the postulated ruptures and essential targets will be established and required pipe whip restraints and jet shield designs will be included. The report will address environmental and flooding effects of cracks in high and moderate energy piping. The as-designed pipe rupture hazards evaluation is prepared on a generic basis to address COL applications referencing the AP1000 design.
The pipe whip restraint and jet shield design includes the properties and characteristics of procured components connected to the piping, components, and walls at identified break and target locations. The design will be completed prior to installation of the piping and connected components.
The as-built reconciliation of the pipe rupture hazards evaluation whip restraint and jet shield design in accordance with the criteria outlined in Subsections 3.6.1.3.2 and 3.6.2.5 will be completed prior to fuel load (in accordance with DCD Tier 1 Table 3.3-6, item 8).
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.6-32 Revision 9
VEGP 3&4 - UFSAR This COL item is also addressed in Subsection 14.3.3.
3.6.4.2 Leak-before-Break Evaluation of As-Designed Piping The leak-before-break evaluation of the as-designed piping is addressed in APP-GW-GLR-022 (Reference 15), and the applicable changes are incorporated into the UFSAR.
3.6.4.3 Leak-before-Break Evaluation of As-Built Piping Not used.
3.6.4.4 Primary System Inspection Program for Leak-before-Break Piping Alloy 690 is not used in leak-before-break piping. No additional or augmented inspections are required beyond the inservice inspection program for leak-before-break piping. An as-built verification of the leak-before-break piping is required to verify that no change was introduced that would invalidate the conclusion reached in this subsection.
3.6.5 References
- 1. NUREG/CR-2913, Two-Phase Jet Loads, January 1983.
- 2. WCAP-8077 (Proprietary) and WCAP-8078 (Nonproprietary), Ice Condenser Containment Pressure Transient Analysis Methods, March 1973.
- 3. ASME B31.1 1989 Edition, Power Piping, including 1989 Addendum.
- 4. ANSI/ANS-58.2-1988, Design Bases for Protection of Light Water Nuclear Power Plants Against Effects of Postulated Pipe Rupture.
- 5. Moody, F. J., Fluid Reaction and Impingement Loads, paper presented at the ASCE Specialty Conference, Chicago, December 1973.
- 6. MULTIFLEX, A FORTRAN-IV Computer Program for Analyzing Thermal-Hydraulic-Structure System Dynamics, WCAP-8708 (Proprietary) and WCAP-8709 (Nonproprietary), February 1976.
- 7. WCAP-8252, Documentation of Selected Westinghouse Structural Analysis Computer Codes, Revision 1, May 1977.
- 8. Not used.
- 9. Biggs, J. M., Introduction to Structural Dynamics, McGraw-Hill Book Company, New York, 1964.
- 10. EPRI NP-5630, Guidelines for Piping System Reconciliation (NCIG-05, Revision 1),
May 1988.
- 11. NUREG-1061, Volume 3, Report of the U. S. Nuclear Regulatory Commission Piping Review Committee, Evaluation of Potential for Pipe Breaks, November 1984.
- 12. Standard Review Plan 3.6.3, Leak Before Break Evaluation Procedures, Federal Register, Volume 52, Number 167, Friday, August 28, 1987; Notice (Public Comment Solicited), pp. 32626-32633.
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- 13. NUREG/CR-6519, Screening Reactor Steam/Water Systems for Water Hammer, November 1996.
- 15. APP-GW-GLR-022, AP1000 Leak-Before-Break Evaluation of As-Designed Piping, Westinghouse Electric Company LLC.
- 16. Information Systems Laboratories (2006), RELAP5/MOD3.3 (Patch03) Code Manual Volume I to VIII, NUREG/CR-5535/Rev P3-Vol I to VIII, prepared for U.S. Nuclear Regulatory Commission.
- 17. ASTM A530/A530M - 99, Standard Specification for General Requirements for Specialized Carbon and Alloy Steel Pipe.
3.6-34 Revision 9
VEGP 3&4 - UFSAR Table 3.6-1 High-Energy and Moderate-Energy Fluid Systems Considered for Protection of Essential Systems(a)
System High-Energy Moderate-Energy Reactor coolant (RCS)..........................................................................................................
- Steam generator (SGS)(b) ....................................................................................................
- Passive core cooling (PXS)..................................................................................................
- Passive containment cooling (PCS)(c) ...........................................................................................................................
- Main control room habitability (VES) ..................................................................................
- Chemical and volume control (CVS)...................................................................................
- Primary sampling (PSS) .......................................................................................................
- Compressed and instrument air (CAS) ..............................................................................
- Normal residual heat removal (RNS)(a) .........................................................................................................................
- Component cooling water (CCS)....................................................................................................................................
- Spent fuel pool cooling (SFS) .........................................................................................................................................
- Demineralized water (DWS)............................................................................................................................................
- Liquid radwaste (WLS).....................................................................................................................................................
- Radioactive drain (WRS) .................................................................................................................................................
- Central chilled water (VWS) ............................................................................................................................................
- Fire protection (FPS) ........................................................................................................................................................
- Steam generator blowdown (BDS)(d)..................................................................................
- Main and startup feedwater (FWS)(d) .................................................................................
- Main steam (MSS)(d) .............................................................................................................
- Hot water heating (VYS) .................................................................................................................................................
- Notes:
- a. Systems included on this list are high-energy or moderate-energy fluid systems located in the containment or the auxiliary building. Systems that operate at or close to atmospheric pressure such as ventilation and gravity drains are not included.
The normal residual heat removal system lines are classified as moderate-energy based on the 1 percent rule. These lines experience high-energy conditions for less than 1 percent of the plant operating time. The portions of the normal residual heat removal system from the connections to the reactor coolant system and passive core cooling system to the first closed valve in each line are high energy. The spent fuel pool cooling system is classified as moderate energy based on the 2 percent rule. These systems experience high-energy conditions for less than 2 percent of the system operating time. See Subsection 3.6.1.1 Item A and Subsection 3.6.1.2 for additional information.
- b. Main and startup feedwater, main steam, and steam generator blowdown lines located in the containment and auxiliary building are part of the steam generator system.
- c. The essential portion of the system is at atmospheric pressure.
- d. The portion of these systems in the turbine building adjacent to the auxiliary building are evaluated for the effect of a circumferential or longitudinal break on the main control room.
3.6-35 Revision 9
VEGP 3&4 - UFSAR Table 3.6-2 (Sheet 1 of 7)
Subcompartments and Postulated Pipe Ruptures Compartment Lines Evaluated to LBB Lines Not Evaluated to LBB Terminal End Room Break Location Name Number Description Excluded by LBB Description Break Location Steam 11201 22 in. Cold Leg (RCS) RC Pump Nozzles 4 in. Pressurizer Cold Leg Nozzles (2)
Generator (2) Spray (RCS)
Compart- 18 in. Fourth Stage Hot Leg Nozzle 3 in. RC Pump RC Pump Cooling ment 1 ADS (RCS) External Piping Nozzles: A/B (4) to/from Motor Bearing HX 3 in. RC Pump Motor Bearing Water External Piping Cooling HX Nozzles:
to/from Motor A/B (4)
Bearing HX 3 in. RC Pump Intermediate breaks External Piping to/ along elbow fittings:
from Motor pumps A/B (8)
Bearing HX 11301 31 in. Hot Leg (RCS) SG Nozzle 3 in. Purification 3 in. SG Channel Head (CVS) Nozzle 18 in. Surge Line Hot Leg Nozzle (RCS) 18 in. & 14 in. Fourth Valves: V004A/C Stage ADS (RCS) 14 in. PRHR Return SG Channel Head (RCS) Nozzle 11401 None 4 in. SG Blowdown 4 in. SG Nozzle (SGS) 11501 None None 11601 16 in and 20 in. SG Nozzle Feedwater (SGS) 6 in. Startup SG Nozzle Feedwater (SGS) 11701 38 in. Main Steam SG Nozzle None (SGS) 3.6-36 Revision 9
VEGP 3&4 - UFSAR Table 3.6-2 (Sheet 2 of 7)
Subcompartments and Postulated Pipe Ruptures Compartment Lines Evaluated to LBB Lines Not Evaluated to LBB Terminal End Break Room Location Excluded Name Number Description by LBB Description Break Location Steam 11202 22 in. Cold Leg (RCS) RC Pump Nozzles 3 in. RC Pump RC Pump Cooling Generator (2) External Piping Nozzles: A/B (4)
Compart- to/from Motor ment 2 Bearing HX 18 in. Fourth Stage Hot Leg Nozzle 3 in. RC Pump Motor Bearing Water ADS (RCS) External Piping Cooling HX Nozzles:
to/from Motor A/B (4)
Bearing HX 20 in. Normal RHR Hot Leg Nozzle 3 in. RC Pump Intermediate breaks (RCS) External Piping along elbow fittings:
to/from Motor pumps A/B (8)
Bearing HX 12 in. Normal RHR 20 in. x 12 in.
(RCS) Reducer (This is not a terminal end) 11302 31 in. Hot Leg (RCS) SG Nozzle None 18 in. & 14 in. Fourth Valves: V004B/D Stage ADS (RCS) 8 in. Cold Leg to CMT Cold Leg Nozzles (2)
(RCS) 11402 None 4 in. SG Blowdown 4 in. SG Nozzle (SGS) 11502 None None 11602 16 in. and 20 in. SG Nozzle Feedwater (SGS) 6 in. Startup SG Nozzle Feedwater (SGS) 3.6-37 Revision 9
VEGP 3&4 - UFSAR Table 3.6-2 (Sheet 3 of 7)
Subcompartments and Postulated Pipe Ruptures Compartment Lines Evaluated to LBB Lines Not Evaluated to LBB Terminal End Break Room Location Excluded Name Number Description by LBB Description Break Location 11702 38 in. Main Steam SG Nozzle None (SGS)
Reactor 11205 31 in. Hot Leg (RCS) Reactor Vessel None Vessel Nozzles (2)
Nozzle 22 in. Cold Leg (RCS) Reactor Vessel Area Nozzles (4) 8 in. Direct Vessel Reactor Vessel Injection (RCS) Nozzles (2)
PXS Valve 11206 8 in. Accumulator Accumulator Nozzle None and Injection (PXS)
Accumulator 8 in. CMT Injection CMT Nozzle Room A (PXS) 6 in. Line from Normal Valve: V017A RHR (RNS) 8 in. Line from IRWST Valves: V125A &
(PXS) V123A PXS Valve 11207 6 in. Line from Normal Valve: V017B None Room B PXS RHR (RNS) 8 in. Line from IRWST Valves: V125B &
(PXS) V123B Accumulator 11207 8 in. Accumulator Accumulator Nozzle None Room B ACCUM Injection (PXS) 8 in. CMT Injection CMT Nozzle (PXS)
RNS Valve 11208 10 in. Normal RHR Valves: V001A/B None Room (RNS) 3.6-38 Revision 9
VEGP 3&4 - UFSAR Table 3.6-2 (Sheet 4 of 7)
Subcompartments and Postulated Pipe Ruptures Compartment Lines Evaluated to LBB Lines Not Evaluated to LBB Terminal End Room Break Location Name Number Description Excluded by LBB Description Break Location Vertical 11204 None 3 in. Line from Regen Anchor to Wall Access HX to SG 01 (CVS) 3 in. Purification from Anchor to Wall Cold Leg to Regen HX (CVS)
Lower 11303 18 in. Surge Line Pressurizer Nozzle None Pressurizer (RCS)
Compartment Upper 11503 14 in. ADS (RCS) Pressurizer Nozzle 4 in. Pressurizer Pressurizer Nozzle Pressurizer (2) Spray (RCS)
Compartment Lower ADS 11603 14 in. & 8 in. ADS Valves: V012B & 4 in. ADS (RCS) Valve V0011B & 14 Valve Area (RCS) V013B in. x 4 in. Branch 6 in. Pressurizer Valves: V005A and Safety (RCS) V005B 14 in. x 6 in. Tees (2)
Upper ADS 11703 14 in. & 8 in., ADS Valves: V012A & 4 in. ADS (RCS) Valve V0011A &
Valve Area (RCS) V013A 14 in. x 4 in. Branch Maintenance 11400 38 in. Main Steam Non-terminal End 6 in. Startup Anchors (2) at Floor/ (SGS) Location (2) at Feedwater (SGS) Containment Mezzanine Boundary of Break Penetration Exclusion Zone 14 in. Passive RHR PRHR HX Inlet (PXS) Nozzle 8 in. CMT Balance CMT Nozzles (2)
Line Piping 3.6-39 Revision 9
VEGP 3&4 - UFSAR Table 3.6-2 (Sheet 5 of 7)
Subcompartments and Postulated Pipe Ruptures Compartment Lines Evaluated to LBB Lines Not Evaluated to LBB Terminal End Room Break Location Name Number Description Excluded by LBB Description Break Location SG01 Access 11304 None 3 in. CVS Return from Intermediate Room Regen HX breaks along pipe fittings:
valve transition (3), weldolet (1)
Pressurizer 11403 None 4 in. Pressurizer Spray Intermediate Spray Valve to 2 in. Auxiliary Spray breaks along Room pipe fitting: 4x2
reducing tee (9)
Maintenance 11300 14 in. Passive RHR PRHR HX Outlet 1.5 in. PRHR Return Sockolet for flow Floor (PXS) Nozzle flow instrument branch element line (PXS)
Operating 11500 None None Deck CVS Room 11209 None 3 in. Purification from Regen HX Pressurizer Spray to Nozzle Regen HX (CVS) 3 in. Return, Auxiliary Regen HX Spray (CVS) Nozzle 3 in. Return to RNS Valve: V079 from Regen HX (CVS) 3 in. Supply from RNS Valve: V072 to Letdown HX (CVS) 3 in. Supply from Regen Nozzles:
HX to Letdown HX Regen HX, (CVS) Letdown HX CVS Room 11209 None 3 in. Purification from Anchor Pipe Anchor to Regen HX Chase 3 in. Return from Regen Anchor HX to Anchor (CVS) 4 in. SG Blowdown Anchors (2) at (SGS) Containment Penetration 3.6-40 Revision 9
VEGP 3&4 - UFSAR Table 3.6-2 (Sheet 6 of 7)
Subcompartments and Postulated Pipe Ruptures Compartment Lines Evaluated to LBB Lines Not Evaluated to LBB Terminal End Room Break Location Name Number Description Excluded by LBB Description Break Location Reactor Coolant 11104 None None Drain Tank Room Reactor Vessel 11105 None None Cavity MSIV 12504/ None Main Steam Main Longitudinal Compartment B 12404 Feedwater Startup Cracks with Crack (Upper/Lower) Feedwater Lines(a) Flow Areas of 1 Square Foot are Postulated MSIV 12506/ None Main Steam Main Longitudinal Compartment A 12406 Feedwater Startup Cracks with Crack (Upper/Lower) Feedwater Lines(a) Flow Areas of 1 Square Foot are Postulated Valve/Piping 12306 None 4 in. Steam Generator Anchors (2) at Penetration Blowdown(a) Containment Room Penetrations Anchors (2) at Wall to Turbine Building Note:
- a. The piping in these areas is included in break exclusion zones. For additional information on the evaluation of these lines, see Subsection 3.6.1.2.1 for the steam generator blowdown line; Subsection 3.6.1.2.2 for information on the evaluation of lines in MSIV compartment B because of the proximity to the main control room; and Subsection 3.6.2.1.1.4 for general break exclusion zone requirements.
3.6-41 Revision 9
VEGP 3&4 - UFSAR Table 3.6-2 (Sheet 7 of 7)
Subcompartments and Postulated Pipe Ruptures Room # Description Bottom Elevation Top Elevation 11104 RCDT Room 66-6 81-0 11105 Reactor Vessel Cavity 66-6 98 11205 Reactor Vessel Nozzle Area 98 107-2 11201 SG Compartment 1 83 104-7 11202 SG Compartment 2 83 104-7 11204 Vertical Access 83 107-2 11206 PXS Valve and Accumulator Room A 87-6 105-2 11300 Maintenance Floor 107-2 118-6 11301 SG Compartment 1 104-7 116-6 11302 SG Compartment 2 104-7 116-6 11400 Maintenance Floor/Mezzanine 118-6 133-3 11401 SG Compartment 1 116-6 135-3 11402 SG Compartment 2 116-6 135-3 11501 SG Compartment 1 135-3 153-0 11502 SG Compartment 2 135-3 153-0 11601 SG Compartment 1 153-0 166-4 11602 SG Compartment 2 153-0 166-6 11701 SG Compartment 1 166-4 ----
11702 SG Compartment 2 166-4 ----
11500 Operating Deck 135-3 281-8 3/8 11303 Lower Pressurizer Compartment 107-2 135-3 11304 SG01 Access Room 107-2 118-6 11403 Pressurizer Spray Valve Room 118-6 133-3 11503 Upper Pressurizer Compartment 135-3 166-1.5 11603 Lower ADS Valve Area 166-1.5 176-10.5 11703 Upper ADS Valve Area 176-10.5 ----
11207 ACCUM Accumulator Room B 87-6 105-2 11207 PXS PXS Valve Room B 87-6 105-2 11208 RNS Valve Room 94 105-2 11209 CVS Room 80-6 105-2 11209 PIPE CHASE CVS Room Pipe Chase 100-0 105-2 12306 Valve/Piping Penetration Room 100-0 117-6 12504/12404 MSIV Compartment B (Upper/Lower) 117-6 153-0 12506/12406 MSIV Compartment A (Upper/Lower) 117-6 153-0 3.6-42 Revision 9
VEGP 3&4 - UFSAR Table 3.6-3 (Sheet 1 of 10)
NI Rooms With Pipe Whip Restraints, Dual Acting Pipe Supports, Jet Shields and Corresponding Hazard Sources and Essential Targets Protection Room Room Mechanism Number Description (Break ID) Hazard Source Essential Target Description/Room 11201 Steam Generator Pipe Whip Restraint Reactor Coolant System (RCS) RCS spray control valve (RCS-V110A) (Room 11403).
Compartment-01 RCS-SH-W002 Pressurizer Spray Line, 4" RCS pressurizer spray bypass valve (RCS-V210A) (Room 11403).
(RCS-002) L110A: Terminal End Break at RCS spray control valve (RCS-V110B) (Room 11403).
RCS Cold Leg RCS pressurizer spray bypass valve (RCS-V210B) (Room 11403).
RCS-PL-L002A/N11A Pipe Whip Restraint RCS Pressurizer Spray Line, 4" RCS spray control valve (RCS-V110A) (Room 11403).
RCS-SH-W003 L106: Terminal End Break at RCS pressurizer spray bypass valve (RCS-V210A) (Room 11403).
(RCS-003) RCS Cold Leg RCS spray control valve (RCS-V110B) (Room 11403).
RCS-PL-L002B/N11B RCS pressurizer spray bypass valve (RCS-V210B) (Room 11403).
11204 Vertical Access Dual Acting Pipe Chemical and Volume Control SGS blowdown piping (L009B).
Room Support System (CVS) Letdown Line, 3" CCS isolation valve (CCS-V201) (Room 11400).
CVS-PH-11R2270 L002: Terminal End Break at CCS isolation valve (CCS-V207) (Room 11400).
(CVS-004U) In-Line Anchor CCS isolation valve (CCS-V220) (Room 11400).
CVS purification stop valve (CVS-V001) (Room 11403).
CVS purification stop valve (CVS-V002) (Room 11403).
CVS purification stop valve (CVS-V003) (Room 11403).
3.6-43 Revision 9
VEGP 3&4 - UFSAR Table 3.6-3 (Sheet 2 of 10)
NI Rooms With Pipe Whip Restraints, Dual Acting Pipe Supports, Jet Shields and Corresponding Hazard Sources and Essential Targets Protection Room Room Mechanism Number Description (Break ID) Hazard Source Essential Target Description/Room 11204 Vertical Access Dual Acting Pipe CVS Purification Return Line, 3" CCS isolation valve (CCS-V201) (Room 11400).
Room Support L018: Terminal End Break at CCS isolation valve (CCS-V207) (Room 11400).
CVS-PH-11R2271 In-Line Anchor CCS isolation valve (CCS-V220) (Room 11400).
(CVS-007D) CVS regenerative HX shell side outlet check valve (CVS-V080)
(Room 11303).
CVS purification return line stop check valve (CVS-V081)
(Room 11304).
CVS purification return line check valve (CVS-V082) (Room 11303).
CVS letdown containment isolation test connection valve (CVS-V084)
(Room 11304).
CVS auxiliary pressurizer spray line isolation valve (CVS-V085)
(Room 11304).
11209 CVS Room Pipe Whip Restraint Steam Generator System (SGS) SGS blowdown piping (L009B).
(Pipe Chase) SGS-SH-W004 Blowdown Line, 4" L009A: CVS isolation valve (CVS-V042) (Room 11209).
(SGS-004) Terminal End Break at CVS isolation valve (CVS-V045) (Room 11300).
Containment Penetration P27 CVS isolation valve (CVS-V058) (Room 11300).
CVS isolation valve (CVS-V091) (Room 11300).
CVS isolation valve (CVS-V100) (Room 11300).
CVS isolation valve (CVS-V217) (Room 11300).
WLS isolation valve (WLS-V055) (Room 11300).
WLS isolation valve (WLS-V058) (Room 11300).
WLS isolation valve (WLS-V067) (Room 11300).
3.6-44 Revision 9
VEGP 3&4 - UFSAR Table 3.6-3 (Sheet 3 of 10)
NI Rooms With Pipe Whip Restraints, Dual Acting Pipe Supports, Jet Shields and Corresponding Hazard Sources and Essential Targets Protection Room Room Mechanism Number Description (Break ID) Hazard Source Essential Target Description/Room 11209 CVS Room Pipe Whip Restraint SGS Blowdown Line, 4" CVS isolation valve (CVS-V042) (Room 11209).
(cont.) (Pipe Chase) SGS-SH-W008 L009B: Terminal End Break at CVS isolation valve (CVS-V045) (Room 11300).
(SGS-008) Containment Penetration P28 CVS isolation valve (CVS-V058) (Room 11300).
CVS isolation valve (CVS-V091) (Room 11300).
CVS isolation valve (CVS-V100) (Room 11300).
CVS isolation valve (CVS-V217) (Room 11300).
WLS isolation valve (WLS-V067) (Room 11300).
Dual Acting Pipe CVS Makeup Line, 3" L056: SGS blowdown piping (L009A).
Supports Terminal End Break at In-Line SGS blowdown piping (L009B).
CVS-PH-11Y2265 Anchor CVS isolation valve (CVS-V045) (Room 11300).
CVS-PH-11Y2266 CVS isolation valve (CVS-V058) (Room 11300).
(CVS-056U) CVS isolation valve (CVS-V091) (Room 11300).
CVS isolation valve (CVS-V100) (Room 11300).
WLS isolation valve (WLS-V055) (Room 11300).
WLS isolation valve (WLS-V058) (Room 11300).
WLS isolation valve (WLS-V067) (Room 11300).
3.6-45 Revision 9
VEGP 3&4 - UFSAR Table 3.6-3 (Sheet 4 of 10)
NI Rooms With Pipe Whip Restraints, Dual Acting Pipe Supports, Jet Shields and Corresponding Hazard Sources and Essential Targets Protection Room Room Mechanism Number Description (Break ID) Hazard Source Essential Target Description/Room 11209 CVS Room Pipe Whip Restraint CVS Letdown Line, 3" L002: SGS blowdown piping (L009A).
(cont.) (Pipe Chase) CVS-SH-W004D Terminal End Break at In-Line SGS blowdown piping (L009B).
(CVS-004D) Anchor CVS isolation valve (CVS-V042) (Room 11209).
CVS isolation valve (CVS-V045) (Room 11300).
CVS isolation valve (CVS-V058) (Room 11300).
CVS isolation valve (CVS-V091) (Room 11300).
CVS isolation valve (CVS-V100) (Room 11300).
CVS isolation valve (CVS-V217) (Room 11300).
WLS isolation valve (WLS-V055) (Room 11300).
WLS isolation valve (WLS-V058) (Room 11300).
WLS isolation valve (WLS-V067) (Room 11300).
Pipe Whip Restraint CVS Purification Return Line, 3" SGS blowdown piping (L009A).
CVS-SH-W007U L018: Terminal End Break at SGS blowdown piping (L009B).
(CVS-007U) In-Line Anchor CVS isolation valve (CVS-V042) (Room 11209).
CVS isolation valve (CVS-V045) (Room 11300).
CVS isolation valve (CVS-V058) (Room 11300).
CVS isolation valve (CVS-V091) (Room 11300).
CVS isolation valve (CVS-V100) (Room 11300).
CVS isolation valve (CVS-V217) (Room 11300).
WLS isolation valve (WLS-V055) (Room 11300).
WLS isolation valve (WLS-V058) (Room 11300).
WLS isolation valve (WLS-V067) (Room 11300).
3.6-46 Revision 9
VEGP 3&4 - UFSAR Table 3.6-3 (Sheet 5 of 10)
NI Rooms With Pipe Whip Restraints, Dual Acting Pipe Supports, Jet Shields and Corresponding Hazard Sources and Essential Targets Protection Room Room Mechanism Number Description (Break ID) Hazard Source Essential Target Description/Room 11301 Steam Generator 1 Pipe Whip Restraint CVS Purification Return CVS isolation valve (CVS-V084) (Room 11304).
Lower Manway RCS-SH-W001 Nozzle, 3" RCS L112: Terminal SGS blowdown piping (L009A) (Room 11301).
Area A/B (RCS-001) End Break at RCS-MB-01/N05 RCS Passive Residual Heat Removal (PRHR) Heat Exchanger (HX) return piping (L113) (Room 11301).
11303 Lower Pressurizer Pipe Whip Restraint CVS Purification Return Line 3" Column support for pressurizer (RCS-MV-02) (Room 11303).
Compartment CVS-SH-W210 RCS L112: Intermediate Break at SG nozzle RCS-MB-01/nozzle N05 (Room 11301).
(CVS-210-C) Outlet CVS-V082 RCS surge line - line RCS-L003 (Room 11303).
Dual Acting Pipe RCS Pressurizer Spray Line, 4" Lateral support struts for pressurizer (RCS-MV-02) (Room 11303).
Support L215: Intermediate Break at RCS instrument piping (L225A) (Room 11303).
RCS-PH-11Y1132 4" x 2" Reducing Tee RCS level instrument (LT195A) (Room 11300).
(RCS-201-C) RCS pressure transmitter (PT191A) (Room 11300).
Pressurizer nozzle - RCS-MV-02/nozzle N02 (Room 11503).
11400 Maintenance Floor Pipe Whip Restraint SGS Startup Feedwater Line, 6" Steam generator 2 (RCS-MB-02) (Room 11602).
Mezzanine SGS-SH-W006 L005B: Terminal End Break at (SGS-006) SGS-PY-C05B, Containment Penetration P45 Pipe Whip Restraint SGS Startup Feedwater Line, 6" Steam generator 1 (RCS-MB-01) (Room 11601).
SGS-SH-W002 L005A: Terminal End Break at Main steam 1 pressure transmitters - PT030/PT032 (Room 11400).
(SGS-002) SGS-PY-C05A, Containment Steam generator 1 (RCS-MB-01) level instrumentation piping -
Penetration P44 SGS-PL-L039A/SGS-PL-L041A, (Room 11400).
3.6-47 Revision 9
VEGP 3&4 - UFSAR Table 3.6-3 (Sheet 6 of 10)
NI Rooms With Pipe Whip Restraints, Dual Acting Pipe Supports, Jet Shields and Corresponding Hazard Sources and Essential Targets Protection Room Room Mechanism Number Description (Break ID) Hazard Source Essential Target Description/Room 11401 Steam Generator 1 Pipe Whip Restraint SGS Blowdown Line, 4" SGS blowdown penetration (SGS-PY-C03A).
Tube Sheet Area SGS-SH-W003 L009A: Terminal End Break at (SGS-003) RCS-MB-01/N06A 11402 Steam Generator 2 Pipe Whip Restraint SGS Blowdown Line, 4" SGS blowdown penetration (SGS-PY-C03B).
Tube Sheet Area SGS-SH-W007 L009B: Terminal End Break at Main steam 2 pressure transmitters - PT034/PT036 (Room 11402).
(SGS-007) RCS-MB-02/N06A Steam generator 2 (RCS-MB-02) level instrumentation piping -
SGS-PL-L039B/SGS-PL-L041B (Room 11402).
11403 Pressurizer Spray Dual Acting Pipe RCS Pressurizer Spray Valve RCS spray control valve (RCS-V110A) (Room 11403).
Valve Room Support Line, 4" L213: Intermediate Break RCS pressurizer spray bypass valve (RCS-V210A) (Room 11403).
RCS-PH-11R0084 at 4" x 2" Reducing Tee RCS spray control valve (RCS-V110B) (Room 11403).
(RCS-203-C) RCS pressurizer spray bypass valve (RCS-V210B) (Room 11403).
3.6-48 Revision 9
VEGP 3&4 - UFSAR Table 3.6-3 (Sheet 7 of 10)
NI Rooms With Pipe Whip Restraints, Dual Acting Pipe Supports, Jet Shields and Corresponding Hazard Sources and Essential Targets Protection Room Room Mechanism Number Description (Break ID) Hazard Source Essential Target Description/Room 11503 Upper Pressurizer Pipe Whip Restraint RCS Pressurizer Spray Line, 4" ADS Stage 1, 2, and 3 valves (RCS-V001B, RCS-V002B, Compartment RCS-SH-W006 L215: Terminal End Break at RCS-V003B, RCS-V011B, RCS-V012B, and RCS-V013B)
(RCS-006) RCS-MV-02/N02 (Room 11603).
ADS Stage 1, 2, and 3 valves (RCS-V001A, RCS-V002A, RCS-V003A, RCS-V011A, RCS-V012A, and RCS-V013A)
(Room 11703).
Pressurizer safety valves (RCS-V005A, RCS-V005B) (Room 11603).
ADS discharge header B vacuum relief valve (RCS-V010B)
(Room 11603).
ADS discharge header A vacuum relief valve (RCS-V010A)
(Room 11703).
ADS Stage 1, 2, and 3 support steel (Rooms 11503, 11603, and 11703).
RCS spray control valve (RCS-V110A) (Room 11403).
RCS pressurizer spray bypass valve (RCS-V210A) (Room 11403).
RCS spray control valve (RCS-V110B) (Room 11403).
RCS pressurizer spray bypass valve (RCS-V210B) (Room 11403).
3.6-49 Revision 9
VEGP 3&4 - UFSAR Table 3.6-3 (Sheet 8 of 10)
NI Rooms With Pipe Whip Restraints, Dual Acting Pipe Supports, Jet Shields and Corresponding Hazard Sources and Essential Targets Protection Room Room Mechanism Number Description (Break ID) Hazard Source Essential Target Description/Room 11601 Steam Generator 1 Pipe Whip Restraint SGS Startup Feedwater Line, 6" RCS head vent valve (RCS-V150A) (Room 11601).
Feedwater Nozzle SGS-SH-W001 L005A: Terminal End Break at RCS head vent valve (RCS-V150B) (Room 11601).
Area (153'-0") (SGS-001) RCS-MB-01/N20 RCS head vent valve (RCS-V150C) (Room 11601).
RCS head vent valve (RCS-V150D) (Room 11601).
SGS startup feedwater penetration (SGS-PY-C05A).
Pipe Whip Restraint SGS Main Feedwater Line, 16" SGS main feedwater penetration (SGS-PY-C02A).
SGS-SH-W021 L008A/L003A: Terminal End Break exclusion zone: SGS main feedwater piping (L002A).
(SGS-021) Break at RCS-MB-01/N04.
11602 Steam Generator 2 Pipe Whip Restraint SGS Startup Feedwater Line, 6" SGS startup feedwater penetration (SGS-PY-C05B).
Feedwater Nozzle SGS-SH-W005 L005B: Terminal End Break at Area (153'-0") (SGS-005) RCS-MB-02/N20 Pipe Whip Restraint SGS Main Feedwater Line, 16" SGS main feedwater penetration (SGS-PY-C02B).
SGS-SH-W022 L008B/L003B: Terminal End Break exclusion zone: SGS main feedwater piping (L002B).
(SGS-022) Break at RCS-MB-02/N04 Jet Shield SGS Main Feedwater Line, 16" Main steam line snubber for support APP-SGS-PH-11Y7057 SGS-SH-P022 L008B/L003B: Terminal End (Room 11602).
(SGS-022) Break at RCS-MB-02/N04 3.6-50 Revision 9
VEGP 3&4 - UFSAR Table 3.6-3 (Sheet 9 of 10)
NI Rooms With Pipe Whip Restraints, Dual Acting Pipe Supports, Jet Shields and Corresponding Hazard Sources and Essential Targets Protection Room Room Mechanism Number Description (Break ID) Hazard Source Essential Target Description/Room 11603 ADS Valve Pipe Whip Restraint RCS ADS Stage 1 Train B Inlet ADS stage 1, 2, and 3 valves (RCS-V002B, RCS-V003B, Area - Lower Tier RCS-SH-W108 Line, 4" L010B: Terminal End RCS-V012B, and RCS-V013B) (Room 11603).
(RCS-108) Break at First Stage ADS ADS stage 1, 2, and 3 valves (RCS-V001A, RCS-V002A, Isolation Valve RCS-PL-V011B RCS-V003A, RCS-V011A, RCS-V012A, and RCS-V013A)
(Room 11703).
ADS discharge header B vacuum relief valve (RCS-V010B)
(Room 11603).
ADS discharge header A vacuum relief valve (RCS-V010A)
(Room 11703).
ADS stage 1, 2, and 3 support steel 1102-Q6-01 (Rooms 11503, 11603, and 11703).
11703 Upper ADS Valve Pipe Whip Restraint RCS ADS Stage 1 Train A Inlet ADS stage 1, 2, and 3 valves (RCS-V001B, RCS-V002B, Area - 176'-10 1/2" RCS-SH-W106 Line, 4" L010A: Terminal End RCS-V003B, RCS-V011B, RCS-V012B, and RCS-V013B)
(RCS-106) Break at First Stage ADS (Room 11603).
Isolation Valve RCS-PL-V011A ADS stage 1, 2, and 3 valves (RCS-V002A, RCS-V003A, RCS-V012A, and RCS-V013A) (Room 11703).
ADS discharge header B vacuum relief valve (RCS-V010B)
(Room 11603).
ADS discharge header A vacuum relief valve (RCS-V010A)
(Room 11703).
ADS stage 1, 2, and 3 support steel 1102-Q6-01 (Rooms 11503, 11603, and 11703).
3.6-51 Revision 9
VEGP 3&4 - UFSAR Table 3.6-3 (Sheet 10 of 10)
NI Rooms With Pipe Whip Restraints, Dual Acting Pipe Supports, Jet Shields and Corresponding Hazard Sources and Essential Targets Protection Room Room Mechanism Number Description (Break ID) Hazard Source Essential Target Description/Room 12258 Degasifier Column Dual Acting Pipe CVS Makeup Pump Discharge Break exclusion zone: CVS makeup piping (L053).
Room Supports Header, 3" L131: Terminal End CVS-PH-12R2272 Break at Anchor CVS-PH-12R2273 CVS-PH-12A0122 (CVS-061D) 12404 East MSIV Pipe Whip Restraint Main Feedwater Line, 20" Floor of Room 12404 (protection of electrical/I&C components in Compartment SGS-SH-W033 L002B and L003B: Longitudinal Room 12304 - room below east MSIV compartment).
SGS-SH-W034 Splits SGS-SH-W035 Pipe Whip Restraint Main Feedwater Line, 20" Wall 11 (Room 12404) and Wall L (Room 12404 - protection of main SGS-SH-W031 L002B and L003B: Longitudinal control room).
SGS-SH-W032 Splits and Guillotine Breaks 3.6-52 Revision 9
VEGP 3&4 - UFSAR HOT POSITION GAP BEARING BAR SADDLE CLIP SADDLE PLATE STRAP PLATE U-BAR INSULATION TAPPED BAR OR CLEVIS BRACKET PLATE SADDLE PIN &
COTTER PIN BASE PLATE SUPPORTING STRUCTURE Figure 3.6-1 Typical U-Bar Restraint 3.6-53 Revision 9
VEGP 3&4 - UFSAR Process Piping Crushable Pipe Slots to Allow for Adjustability Adjustable Bearing Plate to Overlay Plate Clip Anchorage Bolt Welded to Embedment Slots to Allow for Pipe to Plate Clip Adjustability Process Piping Slots to Allow for Installation and Removal Crushable Pipe Attachment Plate Shim Plate to Allow for Adjustability Anchorage Bolt Welded to Attachment Plate Figure 3.6-2 Typical Structural Components of Crushable Pipe Type Whip Restraints 3.6-54 Revision 9
$ % & '
$+/-$QFKRU
%+/-&ORVHG9DOYH
&+/-&ORVHG9DOYH
'+/-7HUPLQDO(QG
$WR%+/-+LJK(QHUJ\
%WR'+/-0RGHUDWH(QHUJ\
Figure 3.6-3 Terminal Ends Definitions 3.6-55 Revision 9
VEGP 3&4 - UFSAR 3.7 Seismic Design Plant structures, systems, and components important to safety are required by General Design Criterion (GDC) 2 of Appendix A of 10 CFR 50 to be designed to withstand the effects of earthquakes without loss of capability to perform their safety functions.
Each plant structure, system, equipment, and component is classified in an applicable seismic category depending on its function. A three-level seismic classification system is used for the AP1000: seismic Category I, seismic Category II, and nonseismic. The definitions of the seismic classifications and a seismic classifications listing of structures, systems, equipment, and components are presented in Section 3.2.
Seismic design of the AP1000 seismic Categories I and II structures, systems, equipment, and components is based on the safe shutdown earthquake (SSE). The safe shutdown earthquake is defined as the maximum potential vibratory ground motion at the generic plant site as identified in Section 2.5.
The operating basis earthquake (OBE) has been eliminated as a design requirement for the AP1000.
Low-level seismic effects are included in the design of certain equipment potentially sensitive to a number of such events based on a percentage of the responses calculated for the safe shutdown earthquake. Criteria for evaluating the need to shut down the plant following an earthquake are established using the cumulative absolute velocity approach according to EPRI Report NP-5930 (Reference 1) and EPRI Report TR-100082 (Reference 17). For the purposes of the shutdown criteria in Reference 1 the operating basis earthquake for shutdown is considered to be one-third of the safe shutdown earthquake.
Seismic Category I structures, systems, and components are designed to withstand the effects of the safe shutdown earthquake event and to maintain the specified design functions. Seismic Category II and nonseismic structures are designed or physically arranged (or both) so that the safe shutdown earthquake could not cause unacceptable structural interaction with or failure of seismic Category I structures, systems, and components.
3.7.1 Seismic Input The geologic and seismologic considerations of the plant site are discussed in Section 2.5.
The peak ground acceleration of the safe shutdown earthquake, now referred to as the Certified Seismic Design Response Spectra (CSDRS), has been established as 0.30g for the AP1000 design.
The vertical peak ground acceleration is conservatively assumed to equal the horizontal value of 0.30g as discussed in Section 2.5.
3.7.1.1 Design Response Spectra The AP1000 design response spectra of the safe shutdown earthquake, now referred to as the Certified Seismic Design Response Spectra (CSDRS), are provided in Figures 3.7.1-1 and 3.7.1-2 for the horizontal and the vertical components, respectively.
The horizontal design response spectra for the AP1000 plant are developed, using the Regulatory Guide 1.60 spectra as the base and several evaluations to investigate the high frequency amplification effects. These evaluations included:
Comparison of Regulatory Guide 1.60 spectra with the spectra predicted by recent eastern U.S. spectral velocity attenuation relations (References 23, 24, 25, and 26) using a suite of magnitudes and distances giving a 0.3 g peak acceleration 3.7-1 Revision 9
VEGP 3&4 - UFSAR Comparison of Regulatory Guide 1.60 spectra with the 10-4 annual probability uniform hazard spectra developed for eastern U.S. nuclear power plants by both Lawrence Livermore National Laboratory (Reference 27) and Electric Power Research Institute (Reference 28)
Comparison of Regulatory Guide 1.60 spectra with the spectra of 79 additional old and newer components of strong earthquake time histories not considered in the original derivation of Regulatory Guide 1.60 Based on the above described evaluations, it is concluded that the eastern U.S. seismic data exceed Regulatory Guide 1.60 spectra by a modest amount in the 15 to 33 hertz frequency range when derived either from published attenuation relations or from the 10-4 annual probability of exceedance uniform hazard spectra at eastern U.S. sites. This conclusion is consistent with findings of other investigators that eastern North American earthquakes have more energy at high frequencies than western earthquakes. Exceedance of Regulatory Guide 1.60 spectra at the high frequency range, therefore, would be expected since Regulatory Guide 1.60 spectra are based primarily on western U.S. earthquakes. The evaluation shows that, at 25 hertz (approximately in the middle of the range of high frequencies being considered, and a frequency for which spectral amplitudes are explicitly evaluated) the mean-plus-one-standard-deviation spectral amplitudes for 5 percent damping range from about 2.1 to 4 cm/sec and average 2.7 cm/sec. Whereas, the Regulatory Guide 1.60 spectral amplitude at the same frequency and damping value equal just over 2 cm/sec.
It is concluded, therefore, that an appropriate augmented 5 percent damping horizontal design velocity response spectrum for the AP1000 project is one with spectral amplitudes equal to the Regulatory Guide 1.60 spectrum at control frequencies 0.25, 2.5, 9 and 33 hertz augmented by an additional control frequency at 25 hertz with an amplitude equal to 3 cm/sec. This spectral amplitude equals 1.3 times the Regulatory Guide 1.60 amplitude at the same frequency. The additional control points spectral amplitude of other damping values were determined by increasing the Regulatory Guide 1.60 spectral amplitude by 30 percent.
The AP1000 design vertical response spectrum is, similarly, based on the Regulatory Guide 1.60 vertical spectra at lower frequencies but is augmented at the higher frequencies equal to the horizontal response spectrum.
The AP1000 design response spectras relative values of spectrum amplification factors for control points are presented in Table 3.7.1-3.
The design response spectra are applied at the foundation level in the free field at hard rock sites and at the finished grade in the free field at firm rock and soil sites. The resulting peak horizontal ground acceleration values are above 0.1g. This satisfies 10 CFR Part 50, Appendix S, which requires that the horizontal component of the SSE ground motion in the free-field at the foundation elevation (that is, bottom of foundation) has a peak ground acceleration of at least 0.1g together with an appropriate response spectrum. The definitions (characteristics) of hard rock, firm rock, and soil sites are provided in Subsection 3.7.1.4.
3.7.1.1.1 Design Ground Motion Response Spectra The Vogtle site-specific safe shutdown earthquake (SSE) design response spectra (DRS) are the site-specific ground motion response spectra (GMRS) determined in Subsection 2.5.2.6. These response spectra are determined in the free-field on the ground surface.
The Vogtle foundation input response spectra (FIRS) are at an outcrop located at the 40' depth. The development of these FIRS is discussed in Subsection 2.5.2.7. These Vogtle response spectra are compared to the AP1000 SSE design response spectra that are also referred to as the AP1000 certified seismic design response spectra (CSDRS). The CSDRS also represents the AP1000 FIRS.
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VEGP 3&4 - UFSAR This is because: (1) the CSDRS at a hard rock site is essentially the same at grade and at foundation; and (2) the CSDRS envelopes the in-column motions of the other generic soil conditions.
The AP1000 CSDRS are applied at the foundation level in the free field at hard rock sites, and at the finished grade for the other soil generic conditions. The comparisons are shown in Figures 3.7-201 and 3.7-202. As seen from those comparisons, there are exceedances above the CSDRS; therefore, plant specific seismic evaluations are performed that demonstrate that the AP1000 plant designed for the CSDRS is acceptable for the Vogtle site. The results from a Vogtle site specific two-dimensional seismic evaluation that demonstrates the acceptability of the Vogtle site are given in Appendix 2.5E.
Additionally, a Vogtle site specific three-dimensional seismic evaluation that demonstrates the acceptability of the Vogtle site is given in Appendix 3GG. Based on these Vogtle site specific seismic evaluations it can also be concluded that the standard AP1000 plant certified design is fully acceptable to a SSE design response spectra level of the CSDRS at Vogtle's plant grade.
As discussed in Subsection 2.5.4.13, the heavy lift derrick (HLD) counterweight and ring foundation were abandoned in place after construction. The HLD counterweight is outside the defined excavation of Unit 3 and Unit 4 and therefore does not need to be evaluated. Portions of the HLD ring foundation extend over the Unit 3 and Unit 4 excavation slopes within the engineered granular backfill (EGB); but outside the Category 1 and 2 backfill. The presence of the HLD ring foundation has no effect on the VEGP site-specific 3D SASSI SSI analyses of the Nuclear Island (NI) presented in Appendix 3GG based on the following information.
The VEGP site-specific 3D SASSI SSI of the NI is consistent with the accepted DCD 3D SASSI NI modeling approach of not including structure-to-structure interaction of the adjacent structures such as the Annex Building and the Turbine Building; and therefore the more distant abandoned HLD ring foundation has even less structure-to-structure effects on the NI seismic response. Additionally only a portion of the abandoned HLD ring foundation is within a limited area of the non-safety EGB over the slopes of the excavation. It has been demonstrated in the ESP as amended that a large variation of the EGB properties does not significantly affect the site-specific seismic analyses; therefore, it is concluded the abandoned portion of the HLD ring foundation in the EGB has no significant effect on the site-specific seismic analyses.
The operating basis earthquake ground motion (OBE) spectral values are used as one measure of potential damage to those structures, systems, and components designed to the SSE design ground motion to determine the severity of the seismic event and make a determination of whether the plant must be shut down. For the AP1000 certified design, OBE is not an explicit design load; as such it is therefore defined as one-third the CSDRS. Since it has been demonstrated that the Vogtle site characteristics do not limit the AP1000 design to the CSDRS, the Vogtle OBE for the AP1000 is defined as one-third the AP1000 CSDRS.
The FIRS and the CSDRS in the horizontal direction in the free-field at the foundation of the AP1000 Nuclear Island exceed the minimum spectrum requirements of 10 CFR50 Appendix S.
3.7.1.2 Design Time History A "single" set of three mutually orthogonal, statistically independent, synthetic acceleration time histories is used as the input in the dynamic analysis of seismic Category I structures. The synthetic time histories were generated by modifying a set of actual recorded "TAFT" earthquake time histories. The design time histories include a total time duration equal to 20 seconds and a corresponding stationary phase, strong motion duration greater than 6 seconds. The acceleration, velocity, and displacement time-history plots for the three orthogonal earthquake components, "H1,"
"H2," and "V," are presented in Figures 3.7.1-3, 3.7.1-4, and 3.7.1-5. Design horizontal time history, H1, is applied in the north-south (Global X or 1) direction; design horizontal time history, H2, is applied in the east-west (global Y or 2) direction; and design vertical time history is applied in the 3.7-3 Revision 9
VEGP 3&4 - UFSAR vertical (global Z or 3) direction. The cross-correlation coefficients between the three components of the design time histories are as follows:
12 = 0.05, 23 = 0.043, and 31 = 0.140 where 1, 2, 3 are the three global directions.
Since the three coefficients are less than 0.16 as recommended in Reference 30, which was referenced by NRC Regulatory Guide 1.92, Revision 1, it is concluded that these three components are statistically independent. The design time histories are applied at the foundation level in the free field.
The ground motion time histories (H1, H2, and V) are generated with time step size of 0.010 second for applications in soil structure interaction analyses. For applications in the fixed-base mode superposition time-history analyses, the time step size is reduced to 0.005 second by linear interpolation. The maximum frequency of interest in the horizontal and vertical seismic analysis of the nuclear island is 33 hertz. Modes with higher frequencies are included in the analysis so that the mass in these higher modes is included in the member forces. The cutoff frequencies used in the soil structure interaction analyses are 33 hertz. The maximum "cut-off" frequency for the soil structure interaction analyses and the fixed-base analyses is well within the Nyquist frequency limit.
The comparison plots of the acceleration response spectra of the time histories versus the design response spectra for 2, 3, 4, 5, and 7 percent critical damping are shown in Figures 3.7.1-6, 3.7.1-7, and 3.7.1-8. The SRP 3.7.1, Table 3.7.1-1, provision of frequency intervals is used in the computation of these response spectra.
In SRP 3.7.1 the NRC introduced the requirement of minimum power spectral density to prevent the design ground acceleration time histories from having a deficiency of power over any frequency range. SRP 3.7.1, Revision 2, specifies that the use of a single time history is justified by satisfying a target power spectral density (PSD) requirement in addition to the design response spectra enveloping requirements. Furthermore, it specifies that when spectra other than Regulatory Guide 1.60 spectra are used, a compatible power spectral density shall be developed using procedures outlined in NUREG/CR-5347 (Reference 29).
The NUREG/CR-5347 procedures involve ad hoc hybridization of two earlier power spectral density envelopes. Since the modification to the RG 1.60 design spectra adopted for AP1000 (see Subsection 3.7.1.1) is relatively small (compared to the uncertainty in the fit to RG 1.60 of power spectral density-compatible time histories referenced in NUREG/CR-5347) and occurs only in the frequency range between 9 to 33 hertz, a project-specific power spectral density is developed using a slightly different hybridization for the higher frequencies.
Since the original RG 1.60 spectrum and the project-specific modified RG 1.60 spectrum are identical for frequencies less than 9 hertz, no modification to the power spectral density is done in this frequency range. At frequencies above 9 hertz, the third and the fourth legs of the power spectral density are slightly modified as follows:
The frequency at which the design response spectrum inflected towards a 1.0 amplification factor at 33 hertz takes place at 25 hertz in the AP1000 spectrum rather than at 9 hertz as in the RG 1.60 spectrum. The third leg of the power spectral density, therefore, is extended to about 25 hertz rather than 16 hertz.
The lead coefficient to the fourth leg of the power spectral density is changed to connect with the extended third leg.
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VEGP 3&4 - UFSAR The AP1000 augmented power spectral density, anchored to 0.3 g, is as follows:
S0(f) = 58.5 (f/2.5)0.2 in2/sec3, f 2.5 hertz S0(f) = 58.5 (2.5/f)1.8 in2/sec3, 2.5 hertz f 9 hertz S0(f) = 5.832 (9/f)3 in2/sec3, 9 hertz f 25 hertz S0(f) = 0.27 (25/f)8 in2/sec3, 25 hertz f The AP1000 Minimum Power Spectral Density is presented in Figure 3.7.1-9. This AP1000 target power spectral density is compatible with the AP1000 horizontal design response spectra and envelops a target power spectral density compatible with the AP1000 vertical design response spectra. This AP1000 target power spectral density, therefore, is conservatively applied to the vertical response spectra.
The comparison plots of the power spectral density curve of the AP1000 acceleration time histories versus the target power spectral density curve are presented in Figures 3.7.1-10, 3.7.1-11, and 3.7.1-12. The power spectral density functions of the design time histories are calculated at uniform frequency steps of 0.0489 hertz. The power spectral densities presented in Figures 3.7.1-10 through 3.7.1-12 are the averaged power spectral density obtained over a moving frequency band of
+/-20 percent centered at each frequency. The power spectral density amplitude at frequency (f) has the averaged power spectral density amplitude between the frequency range of 0.8 f and 1.2 f as stated in appendix A of Revision 2 of SRP 3.7.1.
3.7.1.3 Critical Damping Values Energy dissipation within a structural system is represented by equivalent viscous dampers in the mathematical model. The damping coefficients used are based on the material, load conditions, and type of construction used in the structural system. The safe shutdown earthquake damping values used in the dynamic analysis of various structures, supports, and equipment are presented in Table 3.7.1-1. The damping values are based on Regulatory Guide 1.61, Revision 0, ASCE Standard 4-98 (Reference 3), except for the damping value of the primary coolant loop piping, which is based on Reference 22, and conduits, cable trays and their related supports.
The damping values for conduits, cable trays and their related supports are shown in Table 3.7.1-1.
The damping value of conduit, empty cable trays, and their related supports is similar to that of a bolted structure, namely 7 percent of critical. The damping value of filled cable trays and supports increases with increased cable fill and level of seismic excitation. Full cable trays use a 10-percent damping value consistent with RG 1.61, Revision 1. The limiting condition for design of the AP1000 standard cable tray supports is for full cable tray weight.
For structures or components composed of different material types, the composite modal damping is calculated using the stiffness-weighted method based on Reference 3. The modal damping values equal:
nc
{n}T i [K t ]i {n}
n =
i =1 {n}T [K t ] {n}
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n
= ratio of critical damping for mode n nc = number of elements
{n} = mode n (eigenvector)
[K t ]i = stiffness matrix of element i i
= ratio of critical damping associated with element i
[K t ] = total system stiffness matrix The linear structural damping values were defined in the modeling codes as a parameter of material property defined for each element. This form of structural damping is used for seismic time history analyses. The structural models analyzed follow the damping criteria stated in Table 3.7.1-1 using 5-percent SSE damping for steel composite (SC) structures, including the shield building wall and modules, and 7-percent SSE damping for the remaining reinforced concrete (RC) structures throughout the nuclear island. A time history non-linear analysis confirms only minor cracking in the nuclear island structure.
3.7.1.4 Supporting Media for Seismic Category I Structures The supporting media will be described consistent with the information items in Subsection 2.5.4.
Seismic analyses for both rock and soil sites are described in Subsection 3.7.2 and Appendix 3G.
The AP1000 nuclear island consists of three seismic Category I structures founded on a common basemat. The three structures that make up the nuclear island are the coupled auxiliary and shield buildings, the steel containment vessel, and the containment internal structures. [The nuclear island is shown in Figure 3.7.1-14.]* The foundation embedment depth, foundation size, and total height of the seismic Category I structures are presented in Table 3.7.1-2.
For the design of seismic Category I structures, a set of six design soil profiles (that include hard rock) of various shear wave velocities is established from parametric studies as described in Appendix 3G.3. The soil cases selected for the AP1000 use parameters from the AP600 design, and the AP600 conclusions are applicable to the AP1000 due to the identical footprint to the AP600 and the similarity in overall mass.
For the AP1000 2D and 3D soil-structure interaction analyses, although some of the parabolic soil profiles are defined using a depth of 240 feet, the actual soil profile defined in SASSI (System for Analysis of Soil-Structure Interaction) (base rock) goes to only elevation 120.
Soil-structure interaction analyses on soil sites for the AP1000 used the latest soil degradation curves recommended by EPRI TR-102293, although these represent more recent soils data and differ slightly for those used for the AP600.
These six profiles are sufficient to envelope sites where the shear wave velocity of the supporting medium at the foundation level exceeds 1000 feet per second (see Subsection 2.5.2). The design soil profiles include a hard rock site, a soft rock site, a firm rock site, an upper bound soft-to-medium soil site, a soft-to-medium soil site, and a soft soil site. The shear wave velocity profiles and related governing parameters of the six sites considered are as follows:
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR For the hard rock site, an upper bound case for rock sites using a shear wave velocity of 8000 feet per second.
For the firm rock site, a shear wave velocity of 3500 feet per second to a depth of 120 feet and base rock at the depth of 120 feet.
For the soft rock site, a shear wave velocity of 2400 feet per second at the ground surface, increasing linearly to 3200 feet per second at a depth of 240 feet, and base rock at the depth of 120 feet.
For the upper bound soft-to-medium soil site, a shear wave velocity of 1414 feet per second at ground surface, increasing parabolically to 3394 feet per second at 240 feet, base rock at the depth of 120 feet, and ground water at grade level. The initial soil shear modulus profile is twice that of the soft-to-medium soil site.
For the soft-to-medium soil site, a shear wave velocity of 1000 feet per second at ground surface, increasing parabolically to 2400 feet per second at 240 feet, base rock at the depth of 120 feet, and ground water is assumed at grade level.
For the soft soil site, a shear wave velocity of 1000 feet per second at ground surface, increasing linearly to 1200 feet per second at 240 feet, base rock at the depth of 120 feet, and ground water is assumed at grade level The strain-dependent shear modulus curves for the foundation materials, together with the corresponding damping curves are taken from References 37 and 38 and are shown in Figures 3.7.1-15 and 3.7.1-16 for rock material and soil material respectively. The different curves for soil in Figure 3.7.1-16 apply to the range of depth within a soil column below grade. The strain-dependent soil material damping is limited to 15 percent of critical damping. The strain-dependent properties used in the SSI analyses for the safe shutdown earthquake are shown in Table 3.7.1-4 and Figure 3.7.1-17 for the firm rock, soft rock, upper bound soft-to-medium soil, soft-to-medium soil, and soft soil properties.
Some variation of soil modeling (water table, soil layering, soil degradation model, and the like) and combinations of these have been demonstrated to have no significant effect on the seismic response of the nuclear island structures. The governing parameters obtained for the AP600 soil studies are also applicable to the AP1000. Each of the parameters deemed not significant has been analyzed.
For instance, the combination of effects of the different strain dependent soil parameters that affect the strain-iterated shear wave velocity profiles was evaluated and shown not to result in exceedances of the envelope of the generic seismic design in-structure response spectra (ISRS).
3.7.2 Seismic System Analysis Seismic Category I structures, systems, and components are classified according to Regulatory Guide 1.29. Seismic Category I building structures of AP1000 consist of the containment building (the steel containment vessel and the containment internal structures), the shield building, and the auxiliary building. These structures are founded on a common basemat and are collectively known as the nuclear island or nuclear island structures. [Key dimensions, such as thickness of the basemat, floor slabs, roofs and walls, of the seismic Category I building structures are shown in Figure 3.7.2-12.]*
Seismic systems are defined, according to SRP 3.7.2,Section II.3.a, as the seismic Category I structures that are considered in conjunction with their foundation and supporting media to form a soil-structure interaction model. The following subsections describe the seismic analyses performed
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR for the nuclear island. Other seismic Category I structures, systems, equipment, and components not designated as seismic systems (that is, heating, ventilation, and air-conditioning systems; electrical cable tray systems; piping systems) are designated as seismic subsystems. The analysis of seismic subsystems is presented in Subsection 3.7.3.
Seismic Category I building structures are on the nuclear island. Other building structures are classified nonseismic or seismic Category II. Nonseismic structures are analyzed and designed for seismic loads according to the Uniform Building Code (Reference 2) requirements for Zone 2A. The main area of the turbine building structure is analyzed and designed for seismic loads in accordance with International Building Code requirements for an earthquake magnitude equivalent to the Uniform Building Code, Zone 3. Seismic Category II building structures are designed for the safe shutdown earthquake using the same methods and design allowables as are used for seismic Category I structures. The acceptance criteria are based on ACI 349 for concrete structures, on ANSI/AISC N690 for steel structures, including the supplemental requirements described in Subsections 3.8.4.4.1 and 3.8.4.5, and on AWS D1.1-2000 for weld design, qualification, fabrication, and inspection as described in Subsections 3.8.3.2 and 3.8.4.2.
Separate seismic analyses are performed for the nuclear island for each of the six design soil profiles defined in Subsection 3.7.1.4. The analyses generate one set of in-structure responses for each of the design soil profiles. The six sets of in-structure responses are enveloped to obtain the seismic design envelope (design member forces, nodal accelerations, nodal displacements, and floor response spectra), which are used in the design and analysis of seismic Category I structures, components, and seismic subsystems.
Appendix 3G summarizes the types of models and analysis methods that are used in the seismic analyses of the nuclear island, as well as the type of results that are obtained and where they are used in the design. The seismic analyses of the nuclear island are summarized in a seismic analysis summary report. This report describes the development of the finite element models, the soil structure interaction and fixed base analyses, and the results thereof. Seismic response spectra are given in Appendix 3G for the six key locations:
Containment internal structures at reactor vessel support elevation 100.00.
Containment internal structures at operating deck elevation 134.25.
Auxiliary shield building north east corner at control room floor elevation 116.50.
Auxiliary shield building corner of fuel building roof at shield building elevation 179.19.
Auxiliary shield building roof area elevation 327.41.
Steel containment vessel near polar crane elevation 224.000.
3.7.2.1 Seismic Analysis Methods Seismic analyses of the nuclear island are performed in conformance with the criteria within SRP 3.7.2.
Seismic analyses - using response spectra analysis, the equivalent static acceleration method, the mode superposition time-history method, and the complex frequency response analysis method -
are performed for the safe shutdown earthquake to determine the seismic force distribution for use in the design of the nuclear island structures, and to develop in-structure seismic responses (accelerations, displacements, and floor response spectra) for use in the analysis and design of seismic subsystems.
3.7.2.1.1 Equivalent Static Acceleration Analysis Equivalent static analyses, using computer program ANSYS (Reference 36), are performed to obtain the seismic forces and moments required for the structural design of the steel containment vessel 3.7-8 Revision 9
VEGP 3&4 - UFSAR and the nuclear island basemat (see Subsection 3.8.2.4.1.1). Equivalent static loads are applied to the finite element models using the maximum acceleration results from the time history analyses for the six design soil profiles. Accidental torsional moments are applied as described in Subsection 3.7.2.11.
Equivalent static analyses are also performed for design of the shield building roof and radial roof beams, PCS tank, tension ring, and air inlet structure (see Subsection 3.8.4.4.1). The equivalent static loads are based on the maximum acceleration results from time history dynamic analysis of the nuclear island in Subsection 3.7.2.1.2.
3.7.2.1.2 Time-History Analysis and Complex Frequency Response Analysis Mode superposition time-history analyses using computer program ANSYS and complex frequency response analysis using computer program SASSI are performed to obtain the in-structure seismic response needed in the analysis and design of seismic subsystems. Three-dimensional finite element shell models of the nuclear island structures are used in conjunction with the design soil profiles presented in Subsection 3.7.1.4 to obtain the in-structure responses. Stick models are coupled to the shell models of the concrete structures for the containment vessel, polar crane, reactor coolant loop, pressurizer, and core makeup tanks. Three models are used. The fine (NI10) model, as described in Subsection 3G.2.2.1, is used to define the seismic response for the hard rock site. The coarse (NI20) model, as described in Subsection 3G.2.2.2, is used for the soil structure interaction (SSI) analyses and is set up in both ANSYS and SASSI. The NI05 model, as described in Appendix 3G.2.2.4, is used to develop amplified seismic response for the envelope of soil profiles presented in Subsection 3.7.1.4 for flexible regions not captured by the coarser NI20 model. The models and analyses are described in Appendix 3G.
For the hard rock site, the soil-structure interaction effect is negligible. Therefore, for the hard rock site, the nuclear island is analyzed as a fixed-base structure, using computer program ANSYS without the foundation media. The three components of earthquake (two horizontal and one vertical time histories) are applied simultaneously in the analysis. Since the NI10 finite element model of the auxiliary and shield building uses shell elements to represent the 6-foot-thick basemat, the nodes of the basemat element are at the center of the basemat (elevation 63-6). The finite element model of the containment internal structures uses solid elements, which extend down to elevation 60-6.
When the finite element models are combined and used in the time history analyses, the auxiliary building finite element model is fixed at the shell element basemat nodes (elevation 63-6) and the base of the containment internal structures is fixed at the bottom of the solid element base nodes (elevation 60-6). This difference in elevation of the base fixity is not significant since the concrete between elevations 60-6 and 63-6, below the auxiliary building, is nearly rigid. There is no lateral support due to soil or hard rock below grade. This case results in higher response than a case analyzed with full lateral support below grade.
For additional information on the method used to calculate displacement, see Appendix 3G.4.1 and Appendix 3G.4.2.
3.7.2.1.3 Response Spectrum Analysis Response spectral analysis is used for the evaluation of the nuclear island structures. Response spectrum analyses are used to perform an analysis of a particular structure or portion of structure using the procedures described in Appendix 3G.4.3.1 and Subsections 3.7.2.6, 3.7.2.7, and 3.7.3.
Seismic response spectrum analysis of the auxiliary building, shield building, and containment internal structure is performed to develop the seismic design loads for these buildings, and the loads generated include the amplified load due to flexibility and the distribution of this load to the surrounding structures.
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VEGP 3&4 - UFSAR 3.7.2.2 Natural Frequencies and Response Loads Modal analyses are performed for the shell and lumped-mass stick models of the seismic Category I structures on the nuclear island, as described in Appendix 3G. Seismic response spectra at the six key locations (Subsection 3.7.2) are given in Appendix 3G.
3.7.2.3 Procedure Used for Modeling Based on the general plant arrangement, three-dimensional, finite element models are developed for the nuclear island structures: a finite element model of the coupled shield and auxiliary buildings, a finite element model of the containment internal structures, a finite element model of the shield building roof, and an axisymmetric shell model of the steel containment vessel. These three-dimensional, finite element models provide the basis for the development of the dynamic model of the nuclear island structures.
The finite element models of the coupled shield and auxiliary buildings, and the containment internal structures are based on the gross concrete section with the modulus based on the specified compressive strength of concrete reduced by a factor of 0.8 to consider the effect of cracking as recommended in Table 6-5 of FEMA 356 (Reference 5). This 80-percent value is supported by non-linear ABAQUS analyses performed on the nuclear island finite element model. The comparison between linear and non-linear models shows that the 80-percent stiffness model response spectra enveloped the non-linear model, providing a conservative approach in terms of response spectra and maximum stresses obtained in the shield building wall.
Seismic subsystems coupled to the overall dynamic model of the nuclear island include the coupling of the reactor coolant loop model to the model of the containment internal structures, and the coupling of the polar crane model to the model of the steel containment vessel. The criteria used for decoupling seismic subsystems from the nuclear island model are according to Section II.3.b of SRP 3.7.2, Revision 2. The total mass of other major subsystems and equipment is less than one percent of the respective supporting nuclear island structures; therefore, the mass of other major subsystems and equipment is included as concentrated lumped-mass only.
Several minor (basic building configuration not modified) design changes and model improvements include the following:
Provision for heavier fuel racks in the spent fuel pool area. Fuel and rack masses are updated, and pool water volumes are modeled as lumped masses.
Changes in the annulus configuration are incorporated into the dish model, and lower shield building and upper containment internal structures basemat nodes and elements are modified for compatibility.
The core makeup tanks were added as stick models.
The polar crane model has an updated weight and updated steel containment vessel local stiffness, and now includes polar crane truck stiffness.
The seismic analysis of the water inside the PCCWST was performed for the AP600. It was concluded that the low-frequency sloshing mode is not significant to the response of the nuclear island away from the shield building roof and that this conclusion could be extended to the AP1000 design. Further analysis indicated that the sloshing mass ratio remained essentially unchanged between the AP600 and AP1000.
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VEGP 3&4 - UFSAR 3.7.2.3.1 Coupled Shield and Auxiliary Buildings and Containment Internal Structures The finite element models of the coupled shield and auxiliary buildings and the reinforced concrete portions of the containment internal structures are based on the gross concrete section with the modulus based on the specified compressive strength of concrete of contributing structural walls and slabs. The properties of the concrete-filled structural modules are computed using the combined gross concrete section and the transformed steel face plates of the structural modules. The modulus is reduced by a factor of 0.8 to consider the effect of cracking. Furthermore, the weight density of concrete plus the uniformly distributed miscellaneous dead weights are considered by adding surface mass or by adjusting the material mass density of the structural elements. An equivalent tributary slab area load of 50 pounds per square foot is considered to represent miscellaneous deadweight such as minor equipment, piping and raceways. 25 percent of the floor live load or 75 percent of the roof snow load, whichever is applicable, is considered as mass in the global seismic models.
Major equipment weights are distributed over the floor area or are included as concentrated lumped masses at the equipment locations. The major equipment supported by the containment internal structures is represented by stick models connected to the containment internal structures, and includes the reactor coolant loop, the pressurizer, and the core makeup tank. The core makeup tank model is used only in the nuclear island fine (NI10) model; the core makeup tank is represented by mass in the nuclear island coarse model (NI20). The finite element models of the coupled shield and auxiliary buildings and the containment internal structures are described in Appendix 3G. The auxiliary and shield building is modeled with shell elements and the base of the finite element model is at the middle of the basemat at elevation 63-6. The bottom of the containment and internal structures are modeled with solid elements and the base of the finite element model is at the underside of the basemat at elevation 60-6. The interface between the models is at a radius of 71-0 at the mid-surface of the shield building.
3.7.2.3.2 Steel Containment Vessel The steel containment vessel is a freestanding, cylindrical, steel shell structure with ellipsoidal upper and lower steel domes. The three-dimensional, lumped-mass stick model of the steel containment vessel is developed based on the axisymmetric shell model. Figure 3G.2-4 presents the steel containment vessel stick model. In the stick model, the properties are calculated as follows:
Members representing the cylindrical portion are based on the properties of the actual circular cross section of the containment vessel.
Members representing the bottom head are based on equivalent stiffnesses calculated from the shell of revolution analyses for static 1.0g in vertical and horizontal directions.
Shear, bending and torsional properties for members representing the top head are based on the average of the properties at the successive nodes, using the actual circular cross section.
These are the properties that affect the horizontal modes. Axial properties, which affect the vertical modes, are based on equivalent stiffnesses calculated from the shell of revolution analyses for static 1.0g in the vertical direction.
The equivalent static acceleration analyses of the containment vessel use a finite element shell model with a refined mesh in the area adjacent to the large penetrations. Comparison of this with a time history analysis for the regions immediately surrounding the large penetrations verifies that the loads from equivalent static analysis are conservative to time history using a representative study.
This method used to construct a stick model from the axisymmetric shell model of the containment vessel is verified by comparison of the natural frequencies determined from the stick model and the shell of revolution model as shown in Table 3G.2-2. The shell of revolution vertical model (n = 0 3.7-11 Revision 9
VEGP 3&4 - UFSAR harmonic) has a series of local shell modes of the top head above elevation 265 between 23 and 30 hertz. These modes are predominantly in a direction normal to the shell surface and cannot be represented by a stick model. These local modes have small contribution to the total response to a vertical earthquake as they are at a high frequency where seismic excitation is small. The only seismic Category I components attached to this portion of the top head are the water distribution weirs of the passive containment cooling system. These weirs are designed such that their fundamental frequencies are outside the 23 to 30 hertz range of the local shell modes.
Additional details of the steel containment vessel stick model are included in Appendix 3G.2.1.3.
The containment air baffle, presented in Subsection 3.8.4.1.3, is supported from the steel containment vessel at regular intervals so that a gap is maintained for airflow. It is constructed with individual panels which do not contribute to the stiffness of the containment vessel. The fundamental frequency of the baffle panels and supports is about twice the fundamental frequency of the containment vessel. The mass of the air baffle is small, equal to approximately 10 percent of the vessel plates to which it is attached. The air baffle, therefore, is assumed to have negligible interaction with the steel containment vessel. Only the mass of the air baffle is considered and added at the appropriate elevations of the steel containment vessel stick model.
The interaction between the polar crane and the containment vessel is significant and is included in the model. This polar crane model reflects the polar crane wheel assemblies. The polar crane is supported on a ring girder which is an integral part of the steel containment vessel at elevation 228-0 as shown in Figure 3.8.2-1. It is modeled as a multi-degree of freedom system attached to the steel containment shell at elevation 224 (midpoint of ring girder) as shown in Figure 3G.2-4. The polar crane is modeled as shown in Figure 3G.2-5A with five masses at the mid-height of the bridge at elevation 236-6 and one mass for the trolley. The polar crane model includes the flexibility of the crane bridge girders and truck assembly. The containment shells local flexibility is considered when combined with the global model. When fixed at the center of containment, the model shows fundamental frequencies of 4.6 hertz transverse to the bridge, 6.9 hertz vertically, and 9.4 hertz along the bridge.
[During plant operating conditions, the polar crane is parked in the plant north-south direction with the trolley located at one end near the containment shell.]* In the seismic model, the crane bridge spans in the north-south direction and the mass eccentricity of the trolley is considered by locating the mass of the trolley at the northern limit of travel of the main hook. Furthermore, the mass eccentricity of the two equipment hatches and the two personnel airlocks are considered by placing their mass at their respective center of mass as shown in Figure 3G.2-4. Any modeling change due to the as-procured crane data is resolved with the COL holder item in Subsection 3.7.5.4, Reconciliation of Seismic Analyses of Nuclear Island Structures.
3.7.2.3.3 Nuclear Island Seismic Model The nuclear island seismic models are described in Appendix 3G. The various building models are interconnected to form the overall dynamic model of the nuclear island. The mass properties of the models include all tributary mass expected to be present during plant operating conditions. This includes the dead weight of walls and slabs, weight of major equipment, and equivalent tributary slab area loads representing miscellaneous equipment, piping and raceways.
The hydrodynamic mass effect of the water within the passive containment cooling system water tank on the shield building roof, the in-containment refueling water storage tank within the containment internal structures, and the spent fuel pool in the auxiliary building is evaluated. Since the water in the PCCS tank responds at a very low frequency (sloshing) and does not affect building response, the PCCS tank water horizontal mass is reduced to exclude the low frequency water sloshing mass. The total mass of the water in the in-containment refueling water storage tank within the containment
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR internal structures, and the spent fuel pool in the auxiliary building is included in the nuclear island seismic model.
Seismic response spectra are developed at the locations of the nodes. These response spectra are grouped and enveloped to define the seismic design response spectra. The nodes associated with a specific elevation and building structure (i.e., auxiliary and shield building and containment internal structures) are grouped. For the auxiliary and shield building where the floor at the elevation of interest is rigid (i.e. frequency > 33 hertz), it is only necessary to envelop the response spectra at edge points and interior nodes at the shield wall to obtain the largest seismic response spectra because of rigid motion. The edge nodes reflect the largest rocking and translational response of the auxiliary building, and the response spectra associated with the nodes on the shield wall will reflect the shield wall dynamic response. It is not necessary to include any nodes between the shield wall and auxiliary building edge since the floor is rigid, and the response cannot be worse than those enveloped.
A refined finite element shell model of the nuclear island concrete structures is reviewed for flexible regions, which may produce amplified response spectra. This model, called the NI05 model, has a tetrahedral mesh size of approximately 5 feet by 5 feet. Each of the principal walls and floors in the auxiliary and shield building as well as the containment internal structures are reviewed. A modal analysis of the NI05 model for both auxiliary and shield building and containment internal structures is reviewed for each of these regions for the existence of out of plane modes, which are considered flexible (less than 33 hertz) with significant participating mass. The survey reveals that some regions, typically in the middle of a floor or wall, exhibit amplified behavior compared to the critical nodes at the corner and edge building locations. [These regions, which have flexible areas, are evaluated in one of two ways:
Flexible areas, which have been previously identified, have amplified response spectra developed directly from the time history analyses for the envelope of soil sites.
Flexible regions, which require a detailed analysis to obtain the amplified response spectra, use input directly from time history analysis. The NI05 finite element model is used to capture out-of-plane flexibilities that, because of mesh refinement, a more course model could not capture.
If equipment or a structure is supported at more than one elevation, then the seismic input as an envelope of multiple groups based on the support locations will be defined. Therefore, if the equipment or structure is supported on rigid and flexible floor areas the response spectra (horizontal and vertical directions) used by the analysts will be the envelope of the rigid and flexible areas that include inside and outside nodes.
If an equipment or structure is supported exclusively by a floor or wall, only that spectra will be used for design.]*
3.7.2.4 Soil-Structure Interaction Soil-structure interaction is not significant for the nuclear island founded on rock with a shear wave velocity greater than 8000 feet per second. The soil-structure interaction analyses for the firm rock and soil sites are described in Appendix 3G.
The computer program SASSI is used to perform the soil-structure interaction analysis. The SASSI model of the nuclear island is based on the NI20 Coarse Finite Element model. Soil-structure interaction analyses are performed based on the nuclear island 3D SASSI model for the three soil conditions established from the AP1000 2D SASSI analyses, in addition to soft rock and soft soil.
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR SASSI uses key frequencies to perform its transfer function calculations. For a large model, resting on a very stiff soil (hard rock), SASSI gives conservative results at high frequencies. The significant responses for AP1000 soil cases occur at less than 10 hertz so the SASSI model is adequate for use.
Analyses are performed with large solid-shell finite element models at two levels. The fine (NI10) model is used to define the seismic response for the hard rock site. The coarse (NI20) model is used for the soil-structure interaction analyses. The NI20 coarse model has fewer nodes and elements than the NI10 model. It captures the essential features of the nuclear island configuration. The nominal shell and solid element dimension is about 20 feet.
3.7.2.5 Development of Floor Response Spectra The design floor response spectra are generated according to Regulatory Guide 1.122.
Seismic floor response spectra are computed using time-history responses determined from the nuclear island seismic analyses. The time-history responses for the hard rock condition are determined from a mode superposition time history analysis using computer program ANSYS.
The time-history responses for the firm rock and soil conditions are determined from a complex frequency response analysis using computer program SASSI. Floor response spectra for damping values equal to 2, 3, 4, 5, 7, 10, and 20 percent of critical damping are computed at the required locations.
The floor response spectra for the design of subsystems and components are generated by broadening the enveloped nodal response spectra determined for the hard rock site and soil sites.
The spectral peaks are broadened by +/-15 percent to account for the variation in the structural frequencies, due to the uncertainties in parameters such as material and mass properties of the structure and soil, damping values, seismic analysis technique, and the seismic modeling technique.
Figure 3.7.2-14 shows the broadening procedure used to generate the design floor response spectra. Spectral peaks at frequencies associated with fundamental soil structure interaction frequencies are reviewed. If there is a valley between peaks due to different soil profiles and not the building modal response, then this valley is filled by extending the broadening of the lower peak horizontally until it meets the broadened upper peak.
Floor response spectra for the auxiliary building are obtained from the three-dimensional model as described in Appendix 3G. These spectra are developed for the specific location in the auxiliary building. Where spectra at a number of nodes have similar characteristics, a single set of spectra may be developed by enveloping the broadened spectra at each of the nodes.
The safe shutdown earthquake floor response spectra for 5 percent damping, at representative locations of the coupled auxiliary and shield buildings, the steel containment vessel, and the containment internal structures are presented in Appendix 3G.
3.7.2.6 Three Components of Earthquake Motion Seismic system analyses are performed considering the simultaneous occurrences of the two horizontal and the vertical components of earthquake.
In mode superposition time-history analyses using computer program ANSYS, the three components of earthquake are applied either simultaneously or separately. In the ANSYS analyses with the three earthquake components applied simultaneously, the effect of the three components of earthquake motion is included within the analytical procedure so that further combination is not necessary.
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VEGP 3&4 - UFSAR In analyses with the earthquake components applied separately and in the response spectrum and equivalent static analyses, the effect of the three components of earthquake motion are combined using one of the following methods:
For seismic analyses with the statistically independent earthquake components applied separately, the time-history responses from the three earthquake components are combined algebraically at each time step to obtain the combined response time-history. This method is used in the SASSI analyses.
The peak responses due to the three earthquake components from the response spectrum and equivalent static analyses are combined using the square root of the sum of squares (SRSS) method.
The peak responses due to the three earthquake components from the equivalent static analyses are combined directly, using the assumption that when the peak response from one component occurs, the responses from the other two components are 40 percent of the peak (100 percent-40 percent-40 percent method). Combinations of seismic responses from the three earthquake components, together with variations in sign (plus or minus), are considered. This method is used in the nuclear island basemat analyses, the containment vessel analyses, and the shield building roof analyses.
The containment vessel is analyzed using axisymmetric finite element models. These axisymmetric building structures are analyzed for one horizontal seismic input from any horizontal direction and one vertical earthquake component. Responses are combined by either the square root of the sum of squares method or by the 100 percent-40 percent-40 percent method in which one component is taken at 100 percent of its maximum value and the other components are taken at 40 percent of their maximum value.
For the seismic responses presented in Appendix 3G, the effect of three components of earthquake are considered as follows:
Mode Superposition Time History Analysis (program ANSYS) and the Complex Frequency Response Analysis (program SASSI) - the time history responses from the three components of earthquake motion are combined algebraically at each time step.
A summary of the dynamic analyses performed and the combination techniques used are presented in Appendix 3G.
3.7.2.7 Combination of Modal Responses The modal responses of the response spectrum system structural analysis are combined using the procedures described in Appendix 3G.4.3. In the fixed base mode superposition time history analysis of the hard rock site, the total seismic response is obtained by superposing the modal responses within the analytical procedure so that further combination is not necessary.
A summary of the dynamic analyses performed and the combination techniques used are presented in Appendix 3G.
3.7.2.8 Interaction of Seismic Category II and Nonseismic Structures with Seismic Category I Structures, Systems, or Components Nonseismic structures are evaluated to determine that their seismic response does not preclude the safety functions of seismic Category I structures, systems or components. This is accomplished by satisfying one of the following:
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VEGP 3&4 - UFSAR The collapse of the nonseismic structure will not cause the nonseismic structure to strike a seismic Category I structure, system or component.
The collapse of the nonseismic structure will not impair the integrity of seismic Category I structures, systems or components.
The structure is classified as seismic Category II and is analyzed and designed to prevent its collapse under the safe shutdown earthquake.
The structures adjacent to the nuclear island are the annex building, the radwaste building, and the turbine building.
3.7.2.8.1 Annex Building The portion of the annex building adjacent to the nuclear island is classified as seismic Category II.
The structural configuration is shown in Figure 3.7.2-19. The annex building is analyzed for the safe shutdown earthquake for the six soil profiles described in Subsection 3.7.1.4. For the hard rock site, a range of soil properties is assumed for the layer above rock at the level of the nuclear island foundation. Seismic input is defined by response spectra applied at the base of a dynamic model of the annex building. The seismic response spectra input at the base of the annex building are the envelopes of the range of soil sites and also envelope the AP1000 design free field ground spectra shown in Figures 3.7.1-1 and 3.7.1-2. The envelope of the maximum building response acceleration values is applied as equivalent static loads to a more detailed static model. See Subsection 3.7.2.8.4 for more discussion of modeling and seismic analysis.
The minimum space required between the annex building and the nuclear island to avoid contact is obtained by absolute summation of the deflections of each structure obtained from either a time history or a response spectrum analysis for each structure. The maximum displacement of the roof of the annex building is 1.6 inches in the east-west direction. The minimum clearance between the structural elements of the annex building above grade and the nuclear island is 3 inches; except that the north-south minimum clearance at Unit 3 between elevations 141'-0 and 154'-0 between structural elements of the annex building and the nuclear island west of column line I is 2-1/16 inches.
3.7.2.8.2 Radwaste Building The radwaste building is classified as nonseismic and is designed to the seismic requirements of the Uniform Building Code, Zone 2A with an Importance Factor of 1.25. As shown in the radwaste building general arrangement in Figure 1.2-22, it is a small steel framed building. If it were to impact the nuclear island or collapse in the safe shutdown earthquake, it would not impair the integrity of the reinforced concrete nuclear island. The gap between the structural elements of the radwaste building above grade and the nuclear island is 4 inches.
Three methods are used to demonstrate that a potential radwaste building impact on the nuclear island during a seismic event will not impair its structural integrity:
The maximum kinetic energy of the impact during a seismic event considers the maximum radwaste building and nuclear island velocities. The total kinetic energy is considered to be absorbed by the nuclear island and converted to strain energy. The deflection of the nuclear island is less than 0.2. The shear forces in the nuclear island walls are less than the ultimate shear strength based on a minus one standard deviation of test data.
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VEGP 3&4 - UFSAR Stress wave evaluation shows that the stress wave resulting from the impact of the radwaste building on the nuclear island has a maximum compressive stress less than the concrete compressive strength.
An energy comparison shows that the kinetic energy of the radwaste building is less than the kinetic energy of tornado missiles for which the exterior walls of the nuclear island are designed.
3.7.2.8.3 Turbine Building The south end of the turbine building is separated from the rest of the turbine building by a 2'-0" thick reinforced concrete wall that provides a robust structure around the first bay. This wall isolates the first bay of the turbine building from the general area of the turbine building and from the adjacent yard area. The main segment of this wall is located on column line 11.2. This wall extends from El.100'-0" basemat to El.169'-0". The first bay of the turbine building is classified as seismic Category II. The other bays are classified as non-seismic. The modeling used to represent the structural configuration in the soil structure interaction analysis is shown in Figure 3.7.2-21.
The first bay of the turbine building is analyzed for the safe shutdown earthquake for the six soil profiles described in Subsection 3.7.1.4. For the hard rock site, a range of soil properties is assumed for the layer above rock at the level of the nuclear island foundation. Seismic input is defined by response spectra applied at the base of a dynamic model of the first bay of the turbine building. The seismic response spectra input at the base of the first bay of the turbine building are the envelopes of the range of soil sites and also envelope the AP1000 design free field ground spectra shown in Figures 3.7.1-1 and 3.7.1-2. See Subsection 3.7.2.8.4 for more discussion of modeling and seismic analysis.
The first bay is designed in accordance with ACI-349 for concrete features, ANSI/AISC N690 for steel features, including the supplemental requirements described in Subsections 3.8.4.4.1 and 3.8.4.5, and AWS D1.1-2000 for weld design, qualification, fabrication, and inspection as described in Subsections 3.8.3.2 and 3.8.4.2.
For the non-seismic portion of the Turbine Building, seismic design is upgraded in order to provide margin against collapse during the safe shutdown earthquake. The turbine building is a braced steel frame structure designed to meet the following criteria:
The turbine building is designed in accordance with the 2006 International Building Code (Reference 40). This references ACI-318 for concrete structures and AISC for steel structures. Seismic loads are defined in accordance with the International Building Code with the maximum considered earthquake spectral parameters SDS = 0.9, SD1 = 0.54 for Site Class D. This is consistent with the 1997 Uniform Building Code provisions for Zone 3 with an Importance Factor of 1.0. For a braced structure that has a mix of eccentric and special concentric bracing, the response modification factor is 6 (ASCE 7-05, Reference 42) using strength design.
The design complies with the seismic requirements for eccentrically braced frames and special concentrically braced frames given in the 2005 AISC Seismic Provisions for Structural Steel Buildings (Reference 41). Quality assurance is in accordance with ASCE 7-05 (Reference 42).
3.7.2.8.4 Seismic Modeling and Analysis of Seismic Category II Building Structures Seismic Category II structures, systems, and components are designed so that the safe shutdown earthquake does not cause unacceptable structural failure or interaction with seismic Category I 3.7-17 Revision 9
VEGP 3&4 - UFSAR items. Therefore, the seismic response of seismic Category II buildings must be obtained so that they can be designed to meet the seismic Category II requirements as given in Subsection 3.2.1.1.2.
Seismic Category II structures are analyzed and evaluated in the same manner as seismic Category I structures. The foundation of the non-seismic portion is modeled with the associated mass distributed on it so that the soil structure interaction during a seismic event is reflected in the analysis.
The seismic analyses performed for the adjacent seismic Category II structures are simulated 3D analyses. The seismic analyses are performed primarily using 2D SASSI models. To properly account for the 3D effect, the response from 2D and 3D SASSI analyses of the seismic Category II buildings on rigid foundations are compared and a 3D effects factor is developed from this comparison. Three soil cases (upper bound soft to medium UBSM, soft to medium SM, and soft soil SS) are used to determine the 3D factor. Shown in Figures 3.7.2-20 and 3.7.2-21 are the 2D SASSI models with adjacent building structures. The seismic Category II buildings are modeled as stick models. The 3D model with adjacent structures is shown in Figure 3.7.2-22.
Seismic Category II buildings are designed using envelope foundation input response spectra (FIRS). The development of these FIRS shall be based on a number of analyses results from the SASSI analyses. The seismic Category II FIRS shall be the envelope of the SASSI seismic Category II foundation response spectra resulting from the following seismic inputs/soil profiles:
AP1000 CSDRS - Hard rock at El. 60.5.
AP1000 CSDRS - Firm rock, soft rock, upper bound soft to medium, soft to medium, and soft soil, soil profiles with AP1000 CSDRS spectra input at plant grade; and AP1000 hard rock high frequency (HRHF) - For rock sites, HRHF at plant grade shall be developed using AP1000 HRHF spectra at El. 60.5 and a range of backfill soil profiles. The backfill soil under the annex and turbine buildings has a parabolic soil profile as a function of depth (El. 100 to El. 60.5) and uses EPRI (1993) strain dependent curves. The HRHF at plant grade spectrum shall be generated using soil profiles corresponding to a shear wave velocity of 500 fps, 750 fps, and 1000 fps at El. 100. The HRHF at plant grade shall be used as input to SASSI analyses to determine the FIRS at the base of the seismic Category II structures.
For each soil case, 2D SASSI analyses shall be performed and the results at three locations at the base of the seismic Category II structures are enveloped. The maximum bearing demand and maximum relative displacement shall be established from the 2D SASSI analyses. The 3D effect factor is applied to the envelope foundation spectra and used for the design of the annex building and turbine building first bay.
Response spectrum analyses (using detailed finite element building models) shall be used to obtain seismic design loads for the seismic Category II building design. The seismic input to the response spectrum analyses is the envelope foundation response spectra obtained from the SASSI analyses.
The COL applicant will perform the following screening criteria to determine if it has to perform further analysis for its site. If the requirements given below are not met, then the site applicant can perform site-specific analyses to demonstrate that its site-specific seismic Category II foundation seismic response spectra are less than the AP1000 annex building and turbine building first bay generic design envelope foundation spectra.
The site meets Subsection 2.5.4.2 DCD soil uniformity requirements.
For soil sites, the site GMRS is enveloped by the AP1000 CSDRS with soil profiles SS, SM, UBSM, SR, FR, and HR.
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VEGP 3&4 - UFSAR For HRHF sites, the site GMRS is enveloped by the AP1000 HRHF response spectra with a minimum backfill surface shear wave velocity of 500 fps, and a minimum lateral extent of the backfill corresponding to a line extending down from the surface at a one horizontal to one vertical (1H:1V) slope from the outside footprint limit of the seismic Category II structure.
The bearing capacity with appropriate factor of safety is greater than or equal to the bearing demand.
3.7.2.9 Effects of Parameter Variations on Floor Response Spectra Seismic model uncertainties due to, among other things, uncertainties in material properties, mass properties, damping values, the effect of concrete cracking, and the modeling techniques are accounted for in the widening of floor response spectra, as described in Subsection 3.7.2.5. The effect of cracking of the concrete-filled structural modules inside containment due to thermal loads is discussed in Subsection 3.8.3.4.2.
3.7.2.10 Use of Constant Vertical Static Factors The vertical component of the safe shutdown earthquake is considered to occur simultaneously with the two horizontal components in the seismic analyses. Therefore, constant vertical static factors are not used for the design of seismic Category I structures.
3.7.2.11 Method Used to Account for Torsional Effects The seismic analysis models of the nuclear island incorporate the mass and stiffness eccentricities of the seismic Category I structures and the torsional degrees of freedom.
For the response spectrum analysis of the nuclear island, the seismic loads are combined by means of the square root of the sum of the squares (SRSS). The equation for SRSS is shown below.
( A NS )2 + ( A EW )2 + (A VT )2
AEW maximum element forces due to SSE response analysis in Y (EW)
AVT maximum element forces due to SSE response analysis in Z (VT) factor to account for accidental torsion effect in NS or EW (1.05)
Alternatively, for equivalent static analysis, the 100-40-40 rule is applicable in order to cover both negative and positive member forces. The equation for the 100-40-40 rule is shown below.
k 1 sign(A NS ) ( A NS ) + k 2 sign(A EW ) ( A EW ) + k 3 A VT
- where, ki combination factors (+/-1.0, +/-0.4, +/-0.4) 3.7-19 Revision 9
VEGP 3&4 - UFSAR sign(X) sign of variable X: X < 0 results -1; X 0 results +1 factor to account for accidental torsion effect in NS or EW (1.05) 3.7.2.12 Methods for Seismic Analysis of Dams Seismic analysis of dams is site specific design.
The evaluation of existing dams whose failure could affect the site interface flood level specified in Subsection 2.4.12, is included in Subsection 2.4.4. As discussed in Subsection 2.4.1.2.4, the U.S.
Army Corps of Engineers has no current plans for the construction of additional reservoirs on the Savannah River.
3.7.2.13 Determination of Seismic Category I Structure Overturning Moments Subsection 3.8.5.5.4 describes the effects of seismic overturning moments.
3.7.2.14 Analysis Procedure for Damping Subsection 3.7.1.3 presents the damping values used in the seismic analyses. [For structures comprised of different material types, the composite modal damping approach utilizing the strain energy method is used to determine the composite modal damping values.]* Subsection 3.7.2.4 presents the damping values used in the soil-structure interaction analysis.
3.7.3 Seismic Subsystem Analysis This subsection describes the seismic analysis methodology for subsystems, which are those structures and components that do not have an interface with the soil-structure interaction analyses.
Structures and components considered as subsystems include the following:
Structures, such as floor slabs, walls, miscellaneous steel platforms and framing Equipment modules consisting of components, piping, supports, and structural frames Equipment including vessels, tanks, heat exchangers, valves, and instrumentation Distributive systems including piping and supports, electrical cable trays and supports, HVAC ductwork and supports, instrumentation tubing and supports, and conduits and supports Subsection 3.9.2 describes dynamic analysis methods for the reactor internals. Subsection 3.9.3 describes dynamic analysis methods for the primary coolant loop support system. Subsection 3.7.2 describes the analysis methods for seismic systems, which are those structures and components that are considered with the foundation and supporting media. Section 3.2 includes the seismic classification of building structures, systems, and components.
3.7.3.1 Seismic Analysis Methods The methods used for seismic analysis of subsystems include, modal response spectrum analysis, time-history analysis, and equivalent static analysis. The methods described in this subsection are acceptable for any subsystem. The particular method used is selected by the designer based on its appropriateness for the specific item. Items analyzed by each method are identified in the descriptions of each method in the following paragraphs.
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR 3.7.3.2 Determination of Number of Earthquake Cycles Seismic Category I structures, systems, and components are evaluated for one occurrence of the safe shutdown earthquake (SSE). In addition, subsystems sensitive to fatigue are evaluated for cyclic motion due to earthquakes smaller than the safe shutdown earthquake. Using analysis methods, these effects are considered by inclusion of seismic events with an amplitude not less than one-third of the safe shutdown earthquake amplitude. The number of cycles is calculated based on IEEE-344-1987 (Reference 16) to provide the equivalent fatigue damage of two full safe shutdown earthquake events with 10 high-stress cycles per event. Typically, there are five seismic events with an amplitude equal to one-third of the safe shutdown earthquake response. Each of the one-third safe shutdown earthquake events has 63 high-stress cycles. [For ASME Class 1 piping, the fatigue evaluation is performed based on five seismic events with an amplitude equal to one-third of the safe shutdown earthquake response. Each event has 63 high-stress cycles.]*
When seismic qualification is based on dynamic testing for structures, systems, or components containing mechanisms that must change position in order to function, operability testing is performed for the safe shutdown earthquake preceded by one or more earthquakes. The number of preceding earthquakes is calculated based on IEEE-344-1987 (Reference 16) to provide the equivalent fatigue damage of one safe shutdown earthquake event. Typically, the preceding earthquake is one safe shutdown earthquake event or five one-half safe shutdown earthquake events.
3.7.3.3 Procedure Used for Modeling The dynamic analysis of any complex system requires the discretization of its mass and elastic properties. This is accomplished by concentrating the mass of the system at distinct characteristic points or nodes, and interconnecting them by a network of elastic springs representing the stiffness properties of the systems. The stiffness properties are computed either by hand calculations for simple systems or by finite element methods for more complex systems.
Nodes are located at mass concentrations and at additional points within the system. They are selected in such a way as to provide an adequate representation of the mass distribution and high-stress concentration points of the system.
At each node, degrees of freedom corresponding to translations along three orthogonal axes, and rotations about these axes are assigned. The number of degrees of freedom is reduced by the number of constraints, where applicable. For equipment qualification, reduced degrees of freedom are acceptable provided that the analysis adequately and conservatively predicts the response of the equipment.
The size of the model is reviewed so that a sufficient number of masses or degrees of freedom are used to compute the response of the system. A model is considered adequate provided that additional degrees of freedom do not result in more than a 10 percent increase in response, or the number of degrees of freedom equals or exceeds twice the number of modes with frequencies less than 33 hertz.
Dynamic models of floor and roof slabs and miscellaneous steel platforms and framing include masses equal to 25 percent of the floor live load or 75 percent on the roof snow load, whichever is applicable.
Dynamic models are prepared for the following seismic Category I steel structures. Response spectrum or time history analyses are performed for structural design.
Passive containment cooling valve room (room number 12701)
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR Steel framing around steam generators Containment air baffle Seismic input for the subsystem and component design are the enveloped floor response spectra described in Subsection 3.7.2.5 or the response time histories as described in Subsection 3.7.2.1.
Where amplified response spectra are required on the subsystem for design of components, such as for use in the decoupled analyses of piping or components described in Subsection 3.7.3.8.3, the amplified response spectra are generated and enveloped as described in Subsection 3.7.2.5.
3.7.3.4 Basis for Selection of Frequencies The effect of the building amplification on equipment and components is addressed by the floor response spectra method or by a coupled analysis of the building and equipment. Certain components are designed for a natural frequency greater than 33 hertz. In those cases where it is practical to avoid resonance, the fundamental frequencies of components and equipment are selected to be less than one-half or more than twice the dominant frequencies of the support structure.
3.7.3.5 Equivalent Static Load Method of Analysis
[The equivalent static load method involves equivalent horizontal and vertical static forces applied at the center of gravity of various masses. The equivalent force at a mass location is computed as the product of the mass and the seismic acceleration value applicable to that mass location. Loads, stresses, or deflections, obtained using the equivalent static load method, are adjusted to account for the relative motion between points of support when significant.]*
3.7.3.5.1 Single Mode Dominant or Rigid Structures or Components For rigid structures and components, or for cases where the response can be classified as single mode dominant, the following procedures are used. Examples of these systems, structures, and components are equipment, and piping lines, instrumentation tubing, cable trays, HVAC, and floor beams modeled on a span by span basis.
For rigid systems, structures, and components (fundamental frequency 33 hertz), an equivalent seismic load is defined for the direction of excitation as the product of the component mass and the zero period acceleration value obtained from the applicable floor response spectra.
A rigid component (fundamental frequency 33 hertz), whose support can be represented by a flexible spring, can be modelled as a single degree of freedom model in the direction of excitation (horizontal or vertical directions). The equivalent static seismic load for the direction of excitation is defined as the product of the component mass and the seismic acceleration value at the natural frequency from the applicable floor response spectra. If the frequency is not determined, the peak acceleration from the applicable floor response spectrum is used.
[ If the component has a distributed mass whose dynamic response will be single mode dominant, the equivalent static seismic load for the direction of excitation is defined as the product of the component mass and the seismic acceleration value at the component natural frequency from the applicable floor response spectra times a factor of 1.5. A factor of less than 1.5 may be used if justified. Static factors smaller than 1.5 are not used for piping systems.]* A factor of 1.0 is used for structures or equipment that can be represented as uniformly loaded cantilever, simply supported, fixed-simply supported, or fixed-fixed beams
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VEGP 3&4 - UFSAR (References 10 and 11) when the fundamental frequency is higher than the peak acceleration frequency associated with the applicable floor response spectrum. If the frequency is not determined, the peak acceleration from the applicable floor response spectrum is used.
3.7.3.5.2 Multiple Mode Dominant Response This procedure applies to piping, instrumentation tubing, cable trays, and HVAC that are multiple span models. The equivalent static load method of analysis can be used for design of piping systems, instrumentation and supports that have significant responses at several vibrational frequencies. In this case, [a static load factor of 1.5 is applied to the peak accelerations of the applicable floor response spectra. For runs with axial supports which are rigid in the axial direction (fundamental frequency greater than or equal to 33 hertz), the acceleration value of the mass of piping in its axial direction may be limited to 1.0 times its calculated spectral acceleration value. The spectral acceleration value is based on the frequency of the piping system along the axial direction. The relative motion between support points is also considered.]*
3.7.3.6 Three Components of Earthquake Motion
[Two horizontal components and one vertical component of seismic response spectra are employed as input to a modal response spectrum analysis.]* The spectra are associated with the safe shutdown earthquake. In the response spectrum and equivalent static analyses, the effects of the three components of earthquake motion are combined using one of the following methods:
[ The peak responses due to the three earthquake components from the response spectrum analyses are combined using the square root of the sum of squares (SRSS) method.
The peak responses due to the three earthquake components are combined directly, using the assumption that when the peak response from one component occurs, the responses from the other two components are 40 percent of the peak (100 percent-40 percent-40 percent method). Combinations of seismic responses from the three earthquake components, together with variations in sign (plus or minus), are considered. This method is not used for piping systems.
One set of three mutually orthogonal artificial time histories is used when time-history analyses are performed. The components of earthquake motion specified in the three directions are statistically independent and applied simultaneously. When this method is used, the responses from each of the three components of motion are combined algebraically at each time step.]*
In addition, an optional method for combining the response of the three components of earthquake motion is presented in the following paragraphs.
[The time-history safe shutdown earthquake analysis of a subsystem can be performed by simultaneously applying the displacements and rotations at the interface point(s) between the subsystem and the system. These displacements and rotations are the results obtained from a model of a larger subsystem or a system that includes a simplified representation of the subsystem. The time-history safe shutdown earthquake analysis of the system is performed by applying three mutually orthogonal and statistically independent, artificial time histories.]* Possible examples of the use of this method of seismic analysis include the following:
The subsystem analysis is a flexible floor or miscellaneous structural steel frame. The corresponding system analysis is the soil-structure interaction analysis of the nuclear island structures.
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR The subsystem analysis is the primary loop piping system and interior concrete building structure. The interface point is the top of the basemat. The corresponding system analysis is the soil-structure interaction analysis of the nuclear island structures.
The subsystem analysis is the reactor coolant pump and internal components. The interface points are the welds on the pump suction and discharge nozzles. The corresponding system analysis is the primary loop piping system and interior concrete building structure.
3.7.3.7 Combination of Modal Responses
[For the seismic response spectra analyses, the zero period acceleration cut-off frequency is 33 hertz. High frequency or rigid modes are considered using the left-out-force method or the missing mass method]* described in Subsection 3.7.3.7.1. The method to combine the low frequency modes is described in Subsection 3.7.3.7.2. [The rigid mode results in the three perpendicular directions of the seismic input are combined by the SRSS method. The resultant response of the rigid modes is combined by SRSS with the flexible mode results.]* The combination of modal responses in time history analyses of piping systems is described in Subsection 3.7.3.17. Modal responses in time history analyses of other subsystems are combined as described in Subsection 3.7.2.6.
3.7.3.7.1 Combination of High-Frequency Modes This subsection describes alternative methods of accounting for high-frequency modes (generally greater than 33 hertz) in seismic response spectrum analysis. Higher-frequency modes can be excluded from the response calculation if the change in response is less than or equal to 10 percent.
3.7.3.7.1.1 Left-Out-Force Method or Missing Mass Correction for High Frequency Modes The left-out-force method is based on the Left-Out-Force Theorem. This theorem states that for every time history load there is a frequency, fr, called the "rigid mode cutoff frequency" above which the response in modes with natural frequencies above fr will very closely resemble the applied load at each instant of time. These modes are called "rigid modes." [The left-out-force method is used in program PIPESTRESS.]*
The left-out-force vector, {Fr} , is calculated based on lower modes:
[ ]
{Fr} = 1 M e j e j T f ( t )
where:
f (t) = the applied load vector M = the mass matrix ej = the eigenvector Note that is only for all the flexible modes, not including the rigid modes.
In the response spectra analysis, the total inertia force contribution of higher modes can be interpreted as:
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{Fr} = A [M ][{r} P j e j ]
where:
Am = the maximum spectral acceleration beyond the flexible modes
[M] = the mass matrix
{r} = the influence vector or displacement vector due to unit displacement Pj
= participation factor
- Since,
[
P j = e j T [M ]{r}, {Fr} = A [ M ]{r} 1 M e j e j T ]
[In PIPESTRESS, the low frequency modes are combined by one of the Regulatory Guide 1.92 methods in the response spectrum analysis.]* For each support level, there is a pseudo-load vector or left-out-force vector in the X, Y and Z directions. These left-out-force vectors are used to generate left-out-force solutions which are multiplied by a scalar amplitude equal to a magnification factor specified by the user. This factor is usually the ZPA (zero period acceleration) of the response spectrum for the corresponding direction. The resultant low frequency responses are combined by square root of the sum of the squares with the high frequency responses (rigid modes results).
[In GAPPIPE, the results from the high frequency responses are also combined by the square root of the sum of the squares with those from the resultant loads contributed by lower modes.]* The missing mass correction for an independent support motion or multiple response spectra analysis is exactly the same as that for the single enveloped response spectrum analysis except that Am used is the envelope of all the zero period accelerations of all the independent support inputs.
3.7.3.7.1.2 SRP 3.7.2 Method
[The method described in SRP Section 3.7.2 may also be used for combination of high-frequency modes.]*
The following is the procedure for incorporating responses associated with high-frequency modes.
Step 1 Determine the modal responses only for those modes having natural frequencies less than that at which the spectral acceleration approximately returns to the zero period acceleration (33 hertz for the Regulatory Guide 1.60 response spectra). Combine such modes according to the methods discussed in Subsection 3.7.3.7.2.
Step 2 For each degree of freedom included in the dynamic analysis, determine the fraction of degree of freedom mass included in the summation of all modes included in Step 1. This fraction di for each degree of freedom is given by:
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di = C n =1 n x n,i where:
n = order of mode under consideration N = number of modes included in Step 1 n ,i
= nth natural mode of the system Cn is the participation factor given by:
(n )T [ ] (1)
Cn =
(n )T [ ] (n )
Next, determine the fraction of degree of freedom mass not included in the summation of these modes:
ei = di - ij where ij is the Kronecker delta, which is 1 if degree of freedom i is in the direction of the earthquake motion and 0 if degree of freedom i is a rotation or not in the direction of the earthquake input motion.
If, for any degree of freedom i, the absolute value of this fraction ei exceeds 0.1, the response from higher modes is included with those included in Step 1.
Step 3 Higher modes can be assumed to respond in phase with the zero period acceleration and, thus, with each other. Hence, these modes are combined algebraically, which is equivalent to pseudostatic response to the inertial forces from these higher modes excited at the zero period acceleration. The pseudostatic inertial forces associated with the summation of all higher modes for each degree of freedom i are given by:
Pi = ZPA x Mi x ei where:
Pi = force or moment to be applied by degree of freedom i Mi = mass or mass moment of inertia associated with degree of freedom i.
The subsystem is then statically analyzed for this set of pseudo static inertial forces applied to all degrees of freedom to determine the maximum responses associated with high-frequency modes not included in Step 1.
Step 4 The total combined response to high-frequency modes (Step 3) is combined by the square root of sum of the squares method with the total combined response from lower-frequency modes (Step 1) to determine the overall structural peak responses.
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VEGP 3&4 - UFSAR 3.7.3.7.2 Combination of Low-Frequency Modes This subsection describes the method for combining modal responses in the seismic response spectra analysis. [The total unidirectional seismic response for subsystems is obtained by combining the individual modal responses using the square root sum of the squares method. For subsystems having modes with closely spaced frequencies, this method is modified to include the possible effect of these modes. For piping systems, the methods in Regulatory Guide 1.92 are used for modal combinations.]* For other subsystems, the methods in Regulatory Guide 1.92 or the following alternative methods may be used. [The groups of closely spaced modes are chosen so that the differences between the frequencies of the first mode and the last mode in the group do not exceed 10 percent of the lower frequency.
Combined total response for systems having such closely spaced modal frequencies is obtained by adding to the square root sum of squares of all modes the product of the responses of the modes in each group of closely spaced modes and coupling factor.]* This can be represented mathematically as:
N S Nj-1 Nj R T2 = Ri2 + 2 R k R k i =1 j=1 k = Mj = k+1 where:
RT = total unidirectional response Ri = absolute value of response of mode i N = total number of modes considered S = number of groups of closely spaced modes
= lowest modal number associated with group j of closely spaced modes Nj
= highest modal number associated with group j of closely spaced modes k = coupling factor, defined as follows:
1 k = 1 +
( wk - w )2 (k w k + w )2
- and,
[
wk = w k 1 - (k )2 ]
1/ 2
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k = k +
w k td where:
wk = frequency of closely spaced mode k k = fraction of critical damping in closely spaced mode k td = duration of the earthquake (= 30 seconds)
[Alternatively, a more conservative grouping method can be used in the seismic response spectra analyses. The groups of closely spaced modes are chosen so that the difference between two frequencies is no greater than 10 percent.]* Therefore, N
R T 2 = R i 2 + 2 k R k R i =1 where:
l w k - w l
< 0.1 w
All other terms for the modal combination remain the same. The 10 percent grouping method is more conservative than the grouping method because the same mode can appear in more than one group.
In addition to the above methods, any of the other methods in Regulatory Guide 1.92 may be used for modal combination.
[ 3.7.3.8 Analytical Procedure for Piping This subsection describes the modeling methods and analytical procedures for piping systems.
The piping system is modeled as beam elements with lump masses connected by a network of elastic springs representing the stiffness properties of the piping system. Concentrated weights such as valves or flanges are also modeled as lump masses. The effects of torsion (including eccentric masses), bending, shear, and axial deformations, and effects due to the changes in stiffness values of curved members are accounted for in the piping dynamic model.
The lump masses are selected so that the maximum spacing is not greater than the length that would produce a natural frequency equal to the zero period acceleration (ZPA) frequency of the seismic input when calculated based on a simply supported beam. As a minimum, the number of degrees of freedom is equal to twice the number of modes with frequencies less than the zero period acceleration frequency.
The piping system analysis model includes the effect of piping support mass when the contributory mass of the support is greater than 10 percent of the total mass of the effected piping spans. The contributory mass of the support is the portion of the support mass that is attached to the piping; such as clamps, bolts, trunnions, struts, and snubbers. Supports that are not directly attached to the
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VEGP 3&4 - UFSAR piping, such as box frames, need not be considered for mass effects. The mass of the applicable support will not affect the response of the system in the supported direction, herefore only the unsupported direction needs to be considered. Based on this reasoning, the mass of full anchors can be neglected. The total mass of each effected piping span includes the mass of the piping, fluid contents, insulation, and any concentrated masses (for example, valves or flanges) between the adjacent supports in each unrestrained direction on both sides of the applicable support. For example; the contributory mass of an X direction support must be compared to the mass of the piping spans in the unrestrained Y and Z directions. A contributory support mass that is less than 10 percent of the masses of the effected spans will have insignificant effect on the response of the piping system and can be neglected.
The stiffness matrix of the piping system is calculated based on the stiffness values of the pipe elements and support elements. Default rigid or calculated support stiffness values are used (see subsections 3.9.3.1.5 and 3.9.3.4). When the support deflections are limited to 1/8 inches for the dynamic combined faulted loads, default rigid support stiffness values are used. If the dynamic combined faulted load deflection for any support exceeds 1/8 inches, calculated support stiffness values are used for the affected support.
Valves, equipment and piping modules are considered as rigid if the natural frequencies are greater than 33 hertz. Valves with lower frequencies are included in the piping system model. See subsection 3.7.3.8.2.1 for flexible equipment and subsection 3.7.3.8.3 for flexible modules.
See subsection 3.9.3.1.4 for the primary loop piping and support system.]*
3.7.3.8.1 Supporting Systems This subsection deals with the analysis of piping systems that provide support to other piping systems. The methods used for the analysis of the primary loop piping are described in Appendix 3C.
[The supported piping system may be excluded from the analysis of the supporting piping system when the ratio of the supported pipe to supporting pipe moment of inertia is less than or equal to 0.04.
If the ratio of the run piping outside diameter to the branch piping outside diameter (nominal pipe size) exceeds or equals 3.0, the branch piping can be excluded from the analysis of the run piping.
The mass and stiffness effects of the branch piping are considered as described below.
Stiffness Effect The stiffness effect of the decoupled branch pipe is considered significant when the distance from the run pipe outside diameter to the first rigid or seismic support on the decoupled branch pipe is less than or equal to one half the deadweight span of the branch pipe (given in ASME III Code Subsection NF).
Mass Effect Considering one direction at a time, the mass effect is significant when the weight of half the span (from the decoupling point) of the branch pipe in one direction is more than 20 percent the weight of the main run pipe span in the same direction. Concentrated weights in the branch pipe are considered. A branch pipe span in x direction is the span between the decoupled branch point and the first seismic or rigid support in the x direction. A main run pipe span in the x direction is the piping bounded by the first seismic or rigid support in the x direction on both sides of the decoupled branch point. Similarly, the same definition applies to the spans in other directions (y and z).
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR If the calculated branch pipe weight is less than 20 percent but more than 10 percent of the main run pipe weight, this weight is lumped at the decoupling point of the run pipe for the run pipe analysis.
This weight can be neglected if it is less than 10 percent of the main run pipe weight.
Required Coupled Branch Piping If the stiffness and/or mass effects are considered significant, the branch piping is included in the piping analysis for the run pipe analysis. The portion of branch piping considered in the analysis adequately represents the behavior of the run pipe and branch pipe. The branch line model ends in one of the following ways:
First six-way anchor Four rigid/seismic supports in each of the three perpendicular directions Rigidly supported zone as described in subsection 3.7.3.13.4.2]*
3.7.3.8.2 Supported Systems This subsection deals with the analysis of piping systems that are supported by other piping systems or by equipment.
3.7.3.8.2.1 Large Diameter Auxiliary Piping
[This subsection deals with ASME Class 1 piping larger than 1-inch nominal pipe size and ASME Class 2 and 3 piping with nominal pipe size larger than 2 inches. The response spectra methodology is used.
When the supporting system is a piping system, the supported pipe (branch) can be decoupled from the supporting pipe (run) when the ratio of the run piping nominal pipe size to branch pipe nominal pipe size is greater than or equal to three to one. Decoupling can also be done when the moment of inertia of the branch pipe is less than or equal to 4 percent of the moment of inertia of the run pipe.
During the analysis of the branch piping, resulting values of tee anchor reactions are checked against the capabilities of the tee.
The seismic inertia effects of equipment and piping that provide support to supported (branch) piping systems are considered when significant. When the frequency of the supporting equipment is less than 33 hertz, then either a coupled dynamic model of the piping and equipment is used, or the amplified response spectra at the equipment connection point is used with a decoupled model of the supported piping. When supported piping is supported by larger piping, one of the following methods is used:
A coupled dynamic model of the supported piping and the supporting piping Amplified response spectra at the connection point to the supporting piping with a decoupled model of the supported piping]*
3.7.3.8.2.2 Small-Diameter Auxiliary Piping
[This subsection deals with ASME Code Class 1 piping equal to or less than 1-inch nominal pipe size and ASME Class 2 and 3 piping with nominal pipe sizes less than or equal to 2 inches. This includes instrumentation tubing. These piping systems may be supported by equipment or primary loop piping or other auxiliary piping or both. The response spectra or equivalent static load methodology is used.
One of the following methods may be used for these systems:
Same method as described in subsection 3.7.3.8.2.1
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VEGP 3&4 - UFSAR Equivalent static analysis based on appropriate load factors applied to the response spectra acceleration values]*
Subsections 3.9.3 and 3.9.8.2 discuss the final design and as built reconciliation of small bore piping.
3.7.3.8.3 Piping Systems on Modules Many portions of the systems for the AP1000 are assembled as modules offsite and shipped to the plant as completed units. This method of construction does not result in any unique requirements for the analysis of these structures, systems, or components. Existing industry standards and regulatory requirements and guidelines are appropriate for the evaluation of structures, systems, and components included in modules.
The modules are constructed using a structural steel framework to support the equipment, pipe, and pipe supports in the module. The structural steel framework is designed as part of the building structure according to the criteria given in Subsection 3.8.4.
One exception is the pressurizer and safety relief valve module, which is attached to the top of the pressurizer. For this module the structures and piping arrangements support valves off the pressurizer and not the building structure. The structural steel frame is designed as a component support according to ASME Code,Section III, Subsection NF. [Piping in modules is routed and analyzed in the same manner as in a plant not employing modules. Piping is analyzed from anchor point to anchor point, which are not necessarily at the boundaries of the module.]* This is consistent with the manner in which room walls are treated in a nonmodule plant.
[The supported piping or component may be decoupled from the seismic analysis of the structural frame based on the following criteria. The mass ratio, Rm, and the frequency ratio, Rf, are defined as follows:
Rm = mass of supported component or piping/mass of supporting structural frame Rf = frequency of the component or piping/frequency of the structural frame Decoupling may be done when:
Rm < 0.01, for any Rf, or Rm 0.01 and 0.10, if Rf 0.8 or if Rf is 1.25.
In addition, supported piping may be decoupled if analysis shows that the effect on the structural frame is small, that is, when the change in response is less than 10 percent. When piping or components are decoupled from the analysis of the frame, the contributory mass of the piping and components is included as a rigid mass in the model of the structural frame.]*
When piping or components are decoupled from the analysis of the frame using the preceding criteria, the effect of the frame is accounted for in the analysis of the decoupled components or piping. Either an amplified response spectra or a coupled model is used. The amplified response spectra are obtained from the time history safe shutdown earthquake analysis of the frame. The coupled model consists of a simplified mass and stiffness model of the frame connected to the seismic model of the components or piping.
Alternative criteria may be applied to simple frames that behave as pipe support miscellaneous steel.
Decoupling may be done when the deflection of the frame due to dynamic combined faulted loading is less than or equal to 1/8 inch. These deflections are defined with respect to the structure to which the structural frame is attached. The stiffness of the intervening elements between the frame and the supported piping or component is considered as follows: Default rigid stiffness values are used for
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VEGP 3&4 - UFSAR supports except that vendor stiffness values are used for snubbers and rigid gapped supports. The mass of the structural frame is evaluated as a self-weight excitation loading on the frame and the structures supporting the frame. The same approach is used for pipe support miscellaneous steel, as described in Subsection 3.9.3.4.
When the supported components or piping cannot be decoupled, they are included in the analysis model of the structural frame. The interaction between the piping and the frame is incorporated by including the appropriate stiffness and mass properties of the components, piping, and frame in the coupled model.
[ 3.7.3.8.4 Piping Systems with Gapped Supports This subsection describes the analysis methods for piping systems with rigid gapped supports.
These supports may be used to minimize the number of pipe support snubbers and the corresponding inservice testing and maintenance activities.
The analysis consists of an iterative response spectra analysis of the piping and support system.
Iterations are performed to establish calculated piping displacements that are compatible with the stiffness and gap of the rigid gapped supports. The results of the computer program GAPPIPE, which uses this methodology, are supported with test data (Reference 13).
The method implemented in GAPPIPE to analyze piping systems supported by rigid gapped supports is based on the equivalent linearization technique. GAPPIPE analysis is performed whenever snubber supports are replaced by rigid gapped supports.
The basis of the concept is to find an equivalent linear spring with a response-dependent stiffness for each nonlinear rigid gapped support, or limit stop, in the mathematical model of the piping system.
The equivalent linearized stiffness minimizes the mean difference in force in the support between the equivalent spring and the corresponding original gapped support. The mean difference is estimated by an averaging process in the time domain, that is, across the response duration, using the concept of random vibration. Details of the design and analysis methods and modeling assumptions are described in Reference 12.]*
3.7.3.9 Combination of Support Responses This subsection describes alternative methods for combining the responses from the individual support or attachment points that connect the supported system or subsystem to the supporting system or subsystem. There are two aspects to the responses from the support or attachment points:
seismic anchor motions and envelope or multiple-input response spectra methodology.
Seismic Anchor Motions - The response due to differential seismic anchor motions is calculated using static analysis (without including a dynamic load factor). In this analysis, the static model is identical to the static portion of the dynamic model used to compute the seismic response due to inertial loading. In particular, the structural system supports in the static model are identical to those in the dynamic model.
[The effect of relative seismic anchor displacements is obtained either by using the worst combination of the peak displacements or by proper representation of the relative phasing characteristics associated with different support inputs. For components supported by a single concrete building (coupled shield and auxiliary buildings, or containment internal structures), the seismic motions at all elevations above the basemat are taken to be in phase. When the component supports are in the same structure, the relative seismic anchor motions are small and the effects are neglected. This is applicable to building structures and to those supplemental steel frames that are rigid in comparison to the components. Supplemental steel frames that are flexible can have
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR significant seismic anchor motions which are considered. When the components supports are in different structures, the relative seismic anchor motion between the structures is taken to be out-of-phase and the effects are considered. The results of the modal spectra analysis (multiple input or envelope) are combined with the results from seismic anchor motion by the absolute sum method or the SRSS method, as described in Tables 3.9-5 and 3.9-6.]*
Response Spectra Methods - The envelope broadened uniform-input response spectra can lead to excessive conservatism and unnecessary pipe supports. The peak shifting method and independent support motion spectra method are used to avoid unnecessary conservatism.
Seismic Response Spectra Peak Shifting The peak shifting method may be used in place of the broadened spectra method, as described below.
Determine the natural frequencies (fe)n of the system to be qualified in the broadened range of the maximum spectrum acceleration peak.
If no equipment or piping system natural frequencies exist in the +/-15 percent interval associated with the maximum spectrum acceleration peak, then the interval associated with the next highest spectrum acceleration peak is selected and used in the following procedure.
Consider all N natural frequencies in the interval fj - 0.15fj (fe)n fj + 0.15fj where:
fj = the frequency of maximum acceleration in the envelope spectra n = 1 to N The system is then evaluated by performing N + 3 separate analyses using the envelope unbroadened floor design response spectrum and the envelope unbroadened spectrum modified by shifting the frequencies associated with each of the spectral values by a factor of +0.15; -0.15; and (f e ) n f j fj where:
n = 1 to N The results of these separate seismic analyses are then enveloped to obtain the final result desired (e.g., stress, support loads, acceleration, etc.) at any given point in the system. If three different floor response spectrum curves are used to define the response in the two horizontal and the vertical directions, then the shifting of the spectral values as defined above may be applied independently to these three response spectrum curves.
Independent Support Response Spectrum Methods The use of multiple-input response spectra accounts for the phasing and interdependence characteristics of the various support points. The following alternative methods are used for the
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR AP1000 plant. These are based on the guidelines provided by the "Pressure Vessel Research Committee Technical Committee on Piping Systems" (Reference 14).
[Envelope Uniform Response Spectra - Method A - The seismic response spectrum that envelopes the supports is used in place of the spectra at each support in the envelope uniform response spectra. Also, the contribution from the seismic anchor motion of the support points is assumed to be in phase and is added algebraically as follows:
N qi = d i Pij j=1 where:
qi = combined displacement response in the normal coordinate for mode i di = maximum value of dij d ij
= displacement spectral value for mode i associated with support "j" Pij
= participation factor for mode i associated with support j N = number of support points Enveloped response spectra are developed as the seismic input in three perpendicular directions of the piping coordinate system to include the spectra at the floor elevations of the attachment points and the piping module or equipment if applicable. The mode shapes and frequencies below the cut-off frequency are calculated in the response spectrum analysis. The modal participation factors in each direction of the earthquake motion and the spectral accelerations for each significant mode are calculated. Based on the calculated mode shapes, participation factors, and spectral accelerations of individual modes, the modal inertia response forces, moments, displacements, and accelerations are calculated. For a given direction, these modal inertia responses are combined based on consideration of closely spaced modes and high frequency modes to obtain the resultant forces, moments, displacements, accelerations, and support loads. The total seismic responses are combined by square-root-sum-of-the-squares method for all three earthquake directions.
Independent Support Motion - Method B - When there are more than one supporting structure, the independent support motion (ISM) method for seismic response spectra may be used.
Each support group is considered to be in a random-phase relationship to the other support groups.
The responses caused by each support group are combined by the square-root-sum-of-the-square method. The displacement response in the modal coordinate becomes:
1/ 2 N
qi = (Pij d ij )
2 j=1 A support group is defined by supports that have the same time-history input. This usually means all supports located on the same floor (or portions of a floor) of a structure.]*
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR 3.7.3.10 Vertical Static Factors Constant static factors can be used in some cases for the design of seismic Category I subsystems and equipment. The criteria for using this method are presented in Subsection 3.7.3.5.
3.7.3.11 Torsional Effects of Eccentric Masses
[The methods used to account for the torsional effects of valves and other eccentric masses (for example, valve operators) in the seismic subsystem analyses are as follows:
When valves and other eccentric masses are considered rigid, the mass of the operator and valve body or other eccentric mass are located at their respective center of gravity. The eccentric components (that is, yoke, valve body) are modeled as rigid members.
When valves and other eccentric masses are not considered rigid, the dynamic models are simulated by the lumped masses in discrete locations (that is, center of gravity of valve body and valve operator), coupled by elastic members with properties of the eccentric components.]*
3.7.3.12 Seismic Category I Buried Piping Systems and Tunnels
[There are no seismic Category I buried piping systems and tunnels in the AP1000 design.]*
3.7.3.13 Interaction of Other Systems with Seismic Category I Systems The safety functions of seismic Category I structures, systems, and components are protected from interaction with nonseismic structures, systems, and components; or their interaction is evaluated.
The safety-related systems and components required for safe shutdown are described in Section 7.4.
This equipment is located in selected areas of the auxiliary building and inside containment. The primary means of protecting safety-related structures, systems, and components from adverse seismic interactions are discussed in the following paragraphs in the order of preference.
Separation - separation with the use of physical barriers Segregation - routing away from location of seismic Category I systems, structures, and components Impact Evaluation - contact with seismic Category I systems, structures, and components may occur, and there is insufficient energy in the impact to cause loss of safety function Support as seismic Category II
[Interaction of connected systems with seismic Category I piping is considered by including the other piping in the analysis of the seismic Category I system.]* Interaction of piping systems that are adjacent to Category I structures, systems, and components is also considered. This is discussed in Subsection 3.7.3.13.4.
The containment and each room outside containment containing safety-related systems or equipment, as identified in Table 3.7.3-1, are reviewed for potential adverse seismic interactions to demonstrate that systems, structures, and components are not prevented from performing their required safe shutdown functions. In addition, the review identifies the protection features required to mitigate the consequences of seismic interaction in an area that contains safety-related equipment.
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR The evaluation steps to address seismic interaction taken for each room or building area containing seismic Category I systems, structures, and components are:
- 1. Define targets susceptible to damage (sensitive targets);
Sensitive targets are those seismic Category I components for which adverse spatial interaction can result in loss of safety function.
- 2. Define sources which can potentially interact in an adverse manner with the target.
- 3. If possible, assure adequate free space to eliminate the possibility of seismically-induced damaging impacts for the sensitive targets.
- 4. Assess impact effects (interaction) when adequate free space is not present.
- 5. Correct adverse seismic interaction conditions.
The three-dimensional computer model and composites developed for the nuclear island are used during the design process of the systems and components in the nuclear island, to aid in evaluating and documenting the review for seismic interactions. This review is performed using the design criteria and guidelines described in Subsections 3.7.3.13.1 through 3.7.3.13.4.
The seismic interaction review is discussed in Subsection 3.7.5.3. This review is performed in parallel with the seismic margin evaluation. The review is based on as-procured data, as well as the as-constructed condition.
3.7.3.13.1 Separation and Segregation Separation - The general plant arrangement provides physical separation between the seismic Category I and nonseismic structures, systems, and components to the maximum extent practicable in the nuclear island. The objective is to assist in the preclusion of a potential adverse interaction if the nonseismic structures, systems and components were to fail during a seismic event. Whenever possible, nonseismic pipe, electrical raceway, or ductwork is not routed above or adjacent to safety-related equipment, pipe, electrical raceway systems, or ductwork, thereby eliminating the possibility of seismic interaction.
Workstations and other equipment in the Main Control Room are separated from piping. Further, as stated in Subsection 3.2.1.1.2, structures, systems, and components that are located overhead in the Main Control Room are supported as seismic Category II.
Segregation - Where separation by physical means cannot be accomplished and it becomes necessary to locate or route nonseismic structures, systems, and components in or through safety-related areas, the nonseismic structures, systems and components are segregated from the seismic Category I items to the extent practicable.
Nonseismic cabinets are separated or segregated from seismic Category I cabinets. Also, if a cabinet is a source or a target, the cabinet doors must be secured by latches or fasteners to assure they do not open during a seismic event.
3.7.3.13.2 Impact Analysis Adverse spatial interaction (i.e., loss of structural integrity or function effecting safety) can potentially occur when two items are in close proximity. Adverse spatial interaction can result from contact or impact from overturning. Seismic Category I systems, structures, and components that are sensitive to seismic interaction are identified as potential targets. Sources are structures or components that 3.7-36 Revision 9
VEGP 3&4 - UFSAR can have adverse spatial interaction with the seismic Category I systems, structures, and components. Identification and evaluation of spatial interactions includes the following considerations:
Proximity of the source to the target. That is, the location of the source within the impact evaluation zone (shown in Figure 3.7.3-1)
If a source is outside the impact evaluation zone, and does not enter this zone if overturning occurs, no adverse spatial interaction can occur with the identified target. If the source is within the impact evaluation zone and the supports of the source fail, the source could free fall, potentially impacting the target.
Robustness of target If a target has significant structural integrity, and its function is not an issue, adverse spatial interaction could not occur with the identified source.
Energy of impact The energy of the source impacting the target may be so low as not to cause adverse spatial interaction with the target.
A specific nonseismic structure, system, or component identified as a source to a specific safety-related component can be acceptable without being supported as seismic Category II, if an analysis demonstrates that the weight and configuration of the source, relative to the target, and the trajectory of the source are such that the interaction would not cause unacceptable damage to the target. For example, a nonseismic instrument tube routed above a seismic Category I electrical cable tray system would not pose a hazard and would be acceptable.
Nonseismic equipment can overturn as a result of a safe shutdown earthquake. The trajectory of its fall is evaluated to determine if it poses a potential impact hazard to a safety-related structure, system, or component. If it poses a hazard, the equipment is relocated, or it is supported as described in Subsection 3.7.3.13.3.
Nonseismic walls, platforms, stairs, ladders, grating, handrail installations, or other structures next to safety-related structures, systems, and components are evaluated to determine if their failure is credible.
Should a nonseismic structure, system, or component be capable of being dislodged from its supports, the trajectory of its fall is evaluated for potential adverse impacts. If these present a hazard, the structure, system or component is relocated or supported as described in Subsections 3.7.3.13.3 and 3.7.3.13.4. Impact is assumed for sources within an impact evaluation zone around the safety-related equipment. The impact evaluation zone is defined as the envelope around the target for which a source, if located outside of the envelope, would not impact the target during a safe shutdown earthquake in the event the supports of the source were to fail and allow the source to fall.
The impact evaluation zone is defined by the volume extending 6 feet horizontally from the perimeter of the seismic Category I object up to a height of 35 feet. The impact evaluation zone above 35 feet is defined by a 10-degree cone radiating vertically from the foot of the object, projected from its perimeter. This definition of the impact evaluation zone is illustrated in Figure 3.7.3-1. The impact evaluation zone need not extend beyond seismic Category I structures such as walls or floor slabs.
Using seismic experience data, the following seismic Category I equipment (potential targets) are not sensitive to piping, HVAC ducts, and cable tray interaction because they are robust to these types of impact:
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VEGP 3&4 - UFSAR Tanks, "heavy" equipment (for example, heat exchangers)
Mechanical or electrical penetrations Heating, ventilation, and air conditioning (HVAC)
Adjacent piping Raceway Systems Structures 3.7.3.13.3 Seismic Category II Supports Where the preceding approaches of separation, segregation, or impact analysis cannot prevent unacceptable interaction, the source is classified and supported as seismic Category II. The seismic Category II designation provides confidence that these nonseismic structures, systems, and components can withstand the forces of a safe shutdown earthquake in addition to the loading imparted on the seismic Category II supports due to failure of the remaining nonseismically supported portions. This includes nozzle loads from the nonseismic piping. Design methods and stress criteria for systems, structures, and components classified as seismic Category II are the same as for seismic Category I systems, structures, and components, except for piping which is described in Subsection 3.7.3.13.4.2. However, the functionality of these seismic Category II sources does not have to be maintained following a safe shutdown earthquake.
HVAC duct and/or cable trays within the impact evaluation zone are seismically supported using the criteria given in Appendices 3F and 3A for seismic Category I assuring that the HVAC and cable tray segments identified as a source will not fall or adversely impact the sensitive target. Adequate free space between the source and target is assured using the load combination that includes the safe shutdown earthquake. The seismic displacement of the HVAC duct and/or cable tray is 6 inches or the calculated displacement.
Nonseismic equipment identified as a source within the impact evaluation zone is supported as seismic Category II. Support seismic loads include seismic inertia loads of the equipment determined as described in Subsection 3.7.3.5 and nozzle loads from attached piping determined as described in Subsection 3.7.3.13.4.2. Adequate free space is assessed considering a 6-inch deflection envelope for equipment identified as a source, or calculated deflections obtained using the safe shutdown earthquake load combination and elastic analysis.
[ 3.7.3.13.4 Interaction of Piping with Seismic Category I Piping Systems, Structures, and Components This subsection describes the design methods for piping to prevent adverse spatial interactions.
3.7.3.13.4.1 Seismic Category I Piping The safe shutdown earthquake piping displacements obtained for the seismic Category I piping are used for the evaluation of seismic interaction with sensitive equipment. Adequate free space between a source and a target is checked adding absolutely the piping safe shutdown earthquake deflection and the safe shutdown earthquake target deflection along with the other loads (e.g., dead weight, thermal) that are in the appropriate design criteria load combinations. Sensitive equipment for piping as the source is seismic Category I equipment shown in Table 3.7.3-2 along with the portion that must be protected ("zone of protection"). Supports may be added to limit seismic movement to eliminate potential adverse interaction.
3.7.3.13.4.2 Seismic Category II Piping This subsection describes the methods and criteria for piping that is connected to seismic Category I piping. Interaction of seismic Category I piping and nonseismic Category I piping connected to it is
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR achieved by incorporating into the analysis of the seismic Category I system a length of pipe that represents the actual dynamic behavior of the complete run of the nonseismic Category I system.
The length considered is classified as seismic Category II and extends to the interface anchor or rigid support as described below.
The seismic Category II portion of the line, up to the interface anchor or interface rigid support (last seismic support), is analyzed according to Equation 9 of ASME Code,Section III, Class 3, with a stress limit equal to the smaller of 4.5 Sh and 3.0 Sy. In either case, the nonseismic piping is isolated from the seismic Category I piping by anchors or seismic supports. The anchor or seismic Category II supports are designed for loads from the nonseismic piping. This includes three plastic moment components (Mp1, Mp2, or Mp3) in each of three local coordinate directions. The responses to the three moments are evaluated independently. The seismic Category II portion of the line is analyzed by the response spectrum or equivalent static load method for safe shutdown earthquake.
Single Interface Anchor The seismic Category II piping may be terminated at a single interface anchor (six-way). This anchor and the supports on the seismic Category II piping are evaluated for safe shutdown earthquake loadings using the rules of ASME III Subsection NF. If the anchor is an equipment nozzle, then the equipment load path through the equipment supports are evaluated to the same acceptance criteria as seismic Category I equipment.
Anchor Followed by a Series of Seismic Supports The seismic Category II piping may be terminated at the last seismic support which follows a six-way anchor on the seismic Category II piping. This last seismic support and the supports on the seismic Category II piping are evaluated for safe shutdown earthquake loadings using the rules of ASME III Subsection NF. From the anchor to the last seismic support, the response to the plastic moments (Mp1, Mp2, or Mp3) is combined with the responses to seismic anchor motions and equivalent static seismic inertia of the piping system by the absolute sum method. The responses to these moments are evaluated independently. The support and anchor loads due to the plastic moments (Mp1, Mp2, or Mp3) of the seismically analyzed and supported section can be reduced if the elbow/bend resultant moments have exceeded the plastic limit moments of the elbow/bend. The value of the reduction factor RF is as follows:
RF = Multiplier used to reduce the interface anchor and support loads RF = < l, (if RF > l, no reduction is applicable)
RF = ML/Ma Ma = Resultant moment at elbow/bend. Use maximum value if several elbows/bends are within seismically supported region.
ML = 0.8h0.6 D2t Sy for h < 1.45 ML = D2t Sy for h > 1.45 h = Flexibility characteristic of elbow/bend D = Outside diameter of elbow/bend t = Thickness of elbow/bend R = Bend radius of elbow/bend
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR Rigid Region The seismic Category II piping may be terminated at the last seismic support of a rigidly supported region of the piping system. The rigid region is typically defined as either four bi-lateral supports around an elbow or six bilateral supports around a tee. The structural behavior of the rigid region is similar to that of a six-way anchor. The frequency of the piping system in the rigid region is greater than or equal to 33 hertz. This last seismic support in the rigid region and the supports on the seismic Category II piping are evaluated for safe shutdown earthquake loadings using the rules of ASME III Subsection NF.
3.7.3.13.4.3 Nonseismic Piping Nonseismic piping within the impact evaluation zone is seismically supported, thereby ensuring that the pipe segment identified as a source will not fall or adversely impact the sensitive target (Table 3.7.3-2). This situation is shown in Figure 3.7.3-2, and the seismic supported piping criteria described below:
Supports within the impact evaluation zone, plus one transverse support in each transverse direction beyond the impact evaluation zone, are classified as seismic Category II and are evaluated for the safe shutdown earthquake loading using the rules of ASME III, Subsection NF.
Piping within the impact evaluation zone plus one transverse support in each transverse direction are evaluated to Equation 9 of ASME Code,Section III, Class 3, with a stress limit equal to the smaller of 4.5 Sh and 3.0 Sy. Outside the impact evaluation zone, the nonseismic piping meets ASME B31.1 requirements.
The nonseismic piping and seismic Category II supports are designed for loads from the nonseismic piping beyond the impact evaluation zone. This includes three plastic moment components (Mp1, Mp2, or Mp3) in each of three local coordinate directions applied at the first and last seismic Category II support. The responses to the three moments are evaluated independently. The response from the moments applied at the first seismic Category II support is combined with the response from the moments applied at the last seismic Category II support and with the responses to seismic anchor motions and equivalent static seismic inertia of the piping system by the absolute sum method. The support and anchor loads due to the plastic moments (Mp1, Mp2, or Mp3) of the seismically analyzed and supported section can be reduced if the elbow/bend resultant moments have exceeded the plastic limit moments of the elbow/bend. The value of the reduction factor RF is the same as the value for connected seismic Category II piping described above.
The piping segment identified as the source has at least one effective axial support.
Adequate free space between a source and a target is checked adding absolutely the piping safe shutdown earthquake deflections (defined following seismic Category II piping analysis methodology) and the safe shutdown earthquake target deflection. Also included are the displacements associated with the appropriate load cases.
When the anchor is an equipment nozzle, the equipment is supported as seismic Category II as described in subsection 3.7.3.13.3.]*
3.7.3.14 Seismic Analyses for Reactor Internals See Subsection 3.9.2 for the dynamic analyses of reactor internals.
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR 3.7.3.15 Analysis Procedure for Damping Damping values used in the seismic analyses of subsystems are presented in Subsection 3.7.1.3.
Safe shutdown earthquake damping values used for different types of analysis are provided in Table 3.7.1-1. For subsystems that are composed of different material types, the composite modal damping approach with the weighted stiffness method is used to determine the composite modal damping value. Alternately, the minimum damping value may be used for these systems. [Composite modal damping for coupled building and piping systems is used for piping systems that are coupled to the primary coolant loop system and the interior concrete building. Composite modal damping is used for piping systems that are coupled to flexible equipment or flexible valves. Piping systems analyzed by the uniform envelope response spectra method with rigid valves can be evaluated with 5 percent damping. Five percent damping is not used in piping systems that are susceptible to stress corrosion cracking.]*
For the time history dynamic analysis and independent support motion response spectra analysis of piping systems, 4 percent, 3 percent, and 2 percent damping values are used as described in Table 3.7.1-1.
When piping systems and nonsimple module steel frames (simple frames are described in Subsection 3.7.3.8.3) are in a single coupled model, composite damping, as described in Subsection 3.7.1.3 is used.
3.7.3.16 Analysis of Seismic Category I Tanks This subsection describes the seismic analyses for the large, atmospheric seismic Category I pools and tanks. These are reinforced concrete structures with stainless steel liners or with structural modules, as discussed in Subsections 3.8.3 and 3.8.4. They include the spent fuel pool in the auxiliary building, the in-containment refueling water storage tank, and the passive containment cooling water tank incorporated into the shield building roof. There are no other seismic Category I tanks.
The seismic analyses of the tank consider the impulsive and convective forces of the water as well as the flexibility of the walls. For the spent fuel pool, cask loading pit, cask washdown pit and fuel transfer canal, the impulsive loads are calculated by considering a portion of the water mass responding with the concrete walls. The impulsive forces are calculated by conventional methods for rigid tanks. The passive containment cooling water tank is analyzed using methods described in Reference 15 for toroidal tanks. It is also analyzed by finite element methods. The in-containment refueling water storage tank is irregular in plan and is analyzed by finite element methods.
3.7.3.17 Time History Analysis of Piping Systems
[The time history dynamic analysis is an alternate seismic analysis method for response spectrum analysis when time history seismic input is used. This method is also used for dynamic analyses of piping systems subjected to time history hydraulic transient loadings or forcing functions induced by postulated pipe breaks. The modal superposition method is used to solve the equations of motion.
The computer programs used are GAPPIPE, PIPESTRESS, ANSYS, and WECAN.
The modal superposition method is based on the equations of motion which can be decoupled as long as the piping system is within its elastic limit. The modal responses are obtained from integrating the decoupled equations. The total responses are obtained by the algebraic sum of the individual responses of the individual modes at each time step. The cutoff frequency is selected based on the frequency content of the input forcing function and the highest significant frequency of the piping system. The integration time step is no larger than 10 percent of the period of the cutoff frequency.
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR For dynamic analysis, including seismic analysis at a hard rock site, three separate analyses are performed for each loading case to account for uncertainties. The three analyses correspond to three different time scales: normal time, time expanded by 15 percent, and time compressed by 15 percent. For time history analysis of piping system models that include a dynamic model of the supporting concrete building either the building stiffness is varied by + and - 30 percent, or the time scale is shifted by + and - 15 percent. Alternately, when uniform enveloping time history analysis is performed, modeling uncertainties are accounted for by the spreading that is included in the broadened response spectra.
For time history analysis using the PIPESTRESS program, the response from the high frequency modes above the cutoff frequency is calculated based on the static response to the left-out-forces.
This response is combined with the response from the low frequency modes by algebraic sum at each time step. Composite modal damping is used with PIPESTRESS program. The damping of the individual components is as listed in Table 3.7.1-1.
Alternately, for time history analysis using the PIPESTRESS, GAPPIPE, ANSYS, or WECAN programs, the number of modes used in the modal analysis is chosen so that the results of the dynamic analysis based on the chosen number of modes are within 10 percent of the results of the dynamic analysis based on the next higher number of modes used. The number of modes analyzed is selected to account for the principal vibration modes of the piping system. The modes are combined by algebraic sum. Composite modal damping is used with the ANSYS or WECAN programs. The damping of the individual components is as listed in Table 3.7.1-1.]*
3.7.4 Seismic Instrumentation 3.7.4.1 Comparison with Regulatory Guide 1.12 Compliance with Regulatory Guide 1.12 is discussed in this section and in Subsection 1.9.1.
Administrative procedures define the maintenance and repair of the seismic instrumentation to keep the maximum number of instruments in-service during plant operation and shutdown in accordance with Regulatory Guide 1.12.
3.7.4.1.1 Safety Design Basis The seismic instrumentation serves no safety-related function and therefore has no nuclear safety design basis.
3.7.4.1.2 Power Generation Design Basis The seismic instrumentation is designed to provide the following:
Collection of seismic data in digital format Analysis of seismic data after a seismic event Operator notification that a seismic event meeting one-third SSE exceedance criteria has occurred Operator notification (after analysis of data) that a predetermined cumulative absolute velocity value has been exceeded
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
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VEGP 3&4 - UFSAR 3.7.4.2 Location and Description of Instrumentation The following instrumentation and associated equipment are used to measure plant response to earthquake motion.Four triaxial acceleration sensor units, located as stated in Subsection 3.7.4.2.1, are connected to a time-history analyzer. The time-history analyzer recording and playback system is located in a panel in the nuclear island in a room near the main control room. Seismic event data from these sensors are recorded on a solid-state digital recording system at 200 samples per second per data channel.
This solid-state recording and analysis system has internal batteries and a charger to prevent the loss of data during a power outage, and to allow data collection and analysis in a seismic event during which the power fails. Normally 120 volt alternating current power is supplied from the non-Class 1E dc and uninterruptible power supply system. The system uses triaxial acceleration sensor input signals to initiate the time-history analyzer recording and main control room alarms. The system initiation value is adjustable from 0.002g to 0.02g.
The time-history analyzer starts recording triaxial acceleration data from each of the triaxial acceleration sensors after the initiation value has been exceeded. Pre-event recording time is adjustable from 1.2 to 15.0 seconds, and will be set to record at least 3 seconds of pre-event signal.
Post-event run time is adjustable from 10 to 90 seconds. A minimum of 25 minutes of continuous recording is provided. Each recording channel has an associated timing mark record with 2 marks per second, with an accuracy of about 0.02 percent.
The instrumentation components are qualified to IEEE 344-1987 (Reference 16).
The sensor installation anchors are rigid so that the vibratory transmissibility over the design spectra frequency range is essentially unity.
3.7.4.2.1 Triaxial Acceleration Sensors Each sensor unit contains three accelerometers mounted in a mutually orthogonal array mounted with one horizontal axis parallel to the major axis assumed in the seismic analysis. The triaxial acceleration sensors have a dynamic range of 1000 to 1 (0.001 to 1.0g) and a frequency range of 0.2 to 50 hertz.
One sensor unit will be located in the free field, as discussed in below. The AP1000 seismic monitoring system will provide for signal input from the free field sensor.
A second sensor unit is located on the nuclear island basemat in the spare battery charger room at elevation 66-6 near column lines 9 and L.
A third sensor unit is located on the shield building structure at elevation 266 near column lines 4-1 and K.
The fourth sensor unit is located on the containment internal structure on the east wall of the east steam generator compartment just above the operating floor at elevation 138 close to column lines 6 and K.
Seismic instrumentation is not located on equipment, piping, or supports since experience has shown that data obtained at these locations are obscured by vibratory motion associated with normal plant operation.
A free-field sensor will be located and installed to record the ground surface motion representative of the site. To be representative of this site in regards to seismic response of structures, systems, and 3.7-43 Revision 9
VEGP 3&4 - UFSAR components, the free-field sensor is located on the ground surface of the engineered backfill. The backfill directly supports the Nuclear Island and the adjacent structures and extends out from these structures a significant distance. The free-field sensor is located where the backfill vertically extends from the top of the Blue Bluff Marl to the ground surface, but horizontally at a distance where possible effects on recorded ground motion associated with surface features, buildings, and components would be minimized. The trigger value is initially set at 0.01g.
3.7.4.2.2 Time-History Analyzer The time-history analyzer receives input from the triaxial acceleration sensors and, when activated as described in Subsection 3.7.4.3, begins recording the triaxial data from each triaxial acceleration sensor and initiates audio and visual alarms in the main control room.
This recorded data will be used to evaluate the seismic acceleration of the structure on which the triaxial acceleration sensors are mounted.
The time-history analyzer is a multichannel, digital recording system with the capability to automatically download the recorded acceleration data to a dedicated computer for data storage, playback, and analysis after a seismic event.
The time-history analyzer can compute cumulative absolute velocity (CAV) and the 5 percent of critical damping response spectrum for frequencies between 1 and 10 Hz. The operator may select the analysis of either CAV or the response spectrum. Analysis results are printed out on a dedicated graphics printer that is part of the system and is located in the same panel as the time-history analyzer.
3.7.4.3 Control Room Operator Notification The time-history analyzer provides for initiation of audible and visual alarms in the main control room when the one-third SSE levels sensed by any of the triaxial acceleration sensors are exceeded and when the system is activated to record a seismic event. In addition to alarming when the system is activated, the analyzer portion of the system will provide a second alarm if the predetermined cumulative absolute velocity value has been exceeded by any of the sensors. Alarms are annunciated in the main control room.
3.7.4.4 Comparison of Measured and Predicted Responses The recorded seismic data is used by the combined license holder operations and engineering departments to evaluate the effects of the earthquake on the plant structures and equipment.
The criterion for initiating a plant shutdown following a seismic event will be exceedance of a specified response spectrum limit and a cumulative absolute velocity limit. The seismic instrumentation system is capable of computing the cumulative absolute velocity as described in EPRI Report NP-5930 (Reference 1) and EPRI Report TR-100082 (Reference 17).
Post-earthquake operating procedures utilize the guidance of EPRI Reports NP-5930, TR-100082, and NP-6695, as modified and endorsed by the NRC in Regulatory Guides 1.166 and 1.167. A response spectrum check up to 10Hz and the cumulative absolute velocity will be calculated based on the recorded motions at the free field instrument. If the operating basis earthquake ground motion is exceeded or significant plant damage occurs, the plant must be shutdown in an orderly manner.
In addition, the procedures address measurement of the post-seismic event gaps between the new fuel rack and walls of the new fuel storage pit, between the individual spent fuel racks, and from the 3.7-44 Revision 9
VEGP 3&4 - UFSAR spent fuel racks to the spent fuel pool walls, and provide for appropriate corrective actions to be taken if needed (such as repositioning the racks or analysis of the as-found condition).
3.7.4.5 Tests and Inspections Periodic testing of the seismic instrumentation system is accomplished by the functional test feature included in the software of the time-history recording accelerograph. The system is modular and is capable of single-channel testing or single channel maintenance without disabling the remainder of the system.
Installation and acceptance testing of the triaxial acceleration sensors described in Subsection 3.7.4.2.1 is completed prior to initial startup. Installation and acceptance testing of the time-history analyzer described in Subsection 3.7.4.2.2 is completed prior to initial startup.
3.7.5 Combined License Information 3.7.5.1 Seismic Analysis of Dams Dams whose failure could affect the site interface flood level are addressed in Subsection 3.7.2.12 and Subsection 2.4.4.
3.7.5.2 Post-Earthquake Procedures Site-specific procedures for activities following an earthquake are addressed in Subsection 3.7.4.4.
3.7.5.3 Seismic Interaction Review The seismic interaction review will be updated for as-built information. This review is performed in parallel with the seismic margin evaluation. The review is based on as-procured data, as well as the as-constructed condition. The as-built seismic interaction review is completed prior to fuel load.
3.7.5.4 Reconciliation of Seismic Analyses of Nuclear Island Structures The seismic analyses described in Subsection 3.7.2 will be reconciled for detailed design changes, such as those due to as-procured or as-built changes in component mass, center of gravity, and support configuration based on as-procured equipment information. Deviations are acceptable based on an evaluation consistent with the methods and procedure of Section 3.7 provided the amplitude of the seismic floor response spectra, including the effect due to these deviations, does not exceed the design basis floor response spectra by more than 10 percent. This reconciliation will be completed prior to fuel load.
3.7.5.5 Free Field Acceleration Sensor The location for the free-field acceleration sensor is addressed in Subsection 3.7.4.2.1.
3.7.6 References
- 1. EPRI Report NP-5930, "A Criterion for Determining Exceedance of the Operating Basis Earthquake," July 1988.
- 2. Uniform Building Code, 1997.
- 3. ASCE Standard 4-98, "Seismic Analysis of Safety-Related Nuclear Structures and Commentary," American Society of Civil Engineers, September 1998.
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- 4. ASME B&PV Code, Code Case N-411.
- 5. FEMA 356, "Prestandard and Commentary for the Seismic Rehabilitation of Buildings,"
Federal Emergency Management Agency, November 2000.
- 6. Not used.
- 7. Not used.
- 8. Not used.
- 9. Not used.
- 10. Hyde, S. J., J. M. Pandya, and K. M. Vashi, "Seismic Analysis of Auxiliary Mechanical Equipment in Nuclear Plants," Dynamic and Seismic Analysis of Systems and Components, ASME-PVP-65, American Society of Mechanical Engineers, Orlando, Florida, 1982.
- 11. Lin, C. W., and T. C. Esselman, "Equivalent Static Coefficients for Simplified Seismic Analysis of Piping Systems," SMIRT Conference 1983, Paper K12/9.
- 12. M. S. Yang, J. S. M. Leung, and Y. K. Tang, "Analysis of Piping Systems with Gapped Supports Using the Response Spectrum Method." Presented at the 1989 ASME Pressure Vessels and Piping Conference at Honolulu, July 23-27, 1989.
- 13. "Impact Response of Piping Systems with Gaps," P. H. Anderson and H. Loey, ASME Seismic Engineering, 1989, Volume 182.
- 14. "Independent Support Motion (ISM) Method of Modal Spectra Seismic Analysis,"
December 1989; by Task Group on Independent Support Motion as Part of the PVRC Technical Committee on Piping Systems Under the Guidance of the Steering Committee.
- 15. J. S. Meserole, A. Fortini, "Slosh Dynamics in a Toroidal Tank," Journal Spacecraft Vol. 24, Number 6, November-December 1987.
- 16. IEEE 344-1987, "Recommended Practices for Seismic Qualification of 1E Equipment for Nuclear Power Generating Stations."
- 17. EPRI Report TR-100082, "Standardization of the Cumulative Absolute Velocity,"
December 1991.
- 18. EPRI Report NP-6695, "Guidelines for Nuclear Plant Response to an Earthquake,"
December 1989.
- 19. "Cable Tray and Conduit Raceway Seismic Test Program, Release 4,"
Report 1053-21.1-4, ANCO Engineers, Inc., December 15, 1978.
- 20. Not used.
- 21. Not used.
- 22. WCAP-7921-AR, Damping Values of Nuclear Power Plant Components, May 1974.
3.7-46 Revision 9
- 23. McGuire, R. K., G. R. Toro, and W. J. Silva, "Engineering Model of Earthquake Ground Motion for Eastern North America," Technical Report NP-6074, Electric Power Research Institute, 1988.
- 24. Boore, D. M. and G. M. Atkinson, "Stochastic prediction of ground motion and spectral response at hard-rock sites in eastern North America," Bull. Seism. Soc. Am., 77:2, pages 440-467.
- 25. Nuttli, O. W., Letter dated September 19, 1986 to J. B. Savy, Reproduced in:
D. Bernreuter, J. Savy, R. Mensing, J. Chen, and B. Davis, "Seismic Hazard Characterization of 69 Nuclear Plant Sites East of the Rocky Mountains:
Questionnaires," U.S. Nuclear Regulatory Commission, Technical Report NUREG/
CR-5250, UCID-21517, 7, Prepared by Lawrence Livermore National Laboratory.
- 26. Trifunac, M. and V. W. Lee, "Preliminary Empirical Model for Scaling Pseudo Velocity Spectra of Strong Earthquake Accelerations in Terms of Magnitude, Distance, Site Intensity and Recording Site Conditions," Report No. CE 85-04, University of Southern California, Department of Civil Engineering, 1985.
- 27. Bernreuter, D., J. Savy, R. Mensing, J. Chen, and B. Davis, "Seismic Hazard Characterization of 69 Nuclear Plant Sites East of the Rocky Mountains:
Questionnaires," U.S. Nuclear Regulatory Commission, Technical Report NUREG/
CR-5250, UCID-21517, Lawrence Livermore National Laboratory, 1989.
- 28. McGuire, R. K., G. R. Toro, J. P Jacobson, T. F. OHara, and W. J. Silva, "Probabilistic Seismic Hazard Evaluations at Nuclear Plant Sites in the Central and Eastern United States: Resolution of the Charleston Earthquake Issue," Technical Report NP-6395-D, Electric Power Research Institute, 1989.
- 29. Philippacoupoulos, A. J., "Recommendations for Resolution of Public Comments on USI A-40, Seismic Design Criteria," Brookhaven National Laboratory Report BNL-NUREG-52191, prepared for the U.S. Nuclear Regulatory Commission, and published as NUREG/CR-5347, 1989.
- 30. C. Chen, "Definition of Statistically Independent Time Histories," Journal of the Structural Division, ASCE, February 1975.
- 31. WCAP-9903, "Justification of the Westinghouse Equivalent Static Analysis Method for Seismic Qualification of Nuclear Power Plant Auxiliary Mechanical Equipment,"
August 1980.
- 32. Letter from James T. Wiggins to John J. Taylor, September 13, 1993.
- 33. Not used.
- 34. "Seismic Provisions for Structural Steel Buildings," American Institute of Steel Construction, April 1997 including Supplement No. 2, November 2000.
- 35. "Minimum Design Loads for Buildings and Other Structures," American Society of Civil Engineers, ASCE 7-98.
- 36. ANSYS Engineering Analysis Users Manual, Releases up to and including ANSYS 5.7, ANSYS Inc.
3.7-47 Revision 9
- 37. H. B. Seed, and I. M. Idriss, "Soil Moduli and Damping Factors for Dynamic Response Analysis," Report No. EERC-70-14, Earthquake Engineering Research Center, University of California, Berkeley, 1970.
- 38. EPRI TR-102293, "Guidelines for Determining Design Basis Ground Motions," 1993.
- 40. International Building Code, 2006.
- 41. AISC 341-05, "Seismic Provisions for Structural Steel Buildings," March 9, 2005, including Supplement No. 1 dated November 16, 2005.
- 42. ASCE 7-05, "Minimum Design Loads for Buildings and Other Structures."
- 43. AISC 360, "Specification for Structural Steel Buildings," March 9, 2005.
3.7-48 Revision 9
VEGP 3&4 - UFSAR Table 3.7.1-1 Safe Shutdown Earthquake Damping Values Percent Welded and friction-bolted steel structures and equipment 4 Bearing bolted structures and equipment 7 Prestressed concrete structures 5 Reinforced concrete structures 7 Concrete filled steel plate structures 5
[Piping (for uniform envelope response spectra analysis) 5 Piping (alternative for time history analysis and independent support motion response spectra analysis)
Less than or equal to 12-inch diameter 2 Greater than 12-inch diameter 3 Primary coolant loop 4]*
Fuel assemblies 20 Control rod drive mechanisms 5 Full cable trays and related supports 10 Empty cable trays and related supports 7 Conduits and related supports 7 HVAC ductwork 7 HVAC welded ductwork 4 Cabinets and panels for electrical equipment 5 Equipment such as welded instrument racks and tanks 3
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.7-49 Revision 9
VEGP 3&4 - UFSAR Table 3.7.1-2 Embedment Depth and Related Dimensions of Category I Structures Foundation Embedment Least Foundation Structure Depth (ft) Width (ft) Structure Height (ft)
Shield Building See Note See Note 268.25 Steel Containment Vessel See Note See Note 215.33 Auxiliary Building See Note See Note 119.50 Note:
- 1. The seismic Category I structures are founded on a common basemat embedded 39.5 feet, [with dimensions shown in Figure 3.7.1-14.]*
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.7-50 Revision 9
VEGP 3&4 - UFSAR Table 3.7.1-3 AP1000 Design Response Spectra Amplification Factors for Control Points HORIZONTAL Percent of Acceleration(1) Displacement(1)
Critical Damping A (33 cps) B' (25 cps)(2) B (9 cps) C (2.5 cps) D (0.25 cps) 2.0 1.0 1.70 3.54 4.25 2.50 3.0 1.0 1.66 3.13 3.76 2.34 4.0 1.0 1.63 2.84 3.41 2.19 5.0 1.0 1.60 2.61 3.13 2.05 7.0 1.0 1.55 2.27 2.72 1.88 VERTICAL Percent of Critical Acceleration(1) Displacement(1)
Damping A (33 cps) B' (25 cps)(2) B (9 cps) C (3.5 cps) D (0.25 cps) 2.0 1.0 1.70 3.54 4.05 1.67 3.0 1.0 1.66 3.13 3.58 1.56 4.0 1.0 1.63 2.84 3.25 1.46 5.0 1.0 1.60 2.61 2.98 1.37 7.0 1.0 1.55 2.27 2.59 1.25 Notes:
- 1. Maximum ground displacement is taken proportional to maximum ground acceleration, and is 36 inches for ground acceleration of 1.0 gravity.
- 2. The 5 percent damping amplification factor for control point B' is derived per discussion in Subsection 3.7.1.1. This 5 percent damping amplification factor equals 1.3 times the RG 1.60 response spectra at 25 hertz. The amplification factors at control point B' for other damping values are determined by increasing the RG 1.60 response spectra at 25 hertz by 30 percent.
3.7-51 Revision 9
VEGP 3&4 - UFSAR Table 3.7.1-4 (Sheet 1 of 5)
Strain Compatible Soil Properties Depth to Total Unit Bottom of Thickness of Layer Weight Initial G Initial Vs Final G Final Vs Layer (ft) Layer (ft) Number (kcf) (ksf) (fps) (ksf) (fps) Damping Firm Rock 0.0 - - - - - - - -
5.0 5.0 1 0.15 57422 3500 57032 3499 0.015 10.0 5.0 2 0.15 57422 3500 56600 3486 0.016 15.0 5.0 3 0.15 57422 3500 55943 3465 0.017 20.0 5.0 4 0.15 57422 3500 55511 3452 0.018 25.0 5.0 5 0.15 56442 3500 55933 3465 0.016 30.0 5.0 6 0.15 56442 3500 55436 3450 0.017 33.5 3.5 7 0.15 57422 3500 56076 3470 0.015 39.5 6.0 8 0.15 57422 3500 55898 3464 0.015 45.0 5.5 9 0.15 57422 3500 55716 3458 0.016 50.0 5.0 10 0.15 57422 3500 55575 3454 0.016 60.0 10.0 11 0.15 56442 3500 55400 3449 0.017 80.0 20.0 12 0.15 56442 3500 54695 3427 0.019 100.0 20.0 13 0.15 56442 3500 53358 3384 0.021 120.0 20.0 14 0.15 56442 3500 52295 3351 0.023 Bedrock - - 0.15 300000 8000 298137 8000 0.02 3.7-52 Revision 9
VEGP 3&4 - UFSAR Table 3.7.1-4 (Sheet 2 of 5)
Strain Compatible Soil Properties Depth to Total Unit Bottom of Thickness of Layer Weight Initial G Initial Vs Final G Final Vs Layer (ft) Layer (ft) Number (kcf) (ksf) (fps) (ksf) (fps) Damping Soft Rock 0.0 - - - - - - - -
5.0 5.0 1 0.15 27660 2429 27425 2426 0.016 10.0 5.0 2 0.15 29180 2495 28318 2466 0.018 15.0 5.0 3 0.15 30262 2541 28819 2487 0.020 20.0 5.0 4 0.15 30620 2556 28589 2477 0.023 25.0 5.0 5 0.15 30920 2568 29290 2508 0.019 30.0 5.0 6 0.15 31384 2588 29481 2516 0.021 33.5 3.5 7 0.15 31932 2610 30768 2570 0.017 39.5 6.0 8 0.15 32464 2632 31144 2586 0.018 45.0 5.5 9 0.15 33042 2655 31314 2593 0.019 50.0 5.0 10 0.15 33668 2680 31598 2604 0.020 60.0 10.0 11 0.15 34341 2707 31826 2614 0.021 80.0 20.0 12 0.15 35021 2733 31738 2610 0.024 100.0 20.0 13 0.15 35708 2760 31585 2604 0.026 120.0 20.0 14 0.15 36401 2787 31585 2604 0.028 Bedrock - - 0.15 300000 8000 298137 8000 0.020 3.7-53 Revision 9
VEGP 3&4 - UFSAR Table 3.7.1-4 (Sheet 3 of 5)
Strain Compatible Soil Properties Depth to Total Unit Bottom of Thickness of Layer Weight Initial G Initial Vs Final G Final Vs Layer (ft) Layer (ft) Number (kcf) (ksf) (fps) (ksf) (fps) Damping Upper Bound Soft to Medium Soil 0 - - - - - - - -
5.0 5.0 1 0.11 6873 1414 6664 1397 0.018 10.0 5.0 2 0.11 9844 1692 9202 1641 0.023 15.0 5.0 3 0.11 13917 2012 12880 1942 0.024 20.0 5.0 4 0.11 14971 2087 13629 1997 0.027 25.0 5.0 5 0.11 15645 2133 14574 2065 0.022 30.0 5.0 6 0.11 16419 2186 15045 2099 0.024 33.5 3.5 7 0.11 17873 2280 16908 2225 0.019 39.5 6.0 8 0.11 19036 2353 17873 2287 0.020 45.0 5.5 9 0.11 20387 2435 18996 2358 0.021 50.0 5.0 10 0.11 21726 2514 20136 2428 0.021 60.0 10.0 11 0.11 23234 2600 21366 2501 0.022 80.0 20.0 12 0.11 24712 2681 22314 2556 0.024 100.0 20.0 13 0.11 26151 2758 23137 2602 0.026 120.0 20.0 14 0.11 27546 2831 24009 2651 0.027 Bedrock - - 0.15 300000 8000 298137 8000 0.020 3.7-54 Revision 9
VEGP 3&4 - UFSAR Table 3.7.1-4 (Sheet 4 of 5)
Strain Compatible Soil Properties Depth to Total Unit Bottom of Thickness of Layer Weight Initial G Initial Vs Final G Final Vs Layer (ft) Layer (ft) Number (kcf) (ksf) (fps) (ksf) (fps) Damping Soft-to-Medium Soil 0.0 - - - - - - - -
5.0 5.0 1 0.11 3438 1000 3222 971 0.023 10.0 5.0 2 0.11 4923 1197 4355 1129 0.031 15.0 5.0 3 0.11 6960 1423 5987 1324 0.035 20.0 5.0 4 0.11 7487 1476 6161 1343 0.040 25.0 5.0 5 0.11 7824 1509 6699 1400 0.031 30.0 5.0 6 0.11 8211 1546 6891 1420 0.033 33.5 3.5 7 0.11 8938 1613 7872 1518 0.026 39.5 6.0 8 0.11 9520 1664 8317 1560 0.027 45.0 5.5 9 0.11 10195 1722 8834 1608 0.028 50.0 5.0 10 0.11 10864 1778 9347 1654 0.029 60.0 10.0 11 0.11 11618 1838 9818 1695 0.031 80.0 20.0 12 0.11 12357 1896 10031 1714 0.036 100.0 20.0 13 0.11 13077 1950 10201 1728 0.040 120.0 20.0 14 0.11 13774 2002 10512 1754 0.043 Bedrock - 0.15 300000 8000 298137 8000 0.020 3.7-55 Revision 9
VEGP 3&4 - UFSAR Table 3.7.1-4 (Sheet 5 of 5)
Strain Compatible Soil Properties Depth to Total Unit Bottom of Thickness of Layer Weight Initial G Initial Vs Final G Final Vs Layer (ft) Layer (ft) Number (kcf) (ksf) (fps) (ksf) (fps) Damping Soft Soil 0.0 - - - - - - - -
5.0 5.0 1 0.11 3438 1000 3222 971 0.023 10.0 5.0 2 0.11 3633 1028 3042 944 0.038 15.0 5.0 3 0.11 3865 1060 2974 933 0.047 20.0 5.0 4 0.11 3921 1068 2752 898 0.059 25.0 5.0 5 0.11 3955 1073 2922 925 0.049 30.0 5.0 6 0.11 3994 1078 2762 899 0.056 33.5 3.5 7 0.11 4065 1088 3022 941 0.046 39.5 6.0 8 0.11 4121 1095 2958 931 0.049 45.0 5.5 9 0.11 4183 1103 2896 921 0.053 50.0 5.0 10 0.11 4244 1111 2851 914 0.056 60.0 10.0 11 0.11 4310 1120 2774 901 0.062 80.0 20.0 12 0.11 4374 1128 2668 884 0.068 100.0 20.0 13 0.11 4434 1136 2691 888 0.069 120.0 20.0 14 0.11 4492 1143 2718 892 0.069 Bedrock - - 0.15 300000 8000 298137 8000 0.020 3.7-56 Revision 9
VEGP 3&4 - UFSAR Tables 3.7.2-1-3.7.2-16 Not Used 3.7-57 Revision 9
VEGP 3&4 - UFSAR Table 3.7.3-1 (Sheet 1 of 3)
Seismic Category I Equipment Outside Containment by Room Number Room No. Room Name Equipment Description 12101 Division A battery room Batteries 12102 Division C battery room 1 Batteries 12103 Spare battery room Spare batteries 12104 Division B battery room 1 Batteries 12105 Division D battery room Batteries 12113 Spare battery charger room 12154 Auxiliary building sump room RCA floodup level sensors 12162 RNS pump room A RNS pressure boundary 12163 RNS pump room B RNS pressure boundary 12201 Division A dc equipment room dc equipment 12202 Division C battery room 2 Batteries 12203 Division C dc equipment room dc equipment 12204 Division B battery room 2 Batteries 12205 Division D dc equipment room dc equipment 12207 Division B dc equipment room dc equipment 12211 Corridor Divisional cables SGS instrumentation tubing 12212 Division B RCP trip switchgear room RCP trip switchgear 12244 Lower annulus valve area CVS/WLS containment isolation valves 12251 Demineralizer/filter access area CVS/DWS isolation valves 12254 SFS penetration room SFS containment isolation valve 12256 Containment isolation valve room RNS containment isolation valves 12259 Pipe chase RNS piping 12262 Piping/Valve room RNS pressure boundary, SFS piping 12265 Waste monitor tank room C SFS piping 12269 Pipe chase RNS pressure boundary 12300 Corridor Divisional cable, MCR load shed panel SGS instrumentation tubing 12301 Division A I&C room Divisional I&C 12302 Division C I&C room Divisional I&C 3.7-58 Revision 9
VEGP 3&4 - UFSAR Table 3.7.3-1 (Sheet 2 of 3)
Seismic Category I Equipment Outside Containment by Room Number Room No. Room Name Equipment Description 12303 Remote shutdown room Divisional cabling 12304 Division B I&C/penetration room Divisional I&C/electrical penetrations 12305 Division D I&C/penetration room Divisional I&C/electrical penetrations 12306 Valve/piping penetration room CCS/CVS/DWS/FPS/SGS containment isolation valves SGS instrumentation tubing 12311 Corridor Divisional cabling 12312 Division C RCP trip switchgear room RCP trip switchgear 12313 Division C I&C/penetration room Divisional I&C/electrical penetrations 12321 Non-1E equipment/penetration room Divisional cabling/electrical penetrations 12341 Middle annulus Class 1E electrical penetrations Various mechanical piping penetrations 12343 Division C middle annulus penetration room Class 1E electrical penetrations 12344 Division B middle annulus penetration room Class 1E electrical penetrations 12345 Division D middle annulus penetration room Class 1E electrical penetrations 12351 Maintenance floor staging area Divisional cabling (ceiling) 12352 Personnel hatch Personnel airlock (interlocks) 12354 Middle annulus access room PSS/SFS containment isolation valves 12362 RNS HX room RNS pressure boundary 12365 Waste monitor tank room B SFS piping 12400 Control room vestibule Control room access 12401 Main control room Dedicated safety panel VBS HVAC dampers VES isolation valves Lighting circuits Mounting for lighting fixtures 12404 Lower MSIV compartment B SGS containment isolation valves, instrumentation 12405 Lower VBS B and D equipment room VWS/PXS/CAS containment isolation valves 12406 Lower MSIV compartment A SGS containment isolation valves, instrumentation 12411 Corridor Divisional cabling 12412 Electrical penetration room Division A Divisional electrical penetrations, MCR load shed panel 3.7-59 Revision 9
VEGP 3&4 - UFSAR Table 3.7.3-1 (Sheet 3 of 3)
Seismic Category I Equipment Outside Containment by Room Number Room No. Room Name Equipment Description 12421 Non 1E equipment/penetration room Divisional cabling/electrical penetrations 12422 Reactor trip switchgear II Reactor trip switchgear 12423 Reactor trip switchgear I Reactor trip switchgear 12452 VFS penetration room VFS containment isolation valves, divisional cabling 12454 VFS/SFS/PSS penetration room SFS/PSS/VFS containment isolation valves, RNS pressure boundary 12462 Cask washdown pit SFS piping 12463 Cask loading pit/Spent fuel storage pit Spent fuel transfer gate 12501 VBS MCR/A&C equipment room Divisional cabling 12504 Upper MSIV compartment B SGS CIVs, instrumentation 12506 Upper MSIV compartment A SGS CIVs, instrumentation 12541 Upper annulus PCS piping and cabling PCS air baffle 12553 Personnel access area Personnel airlock (interlocks) 12555 VES air storage VES high pressure air bottles 12651 VAS Equipment Room VFS containment isolation valves 12563 Spent fuel storage pit Spent fuel storage racks Spent fuel transfer gate Spent fuel cask loading pit gate 12701 PCS valve room PCS isolation valves/instrumentation 12703 PCS water storage tank PCS piping, level and temperature instrumentation 3.7-60 Revision 9
VEGP 3&4 - UFSAR Table 3.7.3-2 Equipment Classified as Sensitive Targets for Seismically Analyzed Piping, HVAC Ducting, Cable Tray Systems Component Discussion Zone of Protection Seismic Category I Valve These are manual valves. The actuator Valve body and actuator No Class 1E Electrical Equipment must be protected from impact. area Not pressure sensitive Seismic Category I Valve These valves have sensitive Class 1E One support (acting in Class 1E Electrical Equipment equipment (e.g., Position indicators, limit direction of impact) on Pressure sensitive switches, motor operator) or solenoid each side of valve valves.
Seismic Category I Dampers The actuator must be protected along with Within one support any Class 1E equipment. (acting in direction of impact) on each side of HVAC Monitors This includes: neutron detectors, radiation Monitors and associated monitors, resistance temperature detectors, wiring speed sensors, thermocouples, and transmitters.
Sensitive Electrical Equipment Housed in This includes: relays, contractors, Cabinets, panels, and Cabinets, Panels or Boards breakers, and switchgear. boards housing sensitive devices Class 1E exposed cables and wiring Cables and wiring which are not housed in Exposed cables and cable trays or conduits must be protected. wiring Device or Instrument Tubing Any device or tubing that could be damaged Device or tubing resulting in the loss of the pressure boundary of a safety class line.
Penetrations Rigid penetrations are considered robust. Floating penetration and Floating penetrations with bellows are associated bellows considered sensitive.
3.7-61 Revision 9
VEGP 3&4 - UFSAR Figure 3.7.1-1 Horizontal Design Response Spectra Safe Shutdown Earthquake 3.7-62 Revision 9
VEGP 3&4 - UFSAR Figure 3.7.1-2 Vertical Design Response Spectra Safe Shutdown Earthquake 3.7-63 Revision 9
VEGP 3&4 - UFSAR Figure 3.7.1-3 Design Horizontal Time History, "H1" Acceleration, Velocity & Displacement Plots 3.7-64 Revision 9
VEGP 3&4 - UFSAR Figure 3.7.1-4 Design Horizontal Time History, "H2" Acceleration, Velocity & Displacement Plots 3.7-65 Revision 9
VEGP 3&4 - UFSAR Figure 3.7.1-5 Design Vertical Time History Acceleration, Velocity & Displacement Plots 3.7-66 Revision 9
VEGP 3&4 - UFSAR Figure 3.7.1-6 Acceleration Response Spectra of Design Horizontal Time History, "H1" 3.7-67 Revision 9
VEGP 3&4 - UFSAR Figure 3.7.1-7 Acceleration Response Spectra of Design Horizontal Time History, "H2" 3.7-68 Revision 9
VEGP 3&4 - UFSAR Figure 3.7.1-8 Acceleration Response Spectra of Design Vertical Time History 3.7-69 Revision 9
VEGP 3&4 - UFSAR Figure 3.7.1-9 Minimum Power Spectral Density Curve (Normalized to 0.3g) 3.7-70 Revision 9
VEGP 3&4 - UFSAR PSD of Design Time History H1, with 20% averaging Target PSD, anchored to 0.3g Figure 3.7.1-10 Power Spectral Density of Design Horizontal Time History, "H1" 3.7-71 Revision 9
VEGP 3&4 - UFSAR PSD of Design Time History H2, with 20% averaging Target PSD, anchored to 0.3g Figure 3.7.1-11 Power Spectral Density of Design Horizontal Time History, "H2" 3.7-72 Revision 9
VEGP 3&4 - UFSAR PSD of Design Time History VERTICAL, with 20% averaging Target PSD, anchored to 0.3g Figure 3.7.1-12 Power Spectral Density of Design Vertical Time History 3.7-73 Revision 9
VEGP 3&4 - UFSAR Figure 3.7.1-13 Not Used 3.7-74 Revision 9
'-6 29'-0" X
(NORTH)
SHIELD BUILDING 91'-0" 117'-6" 88'-6" AUXILIARY BUILDING 138'-0" 118'-0" 256'-0" PLAN Figure 3.7.1-14
[Nuclear Island Structures Dimensions]*
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.7-75 Revision 9
VEGP 3&4 - UFSAR Figure 3.7.1-15 Strain Dependent Properties of Rock Material (Ref. 37) 3.7-76 Revision 9
VEGP 3&4 - UFSAR Figure 3.7.1-16 Strain Dependent Properties of Soil Material (Ref. 38) 3.7-77 Revision 9
VEGP 3&4 - UFSAR Shear Wave Velocity Comparison 0
-50
-100 FR Depth (ft)
SR SMS-UB SMS
-150 SS
-200
-250 0 500 1000 1500 2000 2500 3000 3500 4000 Shear Wave (fps)
Initial Properties Shear Wave Velocity Comparison 0
Soft Rock
-50 Hard Rock Depth (ft)
Vs> 8000 fps Upperbound Soft to Medium Soil
-100 Soft to Firm Rock Soft Soil Medium Soil
-150 0 500 1000 1500 2000 2500 3000 3500 4000 4500 Shear Wave (fps)
Strain-Iterated Shear Wave Velocity Profiles Note: Fixed base analyses were performed for hard rock sites. These analyses are applicable for shear wave velocity greater than 8000 feet per second.
Figure 3.7.1-17 Generic Soil Profiles 3.7-78 Revision 9
VEGP 3&4 - UFSAR Figures 3.7.2-1-3.7.2-11 Not Used 3.7-79 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-12 (Sheet 1 of 12)
[Nuclear Island Key Structural Dimensions Plan at El. 66-6]*
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.7-80 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-12 (Sheet 2 of 12)
[Nuclear Island Key Structural Dimensions Plan at El. 82-6]*
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.7-81 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-12 (Sheet 3 of 12)
[Nuclear Island Key Structural Dimensions Plan at El. 100-0 & 107-2]*
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.7-82 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-12 (Sheet 4 of 12)
[Nuclear Island Key Structural Dimensions Plan at El. 117-6]*
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.7-83 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-12 (Sheet 5 of 12)
[Nuclear Island Key Structural Dimensions Plan at El. 135-3]*
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.7-84 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-12 (Sheet 6 of 12)
[Nuclear Island Key Structural Dimensions Plan at El. 153-0 & 160-6]*
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.7-85 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-12 (Sheet 7 of 12)
[Nuclear Island Key Structural Dimensions Plan at El. 160-6, 180-0, & 329-0]*
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.7-86 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-12 (Sheet 8 of 12)
[Nuclear Island Key Structural Dimensions Section A - A]*
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.7-87 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-12 (Sheet 9 of 12)
[Nuclear Island Key Structural Dimensions Section B - B]*
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.7-88 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-12 (Sheet 10 of 12)
[Nuclear Island Key Structural Dimensions Sections C - C and H - H]*
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.7-89 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-12 (Sheet 11 of 12)
[Nuclear Island Key Structural Dimensions Section G - G]*
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.7-90 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-12 (Sheet 12 of 12)
[Nuclear Island Key Structural Dimensions Section J - J]*
- In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval may be required prior to implementing a change in this information.
3.7-91 Revision 9
VEGP 3&4 - UFSAR Figure 3.7.2-13 Not Used 3.7-92 Revision 9
VEGP 3&4 - UFSAR Figure 3.7.2-14 Typical Design Floor Response Spectrum 3.7-93 Revision 9
VEGP 3&4 - UFSAR Figures 3.7.2-15-3.7.2-18 Not Used 3.7-94 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-19 (Sheet 1 of 10)
Annex Building Key Structural Dimensions Plan at Elevation 100-0 3.7-95 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-19 (Sheet 2 of 10)
Annex Building Key Structural Dimensions Plan at Elevation 107-2 and 117-6 3.7-96 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-19 (Sheet 3 of 10)
Annex Building Key Structural Dimensions Plan at Elevation 135-3 3.7-97 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-19 (Sheet 4 of 10)
Annex Building Key Structural Dimensions Plan at Elevation 158-0 and 150-3 3.7-98 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-19 (Sheet 5 of 10)
Annex Building Key Structural Dimensions Roof Plan at Elevation 154-0 and 181-11 3/4 3.7-99 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-19 (Sheet 6 of 10)
Annex Building Key Structural Dimensions Section A - A 3.7-100 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-19 (Sheet 7 of 10)
Annex Building Key Structural Dimensions Section B - B 3.7-101 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-19 (Sheet 8 of 10)
Annex Building Key Structural Dimensions Section C - C 3.7-102 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-19 (Sheet 9 of 10)
Annex Building Key Structural Dimensions Sections D - D, E - E, & F - F 3.7-103 Revision 9
VEGP 3&4 - UFSAR Security-Related Information, Withheld Under 10 CFR 2.390d Figure 3.7.2-19 (Sheet 10 of 10)
Annex Building Key Structural Dimensions Sections G - G, H - H, & J - J 3.7-104 Revision 9
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=RQH Figure 3.7.3-2 Impact Evaluation Zone and Seismic Supported Piping 3.7-109 Revision 9
VEGP 3&4 - UFSAR Comparisons of VEGP Horizontal Seismic Response Spectra to AP1000 CSDRS 5% Damping 1.00 0.90 0.80 0.70 Acceleration (g) 0.60 GMRS 0.50 CSDRS FIRS 0.40 0.30 0.20 0.10 0.00 0.1 1 10 100 Frequency - Hertz Figure 3.7-201 VEGP AP1000 Horizontal Spectra Comparison 3.7-110 Revision 9
VEGP 3&4 - UFSAR Comparisons of VEGP Vertical Seismic Response Spectra to AP1000 CSDRS 5% Damping 1.00 0.90 0.80 0.70 Acceleration (g) 0.60 GMRS CSDRS 0.50 FIRS 0.40 0.30 0.20 0.10 0.00 0.1 1 10 100 Frequency - Hertz Figure 3.7-202 VEGP AP1000 Vertical Spectra Comparison 3.7-111 Revision 9