ML23135A168

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6 to Updated Safety Analysis Report, Chapter 3, Design of Structures, Components, Equipment, and Systems
ML23135A168
Person / Time
Site: Wolf Creek Wolf Creek Nuclear Operating Corporation icon.png
Issue date: 05/08/2023
From:
Wolf Creek
To:
Office of Nuclear Reactor Regulation
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ML23135A163 List: ... further results
References
WO 23-0006
Download: ML23135A168 (1)
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{{#Wiki_filter:WOLF CREEK TABLE OF CONTENTS CHAPTER 3.0 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT,AND SYSTEMS Section Page 3.1 CONFORMANCE WITH NRC GENERAL DESIGN 3.1-1 CRITERIA 3.1.1 DEFINITION OF SINGLE FAILURE 3.1-1 3.1.1.1 Active Component 3.1-2 3.1.1.2 Active Component Failure 3.1-2 3.1.1.3 Passive Component 3.1-2 3.1.1.4 Passive Component Failure 3.1-3 3.1.2 ADDITIONAL SINGLE FAILURE ASSUMPTIONS 3.1-3 3.1.3 OVERALL REQUIREMENTS 3.1-5 3.1.4 PROTECTION BY MULTIPLE FISSION PRODUCT 3.1-9 BARRIERS 3.1.5 PROTECTION AND REACTIVITY CONTROL 3.1-18 3.1.6 FLUID SYSTEMS 3.1-25 3.1.7 REACTOR CONTAINMENT 3.1-39 3.1.8 FUEL AND REACTIVITY CONTROL 3.1-44 3.

1.9 REFERENCES

3.1-48 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, 3.2-1 AND SYSTEM 3.2.1 SEISMIC CLASSIFICATION 3.2-2 3.2.2 SYSTEM QUALITY GROUP CLASSIFICATION 3.2-3 3.2.3 SAFETY CLASSES 3.2-3 3.2.4 QUALITY ASSURANCE PROGRAM 3.2-3 3.2.5 ENGINEERING CODES AND STANDARDS 3.2-4 3.2.6 LOCATION 3.2-4 3.

2.7 REFERENCES

3.2-4 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 3.3-2 3.0-i Rev. 29

WOLF CREEK TABLE OF CONTENTS (Continued) Section Page 3.3.2 TORNADO LOADINGS 3.3-2 3.3.2.1 Applicable Design Parameters 3.3-2 3.3.2.2 Determination of Forces on Structures 3.3-3 3.3.2.3 Effect of Failure of Structure or 3.3-3 Components Not Designed for Tornado Loads 3.

3.3 REFERENCES

3.3-3 3.4 WATER LEVEL (FLOOD) DESIGN 3.4-1 3.4.1 FLOOD PROTECTION 3.4-1 3.4.1.1 Flood Protection Measures for Seismic 3.4-1 Category I Structures 3.4.1.2 Ground Water Elevations 3.4-3 3.4.1.3 Permanent Dewatering Systems 3.4-3 3.4.2 ANALYSIS PROCEDURES 3.4-3 3.4.2.1 Design Basis Flood for the ESWS 3.4-4 Pumphouse 3.4.2.2 Design Basis Ground Water 3.4-4 3.5 MISSILE PROTECTION 3.5-1 3.5.1 MISSILE SELECTION AND DESCRIPTIONS 3.5-1 3.5.1.1 Internally Generated Missiles (Outside 3.5-2 Containment) 3.5.1.2 Internally Generated Missiles (Inside 3.5-3 Containment) 3.5.1.3 Turbine Missiles 3.5-3 3.5.1.4 Missiles Generated by Natural 3.5-8 Phenomena 3.5.1.5 Missiles Generated by Events Near 3.5-8 the Site 3.5.1.6 Aircraft Hazards 3.5-8 3.5.2 SYSTEMS TO BE PROTECTED 3.5-11 3.5.2.1 Essential Service Water Systems Pumphouse 3.5-11 3.5.2.2 Essential Service Water Systems Pipe, 3.5-12 Electrical Duct Banks and Manholes 3.0-ii Rev. 29

WOLF CREEK TABLE OF CONTENTS (Continued) Section Page 3.5.2.3 Essential Service Water System Valve 3.5-12 House 3.5.2.4 Essential Service Water System Discharge 3.5-12 Structure 3.5.2.5 Diesel Generator Building 3.5-12 3.5.2.6 Diesel Exhaust Stack 3.5-13 3.5.3 BARRIER DESIGN PROCEDURES 3.5-16 3.5.3.1 Tornado Missile Barrier Design 3.5-16 Procedures 3.5.3.2 Barrier Design Procedures for Inter- 3.5-16 nally Generated Missiles 3.

5.4 REFERENCES

3.5-16 3.6 PROTECTION AGAINST THE DYNAMIC EFFECTS 3.6-1 ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING 3.6.1 POSTULATED PIPING FAILURES IN FLUID 3.6-1 SYSTEMS INSIDE AND OUTSIDE CONTAINMENT 3.6.1.1 Design Bases 3.6-1 3.6.1.2 Description 3.6-5 3.6.1.3 Safety Evaluation 3.6-6 3.6.2 DETERMINATION OF BREAK LOCATIONS AND 3.6-10 DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING 3.6.2.1 Criteria Used to Define High/Moderate 3.6-10 Energy Break/Crack Locations and Configurations 3.6.2.2 Analytical Methods to Define Forcing 3.6-18 Functions and Response Models 3.6.2.3 Dynamic Analysis Methods to Verify 3.6-25 Integrity and Operability 3.6.2.4 Protective Assembly Design Criteria 3.6-36 3.6.2.5 Material to be Submitted for the 3.6-36 Operating License Review 3.

6.3 REFERENCES

3.6-38 3.0-iii Rev. 29

WOLF CREEK TABLE OF CONTENTS (Continued) Section Page 3.7(B) SEISMIC DESIGN 3.7(B)-1 3.7(B).1 SEISMIC INPUT 3.7(B)-1 3.7(B).1.1 Design Response Spectra 3.7(B)-1 3.7(B).1.2 Design Time History 3.7(B)-2 3.7(B).1.3 Critical Damping Values 3.7(B)-3 3.7(B).1.4 Supporting Media for Seismic 3.7(B)-3 Category I Structures 3.7(B).2 SEISMIC SYSTEM ANALYSIS 3.7(B)-4 3.7(B).2.1 Seismic Analysis Methods 3.7(B)-4 3.7(B).2.2 Natural Frequencies and Response 3.7(B)-4 Loads 3.7(B).2.3 Procedure Used for Modeling 3.7(B)-5 3.7(B).2.4 Soil/Structure Interaction 3.7(B)-5 3.7(B).2.5 Development of Floor Response 3.7(B)-6 Spectra 3.7(B).2.6 Three Components of Earthquake 3.7(B)-6 Motion 3.7(B).2.7 Combination of Modal Responses 3.7(B)-6 3.7(B).2.8 Interaction of Non-Seismic 3.7(B)-7 Category I Structure With Seismic Cateogry I Structures 3.7(B).2.9 Effects of Parameter Variations on 3.7(B)-7 Floor Response Spectra 3.7(B).2.10 Use of Constant Vertical Static 3.7(B)-7 Factors 3.7(B).2.11 Method Used to Account for Tor- 3.7(B)-8 sional Effects 3.7(B).2.12 Comparison of Responses 3.7(B)-8 3.7(B).2.13 Methods for Seismic Analysis of 3.7(B)-8 Dams 3.7(B).2.14 Determination of Seismic Category I 3.7(B)-8 Structure Overturning Moments 3.7(B).2.15 Analysis Procedure for Damping 3.7(B)-8 3.7(B).3 SEISMIC SUBSYSTEM ANALYSIS 3.7(B)-8 3.7(B).3.1 Seismic Analysis Methods 3.7(B)-8 3.7(B).3.2 Determination of Number of Earth- 3.7(B)-8 quake Cycles 3.7(B).3.3 Procedure Used for Modeling 3.7(B)-9 3.0-iv Rev. 29

WOLF CREEK TABLE OF CONTENTS (Continued) Section Page 3.7(B).3.4 Basis for Selection of Frequencies 3.7(B)-9 3.7(B).3.5 Use of Equivalent Static Load 3.7(B)-9 Method of Analysis 3.7(B).3.6 Three Components of Earthquake 3.7(B)-9 Motion 3.7(B).3.7 Combination of Modal Responses 3.7(B)-9 3.7(B).3.8 Analytical Procedures for Piping 3.7(B)-10 3.7(B).3.9 Multiple Supported Equipment and 3.7(B)-10 Components With Distinct Inputs 3.7(B).3.10 Use of Constant Vertical Static 3.7(B)-11 Factors 3.7(B).3.11 Torsional Effects of Eccentric Masses 3.7(B)-11 3.7(B).3.12 Buried Seismic Category I Piping 3.7(B)-11 Systems & Tunnels 3.7(B).3.13 Interaction of Other Piping With 3.7(B)-11 Seismic Category I Piping 3.7(B).3.14 Seismic Analyses for Reactor 3.7(B)-11 Internals 3.7(B).3.15 Analysis Procedure for Damping 3.7(B)-11 3.7(B).3.16 Seismic Analysis for Cable Trays 3.7(B)-11 3.7(B).4 SEISMIC INSTRUMENTATION 3.7(B)-16 3.7(B).4.1 Comparison with Regulatory Guide 3.7(B)-16 1.12 Rev. 2 (March 1997) 3.7(B).4.2 Location and Description of Instru- 3.7(B)-17 mentation 3.7(B).4.3 Control Room Operator Notification 3.7(B)-19 3.7(B).4.4 Comparison of Measured and Pre- 3.7(B)-19 dicted Responses 3.7(B).5 REFERENCES 3.7(B)-19 App. 3.7(B)A IMPEDANCE FUNCTIONS FOR A RIGID 3.7(B)A-1 CIRCULAR FOUNDATION ON A LAYERED VISCOELASTIC MEDIUM A.1 FORMULATION OF THE PROBLEM 3.7(B)A-1 A.1.1 STATEMENT OF THE PROBLEM 3.7(B)A-1 A.1.2 TYPES OF ENERGY DISSIPATION 3.7(B)A-3 A.1.3 INTEGRAL REPRESENTATION 3.7(B)A-5 A.2 INTEGRAL EQUATIONS AND IMPEDANCE FUNCTIONS 3.7(B)A-7 A.3 NUMERICAL SOLUTION 3.7(B)A-11 A.4 REFERENCES 3.7(B)A-12 3.0-v Rev. 29

WOLF CREEK TABLE OF CONTENTS (Continued) Section Page App. 3.7(B)B SOIL DEPENDENT DISPLACEMENT 3.7(B)B-1 FUNCTIONS FOR THE SOLUTION OF THE EQUATIONS OF MOTION 3.7(N) SEISMIC DESIGN 3.7(N)-1 3.7(N).1 SEISMIC INPUT 3.7(N)-1 3.7(N).1.1 Design Response Spectra 3.7(N)-1 3.7(N).1.2 Design Time History 3.7(N)-1 3.7(N).1.3 Critical Damping Values 3.7(N)-2 3.7(N).1.4 Supporting Media for Seismic 3.7(N)-3 Category I Structures 3.7(N).2 SEISMIC SYSTEM ANALYSIS 3.7(N)-3 3.7(N).2.1 Seismic Analysis Methods 3.7(N)-3 3.7(N).2.2 Natural Frequencies and Response 3.7(N)-13 Loads 3.7(N).2.3 Procedures Used for Modeling 3.7(N)-13 3.7(N).2.4 Soil/Structure Interaction 3.7(N)-13 3.7(N).2.5 Development of Floor Response 3.7(N)-13 Spectra 3.7(N).2.6 Three Components of Earthquake 3.7(N)-13 Motion 3.7(N).2.7 Combination of Modal Response 3.7(N)-14 3.7(N).2.8 Interaction of Non-Category I 3.7(N)-16 Structures With Seismic Category I Structures 3.7(N).2.9 Effects of Parameter Variations on 3.7(N)-17 Floor Response Spectra 3.7(N).2.10 Use of Constant Vertical Static 3.7(N)-17 Factors 3.7(N).2.11 Methods Used to Account for Tor- 3.7(N)-17 sional Effects 3.7(N).2.12 Comparison of Responses 3.7(N)-17 3.7(N).2.13 Methods for Seismic Analysis of 3.7(N)-17 Dams 3.7(N).2.14 Determination of Seismic Category I 3.7(N)-17 Structure Overturning Moments 3.7(N).2.15 Analysis Procedure for Damping 3.7(N)-17 3.7(N).3 SEISMIC SUBSYSTEM ANALYSIS 3.7(N)-17 3.7(N).3.1 Seismic Analysis Methods 3.7(N)-17 3.0-vi Rev. 29

WOLF CREEK TABLE OF CONTENTS (Continued) Section Page 3.7(N).3.2 Determination of Number of Earth- 3.7(N)-18 quake Cycles 3.7(N).3.3 Procedure Used for Modeling 3.7(N)-19 3.7(N).3.4 Basis for Selection of Frequencies 3.7(N)-19 3.7(N).3.5 Used of Equivalent Static Load 3.7(N)-19 Method of Analysis 3.7(N).3.6 Three Components of Earthquake Motion 3.7(N)-20 3.7(N).3.7 Combination of Modal Responses 3.7(N)-20 3.7(N).3.8 Analytical Procedures for Piping 3.7(N)-20 3.7(N).3.9 Multiple Supported Equipment Components 3.7(N)-20 3.7(N).3.10 Use of Constant Vertical Static 3.7(N)-21 Factors 3.7(N).3.11 Torsional Effects of Eccentric 3.7(N)-21 Masses 3.7(N).3.12 Buried Seismic Category I Piping 3.7(N)-21 Systems and Tunnels 3.7(N).3.13 Interaction of Other Piping With 3.7(N)-22 Seismic Category I Piping 3.7(N).3.14 Seismic Analyses for Reactor 3.7(N)-22 Internals 3.7(N).3.15 Analysis Procedure for Damping 3.7(N)-22 3.7(N).4 SEISMIC INSTRUMENTATION 3.7(N)-23 3.7(N).5 REFERENCES 3.7(N)-23 3.7(S) SEISMIC DESIGN 3.7(S)-1 3.7(S).1. SEISMIC INPUT 3.7(S)-1 3.7(S).1.1 Design Response Spectra 3.7(S)-1 3.7(S).1.2 Design Time History 3.7(S)-1 3.7(S).1.3 Supporting Media for Seismic Category I 3.7(S)-2 Structure 3.7(S).2 SEISMIC SYSTEM ANALYSIS 3.7(S)-2 3.7(S).2.1 Seismic Analysis Method 3.7(S)-2 3.7(S).2.2 Natural Frequencies and Response Loads 3.7(S)-2 3.7(S).2.3 Soil/Structure Interaction 3.7(S)-3 3.7(S).2.4 Methods for Seismic Analysis of Dams 3.7(S)-3 3.0-vii Rev. 29

WOLF CREEK TABLE OF CONTENTS (Continued) Section Page 3.8 DESIGN OF CATEGORY I STRUCTURES 3.8-1 3.8.1 CONCRETE CONTAINMENT 3.8-1 3.8.1.1 Description of the Reactor Building 3.8-1 3.8.1.2 Applicable Codes, Standards, and 3.8-4 Specifications 3.8.1.3 Loads and Loading Combinations 3.8-6 3.8.1.4 Design and Analysis Procedures 3.8-7 3.8.1.5 Structural Acceptance Criteria 3.8-10 3.8.1.6 Materials, Quality Control, and Special 3.8-11 Construction Techniques 3.8.1.7 Testing and Inservice Surveillance 3.8-26 Requirements 3.8.2 CONTAINMENT SYSTEM STEEL ITEMS 3.8-27 3.8.2.1 Description of Steel Items 3.8-27 3.8.2.2 Applicable Codes, Standards, and 3.8-30 Specifications 3.8.2.3 Loads and Loading Combinations 3.8-32 3.8.2.4 Design and Analysis Procedure 3.8-34 3.8.2.5 Structural Acceptance Criteria 3.8-36 3.8.2.6 Materials, Quality Control, and Special 3.8-37 Construction Techniques 3.8.2.7 Testing and Inservice Surveillance 3.8-37 Requirements 3.8.3 CONCRETE AND STEEL INTERNAL STRUCTURES 3.8-37 OF STEEL OR CONCRETE CONTAINMENTS 3.8.3.1 Description of the Internal Structures 3.8-37 3.8.3.2 Applicable Codes, Standards, and 3.8-43 Specifications 3.8.3.3 Loads and Loading Combinations 3.8-45 3.8.3.4 Design and Analysis Procedures 3.8-51 3.8.3.5 Structural Acceptance Criteria 3.8-54 3.8.3.6 Materials, Quality Control, and 3.8-54 Special Construction Techniques 3.8.3.7 Testing and Inservice Surveillance 3.8-59 Requirements 3.0-viii Rev. 29

WOLF CREEK TABLE OF CONTENTS (Continued) Section Page 3.8.4 OTHER CATEGORY I STRUCTURES 3.8-59 3.8.4.1 Description of the Structures 3.8-59 3.8.4.2 Applicable Codes, Standards, and 3.8-65 Specifications 3.8.4.3 Loads and Load Combinations 3.8-66 3.8.4.4 Design and Analysis Procedures 3.8-66 3.8.4.5 Structural Acceptance Criteria 3.8-71 3.8.4.6 Materials, Quality Control, and 3.8-71 Special Construction Techniques 3.8.4.7 Testing and Inservice Surveillance 3.8-72 Requirements 3.8.5 FOUNDATIONS 3.8-72 3.8.5.1 Description of the Foundations 3.8-72 3.8.5.2 Applicable Codes, Standards, and 3.8-76 Specifications 3.8.5.3 Loads and Load Combinations 3.8-76 3.8.5.4 Design and Analysis Procedures 3.8-76 3.8.5.5 Structural Acceptance Criteria 3.8-77 3.8.5.6 Materials, Quality Control, and Special 3.8-77 Construction Techniques 3.8.5.7 Testing and Inservice Surveillance 3.8-77 Requirements 3.8.6 RADWASTE BUILDING AND TUNNEL 3.8-77 3.8.6.1 Description of the Structures 3.8-77 3.8.6.2 Applicable Codes, Standards and 3.8-78 Specifications 3.8.6.3 Loads and Load Combinations 3.8-79 3.8.6.4 Design and Analysis Procedures 3.8-79 3.8.6.5 Structural Acceptance Criteria 3.8-81 3.8.6.6 Materials, Quality Control, and Special 3.8-81 Construction Techniques 3.8.6.7 Testing and Inservice Surveillance 3.8-82 Requirements 3.

8.7 REFERENCES

3.8-82 App. 3.8A COMPUTER PROGRAMS USED FOR STRUCTURAL 3.8A-1 AND SEISMIC ANALYSES 3.0-ix Rev. 29

WOLF CREEK TABLE OF CONTENTS (Continued) Section Page 3.8A.1 COMPUTER PROGRAMS USED FOR STRUCTURAL 3.8A-3 AND SEISMIC ANALYSES BY BECHTEL POWER CORPORATION 3.8A.1.1 Bechtel CE 201 Bechtel Structural Analysis 3.8A-3 Program - Post Processor (BSAP-POST) 3.8A.1.2 Bechtel CE 239 Hemispherical Dome Tendon 3.8A-4 Analysis (TENDON) 3.8A.1.3 Bechtel CE 309, Structural Engineering 3.8A-5 Systems Solver (STRESS) 3.8A.1.4 Bechtel CE 316, Finite Element Stress 3.8A-5 Analysis (FINEL) 3.8A.1.5 Bechtel CE 400, Concrete Column Design 3.8A-6 (PCACOL) 3.8A.1.6 Bechtel CE 639, Hemispherical Dome Tendon 3.8A-7 Analysis (STRESS) 3.8A.1.7 Bechtel CE 779, Structural Analysis 3.8A-8 Program (SAP) 3.8A.1.8 Bechtel CE 786, Ground Spectrum Raise 3.8A-9 3.8A.1.9 Bechtel CE 798, Engineering Analysis 3.8A-9 System (ANSYS) 3.8A.1.10 Bechtel CE 800, Bechtel Structural 3.8A-10 Analysis Program (BSAP) 3.8A.1.11 Bechtel CE 801, Finite Element Stress 3.8A-12 Analysis (FINEL) 3.8A.1.12 Bechtel CE 802, Response Spectra Analysis 3.8A-13 (SPECTRA) 3.8A.1.13 Bechtel CE 803, Axisymmetric Shell and 3.8A-13 Solid Computer Program (ASHSD) 3.8A.1.14 Bechtel CE 901, The Structural Design 3.8A-14 Language (ICE STRUDL) 3.8A.1.15 Bechtel CE 915, A Computer Program for 3.8A-15 Earthquake Response Analysis of Horizontally Layered Sites (SHAKE) 3.8A.1.16 Bechtel CE 917, Modal Dynamic Analysis 3.8A-16 3.8A.1.17 Bechtel CE 918, Response Spectrum 3.8A-16 Analysis 3.8A.1.18 Bechtel CE 920, Time-History Analysis 3.8A-17 of Structures 3.8A.1.19 Bechtel CE 921, Response Spectrum 3.8A-17 Calculations 3.8A.1.20 Bechtel CE 933, Fourier Analysis of 3.8A-18 Soils (FASS) 3.8A.1.21 Bechtel CE 935, Earthquake Acceleration 3.8A-18 Time-Histories 3.0-x Rev. 0

WOLF CREEK TABLE OF CONTENTS (Continued) Section Page 3.8A.1.22 Bechtel CE 970, Impedance Functions for a 3.8A-19 Rigid Circular Foundation on a Layered Viscoelastic Medium (LUCON) 3.8A.1.23 Computer Programs for Seismic Soil- 3.8A-20 Structure Interaction Analysis 3.8A.1.24 DISCOM, a FLUSH Postprocessor (Control 3.8A-21 Data Corp. Version) 3.8A.1.25 The Structural Design Language (ICES- 3.8A-22 STRUDL by McDonnell-Douglas Automation Version) 3.8A.1.26 Other Computer Programs Used in Structural 3.8A-22 Analysis 3.8A.2 COMPUTER PROGRAMS USED FOR STRUCTURAL 3.8A-23 ANALYSES BY SUPPLIERS 3.8A.2.1 INRYCO, Nuclear Force Computation (NUCFOR) 3.8A-23 3.8A.2.2 CBI Program 7-81, Shells of Revolution 3.8A-23 3.8A.2.3 CBI Program 1027, Stress Intensities at 3.8A-24 Loaded Attachments for Spheres of Cylinders with Round or Square Attachment 3.8A.2.4 CBI Program 1691 3.8A-25 3.9(B) MECHANICAL SYSTEMS AND COMPONENTS 3.9(B)-1 3.9(B).1 SPECIAL TOPICS FOR MECHANICAL COM- 3.9(B)-1 PONENTS 3.9(B).1.1 Design Transients 3.9(B)-1 3.9(B).1.2 Computer Programs Used in Analyses 3.9(B)-1 3.9(B).1.3 Experimental Stress Analysis 3.9(B)-4 3.9(B).1.4 Considerations for the Evaluation 3.9(B)-4 of the Faulted Condition 3.9(B).2 DYNAMIC TESTING AND ANALYSIS 3.9(B)-5 3.9(B).2.1 Piping Vibration, Thermal Expansion 3.9(B)-5 Dynamic Effects 3.9(B).2.2 Seismic Qualification Testing of 3.9(B)-8 Safety-Related Mechanical Equipment 3.9(B).2.3 Dynamic Response Analysis of 3.9(B)-9 Reactor Internals Under Operational Flow Transients and Steady State Conditions 3.0-xi Rev. 29

WOLF CREEK TABLE OF CONTENTS (Continued) Section Page 3.9(B).2.4 Preoperational Flow Induced Vibra- 3.9(B)-9 tion Testing of Reactor Internals 3.9(B).2.5 Dynamic System Analysis of the 3.9(B)-9 Reactor Internals Under Faulted Condition 3.9(B).2.6 Correlations of Reactor Internals 3.9(B)-9 Vibration Tests With Analytical Results 3.9(B).3 ASME SECTION III CLASS 1, 2, and 3 3.9(B)-10 COMPONENTS, COMPONENT SUPPORTS, AND CORE SUPPORT STRUCTURES 3.9(B).3.1 Loading Combinations, Design 3.9(B)-10 Transients, and Stress Limits 3.9(B).3.2 Pump and Valve Operability Assur- 3.9(B)-11 ance 3.9(B).3.3 Design and Installation Details 3.9(B)-16 for Mounting of Pressure Relief Devices 3.9(B).3.4 Component Supports 3.9(B)-17 3.9(B).4 CONTROL ROD DRIVE SYSTEMS 3.9(B)-20 3.9(B).5 REACTOR PRESSURE VESSEL INTERNALS 3.9(B)-20 3.9(B).6 INSERVICE TESTING OF PUMPS AND 3.9(B)-21 VALVES 3.9(B).6.1 Inservice Testing of Pumps 3.9(B)-21 3.9(B).6.2 Inservice Test of Valves 3.9(B)-21 3.9(B).7 REFERENCES 3.9(B)-21 App. 3.9(B)A ME-632 VERIFICATION REPORT 3.9(B)A-1 3.9(N) MECHANICAL SYSTEMS AND COMPONENTS 3.9(N)-1 3.9(N).1 SPECIAL TOPICS FOR MECHANICAL COM- 3.9(N)-1 PONENTS 3.9(N).1.1 Design Transient 3.9(N)-1 3.9(N).1.2 Computer Programs Used in Analysis 3.9(N)-20 3.9(N).1.3 Experimental Stress Analysis 3.9(N)-21 3.9(N).1.4 Considerations for the Evaluation 3.9(N)-21 of the Faulted Condition 3.0-xii Rev. 29

WOLF CREEK TABLE OF CONTENTS (Continued) Section Page 3.9(N).2 DYNAMIC TESTING AND ANALYSIS 3.9(N)-35 3.9(N).2.1 Preoperational Vibration and Dyna- 3.9(N)-35 mic Effects Testing on Piping 3.9(N).2.2 Seismic Qualification Testing of 3.9(N)-36 Safety-Related Mechanical Equipment 3.9(N).2.3 Dynamic Response Analysis of 3.9(N)-37 Reactor Internals Under Operational Flow Transients and Steady State Conditions 3.9(N).2.4 Preoperational Flow-Induced Vibra- 3.9(N)-39 tion Testing of Reactor Internals 3.9(N).2.5 Dynamic System Analysis of the 3.9(N)-41 Reactor Internals Under Faulted Conditions 3.9(N).2.6 Correlations of Reactor Internals 3.9(N)-47 Vibration Tests With the Analy-tical Results 3.9(N).3 ASME CODE CLASS 1, 2 AND 3 COM- 3.9(N)-47 PONENTS, COMPONENT SUPPORTS AND CORE SUPPORT STRUCTURES 3.9(N).3.1 Loading Combinations Design Tran- 3.9(N)-49 sients, and Stress Limits (For ASME Code Class 2 and 3 Components) 3.9(N).3.2 Pump and Valve Operability Assurance 3.9(N)-50 3.9(N).3.3 Design and Installation Details 3.9(N)-55 in Mounting of Pressure Relief Devices 3.9(N).3.4 Component Supports (ASME Code Class 3.9(N)-55 2 and 3) 3.9(N).4 CONTROL ROD DRIVE SYSTEM (CRDS) 3.9(N)-56 3.9(N).4.1 Descriptive Information of CRDS 3.9(N)-56 3.9(N).4.2 Applicable CRDS Design Specifications 3.9(N)-61 3.9(N).4.3 Design Loads, Stress Limits, and 3.9(N)-63 Allowable Deformations 3.9(N).4.4 CRDS Performance Assurance Program 3.9(N)-65 3.0-xiii Rev. 29

WOLF CREEK TABLE OF CONTENTS (Continued) Section Page 3.9(N).5 REACTOR PRESSURE VESSEL INTERNALS 3.9(N)-67 3.9(N).5.1 Design Arrangements 3.9(N)-67 3.9(N).5.2 Design Loading Conditions 3.9(N)-71 3.9(N).5.3 Design Loading Categories 3.9(N)-72 3.9(N).5.4 Design Bases 3.9(N)-73 3.9(N).6 INSERVICE TESTING OF PUMPS AND VALVES 3.9(N)-75 3.9(N).7 REFERENCES 3.9(N)-75 3.10(B) SEISMIC QUALIFICATION OF CATEGORY I 3.10(B)-1 INSTRUMENTATION AND ELECTRICAL EQUIP-MENT 3.10(B).1 SEISMIC QUALIFICATION CRITERIA 3.10(B)-1 3.10(B).2 METHODS AND PROCEDURES FOR QUALI- 3.10(B)-2 FYING ELECTRICAL EQUIPMENT AND INSTRUMENTATION 3.10(B).2.1 Analysis 3.10(B)-2 3.10(B).2.2 Testing 3.10(B)-2 3.10(B).2.3 Combined Analysis and Testing 3.10(B)-2 3.10(B).2.4 Generic Qualification 3.10(B)-2 3.10(B).3 METHODS AND PROCEDURES OF ANALYSIS 3.10(B)-3 OR TESTING OF SUPPORTS OF ELECTRICAL EQUIPMENT 3.10(B).4 METHODS AND PROCEDURES OF ANALYSIS 3.10(B)-4 OR TESTING OF INSTRUMENTATION PANELS, MOUNTING STRUCTURES FOR FIELD MOUNTED INSTRUMENTS, AND SUPPORTS FOR INSTRU-MENT TUBING 3.10(B).4.1 Instrumentation Panels 3.10(B)-4 3.10(B).4.2 Mounting Structures for Field 3.10(B)-4 Mounted Instruments 3.10(B).4.3 Supports for Instrument Tubing 3.10(B)-4 3.10(B).5 OPERATING LICENSE REVIEW 3.10(B)-5 3.10(N) SEISMIC QUALIFICATION OF SEISMIC CATEGORY I 3.10(N)-1 INSTRUMENTATION AND ELECTRICAL EQUIPMENT 3.10(N).1 SEISMIC QUALIFICATION CRITERIA 3.10(N)-1 3.0-xiv Rev. 29

WOLF CREEK TABLE OF CONTENTS (Continued) Section Page 3.10(N).1.1 Qualification Standards 3.10(N)-1 3.10(N).1.2 Performance Requirements for Seismic 3.10(N)-2 Qualification 3.10(N).1.3 Acceptance Criteria 3.10(N)-2 3.10(N).2 METHODS AND PROCEDURES FOR QUALIFYING 3.10(N)-3 ELECTRICAL EQUIPMENT AND INSTRUMENTATION 3.10(N).2.1 Seismic Qualification by Type Test 3.10(N)-3 3.10(N).2.2 Seismic Qualification by Analysis 3.10(N)-4 3.10(N).3 METHODS AND PROCEDURES FOR QUALIFYING 3.10(N)-5 SUPPORTS OF ELECTRICAL EQUIPMENT AND INSTRUMENTATION 3.10(N).4 OPERATING LICENSE REVIEW 3.10(N)-5 3.10(N).5 REFERENCES 3.10(N)-5 3.11(B) ENVIRONMENTAL DESIGN OF MECHANICAL AND 3.11(B)-1 ELECTRICAL EQUIPMENT 3.11(B).1 EQUIPMENT IDENTIFICATION AND EVIR- 3.11(B)-2 RONMENTAL CONDITIONS 3.11(B).1.1 Equipment and Systems List 3.11(B)-2 3.11(B).1.2 Plant Environments 3.11(B)-4 3.11(B).1.3 Voltage and Frequency 3.11(B)-14 3.11(B).1.4 Environment Design Criteria 3.11(B)-15 3.11(B).2 QUALIFICATION TESTS AND ANALYSES 3.11(B)-16 3.11(B).2.1 Equipment Inside Containment 3.11(B)-17 3.11(B).2.2 Auxiliary and Fuel Building 3.11(B)-20 Equipment 3.11(B).2.3 Control Building Equipment 3.11(B)-21 3.11(B).2.4 Essential Service Water Pump House 3.11(B)-22 3.11(B).2.5 Equipment Located Outside of 3.11(B)-22 Buildings 3.11(B).3 QUALIFICATION TEST RESULTS 3.11(B)-23 3.11(B).4 LOSS OF VENTILATION 3.11(B)-23 3.11(B).5 NUREG-0588 PROGRAM REQUIREMENTS 3.11(B)-24 3.11(B).5.1 Display Instrumentation 3.11(B)-24 3.11(B).5.2 Equipment Operability 3.11(B)-24 3.11(B).5.3 Margins 3.11(B)-25 3.11(B).5.4 Aging 3.11(B)-25 3.11(B).5.5 Exemption from Qualification 3.11(B)-26 3.11(B).5.6 Maintenance and Surveillance Activities 3.11(B)-27 3.0-xv Rev. 29

WOLF CREEK TABLE OF CONTENTS (Continued) Section Page 3.11(B).5.7 Equipment Located in Mild Environments 3.11(B)-29 3.11(B).5.8 Synergistic Effects 3.11(B)-29 3.11(B).6 MECHANICAL EQUIPMENT QUALIFICATION 3.11(B)-31 3.11(B).7 CONTROL SYSTEMS QUALIFICATION 3.11(B)-32 3.11(B).8 REFERENCES 3.11(B)-35 3.11(N) ENVIRONMENTAL DESIGN OF MECHANICAL AND 3.11(N)-1 ELECTRICAL EQUIPMENT 3.11(N).1 EQUIPMENT IDENTIFICATION AND ENVI- 3.11(N)-1 RONMENTAL CONDITIONS 3.11(N).2 QUALIFICATION TESTS AND ANALYSES 3.11(N)-1 3.11(N).3 QUALIFICATION TEST RESULTS 3.11(N)-2 3.11(N).4 LOSS OF VENTILATION 3.11(N)-2 3.11(N).5 ESTIMATED CHEMICAL AND RADIATION 3.11(N)-2 ENVIRONMENT 3.11(N).6 REFERENCES 3.11(N)-2 3.12 COMPUTER PROGRAMS FOR STRUCTURAL 3.12-1 ANALYSIS AND DESIGN 3.12.1 ISBILD 3.12-1 3.12.2 QUAD4 3.12-2 3.12.3 RSG 3.12-4 3.12.4 SEEPAGE 3.12-5 3.12.5 SLOPE 3.12-6 3.12.6 BISHOP 3.12-6 3.

12.7 REFERENCES

3.12-8 App. 3A CONFORMANCE TO NRC REGULATORY GUIDES 3A-1 App. 3B HAZARD ANALYSIS 3B-1 3B.1 INTRODUCTION 3B-1 3B.2 ANALYSIS ASSUMPTIONS 3B-2 3B.2.1 EARTHQUAKE ANALYSIS ASSUMPTIONS 3B-3 3B.2.2 PIPE BREAK ANALYSIS ASSUMPTIONS 3B-3 3B.2.3 MISSILES ANALYSIS ASSUMPTIONS 3B-3 3B.2.4 FLOODING ANALYSIS ASSUMPTIONS 3B-4 3B.3 PROTECTION MECHANISMS 3B-4 3B.4 HAZARDS EVALUATIONS 3B-5 3B.4.1 AUXILIARY FEEDWATER PUMP ROOMS 3B-5 3B.4.2 MAIN STEAM/MAIN FEEDWATER ISOLATION 3B-6 VALVE COMPARTMENT 3.0-xvi Rev. 29

WOLF CREEK TABLE OF CONTENTS (Continued) Section Page 3B.4.2.1 Break Size and Location 3B-6 3B.4.2.2 Method of Analysis 3B-7 3B.4.2.3 Mass and Energy Release 3B-7 3B.4.2.4 Compartment Volumes and Vent Areas 3B-10 3B.4.2.5 Initial Conditions 3B-10 3B.4.2.6 Results 3B-11 3B.4.2.7 Design Provisions 3B-13 3B.4.3 TURBINE BUILDING FLOODING EVALUATION 3B-13 3B.4.3.1 Introduction 3B-14 3B.4.3.2 CWS Rupture Analysis 3B-15 3B.4.3.3 CWS Rupture Evaluation 3B-15 3B.4.4 EVALUATION OF RCS LOOP BRANCH LINE 3B-16 BREAKS 3B.5 REFERENCES 3B-16 App. 3C SEISMIC EVALUATION OF WOLF CREEK 3C-1 GENERATING STATION STRUCTURES USING LIVERMORE SPECTRUM 3C.1 EVALUATION OF STRUCTURES 3C-1 3C.

1.1 INTRODUCTION

3C-1 3C.1.2 DISCUSSION 3C-1 3C.

1.3 CONCLUSION

3C-8 3C.

1.4 REFERENCES

3C-8 3C.2 EVALUATION OF PIPING SYSTEMS AND 3C-8 SUPPORTS 3C.

2.1 INTRODUCTION

3C-8 3C.2.2 STRESS EVALUATION OF PIPING SYSTEMS 3C-9 3C.2.3 STRESS EVALUATION RESULTS 3C-9 3C.

2.4 CONCLUSION

S 3C-10 3C.3 EVALUATION OF EQUIPMENT 3C-10 3C.

3.1 INTRODUCTION

3C-10 3C.3.2 EVALUATION OF SAFETY-RELATED EQUIPMENT 3C-10 3C.3.3 EQUIPMENT EVALUATION RESULTS 3C-11 3C.

3.4 CONCLUSION

3C-14 App. 3D WOLF CREEK RESPONSE TO BEYOND - DESIGN - 3D-1 BASIS EXTERNAL EVENT FUKUSHIMA RELATED REQUIRED ACTIONS 3.0-xvii Rev. 29

WOLF CREEK TABLE OF CONTENTS (CONTINUED) LIST OF TABLES Table No. Title 3.2-1 Classifications of Structures, Components, and Systems 3.2-2 Code Requirements for Components and Quality Groups 3.2-3 Design Comparison to Regulatory Positions of Regulatory Guide 1.29, Revision 3, Dated September 1978, Titled "Seismic Design Classification" 3.2-4 Design Comparison to Regulatory Guide 1.26, Revision 3, Dated February 1976, Titled "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste Containing Components of Nuclear Power Plants" 3.2-5 Design Comparison to Regulatory Guide 1.143, for Comments Dated July 1978, Titled "Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants" 3.3-1 Tornado Resistant Buildings and Enclosures - Standard Plant 3.4-1 PMF, Groundwater, Reference, and Actual Plant Elevations 3.4-2 Site Related, Category I Structures with Penetrations Below Ground Water Elevations 3.4-3 Wind-Generated Wave Data for the ESWS Pumphouse 3.5-1 Characteristics of Postulated Tornado Missiles 3.5-2 Structures Providing Tornado Missile Barrier Protection 3.5-3 Turbine Missile Probabilities High Energy, High and Low Trajectory 3.5-4 Range of Maximum and Minimum Turbine Missile Velocities 3.5-5 NUREG-0121 Applicability to Diesel Stack Design 3.6-1 Safety-Related Systems and High and Moderate Energy Systems 3.0-xviii Rev. 29

WOLF CREEK TABLE OF CONTENTS (Continued) LIST OF TABLES Table No. Title 3.6-2 Design Comparison to Regulatory Positions of Regulatory Guide 1.46, Revision 0, Dated May 1973, Titled "Protection of Pipe Whip Inside Containment" 3.6-3 High-Energy Pipe Break Initial Stress Analysis Results 3.6-4 High-Energy Pipe Break Effects Analysis Results 3.6-5 Primary Plus Secondary Stress Intensity Ranges and Cumulative Usage Factors at Design Break Locations in the Reactor Coolant Loop 3.6-6 Summary of Flood Levels in All Safety-Related Rooms 3.7(B)-1 Damping Values for Category I Structures, Systems, and Components (Percent of Critical Damping) 3.7(B)-2 Depth of Soil Deposited Over Bedrock Major Category I Structures 3.7(B)-3 Foundation Depth Below Grade, Minimum Base Dimension and Method of Analysis for Category I Structures 3.7(B)-4 Summary First Mode Natural Frequencies (Hertz) 3.7(B)-5A Spectrum Response Summary Tables through 8AB 3.7(B)-9 Design Comparison with R. G. 1.12, Revision 1, Dated April 1974, Titled "Instrumentation For Earthquakes" 3.7(N)-1 Damping Values Used for Seismic Systems Analysis for Westinghouse-Supplied Equipment 3.7(S)-1 Depth of Soil Deposited over Bedrock - Site Related Seismic Category I Structures 3.7(S)-2 Foundation Depth, Below Grade Minimum Base Dimension and Methods of Analysis for Site Related Seismic Category I Structures 3.7(S)-3A Spectral Response Summary - ESWS Pumphouse, 0.12GSSE 3.0-xix Rev. 29

WOLF CREEK TABLE OF CONTENTS (CONTINUED) LIST OF TABLES Table No. Title 3.7(S)-3B Spectral Response Summary - ESWS Pumphouse, 0.06G0BE 3.8-1 Control Tests for Concrete 3.8-2 Maximum Allowable Offset in Final Welded Joints of Reactor Building Liner Plate 3.8-3 Stress Limits for Steel Portions of Concrete Containments Designed in Accordance with Subsection NE of the ASME Code 3.8-4 General Design Live Loads 3.8-5 Load Combinations and Load Factors for Category I Concrete Structures 3.8-6 Load Combinations and Load Factors for Seismic Category I Concrete Structures 3.8-7 Load Combinations and Load Factors for Category I Steel Structures 3.8-8 Load Combinations and Load Factors for Seismic Category I Steel Structures 3.8-9 Additional Load Combinations for Sliding, Overturning and Flotation 3.9(B)-1 Computer Programs Used in Analysis 3.9(B)-2 Design Loading Combinations for ASME Code Class 2 and 3 Components 3.9(B)-3 Deleted 3.9(B)-4 Deleted 3.9(B)-5 Stress Criteria for ASME Code Class 2 and Class 3 Vessels 3.9(B)-6 Stress Criteria for ASME Code Class 1, 2, and 3 Valves (Active and Inactive) 3.9(B)-7 Design Criteria for ASME Code Class 2 and 3 Piping 3.9(B)-8 Stress Criteria for ASME Code Class 2 and Class 3 Inactive Pumps 3.0-xx Rev. 29

WOLF CREEK TABLE OF CONTENTS (CONTINUED) LIST OF TABLES Table No. Title 3.9(B)-9 Stress Criteria for ASME Code Class 2 and Class 3 Active Pumps 3.9(B)-10 Design Loading Combinations for Supports for ASME Code Class 1, 2, and 3 Components 3.9(B)-11 Allowable Stress Limits for Class 1 Component Supports 3.9(B)-12 Allowable Stress Limits for Class 2 and 3 Component Supports 3.9(B)-13 Response to Regulatory Guide 1.48 for Components Not Furnished With the NSSS 3.9(B)-14 Response to Regulatory Guide 1.124 for Components Not Furnished with the NSSS 3.9(B)-15 Active Pumps Not Furnished with the NSSS 3.9(B)-16 Active Valves 3.9(B)A-1 Summary of Maximum Deflections, Stresses, and Reactions Core Spray Piping System Monticello Nuclear Generating Plant, Unit 1 3.9(B)A-2 Summary of Maximum Deflections, Stresses, and Reactions SMUD Rancho Seco, Unit 1 Piping System 3.9(B)A-3 Comparison of Natural Periods SMUD Rancho Seco, Unit 1 Piping System 3.9(N)-1 Summary of Reactor Coolant System Design Transients 3.9(N)-2 Loading Combinations for ASME Class 1 Components and Supports (Excluding Pipe Supports) 3.9(N)-3 Allowable Stresses for ASME Code, Section III, Class 1 Components 3.9(N)-4 Design Loading Combinations for ASME Code Class 2 and 3 Components and Supports (Excluding Pipe Supports) 3.9(N)-5 Stress Criteria for Safety-Related ASME Class 2 and Class 3 Vessels 3.9(N)-6 Stress Criteria for Safety-Related Class 2 Vessels 3.0-xxi Rev. 29

WOLF CREEK TABLE OF CONTENTS (CONTINUED) LIST OF TABLES Table No. Title 3.9(N)-7 Stress Criteria for ASME Code Class 2 and Class 3 Inactive Pumps and Pump Supports 3.9(N)-8 Design Criteria for Active Pumps and Pump Supports 3.9(N)-9 Stress Criteria for Safety-Related ASME Code Class 2 and Class 3 Valves 3.9(N)-10 Active Pumps 3.9(N)-11 Active Valves 3.9(N)-12 Maximum Deflections Allowed for Reactor Internal Support Structures 3.10(B)-1 Seismic Qualification of Safety-Related (Class 1E) Equipment 3.10(N)-1 Seismic Category I Instrumentation and Electrical Equipment in Westinghouse NSSS Scope of Supply 3.11(B)-1 Plant Environmental Normal Conditions 3.11(B)-2 Environmental Qualification Parameters for SNUPPS NUREG 0588 Review (LOCA, MSLB and HELB) 3.11(B)-3 Identification of Safety-Related Equipment and Components: Equipment Qualification 3.11(B)-4 Containment Worst Case Radiation Levels 3.11(B)-5 Containment Spray Requirements 3.11(B)-6 Deleted 3.11(B)-7 Specifications Reviewed under the NUREG-0588 Program 3.11(B)-8 Exemptions from NUREG-0588 Qualification 3.11(B)-9 Safety Related System Listings 3.11(B)-10 Equipment Added for NUREG-0737 3.11(B)-11 Typical Examples of Maintenance/Surveillance Requirements 3.0-xxii Rev. 29

WOLF CREEK TABLE OF CONTENTS (CONTINUED) LIST OF TABLES Table No. Title 3.12-1 Comparison of Response Spectra Values from RSG and Brancy Et. AL.(1972) 3.12-2 Results of Problem Solved With Bishop and Ices-Slope (Bishop) 3B-1 Hazards Analysis of Auxiliary Building - Elevation 1974'-0" 3B-2 Main Steam/Main Feedwater Isolation Valve Compartment Design Parameters 3B-2A Transient Summary for the Spectrum of Steamline Breaks at 102% Power 3B-3 Mass and Energy Release Data for Main Steam Line Break in Main Steam/Main Feedwater Isolation Valve Compartment 3B-4 Mass Release Data for Main Feedwater Line Break in Main Steam/Main Feedwater Isolation Valve Compartment 3B-5 Summary of Nodalization Model 3B-6 Missiles 3B-7 Evaluation of RCS Loop Branch Line Breaks 3C-1 Site Specific Safety-Related Equipment Evaluated for a 0.15g SSE Design Spectrum 3C-2 Factors of Safety UHS Dam 3C-3 Computed Factor of Safety for the Finite Element Model of UHS Dam - Vertical and Horizontal Acceleration, 0.15g 3C-4 Results of Slope Stability Analysis for ESWS Intake Channel Excavated Slopes 3C-5 Results of Slope Stability Analysis for UHS Excavated Slopes Using Wedge Analysis 3.0-xxiii Rev. 29

WOLF CREEK CHAPTER 3 - LIST OF FIGURES

  • Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.

Figure # Sheet(s) Title Drawing #* 3.4-1 0 Typical Waterproofing Details 3.5-1 0 Turbine Missile Trajectory 3.5-2 Structural Protection for Diesel Exhaust Stacks from Tornado Missiles 3.6-1 1 High Energy Pipe Break Isometric Main Steam System Inside Containment 3.6-1 2 High Energy Pipe Break Isometric Main Steam System Inside Containment 3.6-1 3 High Energy Pipe Break Isometric Main Feedwater Inside Containment 3.6-1 4-7 Deleted 3.6-1 8 High Energy Pipe Break Isometric Reactor Coolant System Pressurizer Relief 3.6-1 9 High Energy Pipe Break Isometric Reactor Coolant System Pressurizer Relief 3.6-1 10-11 Deleted 3.6-1 12 High Energy Pipe Break Isometric Reactor Coolant System Seal Water Injection Inside Containment 3.6-1 13 High Energy Pipe Break Isometric Reactor Coolant Pump A Seal Water Injection Inside Containment 3.6-1 14 High Energy Pipe Break Isometric Reactor Coolant Pump C Seal Water Injection Inside Containment 3.6-1 15 High Energy Pipe Break Isometric Reactor Coolant Pump B Seal Water Injection Inside Containment 3.6-1 16 Deleted 3.6-1 17 Deleted 3.6-1 18 High Energy Pipe Break Isometric NCP to Regen HX CVCS - Outside Containment 3.6-1 19 High Energy Pipe Break Isometric CCP A & B Discharge CVCS - Outside Containment 3.6-1 20 High Energy Pipe Break Isometric CVCS - Letdown Outside Containment 3.6-1 21 High Energy Pipe Break Isometric CVCS - Seal Water Injection Outside Containment 3.6-1 22 High Energy Pipe Break Isometric CCP A - B Miniflow CVCS - Outside Containment 3.6-1 23 High Energy Pipe Break Isometric Letdown to Reheat HX CVCS - Outside Containment 3.6-1 24 High Energy Pipe Break Isometric Normal & Alternate Charging CVCS - Inside Containment 3.0-xxiv Rev. 29

WOLF CREEK CHAPTER 3 - LIST OF FIGURES

  • Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.

Figure # Sheet(s) Title Drawing #* 3.6-1 25 High Energy Pipe Break Isometric CVCS Letdown Inside Containment 3.6-1 26 High Energy Pipe Break Isometric Charging & Excess Letdown CVCS - Inside Containment 3.6-1 27 High Energy Pipe Break Isometric CVCS - Aux. Spray Inside Containment 3.6-1 28 Deleted 3.6-1 29 High Energy Pipe Break Isometric Stm Gen A & D Blowdown Inside Containment 3.6-1 30 High Energy Pipe Break Isometric Stm Gen B & C Blowdown Inside Containment 3.6-1 31 High Energy Pipe Break Isometric Stm Gen A, B, C, D Blowdown Inside Containment 3.6-1 32 High Energy Pipe Break Isometric Stm Gen A Sample & Tube Sheet Drain Inside Containment 3.6-1 33 High Energy Pipe Break Isometric Stm Gen B Sample & Tube Sht. Drain Inside Containment 3.6-1 34 High Energy Pipe Break Isometric Stm Gen C Sample & Tube Sht. Drain Inside Containment 3.6-1 35 High Energy Pipe Break Isometric Stm Gen D Sample & Tube Sht. Drain Inside Containment 3.6-1 36 High Energy Pipe Break Isometric RHR Suction - Loops 1 & 4 Inside Containment 3.6-1 37 High Energy Pipe Break Isometric Boron Injection Tank Inlet SIS - Outside Containment 3.6-1 38 High Energy Pipe Break Isometric BIT and SI & RHR Recirc SIS - Inside Containment 3.6-1 39 High Energy Pipe Break Isometric SI Discharge - Loops 1 & 4 SIS - Inside Containment 3.6-1 40 High Energy Pipe Break Isometric Accumulator Injection - Loops 1 & 4 Inside Containment 3.6-1 41 High Energy Pipe Break Isometric Accumulator Injection - Loops 2 & 3 Inside Containment 3.6-1 42 Deleted 3.6-1 43 High Energy Pipe Break Isometric Aux Stm Deaerator Feed Pump Disch Outside Containment 3.6-1 44 High Energy Pipe Break Isometric Aux Stm Supply Header Outside Containment 3.6-1 45 High Energy Pipe Break Isometric Aux Stm Condensate Return Outside Containment 3.0-xxv Rev. 17

WOLF CREEK CHAPTER 3 - LIST OF FIGURES

  • Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.

Figure # Sheet(s) Title Drawing #* 3.6-1 46 High Energy Pipe Break Isometric Aux Stm Cond Stor & Recov Tank Overflow & Vent Outside Containment 3.6-1 47 High Energy Pipe Break Isometric Aux Stm Cond Transfer Pump Disch Outside Containment 3.6-1 48 High Energy Pipe Break Isometric Aux Stm Deaerator Feed Pump Recirc Outside Containment 3.6-1 49 High Energy Pipe Break Isometric Main Stm Supply to Turb AFP Outside Containment 3.6-1 50 High Energy Pipe Break Isometric Reactor Coolant - Loop Drains Inside Containment 3.6-1 51 High Energy Pipe Break Isometric Accumulator Tank Drains Inside Containment 3.6-2 0 Loss of Reactor Coolant Accident Boundary Limits 3.6-3 0 Location of Postulated Breaks in Reactor Coolant Loop (Including Pressurizer Surge Line) 3.6-4 0 Typical Piping Guide Installation 3.6-5 0 Typical Isolation Restraint 3.6-6 1 Energy Absorbing Honeycomb Material - Large Gap Restraint 3.6-6 2 Typical Prefabricated Energy Absorbing Honeycomb Material Installation 3.6-7 0 Typical Upset Rod Large Gap Restraint 3.6-8 0 Typical Close Gap Restraint 3.6-9 0 Lumped-Parameter Model Pipe Restraint System 3.7(B)-1 0 SSE Horizontal Ground Spectra 0.2g 3.7(B)-2 0 SSE Vertical Ground Spectra 0.2g 3.7(B)-3 0 Synthesized Time History Horizontal (OBE and SSE) 3.7(B)-4 0 Synthesized Time History Vertical (OBE and SSE) 3.7(B)-5 0 Deleted 3.7(B)-6 0 Deleted 3.7(B)-7 0 Deleted 3.7(B)-8 0 Deleted 3.7(B)-9A 0 Typical Free-Field Base Elevation Spectra Callaway Site 3.7(B)-9B 0 Typical Free-Field Base Elevation Spectra Sterling Site 3.7(B)-9C 0 Typical Free-Field Base Elevation Spectra Tyrone Site 3.7(B)-9D 0 Typical Free-Field Base Elevation Spectra Wolf Creek Site 3.7(B)-10 0 Typical Free-Field Base Elevation Spectra Three Site Envelope 3.0-xxvi Rev. 17

WOLF CREEK CHAPTER 3 - LIST OF FIGURES

  • Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.

Figure # Sheet(s) Title Drawing #* 3.7(B)-11A 0 Deleted 3.7(B)-11B 0 Deleted 3.7(B)-12 0 Mathematical Model for Reactor Building and Internal Structures 3.7(B)-13 0 The Finite-Element Model 3.7(B)-14A 0 Spectra - Containment Building SSE, North-South Direction, Polar Crane Location, Callaway Site 3.7(B)-14B 0 Spectra - Containment Building SSE, North-South Direction, Polar Crane Location, Sterling Site 3.7(B)-14C 0 Deleted 3.7(B)-14D 0 Spectra - Containment Building SSE, North-South Direction, Polar Crane Location, Wolf Creek Site 3.7(B)-14E 0 Spectra - Containment Building SSE, East-West Direction, Polar Crane Location, Callaway Site 3.7(B)-14F 0 Spectra - Containment Building SSE, East-West Direction, Polar Crane Location, Sterling Site 3.7(B)-14G 0 Deleted 3.7(B)-14H 0 Spectra - Containment Building SSE, East-West Direction, Polar Crane Location, Wolf Creek Site 3.7(B)-14I 0 Spectra - Containment Building SSE, Vertical Direction, Polar Crane Location, Callaway Site 3.7(B)-14J 0 Spectra - Containment Building SSE, Vertical Direction, Polar Crane Location, Sterling Site 3.7(B)-14K 0 Deleted 3.7(B)-14L 0 Spectra - Containment Building SSE, Vertical Direction, Polar Crane Location, Wolf Creek Site 3.7(B)-14M 0 Spectra - Containment Building OBE, North-South Direction, Polar Crane Location, Callaway Site 3.7(B)-14N 0 Spectra - Containment Building OBE, North-South Direction, Polar Crane Location, Sterling Site 3.7(B)-14O 0 Deleted 3.7(B)-14P 0 Spectra - Containment Building OBE, North-South Direction, Polar Crane Location, Wolf Creek Site 3.7(B)-14Q 0 Spectra - Containment Building OBE, East-West Direction, Polar Crane Location, Callaway Site 3.7(B)-14R 0 Spectra - Containment Building OBE, East-West Direction, Polar Crane Location, Sterling Site 3.7(B)-14S 0 Deleted 3.7(B)-14T 0 Spectra - Containment Building OBE, East-West Direction, Polar Crane Location, Wolf Creek Site 3.0-xxvii Rev. 17

WOLF CREEK CHAPTER 3 - LIST OF FIGURES

  • Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.

Figure # Sheet(s) Title Drawing #* 3.7(B)-14U 0 Spectra - Containment Building OBE, Vertical Direction, Polar Crane Location, Callaway Site 3.7(B)-14V 0 Spectra - Containment Building OBE, Vertical Direction, Polar Crane Location, Sterling Site 3.7(B)-14W 0 Deleted 3.7(B)-14X 0 Spectra - Containment Building OBE, Vertical Direction, Polar Crane Location, Wolf Creek Site 3.7(B)-15A 0 Spectra - Containment Building SSE, North-South Direction, Steam Generator Upper Support, Callaway Site 3.7(B)-15B 0 Spectra - Containment Building SSE, North-South Direction, Steam Generator Upper Support, Sterling Site 3.7(B)-15C 0 Deleted 3.7(B)-15D 0 Spectra - Containment Building SSE, North-South Direction, Steam Generator Upper Support, Wolf Creek Site 3.7(B)-15E 0 Spectra - Containment Building SSE, East-West Direction, Steam Generator Upper Support, Callaway Site 3.7(B)-15F 0 Spectra - Containment Building SSE, East-West Direction, Steam Generator Upper Support, Sterling Site 3.7(B)-15G 0 Deleted 3.7(B)-15H 0 Spectra - Containment Building SSE, East-West Direction, Steam Generator Upper Support, Wolf Creek Site 3.7(B)-15I 0 Spectra - Containment Building SSE, Vertical Direction, Steam Generator Upper Support, Callaway Site 3.7(B)-15J 0 Spectra - Containment Building SSE, Vertical Direction, Steam Generator Upper Support, Sterling Site 3.7(B)-15K 0 Deleted 3.7(B)-15L 0 Spectra - Containment Building SSE, Vertical Direction, Steam Generator Upper Support, Wolf Creek Site 3.7(B)-15M 0 Spectra - Containment Building OBE, North-South Direction, Steam Generator Upper Support, Callaway Site 3.0-xxviii Rev. 17

WOLF CREEK CHAPTER 3 - LIST OF FIGURES

  • Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.

Figure # Sheet(s) Title Drawing #* 3.7(B)-15N 0 Spectra - Containment Building OBE, North-South Direction, Steam Generator Upper Support, Sterling Site 3.7(B)-15O 0 Deleted 3.7(B)-15P 0 Spectra - Containment Building OBE, North-South Direction, Steam Generator Upper Support, Wolf Creek Site 3.7(B)-15Q 0 Spectra - Containment Building OBE, East-West Direction, Steam Generator Upper Support, Callaway Site 3.7(B)-15R 0 Spectra - Containment Building OBE, East-West Direction, Steam Generator Upper Support, Sterling Site 3.7(B)-15S 0 Deleted 3.7(B)-15T 0 Spectra - Containment Building OBE, East-West Direction, Steam Generator Upper Support, Wolf Creek Site 3.7(B)-15U 0 Spectra - Containment Building OBE, Vertical Direction, Steam Generator Upper Support, Callaway Site 3.7(B)-15V 0 Spectra - Containment Building OBE, Vertical Direction, Steam Generator Upper Support, Sterling Site 3.7(B)-15W 0 Deleted 3.7(B)-15X 0 Spectra - Containment Building OBE, Vertical Direction, Steam Generator Upper Support, Wolf Creek Site 3.7(B)-16 0 Deleted 3.7(B)-17 0 Lumped-Mass/Flush Model, Containment Building 3.7(B)-18 0 Lumped-Mass/Flush Model, Fuel Building 3.7(B)-19 0 Lumped-Mass/Flush Model, Auxiliary/Control Building 3.7(B)-20 0 Lumped-Mass/Flush Model, Diesel Generator Building 3.7(B)-21 0 Damping vs. Input Level for Braced Hanger Systems 3.7(B)-22 0 Lower Bound Damping as a Function of Input ZPA 3.7(B)A-1 0 Description of the Model 3.7(N)-1 0 Multi-Degree-of-Freedom System 3.7(S)-1 0 Safe Shutdown Earthquake Horizontal Ground Spectra (0.12g) 3.7(S)-2 0 Safe Shutdown Earthquake Vertical Ground Spectra (0.12g) 3.0-xxix Rev. 17

WOLF CREEK CHAPTER 3 - LIST OF FIGURES

  • Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.

Figure # Sheet(s) Title Drawing #* 3.7(S)-3 0 Horizontal Design Response Spectra For 0.12g Horizontal Ground Acceleration (10% damping) 3.7(S)-4 0 Horizontal Design Response Spectra For 0.12g Horizontal Ground Acceleration (7% damping) 3.7(S)-5 0 Horizontal Design Response Spectra For 0.12g Horizontal Ground Acceleration (5% damping) 3.7(S)-6 0 Vertical Design Response Spectra For 0.12g Horizontal Ground Acceleration (10% damping) 3.7(S)-7 0 Vertical Design Response Spectra For 0.12g Horizontal Ground Acceleration (7% damping) 3.7(S)-8 0 Vertical Design Response Spectra For 0.12g Horizontal Ground Acceleration (5% damping) 3.7(S)-9 0 Typical Free Field Base Elevation Spectra ESWS Pumphouse 3.7(S)-10 0 Free Field Media Typical Subsurface Profile and Soil Properties SSE and OBE 3.7(S)-11 0 Mathematical Model for ESWS Pumphouse For East- West Vertical Analysis 3.7(S)-12A 0 Spectra-ESWS Pumphouse, SSE, North-South Direction, Top of Penthouse Roof 3.7(S)-12B 0 Spectra-ESWS Pumphouse, SSE, East-West Direction, Top of Penthouse Roof 3.7(S)-12C 0 Spectra-ESWS Pumphouse, SSE, Vertical Direction, Top of Penthouse Roof 3.7(S)-12D 0 Spectra-ESWS Pumphouse, OBE, North-South Direction, Top of Penthouse Roof 3.7(S)-12E 0 Spectra-ESWS Pumphouse, OBE, East-West Direction, Top of Penthouse Roof 3.7(S)-12F 0 Spectra-ESWS Penthouse, OBE, Vertical Direction, Top of Penthouse Roof 3.7(S)-13 0 Safe Shutdown Earthquake Horizontal Ground Spectra (0.15g) 3.7(S)-14 0 Safe Shutdown Earthquake Vertical Ground Spectra (0.15g) 3.8-1 0 Plan and Elevation of Reactor Building 3.8-2 0 Reactor Building Ground Floor Plan - Elev. 2000'-0" and 2001'-4" 3.8-3 0 Reactor Building Intermediate Floor Plan - Elev. 2026'-0" 3.8-4 0 Reactor Building Operating Floor Plan - Elev. 2047'- 6" and 2051'-0" 3.8-5 0 Reactor Building Plan - Elev 2068'-0" 3.8-6 0 Reactor Building East-West Cross Section 3.0-xxx Rev. 17

WOLF CREEK CHAPTER 3 - LIST OF FIGURES

  • Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.

Figure # Sheet(s) Title Drawing #* 3.8-7 0 Reactor Building North-South Cross Section 3.8-8 0 Reactor Building Base Mat Reinforcing - Bottom Layers 3.8-9 0 Reactor Building Base Mat Reinforcing - Top Layers 3.8-10 0 Reactor Building Base Mat Reinforcing - Cross Section 3.8-11 0 Reactor Building Base Mat Reinforcing - Shear Tie 3.8-12 0 Reactor Building Shell Reinforcing 3.8-13 0 Reactor Building Dome Reinforcing - Plan 3.8-14 0 Reactor Building Dome Reinforcing - Elevation 3.8-15 0 Reactor Building Tendon Anchorage System 3.8-16 0 Reactor Building Tendon and Buttress Arrangement 3.8-17 0 Reactor Building Tendons - Sections 3.8-18 0 Reactor Building Tendons - Additional Sections 3.8-19 0 Reactor Building Liner Plate - Typical Wall Sections 3.8-20 0 Reactor Building Liner Plate - Dome Stiffener Plan 3.8-21 0 Reactor Building Liner Plate - Typical Dome Section 3.8-22 0 Reactor Building Liner Plate - Dome Details 3.8-23 0 Anchorage at Reactor Cavity - Plan View 3.8-24 0 Anchorage at Reactor Cavity - Typical Section 3.8-25 0 Typical Anchorage Through Base Mat for NSSS Equipment Supports 3.8-26 0 Reactor Building Polar Crane Brackets 3.8-27 0 Reactor Building Shell Typical Beam Support Brackets 3.8-28 0 Reactor Building - Typical Pipe Support Brackets in Dome 3.8-29 0 Reactor Building Liner Plate Leak Chase - Typical Data 3.8-30 0 Reactor Building Buttress Details 3.8-31 0 Reactor Building Equipment Hatch Opening 3.8-32 0 Reactor Building Equipment Hatch Opening - Typical Section 3.8-33 1 Reactor Building Personnel Hatch Opening - Inside Face 3.8-33 2 Reactor Building Personnel Hatch Opening - Outside Face 3.8-34 0 Reactor Building Main Steam and Main Feedwater Openings - Inside Face 3.8-35 0 Reactor Building Main Steam and Main Feedwater Openings - Outside Face 3.0-xxxi Rev. 17

WOLF CREEK CHAPTER 3 - LIST OF FIGURES

  • Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.

Figure # Sheet(s) Title Drawing #* 3.8-36 0 Temperature Gradients Through Reactor Building Wall for DBA (Postulated Primary Coolant Loop Break) 3.8-37 0 Finite Element Model for Axisymmetric Loads - Str. 3.8-38 0 Finite Element Model for Axisymmetric Loads - Dome 3.8-39 0 Finite Element Model for Axisymmetric Loads - Founda. Medium 3.8-40 0 Finite Element Model for Nonaxisymmetric Loads 3.8-41 0 Finite Element Model for Equipment Hatch - Elevation 3.8-42 0 Finite Element Model for Equipment Hatch - Plan 3.8-43 0 Finite Element Model for Personnel Hatch 3.8-44 0 Reactor Building Equipment Hatch 3.8-45 0 Reactor Building Personnel Hatch 3.8-46 0 Reactor Building Auxiliary Access Hatch 3.8-47 0 Reactor Building Typical Pipe Penetration 3.8-48 0 Reactor Building Fuel Transfer Penetration 3.8-49 0 Reactor Building Electrical Penetration 3.8-50 0 Reactor Building Purge Line Penetrations 3.8-51 0 Reactor Vessel Support System - Elevation 3.8-52 0 Reactor Vessel Support System - Plan 3.8-53 0 Steam Generator Support System - Upper Supports 3.8-54 0 Steam Generator Support System - Lower Supports 3.8-55 0 Steam Generator Support System - Elevation 3.8-56 0 Reactor Coolant Pump Lateral Support Embeds 3.8-57 0 Reactor Coolant Pump Support Details 3.8-58 0 Reactor Cavity Plan - Elevation 1997'-6" to 2005'-7" 3.8-59 0 Reactor Cavity Plan - Elevation 2011'-6" to 2021'-7" 3.8-60 0 Reactor Cavity Elevations 3.8-61 0 Reactor Cavity Neat Line 3.8-61a 0 Reactor Cavity Neutron Shield 3.8-62 0 Secondary Shield Walls - Elevation 2000'-0" to 2025'-0" 3.8-63 0 Secondary Shield Walls - Elevation 2025'-0" to 2047'-0" 3.8-64 0 Secondary Shield Walls - Sections 3.8-65 0 Secondary Shield Walls - Additional Sections 3.8-66 0 Pressurizer Supports 3.8-67 0 Pressurizer Support Details 3.8-68 0 Refueling Canal - Typical Plan 3.8-69 0 Refueling Pool Typical Cross Section 3.8-70 0 Reactor Building Operating Floor 3.0-xxxii Rev. 17

WOLF CREEK CHAPTER 3 - LIST OF FIGURES

  • Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.

Figure # Sheet(s) Title Drawing 3.8-71 0 Reactor Building Operating Floor Supports at Shell 3.8-72 0 Reactor Building Intermediate Floor at Elevation 2026'-0" 3.8-73 0 Reactor Building Intermediate Floor at Elevation 2068'-6" 3.8-74 0 Simplified Head Assembly with Reactor Missile Shield 3.8-75 0 Reactor Building Polar Crane Support System 3.8-76 0 Deleted 3.8-77 0 Refueling Pool Finite Element Model - Isometric 3.8-78 0 Refueling Pool Finite Element Model - Plan Views 3.8-79 0 Secondary Shield Wall East Side Finite Element Model - Plan Views 3.8-80 0 Reactor Building Secondary Shield Wall Finite Element Model - Plan View 3.8-81 0 Wall Finite Element Model - Sections A, B and C 3.8-82 0 Reactor Building Secondary Shield Wall Finite Element Model - Section D 3.8-83 0 Reactor Cavity Finite Element Model 3.8-84 0 General Arrangement of Standard Plant Category I Structures 3.8-85 0 Typical Isolation Joints Between Buildings 3.8-86 0 Auxiliary Building Plan - Elev. 1974'-0" 3.8-87 0 Auxiliary Building Plan - Elev. 1988'-0" and 1989'-6" 3.8-88 0 Auxiliary Building Plan - Elev. 2000'-0" 3.8-89 0 Auxiliary Building Plan - Elev. 2026'-0" 3.8-90 0 Auxiliary Building Plan - Elev. 2047'-6" 3.8-91 0 Auxiliary Building Plan - North-South Cross Section 3.8-92 0 Auxiliary Building Plan - East-West Cross Section 3.8-93 0 Auxiliary Building Plan - East-West Cross Section 3.8-94 0 Fuel Building Plan - Elev. 2000'-0" (UN) 3.8-95 0 Fuel Building Plan - Elev. 2026'-0" (UN) 3.8-96 0 Fuel Building Plan - Elev. 2047'-6" 3.8-97 0 Fuel Building - North-South Cross Section 3.8-98 0 Fuel Building - East-West Cross Section 3.8-99 0 Control Building Plan - Elev. 1974'-0" and 1984'-0" 3.8-100 0 Control Building Plan - Elev. 2000'-0" and 2016'-0" 3.8-101 0 Control Building Plan - Elev. 2032'-0" 3.8-102 0 Control Building Plan - Elev. 2047'-6" and 2073'-6" 3.8-103 1 Control Building - North - South Cross Section 3.8-103 2 ESW Vertical Loop Chase - North - South Cross Section 3.8-104 1 Control Building - Isometric Cross Section ESW Vertical Loop Chase - Isometric Cross Section 3.8-104 2 ESW Vertical Loop Chase - Isometric Cross Section 3.8-105 0 Diesel-Generator Building Plan - Elev. 2000'-0" 3.0-xxxiii Rev. 30

WOLF CREEK CHAPTER 3 - LIST OF FIGURES

  • Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.

Figure # Sheet(s) Title Drawing #* 3.8-106 0 Diesel-Generator Building Plan - Elev. 2024'-0" and 2032'-0" 3.8-107 0 Diesel-Generator Building Plan - Elev. 2047'-2" 3.8-108 0 Diesel-Generator Building - East-West Cross Section 3.8-109 0 Diesel-Generator Building - North-South Cross Section 3.8-110 0 Refueling Water Storage Tank and Valve House - Foundation Plan 3.8-111 0 Refueling Water Storage Valve House Elevation 3.8-112 0 Emergency Oil Storage Tanks and Access Vault Plan 3.8-113 0 Emergency Oil Storage Tanks and Access Vault 3.8-114 0 Buried Duct Banks to Emer. Fuel Oil Storage Tanks 3.8-115 0 Buried Duct Banks to Refueling Storage Valve House 3.8-116 0 Arrangement of Foundation - Plan 3.8-117 0 Arrangement of Foundation - Details 3.8-118 0 Arrangement of Foundation - Additional Details 3.8-119 0 Auxiliary and Control Building Foundation Plan 3.8-120 0 Auxiliary and Control Building Foundation Sections 3.8-121 0 Fuel Building Foundation Plan 3.8-122 0 Fuel Building Foundation Sections 3.8-123 0 Diesel-Generator Building Foundation Plan 3.8-124 0 Radwaste Building and Tunnel - Plan El. 1974'-0" and El. 1976'-0" 3.8-125 0 Radwaste Building - Plan El. 2000'-0" 3.8-126 0 Radwaste Building - Plan El. 2022'-0" 3.8-127 0 Radwaste Building - Plan El. 2031'-6" 3.8-128 0 Radwaste Building - Plan El. 2040'-6" and El. 2047'- 0" 3.8-129 0 Radwaste Building - Section 3.8-130 0 Radwaste Building - Section 3.8-131 0 Plan-ESWS Pumphouse 3.8-132 0 E-W Section ESWS Pumphouse 3.8-133 0 N-S Sections ESWS Pumphouse 3.8-134 0 Plan-ESWS Pipes and Duct Banks 3.8-135 0 Plan-ESWS Pipes and Duct Banks 3.8-135a 0 Plan-ESWS Pipes and Duct Banks 3.8-136 0 Section Through ESWS Pipes and Duct Banks 3.8-137 1 Plan and Sections of Pipe Encasements 3.8-137 2 Circ Water Line Protection Structure 3.8-138 0 30" Diameter Pipe Penetration Details 3.8-139 0 Duct Bank Entrance Details 3.8-140 0 Electrical Manholes 3.0-xxxiv Rev. 28

WOLF CREEK CHAPTER 3 - LIST OF FIGURES

  • Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.

Figure # Sheet(s) Title Drawing #* 3.8-141 0 Deleted 3.8-142 0 Plan and Section Discharge Point 3.8-143 1-5 Plan and Section ESW Access Vaults 3.8-144 0 4 and 18 Diameter Pipe Penetration Details 3.8-145 0 ESWS Pumphouse Tornado Missile Shield 3.9(N)-1 1 Reactor Coolant Loop Support System Dynamic Structural Model 3.9(N)-1 2 Reactor Coolant Piping Model for Loop 1 (Typical) 3.9(N)-2 0 Through-Wall Thermal Gradients 3.9(N)-3 0 Vibration Checkout Functional Test Inspection Points 3.9(N)-4 0 Full-Length Control Rod Drive Mechanism 3.9(N)-5 0 Full-Length Control Rod Drive Mechanism Schematic 3.9(N)-6 0 Nominal Latch Clearance at Minimum and Maximum Temperature 3.9(N)-7 0 Control Rod Drive Mechanism Latch Clearance Thermal Effect 3.9(N)-8 0 Lower Core Support Assembly (Core Barrel Assembly) 3.9(N)-9 0 Upper Core Support Structure 3.9(N)-10 0 Plan View of Upper Core Support Structure 3.11(B)-1 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-2 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-3 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-4 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-5 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-6 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-7 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-7A 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-8 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-9 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-9A 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-10 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-11 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-12 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-13 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-14 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.0-xxxv Rev. 29

WOLF CREEK CHAPTER 3 - LIST OF FIGURES

  • Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.

Figure # Sheet(s) Title Drawing #* 3.11(B)-15 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-16 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-17 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-18 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-19 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-20 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-21 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-22 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-23 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-24 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-25 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-26 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-27 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-28 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-29 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-30 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-31 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-32 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-33 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-34 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-35 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-36 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-37 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.0-xxxvi Rev. 29

WOLF CREEK CHAPTER 3 - LIST OF FIGURES

  • Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.

Figure # Sheet(s) Title Drawing #* 3.11(B)-38 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-39 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-40 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-41 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-42 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-43 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-44 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-45 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-46 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-47 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-48 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.11(B)-49 0 See EQSD-1, Attachment A Cross Reference Table between Figures 3.11(B)-1 through 3.11(B)-49 of USAR Section 3.11(B). 3.12-1 0 Displacement in Otter Brook Dam Compared With Isbild Results 3.12-2 0 Settlement Analysis by Isbild for Homogeneous Dam Due to Water Loading 3.12-3 0 Comparison of Water Load Effects on Movement 3.12-4 0 Soil Profile and Finite Element Representation Used for QUAD 4 Sample Problem 3.12-5 0 Strain-Compatible Damping and Modulus Values Used in Analysis by QUAD 4 3.12-6 0 Distribution of Maximum Shear Stresses and Accelerations 3.12-7 0 Acceleration Spectra for Computed Surface Motion 3.12-8 0 Comparison of Accelerated Time History Plot of the El Centro N-S Earthquake Record from RSG and as Published by California Institute of Technology 3.12-9 0 Comparison of Response Spectra PWTS at Various Damping from RSG and as Published in Brady (1972) 3.12-10 0 Fourier Transform Plot from RSG for a 5 Cycle/ Sec. Sine Wave Time History 3.12-11 0 Comparison of Desired Response Spectrum and Response Spectrum of Compatible Acceleration Time History (Damping - 0.02) from RSG 3.0-xxxvii Rev. 29

WOLF CREEK CHAPTER 3 - LIST OF FIGURES

  • Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.

Figure # Sheet(s) Title Drawing #* 3.12-12 0 Finite Element Mesh for Axisymetric Flow Problem for Seepage 3.12-13 0 Critical Slope of Ultimate Heat Sink for Bishop Validation Problem 3A-1 0 Comparison of Tensile Stress for Bolts 3A-2 0 Factor of Safety Against Failure Under Service Level D as a Function of T-S Ratio 3B-1 0 Auxiliary Building El. 1974 Hazards Analysis Room Locations 3B-2 0 Plan and Elevation View of Main Steam/Main Feedwater Isolation Valve Compartment 3B-3 0 Deleted 3B-4 1 Nodalization Model for Main Steam/Main Feedwater Isolation Valve Compartment Pressure Analysis 3B-4 2 Nodalization Model for Main Steam/Main Feedwater Isolation Valve Compartment Temperature Analysis 3B-5 0 Main Steam/Main Feedwater Isolation Valve Compartment Pressure Transient 3B-6a 0 Main Steam/Main Feedwater Isolation Valve Compartment Temperature Transient 3B-6b 0 Main Steam/Main Feedwater Isolation Valve Compartment Temperature Transient 3B-6c 0 Main Steam/Main Feedwater Isolation Valve Compartment Temperature Transient 3B-6d 0 Main Steam/Main Feedwater Isolation Valve Compartment Temperature Transient 3B-7 0 Turbine Building CWS Rupture 3C-1 0 Comparison of Lawrance Livermore Spectrum with Wolf Creek Design Spectra 3C-2 0 Plan View - ESWS Pumphouse 3C-3 0 N-S Section - ESWS Pumphouse 3C-4 0 E-W Section - ESWS Pumphouse 3C-5 0 Comparison of Flush & Fixed Base Analysis Building Shear and Moments 3C-6 0 Comparison of Flush & Fixed Base Analysis Pumphouse N-S Response Spectrum at El. 2000', 3% Damping 3C-7 0 Comparison of Flush & Fixed Base Analysis Pumphouse E-W Response Spectrum at El. 2000', 3% Damping 3.0-xxxviii Rev. 28

WOLF CREEK CHAPTER 3 - LIST OF FIGURES

  • Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.

Figure # Sheet(s) Title Drawing #* 3C-8 0 Comparison of Flush & Fixed Base Analysis Pumphouse Vertical Response Spectrum at El. 2000', 3% Damping 3C-9 0 Comparison of Flush & Fixed Base Analysis Pumphouse N-S Response Spectrum at El. 2025', 3% Damping 3C-10 0 Comparison of Flush & Fixed Base Analysis Pumphouse E-W Response Spectrum at El. 2025', 3% Damping 3C-11 0 Comparison of Flush & Fixed Base Analysis Pumphouse Vertical Response Spectrum at El. 2025', 3% Damping 3C-12 0 Pumphouse Shear & Moments for 0.15g SSE 3C-13 0 Pumphouse Response Spectrum for 0.15g SSE N-S Direction at El. 2000', 3% Damping 3C-14 0 Pumphouse Response Spectrum for 0.15g SSE E-W Direction at El. 2000', 3% Damping 3C-15 0 Pumphouse Response Spectrum for 0.15g SSE Vertical Direction at El. 2000', 3% Damping 3C-16 0 Pumphouse Response Spectrum for 0.15g SSE N-S Direction at El. 2025', 3% Damping 3C-17 0 Pumphouse Response Spectrum for 0.15g SSE E-W Direction at El. 2025', 3% Damping 3C-18 0 Pumphouse Response Spectrum for 0.15g SSE Vertical Direction at El. 2025', 3% Damping 3C-19 0 Test Response Spectra Front to Back Horizontal Direction ESWS Control Panels EF-155 and EF-156 3C-20 0 Comparison of Artificial Accelerogram and Design Response Spectra for Maximum Horizontal Ground Acceleration of 15% of Gravity and 5% Spectra Damping 3C-21 0 Comparison of Artificial Accelerogram and Design Response Spectra for Maximum Vertical Ground Acceleration of 15% of Gravity and 5% Spectra Damping 3.0-xxxix Rev. 17

WOLF CREEK 3.0 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS This chapter identifies, describes, and discusses the principal architectural and engineering design features of those structures, components, equipment, and systems which are necessary to assure:

a. The integrity of the reactor coolant pressure boundary
b. The capability to shut down the reactor and maintain it in a post-accident safe shutdown condition
c. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline values of 10 CFR 50.67.

3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA This section briefly discusses the extent to which the design criteria for safety-related plant structures, systems, and components comply with Title 10, Code of Federal Regulations, Part 50 (10 CFR 50), Appendix A, "General Design Criteria for Nuclear Power Plants" (GDC). As presented in this section, each criterion is first quoted and then discussed in enough detail to demonstrate compliance with each criterion. For some criteria, additional information may be required for a complete discussion. In such cases, detailed evaluations of compliance with the various general design criteria are incorporated in more appropriate USAR sections, but are located by reference. 3.1.1 DEFINITION OF SINGLE FAILURE The single failure criterion is a constraint used in the design of safety systems to improve the reliability of the system to perform its safety function following a design-basis event or design occurrence. A single failure means an occurrence which results in the loss of the capability of a component to perform its intended safety functions. Multiple failures resulting from a single occurrence are considered to be a single failure. Fluid and electrical systems are considered to be designed against an assumed single failure if neither (1) a single failure of any active component (assuming that passive components function properly) nor (2) a single failure of a passive component (assuming that active components function properly) results in a loss of the capability of the system to perform its safety functions. 3.1-1 Rev. 34

WOLF CREEK Single failures are random occurrences imposed upon safety systems that are required to respond to a design basis event. They are postulated despite the fact that the systems were designed to remain functional under the adverse condition imposed by the accident. No mechanism for the cause of the single failure need be postulated. Single failures of passive components in electrical systems are assumed in designing against a single failure. 3.1.1.1 Active Component An active component is a device characterized by an expected significant change of state or a discernible mechanical motion in response to an imposed design-basis load demand upon the system. Examples are switches, relays, powered valves, check and safety valves, pressure switches, turbines, transistors, motors, dampers, pumps, analog meters, etc. (See Sections 3.9(B).3.2 and 3.9(N).3.2 for discussions and lists of active pumps and valves). The definition of an active component for the purpose of supporting the pump and valve operability program includes the Westinghouse nuclear steam supply system (NSSS) check valves. These check valves, although not powered components, meet the definition of having mechanical motion and are therefore included in Table 3.9(N)-11. At the same time, however, they are not considered to be active (powered) components in the Westinghouse design with respect to the Emergency Core Cooling System (ECCS) failure modes and effects analysis (FMEA) of active components or the single active failure analysis for ECCS components. Refere to Section 6.3.2.5. 3.1.1.2 Active Component Failure An active failure is a failure of an active component to complete its intended function upon demand. Examples of active component failures include the failure of a powered valve to move to its correct position, failure of a pump, fan, or diesel generator to start, failure of a relay to respond, etc. Certain selected valves that are provided with a power supply for proper system functioning must be prevented from unwanted movement in certain situations. Remote manual power lockout of these valves is provided to preclude unwanted valve motion due to an assumed single electrical failure. The valves are identified in their appropriate sections. Where the proper active function of a component can be demonstrated despite any reasonable postulated condition, then that component may be considered exempt from active failure. Examples of such components may include code safety valves and check valves. Where such exemption is taken, the basis for the exemption shall be documented in the single failure analysis. Although Westinghouse NSSS check valves are included in Table 3.9(N)-11, they are not considered to be active components in Tables 6.3-5 and 6.3-6. Refer to Section 3.9(N).3.2.1 and Section 6.3.2.5. 3.1.1.3 Passive Component A passive component is a device characterized by an expected negligible change of state or negligible mechanical motion in response to an imposed design basis demand upon the system. Examples are cables, piping, valves in stationary position, resistors, capacitors, fluid filters, indicator lamps, cabinets, cases, etc. 3.1-2 Rev. 26

WOLF CREEK 3.1.1.4 Passive Component Failures A passive component failure is the structural failure of a static component which limits the component's effectiveness in carrying out its design function. When applied to a fluid system, this means a breach of the pressure boundary is postulated, resulting in abnormal leakage. Such leakage is limited to that which results from a single sprung flange, a single pump seal failure, a single valve stem packing failure, or other single failure mechanisms considered credible by a systematic analysis of system components. The probability of a large break in a piping system (e.g., rupture of ECCS piping), subsequent to the original large LOCA pipe break, is considered to be sufficiently low that it need not be postulated. Single failures of passive components in electrical systems are assumed in designing against a single failure. 3.1.2 ADDITIONAL SINGLE FAILURE ASSUMPTIONS In designing for and analyzing for a DBA (i.e., loss-of-coolant accident, main steam line break, fuel handling accident, or steam generator tube rupture), the following assumptions are made, in addition to postulating the initiating event.

a. The events are assumed not to result from a tornado, hurricane, flood, fire, loss of offsite power, or earthquake.
b. Any one of the following occurs:
1. During the short term of an accident, a single failure of any active mechanical component. The short term is defined as less than 24 hours following an accident, or
2. During the short term of an accident, a single failure of any active or passive electrical component, or
3. A single failure of passive components associated with long-term cooling capability, assuming that a single active failure has not occurred during the short term. Long-term cooling applies to a time duration greater than 24 hours.
c. No reactor coolant system transient is assumed, preceding the postulated reactor coolant system piping rupture.

3.1-3 Rev. 0

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d. No operator action is assumed to be taken by plant operators to correct problems during the first 10 minutes following the accident.
e. All offsite power is simultaneously lost and is restored within 7 days (except that for events postulated to occur during MODE 5, MODE 6, and/or during movement of irradiated fuel assemblies when the plant is in MODE 5 or MODE 6 or with the core fully offloaded, such as a fuel handling accident, a loss of all offsite power is not required to be assumed in addition to a single failure).
f. For a LOCA, for additional safety no credit is taken for the functioning of nonseismic Category I components.

In the design and analysis performed for provision of protection of safety-related equipment from hazards and events (tornadoes, floods, missiles, pipe breaks, fires, and seismic events) which could reasonably be expected, the following assumptions were made:

a. Should the event result in a turbine or reactor trip, loss of offsite power is assumed, and the plant will be placed in a hot standby condition.
b. If required by a limiting condition for operation (per Technical Specifications or if the recovery from the event will cause the plant to be shutdown for an extended period of time, the plant will be taken to a cold shutdown (CSD) condition.
c. Redundancy or diversity of systems and components is provided to enable continued operation at hot standby or to cool the reactor to a CSD condition. If required, it is assumed that temporary repairs can be made to circumvent damages resulting from the hazard. All available systems, including non-safety-related systems and those systems requiring operator action, may be employed to mitigate the consequences of the hazard.

In determining the availability of the systems required to mitigate the consequences of a hazard and those required to place the reactor in a safe condition, the direct consequences of the hazard are considered. The feasibility of carrying out operator actions are based on ample time and adequate access to the controls, motor control center, switchgear, etc., associated with the component required to accomplish the proposed action.

d. When the postulated hazard occurs and results in damage to one of two or more redundant or diverse trains, single failures of components in other trains (and associated 3.1-4 Rev. 27

WOLF CREEK supporting trains) are not assumed. The postulated hazard is precluded, by design, from affecting the opposite train or from resulting in a DBA. For the situation in which a hazard affects a safety-related component, the event and subsequent activities are governed by Technical Specification requirements in effect when that component is not functional.

e. When evaluating the effects of any earthquake, no other major hazard or event is assumed, and no seismic Category I equipment is assumed to fail as a result of the earthquake. Certain nonseismic Category I components are designed and constructed to ensure that their failure will not reduce the functioning of a safety-related component to an unacceptable safety level. This criterion meets the intent of Regulatory Guide 1.29, Position C.2. Evaluation of component failure includes drop impact forces and secondary effects, such as spray and flooding from piping failure.

3.1.3 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 functions 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." DISCUSSION The quality assurance program of the utility, together with the quality assurance, quality engineering, and quality control programs of the major contractors and their vendors, ensure that safety-related structures, systems, and components are designed, fabricated, erected, and tested to quality standards commensurate with the safety functions to be performed. This is accomplished 3.1-5 Rev. 0

WOLF CREEK through the use of recognized codes, standards, and design criteria. As necessary, additional supplemental standards, design criteria, and requirements were developed by SNUPPS and the major contractors' engineering organizations. Appropriate records associated with the engineering and design, fabrication, erection, and testing which document the compliance with recognized codes, standards, and design criteria are maintained throughout the life of the unit. Quality assurance is described in Chapter 17.0. The principal design criteria, design bases, codes, and standards applied to the facility are described in Section 3.2. Additional detail may be found in the pertinent section of the USAR dealing with safety-related structures, systems, and components, e.g., the containment as described in Section 3.8.2. 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." DISCUSSION The safety-related structures, systems, and components are designed either to withstand the effects of natural phenomena without loss of the capability to perform their safety functions, or to fail 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 site, determined from recorded data for the site vicinity, with appropriate margin to account for uncertainties in historical data. Appropriate combinations of structural loadings from normal, accident, and natural phenomena are considered in the plant design. The nature and magnitude of the natural phenomena considered in the design of this plant are discussed in Chapter 2.0. Chapter 3.0 discusses the design of the plant in relationship to 3.1-6 Rev. 0

WOLF CREEK natural events. Seismic and quality group classifications, as well as other pertinent standards and information, are given in the sections discussing individual structures and components. 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. Firefighting 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." DISCUSSION The plant is designed to minimize the probability and effect of fires and explosions. Noncombustible and fire-resistant materials are used in the containment, control room, components of safety features systems, and throughout the unit whenever fire is a potential risk to safety-related systems. For example, electrical cables have a fire retardant jacketing, and fire barriers and fire stops are utilized as described in Section 9.5.1. Equipment and facilities for fire protection, including detection, alarm, and extinguishment, are provided to 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, Halon 1301, and portable extinguishers. Firefighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety 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, and, where required by law, the system and equipment conform to the applicable standards:

a. National Fire Protection Association (NFPA) "National Fire Codes" 3.1-7 Rev. 0

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b. American Nuclear Insurers (ANI) "Basic Fire Protection for Nuclear Power Plants," April, 1976
c. "International Guidelines for the Fire Protection of Nuclear Power Plants" - 1974
d. "Occupational Safety and Health Standards," Federal Register Part 1910, October, 1972
e. BTP-APCSB 9.5-1 "Guidelines for Fire Protection for Nuclear Power Plants," May 1, 1976.

CRITERION 4 - ENVIRONMENTAL AND MISSILE 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." DISCUSSION 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 LOCAs. Criteria are presented in Chapter 3.0, and the environmental conditions are described in Sections 3.11(B) and 3.11(N). These 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 systems, structures, and components are included in Chapters 3.0, 5.0, 6.0, 7.0, 9.0, and 10.0. Evaluation of the performance of the safety features is contained in Chapter 15.0. 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 3.1-8 Rev. 0

WOLF CREEK such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units." DISCUSSION Wolf Creek Generating Station is a one unit site. 3.1.4 PROTECTION BY MULTIPLE FISSION PRODUCT BARRIERS CRITERION 10 - REACTOR DESIGN "The reactor core and associated coolant, control, and protection systems shall be designed with an 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." DISCUSSION The reactor core and associated coolant, control, and protection systems are designed to the following criteria:

a. No fuel damage will occur during normal core operation and operational transients (Condition I) or any transient conditions arising from occurrences of moderate frequency (Condition II) beyond the small fraction of clad defects (1 percent) for which the plant shielding, cleanup, and radwaste systems are designed. Fuel damage, as used here, is defined as penetration of the fission product barrier (i.e., the fuel rod clad). Conditions I and II, as used here, are defined by ANSI N18.2-1973. 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.
b. The reactor can be returned to a post-accident 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. Condition III, as used here, is defined by ANSI N18.2-1973.
c. The core will remain intact with acceptable heat transfer geometry following transients arising from occurrences of limiting faults (Condition IV). Condition IV, as used here, is defined by ANSI N18.2-1973.

3.1-9 Rev. 14

WOLF CREEK The reactor trip system is designed to actuate a reactor trip whenever necessary to ensure that the fuel design limits are not exceeded. The core design, together with the process and decay heat removal systems, provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and anticipated transient situations, including 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 4.0 discusses the design bases and design evaluation of core components. Details of the control and protection systems' instrumentation design and logic are discussed in Chapter 7.0. This information supports the accident analyses of Chapter 15.0 which show that the acceptable fuel design limits are not exceeded for Condition I and II occurrences. 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." DISCUSSION Whenever the reactor is critical, prompt compensatory reactivity feedback is assured by the negative fuel temperature effect (Doppler effect). At full power, compensatory reactivity feedback is assured by the nonpositive operational limit on the moderator temperature coefficient of reactivity. The negative Doppler coefficient of reactivity is assured by the inherent design, using low enrichment fuel. The nonpositive moderator temperature coefficient of reactivity is assured by administratively controlling the dissolved absorber concentration or by using burnable poison. Reactivity coefficients and their effects are discussed in Chapter 4.0. 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." 3.1-10 Rev. 11

WOLF CREEK DISCUSSION Power oscillations of the fundamental mode are inherently eliminated by negative Doppler and nonpositive moderator temperature coefficients of reactivity. Small positive moderator temperature coefficients are allowable at reactor powers <70%. The reactor coolant, control, and protection systems are designed to assure positive MTCs at partial power can be controlled within acceptable fuel design limits. Oscillations, due to xenon spatial effects, in the radial, diametral, and azimuthal overtone modes are heavily damped due to the inherent design and due to the negative Doppler and nonpositive moderator temperature coefficients of reactivity. Oscillations, due to xenon spatial effects, may occur in the axial first overtone mode. Assurance that fuel design limits are not exceeded by xenon axial oscillations is provided by reactor trip functions, using the measured axial power imbalance as an input. If necessary to maintain axial imbalance within the limits of the Technical Specifications, i.e., imbalances which are alarmed to the operator and are within the imbalance trip setpoints, the operator can suppress xenon axial oscillations by control rod motions and/or temporary power reductions. Oscillations, due to xenon spatial effects, in axial modes higher than the first overtone are heavily damped due to the inherent design and due to 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 4.0. Details of the instrumentation design and logic are discussed in Chapter 7.0. 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 associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges." DISCUSSION Instrumentation and controls are provided to monitor and control neutron flux, control rod position, fluid temperatures, pressures, flows, and levels, as necessary, to assure that adequate plant 3.1-11 Rev. 13

WOLF CREEK safety can be maintained. Instrumentation is provided in the reactor coolant system, steam and power conversion system, containment, engineered safety features systems, radiological waste systems, and other auxiliaries. Parameters that must be provided for operator use under normal operating and accident conditions are indicated in the control room in proximity to the controls for maintaining the indicated parameter in the proper range. The quantity and types of process instrumentation provided ensure safe and orderly operation of all systems over the full design range of the plant. These systems are described in Chapters 6.0, 7.0, 8.0, 9.0, 10.0, 11.0, and 12.0. 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." DISCUSSION The reactor coolant pressure boundary is designed to accommodate the system pressures and temperatures attained under all expected modes of plant operation, including all anticipated transients, with stresses within applicable limits. Consideration is given to loadings under normal operating conditions and to abnormal loadings, such as pipe rupture and seismic loadings, as discussed in Chapter 3.0. The piping is protected from overpressure by means of pressure-relieving devices, as required by ASME Section III. Reactor coolant pressure boundary materials and fabrication techniques are such that there is a low probability of gross rupture or significant leakage. Coolant chemistry is controlled to protect the materials of construction of the reactor coolant pressure boundary from corrosion. The reactor coolant pressure boundary is accessible for inservice inspections to assess the structural and leaktight integrity. For the reactor vessel, a material surveillance program conforming to applicable codes is provided. Chapter 5.0 has additional details. Instrumentation is provided to detect significant leakage from the reactor coolant pressure boundary with indication in the control room, as discussed in Chapter 5.0. 3.1-12 Rev. 0

WOLF CREEK 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 any condition of normal operation, including anticipated operational occurrences." DISCUSSION Steady-state and transient analyses are performed to ensure that reactor coolant system design conditions are not exceeded during normal operation. Protection and control setpoints are based on these analyses. Additionally, reactor coolant pressure boundary components have a large margin of safety through application of proven materials and design codes, use of proven fabrication techniques, nondestructive shop testing, and integrated hydrostatic testing of assembled components. The effect of radiation embrittlement is considered in reactor vessel design, and surveillance samples monitor adherence to expected conditions throughout the plant life. Multiple safety and relief valves are provided for the reactor coolant system. These valves and their setpoints meet the ASME criteria for overpressure protection. The ASME criteria are satisfactory, based on a long history of industrial use. Chapter 5.0 discusses the reactor coolant system design. CRITERION 16 - CONTAINMENT DESIGN "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." DISCUSSION A steel-lined, prestressed, post-tensioned concrete reactor containment structure encloses the entire reactor coolant system. It is designed to sustain, without loss of required integrity, the effects of LOCAs up to and including the double-ended rupture of the largest pipe in the reactor coolant system or double-ended rupture of a steam or feedwater pipe. Engineered safety features comprising the emergency core cooling system, containment spray 3.1-13 Rev. 0

WOLF CREEK system, and the containment air coolers serve to cool the reactor core and return the containment to near atmospheric pressure. The reactor containment structure and engineered safety features systems are designed to assure the required functional capability of containing any uncontrolled release of radioactivity. The concrete radiological shielding and the liner within the containment limit the uncontrolled release of radioactivity to the environment. Refer to Chapters 3.0, 6.0, and 15.0. CRITERION 17 - ELECTRIC POWER SYSTEMS "An onsite electric power system and an offsite electric power system shall be provided to permit the functioning of structures, systems, and components important to safety. The safety function for each system (assuming the other system is not functioning) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is cooled and containment 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 the other offsite electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One of these circuits shall 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 3.1-14 Rev. 0

WOLF CREEK nuclear power unit, the loss of power from the transmission network, or the loss of power from the onsite electric power supplies." DISCUSSION An onsite electric power system and an offsite electric power system are provided to permit the functioning of safety-related structures, systems, and components. As discussed in Chapter 8.0, each Class 1E electric power system is designed with adequate independence, capacity, redundancy, and testability to ensure the functioning of engineered safety features (ESF). Independence is provided by physical separation and electrical isolation of components and cables to minimize the vulnerability of the redundant systems to any single credible event. Two physically independent sources of power provide preferred power to the onsite power system. One preferred circuit is connected to a 13.8/4.16-kV ESF transformer which supplies power normally to its associated 4.16-kV Class 1E bus. The second preferred circuit is connected to one secondary winding of a 3-winding startup transformer which supplies power to a second 13.8/4.16-kV ESF transformer. The second ESF transformer supplies power normally to its associated 4.16-kV Class 1E bus. Each ESF transformer normally supplies power to its associated 4.16-kV Class 1E ac bus, but it can simultaneously supply power to the second 4.16-kV Class 1E bus, if required, by the closure of the circuit breaker. A failure of a single component will not prevent the safety-related systems from performing their function. Each of the preferred circuits is designed to be available in sufficient time, following a loss of all onsite power sources and the other offsite electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. The onsite ac power is furnished by two diesel generators. Each diesel generator is connected to a Class 1E bus. The ESF loads are divided between the Class 1E busses in a balanced, redundant load grouping. Each diesel generator is capable of supplying sufficient power in sufficient time for the operation of the engineered safety features required for the unit during a postulated loss-of-coolant accident. During a postulated LOCA, both diesel generators start automatically. If preferred power is available to the Class 1E bus following a loss-of-coolant accident, the ESF loads will be started sequentially. However, in the event that preferred power is lost, the load sequencing system will connect the diesel generator to its associated Class 1E bus and sequentially start the ESF equipment. The associated diesel 3.1-15 Rev. 1

WOLF CREEK generator is so arranged that a failure of a single component will not prevent the post-accident safe shutdown of the reactor. The onsite Class 1E dc power supply consists of four independent battery systems. Failure of a single component in this system will not impair control of the engineered safety features required to maintain the reactor in a safe condition. 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 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." DISCUSSION Class 1E electric power systems are designed as described below in order that the following aspects of the system can be periodically tested:

a. The operability and functional performance of the components of Class 1E electric power systems (diesel generators, engineered safety feature (ESF) busses, dc system)
b. The operability of these electric power systems as a whole and under conditions as close to design as practical, including the full operational sequence that actuates these systems The switchyard circuit breakers will be inspected, maintained, and tested on a routine basis without affecting the rest of the system. Transmission lines and protective relaying on these lines will be periodically tested.

Any one of the ESF transformers and its circuit to the Class 1E busses can be taken out of service and tested periodically. Each transformer has the capacity to supply power to both group 1 and 3.1-16 Rev. 14

WOLF CREEK group 2 Class 1E loads simultaneously. The 4160-V and 480-V circuit breakers and the associated equipment will be tested one at a time only while redundant equipment is operational. The dc system is provided with detectors to indicate and alarm when there is a ground existing on any part of the system. During plant operation, normal maintenance may be performed. Complete provisions for the testing of Class 1E electric power systems and the standby power supplies (diesel generators) are described in Chapter 8.0. 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." DISCUSSION A separate control room is provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain in a safe manner under accident conditions, including LOCAs. Operator action outside of the control room to mitigate the consequences of an accident is permitted. The control room and its post-accident ventilation systems are designed to satisfy seismic Category I requirements, as discussed in Chapter 3.0. Adequate concrete shielding and radiation protection are provided against direct gamma radiation and inhalation doses postulated to result from a release of fission products inside the containment structure. The shielding and the control room standby air-conditioning system allow access to and occupancy of the control rooms under accident conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. Refer to Chapter 15.0. Fission product removal is provided 3.1-17 Rev. 34

WOLF CREEK in the control room recirculation equipment to remove iodine and particulate matter, thereby minimizing the dose which could result from the accident. The control room habitability features are described in Chapter 6.0. In the event that the operators are forced to abandon the control room, panel-mounted local instrumentation and controls are provided to achieve and maintain the plant in the hot shutdown condition (see Chapter 7.0). The capability for bringing the plant to a cold shutdown is also provided outside the control room through the use of local controls. 3.1.5 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." DISCUSSION A fully automatic protection system with appropriate redundant channels is provided to cope with transient events where insufficient time is available for manual corrective action. The design basis for all protection systems is in accordance with the intent of IEEE Standards 279-1971 and 379-1972. The reactor protection system automatically initiates a reactor trip when any variable monitored by the system or combination of monitored variables exceeds the normal operating range. Setpoints are designed to provide an envelope of safe operating conditions with adequate margin for uncertainties to ensure that the fuel design limits are not exceeded. Reactor trip is initiated by removing power to the rod drive mechanisms of all the rod cluster control assemblies. This causes the rods to insert by gravity, thus rapidly reducing the reactor power. The response and adequacy of the protection system have been verified by analysis of anticipated transients. The engineered safety features actuation system automatically initiates emergency core cooling and other safety functions by sensing accident conditions, using redundant analog channels measuring diverse variables. Manual actuation of safety features 3.1-18 Rev. 34

WOLF CREEK may be performed where ample time is available for operator action. The engineered safety features actuation system automatically trips the reactor on a manual or automatic safety injection signal. CRITERION 21 - PROTECTION SYSTEM RELIABILITY AND TESTABILITY "The protection system shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed. Redundancy and independence designed into the protection system shall be sufficient to assure that (1) 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." DISCUSSION The protection system is designed for high functional reliability and in-service testability. The design employs redundant logic trains and measurement and equipment diversity. The reliability of the system has been verified by analysis which is documented by Reference 1. The protection system, including the engineered safety features test cabinet, is designed to meet Regulatory Guide 1.22 and conform to the requirements of IEEE Standards 279-1971 and 379-1972. Functions that cannot be tested with the reactor at power are tested during shutdown, as allowed by the regulatory guide and the above standards. In cases where actuated equipment cannot be tested at power, the channels and logic associated with this equipment, up to the final actuation device, have the capability for testing at power. Such testing discloses failures or reduction in redundancy which may have occurred. Removal from service of any single channel or component does not result in the loss of minimum required redundancy. For example, a two-of-three function is placed in the one-of-two mode when one channel is removed. (Note that distinction is made between channels and trains in this discussion. A train may be removed from service only during testing.) 3.1-19 Rev. 0

WOLF CREEK Semiautomatic testers are built into each of the two logic trains of the protection system. These testers have the capability of testing the system logic very rapidly while the reactor is at power. A self-testing provision is designed into each tester. For a detailed description of reliability and testability of the Westinghouse portion of the protection system, refer to Reference 2. 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 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." DISCUSSION Design of the protection systems includes consideration of natural phenomena, normal maintenance, testing, and accident conditions so that the protection functions are always available. Protection system components are designed, arranged, and qualified so that the environment accompanying any emergency situation in which the components are required to function does not result in the loss of the safety function. Functional diversity has been designed into the system. The extent of this functional diversity has been evaluated for a wide variety of postulated accidents. Diverse protection functions will automatically terminate an accident before intolerable consequences can occur. Sufficient redundancy and independence is designed into the protection systems to assure that no single failure or removal from service of any component or channel of a system would result in loss of the protection function. Functional diversity and consequential location diversity are designed into the system. Automatic reactor trips are based upon neutron flux measurements, reactor coolant loop temperature measurements, pressurizer pressure and level measurements, and reactor coolant pump power supply underfrequency and undervoltage measurements. Trips may also be initiated manually or by a safety injection signal. See Chapter 7.0 for details. 3.1-20 Rev. 0

WOLF CREEK High quality components, conservative design and applicable quality control, inspection, calibration, and tests are utilized to guard against common-mode failure. Qualification testing is performed on the various safety systems to demonstrate functional operation at normal and postaccident conditions of temperature, humidity, pressure, and radiation for specified periods, if required. Typical protection system equipment is subjected to type tests under simulated seismic conditions, using conservatively large accelerations and applicable frequencies. The test results indicate no loss of the protection function. Refer to Sections 3.10(B), 3.10(N), 3.11(B) and 3.11(N) for further details. 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." DISCUSSION The protection system is designed with consideration of the most probable failure modes of the components under various perturbations of the environment and energy sources. Each reactor trip channel is designed on the de-energize-to-trip principle so loss of power, disconnection, open channel faults, and the majority of the internal channel short circuit faults cause the channel to go into its tripped mode. Similarly, that portion of the engineered safety features actuation system provided for actuation of auxiliary feedwater system and containment ventilation isolation is designed to fail into a safe state, except for the final output relays. The relays are energized to actuate as are the pumps and motor-operated valves of the actuated equipment. For a more detailed description of the protection system, refer to Chapter 7.0. CRITERION 24 - SEPARATION OF PROTECTION AND CONTROL SYSTEMS "The protection system shall be separated from control systems to the extent that failure of any single control system component or channel, or failure or removal from service of any single protection system component or channel which is common to the control and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the 3.1-21 Rev. 0

WOLF CREEK protection system. Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired." DISCUSSION The protection system is separate and distinct from the control systems, as described in Chapter 7.0. Control systems are, in some cases, dependent on the protection system in that control signals are derived from protection system measurements, where applicable. These signals are transferred to the control system by isolation devices which are classified as protection components. The adequacy of the system isolation has been verified by testing under conditions of postulated credible faults. The 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 system, leaves intact a system which satisfies the requirements of the protection system. Distinction between channel and train is made in this discussion. The removal of a train from service is allowed only during testing of the train. 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 control rods." DISCUSSION 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 completely independent of the normal control function since the trip breakers interrupt power to the rod mechanisms regardless of existing control signals. Thus, in the postulated accidental withdrawal 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. Analyses of the effects of possible malfunctions are discussed in Chapter 15.0. These analyses show that for postulated boron dilution during refueling, startup, manual or automatic operation at power, hot standby, or cold shutdown, the operator has ample time to determine the cause of dilution, terminate the source of 3.1-22 Rev. 0

WOLF CREEK dilution, and initiate reboration before the shutdown margin is lost. Either manual or automatic controls can be used to terminate dilution and initiate boration. The analyses show that acceptable fuel damage limits are not exceeded even in the event of a single malfunction of either system. 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." DISCUSSION Two reactivity control systems are provided. These are rod cluster control assemblies (RCCAs) and chemical shim (boric acid). The RCCAs 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. The shutdown rod banks, along with the control banks, are designed to shut down the reactor with adequate margin under conditions of normal operation and anticipated operational occurrences, thereby ensuring that specified fuel design limits are not exceeded. The most restrictive period in the core life is assumed in all analyses, and the most reactive rod cluster is assumed to be in the fully withdrawn position. The boron system will maintain the reactor in the cold shutdown state independent of the position of the control rods and can compensate for xenon burnout transients. Details of the construction of the RCCAs are presented in Chapter 4.0, and the operation is discussed in Chapter 7.0. The means of controlling the boric acid concentration is described in Chapter 9.0. Performance analyses under accident conditions are included in Chapter 15.0. 3.1-23 Rev. 0

WOLF CREEK 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." DISCUSSION The facility is provided with means of making and holding the core subcritical under any anticipated conditions and with appropriate margin for contingencies. These means are discussed in detail in Chapters 4.0 and 9.0. Combined use of the rod cluster control system and the chemical shim control system permits the necessary shutdown margin to be maintained during long-term xenon decay and plant cooldown. The single highest worth control cluster is assumed to be stuck full out upon trip for this determination. CRITERION 28 - REACTIVITY LIMITS "The reactivity control system shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently 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 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." DISCUSSION The maximum reactivity worth of the control rods and the maximum rates of reactivity insertion employing control rods and boron removal are limited to values that prevent any reactivity increase from rupturing the reactor coolant system boundary or disrupting the core or vessel internals to a degree that could impair the effectiveness of emergency core cooling. The appropriate reactivity insertion rate for the withdrawal of RCCAs and the dilution of the boric acid in the reactor coolant systems are specified in the Technical Specifications for the facility. The COLR includes appropriate graphs that show the permissible withdrawal limits and overlap of the RCCA banks as 3.1-24 Rev. 13

WOLF CREEK a function of power. These data on reactivity insertion rates, dilution, and withdrawal limits are also discussed in Chapter 4.0. The capability of the chemical and volume control system to avoid an inadvertent excessive rate of boron dilution is discussed in Chapter 9.0. The relationship of the reactivity insertion rates to plant safety is discussed in Chapter 15.0. Core cooling capability following accidents, such as rod ejection, steam line break, etc., is assured 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 features. CRITERION 29 - PROTECTION AGAINST ANTICIPATED OPERATIONAL OCCUR-RENCES "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." DISCUSSION The protection and reactivity control systems have an extremely high probability of performing their required safety functions in any anticipated operational occurrences. Diversity and redundancy, coupled with a rigorous quality assurance program and analyses, support this probability as does operating experience in plants using the same basic design. Failure modes of system components are designed to be safe modes. Loss of power to the protection system results in a reactor trip. Details of system design are covered in Chapters 4.0 and 7.0. 3.1.6 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." DISCUSSION All reactor coolant system components are designed, fabricated, inspected, and tested in conformance with the ASME Boiler and Pressure Vessel Code, Section III. 3.1-25 Rev. 0

WOLF CREEK All components are classified according to ANSI-N18.2-1973 and are accorded all the quality measures appropriate to this classification except for the deviation described in section 3.2.3. The design bases and evaluations of the reactor coolant system are discussed in Chapter 5.0. A number of methods are available for detecting reactor coolant leakage. The reactor vessel closure joint is provided with a temperature monitored leakoff between double gaskets. Leakage inside the reactor containment is drained to the reactor building sump where the level is monitored. Leakage is also detected by measuring the airborne activity and humidity of the containment. Monitoring the inventory of reactor coolant in the system at the pressurizer, volume control tank, and coolant drain collection tank provides an accurate indication of integrated leakage. Refer to Chapter 5.0 for complete description of the reactor coolant pressure boundary leakage detection system. 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 (1) 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 (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state, and transient stresses, and (4) size of flaws." DISCUSSION Close control is maintained over material selection and fabrication for the reactor coolant system to assure that the boundary behaves in a nonbrittle manner. The reactor coolant system materials which are exposed to the coolant are corrosion-resistant stainless steel or Inconel. The reference temperature (RTNDT) of the reactor vessel structural steel is established by Charpy V-notch and drop weight tests in accordance with 10 CFR 50, Appendix G, "Fracture Toughness Requirements." The reactor vessel specification imposes the following requirements which are not specified by the ASME code:

a. The performance of a 100-percent volumetric ultrasonic test of reactor vessel plate for shear wave and a post-hydrotest ultrasonic map of all welds in the pressure 3.1-26 Rev. 19

WOLF CREEK vessel are required. Cladding bond ultrasonic inspection to more restrictive requirements than those specified in the code is also required to preclude interpretation problems during inservice inspection.

b. 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 1/2 T 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-185-79, "Standard Recommended Practice of Surveillance Tests for Nuclear Reactor Vessels," and the requirements of 10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements."
c. 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 equivalent to those used for the reactor vessel. The inspections of reactor vessel, pressurizer, piping, pumps, and steam generators are governed by ASME code requirements. Refer to Chapter 5.0 for details. Allowable pressure-temperature relationships for plant heatup and cooldown rates are calculated, using methods derived from the ASME Code, Section III, Appendix G, "Protection Against Non-ductile Failure." The approach specifies that allowed stress intensity factors for all vessel operating conditions shall not exceed the reference stress intensity factor (KIR) for the metal temperature at any time. Operating specifications include conservative margins for predicted changes in the material reference temperatures (RTNDT) due to irradiation. CRITERION 32 - INSPECTION OF REACTOR COOLANT PRESSURE BOUNDARY "Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of important areas and features to assess their structural and leaktight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel." 3.1-27 Rev. 1

WOLF CREEK DISCUSSION The design of the reactor coolant pressure boundary provides accessibility to the entire internal surfaces of the reactor vessel and most external zones of the vessel, including the nozzle to reactor coolant piping welds, the vessel shell beneath the nozzles, the top and bottom heads, and external surfaces of the reactor coolant piping, except for the area of pipe within the primary shielding concrete. The inspection capability complements the leakage detection systems in assessing the pressure boundary component's integrity. The reactor coolant pressure boundary will be periodically inspected under the provisions of the ASME Code, Section XI. Monitoring of changes in the fracture toughness properties of the reactor vessel core region plates forging, weldments, and associated heat treated zones is performed in accordance with 10 CFR 50, Appendix H. Samples of reactor vessel plate materials are retained and cataloged in case future engineering development shows the need for further testing. The material properties surveillance program includes not only the conventional tensile and impact tests, but also fracture mechanics specimens. The observed shifts in RTNDT of the core region materials with irradiation will be used to confirm the allowable limits calculated for all operational transients. The design of the reactor coolant pressure boundary piping provides for accessibility of all welds requiring inservice inspection under the provisions of the ASME Code, Section XI. Removable insulation is provided at all welds requiring inservice inspection. The inservice inspection program is discussed in detail in Chapter 5.2.4. 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 using the piping, pumps, and valves used to maintain coolant inventory during normal reactor operation." 3.1-28 Rev. 1

WOLF CREEK DISCUSSION The chemical and volume control system provides a means of reactor coolant makeup and adjustment of the boric acid concentration. Makeup is added automatically if the level in the volume control tank falls below a preset level. The high pressure centrifugal charging pumps provided are capable of supplying the required make-up and reactor coolant seal injection flow when power is available from either onsite or offsite electric power systems. These pumps also serve as high head safety injection pumps. Functional reliability is assured by provision of standby components assuring a safe response to probable modes of failure. Details of system design, including descriptions of the effects of small piping and component ruptures, are provided in Sections 6.3 and 9.3 and Chapter 15.0, with details of the electric power system included in Chapter 8.0. 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." DISCUSSION The residual heat removal system, in conjunction with the steam and power conversion system, is designed to transfer the fission product decay heat and other residual heat from the reactor core at a rate which keeps the fuel within acceptable limits. The residual heat removal system functions when temperature and pressure are below approximately 350°F and 425 psig, respectively. The design of the RHRS includes two motor-operated isolation valves that are closed during normal operations. They are provided with both a prevent-open interlock and RHRS-Iso-Valve-Open alarm which are designed to prevent possible exposure of the RHRS to normal RCS operating pressure. The isolation valves are opened for residual heat removal during a plant cooldown after the RCS temperature is reduced to approximately 350°F and RCS pressure is less than approximately 360 psig. During a plant startup, the inlet isolation valves are shut after drawing a bubble in the pressurizer and prior to increasing RCS pressure above approximately 425 psig (alarm setpoint). Redundancy of the residual heat removal system is provided by two residual heat removal pumps (located in separate flood-proof compartments, with means available for draining and monitoring leakage), two heat exchangers, and associated piping, cabling, and 3.1-29 Rev. 11

WOLF CREEK electric power sources. For a more detailed description of residual heat removal system redundancy, refer to Section 5.4.7. The residual heat removal system is able to operate on either the onsite or offsite electrical power system. Heat removal at temperatures above approximately 350°F is provided by the four steam generators, four atmospheric relief valves, and the auxiliary feedwater system. Details of the Residual Heat Removal system design are provided in Section 5.4.7. Refer to sections 7.3.6 and 10.4.9 for discussion of the auxiliary feedwater system. 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 (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts. "Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure." DISCUSSION An emergency core cooling system has the capability to mitigate the effects of any LOCA within the design bases. Cooling water is provided in an emergency to transfer heat from the core at a rate sufficient to maintain the core in a coolable geometry and to assure that clad metal-water reaction is limited to less than 1 percent. Design provisions assure performance of the required safety functions even with a single failure. Emergency core cooling is provided even if there should be a failure of any component in the system. A passive system of four accumulators which do not require any external signals or source of power to operate provide the short-term cooling requirements for large reactor coolant pipe breaks. Two independent and redundant high pressure flow and pumping systems, each capable of the required emergency cooling, are provided for small break protection and to keep the core submerged after the accumulators have 3.1-30 Rev. 10

WOLF CREEK discharged following a large break. These systems are arranged so that the single failure of any active component does not interfere with meeting the short-term cooling requirements. The primary function of the ECCS is to deliver borated cooling water to the reactor core in the event of a LOCA. This limits the fuel-clad temperature, ensures that the core will remain intact and in place, with its essential heat transfer geometry preserved, and prevents a return to criticality. This protection is afforded for:

a. All pipe break sizes up to and including the hypothetical circumferential rupture of the largest pipe of a reactor coolant loop
b. A loss-of-coolant associated with a rod ejection accident The ECCS is described in Chapter 6.0. The LOCA including an evaluation of consequences, is discussed in Chapter 15.0.

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." DISCUSSION The ECCS is accessible for visual inspection and for non-destructive inservice inspection, as required by the ASME Code, Section XI. Components outside the containment are accessible for leaktightness inspection during operation of the reactor. Details of the inspection program for the emergency core cooling system are discussed in Section 6.3. 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 (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, 3.1-31 Rev. 0

WOLF CREEK 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." DISCUSSION The design of the ECCS permits periodic testing of both active and passive components of the ECCS. Preoperational performance tests of the ECCS components are performed by the manufacturer. Initial system hydrostatic and functional flow tests demonstrate structural and leaktight integrity of components and proper functioning of the system. Thereafter, periodic tests demonstrate that components are functioning properly. Each active component of the ECCS may be individually operated on the normal power source or transferred to standby power sources at any time during normal plant operation to demonstrate operability. The centrifugal charging/safety injection pumps are not normally operating but, as part of the charging system, they are available for operation as necessary during plant operation. The test of the safety injection pumps employs the minimum flow recirculation test line which connects back to the refueling water storage tank. Remote-operated valves are exercised and actuation circuits tested. The automatic actuation circuitry, valves, and pump breakers may be checked during integrated system tests performed during a planned cooldown of the reactor coolant system. Design provisions include special instrumentation, testing, and sampling lines to perform certain tests during plant shutdown to help demonstrate proper automatic operation of the ECCS. Several subsystems/components of the ECCS can also be tested during normal plant operation. (Refer to Section 7.1.2.5 & Table 7.1-3 for a discussion of Regualtory Guide 1.22). A test signal is applied to initiate automatic action and verification is made that the safety injection pumps attain required discharge heads. The test demonstrates the operation of the valves, pump circuit breakers, and automatic circuitry. In addition, the periodic testing of the pumps and valves verify the delivery capability of the ECCS. The design provided the capability to test initially, to the extent practical, the full operational sequence up to the design conditions, including transfer to alternate power sources for the ECCS to demonstrate the state of readiness and capability of the system. This functional test was performed with the water level below the reactor pressure vessel flange with the reactor coolant system initially cold and depressurized. 3.1-32 Rev. 12

WOLF CREEK The ECCS valving is set to initially simulate the system alignment for plant power operation. Details of the ECCS are found in Chapter 6.0. Performance under accident conditions is evaluated in Chapter 15.0. Surveillance requirements are identified in the Technical Specifications. CRITERION 38 - CONTAINMENT HEAT REMOVAL "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 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." DISCUSSION The containment spray and containment fan cooler systems, in conjunction with the residual heat removal system, are capable of removing sufficient energy and subsequent decay energy from the containment following the hypothesized LOCA to maintain the containment pressure below the containment design pressure. During the post-accident injection phase, water for the containment spray system and residual heat removal system is drawn from the refueling water storage tank. During the later recirculation phase, spray water and reflood water are pumped from the containment sump. Each of these systems consists of two independent subsystems supplied from separate 1E power busses. No single failure, including loss of onsite or offsite electrical power, can cause loss of more than half of the installed 200-percent cooling capacity. The containment spray system and containment fan coolers are discussed in Chapter 6.0. Electrical facilities are described in Chapter 8.0. A containment pressure and temperature analysis following a LOCA is given in Chapter 6.0 with additional results found in Chapter 15.0. 3.1-33 Rev. 1

WOLF CREEK 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." DISCUSSION The essential equipment of the containment spray system is outside the containment, except for risers, distribution header piping, spray nozzles, and the containment sump. The containment sump, spray piping, and nozzles can be inspected during shutdown. Portions of the containment spray suction piping and the RHR suction piping from the containment recirculation sumps are embedded in concrete and are not accessible for inspection. A portion of the piping from the refueling water storage tank is buried in the ground and not accessible for inspection. Associated equipment outside the containment can be visually inspected. The containment air coolers and associated cooling water system piping inside the containment can be inspected during shutdowns. These periodic inspections assure that the capability of these heat removal systems as specified in the Technical Specification is met. For details on the containment air coolers and containment spray system, see Chapter 6.0. 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 (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole, and under conditions as close to the design as practical the performance of 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." DISCUSSION The containment spray system and the containment fan cooling system are designed to permit periodic testing to assure the 3.1-34 Rev. 0

WOLF CREEK structural and leaktight integrity of their components and to assure the operability and performance of the active components of the systems. All active components of the containment spray system and delivery piping up to the last powered valve before the spray nozzle have the capability to be tested during reactor power operation. In addition, when the unit is shutdown, smoke or air can be blown through the test connections for visual verification of the flow path. All safety-related active components of the containment fan cooling system can be tested to verify operability during reactor power operation. In addition, since the containment fan cooling system is a normally operating system, the performance and operability of portions of the system are continuously verified during normal reactor power operation. The facility design allows, under conditions as close to the design as practicable, the performance of a full operational sequence that brings these systems into operation. More complete discussions of the testing of these systems are in Chapters 6.0, 8.0, and the Technical Specifications. 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." DISCUSSION The containment spray system serves to remove radioiodine and other airborne particulate fission products from the containment atmosphere following a LOCA. The system consists of two independent systems, each supplied from separate electrical power busses, as described in Chapter 8.0. Either subsystem alone can provide the fission product removal capacity for which credit is taken in Chapter 15.0, in compliance with Regulatory Guide 1.183. (See Section 3A for discussion of RG 1.183) 3.1-35 Rev. 34

WOLF CREEK The generation of hydrogen in the containment under post-accident conditions has been evaluated, using the assumptions of Regulatory Guide 1.7 (see Chapter 6.0). A post-accident hydrogen recombiner system is provided with redundancy of vital components so that a single failure does not prevent timely operation of the system. This system is described in Section 6.2.5. A hydrogen purge system is provided as a backup. No single failure causes both subsystems to fail to operate. CRITERION 42 - INSPECTION OF CONTAINMENT ATMOSPHERE CLEANUP SYSTEMS "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." DISCUSSION The containment atmosphere cleanup systems are designed and located so that they can be inspected periodically, as required. The essential equipment of the containment spray system is outside the containment, except for risers, distribution header piping, and spray nozzles in the containment. The hydrogen purge and monitoring components of the hydrogen control system are located outside the containment. The equipment outside the containment may be inspected during normal power operation. Components of the containment spray system and the hydrogen control system located inside the containment can be inspected during refueling shutdowns. See Chapter 6.0 for details on the containment spray system and details of the hydrogen control system. 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 leaktight 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." 3.1-36 Rev. 0

WOLF CREEK DISCUSSION The containment spray system which serves as the containment atmosphere cleanup system can be tested. The operation of the spray pumps can be tested by recirculation to the refueling water storage tank through a test line. The system valves can be operated through their full travel. The system is checked for leaktightness during testing. See Sections 6.2.2 and 6.5.2 for details and Chapter 8.0 for electrical power details. The spray headers and nozzles can be smoke or air tested, as described in the response to Criterion 40. 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." DISCUSSION The component cooling and essential service water systems are provided to transfer heat from plant safety-related components to the ultimate heat sink. These systems are designed to transfer their respective heat loads under all anticipated normal and accident conditions. Suitable redundancy, leak detection, systems interconnection, and isolation capabilities are incorporated in the design of these systems to assure the required safety function, assuming a single failure with either onsite or offsite power. Complete descriptions of the essential service water system and the component cooling water system are given in Chapter 9.0. 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." 3.1-37 Rev. 1

WOLF CREEK DISCUSSION The integrity and capability of the component cooling water system and portions of the essential service water system are monitored during normal operation by alternating operation of the systems between the redundant system components. Normally, inactive portions of the essential service water system are periodically tested. The important components are located in accessible areas with the exception of any underground piping for the essential service water system. These components have suitable manholes, handholes, inspection ports, or other appropriate design and layout features to allow periodic inspection. The integrity of any underground piping will be demonstrated by pressure and functional tests. Piping to and from the containment air coolers is accessible for inspection during reactor shutdown and refueling periods. These systems are discussed in Chapter 9.0. 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 leaktight 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." DISCUSSION The component cooling system operates continuously during normal plant operation and shutdown, under flow and pressure conditions that approximate the accident conditions. The essential service water system distribution piping utilizes the service water system cooling flow, during normal plant operation, at flows and pressures approximating accident conditions. Provisions are incorporated in the design to allow for periodic starting of the essential service water pumps and verification of the required flowpath at pressure conditions approximating the accident conditions. These operations demonstrate the operability, performance, and structural and leaktight integrity of all cooling water system components. 3.1-38 Rev. 0

WOLF CREEK The cooling water system is designed to include the capability for testing through the full operational sequence that brings the system into operation for reactor shutdown and for LOCAs, including operation of applicable portions of the protection system and the transfer between normal and emergency power sources. For a detailed description of the cooling water system, refer to Section 9.2. 3.1.7 REACTOR CONTAINMENT CRITERION 50 - CONTAINMENT DESIGN BASIS "The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, 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 as required by Section 50.44 energy from metal-water and other chemical reactions that may result from degradation but not total failure of emergency core cooling functioning, (2) the limited experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters." DISCUSSION The design of the containment structure is based on the containment design basis accidents which include the rupture of a reactor coolant pipe in the reactor coolant system or the rupture of a main steam line. In either case, the pipe rupture is assumed to be coupled with partial loss of the redundant safety features systems minimum safety features. The maximum pressure and temperature reached for a containment design basis accident are presented in Chapter 6.0. Containment design pressure of 60 psig and the design saturation temperature of 320°F provide ample margin to the design basis limits. See Chapters 3.0 and 6.0 for details. 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 3.1-39 Rev. 1

WOLF CREEK behave in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the containment boundary material during operation, maintenance, testing, and postulated accident conditions, and the uncertainties in determining (1) material properties, (2) residual, steady-state, and transient stresses, and (3) size of flaws." DISCUSSION The containment liner plate is a fully silicon kilned, fine-grain practice, normalized plate 1/4-inch thick. Principal load-carrying components of ferritic materials exposed to the external environment are selected so that their temperatures under normal operating and testing conditions are not less than 30°F above nil ductility transition temperature. Refer to Section 3.8.1 for details. 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." DISCUSSION The containment system is designed and constructed and the necessary equipment is provided to permit periodic integrated leakage rate tests during plant lifetime, in accordance with the requirements of Appendix J of 10 CFR

50. Details concerning the conduct of periodic integrated leakage rate tests are included in Chapter 6.0.

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 leaktightness of penetrations which have resilient seals and expansion bellows." DISCUSSION Provisions exist for conducting individual leakage rate tests on containment penetrations. Penetrations are visually inspected and pressure tested for leaktightness at periodic intervals. Other 3.1-40 Rev. 1

WOLF CREEK inspections are performed as required by Appendix J of 10 CFR 50 as modified by the exemption described in KMLNRC 84-192. Refer to Chapter 6.0. 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." DISCUSSION Piping systems penetrating the primary reactor containment are provided with containment isolation valves. Penetrations which must be closed for containment isolation have redundant valving and associated apparatus. Automatic isolation valves with air or motor operators, which do not restrict normal plant operation, are periodically tested to assure operability. Secondary system piping inside the containment is considered an extension of the containment boundary, as described in Section 6.2.4. The isolation valve arrangements are discussed in Chapter 6.0. Piping that penetrates the containment has been equipped with test connections and test vents or has other provisions to allow periodic leak rate testing to ensure that leakage is within the acceptable limit as defined by the Technical Specifications and Appendix J to 10 CFR 50, as described in Chapter 6.0. The fuel transfer tube is not classified as a fluid system penetration. The blind flange and the portion of the transfer tube inside the containment are an extension of the containment boundary. The blind flange isolates the transfer tube at all times, except when the reactor is shutdown for refueling. This assembly is a penetration in the same sense as are equipment hatches and personnel locks. 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 3.1-41 Rev. 0

WOLF CREEK 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 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 inservice inspection, protection against more severe natural phenomena, and additional isolation valves and containment, shall include consideration of the population density, use characteristics, and physical characteristics of the site environs." DISCUSSION Each line that is a part of the reactor coolant pressure boundary and penetrates the containment is provided with isolation valves meeting the intent of this criterion, except that the reactor shutdown lines (RHR system) which are part of the reactor coolant pressure boundary and which penetrate the containment are provided with two isolation valves in series, both inside the containment. This system is a closed system outside the containment and is constructed to ASME Section III, Class 2, specifications and is considered the second passive barrier to fission product release, as described in Chapter 6.0. The arrangement and type of valves 3.1-42 Rev. 0

WOLF CREEK utilized are discussed in Chapter 6.0. Containment penetrations are seismic Category I and are protected against possible environmental effects, including missiles. CRITERION 56 - PRIMARY CONTAINMENT ISOLATION "Each line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis: (1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or (4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment. "Isolation valves outside containment shall be located as close to the containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety." DISCUSSION Lines which communicate directly with the containment atmosphere and which penetrate 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 above acceptable arrangements. Several penetrations use alternative arrangements which satisfy containment isolation on some other defined bases. Special cases are described in Chapter 6.0. Valving arrangements are combinations of locked shut isolation valves and automatic isolation valves or remote-manual isolation 3.1-43 Rev. 0

WOLF CREEK valves. No simple check valves are utilized as automatic isolation valves outside the containment. Where necessary, provision for leak detection is provided for lines outside the containment. Instrument lines satisfy other acceptable criteria, as described in Chapter 6.0. 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, or locked closed, or capable of remote manual operation. This valve shall be outside containment and located as close to the containment as practical. A simple check valve may not be used as the automatic isolation valve." DISCUSSION All containment penetrations are considered to be covered by either GDC-55 or GDC-56. There are no penetrations to which GDC-57 is considered applicable. For a more detailed discussion of containment isolation, refer to Section 6.2.4. 3.1.8 FUEL AND RADIOACTIVITY 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 normal reactor operation, including anticipated operational occurrences. Sufficient holdup capacity shall be provided for retention of gaseous and liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operational limitations upon the release of such effluents to the environment." DISCUSSION 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 assure that the 3.1-44 Rev. 0

WOLF CREEK discharge of radioactive wastes is maintained as low as practicable below regulatory limits of 10 CFR 20 during normal operation. The radioactive waste processing system, the design criteria, and the amounts of estimated releases of radioactive effluents to the environment are described in Chapter 11.0. 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." DISCUSSION The fuel storage pool and associated cooling system, fuel handling system, and radioactive waste processing system are designed to assure adequate safety under normal and postulated accident conditions. The fuel storage pool cooling system provides cooling to remove residual heat from the fuel stored in the spent fuel pool. The system is designed with redundancy and testability to assure continued heat removal. The fuel storage pool cooling system is described in Section 9.1.3. The fuel storage pool is designed so that no postulated accident could cause excessive loss-of-coolant inventory. Accidents are discussed in Chapter 15.0. Structures, components, and systems are designed and located so that appropriate periodic inspection and testing may be performed. Adequate shielding is provided as described in Chapter 12.0. Radiation monitoring is provided as discussed in Chapters 11.0 and 12.0. Individual components that contain significant radioactivity are in confined areas adequately ventilated through appropriate filtering systems. 3.1-45 Rev. 14

WOLF CREEK 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." DISCUSSION The restraints and interlocks provided for the safe handling and storage of new and spent fuel are discussed and illustrated in Chapter 9.0. Criticality in new and spent fuel storage facilities is prevented both by physical separation of fuel assemblies and, in the fuel storage pool, the presence of borated water and the Boral neutron absorber panels. The center-to-center distance between the adjacent fuel assemblies is sufficient to ensure a keff <0.95, even if unborated water is used to fill the fuel storage pool. New fuel is stored with enough center-to-center distance to ensure a keff <0.98 under conditions of optimum moderation. Layout of the fuel handling area is such that the spent fuel cask cannot traverse the spent fuel storage pool. CRITERION 63 - MONITORING FUEL AND WASTE STORAGE "Appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions." DISCUSSION Instrumentation is provided to detect and alarm, in the control room, excessive temperature or low water level in the spent fuel pool. Area radiation monitors are provided in the fuel storage area for personnel protection and general surveillance. 3.1-46 Rev. 14

WOLF CREEK These area monitors alarm locally and in the control room. Normally, the fuel building ventilation system removes radioactivity from the atmosphere above the fuel storage pool and discharges it by way of the plant vent. The ventilation system is continuously monitored by gaseous, particulate, and radio-iodine radiation monitors. If radiation levels reach a predetermined point, an alarm will sound in the control room and the ventilation discharge path will automatically be transferred through filter adsorber units which provides adequate filtration before discharge from the plant vent. See Chapters 7.0, 9.0, and 12.0 for details. 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." DISCUSSION The containment atmosphere is continually monitored during normal and transient station operations, using the containment particulate, gaseous, and radio-iodine radiation monitors. Under accident conditions, samples of the containment atmosphere provide data on existing airborne radioactive concentrations within the containment. Area radiation monitors located in the auxiliary and radwaste buildings are provided to continually monitor radiation levels in the spaces which contain components for recirculation of LOCA fluids and components for processing radioactive wastes. Radioactivity levels contained in the facility effluent and discharge paths and in the plant environs are continually monitored during normal and accident conditions by the station radiation monitoring systems. In addition to the installed detectors, periodic plant environmental surveillance is established. Measurement capability and reporting of effluents will meet the recommendations of Regulatory Guides 4.1 and 1.21. Radiation monitoring systems are discussed in Sections 11.5, 12.3.4, and Chapter 18. 3.1-47 Rev. 14

WOLF CREEK 3.

1.9 REFERENCES

1. Gangloff, W. C. and Loftus, W. D., "An Evaluation of Solid State Logic Reactor Protection in Anticipated Transients,"

WCAP-7706-L(Proprietary) and WCP-7706 (Non-Proprietary), July, 1971.

2. Katz, D.N., "Solid State Logic Protection System Description,"

WCAP-7488-L (Proprietary), January, 1971 and WCAP-7672 (Non-Proprietary), June, 1971.

3. Westinghouse Electric Corporation Reference Safety Analysis Report, RESAR-3, Chapter 3.1.1, Pages 3.1-3 and 3.1-2 dated June 1972.

3.1-48 Rev. 26

WOLF CREEK 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS Certain structures, components, and systems of the nuclear plant are considered to serve a safety function because they:

a. Assure the integrity of the reactor coolant pressure boundary.
b. Assure the capability to shut down the reactor and maintain it in a safe condition.
c. Assure the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of 10 CFR 50.67.
d. Contain or may contain radioactive material.

The purpose of this section is to classify structures, systems, and components according to the importance of the item in order to provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public. Table 3.2-1 delineates each of the items in the plant which fall under the above-mentioned categories and the respective associated classification that the NRC, ANS, and industrial codes committees have developed. Each of the classification categories in Table 3.2-1 is addressed in the following sections. For identification of system and subsystem boundaries, Table 3.2-1 is supplemented (i.e., referenced to applicable figures) by piping and instrument diagrams which have been marked to clearly show the limits of the seismic Category I and various quality group classifications on a system. The legend for the piping and instrument diagrams is provided in Figure 1.1-1. Classification of power supplies, instrumentation and controls, motors, piping and valves, ductwork and dampers, and associated supports, hangers, and restraints is not delineated in Table 3.2-1 because of the extensive listing required. Their classification, however, is consistent with the boundaries shown on the piping and instrumentation drawings. A listing of the piping and instrumentation drawings and their associated USAR figures is found in Table 1.7-2. For ISFSI system structures and components, refer to the NUHOMS EOS System UFSAR, Docket 72-1042 for details of safety, seismic and quality classification, graded quality program requirements, and applicable engineering codes and standards. 3.2-1 Rev. 35

WOLF CREEK 3.2.1 SEISMIC CLASSIFICATION Seismic classification criteria are set forth in 10 CFR 100 and supplemented by Regulatory Guide 1.29. Clarifications and specific exceptions to Regulatory Guide 1.29 are discussed in Table 3.2-3. All components classified as Safety Class 1, 2, or 3 (classifications are as defined by Reference 1), are seismic Category I. Seismic Category I structures, components, and systems are designed to withstand the safe shutdown earthquake (SSE), as discussed in Sections 3.7(B) and 3.7(N), and other applicable load combinations, as discussed in Sections 3.8.1 through 3.8.5. Seismic Category I structures are sufficiently isolated or protected from the other structures to ensure that their integrity is maintained. Radwaste systems and structures are designated as nonseismic Category I. In accordance with Regulatory Guide 1.143, a simplified seismic analysis is performed for portions of the gaseous radwaste system (which by design are intended to store and delay the release of gaseous radioactive waste), including isolation valves, equipment, interconnecting piping, and components located between the upstream and downstream valves used to isolate these components from the rest of the system. In addition, a simplified seismic analysis is performed for structures housing radioactive waste management systems in accordance with Regulatory Guide 1.143. In addition a simplified seismic analysis is performed for structures housing radioactive waste management systems in accordance with Regulatory Guide 1.143, except for the Mixed Waste Storage Facility located in the Owens Corning Building. Mixed waste is stored in barrels which are precluded from tipping over during a seismic event. Also, the total curie content of this building is limited below the limits of 10CFR20 and 100 (see sections 11.4.2.3.5 and 11.4.3). Nonsafety-related structures, systems, and components that must be designed to retain structural integrity during and after an SSE, but do not have to function, are seismically analyzed to ensure that faulted stress limits are not exceeded. These items (for example: piping and piping supports for nonsafety-related piping located over safety-related items) whose continued function is not required are nonseismic Category I and are not controlled by a 10 CFR 50 Appendix B Quality Assurance Program (not Q-listed). The nonseismic Category I Systems Quality Assurance Program is described in Section 17.D of the SNUPPS Quality Assurance Programs Manual for Design and Construction. 3.2-2 Rev. 6

WOLF CREEK 3.2.2 SYSTEM QUALITY GROUP CLASSIFICATION The quality group classification for each water- and steam-containing pressure component is shown in Table 3.2-1. The components are classified according to their safety significance as dictated by service and functional requirements and by the consequences of their failure. The quality group classifications and code requirements for the quality of plant process systems meet the intent of Regulatory Guides 1.26 and 1.143. Clarifications and specific exceptions to these guides are discussed in Tables 3.2-4 and 3.2-5, respectively. These tables compare the design to each regulatory position. The design, fabrication, inspection, and testing requirements of each classification provide the required degree of conservatism in assuring component pressure integrity and operability. Radioactive waste management systems are designed consistent with Regulatory Guide 1.143, as noted in Tables 3.2-1, 3.2-2 and 3.2-5. The radioactive waste management systems are considered to begin at the interface valve(s) in each line, from other systems provided for collecting wastes that may contain radioactive materials, and to include related instrumentation and control systems. The radioactive waste management systems terminate at the point of controlled discharge to the environment, at the point of recycle back to storage for reuse in the reactor, or at the point of storage of packaged solid wastes prior to shipment offsite to a licensed burial ground. The steam generator blowdown system begins at, but does not include, the outermost isolation valve on the blowdown line and terminates at the point of controlled discharge to the environment, at the point of interface with other liquid waste systems, or at the point of recycle back to the secondary system. The code requirements applicable to each quality group classification are identified in Table 3.2-2. The quality group classifications and the interfaces between classifications in a system having components of different classifications are indicated on the piping and instrumentation diagram or flow diagram of that system. 3.2.3 SAFETY CLASSES Table 3.2-1 lists the safety class assigned to applicable systems and components in accordance with ANSI N18.2 (Ref. 1). The criteria (of Ref. 1) is used in the plant design to provide an added degree of assurance that the plant is designed, constructed, and operated without undue risk to the health and safety of the public. Exceptions are portions of the Containment Purge system that perform isolation of Containment if the isolation valves fail. These sections of piping are safety-related in order to provide a sufficient level of safety to protect the health and safety of the public. All components located within the reactor coolant pressure boundary (as defined by 10CFR50.2) are classified as required by 10CFR50.55a with the exception of the pressurizer upper level instrument lines, the pressurizer safety valve loop seal drain lines, 3/4 and smaller branch lines connected to the pressurizer relief lines, and the associated components. These lines are Safety Class 2 although a rupture of one of these lines may result in a rapid depressurization of the reactor coolant system and ECCS actuation on low pressurizer pressure. See Section 5.2.1.1 for additional information. 3.2-3 Rev. 35

WOLF CREEK 3.2.4 QUALITY ASSURANCE PROGRAM Quality assurance practices, in accordance with the program outlined in 10 CFR 50, Appendix B, have been applied to activities which influence the ability of items in Safety Classes 1, 2, and 3 to perform their intended safety function. The quality assurance programs for design and construction is described in Chapter 17.0 of the SNUPPS PSAR. Those Q-listed items which fall under a quality assurance program are identified in Table 3.2-1. In addition to the 10 CFR 50, Appendix B, quality assurance program for the safety-related items shown as Q-listed on Table 3.2-1, a quality program is implemented for those portions of the nonsafety-related structures, systems, or components whose continued function is not required but whose failure could degrade the performance of safety-related items required to maintain the plant in a post accident safe shutdown condition, for interface points between seismic Category I and nonseismic Category I piping, and for the applicable portions of the fire protection system. 3.2.5 ENGINEERING CODES AND STANDARDS The engineering codes and standards are listed in Table 3.2-1. For those components covered by the system quality group classification and the safety classes, the codes and standards employed meet the given classification requirements. The designs of areas and equipment involving the safety and health of personnel include consideration of the Occupational Safety and Health Administration (OSHA) Requirements, 29 CFR 1910. 3.2.6 LOCATION Table 3.2-1 identifies the location of each item by building. 3.

2.7 REFERENCES

1. "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," ANSI N18.2, November 1973.

3.2-4 Rev. 35

WOLF CREEK INDEX TO TABLE 3.2-1 Sheet 1.0 NSSS AND NUCLEAR AUXILIARY SYSTEMS 1 1.1 Reactor Coolant System 1 1.2 Chemical and Volume Control System 3 1.3 Residual Heat Removal System 4 1.4 Safety Injection System 5 1.5 Containment Spray System 5 1.6 Containment Cooling System 5 1.7 Containment Isolation 6 1.8 Containment Hydrogen Control System 6 2.0 WATER SYSTEMS 6 2.1 Service Water System 6 2.2 Essential Service Water System 6 2.3 Component Cooling Water System 6 2.4 Fuel Pool Cooling and Cleanup System 6 2.5 Ultimate Heat Sink 7 3.0 FUEL HANDLING AND STORAGE 7 4.0 RADWASTE MANAGEMENT SYSTEMS 8 4.1 Boron Recycle System 8 4.2 Liquid Radwaste System 8 4.3 Gaseous Radwaste System 10 4.4 Steam Generator Blowdown System 10 4.5 Solid Radwaste System 11 5.0 SECONDARY CYCLE SYSTEMS 12 5.1 Main Steam System 12 5.2 Main Feedwater System 12 5.3 Chemical Addition System 12 5.4 Auxiliary Feedwater System 12 5.5 Turbine-Gland Sealing System 13 5.6 Condenser Air Removal System 13 5.7 Condensate Demineralizer System 13 5.8 Secondary Liquid Waste System 13 5.9 Condensate Storage and Transfer System 14 Rev. 0

WOLF CREEK INDEX TO TABLE 3.2-1 (Sheet 2) Sheet 6.0 SERVICE SYSTEMS 14 6.1 Auxiliary Steam 14 6.2 Standby Diesel Generator Engine 14 6.3 Emergency Fuel Oil System 15 6.4 Compressed Air 16 6.5 Service Gases 16 6.6 Fire Protection 16 6.7 Floor and Equipment Drainage System 16 6.8 Nuclear Sampling System 16 6.9 Process Sampling System 16 7.0 HEATING, VENTILATING, AND AIR CONDITIONING 17 7.1 Control Building 17 7.2 Fuel Building 18 7.3 Auxiliary Building 18 7.4 Diesel Generator Building Ventilation System 19 7.5 Auxiliary, Fuel, Radwaste, Turbine Buildings, Access Control Exhaust HVAC, and Containment Purge 19 7.6 Essential Service Water Pump House HVAC 19 7.7 Containment Purge System 20 7.8 Miscellaneous Building HVAC 20 8.0 CIVIL/ARCHITECTURAL 20 8.1 Structures and Buildings 20 8.2 Materials for Category I Structures 21 9.0 CONTROL AND INSTRUMENTATION 22 10.0 ELECTRICAL POWER SYSTEMS 22 10.1 Class IE Lower Medium Voltage System 22 10.2 Class IE Low Voltage System 23 10.3 Class IE 125 V DC System 23 10.4 Class IE Instrument AC Power 23 10.5 Reactor Building Cable Penetrations 23 10.6 Conduit Supports and Tray Supports 23 10.7 Raceway Installation 23 10.8 Load Shedding and Emergency Load Sequencing 23 10.9 Auxiliary Relay Racks 23 10.10 Transformers 23 10.11 Status Indicating Systems 23 10.12 Local Control Stations 23 Rev. 0

WOLF CREEK TABLE 3.2-1 CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks 1.0 NSSS AND NUCLEAR AUXILIARY SYSTEMS 1.1 Reactor Coolant System (Figure 5.1-1) Reactor Vessel and Appurtenances Vessel Y A 1 Y-W1 III-1 C Head Y A 1 Y-W1 III-1 C Studs Y A 1 Y-W1 III-1 C Shoes and shims Y A 1 Y-W2 III-1 C Supports Y A 1 Y-B III-1 C Lower internals structure Y NA 2 Y-W3 NA C Upper internals structure Y NA 2 Y-W3 NA C Irradiation specimen baskets Y NA 2 Y-W3 III/NG C Irradiation capsules N NA NNS N NA C Irradiation specimens N NA NNS N NA C Fuel assemblies and appurtenances Y NA NA Y-W3 NA C Control rods Y NA NA Y-W3 NA C Primary source rods Y NA NA Y-W3 NA C Burnable poison rod assemblies Y NA NA Y-W3 NA C Thimble guide tubing Y NA 2 Y-W2 III-2 C Thimble guide couplings Y NA 2 Y-W2 III-2 C Thimble seal table and parts Y NA 1 Y-W3 NA C Flux thimble assembly Y NA 2 Y-W1 NA C Control rod drive mechanism (CRDM), housing only Y NA 1 Y-W3 III-1 C Non-class 1E power supply CRDM head adapter plugs Y NA 1 Y-W3 III-1 C CRDM dummy can assemblies N NA NNS N NA C CRDM air cool baffle The CRDM shroud is assemblies (shroud) N NA NNS N NA C seismically qualified CRDM seismic support platform, spacer plates and tie rods Y NA 1 Y-W1 III-1 C Thermal sleeves Y NA 2 Y-W3 NA C Steam generator Tube side - RC Y A 1 Y-W3 III-1 C Shell side - main Y B 2 Y-W3 III-2(7) steam and feedwater Pressurizer Y A 1 Y-W3 III-1/NEMA C Pressurizer heaters N NA 1/NNS(1) N NA C Power supply is diesel-backed non-Class 1E Flux Mapping Frame N NA NNS N NA C RC Thermowell NR Y A 1 Y-A III-1 C RC thermowell WR Y A 1 Y-W2 III-1 C Pressurizer relief tank N D NNS N VIII C The PRT is a seismically qualified Section VIII component RC pump standpipe and N D NNS N VIII C orifice RC pump: C Casing and supports Y A 1 Y-W3 III-1 Main flange Y A 1 Y-W3 III-1 Thermal barrier Y A 1 Y-W3 NA Thermal barrier heat exchanger Y A 1 Y-W3 III-1 Rev. 16

WOLF CREEK TABLE 3.2-1 (Sheet 2) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks 

 #1 Seal housing            Y             A         1           Y-W3         III-2
 #2 Seal housing            Y             B         2           Y-W3         NA
 #3 Seal housing            N             D         NNS         N            NA Bolting (Pressure-retaining)       Y             A         1           Y-W3         III-1 RC Pump Motors                                                                                       Power supply is non-class 1E Shaft coupling             Y             NA        2           Y-W3         NA               C Spool piece                Y             NA        2           Y-W3         NA Armature                   Y             NA        2           Y-W3         NA Flywheel                   Y             NA        2           Y-W3         NA Motor bolting              Y             NA        2           Y-W3         NA Upper oil cooler Tube side-CCW            Y             NA        3           Y-W3         III-3 Shell side - oil         Y             NA        3           Y-W3         NA Lower oil cooler Tube cooling coil        Y             NA        3           Y-W3         III-3 Oil reservoir            Y             NA        3           Y-W3         NA Air water coolers          Y             C         3           Y-W3         III-3 Motor stand and frame      Y             NA        2           Y-W3         NA Piping RC hot, cold, and          Y             A         1           Y-W1         III-1            C crossover leg piping, fittings and fabrication Surge pipe, fittings       Y             A         1           Y-W1         III-1            C and fabrication System to miscellaneous Y                A         1           Y-B          III-1            C boundary valves Pressurizer spray line     Y             A         1           Y-B          III-1            C Pressurizer relief and     N             D         NNS         N            B31.1            C safety valves to pres-surizer relief tank Pressurizer to relief/     Y             A         1           Y-B          III-1            C safety valves Piping/valves              Y             A         1           Y-W1         III-1            C Piping/valves              Y             B         2           Y-B          III-2            C Piping/valves              Y             C         3           Y-B          III-3            C Piping/valves              N             D         NNS         N            B31.3            A/C/R Valves Pressurizer safety         Y             A         1           Y-W1         III-1            C valves Pressurizer power-oper-    Y             A         1           Y-W1         III-1            C     Class 1E power ated relief valves                                                                               supply PORV Block Valves          Y             A         1           Y-W1         III-1            C     Class 1E power supply Valves to RCS boundary     Y             A         1           Y-W1         III-1            C Pressurizer relief tank    N             D         NNS         N            B31.1            C boundary valves not re-quired for containment isolation or part of RCS boundary Pressurizer relief tank    Y             C         3           Y-W1         III-1            C boundary valves re-quired to preserve dedicated letdown path for post accident safe shutdown Rev. 19

WOLF CREEK TABLE 3.2-1 (Sheet 3) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks 1.2 Chemical and Volume Control System (Figure 9.3-8) Letdown and Charging Loop Regenerative heat exchanger Tube side - letdown Y B 2 Y-W1 III-2/TEMA-R C Shell side - charging Y B 2 Y-W1 III-2/TEMA-R C Letdown heat exchanger Tube side - letdown Y B 2 Y-W1 III-2/TEMA-R A Shell side - CCW Y C 3 Y-W1 III-3/TEMA-R A Letdown orifices Y B 2 Y-W2 III-2 A Excess letdown heat exchanger Tube side - letdown Y B 2 Y-W1 III-2/TEMA-R C Shell side - CCW Y C 3 Y-W1 III-3/TEMA-R C Seal water return heat exchanger Tube side - letdown/ Y B 2 Y-W1 III-2/TEMA-R A sealwater Shell side - CCW Y C 3 Y-W1 III-3/TEMA-R A Mixed bed demineral- N D(A) NNS Y-W2 VIII(7) A izers Cation bed demineral- N D(A) NNS Y-W2 VIII(7) A izers Boron meter N D NNS N B31.1 A RC filter Y B 2 Y-W1 III-2 A Volume control tank Y B 2 Y-W1 III-2 A Centrifugal charging Y B 2 Y-W1 III-2 A Class 1E power pump supply. CCW is required. Suction pulsation Y B 2 Y-B III-2 A dampener Discharge pulsation Y B 2 Y-B III-2 A dampener Normal charging pump Y B 2 Y-W1 III-2 A Non-Class 1E power supply. Seal water injection Y B 2 Y-W1 III-2 A filter Seal water return Y B 2 Y-W1 III-2 A filter Boric Acid Makeup Subsystem Boric acid tank Y C 3 Y-B III-3 A Boric acid transfer Y C 3 Y-W1 III-3 A Diesel backed pump non-Class 1E power supply Boric acid filter Y C 3 Y-W2 III-3 A Boric acid batching N D NNS N VIII A tank Boron injection makeup N D NNS N MS A pump Chemical mixing tank N D NNS N VIII A Rev. 10

WOLF CREEK TABLE 3.2-1 (Sheet 4) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks Boron Thermal Regeneration Subsystem Moderating HX Tube side - letdown N D(A) NNS N VIII(7) A Shell side - letdown N D(A) NNS N VIII(7) A Letdown chiller HX Tube side - letdown N D(A) NNS N VIII(7) A Shell side - chilled N D NNS N VIII(7) A wtr Letdown reheat HX Tube side - normal Y B 2 Y-W1 III-2 A charging Shell side - letdown Y D(A) 2 N VIII(7) A Chiller unit N D NNS N NA A Chiller pump N D NNS N MS A Chiller surge tank N D NNS N VIII A Thermal regeneration N D(A) NNS N VIII(7) A demineralizers Piping/valves Y A 1 Y-W1 III-1 C Piping/valves Y B 2 Y-B III-2 A/C Piping/valves Y C 3 Y-B III-3 A/C Piping/valves N D NNS N B31.1 A/C 1.3 Residual Heat Removal System (Figure 5.4-7) RHR Pumps Y B 2 Y-W1 III-2 A Class 1E power supply. CCW required. RHR Heat Exchanger Tube side - RC Y B 2 Y-W1 III-2 A Shell side - CCW Y C 3 Y-W1 III-3 Recirculation valve en- Y B 2 Y-B III-2 A capsulation Piping/valves Y A 1 Y-W1 III-1 C Piping/valves Y B 2 Y-B III-2 A/C Piping/valves Y C 3 YB III-3 A Piping/valves N D NNS N B31.1 A/C 1.4 Safety Injection System (Figure 6.3-1) Accumulators Y B 2 Y-W1 III-2 C Boron injection tank Y B 2 Y-W1 III-2 A Boron injection surge Y C 3 Y-W2 III-3 A tank Refueling water storage Y B 2 Y-B III-2 O tank Safety injection pumps Y B 2 Y-W1 III-2 A Class 1E power supply. CCW required. Boron injection recirc- Y C 3 Y-W2 III-3 A Diesel-backed, ulation pumps non-Class 1E power supply Boron injection flush Y C 3 Y-W2 III-3 A orifices Piping/valves Y A 1 Y-W1 III-1 C Piping/valves Y B 2 Y-B III-2 A/C Piping/valves Y C 3 YB III-3 A Piping/valves N D NNS N B31.1 A/C Rev. 1

WOLF CREEK TABLE 3.2-1 (Sheet 5) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks 1.5 Containment Spray System (Figure 6.2.2-1) Containment spray addi- Y B 2 Y-B III-2 A tive tank Containment spray pump Y B 2 Y-B III-2 A Class 1E power supply Spray additive eductor Y B 2 Y-B III-2 A Spray headers Y B 2 Y-B III-2 C Nozzles Y B 2 Y-B III-2 C Recirculation valve en- Y B 2 Y-B III-2 A capsulation Containment recircula- Y NA 2 Y-B NA C tion sump screen Piping/Valves Y B 2 Y-B III-2 A/C Piping/Valves N D NNS N B31.1 A/C 1.6 Containment Cooling System (Figure 9.4-6) Containment air cooler cooling coil Tube side - ESW Y C 3 Y-B III-3 C Shell side - air Y NA 2 Y-B NA C Containment air Y NA 2 Y-B NA C cooler fan Containment air Y NA 2 Y-B IEEE-334 C Class 1E cooler fan motor power supply Piping/valves Y C 3 Y-B III-3 C Piping (15) N D NNS N B31.1 C Ductwork dampers N NA NNS N NA C 1.7 Containment Isolation Piping Y B 2 Y-B III-2 C/A Flued heads Y B 2 Y-B III-2 C/A Valves Y B 2 Y-B III-2 C/A 1.8 Containment Hydrogen Control System (Figure 6.2.5-1 and 9.4-1) Containment hydrogen Y NA 2 Y-B NEMA C Class 1E recombiner power supply Containment hydrogen Y NA 2 Y-B NA C Class 1E mixing fans power supply Containment hydrogen Y NA 2 Y-B NEMA C mixing fan motors Containment hydrogen Y B 2 Y-B NA A Class 1E analyzer power supply Piping/valves Y B 2 Y-B III-2 A/C Piping/valves (15) N NA NNS N B31.1 A Rev. 7

WOLF CREEK TABLE 3.2-1 (Sheet 6) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks 2.0 WATER SYSTEMS 2.1 Service Water System (Figure 9.2-1) Service water pumps N D NNS N MS I 2.2 Essential Service Water System (11) (Figure 9.2-2) Essential service Y C 3 Y-B III-3 E Class 1E water pump power supply Essential service Y C 3 Y-B III-3 E water pump prelube storage tank Essential service Y C 3 Y-B III-3 E water self-cleaning strainers Essential service Y NA 3 Y-B NA I Class 1E water traveling power supply screens Essential service Y C 3 Y-B III-3 A/B/D/F/I/O water piping Essential service Y C 3 Y-B III-3 E water prelube storage tank filter Piping/valves Y C 3 Y-B III-3 A/B/C/D/E/F/0/V Piping/valves N D NNS N B31.1 A/B/C/D/E/F/R/T/O 2.3 Component Cooling Water System (Figure 9.2-15) Component cooling Y C 3 Y-B III-3 A Class 1E water pump power supply Component cooling wa-ter heat exchanger Tube side - ESW Y C 3 Y-B III-3/TEMA-R A Shell side - CCW Y C 3 Y-B III-3/TEMA-R A Component cooling Y C 3 Y-B III-3 A water surge tank Component cooling N D NNS N VIII A water chemical addition tank Piping/valves Y C 3 Y-C III-3 A/C/F/R Piping/valves N D NNS N B31.1 A/C/F/R 2.4 Fuel Pool Cooling and Cleanup System (Figure 9.1-3) Fuel pool cooling Y C 3 Y-B III-3 F Class 1E pump power supply Fuel pool skimmer N D NNS N MS F pump Fuel pool cleanup N D NNS N MS F pump Fuel pool cooling heat exchanger Tube side - fuel storage pool Y C 3 Y-B III-3/TEMA-R F water Shell side Y C 3 Y-B III-3/TEMA-R 

       - CCW Rev. 30

WOLF CREEK TABLE 3.2-1 (Sheet 7) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks Fuel pool cleanup N D NNS N VIII R demineralizer Skimmer strainer N D NNS N B31.1 F Fuel pool cleanup N D NNS N VIII R filter Skimmer filter N D NNS N VIII R Piping and valves Y C 3 Y-B III-3 F for fuel pool cooling system and essential service water system intertie piping/valves Y C 3 Y-B III-3 C/F Piping/valves N D NNS N ANSI B31.1 C/F/R 2.5 Ultimate Heat Sink (Section 9.2.5) Excavated cooling Y NA 3 Y-U ACI-318-71 O pond and dam 3.0 FUEL HANDLING AND STORAGE Fuel transfer system Non-Class 1E power supply Conveyor system N NA NNS NA NA C/F Remainder of system N NA NNS N NA C/F RCC changing fixture N NA NNS N NA C Fuel transfer Flange Y B 2 Y-W2 III/MC C Tube Y B 2 Y-W2 III/MC C/F Valve N D NNS N MS F Sleeve Y B 2 Y-B III/MC C/F Spent fuel storage Y NA 3 Y-B NA F racks New fuel storage Y NA 3 Y-W2 NA F racks Reactor vessel head N NA NNS N NA C lifting device Reactor vessel missile Y NA NA Y-W C shield AISC Polar crane S NA 3 Y-B NA C Non-Class 1E power supply Refueling machine N NA NNS Y-W2 NA C Non-Class 1E power supply Cask handling crane S NA 3 Y-B NA F Non-Class 1E power supply Spent fuel pool bridge S NA 3 Y-B NA F Non-Class 1E crane power supply Internals lifting N NA NNS N NA C device Spent fuel pool Y NA 3 Y-W2 NA F handling tool Refueling Cavity N NA NNS Y-U NA C Non-Class 1E Elevator power supply Rev. 30

WOLF CREEK TABLE 3.2-1 (Sheet 8) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks 4.0 RADWASTE MANAGEMENT SYSTEMS 4.1 Boron Recycle System (Figure 9.3-11) Tanks Recycle holdup N D(A) NNS N API-650/III-3 R Recycle evaporator N D(A) NNS N VIII R (21) reagent Pumps Recycle evaporator N D(A) NNS N MS(7) R feed Recycle evaporator N D(A) NNS N MS(7) R (21) concentrates Filters Recycle evaporator N D(A) NNS N VIII(7) R feed Recycle evaporator N D(A) NNS N VIII R (21) condensate Recycle evaporator N D(A) NNS N VIII R (21) concentrate Miscellaneous Recycle evaporator N D(A) NNS N VIII(7) R (21) package Recycle evaporator N D(A) NNS N VIII(7) R feed demineralizer Recycle evaporator N D(A) NNS N VIII R (21) condensate demin-eralizer Recycle holdup tank N D(A) NNS N B31.1(7) R vent eductor Piping/valves N D(A) NNS N B31.1 A/R Piping/valves N D NNS N B31.1 A/R 4.2 Liquid Radwaste System (Figure 11.2-1) Tanks Laundry and hot N D(A) NNS N VIII R shower RC drain N D(A) NNS N VIII C Floor drain N D(A) NNS N VIII R Waste holdup N D(A) NNS N VIII R Waste monitor N D(A) NNS N VIII R Chemical drain N D(A) NNS N VIII R Discharge monitor N D(A) NNS N API-650 0 Waste evap. N D(A) NNS N VIII R reagent Waste evap. N D(A) NNS N VIII R condensate Laundry water N D(A) NNS N VIII R storage Rev. 14

WOLF CREEK TABLE 3.2-1 (Sheet 9) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks Pumps RC drain tank N D(A) NNS N MS R Waste evap. feed N D(A) NNS N MS R Waste evap. con- N D(A) NNS N MS R densate tank Chemical drain N D(A) NNS N MS R tank Laundry and hot N D(A) NNS N MS R shower tank Floor drain tank N D(A) NNS N MS R Waste monitor tank N D(A) NNS N MS R Waste evap. distillate N D(A) NNS N MS R Waste evap. concentrate N D(A) NNS N MS R Spent resin sluice N D(A) NNS N MS R Laundry water N D(A) NNS N MS R storage tank Discharge monitor N D(A) NNS N MS R tank transfer Filters Waste evap. feed N D(A) NNS N VIII R Waste evap. con- N D(A) NNS N VIII R densate Laundry and hot N D(A) NNS N VIII R shower Waste monitor tank N D(A) NNS N VIII R Floor drain tank N D(A) NNS N VIII R Miscellaneous RC drain tank heat exchanger Tube side - N D(A) NNS N VIII C RC drains Shell side - CCW Y C 3 Y-W1 III-3 C Laundry and hot N D(A) NNS N NA R shower strainer Waste evaporator N D(A) NNS N VIII R (21) package Waste monitor tank N D(A) NNS N VIII R demineralizer Waste evap. con- N D(A) NNS N VIII R densate demineral-izer Floor drain tank N D(A) NNS N NA R strainer Liquid waste Charcoal adsorber N D(A) NNS N VIII R Laundry and hot shower N NA NNS N VIII R charcoal adsorber Demineralizer Skid N D(A) NNS N VIII R Rev. 27

WOLF CREEK TABLE 3.2-1 (Sheet 10) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks Penetration piping Y B 2 Y-B III-2 C/A Piping/valves Y C 3 Y-B III-3 C Piping/valves N D(A) NNS N B31.1 A/C/E/R/T Piping/valves N D NNS N B31.1 A/R/T 4.3 Gaseous Radwaste System (Figure 11.3-1) Waste gas decay tanks D D(A) NNS N VIII(7) R Waste gas compressor D D(A) NNS N MS/VIII(7) R package Catalatic hydrogen D D(A) NNS N VIII(7) R recombiner package Gas traps D D(A) NNS N VIII(7) R Waste gas drain D D(A) NNS N VIII R filter Gas decay tank drain D D(A) NNS N MS R pump Gaseous radwaste drain N D(A) NNS N VIII R collection tank Piping/valves D D(A) NNS N B31.1 A/R Piping/valves N D NNS N B31.1 A/R 4.4 Steam Generator Blowdown System (Figure 10.4-8) Tanks Surge tank N D(A) NNS N VIII R Pumps Discharge N D(A) NNS N MS R Drain N D(A) NNS N MS A Recirculation N D NNS N MS T Miscellaneous Blowdown regener- N D(A) NNS N VIII T ative heat exchanger Blowdown nonregen- N D(A) NNS N VIII T erative heat exchanger Mixed-bed N D(A) NNS N VIII(7) R demineralizer Filters N D(A) NNS N VIII R Strainers N D(A) NNS N VIII R Penetration piping Y B 2 Y-B III-2 C/A Recirculation sample N D NNS N MS T cooler Piping/valves Y B 2 Y-B III-2 A/C Piping/valves N D(A) NNS N B31.1 B/R/T Piping/valves N D NNS N B31.1 B/R/T Rev. 31

WOLF CREEK TABLE 3.2-1 (Sheet 11) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks 4.5 Solid Radwaste System (Figure 11.4-1) Caustic addition tank N D NNS N VIII R Evaporator bottoms N D(A) NNS N VIII R tank (primary) Evaporator bottoms N D(A) NNS N MS R tank pump (primary) Spent resin tank N D(A) NNS N VIII(7) R (primary) Spent resin tank N D(A) NNS N VIII R (secondary) Spent resin sluice N D(A) NNS N MS(7) R pump (primary) Spent resin sluice N D(A) NNS N MS R pump (secondary) Evaporator bottoms N D(A) NNS N VIII R tank (sec) Evaporator bottoms N D(A) NNS N MS R tank pump (secondary) Acid addition tank N D NNS N VIII R Acid addition N D NNS N MS R metering pump Caustic addition N D NNS N MS R metering pump Resin charging N D NNS N VIII R tank (CVCS) Resin charging N D NNS N VIII R tank (radwaste) Spent resin sluice N D(A) NNS N VIII R filter (primary) Spent resin sluice N D(A) NNS N VIII R filter (secondary) Dry waste compactor N NA NNS N MS R Solid radwaste N NA NNS N NA R bridge crane Rev. 8

WOLF CREEK TABLE 3.2-1 (Sheet 12) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks 5.0 SECONDARY CYCLE SYSTEMS 5.1 Main Steam System (Figure 10.3-1) Piping Penetration (SG Y B 2 Y-B III-2 C/A to isolation valves) To auxiliary FW Y C 3 Y-B III-3 A pump turbine Turbine bypass N D NNS N B31.1 T Other N D NNS N B31.1 T Valves Main steam isolation Y B 2 Y III-2 A Class 1E valves power supply SG safety and Y B 2 Y-B III-2 A Class 1E atmospheric relief power supply valves Piping/valves Y B 2 Y-B III-2 A/C Piping/valves Y C 3 Y-B III-3 A Piping/valves N D NNS N B31.1 A/T Turbine bypass N D NNS N B31.1 T 5.2 Main Feedwater System (Figure 10.4-6) Feedwater heaters N D NNS N VIII/TEMA-C T Heater drain tank N D NNS N VIII T Heater drain pump N D NNS N MS T Feedwater pump N D NNS N MS T Penetration piping Y B 2 Y-B III-2 C/A (isolation valves to SG) Main feedwater Y B 2 Y III-2 A Class 1E isolation valves power supply Motor-driven feedwater N D NNS N MS T pump Reheater drain tank N D NNS N VIII T Moisture separator N D NNS N VIII T drain tank Piping/valves Y B 2 Y-B III-2 A/C Piping/valves N D NNS N B31.1 A/T 5.3 Chemical Addition System (Figure 10.4-7) N D NNS N VIII T 5.4 Auxiliary Feedwater System (Figures 10.4-9 and 10.4-10) Motor-driven Y C 3 Y-B III-3 A Class 1E auxiliary feed- power supply water pump Turbine-driven Y C 3 Y-B III-3 A Class 1E auxiliary feed- power supply water pump Rev. 24

WOLF CREEK TABLE 3.2-1 (Sheet 13) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks Piping/valves Y B 2 Y-B III-2 A Piping/valves Y C 3 Y-B III-3 A Piping/valves N D NNS N B31.1 A/0 5.5 Turbine Gland Sealing System (Figure 10.4-4) Steam packing N D NNS N NA T exhauster Fans N NA NNS N NA T 5.6 Condenser Air Removal System (Figure 10.4-3) Condensers N D NNS N NA T Vacuum pump N NA NNS N NA T Charcoal adsorber N NA NNS N NA T unit 5.7 Condensate Demineralizer System (Figure 10.4-5) Deep-bed condensate N D NNS N VIII T demineralizers Resin separation and N D NNS N VIII T regeneration tank Anion regeneration N D NNS N VIII T tank 5.8 Secondary Liquid Waste System (Figure 10.4-12) SLW evaporator N D(A) NNS N VIII R (21) SLW charcoal adsorber N D NNS N VIII R SLW demineralizer N D NNS N VIII R SLW oil interceptor N D NNS N NA T SLW drain collector N D NNS N VIII T tank SLW monitor tank N D NNS N VIII R SLW drain collector N D NNS N MS T tank pump SLW discharge pump N D NNS N MS R SLW evaporator feed N D NNS N VIII R filter SLW evaporator N D NNS N VIII R reagent tank High TDS transfer N D NNS N VIII T tank High TDS transfer N D NNS N MS T pump High TDS collector N D NNS N VIII T tank High TDS collector N D NNS N MS T tank pump Low TDS transfer N D NNS N VIII T tank Low TDS collector N D NNS N MS T tank pump Low TDS collector N D NNS N VIII T tank Rev. 14

WOLF CREEK TABLE 3.2-1 (Sheet 14) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks Low TDS collector N D NNS N MS T tank pump Low TDS filters N D NNS N VIII R SLW oil intercepter N D NNS N MS T transfer pump Piping/valves N D(A) NNS N B31.1 A/B/R/T Piping/valves N D NNS N B31.1 A/B/R/T (21) 5.9 Condensate Storage and Transfer System (Figure 9.2-12) Condensate storage N D NNS N API 650 O tank Non-safety auxiliary feedwater pump N D NNS N NA T Built to ASME Section III, Class 3, procured as non-safety Piping/valves N D NNS N B31.1 O/T 6.0 SERVICE SYSTEMS 6.1 Auxiliary Steam (Figure 9.5.9-1) Auxiliary steam N D NNS N I T boiler Auxiliary steam N D NNS N NA T reboiler Auxiliary steam N D NNS N VIII T deaerator Condensate recovery N D NNS N VIII A/R tanks Condensate recovery N D NNS N MS A/R tank transfer pumps 6.2 Standby Diesel Generator Engine (Figures 9.5.5-1 and 9.5.6-1) Lube oil cooler Y C 3 Y-B III-3 D Keep-warm lube oil Y (Note 20) 3 Y-B (Note 20) D pump Main Lube oil strainer Y C 3 Y-B III-3 D (duplex) Fuel oil filter Y C 3 Y-B III-3 D Lube oil heater Y C 3 Y-B III-3 D Lube oil level control Y C 3 Y-B (Note 17) D tank Starting air compressor N NA NNS N MS D filter Diesel rocker lube oil Y (Note 18) NA Y-B MS D strainer Diesel oil separator Y (Note 18) NA Y-B MS D Motor driven rocker Y (Note 18) NA Y-B MS D pre-lube pump Starting air dryer N NA NNS N MS D pre-filter Starting air instrument N NA NNS N MS D distr. filter Lube oil suction Y NA NA Y-B MS D strainer Rev. 27

WOLF CREEK TABLE 3.2-1 (Sheet 15) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks Engine driven fuel oil Y (Note 18) NA Y-B MS D pump Engine driven Y (Note 18) NA Y-B MS D intercooler pump Engine driven jacket Y (Note 18) NA Y-B MS D water pump Engine driven lube Y (Note 18) NA Y-B MS D oil pump Engine driven rocker Y (Note 18) NA Y-B MS D lube pump Ejector Y (Note 18) NA Y-B MS D Rocker reservoir tank Y (Note 18) NA Y-B MS D Fuel rack supply air Y C 3 Y-B III-3 D tank Starting air pulsation N NA NNS N MS D dampener Lube oil filter Y C 3 Y-B III-3 D Starting air tanks Y C 3 Y-B III-3 D Jacket water heat Y C 3 Y-B III-3 D exchanger Jacket water ex- Y C 3 Y-B III-3 D pansion tank Keep-warm jacket Y C 3 Y-B III-3 D water pump Intake air filter Y (Note 18) NA Y-B MS D Intake air silencer Y (Note 18) NA Y-B MS D Exhaust silencer Y (Note 18) NA Y-B MS D Engine/generator Y NA NA Y-B MS D control panels Intercooler water Y C 3 Y-B III-3 D heat exchanger Interconnecting Y C 3 Y-B III-3 D piping Fuel oil strainer Y C 3 Y-B III-3 D Auxiliary lube oil Y C 3 Y-B III-3 D tank Jacket water Y C 3 Y-B III-3 D (keepwarm) heater Engine gauge panel Y NA NA Y-B MS D Starting air compressor N NA NNS N MS D Starting air dryer N NA NNS N MS D Standby diesel engine Y (Note 19) NA Y-B MS D Piping/valves Y C 3 Y-B III-3 D Piping/valves N D NNS N B31.1 D 6.3 Emergency Fuel Oil System (Figure 9.5.4-1) Emergency fuel oil Y C 3 Y-B III-3 O storage tank Emergency fuel oil Y C 3 Y-B III-3 O Class 1E transfer pump power supply Emergency fuel oil Y C 3 Y-B III-3 D day tank Emergency fuel oil Y C 3 Y-B III-3 D strainers Piping/valves Y C 3 Y-B III-3 D/O Piping/valves N D NNS N B31.1 D/O Rev. 7

WOLF CREEK TABLE 3.2-1 (Sheet 16) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks 6.4 Compressed Air (Figure 9.3-1) Instrument air com- N D NNS N NA T pressors Air receivers N D NNS N VIII T Emergency accumu- Y C 3 Y-B III-3 A lators Piping/valves Y C 3 Y-B III-3 A Piping/valves N D NNS N B31.1 A/B/C/D/F/O/R/T 6.5 Service Gases (Figure 9.3-9) N D NNS N NA 6.6 Fire Protection (Figure 9.5.1-1) Standpipes, headers, N NA NNS N NFPA A/B/C/D/ and valves F/O/R/T Sprinkler systems, N NA NA N NFPA/UL/ A/B/C/D/ halogenated extin- ANI/FM F/O/R/T guishing systems, hose racks, portable extinguishers Fire detection and N NA NA N NFPA/UL/ A/B/C/D/ alarm system ANI/FM F/O/R/T Main control room N NA NA N MS B fire protection system annunciator and control panel Fire pumps N NA NNS N NFPA I Non-Class 1E 1 motor driven, 1 diesel Piping/valves Y NA NNS N NFPA D3 6.7 Floor and Equipment Drainage System (Figure 9.3-5) General piping, N NA NA N B31.1 A/B/C/D/ pumps, and sumps F/R/T Auxiliary building Y C 3 Y-B III-3 A isolation valves 6.8 Nuclear Sampling System (Figure 9.3-23, 18.2-15) Nuclear sampling N D NNS N MS A/R panels Piping/valves Y B 2 Y-B III-2 A/C Piping/valves N NA NNS N B31.1 A/C 6.9 Process Sampling System (Figure 9.3-4) Process sampling N D NNS N MS T panels Rev. 11

WOLF CREEK TABLE 3.2-1 (Sheet 17) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks 7.0 HEATING, VENTILATING, AND AIR CONDITIONING 7.1 Control Building 7.1.1 Control Room Air Conditioning System (Figure 9.4-1) Control room air conditioning unit Unit Y NA 3 Y-B MS, NEMA (Motor) B Class 1E power supply Condenser Y C 3 Y-B III-3 (water A Side) VIII, Div 1, (refrigerant side) Control room filtration Y NA 3 Y-B ANSI A system absorber train Control room filtration fan Fan Y NA 3 Y-B MS A Motor Y NA 3 Y-B NEMA A Class 1E power supply Control room pressuri-zation system absorber train Unit Y NA 3 Y-B ANSI A Motor Y NA 3 Y-B UL A Class 1E power supply Control room pressuri-zation fan Fan Y NA 3 Y-B MS A Motor Y NA 3 Y-B NEMA A Class 1E power supply Ductwork/dampers Y NA 3 Y-B See Section A 9.4.1 7.1.2 Class 1E Electrical Equipment Air Conditioning System (Figure 9.4-1) Class 1E electric equipment Air conditioning System Unit Y NA 3 Y-B IEEE-323 B Class 1E power supply Condenser Y NA 3 Y-B III-3 (water B side), VIII Div 1, (refrigerant side) Ductwork/Dampers Y NA 3 Y-B MS, NEMA (Motor) (See Section B 9.4.1) Recirculation fan Y NA 3 Y-B IEEE-323, B Class 1E System NEMA (motor) power supply 7.1.3 Balance of Control Building HVAC Equipment (Figure 9.4-1) Ductwork/dampers Y NA 3 Y-B See Section B Control building 9.4.1 isolation Unit heaters & duct N NA NNS N UL B Non-class 1E heaters power supply Fans & fan motors N NA NNS N NEMA (Motor) B Non-class 1E MS (Fan) power supply Fan coil units N NA NNS N MS B Non-class 1E power supply Booster coils N NA NNS N MS B Non-class 1E power supply Rev. 32

WOLF CREEK TABLE 3.2-1 (Sheet 18) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks Supply air units N NA NNS N MS, NEMA B Non-class 1E power supply Cooling coils N NA NNS N MS B Non-class 1E power supply Ductwork/dampers N NA NNS N See Section B 9.4.1 7.2 Fuel Building (Figure 9.4-2) 7.2.1 Emergency Exhaust System Emergency exhaust fan Y NA 3 Y-B MS F Emergency exhaust fan Y NA 3 Y-B IEEE 323 F Class 1E motor power supply Emergency exhaust char- Y NA 3 Y-B R.G.1.52 F coal adsorber train Emergency exhaust elec- Y NA 3 Y-B IEEE 323 F Class 1E tric heater power supply Ductwork/dampers Y NA 3 Y-B See Section F 9.4.2 7.2.2 Pump Room Coolers Pump room cooler Unit Y NA 3 Y-B MS F Motor Y NA 3 Y-B NEMA F Class 1E power supply Coil Y C 3 Y-B III-3 F 7.2.3 Balance of Fuel Building HVAC Equipment Unit heaters N NA NNS N MS, F Non-class 1E UL power supply (Electrical only) Supply air units N NA NNS N MS F Non-class 1E power supply Heating coil units N NA NNS N MS F Non-class 1E power supply Cooling coils N NA NNS N MS F Non-class 1E power supply Ductwork/dampers Y NA 3 Y-B See Section F Fuel building 9.4.2 isolation Ductwork/dampers N NA NNS N See Section F 9.4.2 7.3 Auxiliary Building (Figure 9.4-3) Pump Room and Penetration Room Coolers Pump/penetration room cooler Unit Y NA 3 Y-B MS A Motor Y NA 3 Y-B IEEE-323 A Class 1E power supply Coil Y C 3 Y-B III-3 A Rev. 1

WOLF CREEK TABLE 3.2-1 (Sheet 19) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks 7.3.2 Balance of Auxiliary Building HVAC Equipment Fans & fan motors N NA NNS N MS (Fans) A Non-class 1E power supply Unit heaters & duct N NA NNS N MS, A Non-class 1E heaters UL (Electrical) power supply Filter adsorber units N NA NNS N ANSI A Non-class 1E power supply Supply air units N NA NNS N MS A Non-class 1E power supply Fan coil units N NA NNS N MS A Non-class 1E power supply Exhaust scrubbers N NA NNS N MS A Non-class 1E power supply Ductwork/dampers Y NA 3 Y-B See Section A Auxiliary building 9.4.3 isolation Ductwork/dampers N NA NNS N See Section A 9.4.3 7.4 Diesel Generator Building Ventilation System (Figure 9.4-7) Diesel generator building ventilation fan Fan Y NA 3 Y-B MS D Motor Y NA 3 Y-B IEEE-323 D Class 1E power supply 7.5 Auxiliary, Fuel, Radwaste, Turbine Buildings, Access Control Exhaust HVAC, and Containment Purge (Figures 9.4-1, 9.4-2, 9.4-3, 9.4-4, 9.4-5, 9.4-6) Exhaust fans N NA NNS N NA A Supply fan N NA NNS N NA A Filter units N NA NNS N NA A Recirculation units N NA NNS N NA A Unit heater N NA NNS N NA A 7.6 Essential Service Water Pump House HVAC (Figure 9.4-8) Unit heaters N NA NNS N MS E/O Non-class 1E power supply Essential service water pump house fan Fan Y NA 3 Y-B MS E Motor Y NA 3 Y-B IEEE 323 E Class 1E power supply Ductwork/dampers Y NA 3 Y-B See Section E/O 9.4.8 Rev. 12

WOLF CREEK TABLE 3.2-1 (Sheet 20) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks 7.7 Containment Purge System HVAC (Figure 9.4-6) Supply air units N NA NNS N MS A Non-class 1E power supply Fans & fan motors N NA NNS N NEMA (motors) A Non-class 1E MS (fans) power supply Filter adsorbers unit N NA NNS N ANSI A Ductwork/dampers Y NA 3 Y-B See Section A Auxiliary building 9.4.6 isolation adiati-monitor mounting Ductwork/dampers N NA NNS N See Section A/C 9.4.6 7.8 Miscellaneous Building HVAC (Figure 9.4-3) Fans & fan motors N NA NNS N NEMA (motors) A/C MS (fans) Supply air unit N NA NNS N MS A Unit heaters & N NA NNS N MS A/C/O/R duct heaters UL (Electrical only) Ductwork/dampers Y NA 3 Y-B See Section A Auxiliary building 9.4.3 isolation Ductwork/dampers N NA NNS N See Section A/C 9.4.3 7.9 ESW Vertical Loop Chase N NA NNS N See Section V Ductwork/Dampers 9.4.11 Unit heaters N NA NNS N MS UL (Electrical only) 8.0 CIVIL/ARCHITECTURAL 8.1 Structures and Buildings Reactor building Y NA 2 Y-B BC-TOP-5A, C III/MC AISC Refueling pool and Y NA NA Y-B ACI-318-71 C other internal RB AISC structures Control building Y NA NA Y-B ACI 318-71 B AISC Auxiliary building Y NA NA Y-B ACI 318-71 A AISC Fuel building Y NA NA Y-B ACI 318-71 F AISC Fuel storage pool Y NA NA Y-B ACI 318-71 F Radwaste building D NA NA N ACI 318-71 R AISC Solid radwaste N NA NA N NA O storage warehouse Turbine building N NA NA N ACI 318-71 T AISC UBC-1973 Mixed Waste N NA NA N NA Storage (Owens Corning) Station Blackout N NA NA N IBC - 2006 0 Diesel Generator ACI 318-05 Missile Barrier AISC Rev. 30

WOLF CREEK TABLE 3.2-1 (Sheet 21) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks Essential service Y NA NA Y-B ACI 318-71 O water system pumphouse AISC Essential service Y NA NA Y-B ACI 318-71 O water system electrical duct banks and manholes Essential service Y NA NA Y-B ACI 318-71 water system AISC O caissons Essential service Y NA NA Y ACI 318-71 0 water system access vaults Moveable tornado Y NA NA Y-B ACI 318-71 O/T missile barriers AISC Site drainage (13) N NA NA N NA O Essential service N NA NA N NA O water system discharge point Diesel generator Y NA NA Y-B ACI 318-71 D building AISC Supports and founda- Y NA NA Y-B ACI 318-71 A/B/C/D/ tions for all non- AISC F/1/O NSSS Category I equipment and tanks Refueling water storage Y B 2 Y-B III-2 O tank Access vault for emer- Y NA NA Y-B ACI 318-71 O gency fuel oil tank AISC ESW Vertical Loop Chase Y NA NA Y ACI 318-71-AISC V 8.2 Materials for Category I Structures Containment liner Y NA NA Y-B III - MC C Refer to Sections plate VIII 3.8.1 and 3.8.2 for additional information Containment personnel Y NA NA Y-B III - MC C and equipment hatches Watertight doors Y NA NA Y-B NA A Pipe whip restraints Y NA NA Y-B NA Missile resistant Y NA NA Y-B NA doors Pressure resistant Y NA NA Y-B NA doors Bullet resistant Y NA NA N NA doors Water stops N NA NA N NA C/A Pool liner plate N NA NA N NA C/F and gates Radiation shielding Y NA NA N NA doors Rev. 30

WOLF CREEK TABLE 3.2-1 (Sheet 22) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks 9.0 CONTROL AND INSTRUMENTATION See Note 14 (Table 7.1-1) BOP engineered safety Y NA NA Y-B IEEE 279 B features actuation system NSSS engineering Y NA NA Y-W3 IEEE 279 A/B/C safety features actu-ation and reactor protection system Reactor control N NA NA N CH-7 A/C system Postaccident con- Y NA NA Y-B CH-7 F/A/B tainment radiation monitors and safe-ty-related air-borne radiation monitors Excore neutron mon- N NA NA N-O CH-7 itoring system Excore neutron monitor Postaccident moni- Y NA NA Y-W3 CH-7 A/B/C/F toring system Main control board Y NA NA Y-B/W3 CH-7 B Safety-related auxi- Y NA NA Y-B/W3 CH-7 liary control panels Instrument piping, Y B 2 Y-B III-2 C/A tubing, fittings, and valves that are connected to qual-ity group Class A or B process systems (9) (10) Instrument piping, Y C 3 Y-B III-3 A/B/C/D/F tubing, fittings, and valves that are connected to safety Class 3 process systems (10) Instrument piping, N D NNS N B31.1 A/B/C/D/F/ tubing, fittings, I/O/R/T and valves that are connected to NNS process systems 10.0 ELECTRICAL POWER SYSTEMS 10.1 Class 1E Lower Medium Voltage System Metal-clad switchgear Y NA NA Y-B IEEE-308, C 4.16 kV 336 5 kV power cable Y NA NA Y-B IEEE-308, C/A/D/I 336 Large induction Y NA NA Y-B IEEE-308, A/I motors, 250 hp 336, NEMA and larger MG-1 Rev. 1

WOLF CREEK TABLE 3.2-1 (Sheet 23) CLASSIFICATIONS OF STRUCTURES, COMPONENTS, AND SYSTEMS (14) Quality Principal Group ANS Construction Seismic Classifi- Safety Quality Codes and Category I cation Class Assurance Standards Location System/Component (1) (2) (3) (4) (5) (6) Remarks 10.2 Class 1E Low Voltage System Load center unit Y NA NA Y-B IEEE-308, C/A/I substations 336 Motor control Y NA NA Y-B IEEE-308, C/A/D/I centers 336 600 Volt power and Y NA NA Y-B IEEE-308, A/C/D/F/ control cable 336 I/R Integral and frac- Y NA NA Y-B IEEE-308, A/C/D/F/ tional hp in- 336, 344 I/R duction motors NEMA MG-1 600 Volt fire-resistive Y NA NA N IEEE-344 A/B/D/E/F power and control I/O/R/T/U cable 10.3 Class 1E 125 V DC System Batteries and battery Y NA NA Y-B IEEE-308, C charger 336 DC distribution Y NA NA Y-B IEEE-308, C panels 336 Emergency lighting Y NA NA Y-B MS C dc 10.4 Class 1E Instrument AC Power Vital ac power Y NA NA Y-B IEEE-308, C supply 336 120 V ac vital Y NA NA Y-B IEEE-308, C panels 336 600 V instrument Y NA NA Y-B IEEE-308, A/C/D/ cable 336 F/I 10.5 Reactor Building Cable Y B 2 Y-B IEEE-317, 336 A/C Penetrations 10.6 Conduit Supports and Y NA NA Y-B ASTM All Tray Supports 10.7 Raceway Installation Y NA NA Y-B IEEE-336 All 10.8 Load Shedding and Y NA NA Y-B IEEE-308, C Emergency Load 336 Sequencing 10.9 Auxiliary Relay Racks Y NA NA Y-B ICEA, NEMA A/C IEEE-336 10.10 Transformers Essential service Y NA NA Y-B IEEE-308 I water Regulating Y NA NA Y-B IEEE-308 C 10.11 Status Indicating Y NA NA Y-B/W3 IEEE-308, C Systems 336 10.12 Local Control Stations Y NA NA Y-B IEEE-308, A/D/F 336 Rev. 24

WOLF CREEK NOTES TO TABLE 3.2-1 (1) Y - Component is functionally and structurally designed and constructed to meet seismic Category I requirements, as defined in Regulatory Guide 1.29. S - Category I for structural integrity only. N - Component is non-Category I. Component is seismically designed and constructed if position C.2 of Regulatory Guide 1.29 applies per Table 3.2-3. D - Designed and constructed to seismic requirements given in Regulatory Guide 1.143. (2) A, B, C, D, D(A) - Quality group classification as defined in Regulatory Guide 1.26. NA - Not applicable to safety classification. Design requirements for components and piping associated with the Quality Group D(A) portions of this system which contain radioactive fluid are augmented by Note 1 of Table 3.2-2. (3) 1, 2, 3, NNS - Safety classifications as defined in ANSI N18.2. Except for the deviation described in section 3.2.3. NA - Not applicable to safety classification. (4) Quality Assurance Program All components with Y indicate that the component is subject to utility Quality Assurance Program during plant operation. Y-B Component was subject to the Bechtel Q-listed Quality Assurance Program during design and construction. Y-U Component was subject to the utility Q-listed Quality Assurance Program during design and construction. Y-W1 Component was subject to "Quality Control System Requirements," Westinghouse QCS-1 during design and construction. Y-W2 Component was subject to "Quality Requirements for Manufacture of Nuclear Plant Equipment," Westinghouse QCS-2 during design and construction. Y-W3 Component was subject to the quality assurance program of one of the Westinghouse manufacturing divisions during design and construction. N Component was subject to the requirements of applicable codes and standards and the manufacturer's standard quality assurance program during design and construction. Y-A Component was subject to the quality assurance program of ABB-Combustion Engineering Nuclear Services during design and construction. (5) The principal construction codes and standards are identified as: I: ASME Boiler and Pressure Vessel Code, Section I III and 1, 2, 3, MC,NG: ASME Boiler and Pressure Vessel Code, Section III, Division 1, Class 1, 2, 3, MC, or NG VIII: ASME Boiler and Pressure Vessel Code, Section VIII, Division 1 Rev. 19

WOLF CREEK NOTES TO TABLE 3.2-1 (Sheet 2) B31.1 ANSI B31.1, Code for Power Piping TEMA C, R Tubular Exchanger Manufacturers Association, Class C or Class R IEEE-279: Institute of Electrical and Electronics Engineers, Criteria for Protection Systems for Nuclear Power Generating Stations - 1971 IEEE-308: Institute of Electrical and Electronics Engineers, Standard Criteria for Class IE Power Systems for Nuclear Power Generating Stations - 1974 IEEE-317: Institute of Electrical and Electronics Engineers, Standard for Electric Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations - 1976, 1983 IEEE-323: Institute of Electrical and Electronics Engineers, Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations - 1974 IEEE-334: Institute of Electrical and Electronics Engineers, Standard for Type Tests of Continuous Duty Class IE Motors for Nuclear Power Generating Stations - 1974 IEEE-344: Institute of Electrical and Electronics Engineers, Recommended Practices for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations - 1975, 1987 IEEE-336: Institute of Electrical and Electronics Engineers, Installation, Inspection, and Testing Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Stations - 1971 IEEE-383: Institute of Electrical and Electronics Engineers, Standard for Type Test of Class IE Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations - 1974, 1983 NFPA: National Fire Protection Association ANI: American Nuclear Insurers ARI: Air Conditioning and Refrigeration Institute ACI 318-71: American Concrete Institute, Building Code Requirements for Reinforced Concrete UBC-1973: Uniform Building Code (state and/or local building codes may be substituted where they supersede UBC-1973) ICEA: Insulated Cable Engineers Association ASTM: American Society for Testing and Materials ANSI: American National Standards Institute NEC: National Electric Code Rev. 12

WOLF CREEK NOTES TO TABLE 3.2-1 (Sheet 3) AISC: American Institute of Steel Construction, Specifications for the Design, Fabrication, and Erection of Structural Steel for Buildings, 7th Edition, adopted February 12, 1969, and Supplement Numbers 1, 2, and 3 BC-TOP-5-A: Prestressed Concrete Nuclear Reactor Containment Structures, Revision 3 NEMA: National Electrical Manufacturers Association UL: Underwriters' Laboratories, Inc. FM: Factory Mutual NA: Design requirements specified by designer with appropriate consideration of the intended service and operating conditions API 650: American Petroleum Institute, Welded Steel Tanks for Oil Storage - Atmospheric Tanks MS: Manufacturer's Standard CH-7: Refer to Chapter 7 (6) Location: A. Auxiliary building B. Control building C. Reactor building D. Diesel generator building E. Essential service water pumphouse F. Fuel building I. Intake structure O. Outdoors onsite R. Radwaste building T. Turbine building U. Fire pumphouse V. ESW Vertical Loop Chase Rev. 30

WOLF CREEK NOTES TO TABLE 3.2-1 (Sheet 4) (7) Table indicates the required code based on its safety-related importance as dictated by service and functional requirements and by the consequences of their failure. Note that the actual equipment may be supplied to a higher principal construction code than required. (8) Access for inspection and test required. However, no formal quality program approval is required. (9) A 3/8-inch restriction is provided for all instrument connections to Quality Group A liquid piping to change the instrument piping Quality Group classification from A to B. A 3/4 instrument connection is used on Quality Group A piping connected to pressurizer steam space to change the instrument piping quality group classification from A to B as described in section 5.2.1.1. (10) Requirements of ASME Boiler and Pressure Vessel Code Section III are met, except that the instrument sensing line between the instrument shutoff valve and the instrument is not hydrostatically tested. The instrument sensing line between the process tap and the instrument shutoff valve will be hydrostatically tested in accordance with the Code. (11) Pressure boundary is Safety Class 1; heaters are electrically NNS. (12) Safety-related instruments and controls are described in USAR Sections 7.1 to 7.6. (13) The site drainage system consists of many components including roof drains, site storm drains, culverts and ditches for which no credit is taken in component roof loading or site flooding analyses. However, major modifications to Category I building roofs and the plant railroad spur, roads, and graded surfaces, which are in Zones 1 and 2 of Figure 2.4-3, will be evaluated to ensure that such modifications will not result in flooding of Category I structures. (14) Almost all of the systems listed in Table 3.2-1 include instrumentation and control (I&C) devices. However, it is not the intent of Table 3.2-1 to address this type of detail. The addition of all instrumentation and control devices could triple the size of the listing, adding unnecessary detail that would tend to confuse instead of enhance the understanding of the table. The electrical equipment qualification list (Appendix A of the NUREG-0588 Submittal) provides a detailed lising of all Class 1E powered I&C devices. These devices are included in each system that they serve (e.g. EG-FT-0108 is a flow transmitter in the component cooling water system [EG]). The I&C devices can be divided into two categories, NSSS and BOP supplied. Each type can be identified in the fourth column of Appendix A. The BOP supplied devices that are purchased by the Bechtel I&C Group have a specification number that begins with the letter "J" (e.g., J-301 for EG-FT-0108). The NSSS supplied devices are identified in the fourth column by the respective Westinghouse number EQDP number (e.g., ESE-4). Classification of power supplies, motors, piping and valves, ductwork and dampers and associated supports, hangers and restraints are not delineated in Table 3.2-1 because of the extensive listing required. Their classification is consistent with the boundaries shown on the Piping and Instrumentation Diagrams. (15) Vents, drains, test connections, ect., only. (16) Deleted (17) The Lube Oil Level Control Tank was fabricated by the manufacturer under a Quality Assurance Program per ASME and is constructed of ASME material. The tank has a rectangular configuration not covered by ASME, but the tank is essentially atmospheric and not pressure retaining. (18) Component is supplied with the standard diesel engine as an integral part of the engine or whose design and reliability have been proven through years of previous diesel engine service. The standards used in design, manufacture, and inspection are the manufacturer's standards, developed by the manufacturer's manufacturing and testing experience. The design is considered equivalent to ASME Section III Class 3 requirements with regard to functional operability and inservice reliability. Rev. 19

WOLF CREEK NOTES TO TABLE 3.2-1 (Sheet 5) (19) The diesel engine and the engine-mounted and separately skid-mounted portions of the auxiliary support systems piping and components normally furnished with the diesel generator package are designed to the guidelines of the Diesel Engine Manufacturers Association (DEMA) standards. The diesel engine and its mounted auxiliary support systems piping and components also conform to the requirements of IEEE Standard 387-1977, "Standard Criteria for Diesel-Generator Units Applied as Standby Power Supplies for Nuclear Power Generating Stations, " which endorses the DEMA Standards, and Regulatory Guide 1.9. The diesel engine and its auxiliary support systems meet the quality control requirements of 10 CFR 50, Appendix B. (20) The component design and reliability has been proven through years of previous service. The standards used in design, manufacture and inspection are the manufacturer's standards developed by manufacturing and testing experience. The design meets seismic category I requirements and is equivalent to the originally supplied ASME Section III component. (21) Equipment is no longer in service. Rev. 27

WOLF CREEK TABLE 3.2-2 CODE REQUIREMENTS FOR COMPONENTS AND QUALITY GROUPS QUALITY GROUPS Component A B C D(1) Pressure vessels ASME B & PV Code ASME B & PV Code ASME B & PV Code ASME B & PV Code Section III, Section III, Section III, Section VIII, Div. 1 Class 1 Class 2 Class 3 or 2, or Section I Reactor containment -- ASME B & PV Code -- -- pressure vessels Section III, (steel) Class MC Pumps ASME B & PV Code ASME B & PV Code ASME B & PV Code Manufacturer's Standard2 Section III, Section III, Section III, Class 1 Class 2 Class 3 Valves ASME B & PV Code ASME B & PV Code ASME B & PV Code ANSI B31.1.0 Power Section III, Section III, Section III, Piping Class 1 Class 2 Class 3 Piping ASME B & PV Code ASME B & PV Code ASME B & PV Code ANSI B31.1.0 Power Section III, Section III, Section III, Piping Class 1 Class 2 Class 3 0-15 psig -- ASME B & PV Code ASME B & PV Code API-620 or storage tanks Section III, Section III, or equivalent Class 2 Class 3 Atmospheric storage -- ASME B & PV Code ASME B & PV Code API-650 or API-620 Section III, Section III, or equivalent Class 2 Class 3 (Section III for stainless steel)2 Heat exchangers ASME B & PV Code ASME B & PV Code ASME B & PV Code ASME B & PV Code Section III, Section III, Section III, Section VIII, Class 1 and TEMA "R" Class 2 and TEMA "R" Class 3 and TEMA "R" Div. 1 and TEMA "C"

1. Construction of portions of systems identified by as D(A) Note 2 of Table 3.2-1 use the following augmenting criteria, to the maximum extent possible:
a. Welded construction. Flanged jointed or suitable rapid disconnect fittings are used only where dictated by maintenance or operational requirements.
b. Process lines 2-1/2 inches nominal pipe size or above are butt welded (no backing rings are used on resin or evaporator bottom lines). Process lines 2 inches or smaller are socket welded. Instrumentation lines are not considered process lines, and screwed connections may be used. Manual valves are butt welded, except where flanges are dictated.

Rev. 0

WOLF CREEK TABLE 3.2-2 (Sheet 2)

c. Material used for construction of pressure-retaining components, primarily carbon steel or austenitic stainless steel, complies with applicable sections of the codes and standards for quality group D. Malleable wrought or cast iron materials and plastic piping are not used. Manufacturer's material certification of compliance is required.
d. All welding constituting the pressure boundary of pressure-retaining components is performed by qualified welders employing qualified welding procedures per ASME Code Section IX.
e. High quality non-metallic hoses are used to connect vendor supplied liquid radwaste processing mobile skids to the liquid radwaste system.
2. No ASME code stamp is required.

Rev. 19

WOLF CREEK TABLE 3.2-3 DESIGN COMPARISON TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.29 REVISION 3, DATED SEPTEMBER 1978, TITLED SEISMIC DESIGN CLASSIFICATION This comparison is presented for the BOP portion of the design. Refer to Appendix 3A for the Westinghouse discussion. Regulatory Guide 1.29 Position WCGS

1. The following structures, 1. All plant items which are systems, and components necessary to cope with a of a nuclear power plant, LOCA, secondary side break including their foundations inside containment, or to and supports, are designated shut the plant down safely as Seismic Category I and following an SSE in the should be designed to with- absence of a LOCA are stand the effects of the designed for the SSE.

SSE and remain functional. There are, however, some The pertinent quality assur- plant items not required ance requirements of Appen- following an SSE but which dix B to 10 CFR Part 50 are required to cope with should be applied to all other natural phenomena. activities affecting the For example, a plant item safety-related functions which is required to of these structures, systems, function only during or and components. following a tornado in order to achieve a safe shutdown must be consid-ered to perform a safety function, but the design of the item for an SSE is unnecessary. Further, there are plant items which serve to mitigate the consequence of certain in-plant occurrences (other than LOCA) which are not considered to occur simultaneously with an SSE. Examples of the latter occurrences are fuel handling or spent fuel cask accidents and loss of control room habitability. Thus, cer-tain items not listed in Regulatory Guide 1.29 Rev. 0

WOLF CREEK TABLE 3.2-3 (Sheet 2) Regulatory Guide 1.29 Position WCGS are considered to serve a safety function and sub-ject to quality assurance coverage in accordance with 10 CFR Part 50, Appendix B. Table 3.2-1 itemizes those safety-related structures, systems, and components which are designed for a safe shutdown earthquake.

a. The reactor coolant a. Complies.

pressure boundary.

b. The reactor core and b. Complies.

reactor vessel internals. 1

c. Systems or portions of c. Complies. See Item 2 systems that are required below.

for (1) emergency core cooling, (2) postaccident containment heat removal, or (3) postaccident con-tainment atmosphere cleanup (e.g., hydrogen removal system). 1

d. Systems or portions of d. Complies. See Item 2 systems that are required below.

for (1) reactor shutdown, (2) residual heat removal, or (3) cooling the spent fuel storage pool.

e. Those portions of the steam e. Not applicable to systems of boiling water WCGS.

reactors. . .

f. Those portions of the f. Complies with the steam and feedwater exception that the systems of pressurized words "or remote water reactors extending manual" are considered from and including the to be inserted after secondary side of steam the word "automatic."

This option is Rev. 0

WOLF CREEK TABLE 3.2-3 (Sheet 3) Regulatory Guide 1.29 Position WCGS generators up to and in- included to avoid an cluding the outermost unnecessary complica-containment isolation tion (leading to de-valves, and connected creased plant reli-piping of 2-1/2 inches ability) in the line or larger nominal pipe which is not normally size up to and including provided with auto-the first valve (in- matic closing valves. cluding a safety or relief valve) that is Note that valves in either normally closed lines emanating from or capable of auto- the steam generator matic closure during are for secondary side all modes of normal isolation, not con-reactor operation. tainment isolation.

g. Cooling water, component g. Complies.

cooling, and auxiliary feedwater systems1 or portions of these systems, including the intake structures, that are required for (1) emergency core cooling, (2) postaccident con-tainment heat removal, (3) postaccident con-tainment atmosphere cleanup, (4) residual heat removal from the reactor, or (5) cooling the spent fuel storage pool.

h. Cooling water and seal h. Complies.

water systems1 or por-tions of these systems that are required for functioning of reactor coolant system components important to safety, such as reactor coolant pumps. Rev. 29

WOLF CREEK TABLE 3.2-3 (Sheet 4) Regulatory Guide 1.29 Position WCGS

i. Systems1 or portions of i. Complies.

systems that are re-quired to supply fuel for emergency equip-ment.

j. All electric and j Complies.

mechanical devices and circuitry between the process and the input terminals of the actuator systems involved in gen-erating signals that initiate protective action. 1

k. Systems or portions k. Complies.

of systems that are re-quired for (1) moni-toring of systems impor-tant to safety and (2) actuation of systems im-portant to safety.

l. The spent fuel storage l. Complies, with the pool structure, including clarification that the the fuel racks. pool liner plate and gates are not desig-nated as seismic Category I. (See Sec-tion 9.1.2)
m. The reactivity control m. Complies.

systems, e.g., control rods, control rod drives, and boron injection system.

n. The control room, including n. Complies.

its associated equipment needed to maintain the control room within safe habitability limits for personnel and safe environ-mental limits for vital equipment. Rev. 0

WOLF CREEK TABLE 3.2-3 (Sheet 5) Regulatory Guide 1.29 Position WCGS

o. Primary and secondary o. Complies. Note that reactor containment. the WCGS design does not incorporate a secondary containment.
p. Systems1, other than p. Complies, except that WCGS radioactive waste manage- is using meteorology as ment systems, not covered recommended by Regulatory by items 1.a through 1.o Guide 1.183, Alternative above that contain radio- Radiological Source Terms active material and whose for Evaluating Design-Basis postulated failure would Accidents at Nuclear Power result in conservatively Reactors. Note that calculated potential Regulatory Guide 1.143 offsite doses (using provides guidance on meteorology as prescribed radioactive waste management by Regulatory Guide 1.4, systems and structural "Assumptions Used for seismic design. Table 3.2-1 Evaluating the Potential indicates those systems Radiological Consequences for which the D (Augmented) of a Loss-of-Coolant design criteria are applied.

Accident for Pressurized The dividing line value of Water Reactors") that are 0.5 rem is inappropriate for more than 0.5 rem to the the types of failures which whole body or its equiva- the guide addresses. lent to any part of the body. Quality Group D or D (Augmented) is applied to such systems unless their failure would result in offsite doses approaching the guide values of 10 CFR Part 100 which were the 10 CFR Part 20 limits at that time. Since then, the value was changed from 0.5 rem whole body to 0.1 rem TEDE.

q. The Class IE electric q. Complies; however in certain systems, including the cases Class lE conduits are auxiliary systems for supported from non-Category the onsite electric power I seismic walls. Although supplies, that provide not Category I, these the emergency electric reinforced block walls are power needed for func- analyzed for SSE loads in tioning of plant features accordance with position Z included in items 1.a and are subject to the QA through 1.p above. program described in Position 4.

Rev. 34

WOLF CREEK TABLE 3.2-3 (Sheet 6) Regulatory Guide 1.29 Position WCGS

2. Those portions of struc- 2. Complies, including the tures, systems, or com- following clarification:

ponents whose continued Those portions of struc-function is not required tures, systems, or compo-but whose failure could nents whose continued reduce the functioning function is not required of any plant feature but whose failure could included in items 1.a reduce the functioning through 1.q above to an of any plant feature to unacceptable safety level an unacceptable level or could result in inca- included in items 1.a pacitating injury to through 1.q above, which occupants of the control is required for safe shut-room should be designed down of the plant, and constructed so that following a DBA, are the SSE would not cause designed and constructed such failure. so that the SSE will not cause such a failure. Although LOCA or major natural phenomenon or DBE is not postulated to occur at the time of an SSE, in addition to those safety-related items required for post-accident safe shutdown all systems required to mitigate the consequences of LOCAs and secondary side breaks inside containment are protected from nonseismic items. Since tornadoes are not postulated to occur with an SSE, the contents of the boric acid tank room are not protected from adverse seismic interactions. This system is only relied upon following a tornado induced loss of the RWST. The system is designed in accordance with position 1.m above. Rev. 14

WOLF CREEK TABLE 3.2-3 (Sheet 7) Regulatory Guide 1.29 Position WCGS For these items, a quality program which includes identification, design, and installation is used to meet the intent of Paragraph C.4.

3. Seismic Category I design 3. Seismic Category I design requirements should extend analysis requirements are to the first seismic extended to the first restraint beyond the seismic restraint beyond defined boundaries. the defined boundaries.

Those portions of struc- Since seismic analysis of tures, systems, or com- a piping system requires ponents that form inter- division of the system faces between Seismic into discrete segments Category I features should terminated by fixed be designed to Seismic points, this means that Category I requirements. the seismic analysis can-not be terminated at a seismic restraint, but is extended to include the interface piping out to the first point in the system which can be treated as an anchor to the plant structure. Inasmuch as the seismic analysis is based upon minimum material proper-ties and documented system hydrostatic and perfor-mance tests are made, the nonsafety-related portion of the system (including supports) past the inter-face boundary valve is not seismic Category I and will not be Q-Listed. Rev. 0

WOLF CREEK TABLE 3.2-3 (Sheet 8) Regulatory Guide 1.29 Position WCGS For these items, a quality program which includes identification, design, and installation is used to meet the intent of Paragraph C.4.

4. The pertinent quality 4. The items covered under assurance requirements Regulatory Positions 2 and of Appendix B to 10 CFR 3 above are not considered Part 50 should be applied to be seismic Category I to all activities affect- and are not considered to ing the safety-related be Q-listed.

functions of those portions of structures, systems, and For these items, a quality components covered under assurance program is Regulatory Positions 2 and applied which is commen-3 above. surate with the safety consideration involved. The following practices adequately meet the in-tended requirements:

a. Design and design control for such items are carried out in the same manner as that for items which directly serve as a safety function. This includes the performance of appro-priate design reviews.
b. Design includes considera-tion of loads imposed

______________________ during an SSE. 1The system boundary includes those portions of the system c. Field work is performed required to accomplish the under the direction of specified safety function and experienced field con-connected piping up to and struction superintendents including the first valve and is inspected under a (including a safety or relief QA program. valve) that is either normally closed or capable of automatic closure when the safety function is required. Rev. 0

WOLF CREEK TABLE 3.2-4 DESIGN COMPARISON TO REGULATORY GUIDE 1.26 REVISION 3, DATED FEBRUARY 1976, TITLED "QUALITY GROUP CLASSIFICATIONS AND STANDARDS FOR WATER-, STEAM-, AND RADIOACTIVE-WASTE CONTAINING COMPONENTS OF NUCLEAR POWER PLANTS" Quality group classifications and standards for plant systems and components meet the intent of Regulatory Guide 1.26. However, certain clarifications and specific exceptions to the guide are necessary. In Paragraphs A and B of the regulatory guide, there is a different usage of the term "important to safety" than that used elsewhere in the regulations and regulatory guides. The guide includes components which fall into quality group D under the definition of "important to safety," which implies that a quality assurance program in accordance with 10 CFR 50, Appendix B, should be applied. These quality assurance requirements are neither applied to quality group D components nor are they applied to quality group D (augmented) components. The definition of the term "important to safety," insofar as quality assurance (Appendix B) is concerned, is considered to be that which appears in the introduction of Regulatory Guide 1.29. Regulatory Guide 1.26 establishes the quality group classification for steam and water containing components. However, the guidance is also used to establish the quality group classification of other systems. These systems are designed, fabricated, erected, and tested to quality standards commensurate with the safety function to be performed. Table 3.2-1 itemizes the classification for these systems and components. Section 3.9.3 discusses design for components not covered by the ASME Code. Below is a comparison of the WCGS design with each of the regulatory guide positions. Regulatory Guide 1.26 Position WCGS

1. The group B quality standards 1. Complies.

given in Table 1 of the guide should be applied to water and steam-containing pressure vessels, heat exchangers (other than turbines and con-densers), storage tanks, piping, pumps, and valves that are either part of the reactor Rev. 0

WOLF CREEK TABLE 3.2-4 (Sheet 2) Regulatory Guide 1.26 Position WCGS coolant pressure boundary de-fined in Section 50.2(v) but excluded from the requirements of Section 50.55a pursuant to footnote 2 of that section or not part of the reactor coolant pressure boundary but part of:

a. Systems or portions of a. Complies.

systems important to safety that are designed for (1) emergency core cooling, (2) postaccident contain-ment heat removal, or (3) postaccident fission prod-uct removal.

b. Systems or portions of b. Systems which perform the systems important to functions of reactor safety that are designed shutdown and residual for (1) reactor shutdown heat removal are placed or (2) residual heat in Quality Group B, as removal. indicated by the guide.

This is limited to in clude only the minimum of those systems which must function in the perfor-mance of an orderly safe shutdown and maintenance of the plant in the safe (hot) shutdown condi-tion. Those systems which may be used in the performance of a normal cold shutdown (such as the reactor coolant pumps) or incidentally in the removal of residual heat from the reactor [i.e., heat removal is not their prime function (such as portions of the CVCS)] are not placed in Quality Group B. Rev. 0

WOLF CREEK TABLE 3.2-4 (Sheet 3) Regulatory Guide 1.26 Position WCGS

c. Those portions of the c. Not applicable to WCGS.

steam systems of boiling water reactors...

d. Those portions of the d. Specific exceptions steam and feedwater systems taken to placing portions of pressurized water of main steam and feed-reactors extending from water lines in quality and including the sec- group B are as follows:

ondary side of steam generators up to and in- (1) The words "or remote cluding the outermost con- manual" are consid-tainment isolation valves ered to be inserted and connected piping up to after the word "auto-and including the first matic." This option valve (including a safety is included to avoid or relief valve) that is an unnecessary comp-either normally closed or lication (leading to capable of automatic clo- decreased plant reli-sure during all modes of ability) in lines normal reactor operation. which would not nor-mally be provided with automatic closing valves. (2) Note that valves in lines emanating from the steam generator are for secondary side isolation, not containment iso-lation.

e. Systems or portions of e. Complies.

systems that are connected to the reactor coolant pressure boundary and are not capable of being iso-lated from the boundary during all modes of normal reactor operation by two valves, each of which is either normally closed or capable of automatic clo-sure. Rev. 0

WOLF CREEK TABLE 3.2-4 (Sheet 4) Regulatory Guide 1.26 Position WCGS

2. The group C quality standards 2. Complies as noted below.

given in Table 1 of the guide should be applied to water-, steam-, and radioactive-waste containing pressure vessels, heat exchangers (other than turbines and condensers), storage tanks, piping, pumps, and valves not part of the reactor coolant pressure boundary or included in quality group B but part of:

a. Cooling water and auxiliary a. Complies.

feedwater systems or portions of these systems important to safety that are designed for (1) emergency core cooling, (2) postaccident containment heat removal, (3) postaccident contain-ment atmosphere cleanup, or (4) residual heat removal from the reactor and from the spent fuel storage pool (including primary and sec-ondary cooling systems). Portions of these systems that are required for their safety functions and that (1) do not oper-ate during any mode of normal reactor operation and (2) cannot be tested adequately should be classified as group B.

b. Cooling water and seal water b. Complies.

systems or portions of these systems important to safety that are designed for func-tioning of components and systems important to safety, such as reactor coolant pumps, diesels, and control room. Rev. 0

WOLF CREEK TABLE 3.2-4 (Sheet 5) Regulatory Guide 1.26 Position WCGS

c. Systems or portions of c. Complies.

systems that are connected to the reactor coolant pressure boundary and are capable of being isolated from that boundary during all modes of normal reactor operation by two valves, each of which is either normally closed or capable of automatic closure.

d. Systems, other than d. Complies, except that WCGS is radioactive waste using meteorology as management systems, not recommended by Regulatory Guide covered by items 2.a 1.183, Alternative through 2.c above that Radiological Source Terms for contain radioactive Evaluating Design-Basis material and whose Addicents at Nuclear Power postulated failure would Reactors. Note that result in conservatively Regulatory Guide 1.143 provides calculated potential guidance on radioactive waste offsite doses (using management system design. Table meteorology as recommended 3.2-1 indicates those systems by Regula tory Guide 1.4, to which the D (Augmented)
  "Assumptions Used for                 design criteria are applied.

Evaluating the Potential The dividing line value of 0.5 Radiological Consequences rem is inappropriate for the of a Loss- of-Coolant types of failures which the Accident for Pressurized guide addresses. Quality Group Water Reactors") that D [or D (Augmented)] is applied exceed 0.5 rem to the whole to such systems unless their body or its equivalent to failure would result in offsite any part of the body. For doses approaching the guideline those systems located in values of 10 CFR Part 100 which Seismic Category I were the 10 CFR Part 20 limits structures, only single at that time. Since then, the component failures need be value was changed from 0.5 rem assumed. However, no credit whole body to 0.1 rem TEDE. for automatic isolation Radwaste systems, except for from other components in portions of the steam generator the system or for treatment blowdown system located in the of released material should turbine building, are located be taken unless the within a seismically designed isolation or treatment building as permitted by capability is designed Regulatory Guide 1.143, and only single component failures are considered. Rev. 34

WOLF CREEK TABLE 3.2-4 (Sheet 6) Regulatory Guide 1.26 Position WCGS to the appropriate seismic quality group standards and can withstand loss of offsite power and a single failure of an active component.

3. The group D quality standards 3. Complies. In addition, given in Table 1 of this guide quality standards for D should be applied to water- (Augmented) systems are and steam-containing compo- consistent with Regulatory nents not part of the reactor Guide 1.143.

coolant pressure boundary or included in quality groups B or C but part of systems or portions of systems that con-tain or may contain radioactive material. Rev. 0

WOLF CREEK TABLE 3.2-5 DESIGN COMPARISON TO REGULATORY GUIDE 1.143, FOR COMMENTS DATED JULY, 1978, TITLED "DESIGN GUIDANCE FOR RADIOACTIVE WASTE MANAGEMENT SYSTEMS, STRUCTURES, AND COMPONENTS INSTALLED IN LIGHT-WATER-COOLED NUCLEAR POWER PLANTS" Design requirements of this regulatory guide are applied to components, systems, and structures which fall under the D (augmented) classification established by Regulatory Guide 1.26, position C.2.d and Regulatory Guide 1.29, Position C.1.p. The design requirements of this guide are therefore applied to the following systems or portions of systems:

a. Purification portion of CVCS
b. Boron thermal regeneration portion of CVCS
c. Boron recycle system
d. Liquid radwaste system
e. Gaseous radwaste system
f. Secondary liquid waste evaporator
g. Steam generator blowdown system
h. Solid radwaste system The radioactive waste management systems are considered to begin at the interface valve(s) in each line from other systems provided for collecting wastes that may contain radioactive materials and to include related instrumentation and control systems. The radioactive waste management systems terminate at the point of controlled discharge to the environment, at the point of recycle back to storage for reuse in the reactor, or at the point of storage of packaged solid wastes prior to shipment offsite to a licensed burial ground. The steam generator blowdown system begins at, but does not include, the outermost isolation valve on the blowdown line, and terminates at the point of controlled discharge to the environment, at the point of interface with other liquid waste systems, or at the point of recycle back to the secondary systems.

Rev. 9

WOLF CREEK TABLE 3.2-5 (Sheet 2) Regulatory 1.143 Position WCGS

1. Systems Handling Radioactive Materials in Liquids 1.1 The liquid radwaste treatment system, 1.1 Applies to the systems including the steam generator blow-down system identified above.

downstream of the second containment isolation valve, should meet the following criteria: 1.1.1 These systems should be designed and 1.1.1 Complies. See Table tested to requirements set forth in the codes 3.2-2. and standards listed in Table 1, supplemented by the provisions in 1.1.2 and in regulatory position 4 of this guide. 1.1.2 Materials for pressure-retaining 1.1.2 Complies. Carbon steel, components should conform to the requirements stainless steel, or similar of the specifications for materials listed in materials compatible with the Section II of the ASME Boiler and Pressure chemical, physical, and Vessel Code, except that malleable, wrought, radioactive environment are or cast iron materials and plastic pipe should used for pressure retaining not be used. Materials should be compatible components. The use of with the chemical, physical and radioactive malleable, wrought, or cast environment of specific applications. iron materials or plastic pipe Manufacturers' material certificates of is not allowed. High quality compliance with material specifications, such non-metallic hoses are used to as those contained in the codes referenced in connect vendor supplied mobile Table 1, may be provided in lieu of certified skids to the liquid radwaste material test reports. system. Material certificates of compliance or certified material test reports are required for the materials purchased. Polyethylene or polypropylene tanks may be used in cases where the corrosive constituents make this material a superior choice. Reinforced hoses may be used at interface points between mobile process equipment and piping. 1.1.3 Foundations and walls of structures 1.1.3 Complies. See Section that house the liquid radwaste system should 3.8.6.4. be designed to the seismic criteria described in regulatory position 5 of this guide, to a height sufficient to contain the maximum liquid inventory expected to be in the building. Rev. 19

WOLF CREEK TABLE 3.2-5 (Sheet 3) Regulatory 1.143 Position WCGS 1.1.4 Equipment and components used to 1.1.4 Complies. Liquid collect, process, and store liquid radioactive contained sources are not waste need not be designed to the seismic seismically designed. criteria given in regulatory position 5 of this guide. 1.2 All tanks located outside the reactor 1.2 See response to 1.2.1 containment and containing radioactive through 1.2.5. materials in liquids should be designed to prevent uncontrolled releases of radioactive materials due to spillage (in buildings or from outdoor tanks). The following design features should be included for tanks that may contain radioactive materials: 1.2.1 All tanks inside and outside the plant, 1.2.1 Complies. See Table including the condensate storage tanks, should 11.2-2. have provisions to monitor liquid levels. Potential overflow conditions should actuate alarms both locally and in the control room. 1.2.2 All tank overflows and drains and 1.2.2 Complies. See Table sample lines should be routed to the liquid 11.2-2. radwaste treatment system1. 1.2.3 Indoor tanks should have curbs or 1.2.3 Complies. See Table elevated thresholds with floor drains routed 11.2-2. to the liquid radwaste treatment system1. 1.2.4 The design should include provisions to 1.2.4 Complies. See Sections prevent leakage from entering unmonitored 9.4 and 11.3. systems and ductwork in the area. 1.2.5 Outdoor tanks should have a dike or 1.2.5 Complies. No outdoor retention pond capable of preventing runoff in tanks fall under the the event of a tank overflow and should have D(augmented) classification, provisions for sampling collected liquids and and no dikes are provided. routing them to the liquid radwaste treatment system.

2. Gaseous Radwaste Systems 2.1 The gaseous radwaste treatment system 2 2.1 See response to 2.1.1 should meet the following criteria: through 2.1.3.

2.1.1 The systems should be designed and 2.1.1 Complies. See Table tested to requirements set forth in the codes 3.2-2. and standards listed in Table 1 supplemented by the provisions noted in 2.1.2 and in regulatory position 4 of this guide. Rev. 9

WOLF CREEK TABLE 3.2-5 (Sheet 4) Regulatory 1.143 Position WCGS 2.1.2 Materials for pressure-retaining 2.1.2 Complies. Carbon steel, components should conform to the requirements stainless steel, or similar of the specifications for materials listed in materials compatible with the Section II of the ASME Boiler and Pressure chemical, physical, and Vessel Code, except that malleable, wrought, radioactive environment are or cast iron materials and plastic pipe should used for pressure-retaining not be used. Materials should be compatible components. The use of with the chemical, physical, and radioactive malleable, wrought, or cast environment of specific applications. iron materials or plastic pipe Manufacturers' material certificates of is not allowed. Material compliance with material specifications, such certificates of compliance or as those contained in the codes referenced in certified material test reports Table 1, may be provided in lieu of certified are required for the material. materials test reports. 2.1.3 Those portions of the gaseous radwaste 2.1.3 Complies as indicated in treatment system that are intended to store or response to position 5. The delay the release of gaseous radioactive gaseous radwaste system waste, including portions of structures operates above 1.5 atmospheres. housing these systems, should be designed to the seismic design criteria given in regulatory position 5 of this guide. For the systems that normally operate at pressures above 1.5 atmospheres (absolute), these criteria should apply to isolation valves, equipment, interconnecting piping, and components located between the upstream and downstream valves used to isolate these components from the rest of the system (e.g., waste gas storage tanks in the PWR) and to the building housing this equipment. For systems that operate near ambient pressure and retain gases on charcoal adsorbers, these criteria should apply to the tank support elements (e.g., charcoal delay tanks in a BWR) and the building housing the tanks.

3. Solid Radwaste System 3.1 The solid radwaste system consists of 3.1 See response to 3.1.1 slurry waste collection and settling tanks, through 3.1.4.

spent resin storage tanks, phase separators, and components and subsystems used to solidify radwastes prior to offsite shipment. The solid radwaste handling and treatment system should meet the following criteria: Rev. 9

WOLF CREEK TABLE 3.2-5 (Sheet 5) Regulatory Guide 1.143 Position WCGS 3.1.1 The system should be designed and 3.1.1 Complies. See Table tested to the requirements set forth in the 3.2-2. Fiberglass reinforced codes and standards listed in Table 1 plastic tanks, in accordance supplemented by the provisions noted in 3.1.2 with appropriate articles of and in regulatory position 4 of the guide. Section 10, ASME BPV Code, are used to dewater wastes prior to storage and offsite shipment. 3.1.2 Materials for pressure-retaining 3.1.2 Complies. Carbon steel, components should conform to the requirements stainless steel, or other of the specifications for materials listed in similar materials compatible Section II of the ASME Boiler and Pressure with the chemical, physical, Vessel Code, except that malleable, wrought, and radioactive environment are or cast iron materials and plastic pipe should used for pressure-retaining not be used. Materials should be compatible components. The use of with the chemical, physical and radioactive malleable, wrought, or cast environment of specific applications. iron material or plastic pipe Manufacturers' material certificates of is not allowed. Material compliance with material specifications, such certificates of compliance or as those contained in the codes referenced in certified material test reports Table 1, may be provided in lieu of certified are required for the material materials test reports. purchased. 3.1.3 Foundations and adjacent walls of 3.1.3 Complies, as described structures that house the solid radwaste in Section 3.8.6.4. system should be designed to the seismic criteria given in regulatory position 5 of this guide to a height sufficient to contain the maximum liquid inventory expected to be in the building. 3.1.4 Equipment and components used to 3.1.4 Complies. Contained collect, process, or store solid radwastes sources are not seismically need not be designed to seismic criteria given designed. in regulatory position 5 of this guide. 4.0 Additional Design, Construction, and Testing Criteria In addition to the requirements inherent in the codes and standards listed in Table 1, the following criteria, as a minimum, should be implemented for components and systems considered in this guide: 4.1 The quality assurance provisions 4.1 Complies, as described in described in regulatory position 6 of this position 6. guide should be applied. Rev. 9

WOLF CREEK TABLE 3.2-5 (Sheet 6) Regulatory 1.143 Position WCGS 4.2 Process piping systems include the first 4.2 Complies. root valve on sample and instrument lines. Pressure-retaining components of process systems should use welded construction to the maximum practicable extent. Flanged joints or suitable rapid disconnect fittings should be used only where maintenance or operational requirements clearly indicate that such construction is preferable. Screwed connections in which threads provide the only seal should not be used, except for instrumentation connections where welded connections are not suitable. Process lines should not be less than 3/4 inch (nominal I.D.). Screwed connections backed up by seal welding, mechanical joints, or socket welding may be used on lines 3/4 inch or larger but less than 2-1/2 inches (nominal I.D.). For lines 2-1/2 inches above, pipe welds should be of the butt-joint type. Non-consumable backing rings should not be used in lines carrying resins or other particulate material. All welding constituting the pressure boundary of pressure-retaining components should be performed in accordance with ASME Boiler and Pressure Vessel Code Section IX. Rev. 19

WOLF CREEK TABLE 3.2-5 (Sheet 7) Regulatory 1.143 Position WCGS 4.3 Piping systems should be 4.3 Complies except that WCGS has hydrostatically tested in their replaced the hydrostatic test with entirety, except at atmospheric equally acceptable alternative tests. tank connections where no Since the NRC has approved the use of isolation valves exist. Pressure Code Case N-416-1 at WCGS (with testing should be performed on as additional conditions which are also large a portion of the in-place included in the D-Augmented Program) to systems as practicable. Testing eliminate hydro test of safety related of piping systems should be piping replacements, modifications and performed in accordance with repairs, WCGS has granted similar relief applicable ASME or ANSI codes, for ANSI B31.1 piping. The ASME Section but in no case at less than 75 XI Code case N-416-1 illustrates the psig. The test pressure should opinion of ASME code experts that be held for a minimum of 30 hydrostatic tests provide little minutes with no leakage additional confidence of system integrity indicated. for replacements, modifications and repairs. The ASME Section III design requirements and acceptance criteria for Class 3 piping (as imposed by ASME Section XI) are similar to B31.1 design requirements and acceptance criteria. Although Section XI has lower pressure test requirements, with the imposition of additional surface examinations, the conclusion is still valid in reviewing ANSI hydrostatic test requirements. The performance of a hydrostatic test typically results in difficulty with existing valve leakage, particularly at existing boundary valves, which can preclude meeting the hydrostatic test requirements without rebuilding these boundary valves and retesting. The purpose of the hydro test has nothing to do with boundary valves, thus hydro testing results in additional work and delays not associated with the reason for performing the hydro test. Additionally, the performance of a hydrostatic test can result in damage to piping systems and components whereas alternative in-service leak test and visual exams and non-intrusive PT or MT, does not. Rev. 19

WOLF CREEK TABLE 3.2-5 (Sheet 8) Regulatory 1.143 Position WCGS 4.4 Testing provisions should be incorporated 4.4 Complies. The systems are to enable periodic evaluation of the in intermittent or continuous operability and required functional use, which demonstrates the performance of active components of the systems' performance and system. structural and leaktight integrity.

5. Seismic Design for Radwaste Management Systems and Structures Housing Radwaste Management Systems.

5.1 Gaseous Radwaste Management System3. 5.1 See 5.1.1 through 5.1.3. 5.1.1 For the evaluation of the gaseous 5.1.1 The gaseous radwaste radwaste system described in regulatory system is seismically analyzed, position 2.1.3, a simplified seismic analysis considering a single degree of procedure to determine seismic loads may be freedom and the floor response used. The simplified procedure consists of spectra discussed in position considering the system as a single-degree-of- 5.2. freedom system and picking up a seismic response value from applicable floor response spectra, after the fundamental frequency of the system is determined. The floor response spectra should be obtained analytically (regulatory position 5.2) from the application of the Regulatory Guide 1.60 design response spectra normalized to the maximum ground acceleration for the operating basis earthquake (OBE), as established in the application, at the foundation of the building housing the gaseous radwaste system. More detailed guidance can be found in Regulatory Guide 1.122, "Development of Floor Design Response spectra for Seismic Design of Floor-Supported Equipment or Components." 5.1.2 The allowable stresses to be used for 5.1.2 Complies. steel system support elements should be those given in "Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings," adopted in February 1969. The one-third allowable stress increase provisions for combinations involving earthquake loads, indicated in Section 1.5.6 of the specification, should be included. For design of concrete structures, use of ACI 349-76 as endorsed in Regulatory Guide 1.142, "Safety-Related Concrete Structures for Nuclear Power Plants (Other Than Reactor Vessels and Containments)," is acceptable. Rev. 19

WOLF CREEK TABLE 3.2-5 (Sheet 9) Regulatory Guide 1.143 Position WCGS 5.1.3 The construction and inspection 5.1.3 Complies. requirements for the support elements should comply with those stipulated in AISC or ACI Codes as appropriate. 5.2 Buildings Housing Radwaste Systems 5.2 Complies. Section 3.8.6.4 addresses the requirements of 5.2.1 through 5.2.6. 5.2.1 Input motion at the foundation of the building housing the radwaste systems should be defined. This motion should be defined by normalizing the Regulatory Guide 1.60 spectra to the maximum ground acceleration selected for the plant OBE. A simplified analysis should be performed to determine appropriate seismic loads and floor response spectra pertinent to the location of the system, i.e., an analysis of the building by a several-degrees-of-freedom mathematical model and the use of an approximate method to generate the floor response spectra for radwaste systems and the seismic loads for the buildings. No time history analysis is required. 5.2.2 The simplified method for determining seismic loads for the building consists of (a) calculating the first several modal frequencies and participation factors for the building, (b) determining modal seismic loads using regulatory position 5.2.1 input spectra, and (c) combining modal seismic loads in one of the ways described in Regulatory Guide 1.92, "Combining Modal Responses and Spatial Components in Seismic Response Analysis." 5.2.3 With regard to generation of floor response spectra for radwaste systems, simplified methods that give approximate floor response spectra without need for performing a time history analysis may be used. 5.2.4 The load factors and load combinations to be used for the building should be those given in ACI 349-76 as endorsed in Regulatory Guide 1.142. The allowable stresses for steel components should be those given in the AISC Manual. (See regulatory position 5.1.2.) Rev. 19

WOLF CREEK TABLE 3.2-5 (Sheet 10) Regulatory Guide 1.143 Position WCGS 5.2.5 The construction and inspection requirements for the building elements should comply with those stipulated in the AISC or ACI Code, as appropriate. 5.2.6 The foundation media of structures housing the radwaste systems should be selected and designed to prevent liquefaction from the effects of the maximum ground acceleration selected for the plant OBE. 5.3 In lieu of the criteria - and procedures 5.3 The criteria and defined above, optional shield structures procedures of 5.2 are used. constructed around and supporting the radwaste systems may be erected to protect the radwaste systems from effects of housing structural failure. If this option is adopted, the procedures described in regulatory position 5.2 need only be applied to the shield structures while treating the rest of the housing structures as nonseismic Category I.

6. Quality Assurance for Radwaste Management 6. The quality assurance Systems program for D (augmented) components and systems was provided in Chapter 17.0 of the PSAR.

Since the impact of these systems on safety is limited, a quality assurance program corresponding to the full extent of Appendix B to 10 CFR Part 50 is not required. However, to ensure that systems will perform their intended function, a quality assurance program sufficient to ensure that all design, construction, and testing provisions are met should be established and documented. The following quality assurance program is acceptable to the NRC staff. It is reprinted by permission of the American Nuclear Society from ANSI N199-1976, "Liquid Radioactive Waste Processing System for Pressurized Water Reactor Plants." Rev. 19

WOLF CREEK TABLE 3.2-5 (Sheet 11) Regulatory Guide 1.143 Position WCGS "4.2.3 Quality Control. The design, procurement, fabrication, and construction activities shall conform to the quality control provisions of the codes and standards specified herein. In addition, or where not covered by the referenced codes and standards, the following quality control features shall be established. "4.2.3.1 System Designer and Procurer (1) Design and Procurement Document Control--Design and procurement documents shall be independently verified for conformance to the requirements of this standard by individual(s) within the design organization who are not the originators of the document. Changes to these documents shall be verified or controlled to maintain conformance to this standard. 

     "(2) Control of Purchased Material, Equipment and Services--Measures to ensure that suppliers of materials, equipment, and construction services are capable of supplying these items to the quality specified in the procurement documents shall be established. This may be done by an evaluation or a survey of the suppliers' products and facilities.
     "(3) Instructions shall be provided in procurement documents to control the handling, storage, shipping, and preservation of material and equipment to prevent damage, deterioration, or reduction of cleanness.

Rev. 19

WOLF CREEK TABLE 3.2-5 (Sheet 12) Regulatory Guide 1.143 Position WCGS "4.2.3.2 System Constructor (1) Inspection. In addition to required code inspections, a program for inspection of activities affecting quality shall be established and executed by, or for, the organization performing the activity to verify conformance with the documented instructions, procedures, and drawings for accomplishing the activity. This shall include the visual inspection of components prior to installation for conformance with procurement documents and the visual inspection of items and systems following installation, cleanness, and passivation(where applied). 

     "(2) Inspection, Test, and Operating Status. Measures should be established to provide for the identification of items which have satisfactorily passed required inspections and tests.
     "(3) Identification and Corrective Action for Items of Nonconformance. Measures should be established to identify items of nonconformance with regard to the requirements of the procurement documents or applicable codes and standards and to identify the action taken to correct such items."

In Section 4.2.3.2(3), "items of nonconformance" should be interpreted to include failures, malfunctions, deficiencies, deviations, and defective material and equipment. Sufficient records should be maintained to furnish evidence that the measures identified above are being implemented. The records should include results of reviews and inspections and should be identifiable and retrievable. Rev. 19

WOLF CREEK TABLE 3.2-5 (Sheet 13) Regulatory Guide 1.143 Position WCGS NOTES: 1 Retention by an intermediate sump or drain tank designed for handling radioactive materials and having provisions for routing to the liquid radwaste system is acceptable. 2 For a BWR, this includes the system provided for treatment of normal offgas releases from the main condenser vacuum system beginning at the point of discharge from the condenser air removal equipment; for a PWR, this includes the system provided for the treatment of gases stripped from the primary coolant. 3 For those systems that require seismic capabilities, as indicated in Regulatory Position 2.1.3. Rev. 19

WOLF CREEK 3.3 WIND AND TORNADO LOADINGS All seismic Category I structures which are required for post-accident safe shutdown, contain equipment required for post-accident safe shutdown, are required to protect reactor coolant system integrity, or which protect stored fuel assemblies are designed to withstand the effects of a tornado and the most severe wind phenomena encountered at the site (see Section 2.3) (GDC-2). These structures are identified in Table 3.3-1. For ISFSI system structures and components, refer to the NUHOMS EOS System UFSAR Sections 2.3 and A.2.3 for details of wind and tornado design bases, which differs from the WCGS design bases, where such differences are reconciled in the WCGS 10 CFR 72.212 Evaluation Report. A tabulation of systems and components and their location by room number, except for the RWST, needed for a post-accident safe shutdown and to ensure the integrity of the reactor coolant pressure boundary is provided in Table 7.4-6. All of the components and systems identified in Table 7.4-6, which include those requiring tornado protection, are housed within the protective structures identified in Table 3.5-2. All of those structures are designed to provide tornado protection. The protective structure requirements for the RWST are discussed in Section 6.3.2.2. Since there are no systems or components within the remaining plant structures whose failure could lead to significant offsite radiological consequences, those buildings have not been designed to provide tornado protection for systems contained therein. The structures, systems, and components identified in Appendix A to Regulatory Guide 1.117 have been provided with tornado protection. BC-TOP-3-A (Ref. 1) defines tornado and extreme wind loadings and criteria, and furnishes data, formulae, and procedures for determining maximum wind loading on structures or parts of structures. 3.3.1 WIND LOADINGS 3.3.1.1 Design Wind Velocity The design wind velocity for all seismic Category I structures is 100 mph at 30 feet above ground for a 100-year recurrence interval. The bases for the wind velocity selection and supporting data and wind histories are contained in Section 2.3 and in Section 2.0 of BC-TOP-3-A. The design wind velocity envelops all of the site wind conditions. As referenced in BC-TOP-3-A, ANSI A58.1 (Ref. 2) is used as the basis for determining the vertical velocity distribution and gust factors. The wind pressure values used are those tabulated in Section 6 of ANSI A58.1 for exposure "C," which is flat, open country. Table 5 of ANSI A58.1 is used to determine the effective velocity pressures on buildings and structures. Table 6 of ANSI 3.3-1 Rev. 35

WOLF CREEK A58.1 is used to determine the effective velocity pressures on parts and portions of buildings and structures. A basic wind speed of 100 mph is used, and the tables take into account the effects of vertical velocity distribution and gust factors. 3.3.1.2 Determination of Applied Forces The procedures used to translate the wind velocity into applied forces on the structures are contained in ANSI A58.1 and Sections 2.0 and 4.0 of BC-TOP-3-A. These procedures include the applicable effects of wind force distribution and shape coefficients. For seismic Category I structures which are designed for tornado loading, the applied forces due to wind are calculated to determine if they are less severe than the applied forces due to tornado loadings. The applied tornado-force magnitude and distribution are determined, as described in Section 3.3.2.2 below. Appropriate load combinations, stress levels, and load factors discussed in Section 3.8 are considered in determining the governing loads. 3.3.2 TORNADO LOADINGS Tornado loadings for structural analysis are obtained in accordance with BC-TOP-3-A. Compliance with Regulatory Guide 1.76 is discussed in Appendix 3A. 3.3.2.1 Applicable Design Parameters Tornado loads are not assumed to be coincident with any accident condition or earthquake. Tornado characteristics are established in accordance with Table I of Regulatory Guide 1.76 for tornado intensity region I. A maximum windspeed of 360 mph, which consists of a maximum rotational speed of 290 mph at a radius of 150 feet combined with a maximum translational speed of 70 mph, is used. In order to maximize transit time of the tornado across exposed plant features, a minimum translational speed of 5 mph is used. An atmospheric pressure drop of 3.0 psi, at a linear rate of 2.0 psi per second, is also used. Tornado-generated missiles are discussed in Section 3.5.1.4. 3.3-2 Rev. 0

WOLF CREEK 3.3.2.2 Determination of Forces on Structures The procedures used to transform the tornado loadings into an effective pressure on exposed surfaces of structures are outlined in Section 3.5 of BC-TOP-3-A. The effects of shape coefficients and pressure distribution are included in these procedures. All seismic Category I structures are designed to prevent venting, with the exception of the main steam tunnel (Area 5 of the auxiliary building above El. 2026) and the fuel building. The main steam tunnel and the fuel building are vented to the atmosphere with the exterior walls and roofs designed to resist the full pressure differential (3.0 psi) due to the design basis tornado. The interior walls and slabs are designed to resist the differential pressures between compartments that occur as a result of venting the structure. The methods employed to determine the differential pressures are found in Section 3.5.2 of BC-TOP-3-A. The procedures used to transform the tornado-generated missile loadings into effective loads are discussed in Section 3.5.3. Tornado wind velocity pressure effects, atmospheric pressure change effects, and missile impact effects are combined in accordance with Section 3.4 of BC-TOP-3-A. These combined effects constitute the total tornado effect (Wt ), which is then combined with other loads as specified in Section 3.8. 3.3.2.3 Effect of Failure of Structure or Components Not Designed For Tornado Loads Non-Category I structures are not designed for tornado loads. Non-Category I structures adjacent to seismic Category I structures include the turbine building and communications corridor. The structural framing of these buildings is designed to preclude gross collapse upon safety-related structures or components under loads imposed by the design basis tornado. Other non-Category I structures are located so that their collapse would not endanger safety-related structures or components. 3.

3.3 REFERENCES

1. Tornado and Extreme Wind Design Criteria for Nuclear Power Plants, Bechtel Power Corporation, BC-TOP-3-A, San Francisco, California, Revision 3, August, 1974.
2. American National Standards Institute (ANSI), Building Code Requirements for Minimum Design Loads in Buildings and Other Structures, A58.1-1972.

3.3-3 Rev. 0

WOLF CREEK TABLE 3.3-1 TORNADO-RESISTANT BUILDINGS AND ENCLOSURES Reactor building Control building Fuel building Auxiliary building Diesel generator building Diesel fuel oil storage tank access vaults Turbine building (for structural framing integrity only) Communications corridor (for structural framing integrity only) ESWS Pumphouse ESWS Electrical Manholes ESWS Access Vaults Station Blackout Diesel Generator Missile Barrier (for structure only, unoccupied by personnel) ESW Vertical Loop Chase Rev. 30

WOLF CREEK 3.4 WATER LEVEL (FLOOD) DESIGN The criteria used to establish the design basis flood levels for the powerblock comply with Regulatory Guides 1.59 and 1.102, to the extent described in Appendix 3A. For ISFSI system structures and components, refer to the NUHOMS EOS System UFSAR Sections 2.3 and A.2.3 for details of flood design bases. 3.4.1 FLOOD PROTECTION 3.4.1.1 Flood Protection Measures for Seismic Category I Structures 3.4.1.1.1 External Flood Protection All seismic Category I structures and the systems they house are designed to withstand the effects of natural phenomena, such as flooding and groundwater level (GDC-2). Flood elevations, including the probable maximum flood (PMF) and the maximum groundwater elevations used in the design of powerblock seismic Category I structures for buoyancy and hydrostatic pressure, are shown in Tables 1.2-1 and 3.4-1 and are discussed in Section 2.4. The seismic Category I essential service water system (ESWS) pumphouse and Access Vault AV6 are subject to the forces resulting from the probable maximum precipitation (PMP) coincident with wave activity in the portion of the cooling lake containing the ultimate heat sink (UHS). The resulting design flood elevation in the UHS under this condition is described in Sections 2.4.3.6 and 2.4.10. The ESWS pumphouse extends into the UHS intake channel and takes suction from it through penetrations below both the design normal lake elevation and flood elevation (see Figures 3.8-131, 3.8-132, and 3.8-133). The only safety-related components in the ESWS pumphouse that are considered covered with flood water in the design flood are the casings, shafts, and impellers of the ESWS pumps and components of the traveling water screens, which are capable of normal function while surrounded by the design flood. The ESWS Access Vault AV6 is located below site grade near the shoreline of the cooling lake. The access vault is protected from wave run-up by revetment and sheet pile wall to the east, south and west. Powerblock seismic Category I structures are not protected above grade for flooding because there are no above-grade floods at the structure locations. Safety-related systems located below grade are protected from groundwater inleakage by a combination of a waterproofing system for the structures and other features such as the location of safety-related systems in watertight compartments, sump pumps, alarms and other water level indications and administrative controls. The waterproofing system will minimize groundwater inleakage. Should groundwater inleakage occur, the design features and administrative controls would protect the safety related systems. Refer to Section 1.2 for figures of systems below grade. In addition, an interior floor drainage system, as described in Section 9.3.3, is provided within the structures. 3.4-1 Rev. 35

WOLF CREEK Although not serving a safety-related function, additional waterproofing is provided below grade by means of waterstops and waterproofing materials to minimize inleakage. Waterstops are provided at expansion and construction joints and electrical duct bank penetraions located below grade. An auxiliary waterproofing system is installed on the vertical exterior surfaces of walls below grade of all powerblock seismic Category I buildings, except the reactor building. The minimum 5-foot thickness of base mats provides adequate waterproofing of floor areas with the exception of the ESW Vertical Loop Chase which has waterproofing below the structures base mat. The minimum 7-foot-thick vertical wall and internal steel liner plate provide sufficient waterproofing of the reactor cavity and instrumentation tunnel. There is no functional requirement for waterproofing of the tendon gallery. Below grade penetrations are provided with waterproof seals to minimize groundwater intrusion. Typical waterproofing details are shown in Figure 3.4-1. 3.4.1.1.2 Internal Flooding Protection All safety-related equipment rooms located below grade are protected from back-flooding by the remote location of waste-processing components in the radwaste building. The floor and equipment drains in powerblock seismic Category I buildings drain to sumps in the lowest level of the building in which they are located. These sumps are pumped to the floor drain tank or the waste hold-up tank located in the radwaste building. Should these tanks rupture or leak, flow into safety-related areas will not occur since the tanks are located below the radwaste building flood level. Equipment and floor drains below the 7-foot flood level of the auxiliary building drain to sumps within the same compartment or are provided with drain caps. Several water tight areas have been established in the auxiliary building to provide protection of all safety-related equipment. Drainage areas and protection of the safety-related equipment in this area is described in Section 9.3.3 and Figure 9.3-6. As described in Sections 9.3.3 and 11.2, the drainage and liquid radwaste systems are designed to preclude backflow from occurring in the safety-related equipment in the auxiliary building. Appendix 3B provides an evaluation of the effect of postulated flooding generated within the plant. 3.4.1.2 Ground Water Elevations The design basis for ground water for buoyancy and subsurface hydrostatic loadings on all site-related, seismic Category I structures is full hydrostatic pressure at all depths below elevation 1999.5 feet, refer to Section 2.4.13.5. Seismic Category I structures are protected below grade by waterproofing, waterstops at construction joints and electrical duct bank penetrations, and boot seals at pipe penetrations, where necessary. With the exception of sump pumps installed in ESW electrical manholes, no permanent dewatering system is provided to relieve the effects of ground water. Table 3.4-2 describes the site-related, seismic Category I structures that house safety-related equipment and identifies exterior penetrations that are below the design basis ground water elevation. 3.4-2 Rev. 32

WOLF CREEK 3.4.1.3 Permanent Dewatering Systems The permanent dewatering systems are not a safety-related system, and their failure does not compromise any safety-related system or prevent a safe shutdown of the reactor. Also, no permanent dewatering sytem, with the exception of the sump pumps installed in the ESW electrical manholes, is required. 3.4.2 ANALYSIS PROCEDURES Natural phenomena, such as flood current, wind wave, or hurricane (tsunamis cannot occur at WCGS), that are associated with dynamic water forces are not applicable to the powerblock seismic Category I structures, since the grades for these structures are located above the probable maximum flood elevations. 3.4.2.1 Design Basis Flood for the ESWS Pumphouse The design of the walls of the ESWS pumphouse, for the static and dynamic effects of the postulated wind-wave activity shown in Table 3.4-3, is in accordance with the load factors and loading combinations stated in Section 3.8 for live loads not coincident with earthquake or tornado loads. The load from the maximum postulated static water elevation in the UHS is applied as a hydrostatic force, and the dynamic effect of the nonbreaking waves in the UHS is converted to an equivalent hydrostatic force to the elevation shown in Table 3.4-3. Refer to Section 2.4 for a description of the bases for the data in Table 3.4-3. 3.4.2.2 Design Basis Groundwater Structures as a whole and component parts are designed for the hydrostatic forces from the maximum groundwater level, in accordance with the load factors and loading combinations stated in Section 3.8. 3.4-3 Rev. 32

WOLF CREEK TABLE 3.4-1 PMF, GROUNDWATER, REFERENCE, AND ACTUAL PLANT ELEVATIONS Probable Max. Design Ground- Reference Actual Flood Level water Elevation Plant Grade Plant Grade Structure ft. - msl ft. ft. ft. - msl Reactor building 1095.00 1999.50 1099.50 1099.50 Control building 1095.00 1999.50 1099.50 1099.50 Fuel building 1095.00 1999.50 1099.50 1099.50 Auxiliary building 1095.00 1999.50 1099.50 1099.50 Diesel generator 1095.55(1) (2) 1999.50 1099.50 1099.50 building ESWS intake pumphouse 1100.20(1) (3) 1999.50 1099.50 1099.50 ESWS discharge 1100.20(1) (3) 1999.50 1099.50 1099.50 point ESWS Access Vault AV6 1100.20(1) 1999.50 1099.50 1099.50 ESW Vertical Loop Chase 1095.00 1999.50 1099.50 1099.50 NOTES:

1) Maximum flooding with wave runup
2) At powerblock
3) At ESWS intake pumphouse - face of vertical wall Rev. 29

WOLF CREEK TABLE 3.4-2 SITE-RELATED, CATEGORY I STRUCTURES WITH PENETRATIONS BELOW THE GROUND WATER ELEVATIONS Structure and Areas Below Ground Safety-Related Compo-Figure Refer- Water Level and nents in Areas Below ences Their Penetrations Ground Water Level Inleakage Protection Discussion ESWS 1. Sumps for ESWS 1. Casings, 1. None 1. Sumps are Pumphouse pumps and trav- shafts, normally eling water impellers full of Figures 3.8-131, screens and 11.17 of the ESWS water 3.8-132, 3.8-133 feet x 6 feet pumps, and penetrations for traveling water entry from water screen UHS intake channel components

2. Pits for ESWS and 2. ESWS pipes, 2. Waterproofing on Any inleakage would chemical addition check valves, exterior faces of be visible and pipes, their and electri- pit walls and slabs, accessible for sleeved pene- cal cables waterstops at con- removal from the trations and ESWS struction joints and pumphouse operating electrical duct electrical duct bank floor.

banks and their penetrations, and penetrations boot seals between through the walls the ESWS pipes and of the pits their sleeves ESWS electrical All manholes which Electrical Waterstops at Cables are Manholes have numerous pene- cable construction joints specified for use trations for elec- and penetrations in wet conditions. Figure 3.8-140 trical duct banks Administrative controls monitor cable tray supports. Sump pumps are installed in ESW electrical manholes. ESWS access Access pits which ESWS pipes Waterproofing on exter- Stilling well with Vaults have ESWS piping ior faces of walls and manhole provided Figure 3.8-143 penetrations thru walls bottom of basemat; for water level waterstops at construction indication outside joints, waterproofing and of vault and waterstops at pipe administrative penetrations. controls to monitor inleakage. ESW Vertical Loop Substructure where ESWS ESWS Pipes Waterproofing on exterior faces Any inleakage would be Chase piping penetrates thru of walls and slabs; waterstops at routed to the sump Figures 3.8-103 Sh Control Building wall construction joints and water 2, 3.8-104 Sh 2, proofing and waterstops at pipe 3.8-118 penetrations. Rev. 29

WOLF CREEK TABLE 3.4-3 WIND-GENERATED WAVE DATA FOR THE ESWS PUMPHOUSE Case 1 Case 2 Wind speed 40 mph 90 mph Significant wave height, Hs 3.0 ft 5.7 ft Maximum wave height, Hm 5.0 ft 9.52 ft Static water surface elevation 1995 ft 1988 ft Dynamic (equivalent static) water surface elevation at the face of the ESWS pumphouse for period of wave motion 2000.2 ft 1998.5 ft Max. period of wave motion 3.32 secs 4.15 secs Rev. 14

WOLF CREEK WATER~OOfiNG (WHERE NOTED) 

                                                                                                                                                         /  WATERSTOP WATERPROOFING                                                                                                                                     RCXJGHENEO SURFACE (WHERE NOTED)

TYPICAL PENETRATION IN WALLS BASE SLAB NOTE ALL WATERPROOF:NG WILL BE CARRI[D TO GRADE ELEVAT!ON. FILLER MATERIAL SHALL BE TYPICAL WATERPROOFING APPLICATION Sl LICONE RUBBER FOAM, FROTH ED IN PLACE (LOW DENSITY)~ \ 

                                             \

SLEEVE ADHESIVE BOND Rev. 0 BLOCKOUT FILLED AFTER INSTALLATION WOLF CREEK 

                                                                                          -WATERPROOFING                     UPDATED SAFETY ANALYSIS REPORT FIGURE 3.4-1
                .A.::c8~.. :.A.Tc
                ~;._; ;._r\t"r-.; i..-

WHEN FLEXIBIL!TY OF PIPING IS DctrdCTD/\Tlr'\~1 

                                           ; ; _ ; , i . . - ; t\,...... i 1\.../l'i 1~1 11'1
                                                                                           \AJAtiC' YYML..L..V REQUIRED TYPICAL WATERPROOFING DETAILS

WOLF CREEK 3.5 MISSILE PROTECTION Adequate protection is provided to ensure that those portions of the essential structures, systems, or components whose failure would result in the failure of the integrity of the reactor coolant system, reduce the functioning to an unacceptable level of any plant feature required for a post-accident safe shutdown, or lead to offsite radiological consequences are designed and constructed so as not to fail or cause such a failure in the event of a postulated credible missile impact. The recommendations of Regulatory Guides 1.13 and 1.115 as they pertain to internally and externally generated missiles are met. The response to Regulatory Guide 1.14 and Regulatory Guide 1.27 in regard to missiles is included in Appendix 3A. Appendix 3B provides an evaluation of the effect of postulated missiles generated within the plant. The following sections provide the bases for the selection of the missiles, protection requirements for external missiles, and details of the barrier design. For ISFSI system structures and components, refer to the NUHOMS EOS System UFSAR Sections 2.3 and A.2.3 for details of tornado missile design bases, which differs from the WCGS design bases, where such differences are reconciled in the WCGS 10 CFR 72.212 Evaluation Report. 3.5.1 MISSILE SELECTION AND DESCRIPTIONS There are four general sources from which missiles are postulated. These are:

a. Rotating component failure
b. Pressurized component failure
c. Tornadoes
d. Missiles associated with activities in the proximity of the site The locations where the missiles may be generated are categorized as follows:
a. Internally generated missiles
b. Turbine missiles
c. Externally generated (outside the plant building) missiles during tornadoes 3.5-1 Rev. 35

WOLF CREEK 3.5.1.1 Internally Generated Missiles (Outside Containment) There are two general sources of postulated missiles within the plant:

a. Rotating component failures
b. Pressurized component failure 3.5.1.1.1 Rotating Component Failure Missiles Missiles generated by postulated failures of rotating components, their source and characteristics, and missile protection provided are discussed in Appendix 3B.

Missile selection is based on the following conditions:

a. All rotating components which are operated during normal operating plant conditions are capable of becoming missiles.
b. The energy in a rotating part associated with component failure is assumed to occur at 120-percent overspeed.
c. The energy of the missile is sufficient to perforate the protective housing.

3.5.1.1.2 Pressurized Component Failure Missiles Missiles generated by postulated failures of pressurized components, their source and characteristics, and missile protection provided are discussed in Appendix 3B. The bases for selection are:

a. Pressurized components in systems whose service temperature exceeds 200qF or whose design pressure exceeds 275 psig are evaluated as to their potential for becoming a missile.
b. Temperature or other detectors installed in high energy piping are evaluated as potential missiles if failure of a single circumferential weld could cause their ejection.
c. Welded dead-end flanges are evaluated as potential missiles if the failure of a single circumferential weld could cause their ejection.

3.5-2 Rev. 0

WOLF CREEK

d. Valves of ANSI 900-psig rating and above, constructed in accordance with Section III of the ASME Boiler and Pressure Vessel Code, are pressure seal, bonnet-type valves. For pressure seal bonnet valves, bonnets are prevented from becoming missiles by the retaining ring, which would have to fail in shear, and by the yoke, which would capture the bonnet or reduce bonnet energy.

Because of the highly conservative design of the retaining ring of these valves, bonnet ejection is highly improbable, and hence bonnets are not considered credible missiles.

e. Most valves of ANSI 600-psig rating and below are valves with bolted bonnets. Valve bonnets are prevented from becoming missiles by limiting stresses in the bonnet-to-body bolting material by rules set forth in the ASME Boiler and Pressure Vessel Code, Section III, and by designing flanges in accordance with applicable code requirements. Even if bolt failure were to occur, the likelihood of all bolts experiencing a simultaneous complete severance failure is very remote. The widespread use of valves with bolted bonnets and the low historical incidence of complete severance valve bonnet failures confirm that bolted valve bonnets need not be considered as credible missiles.
f. Valve stems are not considered as potential missiles if at least one feature, in addition to the stem threads, is included in their design to prevent ejection. Valves with backseats are prevented from becoming missiles by this feature. In addition, air- or motor-operated valve stems will be effectively restrained by the valve operators.
g. Nuts, bolts, nut and bolt combinations, and nut and stud combinations have only a small amount of stored energy and thus are of no concern as potential missiles.

3.5.1.2 Internally Generated Missiles (Inside Containment) Sources of internally generated missiles outside the containment are also applicable to the inside of the containment (see Section 3.5.1.1 for discussion). 3.5.1.3 Turbine Missiles The turbine generator stores large amounts of rotational kinetic energy in its rotors. In the unlikely event of a major mechanical failure, this energy may be transformed into both rotational and translational energy of rotor fragments. These fragments may impact the surrounding stationary parts. If the energy-absorbing capability of these stationary turbine generator parts is insufficient, external missiles will be released. These ejected missiles may impact various plant structures, including those housing safety-related equipment. The plant layout, as shown in the general arrangement drawings (Section 1.2), is a peninsular arrangement for the turbine generator. This layout minimizes the possibility of a turbine missile impacting the other plant structures and equipment essential for post-accident safe shutdown requirements. Section 10.2.3.6 describes the inspection requirements and the testing of valves, which prevent turbine overspeed that would cause the missile generation. 3.5-3 Rev. 25

WOLF CREEK The turbine generator was manufactured by General Electric Company (GE), and is described in Section 10.2. During Refueling 18 (RF18), three replacement LP-steampaths comprised of rotors, inner casings and diaphragms, and a single HP steampath consisting of a rotor and diaphragms were installed. The last stage buckets for the LP rotors increased from 38 to 43. Each rotor was manufactured by GE from a single piece of alloy steel forging, employing integral wheels and couplings (monoblock design), which resulted in a reduced rotor stresses and reduced potential for cracking. Therefore, the probability for turbine missiles was re-evaluated. During Refuel 19 (RF19) and Refuel 24 (RF24), the existing turbine control system, feedwater pump control system, and NSSS control systems were replaced with control systems based on the Ovation platform supplied by Emerson Process Management (EPM). The previous General Electric Mark II Electro-Hydraulic Control (EHC) System and Emergency Trip System (ETS) was replaced with a Westinghouse Ovation Digital EHC Turbine Control System (DCS). The Turbine Control System (TCS) architecture is based on combined functional and hardware redundancy to create a robust and reliable system. In order to increase reliability of the new TCS, the Ovation system is provided with redundancy as follows:

1. Two 100% capable controllers, one primary and one backup dedicated to overspeed protection and trip functions - Ovation Emergency Trip System (ETS).
2. Two 100% capable controllers, one primary and one backup dedicated to turbine control and providing backup overspeed protection and trip functions - Ovation Operator Auto/Overspeed Protection and Control (OA/OPC).
3. The system is configured to provide cross trips between the two sets of redundant controllers.

The ETS and the OA/OPC controllers interface with two sets of diverse and independent speed probes, which measure turbine speed. One set consists of three passive speed probes, which interface with the ETS controller. The other set consists of three active speed probes, which interface with the OA/OPC controller. To provide diverse overspeed protection, three additional passive speed sensors are fed to independent Woodward ProTech GII modules to provide an independent, diverse hardwired overspeed trip from the Ovation. Three independent Woodward ProTech GII modules determine the speed of the turbine by measuring the frequency of the output signal from each of the speed sensors. Woodward ProTech GII modules are also configured to energize an output relay when the speed feedback exceeds the setpoint of 111% of rated speed. This setpoint is set independently in each of the Woodward ProTech GII modules. The normally closed contacts from these output relays are wired to the trip circuit for the ETS Testable Dump Manifolds (TDM) solenoids. When an overspeed condition is detected, the contact from the Woodward ProTech GII module opens, and the ETS TDM solenoid is de-energized. This function is independent from the Ovation ETS controller logic and the ETS Speed Detector Modules (SDM) hardwired trip. Studies of known failures of turbine generator rotating elements have indicated that they be classified into two general types:

a. Failure of rotating components at or near normal operating speed.
b. Failure of components that control the admission of steam to the turbine, resulting in excessive shaft rotational speed and consequent mechanical failure.

3.5-4 Rev. 35

WOLF CREEK 3.5.1.3.1 Low-Speed Missiles Brittle Fracture Failure: The brittle fracture failure mechanism in rotors with shrunk on wheels is due to the initiation and growth of stress corrosion cracks to critical size in the exposed wheel keyway surfaces. The probability of this failure mode is dependent on environment, speed, temperature and material properties, as well as, inspection methods and inspection intervals. For a shrunk-on wheel operated at, or near, normal running speed, the probability of bursting and thus of missile generation, is dominated by this brittle fracture mechanism. The new rotors are of monoblock construction and do not have shrunk on wheels. Therefore, the formerly dominant brittle fracture failure mechanism, above, is eliminated in monoblock rotors. 3.5.1.3.2 High-Speed Missiles Significant steps in mechanical design have been taken in order to prevent turbine overspeed. The turbine generators for SNUPPS are provided with an overspeed protection system employing EHC. Table 1 of Reference 1 lists turbines that have experienced bursts of rotating parts. The only turbine in that list that experienced a high-speed burst is the Uskmouth No. 5 turbine designed and built by a British manufacturer. It was equipped with a control system which, in GE terminology, is described as mechanical-hydraulic controls (MHC). Valve opening actuation for the main steam turbine inlet valves is provided by a high pressure hydraulic system which is totally independent of the bearing lubrication system. Valve closing actuation is provided by springs and aided by steam forces following the reduction or relief of hydraulic pressure. The system is designed so that loss of hydraulic fluid pressure leads to valve closing and consequent shutdown. The main steam turbine inlet valves are provided in series arrangement. A group of stop valves actuated by either of two overspeed-trip signals is followed by a group of control valves modulated by the speed-governing system, and tripped by either overspeed-trip signal. These systems are described in Section 10.2.2.3.2. The intermediate valves are arranged in series-pairs, with an intermediate stop valve and intercept valve in one casing. The closure of either one of the two valves will close off the corresponding steam line. Thus, a single failure of any component will not lead to destructive overspeed. A multiple failure at the instant of load loss would be required, involving combinations of undetected electronic faults and/or mechanically stuck valves and/or hydraulic fluid contamination. The probability of such joint occurrences is extremely low, due both to the inherently high reliability of the design of the components and frequent inservice testing. For further description and functioning of intercept valves, refer to Section 10.2.2.3.2. 3.5-5 Rev. 27

WOLF CREEK 3.5.1.3.3 Overspeed Probability Ductile Failure: The probability of ductile failure for a rotor of any type is a function of speed, temperature and material tensile strength. In order to experience ductile failure, stresses within the rotor need to be close to, or exceed, the tensile strength of the material for the given operating temperature. The consideration of ductile failure herein assumes design component temperatures. The stresses in the rotor increase as speed increases, thus it is important to consider the probability of achieving a speed where ductile failure may occur. The previously approved GE probabilistic analysis of turbine over-speed is applicable to units with monoblock rotors. The over-speed analysis considers the characteristics of the turbine control system, the unit configuration and test requirements for the steam valves and other over-speed protection devices. This over-speed analysis shows that the probability of attaining a given over-speed decreases rapidly as the speed increases. As long as the control system is maintained and tested in accordance with Westinghouses recommendations, the annual probability of attaining a speed greater than or equal to 120% of normal speed is 2.44 x 10-6 (Ref. 8). 3.5.1.3.4 Probability of Damage The evaluation of turbine missile effects is commonly characterized by the following equation: P4 = P1 x P2 x P3 Where: P4 = Annual probability of unacceptable damage resulting from a turbine missile. P1 = Annual probability of turbine failure resulting in the ejection of turbine rotor (or internal structure) fragments through the turbine casing. P2 = The probability that a turbine missile strikes a critical plant target, given generation. P3 = The probability that the critical target is unacceptably damaged, given a missile strike. The NRC licensing guidelines (Regulatory Guide 1.115 and NUREG-1048) use this formulation to describe hypothetical turbine missiles and specifies that the probability of unacceptable damage from turbine missiles should be less than or equal to 1 in 10 million per year (i.e., P4 should be <1 x 10-7 per year per plant). Further definition in the guidelines, due to uncertainties associated with the calculation of P2 and P3, state the product of strike and damage probabilities to be 1 x 10-3 per year for a favorable oriented turbine, and 1 x 10-2 per year for an unfavorable oriented turbine. The total turbine missile generation probability (P1) requirements should be less than 1 x 10-4 per year for a favorable oriented turbine, and 1 x 10-5 per year for an unfavorable oriented turbine. The Wolf Creek turbine is favorably oriented (Ref. 6). This discussion summarizes the NRC approved methodology used to determine the annual missile generation probability (P1) for all GE steam turbines. The methodology summary of this discussion is consistent with the 1984 and 1993 GE Proprietary Missile Probability Reports, which focused on shrunk-on wheels of the low-pressure turbines as the critical source of turbine missiles. The methodology has been updated to include integral (monoblock) rotors. 3.5-6 Rev. 27

WOLF CREEK Methodology: The methodology of missile generation probability analysis deals with one element of the overall missile issue, which is the probability (P1) of generating a turbine missile from the LP turbine external to the LP inner casing and LP hood structure. P1 is defined by the following equation: P1 = PA x PB x Pc Where: PA = The probability of the turbine attaining speeds higher than those occurring during normal operation (overspeed). PB = The estimation of rotor burst probability as a function of speed. Pc = The probability of a rotor fragment penetrating the turbine casing and thus generating an external missile. Probability of Turbine Overspeed (PA): The probability of a rotor burst and the probability that a fragment will penetrate the turbine casing are both dependent on the speed at which rotor burst is assumed to occur. Under normal operating conditions, the turbine speed is close to the rated speed (1800 rpm). When an abnormal event occurs, such as a full load rejection and failure of elements of the control system, turbine speeds significantly higher than the rated speed may occur. A major component of the analysis is to estimate the probability of attaining various overspeed levels. Rotor Burst Probability (PB): One rotor failure mode considered is brittle burst, specifically as the result of a crack located in the radial-axial plane growing to a critical size. Brittle burst scenarios addressed are: 1) an undetected internal forging flaw that grows cyclically to critical size, and 2) a time dependent SCC that initiates on the outer body surface and grows to a critical size. A second failure mode due to tensile failure is also included in the methodology. This ductile failure mode contributes to the rotor burst probability particularly during abnormally high overspeed occurrences. Probability of Casing Penetration (PC): The third major component of the missile probability analysis methodology deals with the probability of a rotor burst fragment penetrating the turbine casing. This method considers the kinetic energy of the assumed fragment at the instant of burst as well as the energy absorbing capability of the stationary components of the low pressure turbine. Results: GE issued a formal missile probability assessment letter (Ref. 6) to WCNOC that verified the new turbine rotor configuration met the NRC annual missile probability limit (P1) of 1 x 10-4, subject to testing, managed inspections and implementation of any specified corrective actions identified as a result of the test/inspections. Westinghouse provided the reliability analysis for their new Ovation Digital EHC Turbine Control System (Ref. 7), which states the total failure probability attributed to the failure of the turbine control system is 1.07x10-9 per year. Based upon References 1, 6 and 7, the new calculated P1 for the Wolf Creek turbines is: 3.5-7 Rev. 27

WOLF CREEK P 2.44x10-6 per year The new calculated P4 (Ref. 8) is provided in Table 3.5-3. 3.5.1.4 Missiles Generated by Natural Phenomena Tornado-generated missiles were considered as the limiting natural-phenomena hazard in the design of all structures which are required for post-accident safe shutdown. The missiles considered in design are as listed in Table 3.5-1. Vertical velocities of 70 percent of the indicated horizontal velocities are considered for all missiles, except the 1-inch-diameter steel rod which is critical for penetration and is assumed to have a vertical velocity equal to the horizontal velocity. These design basis missiles are in accordance with Standard Review Plan 3.5.1.4, Revision 1 (Draft). 3.5.1.5 Missiles Generated by Events Near the Site As described in Section 2.2.3, there are no postulated explosions or military activities in the site vicinity that could generate missiles. 3.5.1.6 Aircraft Hazards The Burlington Municipal Airport has been replaced by the Coffey County Airport which opened in 1989 and is not described in Section 2.2.1.3. The hazards associated with the new airport have been evaluated and do not constitute a significant hazard as defined in Standard Review Plan, Section 3.5.1.6. The following evaluation applies to hazards due to the Coffey County Airport.

a. The following small airport is within 5 miles of the station:
1. the Coffey County Airport located 4.5 miles north-northwest of the site. It is classified as a small aircraft airport servicing primarily single and twin engine piston type aircraft.
b. There are no airports between 5 and 10 miles of the station.
c. There are no airports outside 10 miles of the plant with the projected annual number of operations greater than 1000 d (d=miles from site to airport).
d. There is a low-altitude federal airway and a high-altitude jet route passing within 2 miles of the plant which have widths of 8 and 16 nautical miles, respectively. In addition, there is a low level military training route whose centerline passes within 17 miles of the plant. A probability analysis of the aircraft accidents on these routes has indicated that the probability of accidents leading to radiological consequences worse than the exposure guidelines of 10 CFR 50.67 is less than 10-7 per year. The details of this analysis are given in the following.

V-234 is an east-west low-altitude route, and its centerline passes within 3.9 miles of the plant site. It has a width of 8 nautical miles. V-131 is a north-south route passing within 6.1 miles of the plant site. It has a width of 8 nautical miles. Daily traffic on these routes and from direct flights traversing the airspace overlying Wolf Creek is reported to be 161 flights. 3.5-8 Rev. 34

WOLF CREEK J-110 is a high-altitude east-west jet route passing within 0.5 miles of the plant site. It has a width of 16 nautical miles. Daily traffic on this route and from direct flights traversing the airspace overlying Wolf Creek is reported to be 132 flights. IR-502 is a military low-level training route whose centerline passes within 17 miles of the plant. It has a width of 8 nautical miles, and the annual number of flights on this route is 1,560. Conservatively, the hazard from this route has been accounted for in the analysis with the assumption of an in-flight crash rate 2 times higher than that for general aviation. The traffic count on low-altitude air routes does not include the aircraft operating under visual flight rules (VFR). It is conservatively assumed that the VFR traffic is equal to the IFR (instrumental flight rule) traffic reported above. The increase of traffic in the future is practically offset by a decrease in accident rates. The area of the safety-related structures in the plant whose damaged would lead to unacceptable radiological consequences is calculated to be 0.008 mi2. AIRCRAFT IMPACT PROBABILITY DUE TO AIR TRAFFIC The probability, PFA, of an aircraft crashing into the plant and leading to radiological consequences in excess of 10 CFR 50.67 exposure guidelines is calculated as follows: PFAY = C A WL + W T + WH N 2N N L T H where: C = inflight crash rate per mile for aircraft using airway = 4 x 10-10; A = area of safety-related structures whose damage would lead to unacceptable radiological consequences = 0.008 mi2; NL = number of aircraft movements on low-altitude federal air routes V-234 and V-131 [161 x 2] = 322/day = 117,530/year; WL = width of low-altitude air route = 8 nautical miles = 9.2 miles; NT = number of aircraft movements on the training route IR-502 

       = 1,560/year; WT =  width of training route IR-502 (plus twice the distance from the airway edge to the site since the site is outside the airway) = 34.0 miles; NH =  number of aircraft movements on high-altitude federal jet route J-110 = 132/day = 48,180/year; and WH =  width of jet route J-110 = 18.4 miles.

Using these data, the probability of an aircraft crashing into the plant and causing unacceptable radiological consequences is calculated as 5.0 x 10-8 per year. AIRCRAFT IMPACT PROBABILITY DUE TO COFFEY COUNTY AIRPORT The Standard Review Plan (section 2.2) specifies a method for calculating the probability of aircraft impact due to airports located near a nuclear power 3.5-9 Rev. 34

WOLF CREEK plant. The probability per year of an aircraft crashing into the site (PA) is calculated by using the following expression: PA = C x N x A where: C = probability per square mile of a crash per aircraft movement N = number of aircraft movements A = effective plant area (in square miles) Where multiple aircraft types and/or trajectories are involved this expression may be summed for each separate aircraft type and trajectory considered. Aircraft using the Coffey County Airport are lumped into a single category for the purposes of calculating the effective area as described below. No credit was taken for separate trajectories. Numerical values for "C" are given in the standard review plan as a function of the distance from the airport to the nuclear power plant. The Coffey County Airport is located approximately 4.5 miles from Wolf Creek. The probability for general aviation aircraft for airports from 4 to 5 miles from nuclear power plants is given as 1.2 X 10-8 per aircraft movement. The safety of general aviation has improved significantly in the intervening years and therefore the SRP crash probabilities are now very conservative. The National Transportation Safety Board publishes the "Annual Review of Aircraft Accident Data-U.S. General Aviation." These publications were reviewed for the period of 1982 through 1986 (the most recent 5 years of available data) and the number of aircraft accidents occurring within 5 miles of airports has decreased by more than half in this 5 year period alone. The effective area of the plant "A" is the portion of the plant that is susceptible to impact from a given type of aircraft and could result in radiological consequences greater than 10 CFR 50.67 guidelines. For the large commercial and military aircraft considered in the impact probability due to air traffic, a value of 0.008 miles2 was used. However, for the small general aviation aircraft utilizing the Coffey County Airport a significantly smaller value is appropriate. Calculations show than an effective area of 0.0016 miles2 may conservatively be used for general aviation aircraft with approach speeds of <140 mph and weighing <12,500 pounds. These parameters envelope the expected aircraft usage at Coffey County Airport through the year 2000. Substituting the values described above along with actual usage levels results in: PA = 4.0 x 10-8 When combined with the probability originally calculated for air routes of 5 x 10-8 this results in a total probability of aircraft impacts causing significant radiological releases of 9 x 10-8 per year. This result remains below the value of 1 x 10-7 per year given in the SRP as acceptable for siting of nuclear power plants. Since the aircraft movements at the airports and on the air routes do not pose any undue risk to the safe operation of WCGS Unit No. 1, no design-basis aircraft impact is postulated. 3.5-10 Rev. 34

WOLF CREEK 3.5.2 SYSTEMS TO BE PROTECTED The sources of internal missiles which, if generated, could affect the safety of the plant are considered in Appendix 3B. The turbine and tornado missiles which, if generated, could affect the safety of the plant are discussed in Section 3.5.1. All safety-related systems and components to be protected from tornado missiles are enclosed within protective structures which meet the requirements of Regulatory Guide 1.117. A tabulation of protective structures, their minimum wall thickness, and concrete strength are given in Table 3.5-2. The protective structure requirements for the RWST are discussed in Section 6.3. Openings to these structures are designed to prevent the entry of the design basis missile when the result would preclude the safety functions of the enclosed system or components. Prevention of missile entry includes the use of missile doors, barriers and shields at openings and adjacent buildings as shields in penetration areas. The missile barriers are designed utilizing the procedures given in Section 3.5.3. Further description of the seismic Category I structures is provided in Section 3.8.1 for the reactor building and Section 3.8.4 for other structures. The probability of significant damage (P4) to critical components in the plant due to turbine failure has been assessed by first determining the separate probabilities of turbine failure and missile ejection (P1, Refer to Section 3.5.1.3.4), such as a missile striking an entire structure of safety significance (P2), and significant damage occurring to the component (P3, Refer to Section 3.5.1.3.4). Then the overall annual probability P4 = P1 x P2 x P3. The probability of a high or low trajectory, turbine missile striking a structure housing a critical component (P2) is found to be 3.80 x 10-4 at the Wolf Creek site. Refer to Table 3.5-1. In addition, the probability of a high trajectory turbine missile striking a structure housing a critical component is 2.72 x 10-4 which is based on a total available target area of 110,991 square feet. The annual probability (P4) of a turbine missile damaging a critical component at the Wolf Creek site is found to be 2.49 x 10-9. This value is less than 10-7 and is sufficiently low so that no specific protective measures are required for turbine missiles. Refer to Table 3.5-3. Figure 3.5-1 identifies the safety-related structures, including those outside the power block, within the turbine missile trajectory. Protective measures are provided to minimize the effect of potential tornado-generated missiles. The protective structures, shields, and barriers are designed utilizing the procedures given in Section 3.5.3. The portions of the essential service water system (ESWS) located outside the power block requiring protective structures, shields, and barriers are discussed below. 3.5.2.1 Essential Service Water System Pumphouse The ESWS pumphouse is a tornado-resistant, reinforced concrete structure with an operating floor at elevation 2000 ft. The separation of the trains of the ESWS is provided by an interior barrier wall. A tornado-resistant skimmer wall 3.5-11 Rev. 28

WOLF CREEK at the ultimate heat sink (UHS) interface provides protection for the ESWS traveling water screens and the ESWS pumps, whose suction ends are located 28 feet below the normal surface of the UHS. Tornado-resistant shields protect the inlets and outlets of the ventilation system at the roof elevation and protect the personnel doors at grade level. Tornado-resistant covers protect the roof openings. Tornado-resistant shields cover the ESWS Pumphouse forebay pits to protect the ESWS Warming Lines. Figures 3.8-131, 3.8-132, 3.8-133, and 3.8-145 show the tornado missile protection for the safety-related penetrations in the ESWS pumphouse. 3.5.2.2 Essential Service Water System Pipes, Electrical Duct Banks and Manholes All ESWS pipes are buried a minimum depth of 4.5 feet to resist the effects of tornado-generated missiles and frost penetration. The ESWS discharge piping and warming lines use alternate methods for missile and frost protection. The ESWS discharge piping, from the last access vault to the discharge point, are encased in 4,000 psi compressive strength concrete with a minimum of 2 feet cover above the top of the pipe for missile protection. The discharge piping frost protection is provided by normal flow through the piping. The ESWS warming lines, on the north and south side of the pumphouse are covered with 3-6 of granular compacted backfill (CCF1) and 9 of reinforced concrete for missile protection. The normal warming line frost protection is provided by normal flow through the piping. All ESWS electrical duct banks are reinforced concrete structures which are buried at a minimum depth of 4 feet to resist the effects of tornado-generated missiles and frost penetration. The buried ESWS electrical manholes are tornado-resistant, reinforced concrete structures with missile-resistant manway covers and roofs. Figure 3.8-140 shows the tornado missile protection for the ESWS electrical manholes. 3.5.2.3 Essential Service Water System Access Vaults Essential Service Water System Access Vaults The buried ESWS access vaults are tornado-resistant, reinforced concrete structures with missile-resistanct access and manway covers. Figure (3.8-143) shows the tornado missile protection for the ESWS access vaults. 3.5.2.4 Essential Service Water System Discharge Point The submerged ESWS discharge piping follows the grade of the lakebed from the shoreline to the discharge point at 23 feet below the normal surface of the Ultimate Heat Sink. The discharge point is sufficiently protected from tornado-missile damage by being submerged. 3.5.2.5 Diesel Generator Building The barrier separating the two diesel generators is a 2-foot-thick reinforced concrete wall. The wall reinforcement is such that the wall is capable of withstanding the impact of all the externally generated missiles identified in Table 3.5-1. There are four openings in the wall, but they are located within 3 feet of the north end of the building. This location and the small size of the openings (1 3.5-12 Rev. 28

WOLF CREEK foot square or smaller) effectively prevents any internally generated missiles from passing through the openings and damaging equipment in the adjacent area. In addition, these openings actually serve as penetrations for piping and are sealed. 3.5.2.5.1 Diesel Engine Missiles The WCGS diesel engine is a low speed (514 rpm) engine which has a vented crank case. The engine manufacturer has never experienced nor knows of any crank case explosions or engine failures which resulted in missiles. As noted above, the internal wall separating the two diesel engines is designed to withstand a tornado missile impact. In the highly unlikely event that the engine did generate an external missile, the energy of that missile would be significantly less than that of the tornado missile. 3.5.2.5.2 Air Tank Missiles The air tanks are seismically mounted on their skids, which are in turn seismically anchored to the floor. Rupture of a tank would not generate missiles whose energy exceeds that of a tornado missile. 3.5.2.5.3 Pipe Break Missiles There are no high energy lines in the diesel generator building. The only moderate energy lines are those directly associated with each diesel engine. Therefore, a postulated failure of a moderate energy line would be considered the diesel single failure. There are no open penetrations between rooms, and therefore, flooding of one room will not degrade the opposite diesel engine. 3.5.2.5.4 Fuel Oil Storage Tank The fuel oil storage tank fill and vent lines rise above grade within the diesel generator building and then penetrate the building wall to the outside. The portion of these lines within the building is seismically restrained. Failure of these lines does not jeopardize operation of the diesel. If the fill line is unusable, the tank manhole can be used as the fill and vent connection if the tanks have to be replenished. In addition to the transfer line, the storage tank and the day tank are also interconnected via the overflow and recirculation line. Should the storage tank vent be totally restricted, venting can occur through the day tank. (It should be noted that the vent sizes are based on filling operations and not engine operations. The operating vent requirements are significantly less than those required for filling). In the unlikely event that both tank vents are completely restricted, either tank can be vented by alternative means.. Since failure of the nonseismic storage tank vent and fill lines will not prevent system operation, no tornado protection is provided. 3.5.2.6 Diesel Exhaust Stack Although the WCGS safety-related structures and components are designed for the design basis tornado and the design basis tornado missiles, the diesel exhaust stacks can reasonably be exempt from the requirement for specific missile barriers without jeopardizing the health and safety of the public. During the PSAR review stage, the NRC staff questioned the tornado missile protection provided for the stacks (Question 020.13, 430.38), and reached the same conclusion for the present design. 3.5-13 Rev. 30

WOLF CREEK After the construction permit review was complete, the NRC issued Regulatory Guide 1.117 "Tornado Design Classification," which is applicable to Construction Permit applications docketed after May 30, 1978. Even though this guide is not applicable to WCGS, it was addressed in the USAR, and its provisions are adequately met to state that the design is in compliance with the regulatory recommendations. The basis for this conclusion includes consideration of 1) the exhaust stacks inherent resistance to damage from credible missiles and the acceptability of penetration and/or significant denting, 2) the improbability of design basis tornadoes and the low probability that the design basis missiles could exist at the high elevations required, and 3) the significant protection afforded the stacks by existing plant structures. The arguments below demonstrate that it is extremely unlikely that a tornado missile will damage an exhaust stack and inhibit diesel operation. It is even more unlikely that both stacks could be damaged. 3.5.2.6.1 Design The diesel stacks are seismically supported, 35 feet apart, and inherently resistant to damage from tornado missiles due to their large diameter (42-inch O.D.) and 3/8-inch-thick steel wall construction. High kinetic energy missiles could, however, deform the stack or even penetrate it if the impact area is small relative to the kinetic energy. Penetration or significant deformation will not adversely affect the function of the stack since they are oversized. The total allowable pressure drop for the exhaust system for rated power output is 10 inches of water. The pressure drop from the engine through the exhaust silencer and to the diesel building roof line is approximately 5 inches of water. The exposed portion of pipe above the diesel generator roof is only 50 feet long and has an allowable length of more than 926 feet (corresponding to an allowable pressure drop of 5 inches of water). If this pipe were only 32 inches in diameter, its allowable length would be 280 feet. Thus significant local damage due to denting or penetration by a tornado missile is acceptable because full power diesel operation would not be impaired. 3.5.2.6.2 Missile Selection Criteria The improbability of any missile of high density and high energy being elevated to the heights of the diesel stacks is obvious from NUREG-0121 "An Assessment of the Basis for Selection Criteria for Protection Against Tornado-Entrained Debris." Exerpts from NUREG-0121 and Operating Agent remarks are presented in Table 3.5-5 . These discussions are provided to highlight the low probability of any high density, high energy missile which approaches the characteristics of the current set of design basis missiles. 3.5.2.6.3 Structural Protection The following discussion addresses horizontal missile trajectories and missiles recently ejected from the maximum windfield. Missiles falling from greater heights are not specifically addressed, since it is considered extremely improbable that a design basis missile will exceed the heights of the surrounding buildings. USAR Figures 1.2-26, 1.2-27, and 1.2-28 provide detailed plan and elevation views of the stacks and surrounding structures. 3.5-14 Rev. 29

WOLF CREEK Figure 3.5-2 depicts the plant location of the diesel stacks and the inherent protection provided for them by the surrounding power block structures from tornado missiles which could potentially affect the full power operation of the diesel. For analysis purposes approach of missiles on the diesel stacks has been considered for seven zones. (Zones A through G are indicated on Figure 3.5-2) The exact boundaries of each zone are not strictly defined since the stacks are 35 feet apart and tumbling missiles would affect the zone boundaries. Zone A Protection is provided by the control building to a height of 87 feet. Only the top 10 feet of the stacks are exposed to missiles which must rise above nine stories and traverse the control building roof prior to impacting the stacks. Zone B The turbine building roof is approximately 140 feet high and would effectively preclude design basis missiles from reaching the diesel stacks. The control building again affords protection for most of the stacks. Zone C The containment structure provides complete protection from missiles from this direction. Zone D The fuel building is 106 feet high and provides complete protection from credible missiles possessing sufficient energy to inflict adverse damage. Zone E The diesel generator intake penthouse provides protection up to 66 feet above grade. The radwaste building will also provide protection up to 55 feet above grade and effectively disrupt the funnel and windfield to help eject previously entrained missiles prior to their reaching the diesel building. Zone F The diesel generator intake penthouse provides protection up to 66 feet above grade to effectively shield the exhaust stacks from high energy missiles. Zone G This relatively narrow zone is the least protected direction for which missiles could emanate and impact the stacks. However, missiles of concern would have to be raised over five stories while being accelerated to high velocities. These missiles would have to be ejected from the maximum windfield at a significant distance from the control/diesel building to reach the stacks. Once a funnel reaches these buildings, the windfield will be disturbed, and entrained missiles will be less likely to have been accelerated to high velocities. For a tornado approaching from the east, missiles in the leading edge of the windfield when the funnel reaches the diesel building will be traveling in a north/south direction and not impact the stacks. 3.5-15 Rev. 25

WOLF CREEK 3.5.3 BARRIER DESIGN PROCEDURES The plant layout is based on optimizing 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 hazard occurs within the plant, there is a minimum effect on other systems or components required for a post-accident safe shutdown. Missile-resistant barriers and structures are designed to withstand and absorb missile-impact loads to prevent damage to the protected structures, systems, and components. 3.5.3.1 Tornado Missile Barrier Design Procedures Tornado-resistant structures may sustain local missile damage, such as partial penetration and local cracking and/or permanent deformation, provided that structural integrity is maintained, perforation is precluded, and the contained seismic Category I systems, components, and equipment are not subjected to damage by secondary missiles, such as from concrete spalling and scabbing. The wall and roof thicknesses provided to resist the effects of tornado-generated missiles are considered to be more than adequate. It is considered that a thickness of 24 inches for reinforced concrete with a minimum strength of 4,000 psi for the walls and (either 21 inches for the roof with minimum concrete strength of 4,000 psi or 18 inches for the roof with minimum concrete strength of 5,000 psi) roof slabs of seismic Category I structures are adequate to resist the impact of tornado-generated missiles for both penetration and structural response. This is based on the results of the test program, "Missile Impact Testing of Reinforced Concrete Panels," (Ref. 3) and on the EPRI Report, "Full-Scale Tornado Missile Impact," (Ref. 4). The ESW Vertical Loop Chase structure (walls and roof) is constructed from 1/2 inch thick carbon steel plate that works in conjunction with the hollow steel section (HSS) tubular steel substructure to provide an adequate means to resist the impact of tornado-generated missiles for both penetration and structural response. 3.5.3.2 Barrier Design Procedures for Internally Generated Missiles In general, when separation is not feasible, additional protection from internal missiles is provided by barriers. The procedures and calculations employed in the design of missile-resistant barriers for turbine missiles and other internally generated missiles are described in Reference 5. In the design calculations for missile resistant barriers, ductility ratios never were greater than 10. Therefore additional details are not required here. Appendix 3B discusses the protection required for internally generated missiles. 3.

5.4 REFERENCES

1. Hypothetical Turbine Missiles Probability of Occurrence, General Electric Memo Report Dated March 14, 1973.
2. Delete
3. "Missile Impact Testing of Reinforced Concrete Panels,"

Calspan Report No. HC-5609-D-1, Calspan Corporation, Buffalo, New York, January 1975. 3.5-16 Rev. 32

WOLF CREEK

4. Stephenson, A. E., "Full-Scale Tornado Missile Impact," EPRI Report No. NP-440, July 1977.
5. "Design of Structures for Missile Impact," BC-TOP-9-A, Revision 2, Bechtel Power Corporation, San Francisco, California, September 1974.
6. Turbine Missile Analysis Statement, General Electric Correspondence to WCNOC, dated 10/20/2009. Correspondence Number 10-00055.
7. WNA-AR-00155-SAP, Turbine Control System Upgrade Reliability and Fault Tree Analysis, Rev. 1.
8. Calculation AC-M-005, Turbine Missile Probability Study, Rev. 0.
9. Calculation 020544.14.01-C-003, Tornado Missile Impact Analysis of ESW Piping System Steel Tower, Rev. 0.

3.5-17 Rev. 29

WOLF CREEK TABLE 3.5-1 CHARACTERISTICS OF POSTULATED TORNADO MISSILES Horizontal Missile Weight, lbs Velocity, fps Wood plank, 115 272 4" x 12" x 12 long Steel pipe, 6" diameter, 286 170 schedule 40, 15 long Steel rod, 1" diameter, 9 167 3 long Utility pole, 13.5" 1,123 180 diameter, 35 long Steel pipe, 12" diameter, 749 154 schedule 40, 15 long Automobile, 16.4 3,991 194 x 6.6 x 4.3 Rev. 0

WOLF CREEK TABLE 3.5-2 STRUCTURES PROVIDING TORNADO MISSILE BARRIER PROTECTION Structure Nominal Concrete Thickness 90-Day Strength Reactor building 4 ft - wall 4,000 psi 3 ft - dome 4,000 psi Auxiliary building 2 ft - wall 4,000 psi 1 1/2 ft - roof 5,000 psi Control building 2 ft - wall 4,000 psi 1 1/2 ft - roof 5,000 psi Diesel generator building 2 ft - wall 4,000 psi 1 1/2 ft - roof 5,000 psi Fuel building 2 ft - wall 4,000 psi 1 1/2 ft - roof 5,000 psi ESW Vertical Loop 1/2 is ASTM A36 carbon NA Chase steel plate and 5/8 in HSS tubing - wall 1/2 in ASTM A36 carbon steel plate and 5/8 in HSS tubing - roof Rev. 29

WOLF CREEK TABLE 3.5-3 TURBINE MISSILE PROBABILITIES HIGH ENERGY, HIGH AND LOW TRAJECTORY Missile Source Target Structure Striking Probability, P2 

                                                             -4 Turbine                 Power Block                1.30 x 10
                                                             -4 Turbine                 ESWS Pumphouse             0.17 x 10 (High Trajectory)
                                                             -4 Turbine                 ESWS Pumphouse             1.08 x 10 (Low Trajectory on Vertical Wall)
                                                             -4 Turbine                 Buried ESWS                1.25 x 10 Pipes and Duct Banks
                                                            -4 TOTAL P2 =      3.80 x 10 P4 = P1 x P2 x P3 P1 = Less than or equal to 2.49 x 10-6 per year (Refer to Section 3.5.1.3.4)

P2 x P3 = 1 x 10-3 per year (Refer to Section 3.5.1.3.4) P4 = 2.49 x 10-9 per year*

  • This probability meets the NRC requirements -7 stated in section 3.5.1.3.4, that P4 should be less than 10 per year per plant.

Rev. 28

WOLF CREEK TABLE 3.5-4 RANGE OF MAXIMUM AND MINIMUM TURBINE MISSILE VELOCITIES (FPS) Turbine Wheel Groups(1) I II III Max. Min. Max. Min. Max. Min. Va(2) 470 393 550 413 610 400 Vb(2) 620 484 750 521 780 498 Vc(2) 930 625 880 625 910 625 Vd(2) (3) 1040 (3) 1053 (3) 996 NOTES: (1) The turbine manufacturer has divided the various turbine wheels into three groups for analysis purposes. (2) Velocities have been categorized, Va through Vd, to repre sent four different failure modes considered possible by the manufacturer. (3) The minimum velocity required to perforate the structure roof and strike a target, Vmin, is greater than the maximum velocity achievable for the postulated failure mode, Vmax. Rev. 0

WOLF CREEK TABLE 3.5-5 NUREG 0121 APPLICABILITY TO DIESEL STACK DESIGN Excerpt From NUREG 0121 Remarks__________ Any set of design basis tornado missiles should consider:

1. Object likely to be in the plant vicinity.
2. Objects in the vicinity and likely to become airborne and hurled by a tornado windfield.
3. Airborne objects likely to damage plant structures if impacted at high speed.

Of the present list of seven missiles, only missile "G", the automobile, clearly meets all three tests. For the other high density missiles the potential for lofting and acceleration is questionable." (NUREG-0121, Page 2) The automobile and utility poles are not credibly postulated above 30 feet. The minimum height of the exposed stack is 47.5 feet. The information in WASH 1300 suggests that the design basis tornado is itself no more probable than 10-7 per year and that the median tornado windspeed of all U.S. tornadoes studied was about 45 m/sec. Fewer than 10 percent of these studied tornadoes were deduced to have windspeeds above 70 m/sec., fewer than 1 percent above 90 m/sec., and fewer than 0.1 percent above 130 m/sec. In regard to damage potential, typical missiles in a 10-7 per year tornado become very rare in a 10-6 per year tornado, and are physically impossible in tornadoes having higher incidence rates. (NUREG-0121, Pages 3 and 4) Only the near design basis tornado missiles have sufficient energy to cause adverse damage to the diesel stacks. Rev. 0

WOLF CREEK TABLE 3.5-5 (Sheet 2) A rigid body may be lifted into a windfield by any or all conceivable mechanisms: (1) aerodynamic lift..., (2) drag lift , (3) suction... These last two mechanisms, however, are of great importance only to ojbects directly in the path of the tornado vortex. The first mechanism, aerodynamic lift, can be postulated to affect missiles over a much greater area. (NUREG-0121, Pages 6 and 7) Unless a missile is lifted while in the vortex of the tornado, it will not achieve any significant height unless its aerodynamic properties are unique. Missiles of concern to the diesel stack would have to be lifted between 47 and 97 feet while remaining in the vortex and high wind speed region of the tornado in order to be accelerated to significant velocities. Missiles that do not "fly" will experience predominantly horizontal forces and will be accelerated by them. At some point in their trajectory, a maximum velocity will be reached, after which the missile must decelerate. (NUREG-0121, Page 8) The missiles of concern are dense and usually of poor airfoil design and are therefore unlikely to "fly" to the heights required to damage the diesel stacks. In order to achieve any significant fraction of the maximum tangential wind speed, a massive missile must pass through the maximum wind, and the most significant single parameter of a missile trajectory in determining that missiles maximum velocity is the distance traveled within the maximum windfield... such missiles are lifted and accelerated only by the highest velocity winds but can not be easily deflected from a nearly straight path in order to follow the tornado vortex. (NUREG-0121, Pages 12 and 13) Rev. 0

WOLF CREEK TABLE 3.5-5 (Sheet 3) Massive missiles fly straight and therefore do not remain in the maximum windfield for long durations. Once out of the maximum windfield they fall rapidly due to gravity. Therefore the missiles which could damage the diesel stacks would have to be raised to heights greater than 47.5 feet or be in or near the maximum windspeed section of the tornado at the moment of impact. The study has shown that it is relatively easy for a missile to acquire about 10 percent of the maximum tornado windspeed by a brief passage in the windfield, but to acquire significantly higher velocities, a comparatively long distance must be traveled within the windfield. However, massive missiles cannot stay within a windfield long enough to attain high velocities because of centrifugal forces. (NUREG-0121, Page 23) Only high energy missiles are of concern to the diesel stack functionality. Real tornadoes are not uniform windfield, but distorted helical flows... Should a tornado pass over the barrier while a missile is entrained therefore, there is a significant probability that the missile will not, in fact, impact the barrier. (NUREG-0121, Page 17) The surface of the diesel stack which is normal to the flight of a postulated missile is small. Glancing blows of a missile or impacts of an end of a tumbling missile will not adversely affect the diesel stacks. Rev. 0

29 ESW VERTICAL LOOP CHASE REV. 29

29 ESW VERTICAL LOOP CHASE ZONE E RADWASTE STORAGE ADDITION RADWASTE STORAGE REV.29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.5-2 STRUCTURAL PROTECTION FOR DIESEL EXHAUST STACKS FROM TORNADO MISSILES

WOLF CREEK 3.6 PROTECTION AGAINST THE DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING Pipe failure protection is provided in accordance with the requirements of 10 CFR 50, Appendix A, GDC 4. In the event of a high- or moderate-energy pipe failure within the plant, adequate protection is provided to ensure that those portions of the essential structures, systems, or components whose failure could compromise the integrity of the reactor coolant system or reduce the functioning of any plant feature required for a post-accident safe shutdown to an unacceptable level are designed, constructed, and protected so as not to fail or cause such a failure. Appendix 3B, Hazards Analysis, provides several examples of the evaluations made of the effects of postulated pipe failures within the plant. The following sections provide the bases for selection of the pipe failures, the determination of the resultant effects, and details of the protection requirements. 3.6.1 POSTULATED PIPING FAILURES IN FLUID SYSTEMS INSIDE AND OUTSIDE CONTAINMENT Table 3.6-1 provides a matrix which indicates high-energy systems, moderate-energy systems and safety-related systems. Selection of pipe failure locations for evaluation of the consequences on nearby essential systems, components, and structures, is presented in Section 3.6.2 and, except for the reactor coolant loop, is in accordance with Regulatory Guide 1.46, and NRC BTPs ASB 3-1 and MEB 3-1. The dynamic effects from postulated pipe breaks have been eliminated from the structural design basis of the reactor coolant system primary loop piping, as allowed by revised General Design Criterion 4 (Reference 15). The elimination of these breaks is the result of the application of leak-before-break (LBB) technology, as presented in Reference 16, 17, and 18, and approved for WCGS by the NRC (Reference 19). 3.6.1.1 Design Bases The following design bases relate to the evaluation of the effects of the pipe failures determined in Section 3.6.2.

a. The selection of the failure type is based on whether the system is high- or moderate-energy, based on normal operating conditions of the system.

High-energy piping includes those systems or portions of systems in which the maximum operating temperature exceeds 200 F or the maximum operating pressure exceeds 275 psig, during normal plant conditions. 3.6-1 Rev. 14

WOLF CREEK 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 which exceed 200°F or 275 psig for 2 percent or less of the time the system is in operation or which experience high-energy pressures or temperatures for less than 1 percent of the plant operation time are considered moderate-energy.

b. Except for the reactor coolant system, the worst case operational plant conditions (including startup, operation at power, hot standby, shutdown, and upset conditions) are used to determine the piping system support/restraint requirements and to determine blowdown rates for jet impingement loads. For the reactor coolant system, including all Class 1 branch piping, the normal power operation conditions are used as described in Reference 1.
c. Moderate-energy pipe cracks were evaluated for wetting from spray, flooding, and other environmental effects.
d. Each longitudinal or circumferential break in high-energy fluid system piping or leakage crack in moderate-energy fluid system piping was considered separately as a single postulated initial event occurring during normal plant conditions.
e. Offsite power was assumed to be unavailable if a trip of the turbine-generator system or reactor protection system was a direct consequence of the postulated piping failure, unless it was more conservative to assume that offsite power was available (e.g., a feedwater line break with offsite power available leads to a larger inventory of water for flooding considerations).
f. A single active component failure was assumed in systems used to mitigate the consequences of the postulated piping failure and to safely shut down the reactor, except as noted in Paragraph g below. The single active component failure was 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.

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g. When the postulated piping failure occurs and results in damage to one of two or more redundant or diverse safety-related trains, single failures of components in other trains (and associated supporting trains) are not assumed. Postulated failures are precluded, by design, from affecting the opposite train or from resulting in a DBA. The safety-related systems are designed to the following criteria: a) seismic Category I standards, b) powered from both offsite and onsite sources, and c) constructed, operated, and inspected to quality assurance, testing, and in-service inspection standards appropriate for nuclear safety systems.
h. All available systems, including those actuated by operator actions, are employed to mitigate the consequences of a postulated piping failure to the extent clarified in the following paragraphs:
1. In determining the availability of the systems, account was taken of the postulated failure and its direct consequences, such as unit trip and loss of offsite power, and of the assumed single active component failure and its direct consequences. The feasibility of carrying out operator actions was determined on the basis of ample time and adequate access to equipment being available for the proposed actions. Although a postulated high/moderate-energy line failure outside the containment may ultimately require a cold shutdown, operation at power or hot standby was assumed as allowed by the plant technical specifications. During this period plant personnel would assess the situation and make repairs.
2. The use of nonseismic Category I equipment is clarified in the following paragraphs:

(a) For nonseismic Category I piping failures, it was assumed that a safe shutdown earthquake could be the cause of the failure. Thus, only seismic Category I equipment could be used to bring the plant to a post-accident safe shutdown. (b) For seismic Category I and seismically supported nonseismic Category I piping failures, it was assumed that the failure was caused by some mechanism other than an earthquake. Thus, nonseismic Category I equipment could be used to 3.6-3 Rev. 14

WOLF CREEK bring the plant to a post-accident safe shutdown, subject to the power being available to operate such equipment as discussed in Paragraph h(1) above.

i. A whipping pipe was not considered capable of rupturing impacted pipes of equal or greater nominal pipe diameter and equal or greater thickness, assuming that only "piping" was determined to do the impacting. A whipping pipe was considered capable of developing a through wall leakage crack in a pipe of larger nominal pipe size with thinner wall, assuming that only "piping" was determined to do the impacting. Where the potential existed for valves or other components in the whipping pipe to impact the targets, the above criterion was not utilized and the whipping pipe was not allowed to impact a safety-related component.
j. Pipe whip was 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 was considered to form a plastic hinge and rotate about the nearest rigid restraint, anchor, or wall penetration. 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 was assumed that the pipe comes to rest against that surface, with no pipe whip in other directions. If unrestrained, a whipping pipe without a constant energy source (i.e., a break at a closed valve with only one side subject to pressure) was not considered capable of forming a plastic hinge and rotating, provided that its movement could be defined and evaluated. Pipe whip restraints are provided wherever postulated pipe breaks have any possibility of affecting any system or component required for the mitigation of that break or post-accident safe shutdown of the plant. Unrestrained pipe breaks are limited to those areas of the plant that are physically separated from the systems and components required for pipe break mitigation or post-accident safe shutdown.

k. The calculation of thrust and jet impingement forces considers any line restrictions (e.g., flow limiter) between the pressure source and break location and the absence of energy reservoirs, as applicable.

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l. Initial pipe break events were not assumed to occur in pump and valve bodies because of their greater wall thicknesses.
m. A survey of all potential internal flooding sources was performed for all rooms with safety-related components.

This survey determined the worst case internal flooding event for each room. From this survey, calculations were performed to determine the worst case flood level in each of these rooms. A summary of these flood levels is provided in Table 3.6-6. Additional information on containment flooding is provided in Sections 6.2.2.1.3 and 6.3.2.2. Assumptions used in arriving at the worst case flooding event are as follows:

1. One break or crack occurs at a time
2. Nonseismic lines will experience guillotine breaks during seismic events
3. Drain pipes are assumed to be dry before the break or crack
4. Rooms drain through the floor drain(s). No credit is taken for drainage through uncapped or unsealed equipment drains. Typically, no credit is taken for drainage out under doors.
5. Pipes which are supported II/I and are moderate energy during normal plant operating modes are assumed to develop moderate energy cracks only.

3.6.1.2 Description Systems, components, and equipment required to safely shut down the plant and mitigate the consequences of postulated piping failures (hereinafter called essential) were reviewed, in order to comply with the design bases, to determine their susceptibility to the failure effects. The break and crack locations were determined in accordance with Section 3.6.2. Figure 3.6-1 and 3.6-3 show the high-energy pipe break locations and break types. Those essential systems which are subject to the consequences of pipe failure are summarized in Table 3.6-1. The type of hazard (i.e., whipping, jet impingement, spraying, and flooding) is 3.6-5 Rev. 0

WOLF CREEK shown. This summary was based on the detailed failure mode analysis discussed in Section 3.6.1.3, Section 3.6.2.5, and Appendix 3B. The design comparison to Regulatory Guide 1.46 positions, incorporating the comparison to NRC BTP MEB 3-1 and NRC BTP ASB 3-1, is provided in Table 3.6-2. Pressure response analyses were performed for the subcompartments containing high-energy piping. For a detailed discussion of the line breaks selected, and pressure results, refer to Section 6.2.1.2 and Table 3.6-4 for subcompartments inside the containment and Table 3.6-4 for subcompartments located outside the containment. The analytical methods used for pressure response analysis are in accordance with Reference 12. Appendix 3B discusses hazards analysis and Table 3B-1 shows a typical hazards analysis. In the control building, the effects of postulated failures of high energy lines would not impair the integrity or operability of safety related structures, systems or components. There are no effects upon the habitability of the control room from pipe break or pipe whip. Further discussion of the control room habitability systems is provided in Section 6.4. 3.6.1.3 Safety Evaluation 3.6.1.3.1 General An analysis of postulated pipe failures was performed to identify those safety-related systems, components, and equipment that provide protective actions required to mitigate the consequences of the failure. By means of protective measures such as separation, barriers, and pipe whip restraints, discussed below, the effects of breaks and cracks are prevented from damaging essential items to an extent that would impair their design function or necessary component operability. Typical measures used for protecting the essential systems, components, and equipment are outlined below and discussed in detail in Section 3.6.2.4. The ability of specific safety-related systems to withstand a single active failure concurrent with the postulated event is discussed, as applicable. 3.6-6 Rev. 6

WOLF CREEK When the results of the pipe failure effects analysis showed that the effects of a postulated high-energy break or moderate-energy crack, on a reasonable basis, were isolated, physically remote, or restrained by protective measures, from essential systems or components, no further dynamic hazards analysis was performed. 3.6.1.3.2 Protection Mechanisms 3.6.1.3.2.1 General The plant layout 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 within the plant, there is a minimal effect on other essential systems or components which are required for post-accident safe shutdown of the plant or to mitigate the consequences of the failure. The effects associated with a particular high-energy break or moderate-energy crack must be mechanistically consistent with the failure. Thus, actual pipe dimensions, piping layouts, material properties, and equipment arrangements are considered in defining the specific measures for protection against actual pipe movement and other associated consequences of postulated failures. Protection against the dynamic effects of pipe failures is provided in the form of pipe whip restraints, barriers, equipment shields, and physical separation of piping, equipment, and instrumentation. The precise method chosen depends largely upon considerations such as accessibility, maintenance, and proximity to other pipes. SEPARATION - The plant arrangement provides separation, to the extent practicable, between redundant safety systems (including their auxiliaries and support systems) in order to prevent loss of safety function as a result of hazards different from those for which the system is required to function, as well as for the specific event for which the system is required to be functional. Separation between redundant safety systems, with their related auxiliary supporting features, therefore, was the basic protective measure incorporated in the design to protect against the dynamic effects of postulated pipe failures. In general, layout of the facility followed a multistep process to ensure adequate separation.

a. Safety-related systems were located away from high-energy piping, where practicable.

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b. Redundant (e.g., "A" and "B" trains) safety systems were located in separate compartments.
c. As necessary, specific components were enclosed to retain the redundancy required for those systems that must function as a consequence of specific piping failure.
d. Drainage systems were reviewed to assure their adequacy for flooding prevention.

BARRIERS, SHIELDS, and ENCLOSURES - Protection requirements were met through the protection afforded by the walls, floors, columns, abutments, and foundations, in many cases. Where adequate protection did not already exist due to separation, additional barriers, deflectors, or shields were provided to meet the functional protection requirements. Some of the barriers utilized for protection against pipe whip inside the containment are the following: The secondary shield wall serves as a barrier between the reactor coolant loops and the containment liner. In addition, the refueling cavity walls, operating floor, and secondary shield walls minimize the possibility of an accident, which may occur in any one reactor coolant loop from affecting another reactor coolant loop or the containment liner. That portion of the steam and feedwater lines located within the containment was routed behind barriers which separate these lines from all reactor coolant piping. The barriers described above will withstand loadings caused by jet forces and pipe whip impact forces. Further discussion of barriers and shields is provided in Section 3.6.2.4. PIPING RESTRAINT PROTECTION - Measures for protection against pipe whip, as a result of high-energy pipe breaks, were provided where, following a single break, the unrestrained pipe movement of either end of the ruptured pipe could damage, to an unacceptable level, any structure, system, or component required to place the plant in a post-accident safe shutdown condition or mitigate the consequences of the rupture. The design criteria for and description of restraints are given in Section 3.6.2.3. 3.6.1.3.3 Specific Protection Considerations

a. Nonessential systems and system components are not required for the post-accident safe shutdown of the reactor, nor are they required for the limitation of the offsite release in the 3.6-8 Rev. 14

WOLF CREEK event of a pipe rupture. However, while none of this equipment is needed during or following a pipe break event, pipe whip protection is considered where a high-energy nonessential system or component failure could initiate a pipe break event in an essential system or component, or another nonessential system, whose failure could affect an essential system.

b. High-energy containment penetrations are subject to special protection mechanisms. As shown in Figure 3.6-1, isolation restraints are located as close as practical to the containment isolation valves associated with these penetrations. These restraints are provided in order to maintain the operability of the isolation valves and the integrity of the penetration due to a break either upstream or downstream of the penetration and outside the respective isolation restraints.
c. Instrumentation which is required to function following a pipe rupture is protected.
d. High-energy fluid system piping restraints and protective measures are designed so that a postulated break in one piping system cannot, in turn, lead to a rupture of other nearby piping system or components, if the secondary rupture would result in consequences that would be considered unacceptable for the initial postulated break.
e. For any postulated LOCA, the structural integrity of the containment structure is maintained.
f. The escape of steam, water, combustible or corrosive fluids, gases, and heat in the event of a pipe rupture will not preclude:
1. Subsequent access to any areas, as required, to cope with the postulated pipe rupture
2. Habitability of the control room
3. The ability of essential instrumentation, electric power supplies, components, and controls to perform their safety function to the extent necessary to mitigate the consequences of the pipe rupture and achieve and maintain post-accident safe shutdown.

3.6-9 Rev. 14

WOLF CREEK 3.6.2 DETERMINATION OF BREAK LOCATIONS AND DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING This section describes: the design bases for locating postulated breaks/cracks in high-energy/moderate-energy piping inside and outside of 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; restraint design; and protective assembly design. 3.6.2.1 Criteria Used to Define High/Moderate-Energy Break/Crack Locations and Configurations Except for the reactor coolant loop (RCL), NRC Branch Technical Position (BTP) MEB 3-1 was used as the basis of the criteria for the postulation of high-energy pipe breaks. Specific moderate-energy pipe crack locations were not ascertained and, therefore, they were assumed to occur at any location, except as noted in Section 3.6.2.1.2.4. A postulated 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 (i.e., a guillotine break) or as development of a sudden longitudinal, uncontrolled crack (i.e., a longitudinal split) and is postulated for a high-energy fluid system only. For moderate-energy fluid systems, pipe failures are confined to postulation of controlled cracks in piping. These cracks affect the surrounding environmental conditions only, and do not result in whipping of the cracked piping. WCGS complies with either Rev. 0 or Rev. 2 of MEB 3-1. MEB 3-1, Rev. 0 criteria was used in the initial design and Rev. 2 criteria may be used for subsequent analysis. MEB 3-1, Rev. 2 criteria is included in Table 3.6-2. The allowable stress limits for MEB 3-1, Rev. 0 is based on ASME Section III, 1974 Edition and Rev. 2 is based on 1986 Edition. However, for Class 2 and 3 pipe stress analysis, allowable from ASME Section III, Edition 1974 can be used instead of Edition 1986. 3.6.2.1.1 High-Energy Break Locations With the exception of those portions of the piping identified in Section 3.6.2.1.le, breaks were postulated only in high-energy piping at the following locations:

a. ASME B&PV Code, Section III - Class 1 Piping
1. In the pressurizer surgeline, there are a limited number of locations which are more susceptible to failure by virtue of stress or fatigue than the remainder of the system.

Breaks are eliminated from RCS primary loops. The elimination of these breaks is the result of the application of leak-before-break (LBB) technology (References 16, 17, and 18) allowed by the revised GDC-4. (Reference 15) The discrete break locations and orientations in the surge line are derived on the basis of stress and fatigue analysis. 3.6-10 Rev. 13

WOLF CREEK The postulated break locations for the pressurizer surge line were determined with the use of a detailed ASME Code NB-3200 piping analysis together with the MEB 3-1 Rev. 2, June 1987 break criteria (see Reference 7). The Surge line intermediate break locations were deleted (see Reference 20). The original design basis criteria for the reactor coolant loop (Reference 1) postulated eleven pipe break locations. Eight of these pipe break locations have subsequently been eliminated from the WCGS structural design basis as a result of the application of LBB technology. The detailed fracture mechanics techniques used in this evaluation are discussed in References 16, 17, and 18. Application of LBB allow the elimination of the dynamic effects of pipe rupture for these eight locations. To provide the high margins of safety required by GDC-4, the nonmechanistic pipe rupture design basis is maintatined for containment design and ECCS analyses, and the postulated pipe ruptures are retained for electrical and mechanical equipment environmental qualification. 3.6-11 Rev. 13

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2. Pipe breaks are postulated to occur in the following locations in Class 1 piping runs or branch runs outside the primary reactor coolant loops as follows:

(a) The terminal ends of the piping or branch run. (b) Any intermediate locations between the terminal ends where stresses, calculated using equations (12) and (13) of the ASME B&PV Code, Section III, Subsection NB, exceed 2.4 Sm, where Sm is the design stress intensity, as given in the ASME B&PV Code, and the stress range calculated, using equation (10) of the ASME B&PV Code, exceeds 2.4 Sm. (c) Any intermediate locations between terminal ends where the cumulative usage factor, derived from the piping fatigue analysis, under the loadings associated with the OBE and operational plant conditions, exceeds 0.1. (d) If the stresses and usage factor do not exceed the limits in (b) and (c), intermediate breaks are postulated at points of maximum stresses calculated by using Equation 10 of subarticle NB-3653, ASME B&PV Code, Section III. A complete discussion of the reactor coolant loop break location is provided in Reference 1.

b. ASME B&PV Code, Section III - Class 2 and 3 Piping Within Protective Structures 3.6-12 Rev. 13

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1. Breaks are postulated to occur at terminal ends, including:

(a) Piping-pressure vessel or equipment nozzle intersection (b) High-energy/moderate-energy boundary (c) Pipe to anchor intersection (d) A branch intersection point was not considered a terminal end if: 1) the branch and the main piping systems were modeled in the same static, dynamic, and thermal analyses, 2) the intersection is not rigidly constrained to the building structure, or 3) the branch and main run are of comparable size and fixity (i.e., the nominal size of the branch is at least one-half of that of the main).

2. At intermediate locations between terminal ends, where the maximum stress ranges as calculated by the sum of equations (9) and (10) in Subarticle NC-3652 of the ASME B&PV Code, Section III considering normal and upset plant conditions (i.e., sustained loads, occasional loads, and thermal expansion) including an OBE event, exceed 0.8 (1.2Sh + SA) based on 1974 ASME code or 0.8 (1.8 Sh + SA) based on 1986 ASME code, where Sh and SA are the allowable stress at maximum hot temperature and allowable stress range for thermal expansion, respectively, for Class 2 and 3 piping, as defined in Subarticle NC-3600 of the ASME B&PV Code, Section III.

3.(a)In piping systems where the stresses were lower than the limits in 2. above, a minimum of two intermediate break locations were postulated solely on the basis of highest calculated stress levels. This location may be a pipe to valve weld, pipe to fitting weld, or near clamped support attachment point. Where the piping consisted of a straight run and was shorter than 10 pipe diameters in length with no fittings, welded attachments, or valves, a minimum of one location was chosen, based on the highest stress. (b)However, Branch Technical Position MEB 3-1, Revision 2, issued in 1987 no longer mentions arbitrary intermediate pipe ruptures as described in 3(a) above. Piping stress analyses performed subsequent to the issuance of MEB 3-1 in 1987, do not require arbitrary intermediate break location if the stresses were lower than the limits in 2 above.

c. ASME B&PV Code, Section III - Class 2 and 3 Piping Not Enclosed Within Protective Structures No Class 2 or 3 high-energy piping is located outside the protective structures.
d. Non-Nuclear Piping (i.e., not ASME Section III Class 1, 2, or 3) 3.6-13 Rev. 18

WOLF CREEK Breaks in the seismically analyzed high energy non-nuclear piping were postulated at the following locations in each run or branch run:

1. Terminal ends of the run*
2. At all intermediate fittings (e.g., elbows,* tees, reducers, welded attachments, and valves)

Breaks in non-nuclear seismically analyzed high energy piping subsequent to adopting BTP MEB 3-1 Rev. 2 are postulated according to the criteria for ASME Section III Class 2 & 3 piping as described in subsection 3.6.2.1.1.b.2 Leakage cracks in nonseismic Category I piping are postulated in worse case locations.

e. High-Energy Piping in Containment Penetration Areas The portion of the containment penetration area piping defined above, extending from the outside of the inboard isolation restraint to the outside of the outboard isolation restraint, shall be considered and hereafter referred to as the "no break zone" (NBZ).
        "No break zone" boundaries are shown on Figure 3.6-1.

Breaks were not postulated in this area because stresses did not exceed those specified in Section 3.6.2.1.1.b.2. The maximum stress in the "no break zone," except within the isolation restraints, did not exceed 1.8Sh per equation (9), Subarticle NC-3652 of ASME Section III when subjected to the combined loadings of internal pressure, deadweight, and postulated pipe break beyond the "no break zone." The maximum stress within the isolation restraints in the "no break zone" is limited such that no plastic hinge will form in this region. The number of circumferential and longitudinal piping welds and branch connections was minimized.** Welded attachments for pipe supports or other purposes to these portions of piping were avoided except where detailed stress analyses could be performed to demonstrate compliance with the limits of Section 3.6.2.1.1.

  • With one clarification: On approximately 2.67 feet of pipe on FB-081-HBD-2" and 0.5 feet of pipe on FB-093-HBD-3" between the 8-inch auxiliary steam header and the closed high energy/ moderate energy boundary valves on these lines, breaks were not postulated. It was judged that the runs were short enough to prevent guillotine breaks and that any breaks that did occur would be in the 8-inch auxiliary steam header. Breaks in the 8-inch header were postulated and evaluated in the vicinity of the connections for lines 081 and 093.
    • All four main steam isolation valves were relocated 2 feet south from their original locations to improve accessibility and maintainability (Ref. DCP 9952). To accomplish this, a 2 foot section of the main steam pipe spool was 3.6-14 Rev. 30

WOLF CREEK added to the north of each isolation valve and the pipe spool south of the isolation valve was cut short by 2 feet. As a result, four additional longitudinal pipe welds and four circumferential field welds were introduced where 100% volumetric examination were performed and the No Break Zone stress limits are satisfied. The introduction of these new welds is therefore acceptable. When required for isolation valve operability, structural integrity, or the containment integrity, whip restraints capable of resisting torsional and bending moments produced by a postulated pipe break either upstream or downstream of the "no break zone" were located reasonably close to the isolation valves or penetration. These restraints do not prevent the access required to conduct inservice inspection of the welds within the restraints specified in Section XI of the ASME Code. Inservice examinations completed during each inspection interval provide 100-percent volumetric examination of circumferential and longitudinal pipe welds within the boundary of the "no break zone", with the exception of small piping socket welds which undergo 100-percent surface examination during each inspection interval or as required per the risk-informed process for piping as outlined in EPRI report 1006937, Rev. 0-A. See Section 6.6 for further discussion of inservice inspection. Terminal end breaks were not postulated on the main steam, main feedwater, and steam generator blowdown piping at the flued heads inside the containment. The "no break zone" is considered to extend up to and including the pipe to flued head weld inside containment, therefore, the terminal end location falls within the "no break zone" boundary. Inservice examinations, described in Section 6.6, commensurate with the "no break zone" are performed on the main steam, feedwater, and steam generator blowdown piping inside the containment up to the nearest pipe whip restraint. For postulated breaks beyond the first whip restraint, the stress limit (1.8Sh) given in Section 3.6.2.1.1.e. may be exceeded for the portion of piping from the first pipe whip restraint up to and including the pipe to flued head weld; however, the integrity for this portion of piping is verified. The restraints outside the containment on the main steam, main feedwater, and steam generator blowdown lines were located as close as possible to the containment to accommodate the design for the auxiliary building steam tunnel and minimize stresses. The length of the steam tunnel, the location of 5-way restraints in the north wall of the auxiliary building, and the location of isolation restraints just below the floor penetrations, for connecting piping routed to other areas of the auxiliary building, resulted in low stresses considering: 3.6-15 Rev. 30

WOLF CREEK

1. Seismic differential building movements
2. Space requirements for safety valves, isolation valves, flued heads, and other piping and components
3. Minimum space for maintenance
4. Maximum accessibility for inservice inspections performed every inspection interval 3.6.2.1.2 Types of Breaks/Cracks Postulated 3.6.2.1.2.1 ASME Section III - Class 1 Reactor Coolant Loop Piping - High-Energy The types of breaks postulated in the ASME Section III, Class 1 primary reactor coolant loop are discussed in Reference 1.

3.6.2.1.2.2 ASME Section III Piping Other Than Reactor Coolant Loop Piping - High-Energy The following types of breaks were postulated to occur at the locations determined in accordance with Section 3.6.2.1.1.

a. Breaks were not postulated in piping where nominal diameter is 1 inch or less.
b. At terminal ends, only circumferential breaks were postulated.
c. At intermediate locations where both the stress and usage factors were less than the limits of Section 3.6.2.1.1, only circumferential breaks were postulated.
d. At intermediate locations where the stress and/or usage factor exceeded the limits of Section 3.6.2.1.1, only circumferential breaks were postulated in piping less than 4-inch nominal pipe diameter but greater than the size exemption stated in a. above. In piping 4 inches and larger, circumferential and longitudinal breaks were postulated. However, if the longitudinal stress range was at least 1.5 times the circumferential stress range, only circumferential breaks were postulated. Similarly, if the circumferential stress range was at least 1.5 times the longitudinal stress range, only longitudinal breaks were postulated.

3.6-16 Rev. 13

WOLF CREEK 3.6.2.1.2.3 Non-Nuclear Piping - High-Energy For non-nuclear piping, the following combination of breaks was evaluated:

a. Circumferential breaks in piping larger than 1 inch
b. Longitudinal breaks in piping 4 inches and larger, except at terminal ends.

3.6.2.1.2.4 ASME Section III and Non-Nuclear Piping - Moderate - Energy Through-wall leakage cracks were postulated in moderate-energy piping larger than 1 inch located within, or outside and adjacent to, protective structures, except as noted in the following:

a. Through-wall leakage cracks were not postulated in those portions of piping between containment isolation valves, since this piping meets the requirements of ASME Code, Section III, Subarticle NE-1120 and is designed so that the maximum stress range does not exceed 0.4 (1.2Sh +

SA).

b. Through-wall leakage cracks were not postulated in moderate-energy fluid system piping located in the same area in which a break in high-energy fluid system piping was postulated, provided that such cracks would not result in more limiting environmental conditions than the high-energy pipe break.
c. Through-wall leakage cracks were not postulated in ASME Code, Section III, Class 2 or 3 piping and stress analyzed non-nuclear seismic Category I class piping, provided that the maximum stress range in the piping, as calculated by the sum of EQN(9) and EQN(10) in Subarticle NC-3652 of the ASME Code, Section III, considering normal and upset plant conditions, was less than 0.4 (1.2Sh +

SA)

d. Cracks were not postulated when a review of the piping layout and plant arrangement drawings showed that the effects of through-wall leakage cracks at any location in the piping designed to seismic or nonseismic standards were isolated or physically remote from structures, systems, and components required for post-accident safe shutdown.

3.6-17 Rev. 14

WOLF CREEK Cracks were postulated to occur individually at locations that resulted in the maximum effects from fluid spraying and flooding, with the consequent hazards or environmental conditions. Flooding effects were determined on the basis of a conservatively estimated time period required to effect corrective actions. Further discussion of flooding effects is provided in Appendix 3B. 3.6.2.1.3 Break/Crack Configuration 3.6.2.1.3.1 High-Energy Break Configuration The ends of a circumferentially ruptured pipe were assumed to be displaced laterally by a distance equal to or greater than one pipe diameter until and unless one end was restrained in the lateral direction. Movement was assumed to be in the direction of the jet reaction initially, and total path controlled by the piping geometry. The orientation of a longitudinal break, except when otherwise justified by a detailed stress analysis, was considered to cause piping movement normal to the plane of the piping system. The flow area of such a break was equal to the cross-sectional flow area of the pipe. Longitudinal breaks were assumed to be oriented (but not concurrently) at two diametrically opposed points on the piping circumference. Longitudinal and circumferential breaks were not postulated concurrently. 3.6.2.1.3.2 Moderate-Energy Crack Configuration Moderate-energy crack openings were assumed to be a circular orifice of cross-sectional flow area equal to that of a rectangle one-half the pipe inside diameter in length and one-half pipe wall thickness in width. 3.6.2.2 Analytical Methods to Define Forcing Functions and Response Models 3.6.2.2.1 Forcing Functions for Pipe Whip and Jet Impingement To determine the forcing function, the fluid conditions at the upstream source and at the break exit dictates the analytical approach and approximations that are used. For most applications, one of the following situations exist:

a. Superheated or saturated steam
b. Saturated or subcooled water
c. Cold water (non-flashing) 3.6-18 Rev. 0

WOLF CREEK The following three sections describe simplified models that take into account the fluid conditions. Where more complex analysis is warranted, such as for the main steam line, a RELAP4 analysis can be performed, as described in Section 3.6.2.2.1.4. For a discussion of the jet thrust forcing functions from reactor coolant loop breaks, see Section 3.6.2.2.1.5. 3.6.2.2.1.1 Superheated or Saturated Steam Break Analysis For superheated or saturated steam, steady state thrust forces are calculated from the ideal gas relationship. This relationship has been calculated using Fanno lines, assuming homogeneous flow for superheated steam, in Reference 5, Figure 2-1. When the fluid expands into the wet region, it is treated as having a specific heat ratio of 1.1. Whether the specific heat ratio is 1.1 or 1.3, the values of Figure 2-1 of Reference 5 are used. The initial value of the thrust is Po Ae, where Po is the source pressure in psia and Ae is the exit area in square inches. If the steady state thrust at initial source conditions is higher than Po Ae, no transient time is calculated, and the steady state thrust is assumed for the entire time frame. Where significant friction results in steady state thrusts below Po Ae, Po Ae is applied for the initial transient, and the steady state thrust is applied for the remainder of the time frame. The unsteady state forces due to time-dependent wave and blowdown force during the initial stages persist for several wave propagations. From Reference 8, time is approximated as time to empty the initial contents of the piping at an average flowrate. For choked flow: 2o A eL 2oL 2oL tss = = = ( 144 W + Wf ) W 144 i + f W G i + Gf A e A e k + 1 W 2 K 1 W G Gi = 144 i = Coo Gf = 144 Af = G f Gmax Ae k + 1  ; e max 3.6-19 Rev. 1

WOLF CREEK Where: tss = time to reach steady state, sec Wi = initial flowrate, lbm/sec Wf = final flowrate, lbm/sec Ae = break area, square inches L = length of pipe from break to source, ft Gi = initial mass flowrate per square foot, lbm/sec-ft2 Gf = final mass flowrate per square foot, lbm/sec-ft2 o = source density, lbm/ft3 Co = source sonic velocity, ft/sec k = effective specific heat ratio Gmax = maximum mass flowrate per square foot, lb/sec-ft2 For jet impingement forces, the Moody expansion model is coupled with the Reference 5 steady state thrust to determine jet pressure. For pressure/temperature (P/T) analysis, the blowdown rate is based on steady state flow and is determined from Figure 14 of the ASME steam tables (Ref. 9), or calculated using the perfect gas law. This analysis method is based on a converging nozzle at the entrance to the pipe. If a flow restriction is included, it is assumed that a shock wave exists immediately downstream, and the resultant force is lower than as calculated above (see Figure 2-2 of Reference 5). 3.6.2.2.1.2 Saturated or Subcooled Water Break Analysis For subcooled or saturated water, steady state thrust forces are calculated, using the Henry/Fauske model for frictionless flow. As with steam, the initial value of the thrust is Po Ae. However, since frictionless flow is used, the steady state thrust always exceeds Po Ae and the steady state thrust is applied for the entire time frame, except where upstream restrictions are presented as noted in 3.6.1.1.k. 3.6-20 Rev. 0

WOLF CREEK For jet impingement forces, the Moody expansion model defined in Reference 5 is coupled with the steady state thrust to determine jet pressure. 3.6.2.2.1.3 Cold Water Break Analysis For cold water, steady state thrust is calculated, using Reference 5, Equation 7, coupled with the frictional effects, as demonstrated below: F 2 - 2 (Pa/Po) PoAe = 1 + f (L/D) where: F = steady state thrust, lbf Po = source pressure, psia Ae = break area, in2 Pa = ambient pressure, psia f = Darcy's friction factor L/D = equivalent length of a resistance in pipe diameters The initial value of the thrust is Po Ae. If the steady state thrust at initial source conditions is higher than Po Ae, no transient time is calculated, and the steady state thrust is assumed for the entire time frame. Where significant friction results in steady state thrusts below Po Ae, Po Ae is applied for the initial transient, and the steady thrust is applied for the remainder of the time frame. The unsteady state forces due to time-dependent wave and blowdown forces during the initial stages persist for several wave propagations. From Reference 8, time is approximated as: 1 1 tss = C L 199(1 - Vi/Vss) (Vi/Vss) ln 1 + Vi/Vss o 2 o (Po - Pa) ( o) (2) (32.2) (144) Vi = (144) (32.2) (Po - Pa)  ; Vss = Co 1 + f(L / D) 3.6-21 Rev. 1

WOLF CREEK tss time to reach steady state, sec L length of pipe from break to source or upstream restriction, ft Yo specific volume, ft 3 /lbm Vi initial velocity, ft/sec Vss steady state velocity, ft/sec Jet impingement forces are calculated using the relations of Section 2.3 of Reference 5, assuming a 10 degree expansion throughout the entire jet expansion. For flooding analysis, the blowdown rate is based on a basis derivation of the Bernoulli Theorem. The blowdown rate is: W = Ae P0-Pa 144 Z;Zo J2 e 144 ( 1 + f(L/D) Po - 1 + f(L/D) g where: We steady state blowdown rate, ft3/sec Ze-Zo may be neglected when it is positive, for conservatism. 1+ f(L/D) 3.6.2.2.1.4 RELAP4 Analysis RELAP4 (Ref. 10) is a computer program developed primarily to describe the thermal-hydraulic transient behavior of water-cooled nuclear reactors subjected to a loss of coolant. This code was used to describe transients resulting from breaks in both main steam and feedwater lines. For the main steam lines, breaks inside and outside the containment were postulated. For the feedwater lines, only breaks outside the containment were considered. Both types of breaks, i.e., double-ended guillotine and slot breaks, were analyzed. For the calculation of the loading history, resulting from the above breaks on the piping elbows, the approach suggested in Ref. 8 and 11 was followed. For this purpose, the fluid properties calculated by RELAP4 were utilized. 3.6-22 Rev. 29

WOLF CREEK 3.6.2.2.1.5 Time Functions of Jet Thrust Force on Ruptured and Intact Reactor Coolant Loop Piping In order to determine the thrust and reactive force loads to be applied to the reactor coolant loop during the postulated LOCA, it is necessary to have a detailed description of the hydraulic transient. Hydraulic forcing functions are calculated for the ruptured and intact reactor coolant loops as a result of a postulated LOCA. These forces result from the transient flow and pressure histories in the reactor coolant system. The calculation is performed in two steps. The first step is to calculate the transient pressure, mass flow rates, 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 (e.g., elbows) in the reactor coolant loops. The hydraulic model represents the behavior of the coolant fluid within the entire RCS. Key parameters calculated by the hydraulic model are pressure, mass flow rate, 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 LOCA, the pressure and momentum flux terms are dominant. The inertia and gravitational terms are taken into account in evaluation of the local fluid conditions in the hydraulic model. The blowdown hydraulic analysis is required to provide 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. The MULTIFLEX Code (Ref. 2) was developed with a capability to provide this information. The MULTIFLEX Code 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 homogeneous fluid in thermal equilibrium. The ability to treat multiple flow branches and a large number of mesh points gives the MULTIFLEX Code the required flexibility to represent the various flow passages within the primary RCS. The system geometry is represented by a network of one-dimensional flow passages. 3.6-23 Rev. 0

WOLF CREEK The THRUST computer program was developed to compute the transient (blowdown) hydraulic loads resulting from a LOCA. The blowdown hydraulic loads on primary loop components are computed from the equation. m. 2 F = 144A (P - 14.7) + 144 g A c m Which includes both the static and dynamic effects. The symbols and units are: F = force, lbf A = aperture area, ft2 P = system pressure, psia m = mass flow rate, lbm/sec 

        =   density, lbm/ft3 gc   =   gravitational constant (32.174 ft-lbm/lbf-sec2)

Am = mass flow area, ft2 In the model used 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 fully described by: 1) blowdown hydraulic information and 2) the orientation of the streamlines 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, total y force, and total z force. These thrust forces serve as input to the piping/restraint dynamic analysis. The THRUST Code (which uses MULTIFLEX results as input) calculates forces exactly the same way as the STHRUST Code (which uses SATAN- IV [Ref. 3] results as input). 3.6-24 Rev. 1

WOLF CREEK The STHRUST Code is described in Reference 4. 3.6.2.2.2 Response Models 3.6.2.2.2.1 Response Model for Other Than Reactor Coolant Loop The dynamic analysis of system piping is described in Section 3.9(B). 3.6.2.2.2.2 Response Model of the Reactor Coolant Loop Piping, Equipment Supports, and Pipe Whip Restraints The dynamic analysis of the reactor coolant loop piping for LOCA loadings is described in Section 3.9(N) and Reference 1. 3.6.2.3 Dynamic Analysis Methods to Verify Integrity and Operability 3.6.2.3.1 Dynamic Analysis Methods to Verify Integrity and Operability for Other Than Reactor Coolant Loop The analytical methods of Reference 5, with the amplifying clarifications and assumptions discussed in Section 3.6.2.2, were used to determine the jet impingement effects and loading effects applicable to components and systems resulting from postulated pipe breaks and cracks. This information was then used in the protection evaluation described in this section, Section 3.6.2.3. This section describes the design of restraints used to protect the essential systems, components, and equipment from the effects of pipe whip. 3.6.2.3.2 Dynamic Analysis Methods to Verify Integrity and Operability for the Reactor Coolant Loop 3.6.2.3.2.1 General A LOCA is assumed to occur for a branch line break down to the restraint of the second normally open automatic isolation valve (Case II in Figure 3.6-2) on outgoing lines1 and down to and including the second check valve (Case III in Figure 3.6-2) on incoming lines normally with flow. A pipe break beyond the 1It is assumed that motion of the unsupported line containing the isolation valves could cause failure of the operators of both valves to function. 3.6-25 Rev. 0

WOLF CREEK restraint or second check valve will not result in an uncontrolled loss of reactor coolant if either of the two valves in the line closes. Accordingly, both of the automatic isolation valves are suitably protected and restrained as close to the valves as possible so that a pipe break beyond the restraint will not jeopardize the integrity and operability of the valves. Further, periodic testing capability of the valves to perform their intended function is essential. This criterion takes credit for only one of the two valves performing its intended function. For normally closed isolation or incoming check valves (Cases I and IV in Figure 3.6-2), a LOCA is assumed to occur for pipe breaks on the reactor side of the valve. Branch lines connected to the reactor coolant loop (RCL) are defined as "large" for the purpose of this criteria and as having an inside diameter greater than 4 inches up to the largest connecting line, generally the pressurizer surge line. Rupture of these lines results in a rapid blowdown from the RCL, and protection is basically provided by the accumulators and the low head safety injection pumps (residual heat removal pumps). Branch lines connected to the RCL are defined as "small" if they have an inside diameter equal to or less than 4 inches. This size is such that emergency core cooling system analyses, using realistic assumptions, show that no clad damage is expected for a break area of up to 12.5 square inches, corresponding to 4-inch inside diameter piping. Engineered safety features are provided for core cooling and boration, pressure reduction, and activity confinement in the event of a LOCA or steam or feedwater line break accident to ensure that the public is protected in accordance with 10 CFR 50.67 guidelines. These safety systems are designed to provide protection for a reactor coolant system pipe rupture of a size up to and including a double-ended severance of a reactor coolant loop. In order to assure the continued integrity of the vital components and the engineered safety systems, consideration is given to the consequential effects of the pipe break itself to the extent that:

a. The minimum performance capabilities of the engineered safety systems are not reduced below that required to protect against the postulated break.

3.6-26 Rev. 34

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b. The containment leaktightness is not decreased below the design value, if the break leads to a LOCA.2
c. Propagation of damage is limited in type and/or degree to the extent that:
1. A pipe break which is not a LOCA will not cause a LOCA or steam or feedwater line break.
2. An RCS pipe break will not cause a steam or feedwater system pipe break, and vice versa.

3.6.2.3.2.2 Large Reactor Coolant System Piping Propagation of damage resulting from the rupture of a reactor coolant loop is permitted to occur but must not exceed the design basis for calculating containment and subcompartment pressures, loop hydraulic forces, reactor internals reaction loads, primary equipment support loads, or emergency core cooling system performance. Large branch line piping, as defined in Section 3.6.2.3.2.1, is restrained to meet the following criteria, in addition to items a through c of Section 3.6.2.3.2.1, for a pipe break resulting in a LOCA.

a. Propagation of the break to the unaffected loops is prevented to assure the delivery capacity of the accumulators and low head pumps.
b. Propagation of the break in the affected loop is permitted to occur but does not exceed 20 percent of the flow area of the line which initially ruptured. This criterion has been voluntarily applied so as not to substantially increase the severity of the LOCA.

3.6.2.3.2.3 Small Branch Lines In the unlikely event that one of the small pressurized lines, as defined in Section 3.6.2.3.2.1, should fail and result in a LOCA, the piping is restrained or arranged to meet the following criteria in addition to items a through c of Section 3.6.2.3.2.1. 2The containment is here defined as the containment structure liner and penetrations and the steam generator shell, the steam generator steam side instrumentation connections, the steam, feedwater, blowdown, and steam generator drain pipes within the containment structure. 3.6-27 Rev. 0

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a. Break propagation is limited to the affected leg, i.e.,

propagation to the other leg of the affected loop and to the other loops is prevented.

b. Propagation of the break in the affected leg is permitted but must be limited to a total break area of 12.5 square inches (4 inches inside diameter). The exception to this case is when the initiating small break is a cold leg high head safety injection line. Further propagation is not permitted for this case.
c. Damage to the high head safety injection lines connected to the other leg of the affected loop or to the other loops is prevented.
d. Propagation of the break to a high head safety injection line connected to the affected leg is prevented if the line break results in a loss of core cooling capability due to a spilling injection line.

3.6.2.3.2.4 Design and Verification of Adequacy of RCL Components and Supports The methods described below are used in the Westinghouse design and verification of the adequacy of primary reactor coolant loop (RCL) components and supports. It is emphasized that these methods are used only to determine jet impingement loads on RCL components and supports. The design basis postulated pipe rupture locations for the reactor coolant loop piping are determined, using the criteria given in Section 3.6.2. These design basis ruptures are used here as the rupture locations for consideration of jet impingement effects on primary equipment and supports. The dynamic analysis, as discussed in Section 3.6.2.2.2, is used to determine maximum piping displacements at each design basis rupture location. These maximum piping displacements are used to compute the effective rupture flow area at each location. This area and rupture orientation are then used to determine the jet flow pattern and to identify any primary components and supports which are potential targets for jet impingement. The jet thrust at the point of rupture is based on the fluid pressure and temperature conditions occurring during normal (100 percent) steady state operating conditions of the plant. At the point of rupture, the jet force is equal and opposite to the jet thrust. The force of the jet is conservatively assumed to be constant throughout the jet flow distance. The subcooled jet is 3.6-28 Rev. 0

WOLF CREEK assumed to expand uniformly at a half angle of 10 degrees, from which the area of the jet at the target and the fraction of the jet intercepted by the target structure can be readily determined. The shape of the target affects the amount of momentum change in the jet and thus affects the impingement force on the target. The target shape factor is used to account for target shapes which do not deflect the flow 90 degrees away from the jet axis. The method used to compute the jet impingement load on a target is one of the following:

a. The dynamic effect of jet impingement on the target structure is evaluated by applying a step load whose magnitude is given by:

Fj = KoPoAmBRS where: Fj = jet impingement load on target, lbf Ko = dimensionless jet thrust coefficient based on initial fluid conditions in the broken loop Po = initial system pressure, lb/in.2 AmB = calculated maximum break flow area, in.2 R = fraction of jet intercepted by target S = target shape factor Discharge flow areas for limited flow area circumferential breaks are obtained from reactor coolant loop analyses performed to determine the axial and lateral displacements of the broken ends as a function of time. AmB is the maximum break flow area occurring during the transient, and is calculated as the total surface area through which the fluid must pass to emerge from the broken pipe. Using geometrical formulations, this surface area is determined to be a function of the pipe separation (axial and transverse) and the dimensions of the pipe (inside and outside diameter). 3.6-29 Rev. 0

WOLF CREEK If simplified static analysis is performed instead of a dynamic analysis, the above jet load (Fj) 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 2.0. This factor assumes that the target can be represented as essentially a one degree of freedom system, and the impingement force is conservatively applied as a step load. The calculation of the dimensionless jet thrust coefficient and break flow area is discussed in Section 3.6.2.5.

b. The dynamic effect of jet impingement is evaluated by applying the following time-dependent load to the target structure.

Fj = K P AB RS where the system pressure P is a function of time; the jet thrust coefficient K is evaluated as a function of system pressure and enthalpy; and the break flow area AB is a function of time. 3.6.2.3.3 Types of Restraints 3.6.2.3.3.1 Restraints Other Than Reactor Coolant Loop Restraints To satisfy varying requirements of available space, permissible pipe deflection, and equipment operability, the restraints are generally located as close as possible to the postulated breaks, in order to limit whipping of the failed pipe in a direction away from the break. Where necessary, guides were used to prevent uncontrolled motion of the pipe in a direction other than that caused by the primary motion generated by the blowdown force. A typical example is shown in Figure 3.6-4. Restraints identified as isolation restraints are located to protect an essential portion of a piping system from postulated leaks either upstream or downstream of the protected area. These restraints limit pipe motion in all directions. A typical example of an isolation restraint is shown in Figure 3.6-5. The restraints are of three design types. These include two types of large gap restraints and one type of close gap restraint. 3.6-30 Rev. 0

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a. Large gap restraints In order to account for dynamic and gap effects, restraints utilizing either energy absorbing honeycomb material (EAHM) or stainless steel upset rods are used.

Large gap restraints were employed where the resulting piping motion may be tolerated without causing a rupture elsewhere in the piping system. EAHM restraints are the large gap restraints most frequently used. This type of restraint consists of substructures which are allowed to behave plastically within acceptable ductility ratios and have an energy dissipating material (stainless steel honeycomb) between the pipe and the substructure. A typical example of an EAHM restraint is shown in Figure 3.6-6. The upset rod restraint prevents uncontrolled pipe motion by using its capacity to undergo considerable plastic deformation, thereby absorbing the kinetic energy of the whipping pipe. A typical example of a rod restraint is shown in Figure 3.6-7.

b. Close Gap Restraints Close gap restraints were installed where large piping motions permitted by large gap restraints could not be tolerated. The primary purpose of close gap restraints is to limit pipe stresses in areas which are designated as no-break zones. A typical example of a close gap restraint is shown in Figure 3.6-8.

3.6.2.3.3.2 Restraints for Reactor Coolant Loop Pipe restraints and locations are discussed in Section 5.4.14. 3.6.2.3.4 Analytical Methods 3.6.2.3.4.1 Restraints Other Than Reactor Coolant Loop Restraints

a. Location of restraints For purposes of locating restraints, the collapse moment of the pipe is determined in the following manner:

Mp = kSyS for stainless steel pipe 3.6-31 Rev. 0

WOLF CREEK where: k = 2.5 Sy = yield stress at pipe operating temperature S = elastic modulus of pipe 3.14 3.14 Ro - Ri Mp = 1.07 Su 0.14 for carbon steel pipe (Ref. 13) Ro where: Su = ultimate stress at pipe operating temperature Ro = outside radius of pipe Ri = inside radius of pipe Restraints (with the exception of isolation restraints) are located as close to the postulated break as practicable. Restraints located so that a collapse moment will not form in the pipe required no further evaluation because the pipe whip is limited by the rigidity of the piping. If, due to physical limitations, restraints were located so that collapse mechanisms in the pipe may form, the consequences of the whipping pipe and the jet impingement effect were further investigated. Guides were provided where necessary to control pipe motion.

b. Design of Restraints One of the following three methods, depending upon the type of restraint, was used to determine the response of the piping/restraint/supporting structure to the jet thrust developed by the postulated pipe rupture. These methods are energy balance, jet thrust with dynamic load factor of 2, and dynamic analysis using a lumped parameter model. All methods address the following effects, as appropriate:
1. Stiffness characteristics of the piping system, restraint system, major components, and supporting walls and structures
2. Transient forcing functions acting on the piping system, and jet thrusts on structures 3.6-32 Rev. 0

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3. Elastic and inelastic deformation of piping and/or restraints
4. Insulation thickness
5. Seismic and thermal movements (for determination of clearance values)

The energy balance method of analysis is discussed in Section 3.0 of Reference 5. This method is the primary method used for large gap restraints as described below: Forcing Function - obtained from Reference 5. Resistance Response of Piping System - the resistance of piping system (load-deflection response) was achieved by a static analysis (by inputting the force at the postulated pipe break location). The displacement obtained for a corresponding force gave the force-deflection response of the piping system in the elastic range. A perfectly plastic response for the piping system was assumed when the intensified stress (due to the stress intensification factor of the fitting) at the first elbow beyond the pipe whip restraint reached yield stress of the material. Restraint Response: EAHM Restraints - This is basically an energy dissipating material which is supported by a substructure. This substructure is allowed to behave plastically within acceptable ductility ratios as defined in BC-TOP-9A. The kinetic energy of the impacting pipe is absorbed by the collapse of the crushable honeycomb core. The substructure, in turn, is designed to absorb the sudden, impulsive dynamic loading created by the crushing EAM (Energy Absorbing Material). The properties as a function of cell size and web size of the honeycomb core were obtained by test by the manufacturer for the specific material used. The EAHM restraint resistance Rr was determined from equation (1) below: Y FY = Rr (Y - Yg) + Rp 2p + Rp (Y - Yp) (1) Y FY - Rp Y - 2p Rr = Y - Yg 3.6-33 Rev. 0

WOLF CREEK where: F = pipe jet thrust Y = total pipe displacement Yg = gap between pipe and restraint RP = maximum pipe resistance YP = elastic displacement of pipe where: Rr < AmPc and t > (Y - Yg) (2) where: Am = cross sectional area of energy absorbing honeycomb material Pc = crushing strength of the energy absorbing honeycomb material 

         =   allowable deformation in percent of total thickness (t) t  =   total thickness of the energy absorbing honeycomb material For a suitable value of Pc, Am is determined from equation (2).

Where crushable honeycomb energy absorbing material is used, the material will not experience a deflection in excess of that which is defined by the horizontal portion of its load deflection curve as determined by test, under designed loads. Upset Rod Restraints The analytical procedures used to size the upset rod restraint are based on an energy balance method similar to that used for the EAHM restraint design. These are illustrated using a simplified example. Assuming the jet thrust force as constant with time, the strain energy absorbed by the rod in deflecting from its initial configuration to the maximum allowable strain (50% ultimate strain) is equal to the work generated by jet thrust force. In equation form this becomes: 2nd2 W = F (Yg + Lee) = Le(u ) 4 where: W = total strain energy Le = effective length of the restraint determined by test 3.6-34 Rev. 0

WOLF CREEK n = number of upset rods per restraint d = diameter of rod u = strain energy per unit volume conservatively idealized to represent the material properties e = maximum strain allowed Assuming a plastic collision between the pipe and the restraint and ignoring the energy absorbed by the pipe (in this example) the rod can be sized by solving for (d). Substructures for both the EAHM and upset rod restraints are allowed to behave plastically throughout a postulated pipe break event. Ductility ratios are in accordance with BC-TOP-9A. A ductility ratio of three is used for anchor bolts and welded studs, based on test data. Design methods are in accordance with Sections 3.8.3 and 3.8.4. For some close-gap restraints, the simplified jet thrust with load factor method was used. Briefly, the force on the restraint was taken as equal to the jet thrust (pressure x area x thrust coefficient) multiplied by a dynamic load factor. This load factor was conservatively assumed to be 2, the largest possible for a restraint which was virtually in contact with the pipe. (If the clearance between pipe and restraint was large enough to permit the whipping pipe to attain significant velocity before contacting the restraint, thus causing impact effects, other analytical methods were used.) As an alternate to the energy balance method of analysis, a dynamic analysis of the isolation restraints using a lumped parameter model is employed. The model is shown in Figure 3.6-9. To calculate the isolation restraint design loads, resulting from a postulated piping failure, a dynamic analysis is performed. PIPE RUP (see 3.9(B).7, Ref.

4) was used to perform this analysis. The isolation restraint is designed such that in the event of a postulated piping failure, inside or outside containment, the "no break zone" criteria per Section 3.6.2.1.le is met.

3.6.2.3.4.2 Reactor Coolant Loop Restraints As described in Section 3.9(N), the forces associated with the rupture of reactor piping systems are considered in combination 3.6-35 Rev. 0

WOLF CREEK with normal operating loads and earthquake loads for the design of supports and restraints in order to assure the continued integrity of vital components and engineered safety features. The stress limits for reactor coolant piping and supports are discussed in Section 3.9(N). 3.6.2.4 Protective Assembly Design Criteria 3.6.2.4.1 Jet Impingement Barriers and Shields Barriers and shields, which may be either of steel or concrete construction, are provided to protect essential equipment 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 and floors and structures specifically designed to provide protection from postulated pipe breaks. Barrier and shield design is based on the methods of Reference 5, Section 3.0, and the elastic- plastic methods for dynamic analysis included in Reference 14. Design criteria and loading combinations are in accordance with Sections 3.8.3 and 3.8.4. 3.6.2.4.2 Auxiliary Guardpipes The use of guardpipes has been minimized by plant arrangement and routing of high-energy piping. Where they are used, guardpipes are designed to withstand all environmental, jet impingement, and impact effects of postulated breaks of the enclosed pipe. Design criteria, loading combinations, and methods of analysis are similar to those for barriers and shields described in Section 3.6.2.4.1. 3.6.2.5 Material to be Submitted for the Operating License Review 3.6.2.5.1 Piping Systems Other Than Reactor Coolant Loop Pipe break locations were obtained in accordance with the criteria of Section 3.6.2.1. Pipe crack locations were postulated to occur at any location, as stated in Section 3.6.2.1. High-energy piping with break locations identified are provided in isometric drawings, Figure 3.6-1. Break types are also shown (i.e., circumferential or longitudinal). The stress results which were utilized to determine the break types and locations are given in Table 3.6-3. If there are changes in the pipe stress analysis, the stress tables will be updated only when those changes affect 3.6-36 Rev. 0

WOLF CREEK the break locations shown on the figures previously mentioned. Associated stress nodes are shown in Figure 3.6-1. High-energy pipe break effects analysis is discussed room-by-room in Table 3.6-4. Each piping isometric (Figure 3.6-1) references the appropriate sheet of Table 3.6-4 by which the effects analysis is discussed for all breaks on that isometric drawing. Table 3.6-4 discussion includes pipe whip, jet impingement, flooding, room pressurization, temperature effect, and humidity effects. Moderate-energy piping crack locations are defined in Section 3.6.2.1.2.4. Evaluation of the effects of moderate-energy cracks is discussed in Appendix 3B. The augmented inservice inspection plan is discussed in Section 6.6.8. Pipe whip restraints are designed in accordance with Section 3.6.2.3. Restraint locations and orientation for each high-energy break are shown in Figure 3.6-1. Barriers and shields are designed in accordance with the criteria of Section 3.6.2.4. Jet thrust and impingement forces were determined in accordance with Section 3.6.2.2. Thrust forces for each break are presented in Figure 3.6-1. 3.6.2.5.2 Reactor Coolant Loop

a. Figure 3.6-3 identifies the design basis break locations and orientations for the reactor coolant loops.

The primary plus secondary stress intensity ranges and the fatigue cumulative usage factors at the design break locations specified in Reference 1 are given in Table 3.6-5 for a reference fatigue analysis. The reference analysis was prepared to be applicable for many plants. It uses seismic umbrella moments which are higher than those used in Reference 1, in which the primary stress is equal to the limits of equation 9 in NB-3650 (Section III of the ASME Boiler and Pressure Vessel Code) at many locations in the system, where in Reference 1 one location was at the limit. Therefore, the results of the reference analysis may differ slightly from Reference 1, but the philosophy and conclusions of Reference 1 are valid. There are no other locations in the model used in the reference fatigue analysis, consistent with Reference 1, where the stress intensity ranges and/or usage factors exceed the criteria of 2.4 Sm and 0.2, respectively. 3.6-37 Rev. 0

WOLF CREEK Actual plant moments for WCGS are also given in Table 3.6-5 at the design basis break locations. As noted in Table 3.6-5, the reference analysis thermal moments are exceeded by the WCGS moments at three locations. The change in the usage factors was insignificant. Thus, there are no other locations in the reactor coolant loop, consistent with Reference 1, where additional breaks are required to be postulated.

b. Pipe whip restraints associated with the main reactor coolant loop are described in Section 5.4.14.
c. The methods and analysis procedures used to determine jet impingement loads associated with the rupture of the reactor coolant loop piping are discussed in Section 3.6.2.3. These loads are used to determine the adequacy of the primary equipment and supports.
d. Design loading combinations and applicable criteria for ASME Class 1 components and supports are provided in Section 3.9(N).1.4. Pipe rupture loads include not only the jet thrust forces acting on the piping but also jet impingement loads on the primary equipment and supports.
e. The original design basis criteria for the reactor coolant loop (Reference 1) postulated eleven pipe break locations.

Eight of these pipe break locations have subsequently been eliminated from the WCGS structural design basis as a result of the application of LBB technology. The detailed fracture mechanics techniques used in this evaluation are discussed in References 16, 17, and 18. Application of LBB allow the elimination of the dynamic effects of pipe rupture for these eight locations. To provide the high margins of safety required by GDC-4, the nonmechanistic pipe rupture design basis is maintained for containment design, ECCS analyses, and the postulated pipe ruptures are retained for electrical and mechanical equipment environmental qualification. 3.

6.3 REFERENCES

1. "Pipe Breaks for the LOCA Analysis of the Westinghouse Primary Coolant Loop," WCAP-8082-P-A (Proprietary) and WCAP-8172-A (Non-Proprietary), January 1975.
2. Takeuchi, K., et al., "MULTIFLEX-A FORTRAN-IV Computer Program for Analyzing Thermal-Hydraulic-Structure System Dynamics," WCAP-8708-P-A", Volumes 1 and 2 (Proprietary) and WCAP-8709-A, Volumes 1 and 2 (Non-Proprietary), February 1976.
3. Bordelon, F. M., "A Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant (SATAN-IV Digital Code)," WCAP-7750, August 1971.
4. "Documentation of Selected Westinghouse Structural Analysis Computer Codes," WCAP-8252, Revision l, May 1977.
5. "Design for Pipe Break Effects," BN-TOP-2, Revision 2, Bechtel Power Corporation, May 1974.
6. NRC Branch Technical Position ASB 3-1, "Protection Against Postulated Piping Failures in Fluid Systems Outside Containment," Revision 0 November 24, 1975, SPLB 3-1, Revision 2, October 1990.

3.6-38 Rev. 13

WOLF CREEK

7. NRC Branch Technical Position MEB 3-1, "Postulated Break and Leakage Locations in Fluid System Piping Outside Containment," Revision 0 November 24, 1975, and Revision 2, June 1987.
8. Moody, F. J., "Fluid Reaction and Impingement Loads,"

presented at the ASCE Specialty Conference, Chicago, Ill., December 1973.

9. American Society of Mechanical Engineers, "Thermodynamic and Transport Properties of Steam Comprising Tables and Charts for Steam and Water," 1967 Edition.
10. Aerojet Nuclear Company, "RELAP4/MOD5 - A Computer Program for Transient Thermal-Hydraulic Analysis of Nuclear Reactors and Related Systems," Volumes I-III, ANCR-NUREG-1335, September 1976.
11. Moody, F. J., "Time-Dependent Pipe Forces Caused by Blowdown and Flow Stoppage," ASME Paper No. 73-FE-23, June 1973.
12. "Subcompartment Pressure Analyses," BN-TOP-4, Revision 1, Bechtel Power Corporation, October 1977.
13. Gerber, T. L., "Plastic Deformation of Piping Due to Pipe-Whip Loading," ASME Paper No. 74-NE-1, June 1974.
14. Biggs, J. M., Introduction to Structural Dynamics, McGraw-Hill Book Company, New York, 1964.
15. Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures -- Final Rule (Broad Scope), 52 FR 41288, October 27, 1987.
16. WCAP-10691, "Technical Basis for Eliminating Large Primary Loop Pipe Rupture as a Structural Design Basis for Callaway and Wolf Creek Plants," October, 1984.
17. WCAP-9558, "Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing A Postulated Circumferential Through-Wall Crack," Revision 2, June, 1981.
18. WCAP-10456, "The Effects of Thermal Aging on the Structural Integrity of Cast Stainless Steel Piping for Westinghouse NSSS," November, 1983.
19. NUREG-0881, Supplement No. 5, "Safety Evaluation Report related to the operation of Wolf Creek Generating Station, Unit No. 1," USNRC, March, 1985.
20. Structural Evaluation of the Wolf Creek and Callaway Pressurizer Surge Lines, Considering the Effects of Thermal Stratification, WCAP-12893, Rev. 0, March 1991.

3.6-39 Rev. 28

WOLF CREEK

21. Structural Analysis of the Reactor Coolant Loop for Standard Nuclear Unit Power Plant System, WCAP-9728.
22. Generic Letter 87-11, Relaxation in Arbitrary Intermediate Pipe Rupture Requirements, June 19, 1987.
23. WCAP-9728, Volume 1, Revision 5, Supplement 1, Wolf Creek Cold Leg/Reactor Vessel Nozzle Safe-end Thickness Nonconformance Evaluation, September 2010 (calculation BB-S-019).
24. WCAP-17592-P, Revision 0, Wolf Creek Stress Report Addendum for the Reactor Vessel Inlet Nozzle (calculation BB-S-018).

3.6-40 Rev. 28

WOL' CAIIK TABLE 3.1*1 IA,ITV AILATID IVITIMI AND HIGH AND MODIAATI INIAGV IVSTIMS LOCATIO IN I SA, ETV I HIGHt 

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WOLI' CIIIEK TAibi 3.8*1 (IHIIT 21 1 LOCATIO IN HIGH/ IAI'ITY 1 lSI' I AILATIO MOOEAATE iSTAUCTUA&Si ENI!AGY I 1- .. 2 Ill 

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' 01 LY 'lAST I** it . I' CHEMICA II I I* IUII"MINT OAAINI AO I ION ro AU X ILIA I **  : I CONTAIN MIN II. It' I U ' _It t! WI M I t!A !ITO A Alii I I"AOCISS SAMI" lNG ' II! lATHING All!

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  • No Looeled In dleeel eulldlne II Hlgt\ *r*aeure aaaoolatad with reaotor ooolant oraaaura tloundary L.ocataCI in a aataty*ralate41 atea Rev. 6
  • WOLF CREEK TABLE 3.6-2 DESIGN COMPARISON TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.46, REVISION 0, DATED MAY 1973, TITLED "PROTECTION OF PIPE WHIP INSIDE CONTAINMENT" The basis for compliance to Regulatory Guide 1.46 is the implementation of NRC Branch Technical Position (BTP) MEB 3-1, NRC BTP ASB 3-1, WCAP-8082-P-A, and WCAP-8172-A. The following provides a summary of the compliance with MEB 3-1 and ASB 3-1.

BTP ASB 3-1 Position WCGS Compliance B.1 Plant Arrangement B.1 Complies. See Section 3.6.1.3 Protection of essential systems and components against postulated piping failures in high or moderate energy fluid systems that operate during normal plant conditions and that are located outside of containment should be provided by one of the following plant arrangement considerations: B.1.a. Plant arrangements should separate fluid system piping from essential systems and components. Separation should be achieved by plant physical layouts that provide sufficient distances between essential systems and components and fluid system piping such that the effects of any postulated piping failure therein (e.g., pipe whip, jet impingement, and the environmental conditions resulting from the escape of contained fluids as appropriate to high- or moderate-energy fluid system piping) cannot impair the integrity or operability of essential systems and components. B.1.b Fluid system piping or portions thereof not satisfying the provisions of B.1.a should be enclosed within structures or compartments designed to protect nearby essential systems and components. Alternatively, essential systems and components may be enclosed within structures or compartments designed to withstand the effects of postulated piping failures in nearby fluid systems. B.1.c Plant arrangements or system features that do not satisfy the provisions of either B.1.a or B.1.b should be limited to those for which the above provisions are impractical because of the stage of design or construction of the plant; because the plant design is based upon that of an earlier plant accepted by the staff as a base plant under the Commission's standardization and replication policy; or for other substantive reasons such as particular design features of the fluid systems. Such cases may arise, for example, (1) at inter-connections between fluid systems and essential systems and components, or (2) in fluid systems having dual functions (i.e., required to operate during normal plant conditions as well as to shut down the reactor). In these cases, redundant design features that are separated or otherwise protected from postulated piping failures, or Rev. 0

WOLF CREEK TABLE 3.6-2 (Sheet 2) BTP ASB 3-1 Position WCGS Compliance additional protection, should be provided so that the effects of postulated piping failures are shown by the analyses and guidelines of B.3 to be acceptable. Additional protection may be provided by restraints and barriers or by designing or testing essential systems and components to withstand the effects asso-ciated with postulated piping failures. B.2 Design Features B.2.a Essential systems and components should be designed to B.2.a Complies, as described in Table meet the seismic design requirements of Regulatory Guide 3.2-3. 1.29. B.2.b Protective structures or compartments, fluid system piping restraints, and other protective measures should be designed in accordance with the following: (1) Protective structures or compartments needed to B.2.b.(1) Complies. See Sections 3.8.3 and implement B.1 should be designed to seismic Category 3.8.4 for loading combinations. I requirements. The protective structures should be designed to withstand the effects of a postulated piping failure (i.e., pipe whip, jet impingement, pressurization of compartments, water spray, and flooding, as appropriate) in combination with loadings associated with the operating basis earthquake and safe shutdown earthquake within the respective design load limits for structures. Piping restraints, if used, may be taken into account to limit effects of the postulated piping failure. (2) High-energy fluid system piping restraints and B.2.b(2) Complies. See Section 3.6.1.1i. protective measures should be designed such that a postulated break in one pipe cannot, in turn, lead to rupture of other nearby pipes or components if the secondary rupture could result in consequences that would be considered unacceptable for the initial postulated break. An unrestrained whipping pipe should be considered capable of (a) rupturing impacted pipes of smaller nominal pipe sizes and (b) developing through-wall leakage cracks in larger nominal pipe sizes with thinner wall thickness, except where experimental or analytical data for the expected range of impact energies demonstrate the capability to withstand the impact without failure. Rev. 0

WOLF CREEK TABLE 3.6-2 (Sheet 3) BTP ASB 3-1 Position WCGS Compliance B.2.c Fluid system piping in containment penetration areas should B.2.c All high energy fluid system and meet the following design provisions: selected moderate energy fluid piping in the containment penetration areas com-ply with the following criteria: (1) Portions of fluid system piping between the required B.2.c.(1) High-energy (H-E) piping systems restraints located inside and outside containment associated with the steam tunnel, beyond the isolation valves of single barrier con- i.e., main steam, feedwater, and tainment structures (including any rigid connection steam generator blowdown, are pro-to the containment penetration) that connect, on a vided with isolation restraints continuous or intermittent basis, to the reactor which protect the penetration piping coolant pressure boundary, or the steam and feedwater in the steam tunnel. For further systems of PWR plants, should be designed to the discussion of the main steam, feed-stress limits specified in B.1.b or B.2.b of Branch water and steam generator blowdown Technical Position (BTP) MEB 3-1, attached to piping penetration areas, see Standard Review Plan 3.6.2. Section 3.6.2.1.1.e. For all other H-E piping penetrations, isolation restraints have been pro-vided reasonably close to the contain-These portions of high-energy fluid system piping ment isolation valves to protect the should be provided with pipe whip restraints that are "no break zone" piping, protect the capable of resisting bending and torsional moments integrity of the penetration, and produced by a postulated piping failure either protect the operability of the iso-upstream or downstream of the containment isolation lation valves (when present), assuming valves. The restraints should be located reasonably a rupture at the postulated intermed-close to the containment isolation valves and should iate breakpoints or terminal ends be designed to withstand the loadings resulting from a outside the regions defined as "no postulated piping failure beyond these portions of break zone." For further discussion piping so that neither isolation valve operability nor see Section 36.2.1.1.e. the leaktight integrity of the containment will be impaired. (2) Portions of fluid system piping between the required B.2.c.(2) Not applicable to WCGS. restraints located inside and outside containment beyond the isolation valves of dual barrier contain-ment structures should also meet the design provisions of B.2.c.(1). In addition, those portions of piping that pass through the containment annulus, and whose postulated failure could affect the leaktight integrity of the containment structure or result in pressurization of the containment annulus beyond the design limits should be provided with an enclosing protective structure. For the purpose of establishing the design parameters (i.e., pressure, temperature) of the enclosing protective structure, a full flow area opening should be assumed in that portion of piping within the enclosing structure and vent areas should be taken into account, if provided, in the enclosing structure. Where guard pipes for individual process Rev. 0

WOLF CREEK TABLE 3.6-2 (Sheet 4) BTP ASB 3-1 Position WCGS Compliance pipes are used as an enclosing protective structure, such guard pipes should be de-signed to meet the requirements specified in B.1.b(6) of BTP MEB 3-1. (3) Terminal ends of the piping runs extending beyond B.2.c.(3) Terminal ends of H-E piping fall these portions of high-energy fluid system piping within the "no break zone" boundary; should be considered to originate at a point therefore, no terminal end breaks adjacent to the required pipe whip restraints are postulated except to calculate located inside and outside containment. the design load for the isolation restraint. (4) Piping classification as required by Regulatory B.2.c.(4) Complies. Guide 1.26 should be maintained without change until beyond the outboard restraint. If the restraint is located at the isolation valve, a classification change at the valve interface is acceptable. B.2.d. Inservice examination and related design provisions should be in accordance with the following: (1) The protective measures, structures, and guard B.2.d.(1) Complies. pipes should not prevent the access required to conduct the inservice examinations specified in the ASME Boiler and Pressure Vessel Code, Section XI, Division 1, "Rules for Inspection and Testing of Components in Light-Water Cooled Plants." (2) For those portions of fluid system piping identified B.2.d.(2) Complies, with the exception of of small in B.2.c, includes piping running from inboard to piping socket welds which will undergo outboard restraints in containment penetration areas, 100 percent surface examination during the extent of inservice examinations completed during each inspection interval or the extent of each inspection interval (IWA-2400, ASME Code, inservice examinations completed during Section XI) should provide 100 percent volumetric each inspection interval are as required examination of circumferential and longitudinal pipe per the risk-informed process for piping as welds within the boundary of these portions of outlined in EPRI Report 1006437, Rev. 0-A. piping. See Section 6.6 and 3.6.2.1.1e. (3) For those portions of fluid system piping enclosed in B.2.d.(3) WCGS has no guard pipes guard pipes, inspection ports should be provided in located in the penetration areas. guard pipes to permit the required examination of Guard pipes utilized in other circumferential pipe welds. Inspection ports should areas comply with this position. not be located in that portion of the guard pipe passing through the annulus of dual barrier containment structures. (4) The areas subject to examination should be defined in B.2.d.(4) Complies. See Section 6.6. accordance with Examination Categories C-F and C-G for Class 2 piping welds in Tables IWC-2520. Rev. 19

WOLF CREEK TABLE 3.6-2 (Sheet 5) BTP ASB 3-1 Position WCGS Compliance B.3 Analyses and Effects of Postulated Piping Failures B.3.a To show that the plant arrangement and design features B.3.a Complies. See Section 3.6.1.1d, 3.6.1.1k, provide the necessary protection of essential systems and Table 3.6-4. and components, piping failures should be postulated in accordance with BTP MEB 3-1, attached to Standard Review Plant 3.6.2. In applying the provisions of BTP MEB 3-1, each longitudinal or circumferential break in high-energy fluid system piping or leakage crack in moderate-energy fluid system piping should be considered separately as a single postulated initial event occurring during normal plant conditions. An analysis should be made of the effects of each such event, taking into account the provisions of BTP MEB 3-1 and of the system and component operability considerations of B.3.b below. The effects of each postulated piping failure should be shown to result in offsite consequences within the guidelines of 10 CFR Part 100 and to meet the provisions of B.3.c and d below. B.3.b In analyzing the effects of postulated piping failures, the following assumptions should be made with regard to the operability of systems and components: (1) Offsite power should be assumed to be unavailable if B.3.6.(1) Complies. See Section 3.6.1.1e. a trip of the turbine-generator system or reactor protection system is a direct consequence of the postulated piping failure. (2) A single active component failure should be assumed B.3.6.(2) Complies. See Section 3.6.1.1f. in systems used to mitigate consequences of the postulated piping failure and to shut down the reactor, except as noted in B.3.b(3) below. 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. (3) Where the postulated piping failure is assumed to B.3.b.(3) Complies. Section 3.61.1g defines a occur in one of two or more redundant trains of a train to include those systems which dual-purpose moderate-energy essential system, i.e., support its function. Note that the one required to operate during normal plant conditions criteria is also applied to single-as well as to shut down the reactor and mitigate the purpose and high energy systems, since consequences of the piping failure, single failures the same quality, design, construction, of components in the other train or trains of that and inspection standards are used. system only need not be assumed provided the system is designed to seismic Category I standards, is powered The only applicable H-E piping system from both offsite and onsite sources, and is con- is CVCS charging. structed, operated, and inspected to quality assurance, testing, and inservice inspection standards appropriate for nuclear safety systems. Examples of Rev. 0

WOLF CREEK TABLE 3.6-2 (Sheet 6) BTP ASB 3-1 Position WCGS Compliance systems that may, in some plant designs, qualify as dual-purpose essential systems are service water systems, component cooling systems, and residual heat removal systems. (4) All available systems, including those actuated by B.3.b.(4) Complies. See Section 3.6.1.1h. operator actions, may be employed to mitigate the consequences of a postulated piping failure. In judging the availability of systems, account should be taken of the postulated failure and its direct consequences such as unit trip and loss of offsite power, and of the assumed single active component failure and its direct consequences. The feasi-bility of carrying out operator actions should be judged on the basis of ample time and adequate access to equipment being available for the proposed actions. B.3.c. The effects of a postulated piping failure, including B.3.c Complies. environmental conditions resulting from the escape of contained fluids, should not preclude habitability of the control room or access to surrounding areas important to the safe control of reactor operations needed to cope with the consequences of the piping failure. B.3.d A postulated failure of piping not designed to seismic B.3.d Complies. See Section 3B.2.1. Category I standards should not result in any loss of capability of essential systems and components to withstand the further effects of any single active component failure and still perform all functions required to shut down the reactor and mitigate the consequences of the postulated piping failure. Rev. 0

WOLF CREEK TABLE 3.6-2 (Sheet 7) BTP MEB 3-1 Position WCGS Compliance B.1 High-Energy Fluid System Piping B.1.a Fluid Systems Separated from Essential Systems and B.1.a Complies. See Section 3.6.1.3.2 Components For the purpose of satisfying the separation provisions of plant arrangement as specified in B.1.a of Branch Technical Position BTP ASB 3-1, a review of the piping layout and plant arrangement drawings should clearly show the effects of postulated piping breaks at any location are isolated or physically remote from essential systems and components. At the designer's option, break locations as determined from 1.c and 1.d of this position may be assumed for this purpose. B.1.b Fluid System Piping In Containment Penetration Areas B.1.b Complies. Breaks need not be postulated in those portions of piping identified in B.2.c of BTP ASB 3-1 provided they meet the requirements of the ASME Code, Section III, Subarticle NE-1120 and the following additional design requirements: (1) The following design stress and fatigue limits B.1.b(1)(a)-(d) There is no Class 1 piping should not be exceeded. in containment penetration areas on WCGS. For ASME Code, Section III, Class 1 Piping (a) The maximum stress range should not exceed 2.4S m (b) The maximum stress range between any two load sets (including the zero load set) should be calculated by Eq. (10) in Paragraph NB-3653, ASME Code, Section III, for normal and upset plant conditions and an operating basis earthquake (OBE) event transient. If the calculated maximum stress range of Eq. (10) exceeds the limit of B.1.b(1)(a) but is not greater than 3Sm, the limit of B.1.b(1)(c) should be met. If the calculated maximum stress range of Eq. (10) exceeds 3Sm, the stress ranges calculated by both Eq. (12) and Eq. (13) in Paragraph NB-3653 should meet the limit of B.1.b(1)(a) and the limit of B.1.b(1)(c). Rev. 0

WOLF CREEK TABLE 3.6-2 (Sheet 8) BTP MEB 3-1 Position WCGS Compliance (c) The cumulative usage factor should be less than 0.1 if consideration of fatigue limits is required according to B.1.b(1)(b). (d) The maximum stress, as calculated by Eq. (9) in Paragraph NB-3652 under the loadings resulting from a postulated piping failure beyond these portions of piping should not exceed 2.25Smexcept that following a failure outside containment, the pipe between the outboard isolation valve and the first restraint may be permitted higher stresses provided a plastic hinge is not formed and operability of the valves with such stresses is assured in accordance with the requirements specified in SRP 3.9.3. Primary loads include those which are deflection limited by whip restraints. For ASME Code, Section III, Class 2 Piping (e) The maximum stress ranges as calculated by the sum B.1.b(1)(e) Complies. of Eq. (9) and (10) in Paragraph NC-3652, ASME Code, Section III, considering normal and upset plant conditions (i.e., sustained loads, occasional loads, and thermal expansion) and an OBE event should not exceed 0.8(1.2Sh + SA). (f) The maximum stress, as calculated by Eq. (9) in B.1.b.(1)(f) Complies. For further discussion Paragraph NC-3652 under the loadings resulting see Section 3.6.2.1.1.e. from a postulated piping failure of fluid system piping beyond these portions of piping should not exceed 1.8Sh. Primary loads include those which are deflection limited by whip restraints. The exceptions per-mitted in (d) may also be applied provided that when the piping between the outboard isolation valve and the restraint is constructed in accordance with the Power Piping Code ANSI B31.1 (see ASB 3-1 B.2.c[4]), the piping shall either be of seamless construction with full radiography of all circumferential welds, or all longitudinal and circumferential welds shall be fully radiographed. (2) Welded attachments, for pipe supports or other B.1.b.(2) Welded attachments to these portions purposes, to these portions of piping should be avoided of the piping are minimized. Attach-except where detailed stress analyses, or tests, are ments for welded pipe supports are performed to demonstrate compliance with the limits of reviewed separately for local stresses B.1.b(1). and the limits of B.1.b(1) are met. Rev. 0

WOLF CREEK TABLE 3.6-2 (Sheet 9) BTP MEB 3-1 Position WCGS Compliance Stress Analysis is performed to demon-strate that Eq. (9) and (10) stresses do not exceed 0.8 (1.2 Sh + SA). (3) The number of circumferential and longitudinal piping B.1.b.(3) Complies. Guard pipes are not used in welds and branch connections should be minimized. the containment penetration areas. Where guard pipes are used, the enclosed portion of fluid system piping should be seamless construction unless specific access provisions are made to permit inservice volumetric examination of the longitudinal welds. (4) The length of these portions of piping should be B.1.b.(4) See compliance statement to BTP ASB reduced to the minimum length practical. 3-1 position B.2.c.(1). (5) The design of pipe anchors or restraints (e.g., B.1.b.(5) All high-energy containment penetra-connections to containment penetrations and pipe whip tions are flued integrally-forged restraints) should not require welding directly to the piped fittings. Pipe whip restraints outer surface of the piping (e.g., flued integrally- do not require welding directly to the forged pipe fittings may be used) except where such outer surface of the piping, except welds are 100 percent volumetrically examinable in where 100-percent volumetric examin-service and a detailed stress analysis is performed to ation and a review for local stresses demonstrate compliance with the limits of B.1.b(1). are performed. The main steam and main feedwater lines outside the containment have flued integrally-forged pipe fitting whip restraints. (6) Guard pipes provided for those portions identified in B.1.b.(6) WCGS has no guard pipes located in B.2.c(2) of BTP ASB 3-1 should be constructed in the containment penetration areas. accordance with the rules of Class MC, Subsection NE of the ASME Code, Section III, where the guard pipe is part of the containment boundary. In addition, the entire guard pipe assembly should be designed to meet the following requirements and tests: (a) The design pressure and temperature should not be less than the maximum operating pressure and temperature of the enclosed pipe under normal plant conditions. (b) The design stress limits of Paragraph NE-3131(c) should not be exceeded under the loading associated with containment design pressure and temperature in combination with the safe shutdown earthquake. (c) Guard pipe assemblies should be subjected to a single pressure test at a pressure not less than its design pressure. Rev. 0

WOLF CREEK TABLE 3.6-2 (Sheet 10) BTP MEB 3-1 Position WCGS Compliance B.1.c. Fluid Systems Enclosed Within Protective Structures (1) With the exceptions of those portions of piping identi-fied in B.1.b, breaks in Class 2 and 3 piping (ASME Code, Section III) should be postulated at the following loca-tions in those portions of each piping and branch run within a protective structure or compartment designed to satisfy the plant arrangement provisions of B.1.b or B.1.c of BTP ASB 3-1. (a) At terminal ends of the run if located within the B.1.c.(1)(a) Complies. See Section 3.6.2.1.1b. protective structure. Terminal ends are identified and compliance statement to BTP in ASB 3-1 B.2.c.(3). ASB 3-1 position B.2.c.(3). (b) At intermediate locations selected by one of the B.1.c.(1)(b) Complies. Intermediate breaks are following criteria: selected solely on the basis of highest calculated stress (i.e., (i) At each pipe fitting (e.g., elbow, tee, breaks may not be separated by a cross, flange, and nonstandard fitting), change in direction of the piping welded attachment, and valve. Where the run or located at a weld). piping contains no fittings, welded attach-ments, or valves, at one location at each extreme of the piping within the protective structure. A terminal end, as determined by B.1.c(1)(a), may be considered as one of these extremes. (ii) At each location where the stresses 1) exceed 0.8(1.2Sh + SA) but at not less than two separated locations chosen on the basis of highest stress, 2) Where the piping consists of a straight run without fittings, welded attachments, and valves, and all stresses are below 0.8(1.2Sh + SA), a mini-mum of one location chosen on the basis of highest stress. (2) Breaks in non-nuclear class piping should be postu- B.1.c.(2) Break postulation in non-nuclear class lated at the following locations in each piping or piping complies. See Section 3.6.2.1.1d. branch run: Non-nuclear, high-energy pipes are either refrained from impacting or af-(a) At terminal ends of the run if located within fecting the separating structure or the protective structure. the separating structure are designed for full effects. (b) At each intermediate pipe fitting, welded attachment, and valve. If a structure separates a high energy line from an essential component, that separating structure should be designed to withstand the consequences of the pipe Rev. 0

WOLF CREEK TABLE 3.6-2 (Sheet 11) BTP MEB 3-1 Position WCGS Compliance break in the high energy line which produces the greatest effect at the structure irrespective of the fact that the above criteria might not require such a break location to be postulated. (3) Applicable to (1) and (2) above: B.1.c.(3) Separating structures are analyzed to If a structure separates a high energy line from an withstand the dynamic effects of the essential component, that separating structure should postulated pipe breaks as defined be designed to withstand the consequences of the pipe in B.1.c.(1) and B.1.c.(2) above. break in the high energy line which produces the greatest effect at the structure irrespective of the fact that the above criteria might not require such a break location to be postulated. B.1.d Fluid Systems Not Enclosed Within Protective Structures (1) With the exceptions of those portions of piping iden- B.1.d.(1) No Class 2 or 3 high-energy piping is tified in B.1.b, breaks in Class 2 and 3 piping (ASME located outside of the protective Code, Section III) should be postulated at the structures. following locations in those portions of each piping and branch run routed outside of, but alongside, above, or below, a protective structure or compart-ment containing essential systems and components and designed to satisfy the plant arrangement provisions of B.1.b or B.1.c or BTP ASB 3-1. Such piping should be considered as located adjacent to a protective structure if the distance between the piping and structure is insufficient to preclude impairment of the integrity of the structure from the effects of a postulated piping failure assuming the piping is unrestrained. (a) At terminal ends of the run if located adjacent to the protective structure. Terminal ends are identified in ASB 3-1 B.2.c.(3). (b) At intermediate locations selected by one of the following criteria: (i) At each pipe fitting (e.g., elbow, tee, cross, flange, and nonstandard fitting), welded attachment, and valve. Where the piping contains no fittings, welded attachments, or valves, at one location at each extreme of the piping run adjacent to the protective structure. Rev. 0

WOLF CREEK TABLE 3.6-2 (Sheet 12) BTP MEB 3-1 Position WCGS Compliance (ii) At each location where the stresses 1) exceed 0.8(1.2S + SA) but at not less than h two separated locations chosen on the basis of highest stress. 2) Where the piping consists of a straight run without fittings, welded attachment, or valves, and all stresses are below 0.8(1.2S + SA), a mini-h mum of one location chosen on the basis of highest stress. (2) Breaks in non-nuclear class piping should be postulated B.1.d.(2) Complies. With one clarification: On at the following locations in each piping or branch run: approximately 2.67 feet of pipe on FB-081-HBD-2" and 0.5 feet of pipe on (a) At terminal ends of the run if located adjacent to FB-093-HBD-3" between the 8-inch aux-the protective structure. iliary steam header and the normally closed high energy/moderate energy (b) At each intermediate pipe fitting, welded boundary valves, breaks were not attachment, and valve. postulated. It was judged that the runs were short enough to prevent guillotine breaks and that any breaks that did occur would be in the 8-inch auxiliary steam header. Breaks in the 8-inch header were postulated and evaluated in the vicinity of the connections for lines 081 and 093. (3) Applicable to (1) and (2) above: B.1.d.(3) Complies. If a structure separates a high energy line from an essential component, that separating structure should be designed to withstand the consequences of the pipe break in the high energy line which produces the greatest effect at the structure irrespective of the fact that the above criteria might not require such a break location to be postulated. B.1.e. The designer should identify each piping run he has con- B.1.e. Complies. See Section 3.6.2.5. sidered to postulate the break locations required by B.1.c and B.1.d above. In complex systems such as those containing arrangements of headers and parallel piping running between headers, the designer should identify and include all such piping within a designated run in order to postulate the number of breaks required by these criteria. Rev. 0

WOLF CREEK TABLE 3.6-2 (Sheet 12a) BTP MEB 3-1 Position WCGS Compliance B.2. Moderate-Energy Fluid System Piping B.2.a. Complies. See Section 3.6.1.3 and Appendix 3B. B.2.a. Fluid Systems Separated from Essential Systems and Components For the purpose of satisfying the separation provisions of plant arrangement as specified in B.1.a of BTP ASB 3-1, a review of the piping layout and plant arrangement drawings should clearly show that the effects of through-wall leakage cracks at any location in piping designed to seismic and non-seismic standards are isolated or physically remote from essential systems and components. Rev. 0

WOLF CREEK TABLE 3.6-2 (Sheet 13) BTP MEB 3-1 Position WCGS Compliance B.2.b Fluid System Piping Between Containment Isolation Valves B.2.b. Complies. See Section 3.6.2.1.2.4. Leakage cracks need not be postulated in those portions of piping identified in B.2.c. of (BTP) ASB 3-1 provided they meet the requirements of the ASME Code, Section III, Subarticle NE-1120, and are designed such that the maximum stress range does not exceed 0.4(1.2S + SA) for ASME Code, h Section III, 1985 2 piping. B.2.c Fluid Systems Within, or Outside and Adjacent to, Protective B.2.c. See compliance statement to B.2.b above. Structures

i. Through-wall leakage cracks should be postulated in seismic Category I fluid system piping located within, or outside and adjacent to, protective structures designed to satisfy the plant arrangement provisions of B.1.b. or B.1.c of BTP ASB 3-1, except (1) where exempted by B.2.b and B.2.d, or (2) where the maximum stress range in these portions of Class 2 or 3 piping (ASME Code, Section III), or non-nuclear piping is less than 0.4(1.2S + SA). The cracks should be postulated h

to occur individually at locations that result in the maximum effects from fluid spraying and flooding, with the consequent hazards or environmental conditions developed. ii. Through-wall leakage cracks should be postulated in fluid system piping designed to non-seismic standards as necessary to satisfy B.3.d of BTP ASB 3-1. B.2.d Moderate-Energy Fluid Systems in Proximity to High-Energy B.2.d. Complies. See compliance statement to Fluid Systems B.2.b above. Cracks need not be postulated in moderate-energy fluid system piping located in an area in which a break in high-energy fluid system piping is postulated, provided such cracks would not result in more limiting environmental conditions than the high-energy piping break. Where a postulated leakage crack in the moderate-energy fluid system piping results in more limiting environmental conditions than the break in proximate high-energy fluid system piping, the provisions of B.2.c should be applied. Rev. 0

WOLF CREEK TABLE 3.6-2 (Sheet 14) BTP MEB 3-1 Position WCGS Compliance B.2.e. Fluid Systems Qualifying as High-Energy or Moderate- B.2.e. Complies. See Section 3.6.1.1a Energy Systems Through-wall leakage cracks instead of breaks may be postu-lated in the piping of those fluid systems that qualify as high-energy fluid systems for only short operational periods3 but qualify as moderate-energy fluid systems for the major operational period. B.3. Type of Breaks and Leakage Cracks in Fluid System Piping B.3.a Circumferential Pipe Breaks The following circumferential breaks should be postulated in high-energy fluid system piping at the locations specified in B.1 of this position: (1) Circumferential breaks should be postulated in fluid B.3.a.(1) Complies. See Section 3.6.2.1.2.2. system piping and branch runs exceeding a nominal pipe size of 1 inch, except where the maximum stress range1 exceeds the limits specified in B.1.c(1)(b)(ii) and B.1.d(1)(b)(ii) but the circumferential stress range is at least 1.5 times the axial stress range. Instrument lines, one inch and less nominal pipe or tubing size should meet the provisions of Regulatory Guide 1.11. (2) Where break locations are selected without the benefit B.3.a.(2) Complies. All high-energy Class 1, 2, of stress calculations, breaks should be postulated at and 3 piping is analyzed by stress the piping welds to each fitting, valve, or welded calculations. Non-nuclear class attachment. Alternatively, a single break location at high-energy piping breaks are postu-the section of maximum stress range may be selected as lated at all welds, fittings, welded determined by detailed stress analyses (e.g., finite attachments, etc. element analyses) or test on a pipe fitting. (3) Circumferential breaks should be assumed to result in B.3.a.(3) Complies. pipe severance and separation amounting to at least a one-diameter lateral displacement of the ruptured piping sections unless physically limited by piping restraints, structural members, or piping stiffness as may be demonstrated by inelastic limit analysis (e.g., a plastic hinge in the piping is not developed under loading). (4) The dynamic force of the jet discharge at the break B.3.a.(4) See Section 3.6.2.2.1. location should be based on the effective cross-sectional flow area of the pipe on a calculated fluid pressure as modified by an analytically or experi-mentally determined thrust coefficient. Limited pipe Rev. 13

WOLF CREEK TABLE 3.6-2 (Sheet 15) BTP MEB 3-1 Position WCGS Compliance displacement at the break location, line restrictions, flow limiters, positive pump-controlled flow, and the absence of energy reservoirs may be taken into account, as applicable, in the reduction of jet discharge. (5) Pipe whipping should be assumed to occur in the plane B.3.a.(5) Complies. See Section 3.6.1.1j. defined by the piping geometry and configuration, and to cause pipe movement in the direction of the jet reaction. B.3.b. Longitudinal Pipe Breaks The following longitudinal breaks should be postulated in high-energy fluid system piping at the locations of the circumferential breaks specified in B.3.a: (1) Longitudinal breaks in fluid system piping and branch B.3.b.(1) Complies. See Section 3.6.2.1.2.2. runs should be postulated in nominal pipe sizes 4-inch and larger, except where the maximum stress range1 exceeds the limits specified in B.1.c(1)(b)(ii) and B.1.d(1)(b)(ii) but the axial stress range is at least 1.5 times the circumferential stress range. (2) Longitudinal breaks need not be postulated at: B.3.b.(2) Per Section 3.6.2.1.2.2, only circum-ferential breaks are postulated at (a) Terminal ends provided the piping at the terminal terminal ends, even if a longitudinal ends contains no longitudinal pipe welds (if pipe weld is present at that point. longitudinal welds are used, the requirements of At intermediate locations, the excep-B.3.b(1) apply). tion of this position was complied with. (b) At intermediate locations where the criterion for a minimum number of break locations must be satisfied. (3) Longitudinal breaks should be assumed to result in an B.3.b.(3) Complies. See Section 3.6.2.1.3.1. axial split without pipe severance. Splits should be oriented (but not concurrently) at two diametrically-opposed points on the piping circumference such that the jet reaction causes out-of-plane bending of the piping configuration. Alternatively, a single split may be assumed at the section of highest tensile stress as determined by detailed stress analysis (e.g., finite element analysis). (4) The dynamic force of the fluid jet discharge should be B.3.b.(4) See Section 3.6.2.2.1. based on circular elliptical (2D x 1/2D) break area equal to the effective cross-sectional flow area of the pipe at the break location and on a calculated Rev. 0

WOLF CREEK TABLE 3.6-2 (Sheet 16) BTP MEB 3-1 Position WCGS Compliance fluid pressure modified by an analytically or exper-imentally determined thrust coefficient as determined for a circumferential break at the same location. Line restrictions, flow limiters, positive pump-controlled flow, and the absence of energy reservoirs may be taken into account, as applicable, in the reduction of jet discharge. (5) Piping movement should be assumed to occur in the B.3.b.(5) Complies. direction of the jet reaction unless limited by struc-tural members, piping restraints, or piping stiffness as demonstrated by inelastic limit analysis. B.3.c Through-Wall Leakage Cracks The following through-wall leakage cracks should be postu-lated in moderate-energy fluid system piping at the locations specified in B.2 of this position: (1) Cracks should be postulated in moderate-energy fluid B.3.c.(1) Complies. system piping and branch runs exceeding a nominal pipe size of 1 inch. (2) Fluid flow from a crack should be based on a circular B.3.c.(2) Complies. opening of area equal to that of a rectangle one-half pipe-diameter in length and one-half pipe wall thickness in width. (3) The flow from the crack should be assumed to result in B.3.c.(3) Complies. an environment that wets all unprotected components within the compartment, with the consequent flooding in the compartment and communicating compartments. Flooding effects should be determined on the basis of a conservatively estimated time period required to effect corrective actions. Rev. 0

WOLF CREEK TABLE 3.6-2 (Sheet 17) BTP MEB 3-1 Position (footnotes)

1. Stresses under normal and upset plant conditions, and an OBE event as calculated by Eq. (9) and (10), Para. NC-3652 of the ASME Code, Section III.
2. Select two locations with at least 10% difference in stress, or, if stresses differ by less than 10%, two locations separated by a change of direction of the pipe run.
3. An operational period is considered "short" if the fraction of time that the system operates within the pressure-temperature conditions specified for high-energy fluid systems is about 2 percent of the time that the system operates as a moderate-energy fluid system (e.g., systems such as the reactor decay heat removal system qualify as moderate-energy fluid systems; however, systems such as auxiliary feedwater systems operated during PWR reactor startup, hot standby, or shutdown qualify as high-energy fluid systems).

Rev. 0

WOLF CREEK TABLE 3.6-2a DESIGN COMPARISON TO REGULATORY POSITIONS OF BRANCH TECHNICAL POSITION MEB 3-1, REVISION 2, DATED JUNE 1987, TITLED POSTULATED RUPTURE LOCATIONS IN FLUID SYSTEM PIPING INSIDE AND OUTSIDE CONTAINMENT*

  • NOTE: Regulatory Guide 1.46, Revision 0, dated 1973, titled Protection of Pipe Whip Inside Containment, was withdrawn per 50FR9732 March 11, 1985.

BTP 3-1 Position WCGS Compliance B.1 High-Energy Fluid Systems Piping B.1.a Fluid Systems Separated From Essential Systems and Components B.1.a Complies. See Section 3.6.1.3.2. For the purpose of satisfying the separation provisions of plant arrangement as specified in B.1.a of Branch Technical Position (BTP) ASB 3-1, a review of the piping layout and plant arrangement drawings should clearly show the effects of postulated piping breaks at any location are isolated or physically remote from essential systems and components.1 At the designer's option, break locations as determined from B.1.c. of this position may be assumed for this purpose. B.1.b Fluid System Piping in Containment Penetration Areas B.1.b Complies. Arbitrary intermediate Breaks and cracks need not be postulated in those portions of breaks postulated piping from containment wall to and including the inboard or envelope all outboard isolation valves provided they meet the requirements environmental effects of the ASME Code, Section III, Subarticle NE-1120 and the from breaks. following additional design requirements: B.1.b.(1) The following design stress and fatigue limits should not be exceeded: For ASME Code, Section III, Class 1 Piping B.1.b.(1).(a) through B.1.b.(1).(c). There B.1.b.(1).(a) The maximum stress range between any two load sets is no piping in (including the zero load set) should not exceed 2.4 Containment Sm, and should be calculated2 by Eq. (10) in NB-3653, penetration area at ASME Code, Section III. WCGS If the calculated maximum stress range of Eq. (10) exceeds 2.4 Sm, the stress ranges calculated by both Eq. (12) and Eq. (13) in Paragraph NB-3653 should meet the limit of 2.4 Sm. B.1.b.(1).(b) The cumulative usage factor should be less than 0.1. B.1.b.(1).(c) The maximum stress, as calculated by Eq. (9) in NB-3652 under the loadings resulting from a postulated piping failure beyond these, portions of piping should not exceed the lesser of 2.25 Sm and 1.8 Sy, except that following a failure outside containment, the pipe between the outboard isolation valve and the first restraint may be permitted higher stresses provided a plastic hinge is not formed and operability of the valves with such stresses is assured in accordance with the requirements specified in SRP Section 3.9.3. Primary loads include those which are deflection limited by whip restraints. Rev. 13

WOLF CREEK TABLE 3.6-2a (sheet 2) BTP 3-1 Position WCGS Compliance For ASME Code, Section III, Class 1 Piping B.1.b.(1).(d) The maximum stress as calculated by the sum of Eqs. B.1.b.(1).(d) (9) and (10) in Paragraph NC-3652, ASME Code, Section Complies. III, considering those loads and conditions thereof for which level A and level B stress limits have been specified in the system's Design Specification (i.e., sustained loads, occasional loads, and thermal expansion), including an OBE event should not exceed 0.8(1.8 Sh + SA ). The Sh and SA are allowable stresses at maximum (hot) temperature and allowable stress range for thermal expansion, respectively, as defined in Article NC-3600 of the ASME Code, Section III. B.1.b.(1).(e) B.1.b.(1).(e) The maximum stress, as calculated by Eq. (9) in NC- Complies. 3653 under the loadings resulting from a postulated See Section B.1.b.1.a piping failure of fluid system piping beyond these portions of piping should not exceed the lesser of 2.25 Sh and 1.8 Sy. Primary loads include those which are deflection limited by whip restraints. The exceptions permitted in (c) above may also be applied, provided that when the piping between the outboard isolation valve and the restraint is constructed in accordance with the Power Piping Code ANSI B31.1 (see ASB 3-1 B.2.c(4)), the piping shall either be of seamless construction with full radiography of all circumferential welds, or all longitudinal and circumferential welds shall be fully radiographed. B.1.b.(2) Welded attachments, for pipe supports or other purposes, Welded attachments to to these portions of piping should be avoided, except these portions of piping where detailed stress analyses, or tests, are performed are minimized. Attach-ments for welded pipe to demonstrate compliance with the limits of B.1.b.(1). supports are reviewed separately for local stress and the limits of B.1.b.(1) are met. Stress analysis is performed to demonstrate that Eq.(9) and (10) do not exceed stress limits of B.1.b.(1).(d). B.1.b.(3) The number of circumferential and longitudinal piping B.1.b.(3) Complies. welds and branch connections should be minimized. Where Guard pipes are not guard pipes are used, the enclosed portion of fluid used in the system piping should be seamless construction and without containment circumferential welds unless specific access provisions penetration area. are made to permit inservice volumetric examination of the longitudinal and circumferential welds. B.1.b.(4) The length of these portions of piping should be reduced B.1.b.(4) Complies. to the minimum length practical. B.1.b.(5) The design of pipe anchors or restraints (e.g., connec- B.1.b.(5) All high-energy tions to containment penetrations and pipe whip containment penetrations restraints) should not require welding directly to the are flued integrally-forged piped fittings. outer surface of the piping (e.g., flued integrally Pipe whip restraints do forged pipe fittings may be used), except where such not require welding welds are 100 percent volumetrically examinable in directly to the outer service and a detailed stress analysis is performed to surface of the piping, demonstrate compliance with the limits of B.1.b.(1). except where 100 percent volumetric examinations are performed. The main steam and main feedwater lines outside containment have flued integrally-forged pipe fitting whip restraints. Rev. 13

WOLF CREEK TABLE 3.6-2a (sheet 3) BTP 3-1 Position WCGS Compliance B.1.b.(6) Guard pipes provided for those portions of piping in the B.1.b.(6) WCGS has no containment penetration areas should be constructed in guard pipes located in accordance with the rules of Class MC, Subsection NE of the contianment the ASME Code, Section III, where the guard pipe is part penetration areas of the containment boundary. In addition, the entire guard pipe assembly should be designed to meet the following requirements and tests: B.1.b.(6).(a) The design pressure and temperature should not be less than the maximum operating pressure and temperature of the enclosed pipe under normal plant conditions. B.1.b.(6).(b) The Level C stress limits in NE-3220, ASME Code, Section III, should not be exceeded under the loadings associated with containment design pressure and temperature in combination with the safe shutdown earthquake. B.1.b.(6).(c) Guard pipe assemblies should be subjected to a single pressure test at a pressure not less than its design pressure. B.1.b.(6).(d) Guard pipe assemblies should not prevent the access required to conduct the inservice examination specified in B.1.b.(7). Inspection ports, if used, should not be located in that portion of the guard pipe through the annulus of dual barrier containment structures. B.1.b.(7) A 100% volumetric inservice examination of all pipe welds B.1.b.(7) Complies, with should be conducted during each inspection interval as the exception of small defined in IWA-2400, ASME Code, Section XI. bore socket welds which undergo 100 percent surface examination, or the extent of inservice examinations completed during each inspection interval are as required per the risk informed process for piping as outlined in EPRI report 1006437, Rev. 0-A B.1.c Postulation of Pipe Breaks in Areas Other Than Containment Penetration B.1.c.(1) With the exceptions of those portions of piping B.1.c.(1) Complies.* identified in B.1.b, breaks in Class 1 piping (ASME Code, See Section Section III) should be postulated at the following 3.6.2.1.1(a)2 locations in each piping and branch run: (a) At terminal ends3. (b) At intermediate locations where the maximum stress

  • No Class 1 piping is range2 as calculated by Eq. (10) exceeds 2.4 Sm. located outside the protective structures.

(c) At intermediate locations where the cumulative usage factor exceeds 0.1. As a result of piping reanalysis due to differences between the design configuration and the as-built configuration, the highest stress or cumulative usage factor locations may be shifted; however, the initially determined intermediate break locations need not be changed unless one of the following conditions exist: (i) The dynamic effects from the new (as-built) intermediate break locations are not mitigated by the original pipe-whip restraints and jet shields. (ii) A change is required in pipe parameters such as major differences in pipe size, wall thickness, and routing. Rev. 19

WOLF CREEK TABLE 3.6-2a (sheet 4) BTP 3-1 Position WCGS Compliance B.1.c.(2) With the exceptions of those portions of piping No Class 2 or 3 high-identified in B.1.b, breaks in Class 2 and 3 piping (ASME energy piping is located Code, Section III) should be postulated at the following outside the protective locations in those portions of each piping and branch structures run: B.1.c.(2).(a) At terminal ends B.1.c.(2).(a) Complies.* See Sections 3.6.2.1. and compliance statement to BTP ASB 3-1 position B.2.c.(3). B.1.c.(2).(b) At intermediate locations selected by one of the following criteria: B.1.c.(2).(b).(i) At each pipe fitting (e.g., elbow, tee, cross, B.1.c.(2).(b).(i) flange, and nonstandard fitting), welded Complies. Intermediate attachment, and valve. Where the piping breaks are selected on the basis of high contains no fittings, welded attachments, or stresses but arbitrary valves, at one location at each extreme of the intermediate breaks are piping run adjacent to the protective structure. postulated. See Section 3.6.2.1.1.b.3. B.1.c.(2).(b).(ii) At each location where stresses calculated2 by B.1.c.(2).(b).(ii) the sum of Eqs. (9) and (10) in NC/ND-3653, ASME Complies. See Code, Section III, exceed 0.8 times the sum of Sections the stress limits given in NC/ND-3653. 3.6.2.1.1.a.2.(e) and 3.6.2.1.1.b.2. As a result of piping reanalysis, due to differences between the design configuration and the as-built configuration, the highest stress locations may be shifted; however, the initially determined intermediate break locations may be used unless a redesign of the piping resulting in a change in pipe parameters (diameter, wall thickness, routing) is required, or the dynamic effects from the new (as-built) intermediate break locations are not mitigated by the original pipe-whip restraints and jet shields. B.1.c.(3) Breaks in seismically analyzed non-ASME Class piping are B.1.c.(3) Complies. postulated according to the same requirements as for ASME See Section Class 2 and 3 piping above.4 3.6.2.1.1.a B.1.c.(4) Applicable to (1), (2), and (3) above: B.1.c.(4) Complies. Separating structures If a structure separates a high-energy line from an are analyzed to essential component, that separating structure should withstand the dynamic be designed to withstand the consequences of the pipe effects of the break in the high-energy line which produces the postulated pipe breaks greatest effect at the structure irrespective of the as defined in fact that the above criteria might not require such a B.1.c.(1) and break location to be postulated. B.1.c.(2). B.1.c.(5) Safety-related equipment must be environmentally B.1.c.(5) Complies. qualified in accordance with SRP Section 3.11. Required See Sections 3.11(B) pipe ruptures and leakage cracks (whichever controls) and 3.11(N). must be included in the design bases for environmental qualification of electrical and mechanical equipment both inside and outside the containment. B.1.d The designer should identify each piping run he has B.1.d Complies. See considered to postulate the break locations required by B.1.c Section 3.6.2.5. above. In complex systems such as those containing arrangements of headers and parallel piping running between headers, the designer should identify and include all such piping within a designated run in order to postulate the number of breaks required by these criteria. Rev. 19

WOLF CREEK TABLE 3.6-2a (sheet 5) BTP 3-1 Position WCGS Compliance B.1.e With the exceptions of those portions of piping identified in B.1.e.(1) through (3): B.1.b, leakage cracks should be postulated as follows: Leakage cracks need not be postulated for (1) For ASME Code, Section III, Class 1 piping, at axial Class 1 piping locations where the calculated stress range2 by Eq. (10) analyzed initially in in NB-3653 exceeds 1.2 Sm. the design stage. Environmental effects (2) For ASME Code, Section III, Class 2 and 3 or resulting from AIBs nonsafety class (not ASME Class 1, 2, or 3) piping, at were considered. axial locations where the calculated stress2 by the sum of Eqs. (9) and (10) in NC/ND-3653 exceeds 0.4 times the sum of the stress limits given in NC/ND-3653. (3) Nonsafety class piping which has not been evaluated to obtain stress information should have leakage cracks postulated at axial locations that produce the most severe environmental effects. B.2 Moderate-Energy Fluid System Piping B.2.a Fluid Systems Separated from Essential Systems and Components B.2.a Complies. See Section 3.6.1.3 and For the purpose of satisfying the separation provisions of Appendix 3B. plant arrangement as specified in B.1.a. of BTP ASB 3-1, a review of the piping layout and plant arrangement drawings should clearly show that the effects of through-wall leakage cracks at any location in piping designated to seismic and non-seismic standards are isolated or physically remote from essential systems and components. B.2.b Fluid System Piping in Containment Penetration Areas B.2.b Complies. See Section 3.6.2.1.2.4. Leakage cracks need not be postulated in those portions of piping from containment wall to and including the inboard or outboard isolation valves provided they meet the requirements of the ASME Code, Section III, NE-1120, and the stresses calculated2 by the sum of Eqs. (9) and (10) in ASME Code, Section III, NC-3653 do not exceed 0.4 times the sum of the stress limits given in NC-3653. B.2.c Fluid Systems in Areas Other Than Containment Penetration B.2.c.(1) Leakage cracks should be postulated in piping located adjacent to structures, systems, or components important to safety, except: B.2.c.(1).(a) Where exempted by B.2.b or B.2.d, B.2.c.(1).(a) Complies B.2.c.(1).(b) For ASME Code, Section III, Class 1 piping, the B.2.c.(1).(b) Complies stress range calculated2 by Eq. (10) in NB-3653 is See compliance less than 1.2 Sm, and statement to B.1.e.(1) B.2.c.(1).(c) For ASME Code, Section III, Class 2 or 3 and B.2.c.(1).(c) Complies nonsafety-class piping, the stresses calculated2 by See Section the sum of Eqs. (9) and (10) in NC/ND-3653 are less 3.6.2.1.2.4(c) than 0.4 times the sum of the stress limits given in NC/ND-3653. B.2.c.(2) Leakage cracks, unless the piping system is exempted by B.2.c.(2) Complies. (1) above, should be postulated at axial and See Section 3.6.2.1.1 circumferential locations that result in the most severe environmental consequences. B.2.c.(3) Leakage cracks should be postulated in fluid system B.2.c.(3) Complies. piping designed to non-seismic standards as necessary to See compliance satisfy B.3.d of BTP ASB 3-1. statement to B.2.b. Rev. 13

WOLF CREEK TABLE 3.6-2a (sheet 6) BTP 3-1 Position WCGS Compliance B.2.d Moderate-Energy Fluid Systems in Proximity to High-Energy B.2.d Complies. See Fluid Systems compliance statement to B.2.b. Leakage cracks need not be postulated in moderate-energy fluid system piping located in an area in which a break in high-energy fluid system piping is postulated, provided such leakage cracks would not result in more limiting environmental conditions than the high-energy piping break. Where a postulated leakage crack in the moderate-energy fluid system piping results in more limiting environmental conditions than the break in proximate high-energy fluid system piping, the provisions of B.2.c. should be applied. B.2.e Fluid Systems Qualifying as High-Energy or Moderate-Energy B.2.e Complies. See Systems Section 3.6.1.1.a Leakage cracks instead of breaks may be postulated in the piping of those fluid systems that qualify as high-energy fluid systems for only short operational periods5 but qualify as moderate-energy fluid systems for the major operational period. B.3 Type of Breaks and Leakage Cracks in Fluid System Piping B.3.a Circumferential Pipe Breaks The following circumferential breaks should be postulated individually in high-energy fluid system piping at the locations specified in B.1 of this position: B.3.a.(1) Circumferential breaks should be postulated in fluid B.3.a.(1) Complies. system piping and branch runs exceeding a nominal pipe See Section size of 1 inch, except where the maximum stress range2 3.6.2.1.2.2 exceeds the limits specified in B.1.c.(1) and B.1.c.(2), but the circumferential stress range is at least 1.5 times the axial stress range. Instrument lines, 1 inch and less nominal pipe or tubing size should meet the provisions of Regulatory Guide 1.11. B.3.a.(2) Where break locations are selected without the benefit of B.3.a.(2) Complies. All stress calculations, breaks should be postulated at the high-energy Class 1, 2 piping welds to each fitting, valve, or welded and 3 piping is analyzed by stress calculations. attachment. Non-nuclear class high energy piping breaks are postulated at all welds, fittings, welded attachments, etc. B.3.a.(3) Circumferential breaks should be assumed to result in B.3.a.(3) Complies. pipe severance and separation amounting to at least a one-diameter lateral displacement of the ruptured piping sections unless physically limited by piping restraints, structural members, or piping stiffness as may be demonstrated by inelastic limit analysis (e.g., a plastic hinge in the piping is not developed under loading). B.3.a.(4) The dynamic force of the jet discharge at the break B.3.a.(4) Complies. location should be based on the effective cross-sectional See Section 3.6.2.2.1. flow area of the pipe and on a calculated fluid pressure as modified by an analytically or experimentally determined thrust coefficient. Limited pipe displacement at the break location, line restrictions, flow limiters, positive pump-controlled flow, and the absence of energy reservoirs may be taken into account, as applicable, in the reduction of jet discharge. Rev. 13

WOLF CREEK TABLE 3.6-2a (sheet 7) BTP 3-1 Position WCGS Compliance B.3.a.(5) Pipe whipping should be assumed to occur in the plane B.3.a.(5) Complies. defined by the piping geometry and configuration, and to See Section 3.6.1.1.j initiate pipe movement in the direction of the jet reaction. B.3.b Longitudinal Pipe Breaks The following longitudinal breaks should be postulated in high-energy fluid system piping at the locations of the circumferential breaks specified in B.3.a.: B.3.b.(1) Longitudinal breaks in fluid system piping and branch B.3.b.(1) Complies. runs should be postulated in nominal pipe sizes 4-inch See Section and larger, except where the maximum stress range2 exceeds 3.6.2.1.2.2. the limits specified in B.1.c.(1) and B.1.c.(2), but the axial stress range is at least 1.5 times the circumferential stress range. B.3.b.(2) Longitudinal breaks need not be postulated at terminal B.3.b.(2): Per Section ends. 3.6.2.1.2.2, only breaks are postulated at terminal end, even if a longitudinal pipe weld is present at that point. B.3.b.(3) Longitudinal breaks should be assumed to result in an B.3.b.(3) Complies. axial split without pipe severance. Splits should be See Section oriented (but not concurrently) at two diametrically 3.6.2.1.3.1. opposed points on the piping circumference such that the jet reactions causes out-of-plant bending of the piping configuration. Alternatively, a single split may be assumed at the section of highest tensile stress as determined by detailed stress analysis (e.g., finite element analysis). B.3.b.(4) The dynamic force of the fluid jet discharge should be B.3.b.(4) Complies. based on a circular or elliptical (2D x 1/2D) break area See Section 3.6.2.2.1. equal to the effective cross-sectional flow area of the pipe at the break location and on a calculated fluid pressure modified by an analytically or experimentally determined thrust coefficient as determined for a circumferential break at the same location. Line restrictions, flow limiters, positive pump-controlled flow, and the absence of energy reservoirs may be taken into account, as applicable, in the reduction of jet discharge. B.3.b.(5) Piping movement should be assumed to occur in the B.3.b.(5) Complies. direction of the jet reaction unless limited by structural members, piping restraints, or piping stiffness as demonstrated by inelastic limit analysis. B.3.c Leakage Crack Leakage cracks should be postulated at those axial locations specified in B.1.e for high-energy fluid system piping and in those piping systems not exempted in B.2.c (1) for moderate-energy fluid system piping. B.3.c.(1) Leakage cracks need not be postulated in 1-inch and B.3.c.(1) Complies. smaller piping. Rev. 13

WOLF CREEK TABLE 3.6-2a (sheet 8) BTP 3-1 Position WCGS Compliance B.3.c.(2) For high-energy fluid system piping, the leakage cracks Leakage cracks for high-should be postulated to be in those circumferential energy piping need not be locations that result in the most severe environmental postulated for those analyzed initially in the consequences. For moderate-energy fluid system piping, design stage. see B.2.c.(2). Environmental effects resulting from AIBs were considered. B.3.c.(3) Fluid flow from a leakage crack should be based on a B.3.c.(3) Complies. circular opening of area equal to that of a rectangle one-half pipe diameter in length and one-half pipe wall thickness in width. B.3.c.(4) The flow from the leakage crack should be assumed to B.3.c.(4) Complies. result in an environment that wets all unprotected components within the compartment, with consequent flooding in the compartment and communicating compartments. Flooding effects should be determined on the basis of a conservatively estimated time period required to effect corrective actions. BTP MEB 3-1, REV. 2 FOOTNOTES 1 Systems and components required to shut down the reactor and mitigate the consequences of a postulated pipe rupture without offsite power. 2 For those loads and conditions in which Level A and Level B stress limits have been specified in the Design Specification (including the operating basis earthquake). 3 Extremities of piping runs that connect to structures, components (e.g., vessels, pumps, valves), or pipe anchors that act as rigid constraints to piping motion and thermal expansion. A branch connection to a main piping run is a terminal end of the branch run, except where the branch run is classified as part of a main run in the stress analysis and is shown to have a significant effect on the main run behavior. In piping runs which are maintained pressurized during normal plant conditions for only a portion of the run (i.e., up to the first normally closed valve), a terminal end of such runs is the piping connection to this closed valve. 4 Note that, in addition, breaks in non-seismic (i.e., non-Category I) piping are to be taken into account as described in Section II.2.k, "Interaction of Other Piping with Category I Piping," of SRP Section 3.9.2. 5 The operational period is considered "short" if the fraction of time that the system operates within the pressure-temperature conditions specified for high-energy fluid systems is about 2 percent of the time that the system operates as a moderate-energy fluid system (e.g., systems such as the reactor decay heat removal system qualify as moderate-energy fluid systems; however, systems such as auxiliary feedwater systems operated during PWR reactor startup, hot standby, or shutdown qualify as high-energy fluid systems). Rev. 13

WOLF CREEK TABLE 3.6-3 Historical Information HIGH-ENERGY PIPE BREAK INITIAL STRESS ANALYSIS RESULTS SYSTEM - MAIN STEAM SYSTEM Prob. No. P-001 Pipebreak Isometric No.: Figure 3.6-1(AB01) Issue - 8 Sheet 1 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 1** 11,227 5,286 16,563 37,800 5B 14,595 16,696 31,291 37,800 Bend 5M 14,460 16,976 31,436 37,800 Bend 20B 14,719 13,441 28,160 37,800 Bend 40B 16,003 5,745 21,748 37,800 Bend 50B 17,627 6,282 23,909 37,800 Bend 50E 18,414 6,137 24,551 37,800 80B 19,919 12,338 32,257 37,800 Bend 80E 20,470 10,814 31,284 37,800 90B 15,884 7,346 23,230 37,800 Bend 90M 16,122 7,086 23,208 37,800 Bend 101* 10,708 5,734 16,442 37,800

  • - Indicates Terminal End
    • - Indicates Terminal End Break Rev. 20

WOLF CREEK TABLE 3.6-3 (Sheet 1A) SYSTEM - MAIN STEAM SYSTEM (Loop 1) Prob. No. P-001 (0520511-C-0001) Pipebreak Isometric No.: Figure 3.6-1(AB01) Sheet 1 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 1** 10,461 6,612 17,073 37,800 5 B Bend 13,273 19,617 32,889 37,800 5 M + 11,560 19,865 31,425 37,800 Bend 20 B Bend 13,267 15,743 29,010 37,800 20 M 12,905 15,635 28,540 37,800 40 B Bend 17,351 8,009 25,360 37,800 50 B Bend 18,576 8,663 27,238 37,800 50 E 18,760 8,680 27,440 37,800 80 B + 15,240 12,284 27,524 37,800 Bend 80 E [180] 15,708 11,439 27,147 37,800 90 B [88] 12,907 14,833 27,740 37,800 Bend 90 AM Bend 13,144 15,533 28,676 37,800 101* 9,560 10,704 20,264 37,800

  • - Indicates Terminal End
    • - Indicates Terminal End Break

+ - Arbitrary Break Location,, deleted per MEB 3-1, Rev. 2 [] Indicates the new Node Points from calc 0520511-C-001 Rev. 20

WOLF CREEK TABLE 3.6-3 (Sheet 2) Historical Information SYSTEM - MAIN STEAM SYSTEM Prob. No. P-001A Pipebreak Isometric No.: Figure 3.6-1(AB01) Issue - 8 Sheet 1 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 1** 11,568 5,289 16,857 37,800 5B 15,241 16,546 31,787 37,800 Bend 5M 15,094 16,769 31,863 37,800 Bend 20B 15,681 14,774 30,455 37,800 Bend 20M 15,573 15,088 30,661 37,800 40B 16,700 7,575 24,275 37,800 Bend 40E 17,401 8,352 25,753 37,800 50B 18,676 4,671 23,347 37,800 Bend 50E 18,566 4,812 23,378 37,800 80B 21,192 13,457 34,649 37,800 Bend 80M 21,450 12,702 34,152 37,800 Bend 90B 16,620 7,785 24,405 37,800 Bend 90M 16,624 8,441 25,065 37,800 Bend 101* 11,239 5,830 17,069 37,800

  • - Indicates Terminal End
    • - Indicates Terminal End Break Rev. 20

WOLF CREEK TABLE 3.6-3 (Sheet 2A) SYSTEM - MAIN STEAM SYSTEM (Loop 2) Prob. No. P-001A (0520511-C-002) Pipebreak Isometric No.: Figure 3.6-1(AB01) Sheet 1 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 1** 10,689 6,658 17,347 37,800 5 B Bend 14,085 19,761 33,845 37,800 5 M Bend + 11,762 19,896 31,657 37,800 20 B Bend 13,644 17,467 31,111 37,800 20 M 13,276 17,636 30,912 37,800 40 B Bend 17,758 8,704 26,461 37,800 40 E 13,008 10,629 23,637 37,800 50 B Bend 18,526 9,475 28,001 37,800 50 E 18,016 8,718 26,733 37,800 80 B Bend + 16,103 11,529 27,632 37,800 80 M Bend 16,680 10,696 27,375 37,800 80 E [180] 16,374 10,489 26,863 37,800 90 B [88] 13,420 15,944 29,373 37,800 Bend 90AM Bend 13,702 17,349 31,051 37,800 101* 9,089 11,541 20,631 37,800

  • - Indicates Terminal End
    • - Indicates Terminal End Break

+ - Arbitrary Break Location,, deleted per MEB 3-1, Rev. 2 [] Indicates the new Node Points from calc 0520511-C-002 Rev. 20

WOLF CREEK TABLE 3.6-3 (Sheet 3) Historical Information SYSTEM - MAIN STEAM SYSTEM Prob. No. P-002 Pipebreak Isometric No.: Figure 3.6-1(AB01) Issue - 8 Sheet 1 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 1** 11,945 3,697 15,642 37,800 5B 15,639 11,633 27,272 37,800 Bend 5M 15,518 11,803 27,321 37,800 20B 16,599 9,296 25,895 37,800 Bend 20M 16,430 9,430 25,860 37,800 Bend 40B 17,366 7,135 24,501 37,800 Bend 40E 16,185 7,560 23,745 37,800 60B 26,965 10,054 37,019 37,800 Bend 60E 23,473 10,860 34,333 37,800 101* 16,315 1,782 18,097 37,800

  • - Indicates Terminal End
    • - Indicates Terminal End Break Rev. 20

WOLF CREEK TABLE 3.6-3 (Sheet 3A) SYSTEM - MAIN STEAM SYSTEM (Loop 4) Prob. No. P-002 (0520511-C-004) Pipebreak Isometric No.: Figure 3.6-1(AB01) Sheet 1 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 1** 13,568 9,793 23,361 37,800 5 B Bend 13,992 23,569 37,562 37,800 5 M + 10,440 21,199 31,639 37,800 20 B Bend 15,026 11,679 26,705 37,800 20 M Bend 14,241 13,231 27,472 37,800 40 B Bend 14,858 12,984 27,842 37,800 40 E 14,682 14,822 29,504 37,800 60 B Bend + 17,458 11,113 28,571 37,800 60 E 15,656 11,987 27,643 37,800 101* 11,587 1,906 13,493 37,800

  • - Indicates Terminal End
    • - Indicates Terminal End Break

+ - Arbitrary Break Location,, deleted per MEB 3-1, Rev. 2 Rev. 20

WOLF CREEK TABLE 3.6-3 (Sheet 4) Historical Information SYSTEM - MAIN STEAM SYSTEM Prob. No. P-002A Pipebreak Isometric No.: Figure 3.6-1(AB01) Issue - 8 Sheet 1 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 1** 12,581 3,760 16,341 37,800 5B 16,436 11,870 28,306 37,800 Bend 5M 16,333 12,073 28,406 37,800 20B 17,813 9,169 26,982 37,800 Bend 20M 17,600 9,220 26,820 37,800 40B 16,643 7,828 24,471 37,800 Bend 40E 16,995 8,152 25,147 37,800 60B 24,608 10,520 35,128 37,800 Bend 60M 24,467 11,043 35,510 37,800 101* 15,677 1,999 17,676 37,800

  • - Indicates Terminal End
    • - Indicates Terminal End Break Rev. 20

WOLF CREEK TABLE 3.6-3 (Sheet 4A) SYSTEM - MAIN STEAM SYSTEM (Loop 3) Prob. No. P-002A (0520511-C-003) Pipebreak Isometric No.: Figure 3.6-1(AB01) Sheet 1 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 1** 10,921 8,212 19,133 37,800 5 B 13,727 23,749 37,476 37,800 Bend 5 M + 10,228 21,371 31,599 37,800 20 B 15,432 11,440 26,873 37,800 Bend 20 M 14,744 12,902 27,647 37,800 40 B 14,150 13,322 27,472 37,800 Bend 40 E 16,029 15,425 31,455 37,800 60 B 15,926 11,531 27,457 37,800 Bend 60 M 15,247 12,212 27,459 37,800 

   +

101* 10,484 1,811 12,296 37,800

  • - Indicates Terminal End
    • - Indicates Terminal End Break

+ - Arbitrary Break Location,, deleted per MEB 3-1, Rev. 2 Rev. 20

WOLF CREEK TABLE 3.6-3 (Sheet 5) Historical Information SYSTEM - MAIN FEEDWATER SYSTEM INSIDE CONTAINMENT Prob. No. P-003 Pipebreak Isometric No.: Figure 3.6-1(AE04) Issue - 5 Sheet 2 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 5* 5,486 5,662 11,148 32,400 20E 6,383 11,123 17,506 32,400 20M 6,161 11,708 17,869 32,400 27B 5,230 18,658 23,888 32,400 Bend 27M 5,543 19,604 25,147 32,400 35M 6,099 11,578 17,677 32,400 35E 6,223 12,124 18,347 32,400 75M 5,194 13,530 18,724 32,400 95M 4,987 12,747 17,734 32,400 95E 4,964 13,095 18,059 32,400 100+ 4,975 13,294 18,269 32,400 125**,++ 5,323 17,612 22,935 32,400

  • - Indicates Terminal End
    • - Indicates Terminal End Break

++ - Terminal End Break includes Break at reducer Rev. 20

WOLF CREEK TABLE 3.6-3 (Sheet 5A) Historical Information SYSTEM - MAIN FEEDWATER SYSTEM (Loop 1) Prob. No. P-003 (0520511-C-005) Pipebreak Isometric No.: Figure 3.6-1(AE04) Sheet 2 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 5* 7,993 2,638 10,631 32,400 20E 6,674 6,314 12,988 32,400 20AM 6,648 6,657 13,305 32,400 27 B (Bend) 5,725 12,039 17,764 32,400 27 M [27A] + 6,037 12,538 18,575 32,400 35 M 6,274 6,286 12,560 32,400 35 E 6,393 6,457 12,850 32,400 75 AM 10,528 15,097 25,624 32,400 95 M [95C] 13,845 10,985 24,829 32,400 95 E 9,893 16,120 26,013 32,400 100 [10R] + 10,405 19,074 29,479 32,400 125 [12N]**,++ 11,002 23,373 34,376 32,400

  • - Indicates Terminal End
    • - Indicates Terminal End Break

+ - Arbitrary Break Location, deleted per MEB 3-1, Rev. 2 ++ - Terminal End Break includes Break at Reducer [] - Indicates the new Node Points from calc 0520511-C-005 Rev. 23

WOLF CREEK TABLE 3.6-3 (Sheet 5B) SYSTEM - MAIN FEEDWATER SYSTEM (Loop 1) Prob. No. P-003 P0001 Pipebreak Isometric No.: Figure 3.6-1(AE04) Sheet 2 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 5* 8,470 2,638 11,108 32,400 20E 7,048 6,314 13,362 32,400 20AM 6,929 6,657 13,586 32,400 27 B (Bend) 6,092 12,039 18,131 32,400 27 M [27A] + 6,524 12,538 19,062 32,400 35 M 6,737 6,286 13,024 32,400 35 E 6,864 6,457 13,320 32,400 55 M 15,176 8,328 23,504 32,400 55 E 11,294 3,740 15,034 32,400 75 AM 11,003 15,097 26,099 32,400 95 M [95C] 14,334 10,985 25,319 32,400 95 E 10,372 16,120 26,492 32,400 100 [10R] + 10,875 19,074 29,949 32,400 125 [12N]**,++ 11,482 23,373 34,855 32,400

  • - Indicates Terminal End
    • - Indicates Terminal End Break

+ - Arbitrary Break Location, deleted per MEB 3-1, Rev. 2 ++ - Terminal End Break includes Break at Reducer [] - Indicates the new Node Points from calc P0001 Rev. 23

WOLF CREEK TABLE 3.6-3 (Sheet 6) Historical Information SYSTEM - MAIN FEEDWATER SYSTEM INSIDE CONTAINMENT Prob. No. P-003A Pipebreak Isometric No.: Figure 3.6-1(AE04) Issue - 8 Sheet 2 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 5* 5,486 5,662 11,148 32,400 20E 6,383 11,123 17,506 32,400 20M 6,161 11,708 17,869 32,400 27B 5,230 18,658 23,888 32,400 Bend 27M 5,543 19,604 25,147 32,400 35M 6,099 11,578 17,677 32,400 35E 6,223 12,124 18,347 32,400 75M 5,194 13,530 18,724 32,400 95M 4,987 12,747 17,734 32,400 95E 4,964 13,095 18,059 32,400 100 4,975 13,294 18,269 32,400 125**, ++ 5,323 17,612 22,935 32,400

  • - Indicates Terminal End
    • - Indicates Terminal End Break

++ - Terminal End Break includes Break at Reducer Rev. 20

WOLF CREEK TABLE 3.6-3 (Sheet 6A) Historical Information SYSTEM - MAIN FEEDWATER SYSTEM (Loop 2) Prob. No. P-003A Pipebreak Isometric No.: Figure 3.6-1(AE04) Sheet 2 Pipe Break Node No. Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 5* 8,001 2,638 10,639 32,400 20 E 6,685 6,310 12,995 32,400 20AM 6,670 6,653 13,323 32,400 27B (Bend) 5,717 12,036 17,753 32,400 27M [27A] + 6,024 12,535 18,559 32,400 35M 6,239 6,102 12,341 32,400 35E 6,362 6,415 12,777 32,400 75AM 10,588 14,920 25,509 32,400 95M [95C] 13,933 10,928 24,861 32,400 95E 10,047 15,728 25,775 32,400 100 [10R] + 10,592 18,514 29,105 32,400 125 [12N]**,++ 11,181 22,579 33,760 32,400

  • - Indicates Terminal End
    • - Indicates Terminal End Break

+ - Arbitrary Break Location, deleted per MEB 3-1, Rev. 2 ++ - Terminal End Break includes Break at Reducer [] - Indicates the new Node Points from calc 0520511-C-006 Rev. 23

WOLF CREEK TABLE 3.6-3 (Sheet 6B) SYSTEM - MAIN FEEDWATER SYSTEM (Loop 2) Prob. No. P-003A (P0002) Pipebreak Isometric No.: Figure 3.6-1(AE04) Sheet 2 Pipe Break Node No. Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 5* 8,485 2,638 11,122 32,400 20 E 7,113 6,310 13,423 32,400 20AM 7,052 6,653 13,705 32,400 27B (Bend) 6,155 12,036 18,191 32,400 27M [27A] + 6,513 12,535 19,048 32,400 35M 6,715 6,102 12,818 32,400 35E 6,844 6,415 13,259 32,400 55M 15,458 8,497 23,956 32,400 55E 11,504 3,888 15,392 32,400 75AM 11,077 14,920 25,998 32,400 95M [95C] 14,425 10,928 25,354 32,400 95E 10,536 15,728 26,265 32,400 100 [10R] + 11,078 18,514 29,592 32,400 125 [12N]**,++ 11,677 22,579 34,256 32,400

  • - Indicates Terminal End
    • - Indicates Terminal End Break

+ - Arbitrary Break Location, deleted per MEB 3-1, Rev. 2 ++ - Terminal End Break includes Break at Reducer [] - Indicates the new Node Points from calc P0002 () - Stress values increased to consider min wall thickness at that location Rev. 23

WOLF CREEK TABLE 3.6-3 (Sheet 7) Historical Information SYSTEM - MAIN FEEDWATER SYSTEM Prob. No. P-004 Pipebreak Isometric No.: Figure 3.6-1(AE05) Issue - 5 Sheet 3 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 10* 5,991 4,776 10,767 32,400 20B 5,468 17,391 22,859 32,400 Bend 20M 5,660 18,194 23,854 32,400 30M 5,271 18,237 23,508 32,400 30E 5,360 18,455 23,815 32,400 45B 5,417 20,764 26,181 32,400 Bend 45M 5,670 20,672 26,342 32,400 71M 5,625 13,151 18,776 32,400 71E 5,842 14,116 19,958 32,400 90E 5,394 13,537 18,931 32,400 95 5,576 14,747 20,323 32,400 100**, ++ 5,656 16,297 21,953 32,400

  • - Indicates Terminal End
    • - Indicates Terminal End Break

++ - Terminal End Break includes break at the reducer Rev. 20

WOLF CREEK TABLE 3.6-3 (Sheet 7A) Historical Information SYSTEM - MAIN FEEDWATER SYSTEM (Loop 4) Prob. No. P-004 (0520511-C-008) Pipebreak Isometric No.: Figure 3.6-1(AE05) Sheet 3 Pipe Break Node No. Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 10* 15,071 10,333 25,403 32,400 20B[19A](Bend) 10,241 20,915 31,155 32,400 20M + 10,913 21,052 31,965 32,400 30 M 9,396 21,270 30,667 32,400 30 E [32] 8,523 20,433 28,956 32,400 45B[D45](Bend) 10,157 10,092 20,249 32,400 45 M + 10,580 10,072 20,651 32,400 71M [71A] 8,575 13,939 22,514 32,400 71 E 8,859 12,770 21,629 32,400 90 E [91] 8,650 18,098 26,748 32,400 95 9,955 20,380 30,335 32,400 100[101]**,++, 10,665 23,861 34,526 32,400

  • - Indicates Terminal End
    • - Indicates Terminal End Break

+ - Arbitrary Break Location, deleted per MEB 3-1, Rev. 2 ++ - Terminal End Break includes Break at the Reducer [] Indicates the new Node Points from calc 0520511-C-008 Stress values increased to consider min wall thickness at that location Rev. 23

WOLF CREEK TABLE 3.6-3 (Sheet 7B) SYSTEM - MAIN FEEDWATER SYSTEM (Loop 4) Prob. No. P-004 P0004 Pipebreak Isometric No.: Figure 3.6-1(AE05) Sheet 3 Pipe Break Node No. Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 10* 10,350 10,169 20,519 32,400 20B[19A](Bend) 9,293 20,659 29,953 32,400 20M + 9,386 20,836 30,223 32,400 30 M 9,118 21,374 30,492 32,400 30 E [32] 8,157 20,515 28,672 32,400 45B[D45](Bend) 10,444 10,128 20,573 32,400 45 M + 10,898 10,095 20,993 32,400 71M [71A] 9,027 13,939 22,966 32,400 71 E 9,304 12,771 22,075 32,400 90 E [91] 9,107 18,101 27,208 32,400 95 10,411 20,385 30,796 32,400 100[101]**,++, 10,341 22,112 32,453 32,400

  • - Indicates Terminal End
    • - Indicates Terminal End Break

+ - Arbitrary Break Location, deleted per MEB 3-1, Rev. 2 ++ - Terminal End Break includes Break at the Reducer [] Indicates the new Node Points from calc P0004 Stress values increased to consider min wall thickness at that location Rev. 23

WOLF CREEK TABLE 3.6-3 (Sheet 8) Historical Information SYSTEM - MAIN FEEDWATER SYSTEM Prob. No. P-004A Pipebreak Isometric No.: Figure 3.6-1(AE05) Issue - 7 Sheet 3 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 10* 5,991 4,776 10,767 32,400 20B 5,468 17,391 22,859 32,400 Bend 20M 5,660 18,194 23,854 32,400 30M 5,271 18,237 23,508 32,400 30E 5,360 18,455 23,815 32,400 45B 5,417 20,764 26,181 32,400 Bend 45M 5,670 20,672 26,342 32,400 71M 5,625 13,151 18,776 32,400 71E 5,842 14,116 19,958 32,400 90E 5,394 13,537 18,931 32,400 95 5,576 14,747 20,323 32,400 100**, ++ 5,656 16,297 21,953 32,400

  • - Indicates Terminal End
    • - Indicates Terminal End Break

++ - Terminal End Break includes break at the reducer Rev. 20

WOLF CREEK TABLE 3.6-3 (Sheet 8A) Historical Information SYSTEM - MAIN FEEDWATER SYSTEM (Loop 3) Prob. No. P-004A (0520511-C-007) Pipebreak Isometric No.: Figure 3.6-1(AE05) Sheet 3 Pipe Break Node No. Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 10* 15,013 10,381 25,394 32,400 20B Bend 10,257 21,160 31,416 32,400 20M + 10,918 21,295 32,212 32,400 30M 9,424 21,362 30,785 32,400 30E 8,550 20,511 29,061 32,400 45B Bend 10,183 10,327 20,511 32,400 45M + 10,599 10,733 21,332 32,400 71M [71A] 8,584 13,303 21,887 32,400 71E 8,927 12,833 21,760 32,400 90E 8,690 18,347 27,037 32,400 95 9,998 20,709 30,707 32,400 100[101]**, 9,940 22,504 32,445 32,400 ++

  • - Indicates Terminal End
    • - Indicates Terminal End Break

+ - Arbitrary Break Location, deleted per MEB 3-1, Rev. 2 ++ - Terminal End Break includes break at the reducer [] - Indicates the new Node Points from calc 0520511-C-007 Rev. 23

WOLF CREEK TABLE 3.6-3 (Sheet 8B) SYSTEM - MAIN FEEDWATER SYSTEM (Loop 3) Prob. No. P-004A P0003 Pipebreak Isometric No.: Figure 3.6-1(AE05) Sheet 3 Pipe Break Node No. Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 10* 10,265 10,163 20,428 32,400 20B Bend 9,282 20,829 30,111 32,400 20M + 9,378 21,013 30,392 32,400 30M 9,106 21,516 30,622 32,400 30E 8,167 20,632 28,799 32,400 45B Bend 10,470 10,360 20,831 32,400 45M + 10,919 10,761 21,680 32,400 71M [71A] 9,045 13,303 22,348 32,400 71E 9,403 12,834 22,237 32,400 90E 9,147 18,350 27,497 32,400 95 10,457 20,714 31,171 32,400 100[101]**, ++ 10,408 22,511 32,919 32,400

  • - Indicates Terminal End
    • - Indicates Terminal End Break

+ - Arbitrary Break Location, deleted per MEB 3-1, Rev. 2 ++ - Terminal End Break includes break at the reducer [] - Indicates the new Node Points from calc P0003 Rev. 23

WOLF CREEK TABLE 3.6-3 (Sheet 9) SYSTEM - HIGH PRESSURE COOLANT INJECTION Prob. No. P-21 Pipebreak Isometric No.: Figure 3.6-1(EM02) Issue - 5 Sheet 27 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 5 9,870 8,286 18,156 39,448 TNGT 30 TNGT 6,465 15,943 22,408 39,448 50* TNGT 6,905 2,214 9,119 39,448 164 TNGT 10,870 1,806 12,676 39,448 180* TNGT 7,065 2,669 9,734 39,448 67 8,481 2,353 10,834 39,448 100M Bend 7,440 2,849 10,289 39,448 116* 8,440 1,732 10,172 39,448 255E 5,421 6,884 12,305 39,448 266 TNGT 13,353 1,280 14,633 39,448 320B Bend 3,975 24,019 27,994 39,448 340* 5,264 2,834 8,098 39,448

  • - Indicates Terminal End Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 10) SYSTEM - MAIN STEAM - AUXILIARY BUILDING Prob. No. P-026 Pipebreak Isometric No.: Figure 3.6-1(AB01) Issue - 7 Sheet 1 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 5 9,170 16,814 25,984 37,800 25 7,063 3,446 10,509 37,800 33 8,902 9,195 18,097 37,800 45F 17,924 0 17,924 38,700 60 12,519 2,299 14,818 37,800 83 6,924 2,309 9,233 37,800 300 8,922 18,909 27,831 37,800 294 8,722 9,636 18,408 37,800 291 9,004 5,890 14,894 37,800 289 8,768 6,571 15,339 37,800 287 11,351 1,228 12,579 37,800 282 9,313 978 10,291 37,800 NOTE: This problem meets No Break Zone Criteria

  • - Indicates Terminal End Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 11) SYSTEM - MAIN STEAM Prob. No. P-27BY Pipebreak Isometric No.: Figure 3.6-1(AB01) Issue - 7 Sheet 1 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 100 16,223 12,234 28,457 32,400 105 11,667 8,978 20,645 32,400 106 12,783 10,526 23,309 32,400 160 15,292 10,316 25,608 32,400 170 15,535 9,142 24,677 32,400 185 14,959 5,561 20,520 32,400 200 11,764 14,004 25,768 32,400 202 9,127 6,284 15,411 32,400 210 9,000 15,582 24,582 32,400 215 10,651 20,208 30,859 32,400 145 16,002 14,908 30,910 32,400 142 12,292 6,952 19,244 32,400 190 9,239 20,320 29,559 32,400 205 9,322 17,696 27,018 32,400 NOTE: This problem meets No Break Zone Criteria

  • - Indicates Terminal End Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 12) SYSTEM - MAIN FEEDWATER SYSTEM Prob. No. P-028 Pipebreak Isometric No.: Figure 3.6-1(AE04) Issue - 2 Sheet 2 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 675 10,325 2,511 12,836 32,400 720 10,658 5,742 16,400 32,400 775 9,992 9,295 19,287 32,400 820 10,239 5,684 15,923 32,400 575 9,941 7,507 17,448 32,400 620 10,007 6,673 16,680 32,400 875 10,415 3,109 13,524 32,400 920 10,028 6,765 16,793 32,400 NOTE: This problem meets No Break Zone Criteria

  • - Indicates Terminal End Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 13) SYSTEM - CVCS - LETDOWN TO REHEAT HEAT EXCHANGER Prob. No. P-29B1 Pipebreak Isometric No.: Figure 3.6-1(BG11) Issue - 5 Sheet 23 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 815* 4,568 4,363 8,931 37,712 840M Bend 6,486 19,775 26,261 37,712 860M Bend 7,607 14,689 22,296 37,712 980M Bend 6,722 11,961 18,683 37,712 878* 4,607 1,121 5,728 37,712 840B Bend 6,402 18,665 25,067 37,712 860B Bend 7,478 14,535 22,013 37,712

  • - Indicates Terminal End Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 14) SYSTEM - CVCS LETDOWN TO REHEAT Prob. No. P-29B2 HEAT EXCHGR-AUX BLDG Issue - 4 Pipebreak Isometric No.: Figure 3.6-1(BG11) Sheet 23 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 716* 3,232 739 3,971 37,710 774B Bend 8,047 16,567 24,614 37,710 774M Bend 8,568 16,075 24,643 37,710 774E Bend 9,078 13,917 22,995 37,710 778E Bend 8,635 13,483 22,118 37,710 778M Bend 8,377 13,614 21,991 37,710 804M Bend 4,296 9,459 13,755 37,710 818* 4,269 1,444 5,713 37,710 752M Bend 5,935 14,275 20,210 37,710

  • - Indicates Terminal End Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 15) Historical Information SYSTEM - CVCS LETDOWN FLOW - AUX BLDG Prob. No. P-29B3 Pipebreak Isometric No.: Figure 3.6-1(BG11) Issue - 6 Sheet 23 (BG03) Sheet 20 (BG22) Sheet 25 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 450* 4,860 2,692 7,553 37,685 415M Bend 4,961 8,844 13,805 37,685 395 11,182 6,518 17,700 37,685 390 13,332 16,192 29,524 37,685 385* 15,129 13,726 28,855 37,685 705E Bend 5,497 19,064 24,501 37,685 507 11,814 7,909 19,723 37,685 515 12,289 11,476 23,764 37,685 485 12,871 13,855 26,727 37,685 415 4,961 8,844 13,805 37,685

  • - Indicates Terminal End Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 15A) SYSTEM - CVCS LETDOWN TO FLOW - AUX BLDG Prob. No P-029B3 Pipebreak Isometric No: Figure 3.6-1 (BG11) Sheet 23 Altran Calc. No. 02101-C-003, Rev. 0 (BG03) Sheet 20 (BG22) Sheet 25 Stress (psi) Pipe Break Node Primary Secondary Total Stress Limit (psi) (EQN 9) (EQN 10) (EQN 9 + EQN 10) 0.8 (SA + 1.2 S h) 450* 5,077 4,822 9,899 37,685 415 M Bend 4,084 12,960 17,043 37,685 395 ** 7,647 8,743 16,390 37,685 390 ** 13,284 19,176 32,460 37,685 385* 12,344 12,329 24,673 37,685 705 E Bend ** 5,301 19,103 24,404 37,685 507 ** 7,471 7,423 14,894 37,685 515 ** 6,780 11,856 18,637 37,685 485 8,902 24,322 33,224 37,685 415 4,084 12,960 17,043 37,685

  • - Indicates Terminal End
    • - Indicates that intermediate Break is deleted (per MEB 3-1, Rev. 2)

Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 16) Historical Information SYSTEM - CVCS LETDOWN TO REHEAT BLDG Prob. No. P-29B3 Pipebreak Isometric No.: Figure 3.6-1(BG11) Issue - 5 Sheet 23 (BG03) Sheet 20 (BG22) Sheet 25 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 818* 7,489 5,231 12,720 37,685 834M Bend 4,085 24,439 28,523 37,685 838B Bend 4,027 24,530 28,557 37,685 815* 6,347 2,843 9,190 37,685 790M Bend 4,089 21,269 25,357 37,685 720B 5,567 17,335 22,902 37,685 868 6,653 13,257 19,910 37,685

  • - Indicates Terminal End Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 16A) SYSTEM - CVCS LETDOWN TO REHEAT BLDG Prob. No P-029B3 Pipebreak Isometric No: Figure 3.6-1 (BG11) Sheet 23 Altran Calc. No. 02101-C-003, Rev. 0 (BG03) Sheet 20 (BG22) Sheet 25 Stress (psi) Pipe Break Node Primary Secondary Total Stress Limit (psi) (EQN 9) (EQN 10) (EQN 9 + EQN 10) 0.8 (SA + 1.2 S h) 818* 10,021 17,863 27,884 37,685 834 M Bend ** 4,672 24,440 29,112 37,685 838 B Bend** 5,038 23,808 28,846 37,685 815* 11,367 10,977 22,344 37,685 790 M Bend 9,196 21,952 31,147 37,685 720 B 7,731 18,862 26,593 37,685 868 6,453 12,341 18,793 37,685

  • - Indicates Terminal End
    • - Indicates that Intermediate Break is deleted per MEM 3-1, Rev. 2 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 17) SYSTEM - CHEMICAL AND VOLUME CONTROL Prob. No. P-31 Pipebreak Isometric No.: Figure 3.6-1(BG09) Issue - 7 Sheet 21 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 780 12,017 3,373 15,390 39,657 785* Bend 7,297 3,256 10,553 39,657 805 8,187 4,923 13,110 39,657 810M Bend 8,024 7,057 15,081 39,657 815 7,414 6,423 13,837 39,657 874M Bend 6,676 7,385 14,061 39,657 875T 8,333 748 9,081 39,657 873M Bend 6,395 5,078 11,473 39,657 903 6,688 906 7,594 39,657 906* 6,825 1,108 7,933 39,657 891* 6,451 2,228 8,679 39,657 995* 6,653 45 6,698 39,657 932* 6,712 69 6,781 39,657

  • - Indicates Terminal End Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 18) SYSTEM - CHEMICAL AND VOLUME CONTROL Prob. No. P-33 Pipebreak Isometric No.: Figure 3.6-1(BG09) Issue - 7 Sheet 21 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 140* 4,876 9,159 14,035 39,680 130 TNGT 5,265 12,715 17,980 39,680 95 TNGT 10,483 3,872 14,355 39,680 385M Bend 9,739 7,415 17,154 39,680 85T 11,178 5,113 16,291 39,680 465 8,537 3,706 12,243 39,680 425** 12,930 9,416 22,346 39,680 505** 12,091 12,432 24,523 39,680 580** 14,075 11,878 25,953 39,680 25T** 11,131 17,068 28,199 39,680

  • - Indicates Terminal End Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 18A) SYSTEM - CHEMICAL AND VOLUME CONTROL Prob. No. P-033-007-CN005 Pipebreak Isometric No.: Figure 3.6-1(BG09) Sheet 21 Pipe Break Stress (psi) Stress Limit (psi) Node Primary Secondary Total 0.8 (SA + 1.2Sh) 140* 3,235 8,078 11,312 39,610 130 TNGT 3,393 11,240 14,633 39,610 95 TNGT 4,878 1,618 6,496 39,610 385M Bend 4,953 5,326 10,279 39,610 85T 5,330 3,700 9,030 39,610 465 4,288 2,518 6,806 39,610 425** 6,406 6,578 12,984 39,610 505** 7,779 19,100 26,879 39,610 580** 8,838 19,993 28,830 39,610 25T** 6,381 16,697 23,078 39,610

  • - Indicates Terminal End
    • - No Break Zone Rev. 32

WOLF CREEK TABLE 3.6-3 (Sheet 19) SYSTEM - CHEMICAL AND VOLUME CONTROL Prob. No. P-033A Pipebreak Isometric No.: Figure 3.6-1(BG09) Issue - 6 Sheet 21 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 140* 3,249 677 3,926 39,448 164 5,627 25,146 30,773 39,448 165M** Bend 3,446 14,055 17,501 39,448 181 8,686 10,360 19,046 39,448 240 3,703 5,056 8,759 39,448 305* 5,651 594 6,245 39,448 380* 5,714 1,109 6,823 39,448

  • - Indicates Terminal End
    • - Indicates Intermediate Breaks That Have Been Replaced Rev. 32

WOLF CREEK TABLE 3.6-3 (Sheet 20) SYSTEM - CVCS LETDOWN - AUX BLDG Prob. No. 0720515-C-003 Pipebreak Isometric No.: Figure 3.6-1(BG0) (P-036 Issue - 6) Sheet 20 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 80* 3772 3099 6872 40376 163** 4661 6374 11035 40376 180M** Bend 6902 1927 8828 40376 190* 3590 1134 4724 40376 255* 9732 2518 12250 40376 225* 5543 1141 6684 40376

  • - Indicates Terminal End
    • - Indicates Intermediate Break Deleted per MEB 3-1, Rev. 2 Rev. 32

WOLF CREEK TABLE 3.6-3 (Sheet 21) SYSTEM - TURBINE DRIVEN AUXILIARY FEEDWATER PUMP Prob. No. P-060 Pipebreak Isometric No.: Figure 3.6-1(FC01) Issue - 10 Sheet 49 (AB01) Sheet 1 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 15* 4,140 2,006 6,146 32,400 30M Bend 4,433 2,839 7,272 32,400 35E Bend 5,288 3,945 9,233 32,400 35M Bend 4,968 3,732 8,700 32,400 48T 7,820 10,955 18,775 32,400 50* 8,759 15,837 24,596 32,400 215T** 4,374 4,294 8,668 32,400 260** 6,260 8,444 14,704 32,400 275** 6,295 12,807 17,102 32,400 285** 5,861 16,742 22,603 32,400 410** 4,265 9,537 14,162 32,400 Bend 410M** 4,208 8,496 13,704 32,400 Bend 

* - Indicates Terminal End
    • - Meets No Break Zone Criteria Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 22) Historical Information SYSTEM - CHEMICAL AND VOLUME CONTROL SYSTEM Prob. No. P-069 Pipebreak Isometric No.: Figure 3.6-1(BG02) Sheet 19 (BG10) Sheet 22 (BG09) Sheet 21 (EM02) Sheet 37 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 170* 3,824 16,258 20,082 39,610 93 TNGT 4,806 273 5,079 39,610 140* 5,252 2,086 7,338 39,610 955 E 5,283 5,649 10,932 39,610 900 5,606 2,438 7,932 39,610 870* TNGT 4,596 4,181 8,777 39,610 650* 4,746 10 4,756 39,610 310* 4,928 362 5,290 39,610 266 5,445 14,584 20,029 39,610 270 M Bend 4,679 14,656 19,335 39,610 730 TNGT 13,020 5,658 18,678 39,610 A75 TNGT 13,736 3,119 16,855 39,610 745 B Bend 4,927 1,325 6,252 39,610 155 M 4,025 9,259 13,284 39,610 50 5,912 8,091 14,003 39,610 970* 5,361 11,601 16,962 39,610

  • - Indicates Terminal End Rev. 18

WOLF CREEK TABLE 3.6-3 (Sheet 22A) SYSTEM - CHEMICAL AND VOLUME CONTROL SYSTEM Prob. No. P-069-005-CN001 Pipebreak Isometric No.: Figure 3.6-1 (BG02) Sheet 19 (BG10) Sheet 22 (BG09) Sheet 21 (EM02) Sheet 37 Stress (psi) Pipe Break Node Primary Secondary Total Stress Limit (psi) (EQN 9) (EQN 10) (EQN 9 + EQN 10) 0.8 (SA + 1.2Sh) 170* 6,364 26,619 32,983 39,610 93 TNGT** 8,372 550 8,922 39,610 140* 8,844 3,174 12,019 39,610 650* 4,751 20 4,771 39,610 310* 6,028 730 6,758 39,610 266** 5,767 12,516 18,283 39,610 270 M Bend** 4,956 12,241 17,197 39,610 730 TNGT** 19,235 7,270 26,504 39,610 A75** 5,141 590 5,730 39,610 745** 6,269 1,584 7,852 39,610 155 M** 4,978 11,573 16,551 39,610 50** 7,152 7,681 14,833 39,610 970* 5,312 182 5,494 39,610 

  * - Indicates Terminal End
  ** - Indicates that Intermediate Break is deleted (per MEB 3-1, R/2)

Rev. 18

WOLF CREEK TABLE 3.6-3 (Sheet 23) (Historical Information) SYSTEM - CHEMICAL VOLUME CONTROL SYSTEM Prob. No. P-069 Pipebreak Isometric No.: Figure 3.6-1(BG02) Sheet 19 (BG10) Sheet 22 (BG09) Sheet 21 (EM02) Sheet 37 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 95F TNGT 13,338 22,072 33,410 39,610 95C 16,645 17,209 35,854 39,610 885M 5,173 3,685 8,858 39,610 620* 5,310 541 5,851 39,610 625 TNGT 10,642 5,560 16,202 39,610 604 E 5,119 14,293 19,412 39,610 601 TNGT 5,615 17,256 22,871 39,610 641 12,488 2,385 14,873 39,610 626 B Bend 10,120 5,459 15,579 39,610 574 M 11,868 1,861 13,729 39,610 573* TNGT 8,461 1,409 9,870 39,610 545 TNGT 10,951 10,004 20,955 39,610 62A* 7,741 1,667 9,408 39,610 75 9,631 7,767 17,398 39,610 91 14,254 7,365 21,619 39,610

  • - Indicates Terminal End Rev. 18

WOLF CREEK TABLE 3.6-3 (Sheet 23A) SYSTEM - CHEMICAL VOLUME CONTROL SYSTEM Prob. No. P-069-005-CN001 Pipebreak Isometric No.: Figure 3.6-1 (BG02) Sheet 19 (BG10) Sheet 22 (BG09) Sheet 21 (EM02) Sheet 37 Stress (psi) Pipe Break Node Primary Secondary Total Stress Limit (psi) (EQN 9) (EQN 10) (EQN 9 + EQN 10) 0.8 (SA + 1.2Sh) 574** 8,282 1,200 9,482 39,610, 573* TNGT 6,509 934 7,443 39,610, 545** 9,414 14,221 23,635 39,610, 75** 7,470 9,504 16,974 39,610, 91** 16,898 11,793 28,691 39,610, 575** 9,102 502 9,603 39,610, 

  * - Indicates Terminal end
  ** - Indicates that Intermediate Break is deleted (per MEB 3-1, R/2)

Rev. 18

WOLF CREEK TABLE 3.6-3 (Sheet 24) Historical Information SYSTEM - CHEMICAL AND VOLUME CONTROL SYSTEM Prob. No. P-069 Pipebreak Isometric No.: Figure 3.6-1 (BG02) Sheet 19 (BG10) Sheet 22 (BG09) Sheet 21 (EM02) Sheet 37 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) A92 8,117 11,489 19,609 39,610 425B 4,506 11,901 16,407 39,610 C92B* 4,924 1,952 6,876 39,610 D92* TNGT 13,774 6,720 20,494 39,610 575 13,044 1,384 14,428 39,616

  • - Indicates Terminal End Rev. 28

WOLF CREEK TABLE 3.6-3 (Sheet 24A) SYSTEM - CHEMICAL AND VOLUME CONTROL SYSTEM Prob. No P-069B Pipebreak Isometric No: Figure 3.6-1 (BG02) Sheet 19 Altran calc. No. 02101-C-004, Rev. 0 (BG10) Sheet 22 (BG09) Sheet 21 (EM02) Sheet 37 Stress (psi) Pipe Break Node Primary Secondary Total Stress Limit (psi) (EQN 9) (EQN 10) (EQN 9 + EQN 10) 0.8 (SA + 1.2 S h) 970* 6950 26093 33043 39,610 955 E 4797 3997 8794 39,610 900 5307 1992 7299 39,610 870 TNGT* 6203 5160 11362 39,610 95F TNGT ** 9678 18198 27876 39,610 95C 14274 14910 29184 39,610 885 M ** 4868 2597 7465 39,610 620* 6681 668 7350 39,610 625 TNGT ** 4756 962 5718 39,610 603 E 4684 9386 14069 39,610 601 TNGT ** 5105 11939 17044 39,610 641 6074 2071 8145 39,610 626 7077 5369 12447 39,610 62A* 7233 3407 10641 39,610 A92 ** 5602 1700 7302 39,610 425 3002 12832 15834 39,610 C92 B 7353 1163 8516 39,610 D92* TNGT 6924 1031 7954 39,610

  • - Indicates Terminal End
    • - Indicates that Intermediate Break is deleted (per MEB 3-1, Rev. 2)

Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 25) SYSTEM - CVCS MINIMUM CHGNG FLOW - AUX BLDG Prob. No. BG-5-007 Pipebreak Isometric No.: Figure 3.6-1 Issue - 0 (BG01) Sheet 18 (BG02) Sheet 19 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2 Sh) 22* 9,941 1863 11,804 39,448 265 8,460 3,065 11,525 39,448 16E 5,432 2,617 8,049 39,448 810* 4,846 2,313 7,159 39,448 800* 5,742 2,993 8,735 39,448 A50* 9,685 628 10,313 39,448 170* 12,575 5,409 17,984 39,448 185M Bend 5,485 4,656 10,141 39,448 265 7,448 1,941 9,389 39,448 230 8,454 5,333 13,787 39,448 18 7,348 13,494 21,177 39,448 

                           * - Indicates Terminal End Rev. 31

WOLF CREEK TABLE 3.6-3 (Sheet 25A) SYSTEM - CVCS MINIMUM CHGNG FLOW - AUX BLDG Prob. No. Pipebreak Isometric No.: Figure 3.6-1 BG-S-007-001-CN001 BG01 Sheet 18 BG02 Sheet 19 Node Stress (psi) Pipe Break Stress Limit Point Primary Secondary Total 0.8 (SA + 1.2Sh) (psi) 22* 6707 2029 8736 39448 265 5541 3049 8591 39448 16E 5136 2607 7743 39448 810B 5003 2580 7583 39448 800* 5384 3143 8527 39448 A50* 6975 6756 13731 39448 170* 10449 3121 13570 39448 185M 4622 4351 8973 39448 265 5783 1992 7776 39448 230 7309 4780 12089 39448 18 5859 13054 18913 39448 

  * - Indicated Terminal end NOTE:  All intermediate breaks are deleted per MEB 3-1, Rev. 2 criteria.

Rev. 32

WOLF CREEK TABLE 3.6-3 (Sheet 26) SYSTEM - CVCS MINIMUM CHGNG FLOW - AUX BLDG Prob. No. P-73B Pipebreak Isometric No.: Figure 3.6-1(BG01) Issue - 7 Sheet 18 (BG09) Sheet 21 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 22* TNGT 9,762 506 10,268 39,564 27 TNGT 7,186 654 7,840 39,564 28 TNGT 8,006 5,505 13,511 39,564 48 6,423 6,406 12,829 39,564 74A** 10,028 2,180 12,208 39,564 86** M Bend 9,314 5,494 14,808 39,564 193** 15,745 6,730 22,475 39,564 995* 5,079 31 5,110 39,564 834 8,204 7,578 15,782 39,564 846 TNGT 9,650 5,852 15,502 39,564 68M 4,801 2,329 7,130 39,564 56M 4,410 6,788 11,198 39,564 821 12,261 6,779 19,040 39,564

  • - Indicates Terminal End
    • - No Break Zone Rev. 32

WOLF CREEK TABLE 3.6-3 (Sheet 27) SYSTEM - CHEMICAL AND VOLUME CONTROL SYSTEM Prob. No. P-119 Pipebreak Isometric No.: Figure 3.6-1(BG22) Issue - 6 Sheet 25 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 45BB 5,349 13,958 19,307 37,244 47M 5,017 19,116 24,133 37,244 49 18,238 15,476 34,395 37,244 60T 15,515 7,989 23,504 37,244 145M 9,021 19,464 28,485 37,244 160M 10,232 21,632 31,864 37,244 220M 4,066 18,211 22,277 37,244 245E 4,087 19,998 24,085 37,244 270* 3,980 3,553 7,533 37,244

  • - Indicates Terminal End Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 28) SYSTEM - CHEMICAL AND VOLUME CONTROL Prob. No. P-139 Pipebreak Isometric No.: Figure 3.6-1(BG21) Issue - 5 Sheet 24 (BG24) Sheet 27 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 20* 13,641 9,820 23,461 37,240 65 TNGT 8,565 8,504 17,069 37,240 90 TNGT 8,540 11,166 19,706 37,240 100* 10,452 9,764 20,216 37,240 240M Bend 7,136 14,007 21,143 37,240 297* 7,146 3,900 11,046 37,240 215 TNGT 10,092 14,558 24,650 37,240 225M Bend 6,782 5,103 11,885 37,240 250B Bend 6,211 13,513 19,724 37,240

  • - Indicates Terminal End Rev. 11

WOLF CREEK TABLE 3.6-3 (Sheet 29) SYSTEM - CHEMICAL AND VOLUME CONTROL Prob. No. P-139 Pipebreak Isometric No.: Figure 3.6-1(BG21) Issue - 5 Sheet 24 (BG24) Sheet 27 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 280M Bend 7,992 6,451 14,443 37,240 444* 8,019 18,804 26,823 37,240 440 Bend 6,562 18,647 25,209 37,240 405 Bend 11,567 11,023 22,590 37,240 400T 15,646 15,351 30,997 37,240 70 TNGT 7,727 9,066 16,793 37,240 285M Bend 8,932 9,331 18,263 37,240

  • - Indicates Terminal End Rev. 11

WOLF CREEK TABLE 3.6-3 (Sheet 30) SYSTEM - CVCS AUXILIARY SPRAY Prob. No. P-140 Pipebreak Isometric No.: Figure 3.6-1(BG24) Issue - 5 Sheet 27 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 444T* TNGT 7,205 8,198 15,403 37,240 450M Bend 5,176 11,886 17,062 37,240 670 Bend 5,129 14,730 19,859 37,240 735M Bend 6,303 15,643 21,946 37,240 770E 4,961 2,141 7,102 37,240 771* 8,247 12,883 21,130 37,240 645 9,178 5,768 14,946 37,240 716A 7,972 20,565 28,587 37,240 620A 6,850 19,919 26,779 37,240

  • - Indicates Terminal End Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 31) SYSTEM - CHEMICAL AND VOLUME CONTROL Prob. No. P-145 Pipebreak Isometric No.: Figure 3.6-1(BG22) Issue - 5 Sheet 25 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 5* TNGT 9,781 333 10,114 37,200 77 7,599 1,377 8,976 37,200 25 13,788 19,220 33,008 37,200 40 17,437 20,592 38,029 37,200 45B Bend 7,419 5,020 12,439 37,200 105* 6,218 885 7,103 37,200 90 7,484 1,246 8,730 37,200 10 7,334 1,087 8,421 37,200

  • - Indicates Terminal End Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 32) Historical Information SYSTEM - CHEMICAL AND VOLUME CONTROL Prob. No. P-146 Pipebreak Isometric No.: Figure 3.6-1(BG22) Issue - 5 Sheet 25 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 5* 5,469 5,201 10,670 37,648 30T 7,194 23,858 31,052 37,648 35T 8,722 13,492 22,214 37,648 40T 8,741 10,360 19,101 37,648 44T 10,526 12,356 22,882 37,648 48 11,182 8,537 19,719 37,648 80T 12,578 6,878 19,456 37,648 102T 12,189 22,635 34,824 37,648 106 12,288 10,262 22,550 37,648 130T 15,138 6,132 21,270 37,648 202T 11,671 8,697 20,368 37,648 401* 15,986 22,849 38,835 37,648 315* 7,077 630 7,707 37,648

  • - Indicates Terminal End Rev. 18

WOLF CREEK TABLE 3.6-3 (Sheet 32A) SYSTEM - CHEMICAL AND VOLUME CONTROL Prob. No. CN-SMT-00-67 Pipebreak Isometric No.: Figure 3.6-1(BG22) Sheet 25 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.8Sh) 5* ++ ++ 17703 44900 30T** - - 35353 44900 35T** - - 38168 44900 40T** - - 30511 44900 44T** - - 13967 44900 48** - - 27826 44900 80T** - - 19441 44900 102T** - - 19867 44900 106** - - 28425 44900 130T** - - 17219 44900 202T** - - 12706 44900 401* ++ ++ ++ 44900 315* ++ ++ ++ 44900

  • Indicates Terminal End
             **              Indicates, that Intermediate Break is deleted (MEB 3-1, R/2)
             ++              Break as required by MEB 3-1.

Rev. 18

WOLF CREEK TABLE 3.6-3 (Sheet 33) SYSTEM - CVCS CHGNG AND EXCESS LETDOWN Prob. No. P-147 Pipebreak Isometric No.: Figure 3.6-1(BG23) Issue - 4 Sheet 26 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 5* 5,987 754 6,741 40,240 150M Bend 7,001 4,015 11,016 40,240 150E Bend 6,379 4,240 10,619 40,240 157 8,399 4,473 12,872 40,240 160E Bend 6,617 4,062 10,679 40,240 250M** Bend 5,993 2,700 8,693 40,240 300** 10,369 3,212 13,581 40,240

  • - Indicates Terminal End Rev. 23

WOLF CREEK TABLE 3.6-3 (Sheet 34) SYSTEM - STEAM GENERATOR BLOWDOWN Prob. No. P-196(1) Pipebreak Isometric No.: Figure 3.6-1(BM21) Issue - 2 Sheet 29 Prob. No. P-196(2) Issue - 1 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 5 TNGT 6,814 5,314 12,128 32,400 35 5,529 330 5,859 32,400 50 5,468 206 5,674 32,400 NOTE: This problem meets no Break Zone Criteria

  • - Indicates Terminal End Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 35) SYSTEM - STEAM GENERATOR BLOWDOWN Prob. No. P-197(1) Pipebreak Isometric No.: Figure 3.6-1(BM01) Issue - 2 Sheet 29 Prob. No. P-197(2) Issue - 1 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 5T 6,020 8,200 14,220 32,400 15 5,799 7,290 13,089 32,400 20 6,796 3,440 10,236 32,400 35 5,394 472 5,866 32,400 50 5,323 306 5,629 32,400 NOTE: This problem meets no Break Zone Criteria

  • - Indicates Terminal End Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 36) Historical Information SYSTEM - STEAM GENERATOR BLOWDOWN Prob. No. P-219 Pipebreak Isometric No.: Figure 3.6-1(BM01) Issue - 7 Sheet 29 (BM20) Sheet 35 (BM05) Sheet 31 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 5 4,755 16,362 21,117 32,400 20B 3,085 6,435 9,520 32,400 95 6,559 412 6,971 32,400 178* 17,893 20,834 38,727 32,400 A20 4,491 33,600 38,091 32,400 A30 3,863 16,605 20,468 32,400 B38 8,283 9,143 17,426 32,400 B60 8,162 11,297 19,459 32,400 A80 4,765 1,877 6,642 32,400 A63 7,333 1,011 8,344 32,400 192T 6,677 28,319 34,996 32,400 240* 3,614 383 3,997 32,400 203 2,783 4,180 6,963 32,400 260 4,818 1,328 6,146 32,400 255 5,480 1,233 6,713 32,400 E80 7,237 3,362 10,599 32,400

  • - Indicates Terminal End Rev. 18

WOLF CREEK TABLE 3.6-3 (Sheet 36A) SYSTEM - STEAM GENERATOR BLOWDOWN Prob. No. CN-SMT-01-9 Pipebreak Isometric No.: Figure 3.6-1(BM01) Sheet 29 (BM20) Sheet 35 (BM05) Sheet 31 Pipe Break Stress (psi) Stress Limit (psi) Node Primary Secondary Total 0.8 (SA + 1.8Sh) 5 - - 20533 39600 20B - - 12029 39600 95 - - 11085 39600 178* ++ ++ 30571 39600 A20** - - 34571 39600 A30 - - 24366 39600 B38 - - 21529 39600 B60 - - 24642 39600 A80 - - 7968 39600 A63 - - 9793 39600 192T** - - 37778 39600 240* ++ ++ 3608 39600 203 - - 6658 39600 260 - - 4934 39600 255 - - 5486 39600 E80 - - 7331 39600

  • - Indicates Terminal End
    • - Indicates, that Intermediate Break is deleted. (MEB 3-1, Rev. 2)

++ - Break as required by MEB 3-1. Rev. 25

WOLF CREEK TABLE 3.6-3 (Sheet 37) Historical Information SYSTEM - STEAM GENERATOR BLOWDOWN Prob. No. BM-S-002 (P-219) Pipebreak Isometric No.: Figure 3.6-1(BM01) Issue - 0 Sheet 29 (BM20) Sheet 25 (BM05) Sheet 31 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) F20 4,414 9,291 13,705 32,400 F25* 5,868 19,607 25,475 32,400 C40* 12,973 49,035 62,008 32,400 C35 8,364 26,846 35,210 32,400

  • - Indicates Terminal End Rev. 18

WOLF CREEK TABLE 3.6-3 (Sheet 37A) SYSTEM - STEAM GENERATOR BLOWDOWN Prob. No. CN-SMT-01-9 Pipebreak Isometric No.: Figure 3.6-1(BM01) Sheet 29 (BM20) Sheet 35 (BM05) Sheet 31 Pipe Break Stress (psi) Stress Limit (psi) Node Primary Secondary Total 0.8 (SA + 1.8Sh) F20 - - 10133 39600 F25* ++ ++ 20181 39600 C40* ++ ++ 46789 39600 C35 - - 24750 39600

  • - Indicates Terminal End

++ - Break as required by MEB 3-1. Rev. 25

WOLF CREEK TABLE 3.6-3 (Sheet 38) Historical Information SYSTEM - STEAM GENERATOR BLOWDOWN Prob. No. BM-S-003 (P-220) Pipebreak Isometric No.: Figure 3.6-1(BM01) Issue - 0 Sheet 29 (BM03) Sheet 31 (BM17) Sheet 32 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 5 6,796 2,597 9,393 32,400 85 9,521 7,261 16,782 32,400 145 3,372 2,208 5,580 32,400 155 9,035 29,447 38,482 32,400 165* 11,311 31,678 42,989 32,400 45B 5,662 29,264 34,926 32,400 195* 6,471 1,921 8,392 32,400 210 6,931 1,104 8,035 32,400 220 15,047 1,948 16,995 32,400 345* 5,211 14,312 19,523 32,400 349 8,781 31,728 40,509 32,400 350 5,613 14,057 19,670 32,400 352 4,334 13,963 18,297 32,400 365 5,026 16,317 21,343 32,400 400 16,645 9,399 26,044 32,400 435 13,446 37,686 51,132 32,400 467 10,115 23,717 33,832 32,400 470* 10,385 33,637 44,022 32,400

  • - Indicates Terminal End Rev. 18

WOLF CREEK TABLE 3.6-3 (Sheet 38A) SYSTEM - STEAM GENERATOR BLOWDOWN Prob. No. CN-SMT-01-10 Pipebreak Isometric No.: Figure 3.6-1(BM01) Sheet 29 (BM03) Sheet 31 (BM17) Sheet 32 Pipe Break Stress (psi) Stress Limit (psi) Node Primary Secondary Total 0.8 (SA + 1.8Sh) 5 - - 6844 39600 85 - - 38607 39600 145 - - 5303 39600 155*** - - 43342 39600 165* ++ ++ 37103 39600 45B*** - - 48992 39600 195* ++ ++ 4678 39600 210 - - 3329 39600 220* ++ ++ 13824 39600 345* ++ ++ 16760 39600 349*** - - 46918 39600 350 - - 22519 39600 352 - - 21063 39600 365 - - 18843 39600 400 - - 15186 39600 435** - - 11778 39600 467 - - 2411 39600 470* ++ ++ 32967 39600

  • - Indicates Terminal End

++ - Break as required by MEB 3-1.

    • - Indicates, that Intermediate Break is deleted.

(MEB 3-1, Rev. 2)

      • - Indicates, that Intermediate Break is required Rev. 25

WOLF CREEK TABLE 3.6-3 (Sheet 39) Historical Information SYSTEM - STEAM GENERATOR BLOWDOWN Prob. No. P-221 Pipebreak Isometric No.: Figure 3.6-1(BM02) Issue - 7 Sheet 30 (BM18) Sheet 33 (BM03) Sheet 18 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 10 4,776 4,836 9,612 32,400 20 8,810 6,180 14,990 32,400 25 8,026 3,388 11,414 32,400 105 4,909 2,319 7,228 32,400 135 13,731 1,136 14,867 32,400 140 11,955 1,434 13,389 32,400 165 10,705 3,063 13,768 32,400 175 11,891 2,239 14,130 32,400 225 4,591 2,539 7,130 32,400 240 4,923 4,748 9,671 32,400 255 4,019 5,380 9,399 32,400 100* 6,600 5,503 12,103 32,400 300 12,530 4,672 17,202 32,400 306 4,956 5,570 10,526 32,400 313 5,306 5,269 10,575 32,400 325 4,364 5,350 9,714 32,400 385 5,625 6,652 12,277 32,400 350 4,964 5,013 9,977 32,400 390* 12,459 22,786 35,245 32,400

  • - Indicates Terminal End Rev. 18

WOLF CREEK TABLE 3.6-3 (Sheet 39A) SYSTEM - STEAM GENERATOR BLOWDOWN Prob. No. CN-SMT-00-50 Pipebreak Isometric No.: Figure 3.6-1(BM02) Sheet 30 (BM18) Sheet 33 Pipe Break Stress (psi) Stress Limit (psi) Node Primary Secondary Total 0.8 (SA + 1.8Sh) 10** - - 7120 39600 20** - - 8959 39600 25** - - 8627 39600 105** - - 13275 39600 135** - - 6006 39600 140A** - - 7857 39600 165** - - 7988 39600 175** - - 36504 39600 225** - - 8276 39600 240** - - 10138 39600 255** - - 8546 39600 100* ++ ++ 20756 39600 300** - - 38600 39600 325** - - 17725 39600 385** - - 11946 39600 350** - - 10287 39600 390* ++ ++ 35779 39600

  • Indicates Terminal End
    • Indicates, that Intermediate Break is deleted.
      • Break as required by MEB 3-1.

Rev. 25

WOLF CREEK TABLE 3.6-3 (Sheet 40) Historical Information SYSTEM - STEAM GENERATOR BLOWDOWN Prob. No. CN-SMT-00-50 Pipebreak Isometric No.: Figure 3.6-1(BM02) Sheet 20 (BM18) Sheet 33 (BM03) Sheet 21 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 360 3,498 3,340 6,838 32,400 375 6,836 4,890 11,726 32,400 380 6,365 7,164 13,529 32,400 330 12,099 6,056 18,155 32,400 345 11,927 1,446 13,373 32,400 400 11,055 9,507 20,562 32,400 410 6,577 4,858 11,435 32,400 423 7,503 4,089 11,592 32,400 455 6,307 12,139 18,446 32,400 475 4,936 12,702 17,638 32,400 500 16,024 2,383 18,407 32,400 501 16,421 5,300 21,721 32,400 555* 15,386 30,025 45,411 32,400 520 10,931 2,136 13,067 32,400 523 6,026 1,848 7,874 32,400 545 5,313 13,634 18,947 32,400 550 9,409 20,609 30,018 32,400 440 8,542 11,117 19,659 32,400 462 6,233 29,455 35,688 32,400

  • - Indicates Terminal End Rev. 18

WOLF CREEK TABLE 3.6-3 (Sheet 40A) SYSTEM - STEAM GENERATOR BLOWDOWN Prob. CN-SMT-00-50 Pipebreak Isometric No.: Figure 3.6-1(BM02) Sheet 30 (BM18) Sheet 33 Pipe Break Stress (psi) Stress Limit (psi) Node Primary Secondary Total 0.8 (SA + 1.8Sh) 360** - - 6163 39600 375** - - 10762 39600 380** - - 13875 39600 330** - - 15705 39600 345** - - 10590 39600 400** - - 20916 39600 410** - - 11096 39600 423** - - 11214 39600 455** - - 15016 39600 475A** - - 10812 39600 500** - - 17879 39600 501A** - - 25294 39600 555* ++ ++ 36339 39600 520** - - 13725 39600 523** - - 14630 39600 545** - - 11021 39600 550** - - 17681 39600 440** - - 10177 39600 462** - - 19837 39600 550A** - - 23369 39600

  • Indicates Terminal End
    • Indicates, that Intermediate Break is deleted (MEB 3-1, R/2)

++ Break as required by MEB 3-1. Rev. 25

WOLF CREEK TABLE 3.6-3 (Sheet 41) Historical Information SYSTEM - STEAM GENERATOR BLOWDOWN Prob. No. P-221 Pipebreak Isometric No.: Figure 3.6-1(BM02) Issue - 7 Sheet 20 (BM18) Sheet 33 (BM03) Sheet 21 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 601 17,610 7,171 24,781 32,400 602 5,785 5,818 11,603 32,400 615 4,411 3,523 7,934 32,400 640* 8,409 1,733 10,142 32,400 605 6,239 5,663 11,902 32,400 637 7,774 1,167 8,941 32,400

  • - Indicates Terminal End Rev. 18

WOLF CREEK TABLE 3.6-3 (Sheet 41A) SYSTEM - STEAM GENERATOR BLOWDOWN Prob. CN-SMT-00-50 Pipebreak Isometric No.: Figure 3.6-1(BM02) Sheet 30 (BM03) Sheet 31 Pipe Break Stress (psi) Stress Limit (psi) Node Primary Secondary Total 0.8 (SA + 1.8Sh) 602** - - 9435 39600 615** - - 5963 39600 640* ++ ++ 6438 39600 605** - - 9377 39600 637** - - 6550 39600 615A** - - 6648 39600

  • Indicates Terminal End
    • Indicates, that Intermediate Break is deleted (MEB 3-1, R/2)

++ Break as required by MEB 3-1. Rev. 25

WOLF CREEK TABLE 3.6-3 (Sheet 42) Historical Information SYSTEM - STEAM GENERATOR BLOWDOWN Prob. No. CN-SMT-00-65 Pipebreak Isometric No.: Figure 3.6-1(BM02) Sheet 30 (BM03) Sheet 31 (BM19) Sheet 34 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) F10 12,803 3,033 15,836 32,400 A20T 16,282 20,916 37,198 32,400 305BB 8,638 11,797 20,435 32,400 C25 12,505 1,785 14,290 32,400 610 13,678 2,908 16,586 32,400 D20 14,305 10,435 24,741 32,400 302* 12,579 26,968 39,547 32,400 C77* 10,058 40,831 50,889 32,400 F60* 3,472 32,927 36,399 32,400 590 5,493 2,691 8,134 32,400 556T 17,337 8,672 26,009 32,400 C75M 6,707 26,877 33,584 32,400

  • - Indicates Terminal End Rev. 18

WOLF CREEK TABLE 3.6-3 (Sheet 42A) SYSTEM - STEAM GENERATOR BLOWDOWN Prob. No. CN-SMT-00-65 Pipebreak Isometric No.: Figure 3.6-1(BM02) Sheet 30 (BM03) Sheet 31 Pipe Break Stress (psi) Stress Limit (psi) Node Primary Secondary Total 0.8 (SA + 1.8Sh) F10** - - 9612 39600 A20T** - - 29753 39600 305BB** - - 23513 39600 C25** - - 7636 39600 610** - - 7718 39600 D20** - - 25518 39600 303* ++ ++ 43441 39600 C77 ++ ++ 32550 39600 F60* ++ ++ 32516 39600 590** - - 7469 39600 556T** - - 10855 39600 C75M** - - 19833 39600

  • Indicates Terminal End
    • Indicates, that Intermediate Break is deleted (MEB 3-1, R/2)

++ Break as required by MEB 3-1. Rev. 25

WOLF CREEK TABLE 3.6-3 (Sheet 43) Historical Information SYSTEM - MAIN STEAM ATMOSPHERIC DUMP LINE Prob. No. P-225 Pipebreak Isometric No.: Figure 3.6-1(AB01) Issue - 6 Sheet 1 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 520B Bend 7,816 19,035 26,851 32,400 545T 8,980 5,552 14,532 32,400 555B Bend 10,257 8,012 18,269 32,400 575 4 0 4 32,400 580T 7,945 10,808 18,753 32,400 NOTE: This problem meets no Break Zone Criteria.

  • - Indicates Terminal End Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 43A) SYSTEM - MAIN STEAM ATMOSPHERIC DUMP LINE Prob. No P-225 Pipebreak Isometric No: Figure 3.6-1 (AB01) Sheet 1 Altran Calc. No. 02101-C-006, Rev. 0 Stress (psi) Pipe Break Node Primary Secondary Total Stress Limit (psi) (EQN 9) (EQN 10) (EQN 9 + EQN 10) 0.8 (SA + 1.2 S h) 520B Bend 8,556 10,317 18,873 32,400 545T 9,390 5,721 15,112 32,400 555B Bend 8,091 5,664 13,755 32,400 575 0 0 0 32,400 580T 6,842 10,754 17,596 32,400 NOTE: This problem meets no Break Zone Criteria - Indicates Terminal End Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 44) SYSTEM - CHEMICAL AND VOLUME CONTROL Prob. No. P-254A Pipebreak Isometric No.: Figure 3.6-1(BG21) Issue - 3 Sheet 24 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 175* 7,033 3,674 10,707 37,244 170M Bend 7,523 3,240 10,763 37,244 170B Bend 7,214 2,869 10,083 37,244 155 6,784 6,413 13,197 37,244 140 6,907 13,563 20,470 37,244 130M Bend 5,388 21,301 26,689 37,244 105 6,526 9,440 15,966 37,244 95* 7,931 1,664 9,595 37,244

  • - Indicates Terminal End Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 45) SYSTEM - REACTOR COOLANT SYSTEM - REACTOR BLDG Prob. No. P-276 Issue - 2 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 40* 5,489 1,169 6,658 39,656 85 6,745 12,128 18,873 39,656 125 6,469 22,604 29,073 39,656 70 7,585 8,461 16,046 39,656 100 8,657 6,634 15,291 39,656 50B 7,387 6,998 14,385 39,656 55E 7,389 6,983 14,372 39,656 75E 5,140 8,352 13,492 39,656 80B 5,899 6,720 12,619 39,656

  • - Indicates Terminal End Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 46) SYSTEM - REACTOR COOLANT SYSTEM Prob. No. P-277 Pipebreak Isometric No.: Figure 3.6-1(BB09) Issue - 7 Sheet 14 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 35* 5,702 239 5,941 39,610 70M** Bend 6,432 5,898 12,330 39,610 70E** Bend 6,343 5,520 11,863 39,610 75 7,694 2,933 10,627 39,610 47E Bend 6,357 772 7,129 39,610 105T** 6,168 826 6,994 39,610 65B 6,467 2,251 8,718 39,610 55 5,786 2,750 8,536 39,610 

* - Indicates Terminal End
    • - Meets No Break Zone Criteria Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 47) SYSTEM - REACTOR COOLANT SYSTEM Prob. No. P-278 Pipebreak Isometric No.: Figure 3.6-1(BB11) Issue - 6 Sheet 15 Pipe Break Node Stress (psi) Stress Limit (psi) Primary Secondary Total 0.8 (SA + 1.2Sh) 20* 13,745 4,760 18,505 39,610 55M Bend 13,287 2,895 16,182 39,610 55E Bend 13,311 2,817 16,128 39,610 100M Bend 6,518 5,115 11,633 39,610 100B Bend 6,424 4,680 11,104 39,610 165 8,832 1,393 10,225 39,610 50 8,585 2,253 10,838 39,610 95 7,239 938 8,177 39,610 135 6,781 1,050 7,831 39,610 31 12,106 8,039 20,145 39,610

  • - Indicates Terminal End Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 48) (Historical Information) SYSTEM - ACCUMULATOR SAFETY INJECTION (LOOP 1) Prob. No. 234 Pipebreak Isometric No.: Figure 3.6-1 (EP01) Sheet 40 See SNP-6566 (AB27) Sheet 51 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 15* ++ ++ ++ ++ 30** 16,220 40,930 0.40 19,360 115** 49,838 29,247 0.47 46,440 450* ++ ++ ++ ++ 455** 39,187+ 13,206 0.001 46,440 485** 24,860 41,796 0.30 46,440 495* ++ ++ ++ ++ 665* ++ ++ ++ ++ 210* ++ ++ ++ ++ 955* ++ ++ ++ ++ 960** ++ ++ ++ ++ 975* ++ ++ ++ ++ 

*Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1, Rev 0 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 48A) Revised Stress Analysis Results SYSTEM - ACCUMULATOR SAFETY INJECTION (LOOP 1) Prob. No. 234 Pipebreak Isometric NO.: Figure 3.6-1 (EP01) Sheet 40 Issue - N/A Figure 3.6-1 (HB27) Sheet 51 See SAP-96-129 BECHTEL CUM. WESTINGHOUSE EQUATION 12 EQUATION 13 ALLOWABLE NODE USAGE NODE NUMBERS STRESS (KSI) STRESS (KSI) STRESS (KSI) NUMBERS FACTOR 15* 3020 ++ ++ ++ ++ 30** 3050 16.8 43.50 0.24 50.50 115** 3150 59.60 48.10 0.98 58.10 450* 5000 ++ ++ ++ ++ 455** 5003 485** 5070 53.6 48.00 0.33 58.10 495* 5100 ++ ++ ++ ++ 665* 4100 ++ ++ ++ ++ 210* 3340 ++ ++ ++ ++ 955* 6500 ++ ++ ++ ++ 960/970** 6510 ++ ++ ++ ++ 975* 6520 ++ ++ ++ ++

  • Terminal End Break
  **       Intermediate Break
  +        Break can be deleted per Arbitrary Break Elimination of MEB 3-1, Rev 2
  ++       Break as required by MEB 3-1 Rev. 0 Rev. 27

WOLF CREEK TABLE 3.6-3 (Sheet 49) (Historical Information) SYSTEM - ACCUMULATOR SAFETY INJECTION (LOOP 4) Prob. No. 235 Pipebreak Isometric No.: Figure 3.6-1 (EP01)Sheet 40 See SNP-6566 (HB27) Sheet 57 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 15* ++ ++ ++ ++ 35** 11,821 40,930 0.40 39,360 65** 40,262 30,865 0.47 46,440 348* ++ ++ ++ ++ 860* ++ ++ ++ ++ 360** 53,022 41,796 0.30 46,440 365* 16,175 10,800 0.001 46,440 720* ++ ++ ++ ++ 300* ++ ++ ++ ++ 405* ++ ++ ++ ++ 410** ++ ++ ++ ++ 425** ++ ++ ++ ++ 430* ++ ++ ++ ++ 

*Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1, Rev. 0 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 49A) Revised Stress Analysis Results System - Accumulator Safety Injection (Loop 4) Prob. No. 235 Pipebreak Isometric No: Figure 3.6-1(EP01)Sheet 40 Issue - N/A Figure 3.6-1 (HB27) Sheet 51 See SAP-96-129 BECHTEL CUM. WESTINGHOUSE EQUATION 12 EQUATION 13 ALLOWABLE NODE USAGE NODE NUMBERS STRESS (KSI) STRESS (KSI) STRESS (KSI) NUMBERS FACTOR 15* 3020 ++ ++ ++ ++ 35** 3080 18.8 40.9 0.24 50.50 65** 3590 42.2 35.0 0.98 58.10 348* 4070 ++ ++ ++ ++ 860* 3840 ++ ++ ++ ++ 360** 4320 39.5 58.9 0.33 58.10 365 + 4170 720* 4200 ++ ++ ++ ++ 300* 3542 ++ ++ ++ ++ 405* 5790 ++ ++ ++ ++ 410/420 5800 ++ ++ ++ ++ 425** 5820 ++ ++ ++ ++ 430* 5840 ++ ++ ++ ++

  • Terminal End Break
    • Intermediate Break

++ Break as Require by MEB 3-1 Rev. 0 + Break can be eliminate per Arbitrary Intermediate Break elimination (MEB 3 1, Revision 2) Rev. 27

WOLF CREEK TABLE 3.6-3 (Sheet 50) (Historical Information) SYSTEM - ACCUMULATOR SAFETY INJECTION (LOOP 3) Prob. No. 236 Pipebreak Isometric No.: Figure 3.6-1 (EP02) Sheet 41 See SNP-6566 (HB27) Sheet 51 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 15* ++ ++ ++ ++ 35** 12,014 40,930 0.40 39,360 85** 43,197 27,785 0.47 46,440 525* ++ ++ ++ ++ 450* ++ ++ ++ ++ 535** 9,326 11,690 0.001 46,440 550** 17,241 41,796 0.30 46,440 610* ++ ++ ++ ++ 205* ++ ++ ++ ++ 955* ++ ++ ++ ++ 960** ++ ++ ++ ++ 972** ++ ++ ++ ++ 975* ++ ++ ++ ++ 

*Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1, Rev. 0 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 50A) Revised Stress Analysis Results SYSTEM: ACCUMULATOR SAFETY INJECTION (LOOP 3) Prob. No. 236 Pipebreak Isometric No: Figure 3.6-1(EP02)Sheet 41 Issue - N/A Figure 3.6-1 (HB27) Sheet 51 See SAP-96-129 BECHTEL WESTINGHOUSE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE NODE NODE NUMBERS STRESS (KSI) STRESS (KSI) FACTOR STRESS (KSI) NUMBERS 15* 3020 ++ ++ ++ ++ 35** 3050 24.4 41.00 0.24 50.50 85** 3160 / 4000 41.2 30.50 0.98 58.10 525* 5000 ++ ++ ++ ++ 450* 4050 ++ ++ ++ ++ 535+ 5012 58.10 550 (1) 5040 55.3 59.60 0.33 ++ 610* 5070 ++ ++ ++ ++ 205* 3395 ++ ++ ++ ++ 955* 6500 ++ ++ ++ ++ 960/970** 6510 ++ ++ ++ ++ 972** 6525 ++ ++ ++ ++ 975* 6540 ++ ++ ++ ++

  • Terminal End Break
** Intermediate Break
+   Break can be deleted per Arbitrary Break Elimination of MEB 3-1, Rev. 2
++ Break as required by MEB 3-1 Rev. 0 (1)    Break reinstated. Usage Factor 7.1 and Stresses are above the allowable.

Rev. 27

WOLF CREEK TABLE 3.6-3 (Sheet 51) (Historical Information) SYSTEM - ACCUMULATOR SAFETY INJECTION (LOOP 2) Prob. No. 237 Pipebreak Isometric No.: Figure 3.6-1 (EPO2) Issue - N/A Sheet 41 See SNP-6566 (HB27) Sheet 51 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 12* ++ ++ ++ ++ 25** 15,142 40,930 0.40 39,408 110** 47,117 29,997 0.47 46,446 485* ++ ++ ++ ++ 445* ++ ++ ++ ++ 500* 30,524 16,451 0.001 46,440* 508* 47,747 51,850 0.30 46,440 570* ++ ++ ++ ++ 220* ++ ++ ++ ++ 905* ++ ++ ++ ++ 910** ++ ++ ++ ++ 925** ++ ++ ++ ++ 930* ++ ++ ++ ++ 

*Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 51A) Revised Stress Analysis Results SYSTEM - ACCUMULATOR SAFETY INJECTION (LOOP 2) Prob. No. 237 Pipebreak Isometric No: Figure 3.6-1 (EP02) Sheet 41 Issue - N/A Figure 3.6-1 (HB27) Sheet 51 See SAP-6566 BECHTEL WESTINGHOUSE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE NODE NODE NUMBERS STRESS (KSI) STRESS (KSI) FACTOR STRESS (KSI) NUMBERS 12* 3020 ++ ++ ++ ++ 25** 3050 14.3 41.00 0.24 50.50 110** 3160 45.2 43.10 0.98 58.10 485* 5000 ++ ++ ++ ++ 445* 4050 ++ ++ ++ ++ 500* 5030 508** 5050 40.9 59.80 0.33 58.10 570* 5060 ++ ++ ++ ++ 220* 3385 ++ ++ ++ ++ 905* 6500 ++ ++ ++ ++ 910** 6510 ++ ++ ++ ++ 925** 6530 ++ ++ ++ ++ 930* 6550 ++ ++ ++ ++

  • Terminal End Break
    • Intermediate Break

+ Break can be deleted per Arbitrary Break elimination of MEB 3-1, Rev. 2 ++ Break as required by MEB 3-1, Rev. 0 Rev. 27

WOLF CREEK TABLE 3.6-3 (Sheet 52) THIS SHEET HAS BEEN DELETED Rev. 6

WOLF CREEK TABLE 3.6-3 (Sheet 53) THIS SHEET HAS BEEN DELETED Rev. 6

WOLF CREEK TABLE 3.6-3 (Sheet 54) THIS SHEET HAS BEEN DELETED Rev. 6

WOLF CREEK TABLE 3.6-3 (Sheet 55) THIS SHEET HAS BEEN DELETED Rev. 6

WOLF CREEK TABLE 3.6-3 (Sheet 56) (Historical Information) SYSTEM - PRESSURIZER SPRAY Prob. No. 242 Pipebreak Isometric No.: Figure 3.6-1(BB04) Issue - N/A Sheet 9 See SNP-6566 Sheet 27 (BG24) NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 520* ++ ++ ++ ++ 10* ++ ++ ++ ++ 310* ++ ++ ++ ++ 580** 10,422 43,543 0.039 39,413 270* ++ ++ ++ ++ 270** 7,894 53,785 0.03 39,413 285** 38,200 27,023 0.126 39,413 285 to 305 20,174 10,070 0.390 39,413 305** 38,200 27,023 0.126 39,413 600** 4,664 48,374 0.21 39,413 

*Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 56A) Revised Stress Analysis Results SYSTEM: PRESSURIZER SPRAY PROBLEM NO. P - 242 PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 BB-S-036-000-CN005 (BB04) SHEET 9 WESTINGHOUSE CALC. NO. WCAP-9728, (BG24) SHEET 27 VOLUME IV (REVISION 2, ADDENDA) Westinghouse Maximum Maximum Cumulative Allowable Bechtel Westinghouse Westinghouse Piping Eq. 12 Stress Eq. 13 Stress Usage Stress 3Sm Node Nos. Node Nos. Piping Section Component (psi) (psi) Factor (psi) No. Name 520* 3020 1 4 TRA ++ ++ ++ ++ 10* 4020 1 4 TRA ++ ++ ++ ++ 310* 27 5 4 TRA Nozzle ++ ++ ++ ++ 580** 5100 2 2 x 3/4 TEA 7,000 52,400 0.883 52,500 270* 5010 2 6 x 2 BRA ++ ++ ++ ++ 270** 3570 4 6 RUP 14,400 45,700 0.168 50,500 285** 3640 - 3645 5 4 ELL 37,000 18,300 0.126 48,530 285 to 305 3645 to 3735 5 Various 23,300 14,000 0.400 48,530 305** 3730 5 4 ELL 37,000 18,300 0.126 48,530 600**+ 5150 2 2 TRA 7,900 35,400 0.093 52,500

  • Terminal End Break
    • Intermediate Break

++ Break as Require by MEB 3-1 Rev. 0 + Break can be eliminated per Arbitrary Intermediate Break elimination (MEB 3-1, Revision 2) Note: Westinghouse piping section numbers as specified above include more than one components with maximum cumulative usage factor. The stress values and cumulative usage factors listed above are the maximum values for a component. Rev. 26

WOLF CREEK TABLE 3.6-3 (Sheet 57) (Historical Information) SYSTEM - AUXILIARY PRESSURIZER SPRAY Prob. No. P-242 Pipebreak Isometric No.: Figure 3.6-1(BB04) Issue - N/A Sheet 9 See SNP-6566 (BG04) Sheet 27 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 615** 4,669 48,374 0.20 39,413 See Note 1 771* ++ ++ ++ ++ NOTE 1: Class 1 equations and allowables used although this break is on the Class 2 portion of the line. 

*Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 57A) Revised Stress Analysis Results 05 SYSTEM: PRESSURIZER SPRAY PROBLEM NO. P - 242 PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 BB-S-036-000-CN0 (BB04) SHEET 9 WESTINGHOUSE CALC. NO. WCAP-9728, (BG24) SHEET 27 VOLUME IV (REVISION 2, ADDENDA) Westinghouse Maximum Maximum Cumulative Allowable Bechtel Westinghouse Westinghouse Piping Eq. 12 Stress Eq. 13 Stress Usage Stress 3Sm Node Nos. Node Nos. Piping Section Component (psi) (psi) Factor (psi) No. Name 52,5 615** 5180 2 2 TRA 7,900 35,400 0.093 00 See Note 1 2 CLASS II 771* 5860 N/A ++ ++ ++ ++ PIPE NOTE 1: Class 1 equations and allowables used although this break is on the Class 2 portion of the line.

  • Terminal End Break
    • Intermediate Break

++ Break as Require by MEB 3-1 Rev. 0 Note: Westinghouse piping section numbers as specified above include more than one components with maximum cumulative usage factor. The stress values and cumulative usage factors listed above are the maximum values for a component. Rev. 26

WOLF CREEK TABLE 3.6-3 (Sheet 58) (Historical Information) SYSTEM - PRESSURIZER RELIEF Prob. No. 243A&B PIPE BREAK ISOMETRIC NO.: FIG. 3.6-1 (BB02) Issue - 0 Sheet 8 See SNP-6566 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 415** 17,900 38,800 0.8022 38,640 395 12,500 14,100 0.4112 38,640 to 375** 375 12,500 14,100 0.4112 38,640 to 340** 300* ++ ++ ++ ++ 415 12,500 14,100 0.4112 38,640 to 465** 465** 12,500 14,100 0.4112 28,640 to 500** 500** ++ ++ ++ ++ 

*Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 58A) Revised Stress Analysis Results SYSTEM: PRESSURIZER RELIEF PROBLEM NO. P-243A&B PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 BB-S-036-000-CN004 (BB02) SHEET 8 WESTINGHOUSE CALC. NO. WCAP-9728, VOLUME IV (REVISION 2, ADDENDA) Maximum Maximum Westinghouse Cumulative Allowable Bechtel Westinghouse Westinghouse Eq. 12 Eq. 13 Piping Component Usage Stress 3Sm Node Nos. Node Nos. Piping Section Stress Stress Name Factor (psi) No. (psi) (psi) 415** 4740 4 6 x 3 TEA 29,100 38,800 0.850 48,300 395 to 375** 4800 to 4850 4 3 ELL 37,500 28,400 0.100 48,300 375 to 340** 4850 to 4950 4 3 RUV 46,800 39,700 0.970 48,300 340* 4950 4 ++ ++ ++ ++ ++ 415 to 465** 4740 to 5150 4 6 x 3 TEA 29,100 38,800 0.850 48,300 465 to 500** 5150 to 5250 4 3 RUV 46,800 39,700 0.970 48,300 500* 5250 4 ++ ++ ++ ++ ++

  • Terminal End Break
 **   Intermediate Break location
 ++ Break as Required by MEB 3-1 Rev. 0 Note: Westinghouse piping section numbers as specified above include more than one components with maximum cumulative usage factor. The stress values and cumulative usage factors listed above are the maximum values for a component.

Rev. 26

WOLF CREEK TABLE 3.6-3 (Sheet 59) (Historical Information) SYSTEM - PRESSURIZER RELIEF Prob. No. 243A&B Pipebreak Isometric No.: Figure 3.6-1(BB02) Issue - N/A Sheet 8 See SNP-6566 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 175* ++ ++ ++ ++ 170** 60,000(a) 60,000(a) 0.163 38,640 165** 22,800 20,900 0.1054 38,640 165 8,500 14,200 0.294 38,640 to 160** 160 29,300 39,300 0.931 38,640 to 150** 150 8,500 20,900 0.1054 38,640 to 145** 145* ++ ++ ++ ++ 285* ++ ++ ++ ++ 275** 11,800 20,900 0.1054 38,640 275 8,500 14,200 0.294 38,640 to 270** 270 29,300 39,300 0.931 38,640 to 260** 260 8,500 14,200 0.294 38,640 to 255** 255* ++ ++ ++ ++ 5* ++ ++ ++ ++ 15** 22,800 20,900 0.1054 38,640 15 8,500 14,200 0.294 38,640 to 20** 20 29,000 39,300 0.931 38,640 to 30** 

*Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1 (a)As a result of constrained thermal expansion cycles. Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 59A) Revised Stress Analysis Results SYSTEM: PRESSURIZER RELIEF PROBLEM NO. P-243A&B PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 BB-S-036-000-CN004 (BB02) SHEET 8 WESTINGHOUSE CALC. NO. WCAP-9728, VOLUME IV (REVISION 2, ADDENDA) Maximum Maximum Westinghouse Cumulative Allowable Bechtel Westinghouse Westinghouse Eq. 12 Eq. 13 Stress Piping Component Usage 3Sm Node Nos. Node Nos. Piping Section Stress Stress Name Factor (psi) No. (psi) (psi) 175* 3010 1 ++ ++ ++ ++ ++ 170** 3020 1 ++ ++ ++ ++ ++ 165** 3040 1 6 ELL 47,400 25,400 0.700 48,300 165 to 160** 3040 to 3060 2 6 RUP 27,200 17,300 0.310 48,300 160 to 150** 3060 to 3090 2 6x 3/4 BRA 36,700 29,400 0.975 48,300 150 to 145** 3090 to 3130 2 6 RUF 24,700 35,800 0.910 48,300 145* 3130 2 ++ ++ ++ ++ ++ 285* 3810 1 ++ ++ ++ ++ ++ 275** 3840 1 6 ELL 47,400 25,400 0.700 48,300 275 to 270** 3840 to 3860 2 6 RUP 27,200 17,300 0.310 48,300 270 to 260** 3860 to 3890 2 6x 3/4 BRA 36,700 29,400 0.975 48,300 260 to 255** 3890 to 3930 2 6 RUF 24,700 35,800 0.910 48,300 255* 3930 2 ++ ++ ++ ++ ++

  • Terminal End Break
**   Intermediate Break location
++ Break as Required by MEB 3-1 Rev. 0 Note: Westinghouse piping section numbers as specified above include more than one components with maximum cumulative usage factor. The stress values and cumulative usage factors listed above are the maximum values for a component.

Rev. 26

WOLF CREEK TABLE 3.6-3 (Sheet 59B) (Historical Information) SYSTEM - PRESSURIZER RELIEF Prob. No. 243A&B Pipebreak Isometric No.: Figure 3.6-1(BB02) Sheet 8 Issue - N/A See SNP-6566 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) (PSI) 30 to 35** 8,500 14,200 0.294 38,640 35* ++ ++ ++ ++ 450* ++ ++ ++ ++

  • Terminal End Break
    • Intermediate Break

++ Break as required by MEB 3-1 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 59C) Revised Stress Analysis Results SYSTEM: PRESSURIZER RELIEF PROBLEM NO. P-243A&B PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 BB-S-036-000-CN004 (BB02) SHEET 8 WESTINGHOUSE CALC. NO. WCAP-9728, VOLUME IV (REVISION 2, ADDENDA) Maximum Maximum Westinghouse Cumulative Allowable Bechtel Westinghouse Westinghouse Eq. 12 Eq. 13 Piping Component Usage Stress 3Sm Node Nos. Node Nos. Piping Section Stress Stress Name Factor (psi) No. (psi) (psi) 5* 4300 1 ++ ++ ++ ++ ++ 15** 4340 2 6 ELL 47,400 25,400 0.700 48,300 15 to 20** 4340 to 4370 2 6 RUP 27,200 17,300 0.310 48,300 20 to 30** 4370 to 4400 2 6x 3/4 BRA 36,700 29,400 0.975 48,300 30 to 35** 4400 to 4440 2 6 RUF 24,700 35,800 0.910 48,300 35* 4440 2 ++ ++ ++ ++ ++ 450* 4640 3 ++ ++ ++ ++ ++

  • Terminal End Break
**   Intermediate Break location
++ Break as Required by MEB 3-1 Rev. 0 Note: Westinghouse piping section numbers as specified above include more than one components with maximum cumulative usage factor. The stress values and cumulative usage factors listed above are the maximum values for a component.

Rev. 26

WOLF CREEK TABLE 3.6-3 (Sheet 60) (Historical Information) SYSTEM - CVCS EXCESS LETDOWN Prob. No. 244 Pipebreak Isometric No.: Figure 3.6-1(BG23) Issue - N/A Sheet 26 See SNP-6566 (HB24) Sheet 50 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 5* ++ ++ ++ ++ 20** 36,138 22,210 0.290 39,413 30** 34,006 32,576 0.439 39,413 40** 200** 36,168 13,324 0.001 39,413 205* ++ ++ ++ ++ 415* ++ ++ ++ ++ 410** 32,796 14,631 0.005 39,413 400* ++ ++ ++ ++ 15** 36,138 32,355 0.367 39,413 

*Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 60A) Revised Stress Analysis Results SYSTEM: EXCESS LETDOWN / DRAIN LOOP 1 & 4 PROBLEM NO. P - 244 PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 BB-S-036-000-CN005 (BG23) SHEET 26 WESTINGHOUSE CALC. NO. WCAP-9728, (HB24) SHEET 50 VOLUME IV (REVISION 2, ADDENDA) Westinghous Maximum Maximum Cumulative Allowable Loop Bechtel Westinghouse Westinghouse e Piping Eq. 12 Eq. 13 Usage Stress 3Sm No. Node Nos. Node Nos. Piping Section Component Stress Stress Factor (psi) No. Name (psi) (psi) 5* 4015 1 2 TRA ++ ++ ++ ++ 2 x 3/4 15** 4050 1 23,100 46,600 0.099 50,500 TEE 2 X 2 20** 4080 1 43,800 20,800 0.099 50,500 4 TEE 30** & 40** 4100 & 4140 1 2 TRA 16,500 32,600 0.010 50,500 200** 6005 4 2 ELL/EL5 34,900 15,700 0.040 50,500 205* 6060 4 2 TRA ++ ++ ++ ++ 415* 3010 4 2 TRA ++ ++ ++ ++ 1 410** 3040 4 2 ELL 21,200 16,700 0.004 50,500 400* 3060 4 2 TRA ++ ++ ++ ++

  • Terminal End Break
    ** Intermediate Break
    ++ Break as Require by MEB 3-1 Rev. 0 Note: Westinghouse piping section numbers as specified above include more than one components with maximum cumulative usage factor. The stress values and cumulative usage factors listed above are the maximum values for a component.

Rev.26

WOLF CREEK TABLE 3.6-3 (Sheet 61) (Historical Information) SYSTEM - CVCS LETDOWN Prob. No. 245 Pipebreak Isometric No.: Figure 3.6-1(BG22) Issue - N/A Sheet 25 SEE SNP-6566 (HB24) Sheet 50 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 5* ++ ++ ++ ++ 10** 24,384 13,570 0.018 39,413 30** 50,693 <54,600 <1.0 39,413 50** 15,941 18,890 0.09 39,413 100** 7,476 18,890 0.09 39,413 205* ++ ++ ++ ++ 440* ++ ++ ++ ++ 435** 26,473 9,678 .012 39,413 430* ++ ++ ++ ++ 195** 50,693 <54,600 <1.0 39,413 125* ++ ++ ++ ++ 

*Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 61A) Revised Stress Analysis Results SYSTEM: NORMAL LETDOWN DRAIN LOOP 2 & 3 PROBLEM NO. P - 245 PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 BB-S-036-000-CN005 (BG22) SHEET 25 WESTINGHOUSE CALC. NO. WCAP-9728, (HB24) SHEET 50 VOLUME IV (REVISION 2, ADDENDA) Westinghouse Westinghouse Maximum Maximum Cumulative Allowable Bechtel Westinghouse Piping Piping Section Eq. 12 Stress Eq. 13 Stress Usage Stress 3Sm Node Nos. Node Nos. Component No. (psi) (psi) Factor (psi) Name 5* 3020 1 3 TRA ++ ++ ++ ++ 10** 3040 1 3 ELL 23,800 22,156 0.015 50,500 30** 3100 2 3 x 2 BRA 42,100 44,584 0.968 50,500 50** 3200 2 3 TRA 17,200 37,000 0.098 50,500 100** 3380 2 3 TRA 17,200 37,000 0.098 50,500 205* 3940 4 2 TRA ++ ++ ++ ++ 440* 3520 4 2 TRA ++ ++ ++ ++ 435** 3530 4 2 EL5 24,300 21,356 0.031 50,500 430* 3550 5 2 TRA ++ ++ ++ ++ 195** 3910 4 2 RUP 22,700 20,603 0.014 50,500 125* 3430 2 3 TRA ++ ++ ++ ++

  • Terminal End Break
 ** Intermediate Break
 ++ Break as Require by MEB 3-1 Rev. 0 Note: Westinghouse piping section numbers as specified above include more than one components with maximum cumulative usage factor. The stress values and cumulative usage factors listed above are themaximum values for a component.

Rev. 26

WOLF CREEK TABLE 3.6-3 (Sheet 61B) THIS SHEET HAS BEEN DELETED Rev. 26

WOLF CREEK TABLE 3.6-3 (Sheet 62) THIS SHEET HAS BEEN DELETED Rev. 0

WOLF CREEK TABLE 3.6-3 (Sheet 63) (Historical Information) SYSTEM - HPCI Prob. No. 247 & 247A Pipebreak Isometric No.: Figure 3.6-1(EM03) Issue - N/A Sheet 38 See SNP-6566 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 305* ++ ++ ++ 40,560 310** 20,400 20,000 0.052 40,416 320** 20,400 20,000 0.052 40,416 325* ++ ++ ++ ++ 210* ++ ++ ++ ++ 215** 20,400 20,000 0.052 40,416 225** 20,400 20,000 0.052 40,416 235* ++ ++ ++ ++ 105* ++ ++ ++ ++ 120** 20,400 20,000 0.052 40,416 125** 20,400 20,000 0.052 40,416 130* ++ ++ ++ ++ 10* ++ ++ ++ ++ 25** 20,400 20,000 0.052 40,416 30** 20,400 20,000 0.052 40,416 35* ++ ++ ++ ++ 

*Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 63A) Revised Stress Analysis Results SYSTEM: BIT LOOP 1, 2, 3 & 4 PROBLEM NO. P - 247A PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 BB-S-036-000-CN004 (EM03) SHEET 38 WESTINGHOUSE CALC. NO. WCAP - 9728, VOLUME IV (REVISION 2, ADDENDA) Westinghouse Maximum Maximum Cumulative Allowable Bechtel Westinghouse Westinghouse Piping Eq. 12 Stress Eq. 13 Stress Usage Stress 3Sm Node Nos. Node Nos. Piping Section Component (psi) (psi) Factor (psi) No. Name 305* Loop 1 3020 1 ++ ++ ++ ++ ++ 310** Loop 1 3035 1 1 1/2 EL5 33,400 25,605 0.050 50,500 320** Loop 1 3060 1 1 1/2 EL5 33,400 25,605 0.050 50,500 325* Loop 1 3080 1 ++ ++ ++ ++ ++ 210* Loop 2 4020 1 ++ ++ ++ ++ ++ 215** Loop 2 4040 1 1 1/2 EL5 33,400 25,605 0.050 50,500 225** Loop 2 4080 1 1 1/2 EL5 33,400 25,605 0.050 50,500 235* Loop 2 4110 1 ++ ++ ++ ++ ++ 105* Loop 3 5020 1 ++ ++ ++ ++ ++ 120** Loop 3 5080 1 1 1/2 EL5 33,400 25,605 0.050 50,500 125** Loop 3 5110 1 1 1/2 EL5 33,400 25,605 0.050 50,500 130* Loop 3 5140 1 ++ ++ ++ ++ ++ 10* Loop 4 6020 1 ++ ++ ++ ++ ++ 25** Loop 4 6060 1 1 1/2 EL5 33,400 25,605 0.050 50,500 30** Loop 4 6080 1 1 1/2 EL5 33,400 25,605 0.050 50,500 35* Loop 4 6110 1 ++ ++ ++ ++ ++

  • Terminal End Break
 ** Intermediate Break
 ++ Break as Required by MEB 3-1 Rev. 0 Note: Westinghouse piping section numbers as specified above include more than one components with maximum cumulative usage factor. The stress values and cumulative usage factors listed above are the maximum values for a component.

Rev. 26

WOLF CREEK TABLE 3.6-3 (Sheet 64) (Historical Information) SYSTEM - RHR AND HPCI Prob. No. 248A Pipebreak Isometric No.: Figure 3.6-1(EM03) Issue - N/A Sheet 38 See SNP-6566 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 175* ++ ++ ++ ++ 170** 15,854 15,340 <0.1 38,880 165* ++ ++ ++ ++ 290* ++ ++ ++ ++ 285** 18,638 16,848 <0.1 38,880 280* ++ ++ ++ ++ 

*Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 64A) Revised Stress Analysis Results SYSTEM: SI HOT LEG LOOP 2 AND 3 PROBLEM NO. P - 248A PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 BB-S-036-000-CN005 (EM03) SHEET 38 WESTINGHOUSE CALC. NO. WCAP- 9728, VOLUME IV (REVISION 2, ADDENDA) Westinghouse Maximum Maximum Allowable Bechtel Westinghouse Westinghouse Piping Cumulative Eq. 12 Stress Eq. 13 Stress Stress 3Sm Node Nos. Node Nos. Piping Section Component Usage Factor (psi) (psi) (psi) No. Name 175* 3020 1 6 TRA ++ ++ ++ ++ 170** 3040 1 6 ELL 24,600 22,300 0.011 50,500 165* 3060 1 6 TRA ++ ++ ++ ++ 290* 3620 1 6 TRA ++ ++ ++ ++ 285** 3640 1 6 ELL 24,600 22,300 0.011 50,500 280* 3670 1 6 TRA ++ ++ ++ ++

  • Terminal End Break
  ** Intermediate Break
  ++ Break as Require by MEB 3-1 Rev. 0 Note: Westinghouse piping section numbers as specified above include more than one components with maximum cumulative usage factor. The stress values and cumulative usage factors listed above are the maximum values for a component.

Rev. 26

WOLF CREEK TABLE 3.6-3 (Sheet 65) (Historical Information) SYSTEM - SEAL INJECTION (LOOP 4) Prob. No. 249 Pipebreak Isometric No.: Figure 3.6-1 (BB07) Sheet 12 See SNP-6566 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 10* ++ ++ ++ ++ 35** 12,043 10,915 0.0302 48,000 65** 12,043 10,915 0.0302 48,000 

*Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1, Rev. 0 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 65A) Revised Stress Analysis Results SYSTEM: SEAL INJECTION (LOOP 4) PROBLEM NO. P - 249 BB-S-036-000-CN004 PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 WESTINGHOUSE CALC. NO. WCAP-9728, (BB07) SHEET 12 VOLUME IV (REVISION 2, ADDENDA) Westinghouse Maximum Maximum Bechtel Westinghouse Cumulative Allowable Westinghouse Piping Eq. 12 Eq. 13 Node Piping Usage Stress 3Sm Node Nos. Component Stress Stress Nos. Section No. Factor (psi) Name (psi) (psi) 10

  • 3040 1 11/2" RUP ++ ++ ++ ++

35 **+ 3140 1 2 ELL 11,300 34,274 0.010 60,000 65**+ 3260 1 2 ELL 11,300 34,274 0.010 60,000 2 CLASS 2 Pen 40 4010 N/A PIPE Pen 40, this is a terminal end but is in the No Break Zone

  • Terminal End Break
  ** Intermediate Break
  ++ Break as Required by MEB 3-1 Rev. 0
  + Break can be eliminate per Arbitrary Intermediate Break elimination (MEB 3-1, Revision 2)

Note: Westinghouse piping section numbers as specified above include more than one components with maximum cumulative usage factor. The stress values and cumulative usage factors listed above are the maximum values for a component. Rev. 26

WOLF CREEK TABLE 3.6-3 (Sheet 66) (Historical Information) SYSTEM - RCP SEAL INJECTION (LOOP 1) Prob. No. 250 Pipebreak Isometric No.: Figure 3.6-1 (BB08) Sheet 13 See SNP-6566 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 7* ++ ++ ++ ++ 20** 6,375 34,200 0.0456 48,000 40** 6,375 34,200 0.0456 48,000 140** ++ ++ ++ ++

  • Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1, Rev. 0 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 66A) Revised Stress Analysis Results SYSTEM: RCP SEAL INJECTION (LOOP 1) PROBLEM NO. P - 250 BB-S-036-000-CN004 PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 WESTINGHOUSE CALC. NO. WCAP-9728, (BB08) SHEET 13 VOLUME IV (REVISION 2, ADDENDA) Westinghouse Maximum Maximum Cumulative Allowable Bechtel Westinghouse Westinghouse Piping Eq. 12 Stress Eq. 13 Stress Usage Stress 3Sm Node Nos. Node Nos. Piping Section Component (psi) (psi) Factor (psi) No. Name 7* 3120 1 ++ ++ ++ ++ ++ 20**+ 3160 1 2 TEA 26,600 23,700 0.015 60,000 40**+ 3240 1 2 TEA 26,600 23,700 0.015 68,000 2 CLASS 2 140* 3580 N/A ++ ++ ++ ++ PIPE

  • Terminal End Break
   **   Intermediate Break
   ++   Break as Require by MEB 3-1 Rev. 0
    +   Break can be eliminated per Arbitrary Intermediate Break elimination (MEB 3-1, Revision 2)

Note: Westinghouse piping section numbers as specified above include more than one components with maximum cumulative usage factor. The stress values and cumulative usage factors listed above are the maximum values for a component. Rev. 26

WOLF CREEK TABLE 3.6-3 (Sheet 67) (Historical Information) SYSTEM - RCP SEAL INJECTION (LOOP 3) Prob. No. 251 Pipebreak Isometric No.: Figure 3.6-1 (BB09) Sheet 14 See SNP-6566 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 7* ++ ++ ++ ++ 15** 8,988 47,035 0.0456 48,000 25** 8,988 47,035 0.4564 48,000 130** ++ ++ ++ ++

  • Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1, Rev 0 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 67A) Revised Stress Analysis Results SYSTEM: RCP SEAL INJECTION (LOOP 3) PROBLEM NO. P - 251 PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 WESTINGHOUSE CALC. NO. WCAP-9728, (BB09) SHEET 14 VOLUME IV (REVISION 2, ADDENDA) Westinghouse Maximum Maximum Cumulative Allowable Bechtel Westinghouse Westinghouse Piping Eq. 12 Stress Eq. 13 Stress Usage Stress 3Sm Node Nos. Node Nos. Piping Section Component (psi) (psi) Factor (psi) No. Name 7* 3020 1 ++ ++ ++ ++ ++ NEW + 3050 1 1 1/2 x 2 REA 19,800 59,400 0.076 60,000 15** + 3070 1 2 x 2 TEE 5,800 36,300 0.016 60,000 25** + 3100 1 2 x 2 TEE 5,800 36,300 0.016 60,000 2 CLASS 2 130* 4140 N/A ++ ++ ++ ++ PIPE

  • Terminal End Break
    **   Intermediate Break
    ++   Break as Require by MEB 3-1 Rev. 0
    +    Break can be eliminated per Arbitrary Intermediate Break elimination (MEB 3-1, Revision 2)

Note: Westinghouse piping section numbers as specified above include more than one components with maximum cumulative usage factor. The stress values and cumulative usage factors listed above are the maximum values for a component. Rev. 26

WOLF CREEK Table 3.6-3 (Sheet 68) (Historical Information) System - RCP Seal Injection (Loop 2) Prob. No. 252 Pipe Break Isometric No.: Figure 3.6-1(BB11) Sheet 15 Issue - N/A NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 7* ++ ++ ++ ++ 25** 7,845 41,114 0.0254 48,000 20** 14,330 32,699 0.0157 48,000

  • Terminal End Break
    • Intermediate Break

++ Terminal Break as Required by MEB 3-1 Rev. 19

WOLF CREEK Table 3.6-3 (Sheet 68A) Revised Stress Analysis Results System - RCP Seal Injection (Loop 2) Prob. No. 252 Pipe Break Isometric No.: Figure 3.6-1(BB11) Sheet 15 Issue - N/A NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE NO. STRESS (KSI) STRESS (KSI) FACTOR STRESS (KSI) 7* ++ ++ ++ ++ 20+ 25+ 110* ++ ++ ++ ++

  • Terminal End Break

+ Break can be deleted per Arbitrary Break Elimination of MEB 3-1, Rev. 2 ++ Break as Required by MEB 3-1, Rev. 0 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 69) (Historical Information) SYSTEM - CVCS Prob. No. 253 Pipebreak Isometric No.: Figure 3.6-1(BG21) Issue - N/A Sheet 24 See SNp-6566 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 5* ++ ++ ++ ++ 10** 26,126 40,000 0.4384 40,469 20** 50** 22,023 40,000 0.7734 40,469 30** 8,024 31,820 0.8337 40,469 45** 8,024 31,820 0.8337 40,469 60** 22,023 40,000 0.7734 40,469 See Note 1 105* ++ ++ ++ ++ See Note 2 NOTE 1: Class 1 equations and allowables used although this break is on the Class 2 portion of the line. NOTE 2: Class 2 break 

*Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 69A) Revised Stress Analysis Results SYSTEM: ALTERNATE CHARGING LOOP 4 PROBLEM NO. P - 253 PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 BB-S-036-000-CN004 (BG21) SHEET 24 WESTINGHOUSE CALC. NO. WCAP-9728, VOLUME IV (REVISION 2, ADDENDA) Westinghouse Maximum Maximum Cumulative Allowable Bechtel Westinghouse Westinghouse Piping Component Eq. 12 Stress Eq. 13 Stress Usage Stress 3Sm Node Nos. Node Nos. Piping Section Name (psi) (psi) Factor (psi) No. 5* 3020 1 3 TRA ++ ++ ++ ++ 10** & 20** 3030 & 3050 1 3 TRA 24,900 40,000 0.910 50,500 50** 3150 2 3 TRA 18,800 40,000 0.910 50,500 30** 4010 2 3 x 3/4 BRA 29,600 33,500 0.930 50,500 45** 5010 2 3 x 3/4 BRA 29,600 33,500 0.930 50,500 60** 3170 2 3 TRA 18,800 40,000 0.910 50,500 See Note 1 105* 3320 2 3 TRA ++ ++ ++ ++ See Note 2 NOTE 1: Class 1 equations and allowables used although this break is on the Class 2 portion of the line. NOTE 2: Class 2 break

  • Terminal End Break
  ** Intermediate Break
  ++ Break as Require by MEB 3-1 Rev. 0 Note: Westinghouse piping section numbers as specified above include more than one components with maximum cumulative usage factor. The stress values and cumulative usage factors listed above are the maximum values for a component.

Rev. 26

WOLF CREEK TABLE 3.6-3 (Sheet 70) (Historical Information) SYSTEM - CVCS Prob. No. 254 Pipebreak Isometric No.: Figure 3.6-1(BG21) Issue - N/A Sheet 24 See SNP-6566 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 10* ++ ++ ++ ++ 35** 23,225 40,000 0.4384 40,469 45** 55** 27,156 40,000 0.7734 40,469 17** 7,629 31,820 0.8337 40,469 52** 31,087 31,820 0.736 40,469 65** 27,156 40,000 0.7734 40,469 See Note 1 95* ++ ++ ++ ++ See Note 2 Note 1: Class 1 equations and allowables used although this break is on the Class 2 portion of the line. Note 2: Class 2 break 

*Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 70A) Revised Stress Analysis Results SYSTEM: NORMAL CHARGING LOOP 1 PROBLEM NO. P - 254 PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 BB-S-036-000-CN005 (BG21) SHEET 24 WESTINGHOUSE CALC. NO. WCAP-9728, VOLUME IV (REVISION 2, ADDENDA) Westinghouse Maximum Maximum Cumulative Allowable Bechtel Westinghouse Westinghouse Piping Component Eq. 12 Stress Eq. 13 Stress Usage Stress 3Sm Node Nos. Node Nos. Piping Section Name (psi) (psi) Factor (psi) No. 10* 3020 1 3 TRA ++ ++ ++ ++ 35** & 45** 3120 & 3150 1 3 TRA 28,100 40,000 0.910 50,500 55** 3230 2 3 TRA 28,200 40,000 0.910 50,500 17** 4010 1 3 x 3/4 BRA 9,200 31,800 0.900 50,500 52** 5010 2 3 X 1 BRA 31,700 31,800 0.930 50,500 65** 3260 2 3 TRA 28,200 40,000 0.910 50,500 See Note 1 95* 3400 2 3 TRA ++ ++ ++ ++ See Note 2 NOTE 1: Class 1 equations and allowables used although this break is on the Class 2 portion of the line. NOTE 2: Class 2 break

  • Terminal End Break
  ** Intermediate Break
  ++ Break as Require by MEB 3-1 Rev. 0 Note: Westinghouse piping section numbers as specified above include more than one components with maximum cumulative usage factor. The stress values and cumulative usage factors listed above are the maximum values for a component.

Rev. 26

WOLF CREEK TABLE 3.6-3 (Sheet 71) (Historical Information) SYSTEM - RHR Prob. No. 255 Pipebreak Isometric No.: Figure 3.6-1(EJ04) Issue - N/A Sheet 36 See SNP-6566 (EM05) Sheet 36 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 10* ++ ++ ++ ++ 15** 20,386 18,673 <0.1 38,880 20** 20,386 18,673 <0.1 38,880 195* ++ ++ ++ ++ 40* ++ ++ ++ ++ 200** 2,224 18,788 <0.1 38,880 205* ++ ++ ++ ++ 

*Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 71A) Revised Stress Analysis Results SYSTEM: RHR LOOP 1 PROBLEM NO. P - 255 PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 BB-S-036-000-CN004 (EJ04) SHEET 36 WESTINGHOUSE CALC. NO. WCAP-9728, (EM05) SHEET 39 VOLUME IV (REVISION 2, ADDENDA) Westinghouse Maximum Maximum Cumulative Allowable Bechtel Westinghouse Westinghouse Piping Eq. 12 Stress Eq. 13 Stress Usage Stress 3Sm Node Nos. Node Nos. Piping Section Component (psi) (psi) Factor (psi) No. Name 10* 3020 1 12 TRA ++ ++ ++ ++ 15** 3030 1 12 ELL 27,000 25,925 0.03 48,600 20** 3050 1 12 ELL 27,000 25,925 0.03 48,600 195* 4000 1 6 RUP ++ ++ ++ ++ 40* 3090 1 6 TRA ++ ++ ++ ++ 200** 4020 1 6 ELL 10,500 23,425 0.011 48,600 205* 4040 1 6 TRA ++ ++ ++ ++

  • Terminal End Break
    ** Intermediate Break
    ++ Break as Require by MEB 3-1 Rev. 0 Note: Westinghouse piping section numbers as specified above include more than one components with maximum cumulative usage factor. The stress values and cumulative usage factors listed above are the maximum values for a component.

Rev. 26

WOLF CREEK TABLE 3.6-3 (Sheet 72) (Historical Information) SYSTEM - RHR Prob. No. 256 Pipebreak Isometric No.: Figure 3.6-1(EJ04) Issue - N/A Sheet 36 See SNP-6566 (EMO5) Sheet 39 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 10* ++ ++ ++ ++ 15** 35,839 19,821 <0.1 38,880 20** 35,839 19,821 <0.1 38,880 195* ++ ++ ++ ++ 45* ++ ++ ++ ++ 200** 4,413 16,965 <0.1 38,880 220* ++ ++ ++ ++ 

*Terminal End Break
    • Intermediate Break

++Break as required by MEB 3-1 Rev. 19

WOLF CREEK TABLE 3.6-3 (Sheet 72A) Revised Stress Analysis Results SYSTEM: RHR LOOP 1 PROBLEM NO. P - 256 PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 BB-S-036-000-CN005 (EJ04) SHEET 36 WESTINGHOUSE CALC. NO. WCAP-9728, (EM05) SHEET 39 VOLUME IV (REVISION 2, ADDENDA) Westinghouse Maximum Maximum Cumulative Allowable Bechtel Westinghouse Westinghouse Piping Eq. 12 Stress Eq. 13 Stress Usage Stress 3Sm Node Nos. Node Nos. Piping Section Component (psi) (psi) Factor (psi) No. Name 10* 3020 1 12 TRA ++ ++ ++ ++ 15** 3030 1 12 ELL 31,300 29,700 0.030 48,600 20** 3060 1 12 ELL 31,300 29,700 0.030 48,600 195* 4000 1 6 RUP ++ ++ ++ ++ 45* 3180 2 12 TRA ++ ++ ++ ++ 200** 4020 1 6 ELL 4,000 18,456 0.011 48,600 220* 4400 1 6 TRA ++ ++ ++ ++

  • Terminal End Break
    ** Intermediate Break
    ++ Break as Require by MEB 3-1 Rev. 0 Note: Westinghouse piping section numbers as specified above include more than one components with maximum cumulative usage factor. The stress values and cumulative usage factors listed above are the maximum values for a component.

Rev. 26

WOLF CREEK TABLE 3.6-3 (Sheet 73) SYSTEM - REACTOR COOLANT SYSTEM PRIMARY LOOP Prob. No. 257 Pipebreak Isometric No.: Figure 3.6-3(BB01) Issue - N/A See SAP-91-165 NODE EQUATION 12 EQUATION 13 CUM. USAGE ALLOWABLE STRESS NO. STRESS (PSI) STRESS (PSI) FACTOR (2.4S_m_) PSI 2020* ++ ++ ++ ++ 2490* ++ ++ ++ ++ 

*Terminal End Break
    • Intermediate Break Location, Deleted Per MEB 3-1 Rev. 2-June 1987

++Break as required by MEB 3-1 Rev. 6

WOLF CREEK TABLE 3.6-3 (Sheet 73A) SYSTEM: PRESSURIZER SURGE LINE PROBLEM NO. P - 257 PIPE BREAK ISOMETRIC NO. FIGURE 3.6-3 (BB01) WESTINGHOUSE CALC. NO. WCAP-9728, VOLUME IV (REVISION 2, ADDENDA) Westinghouse Bechtel Maximum Maximum Cumulative Allowable Westinghouse Westinghouse Piping Node Nos. Eq. 12 Stress Eq. 13 Stress Usage Stress 2.4Sm Node Nos. Piping Section Component Note 1 (psi) (psi) Factor (psi) No. Name 2020* 3030 N/A RCL Nozzle ++ ++ ++ ++ N/A ** N/A N/A N/A ** N/A N/A 2490* 3530 N/A PZR Nozzle ++ ++ ++ ++

  • Terminal End Break
  **   Intermediate Break location, deleted per MEB 3-1 Rev. 2 June 1987
  ++ Break as Required by MEB 3-1 Rev. 0 Note: These node numbers are based on Westinghouse original loop analysis not Bechtel.

Rev. 26

WOLF CREEK TABLE 3.6-3 (Sheet 74) PROBLEM NO. FB-S-011 R/0 SYSTEM - AUX STEAM CONDENSATE TRANSFER PUMP DISCHARGE OUTSIDE CONTAINMENT PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 (FB10) SHEET 47 Node Stress (PSI) Pipe Break Primary Secondary Total Stress Limit (PSI) (EQN 12) (EQN 13) (EQN 12 + EQN 13) 0.8 (SA + 1.2 Sh) 5*TNGT 940 2,467 3,407 32,400 10 907 4,696 5,602 32,400 55 1,082 8,135 9,216 32,400 70 TNGT 1,164 2,388 3,552 32,400 75 610 1,292 1,902 32,400 80* 436 1,185 1,620 32,400 

       * - Indicates Terminal End Note:  Arbitrary Intermediate Breaks are not required.

Rev. 30

WOLF CREEK TABLE 3.6-3 (Sheet 75) PROBLEM NO. FB-S-010 R/0 SYSTEM - AUX STEAM DE-AERATOR FEED PUMP DISCHARGE OUTSIDE CONTAINMENT PIPE BREAK ISOMETRIC NO. FIGURE 3.6-1 (FB04) SHEET 43 AND (FB13) SHEET 48 Node Stress (PSI) Pipe Break Primary Secondary Total Stress Limit (PSI) (EQN 12) (EQN 13) (EQN 12 + EQN 13) 0.8 (SA + 1.2 Sh) 160* 5,522 6,460 11,982 32,400 180 TNGT 5,371 2,295 7,667 32,400 220* 6,260 6,411 12,671 32,400 240 TNGT 5,832 2,592 8,424 32,400 338 TNGT 2,470 1,072 3,543 32,400 342* 1,627 5,457 7,084 32,400 730* 12,363 9,439 21,802 32,400 724 TNGT 6,173 8,089 14,262 32,400 

       * - Indicates Terminal End Note: Arbitrary Intermediate Breaks are not required per MEB 3-1, Rev 2.

Rev. 31

WOLF CREEK TABLE 3.6-4 HIGH-ENERGY PIPE BREAK EFFECTS ANALYSIS RESULTS Room No. 1101 Elev. 1974'-0" General Floor Area No. 1 I. Sheets of Figure 3.6-1 showing high-energy 44, 45 (H-E) piping in this room II. Effects Analysis A. Room 1101; non-LOCA Breaks.

1. General: Breaks at all intermediate fittings (e.g.,

elbows, tees, reducers, welded attachments, and valves) in lines FB-032-HBD-8" having an auxiliary steam supply source and FB-050-HBD-3" having a condensate return source. No restrictions are considered in the calculation of thrust forces.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: 8-inch and 3-inch auxiliary steam piping restrained per Figure 3.6-1, Sheets 44, 45 such that no whipping occurs.
4. Jet impingement: Auxiliary steam and condensate jets impact safety-related targets required for post accident safe shutdown. Function of the essential targets is ensured.
5. Room pressurization: Breaks in the auxiliary steam supply header will result in peak local pressures greater than 0.2 psid; however, no structures, systems, or components required for post-accident safe shutdown will be adversely affected due to the short duration of the blowdown.
6. Temperature and humidity: Humidity is 100 percent following the breaks. The transient temperature is harsh and provides a limiting case for equipment qualification.

Rev. 19

WOLF CREEK TABLE 3.6-4 (Sheet 2) Room No. 1102 Elev. 1974'-0" Chiller and Surge Tank Area I. Sheet of Figure 3.6-1 showing high-energy 44, 45 (H-E) piping in this room II. Effects Analysis A. Room 1102; non-LOCA Breaks.

1. General: Breaks at all intermediate fittings (e.g.,

elbows, tees, reducer, welded attachments, and elbows) as follows: FB-032-HBD-8", with auxiliary steam supply source and FB-050-HBD-3" with condensate return source. No restrictions are considered in the calculation of thrust forces.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Nonsafety-related auxiliary steam piping whips such that no safety-related items are impacted.

Whip restraints are, therefore, not required.

4. Jet impingement: Jets do not impact any safety-related equipment in the area.
5. Room pressurization: Breaks in the auxiliary steam supply header will result in peak local pressures greater than 0.2 psid; however, no structures, systems, or components required for post-accident safe shutdown will be adversely affected due to the short duration of the blowdown.
6. Temperature and humidity: Humidity is 100 percent following the breaks. The transient temperature is harsh and provides a limiting case for equipment qualification.

Rev. 14

WOLF CREEK TABLE 3.6-4 (Sheet 3) Room No. 1104 Elev. 1974'-0" Letdown Reheat Heat Exchanger Room I. Sheets of Figure 3.6-1 showing high-energy 23 (H-E) piping in this room II. Effects Analysis A. Room 1104; non-LOCA Breaks.

1. General: Breaks BGll-07, 08 are non-LOCA breaks.

BGll-07 has sources from CVCS letdown off Loop 3 and from letdown reheat HX. BG11-08 has letdown reheat HX source with no loop source because of check valve 7039. No restrictions are considered in the calculation of thrust forces.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Nothing in this room is required for safe shutdown. Therefore, pipes are unrestrained and free to whip.
4. Jet impingement: No jet targets are required for safe shutdown.
5. Room pressurization: Adequate vent area is provided to ensure the integrity of all structures, systems, and components required for post-accident safe shutdown.

Items not required for post-accident safe shutdown will not fail in a manner that could affect post-accident safe shutdown equipment.

6. Temperature and humidity: No safety-related equipment is in the area; therefore, these breaks do not result in limiting temperature and humidity conditions for equipment qualification.

Rev. 14

WOLF CREEK TABLE 3.6-4 (Sheet 4) Room No. 1105 Elev. 1974'-0" Auxiliary Heat Exchanger Valve Compartment I. Sheets of Figure 3.6-1 showing high-energy 23 (H-E) piping in this room II. Effects Analysis A. Room 1105; non-LOCA Breaks.

1. General: Break BGll-06 is a non-LOCA break having sources from CVCS letdown off Loop 3 and from letdown reheat HX. No restrictions are considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Whip targets are uniquely associated with CVCS letdown flow path. Redundant letdown path available to ensure post-accident safe shutdown. Whip restraints are, therefore, not required.
4. Jet impingement: No jet targets are required for safe shutdown.
5. Room pressurization: Adequate vent area is provided to ensure the integrity of all structures, systems, and components required for post-accident safe shutdown.

Items not required for safe shutdown will not fail in a manner that could affect post-accident safe shutdown equipment.

6. Temperature and humidity: No safety-related equipment is in the area; therefore, these breaks do not result in limiting temperature and humidity conditions for equipment qualification.

Rev. 14

WOLF CREEK TABLE 3.6-4 (Sheet 5) Room No. 1107 Elev. 1974'-0" Centrifugal Charging Pump Room B I. Sheets of Figure 3.6-1 showing high-energy 19, 21, 22, 37 (H-E) piping in this room II. Effects Analysis A. Room 1107; non-LOCA Breaks.

1. General: Breaks BG02-04, 12* have CCP B source; no source from CCP A because of check valve 8481B. Breaks BG09-31*, 32, 33 have one source from CCP A/CCP B and no other source due to check valve BG-V589. Breaks BG02-13* and BG09-33 have sources from CCP B and from CCP A.

Breaks BG10-04, 05* on miniflow line have CCP B source; downstream source is moderate energy. Break EM02-07* has CCP B/CCP A source and moderate energy source downstream. No restrictions are considered in the calculation of thrust forces.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe Whip: EJFIS0611 is located such that it is not impacted. All other equipment in this room is uniquely associated with CCP B. Redundant charging path is available through CCP A. Whip restraints are, therefore, not required.
4. Jet impingement: No jet targets are required for safe shutdown.
5. Room pressurization: Cold water breaks only, P/T analysis not applicable.
6. Temperature and humidity: See 5 above.
  • The following intermediate breaks are deleted: BG02-12, BG02-13, BG09-31, BG10-05 & EM02-07.

Rev. 27

WOLF CREEK TABLE 3.6-4 (Sheet 6) Room No. 1114 Elev. 1974'-0" Centrifugal Charging Pump Room A I. Sheets of Figure 3.6-1 showing high-energy 19, 21, 22 (H-E) piping in this room II. Effects Analysis A. Room 1114; non-LOCA Breaks.

1. General: Breaks BG02-01 and BG09-35*, 36* have CCP A source; no source from CCP B because of check valves 8481A and V590, respectively. Break BG02-18 has both CCP A and CCP B source. Breaks BG10-01*, 03, & 06* on Miniflow line have CCP A source; downstream source is moderate energy. No restrictions are considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: All equipment in this room is uniquely associated with CCP A. Redundant charging path is available through CCP B. Whip restraints are, therefore, not required.
4. Jet impingement: No jet targets are required for safe shutdown.
5. Room pressurization: Cold water breaks only, P/T analysis not applicable.
6. Temperature and humidity: See 5 above.
  • Intermediate Breaks BG02-18, BG09-35, BG09-36, BG10-01, & BG10-06 are deleted.

Rev. 32

WOLF CREEK TABLE 3.6-4 (Sheet 7) Room No. 1115 Elev. 1974'-0" Normal Charging Pump Room I. Sheets of Figure 3.6-1 showing high-energy 18, 19, 21 (H-E) piping in this room II. Effects Analysis A. Room 1115; non-LOCA Breaks.

1. General: Breaks BG01-01, and 04* have a NCP source with no CCP A/B source because of check valve 8497. Break BG01-06 has a NCP source with a moderate energy source downstream of valve HV 8109. Breaks BG01-07*, 08, 09, 11, and 14 have both NCP and CCP A/B sources with no regenerative HX source because of check valve 8381. Breaks BG02-07, 10, and 11 and the downstream break on BG09-34 have both NCP and CCP A/B sources. The upstream break on BG09-34 has a NCP source only. No restrictions are considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: All equipment in this room is uniquely associated with Normal charging path. Redundant charging path is available through either CCP path. Whip restraints are, therefore, not required.
4. Jet impingement: No jet targets are required for safe shutdown.
5. Room pressurization: Cold water breaks only, P/T analysis not applicable.
6. Temperature and humidity: See 5 above.
  • Intermediate breaks BG01-04 and BG01-07 are deleted per MEB 3-1, Rev.

2, criteria. Rev. 32

WOLF CREEK TABLE 3.6-4 (Sheet 8) Contents of this page were moved to TABLE 3.6-4 (Sheet 7) Rev. 32

WOLF CREEK TABLE 3.6-4 (Sheet 9) Room No. 1117 Elev. 1974'-0" Boric Acid Tank Room B I. Sheets of Figure 3.6-1 showing high-energy 44, 45 (H-E) piping in this room II. Effects Analysis A. Room 1117; non-LOCA Breaks.

1. General: Breaks at all intermediate fittings (e.g.,

elbows, tees, reducers, welded attachments, and valves) in line FB-082-HBD-2" with a condensate return source. No restrictions are considered in the calculation of thrust forces.

2. Criteria: The non-LOCA break criteria has been met (See Note C)
3. Pipe whip: No items required for post-accident safe shutdown are impacted. Whip restraints are, therefore, not required.
4. Jet impingement: No jet targets are required for safe shutdown.
5. Room pressurization: Breaks in condensate return lines will not pressurize the area.
6. Temperature and humidity: No equipment in this room is required for post-accident safe shutdown; additionally, breaks in condensate return lines do not generate a harsh temperature or humidity environment.

Rev. 14

WOLF CREEK TABLE 3.6-4 (Sheet 10) Room No. 1122 Elev. 1974'-0" General Floor Area No. 3 I. Sheets of Figure 3.6-1 showing high-energy 37, 45 (H-E) piping in this room II. Effects Analysis A. Room 1122; non-LOCA Breaks.

1. General: Breaks at all intermediate fittings (e.g.,

elbows, tees, reducers, welded attachments, and valves) in lines FB-095-HBD-3" and FB-050-HBD-3" with condensate return source. No restrictions are considered in calculation of thrust forces.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: No items required for post-accident safe shutdown are impacted. Whip restraints are, therefore, not required.
4. Jet impingement: An 8-inch ESW line to the auxiliary feedwater system, et al, is impacted. Function of this essential line is ensured.
5. Room pressurization: Breaks in condensate return lines will not pressurize the area.
6. Temperature and humidity: Breaks in condensate return lines do not generate a harsh temperature or humidity environment.

Rev. 14

WOLF CREEK TABLE 3.6-4 (Sheet 11) Room No. 1124 Elev. 1974'-0" Letdown Heat Exchanger Valve Compartment I. Sheets of Figure 3.6-1 showing high-energy 20 (H-E) piping in this room II. Effects Analysis A. Room No. 1124; non-LOCA Breaks.

1. General: Breaks BG03-01, 02* and the branch break on BG03-03* have a combined source from CVCS letdown/

letdown. Break BG03-12* and the upstream and downstream breaks on BG03-03* have one source from CVCS letdown and one limited source from the letdown HX. Breaks BG03-09*, 10, 11, 13* and 16* have one source only - from CVCS letdown. The downstream source is moderate energy. No restrictions are considered in the calculation of thrust forces.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: All equipment in this room is uniquely associated with normal letdown. Redundant letdown is available for shutdown. Whip restraints are, therefore, not required.
4. Jet impingement: No jet targets are required for safe shutdown.
5. Room pressurization: Breaks in the CVCS letdown lines will result in peak local pressures greater than 0.2 psid; however, no structures, systems, or components required for post-accident safe shutdown will be adversely affected due to the short duration of the blowdown. Items not required for post-accident safe shutdown will not fail in a manner that could affect post-accident safe shutdown equipment.
6. Temperature and humidity: No post-accident safe shutdown equipment is in the area; therefore, these breaks do not result in limiting temperature and humidity conditions for equipment qualification.
  • The following intermediate breaks are deleted per MEB 3-1, Rev 2: BG03-02, BG03-03, BG03-12, BG03-09, BG03-13, & BG03-16.

Rev. 32

WOLF CREEK TABLE 3.6-4 (Sheet 12) Room No. 1125 Elev. 1974'-0" Letdown Heat Exchanger Room I. Sheets of Figure 3.6-1 showing high-energy 20, 23 (H-E) piping in this room II. Effects Analysis A. Room No. 1125; non-LOCA Breaks.

1. General: Break BG03-05 has one source from CVCS letdown and one limited source from the letdown HX.

Break BG03-15* has one combined source from CVCS letdown/letdown HX. Break BG03-06 has one source from CVCS letdown. No restrictions are considered in the calculation of thrust forces.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: All equipment in this room is uniquely associated with normal letdown. Redundant letdown is available for shutdown. Whip restraints are, therefore, not required.
4. Jet impingement: No jet targets are required for safe shutdown.
5. Room pressurization: Breaks in the CVCS letdown supplied lines will result in peak local pressures greater than 0.2 psid; however, no structures, systems, or components required for post-accident safe shutdown will be adversely affected due to the short duration of the blowdown. Items not required for post-accident safe shutdown will not fail in a manner that could affect post-accident safe shutdown equipment.
6. Temperature and humidity: No safety-related equipment is in the area; therefore, these breaks do not result in limiting temperature and humidity conditions for equipment qualification.
  • The intermediate break BG03-15 is deleted.

Rev. 19

WOLF CREEK TABLE 3.6-4 (Sheet 13) Room No. 1126 Elev. 1974'-0" Boron Injection Tank Room I. Sheets of Figure 3.6-1 showing high-energy 37 (H-E) piping in this room II. Effects Analysis A. Room 1126, non-LOCA Breaks.

1. General: Breaks EM02-06, 16 have CCP A source; EM02-05 has CCP B source. For all breaks, the BIT source downstream is moderate energy. No restrictions are considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Breaks EM02-05 and 06 are restrained per Figure 3.6-1, Sheet 37, such that whipping is prevented.
4. Jet impingement: All equipment in this room is uniquely associated with BIT and redundant means of boration exist for shutdown. Therefore, since any high energy break in the room will flood all the essential BIT equipment, jet impingement is not applicable.
5. Room pressurization: Cold water breaks only, P/T analysis not applicable.
6. Temperature and humidity: See 5 above.

Rev. 0

WOLF CREEK TABLE 3.6-4 (Sheet 14) Room No. 1127 Elev. 1974'-0" Stairwell A-2 I. Sheet of Figure 3.6-1 showing high-energy 45 (H-E) piping in this room II. Effects Analysis A. Room No. 1127; non-LOCA Breaks.

1. General: Breaks at all intermediate fittings (e.g.,

elbows, tees, reducers, welded attachments, and valves) in lines FB-050-HBD-3" and FB-095-HBD-3" having source from condensate return. The calculation of thrust forces is not required, since these condensate lines are open to atmospheric pressure.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: See Item 1 above.
4. Jet impingement: See Item 1 above.
5. Room pressurization: Breaks in the condensate return lines will not pressurize the area.
6. Temperature and humidity: Condensate water in these lines will not adversely affect any safety-related equipment required for post-accident safe shutdown.

Rev. 14

WOLF CREEK TABLE 3.6-4 (Sheet 15) Room No. 1128 Elev. 1974'-0" General Area No. 5

1. Sheets of Figure 3.6-1 showing high-energy 45 (H-E) piping in this room II. Effects Analysis A. Room No. 1128; non-LOCA Breaks.
1. General: Breaks at all intermediate fittings (e.g.,

elbows, tees, reducers, welded attachments, and valves) in lines FB-050-HBD-3" and FB-095-HBD-3" having source from condensate return. The calculation of thrust forces is not required, since these condensate lines are open to atmospheric pressure.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: See Item 1 above.
4. Jet impingement: See Item 1 above.
5. Room pressurization: Breaks in the condensate return lines will not pressurize the area.
6. Temperature and humidity: Condensate water in these lines will not adversely affect any safety-related equipment required for post-accident safe shutdown.
7. Any rise in temperature and humidity due to the postulated breaks in Room 1129 will not affect adversely any safety related equipment required for post-accident safe shutdown.

Rev. 31

WOLF CREEK TABLE 3.6-4 (Sheet 16) Room No. 1129 Elev. 1974'-0" Auxiliary Steam Con-densate Recovery and Storage Tank Room I. Sheet of Figure 3.6-1 showing high-energy 43, 45, 46, 47, 48 (H-E) piping in this room II. Effects Analysis A. Room No. 1129; non-LOCA Breaks.

1. General: Breaks at all intermediate fittings (e.g.,

elbows, tees, reducers, welded attachments, and valves) as follows: *line FB-110-HBD-2", lines FB-050, 095, 116-HBD-3", line FB-078-HBD-4", and lines FB-051, 052, 053-HBD-6" have condensate sources. The calculation of thrust forces is not required, since these lines carry condensate at low or atmospheric pressure. Lines FB-001, 054, 055-HBD-4" and lines FB-056, 057-HBD-2" have source from auxiliary steam deaerator feed pumps. These lines and FB-110-HBD-2 have been seismically analyzed and their stresses meet subsection 3.6.2.1.1.b.2 criteria. Hence, arbitrary intermediate breaks are not required to be postulated. Non-LOCA terminal end breaks FB04-02, FB04-03, FB13-01 and FB10-01 were postulated. No restrictions were considered in the calculation of thrust forces.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Pipe support FB10-H501 is designed to take pipe break load. Pipe supports FB13-H512, FB04-H001 &

FB04-H002 are designed to take pipe break loads. No essential equipment is impacted.

4. Jet impingement: No essential equipment is impacted by jets.
5. Room pressurization: Breaks in the condensate return lines will not pressurize the area. The pressure increases marginally due to terminal end breaks and has no impact on any essential equipment.
6. Temperature and humidity: Condensate water in these lines, a steam/condensate mixture in line FB-110-HBD-2 and lines FB-001, 054, 055-HBD-4" and lines FB-056, 057-HBD-2" will not adversely affect any safety-related equipment required for post-accident safe shutdown.
            *Line FB-110-HBD-2 has the potential to carry a mixture of condensate and steam and has been reclassified as high-energy line.

Rev. 36

WOLF CREEK TABLE 3.6-4 (Sheet 17) Room No. 1130 Elev. 1974'-0" North Corridor I. Sheets of Figure 3.6-1 showing high-energy 44, 45 (H-E) piping in this room II. Effects Analysis A. Room 1130; non-LOCA Breaks.

1. General: Breaks at all intermediate fittings (e.g.,

elbows, tees, reducers, welded attachments, and valves) as follows: FB-032-HBD-8" with auxiliary steam supply source, FB-095-HBD-3" and FB-050-HBD-3" with condensate return source. No restrictions are used in the calculation of thrust forces.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: No essential equipment is impacted. Whip restraints are, therefore, not required.
4. Jet impingement: No jet targets are required for safe shutdown.
5. Room pressurization: Breaks in the auxiliary steam supply header will result in peak local pressures greater than 0.2 psid; however, no structures, systems, or components required for post-accident safe shutdown will be adversely affected, due to the short duration of the blowdown.
6. Temperature and humidity: Humidity is 100 percent following the breaks. The transient temperature is harsh and provides a limiting case for equipment qualification.

Rev. 14

WOLF CREEK TABLE 3.6-4 (Sheet 18) High energy lines which were formerly in Room 1201 (Sheet 17), Room 1202 (Sheet 18), and Room 1321 (Sheet 27) have been declassified to moderate energy. Rev. 0

WOLF CREEK TABLE 3.6-4 (Sheet 19) Room No. 1203 Elev. 1988'-0" Pipe Space B I. Sheets of Figure 3.6-1 showing high-energy 19, 20, 21, 22, 23 (H-E) piping in this room II. Effects Analysis A. Room No. 1203; non-LOCA Breaks.

1. General: Breaks BG02-14, 15 have a CCP B source and a CCP A source. Breaks BG09-06, 07, 08, 19, 20, 23, 29, and 30; Break BG09-38, which is a Callaway only break; and Breaks BG09-41 and 42, which are Wolf Creek only breaks; have a charging pumps source only, since check valves BB-V118, V148, V178, and V208 are between the breaks and downstream source. Breaks BGll-02*, 03, 04, 05, and 13* have a CVCS letdown from Loop 3 source and a limited source from the letdown reheat heat exchanger.

Breaks BG09-01, 02, 12, 13 have CCP A and CCP B sources. Breaks BG11-09, 10, 11, 12 have a CVCS letdown from Loop 3 source only. There is no source in the opposite direction due to closed valve TCV-381A. No restrictions are considered in the calculation of thrust forces.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Breaks BG02-14, 15; BG09-01, 02, 12, 13; and BGll-02*, 03, 04, 10, 11, 12, 13* and the downstream break on BGll-09 are restrained per Figure 3.6-1, Sheets 21 and 23, such that no essential equipment is impacted.
4. Jet impingement: Two CVCS lines to the seal water injection filters, a CVCS CCP charging line, a CVCS CCP miniflow line, an RHR heat exchanger discharge line, and an RHR SI suction line are impacted by jets.

Function of all these essential lines is ensured.

  • The following intermediate breaks are deleted: BG11-02, BG11-13.

Rev. 19

WOLF CREEK TABLE 3.6-4 (Sheet 19 - cont)

5. Room pressurization: Breaks in the CVCS letdown line will result in pressures greater than 0.2 psid.

However, no post-accident safe shutdown equipment will be adversely affected due to the short duration of the blowdown.

6. Temperature and humidity: No equipment in this room is required for post-accident safe shutdown; therefore, the resultant temperature and humidity do not affect the qualification of any equipment.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 20) Room No. 1204 Elev. 1988'-0" Pipe Space A I. Sheets of Figure 3.6-1 showing high-energy 18, 19, 21, 37 (H-E) piping in this room II. Effects Analysis A. Room No. 1204; non-LOCA Breaks.

1. General: Break BG02-16* is located at a 3-inch tee.

The sources for the three break points are as follows: upstream - CCP B and CCP A, Branch (B) - CCP A and CCP B, downstream - CCP A/CCP B. Break BG02-17* has one CCP A/CCP B source. Break BG09-11 has two CCP A/CCP B combined sources. Breaks EM02-08, 09, 10, and 11 have a CCP B source with a moderate energy source downstream. Break BG09-40 is a Wolf Creek only break with a combined CCP A/CCP source from both directions. Breaks EM02-12*, 13, 14, and 15 have a CCP A source with a moderate energy source downstream. No restrictions are considered in the calculation of thrust forces.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: All three breaks on BG02-16*, the upstream break on BG02-17*, and break BG09-11 are restrained per Figure 3.6-1, Sheets 19, 21, and 23, such that no essential equipment is impacted.
4. Jet impingement: A CVCS CCP charging line, a CVCS CCP miniflow line, an ESW room cooler return line, and a CCW room cooler return line are impacted by jets.

Function of all these essential lines is ensured.

5. Room pressurization: Cold water breaks only, P/T analysis not applicable.
6. Temperature and humidity: See 5 above.
  • Intermediate Breaks BG02-16, BG02-17, EM02-12 are deleted.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 21) Room No. 1207 Elev. 1989'-0" Pipe Chase I. Sheets of Figure 3.6-1 showing high-energy 43, 46, 47 (H-E) piping in this room II. Effects Analysis A. Room No. 1207; non-LOCA Breaks.

1. General: Breaks at all intermediate fittings (e.g.,elbows, tees, reducers, welded attachments, and valves) as follows: FB-078-HBD-4" has a condensate return source with a moderate energy source downstream.

FB-110-HBD-2" has a condensate return source from both ends. It has potential to carry steam. FB-001-HBD-4" has an auxiliary steam deaerator feed pumps source with a moderate energy source from the auxiliary steam deaerator. Both lines have been reclassified as high-energy lines and analyzed seismically to meet stresses per subsection 3.6.2.1.1.b.2 criteria. Therefore, arbitrary intermediate breaks are not postulated on these lines. No restrictions are considered in the calculation of thrust forces. Terminal end break FB10-02 has a penetration designed such that no condensation/steam released due to a break will enter Room 1207. Non-LOCA terminal end break FB04-01 was postulated in adjacent Room 1329, which has an access opening for the ladder to Room 1207. However, the anchor at the penetration is designed such that no steam will enter Room 1329 in the event of a terminal end break.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: No essential equipment is impacted. Whip restraints are, therefore, not required. Pipe support FB10-H502 is designed to take pipe break load.
4. Jet impingement: No essential equipment is impacted by jets.
5. Room pressurization: - Cold water breaks only, P/T analysis not applicable. Penetration and anchor are designed such that no steam will enter Room 1207 in the event of a postulated terminal end break.
6. Temperature and humidity: See 5 above.

Rev. 32

WOLF CREEK TABLE 3.6-4 (Sheet 22) Room No. 1301 Elev. 2000'-0" Corridor No. 1 I. Sheets of Figure 3.6-1 showing high-energy 44, 45 (H-E) piping in this room II. Effects Analysis A. Room No. 1301; non-LOCA Breaks.

1. General: Breaks at all intermediate fittings (e.g.,

elbows, tees, reducers, welded attachments, and valves) as follows: FB-032-HBD-8" has an auxiliary steam supply source, FB-082-HBD-2", FB-096-HBD-3", and FB-095-HBD-3" have a condensate return source. No restrictions are considered in the calculation of thrust forces.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: The 8-inch auxiliary steam piping whips into non-safety-related equipment. Whip restraints are, therefore, not required.
4. Jet impingement: No essential equipment is impacted by jets.
5. Room pressurization: Breaks in the auxiliary steam supply header will result in peak local pressures greater than 0.2 psid; however, no structures, systems, or components required for post-accident safe shutdown will be adversely affected due to the short duration of the blowdown.
6. Temperature and humidity: Humidity is 100 percent following the breaks. The transient temperature is harsh and provides a limiting case for equipment qualification.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 23) Room No. 1302 Elev. 2000'-0" Filter Compartments 

                           - (5)

I. Sheets of Figure 3.6-1 showing high-energy 21 (H-E) piping in this room II. Effects Analysis A. Room No. 1302; non-LOCA Breaks.

1. General: Breaks BG09-09, 10, 14, and 15 have a charging pump source only since check valves BB-V118, V148, V178, and V208 are between the breaks and the downstream source. No restrictions are considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: All equipment in each compartment is uniquely associated with the seal water filters and a redundant path through CCW is available to the seals.

Whip restraints are, therefore, not required.

4. Jet impingement: No jet targets are required to ensure post-accident safe shutdown.
5. Room pressurization: Cold water breaks only, P/T analysis not applicable.
6. Temperature and humidity: See 5 above.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 24) Room No. 1304 Elev. 2013'-6" Auxiliary Feedwater Pipe Chase I. Sheets of Figure 3.6-1 showing high-energy (H-E) piping in this room NA II. Effects Analysis Safety-related piping in this room is associated with portions of the auxiliary feedwater system which experience high-energy conditions less than l percent of the plant operation time. This piping is therefore considered moderate energy per Section 3.6.1.la, and high-energy line breaks are not applicable.

1. General: No breaks are postulated in this room as noted above.
2. Criteria: N/A
3. Pipe whip: N/A
4. Jet impingement N/A
5. Room pressurization: The subcompartment pressurization analysis for the auxiliary feedwater valve compartments and pipe chases is based on a maximum break size of 3/4-inch nozzle on a back-up gas accumulator tank pressurized with N2 gas.

The results of the analysis indicate that the existing vent area is adequate to limit the room pressure to the design value of 1.5 psig.

6. Temperature and humidity: No extreme temperature or humidity environments are experienced as a result of the back-up gas accumulator tank nozzle break.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 25) Room No. 1305 Elev. 2013'-6" Auxiliary Feedwater Pipe Chase I. Sheets of Figure 3.6-1 showing high-energy NA (H-E) piping in this room II. Effects Analysis Safety-related piping in this room is associated with portions of the auxiliary feedwater system which experience high-energy conditions less than 1 percent of the plant operation time. This piping is therefore considered moderate energy per Section 3.6.1.la, and high-energy line breaks are not applicable.

1. General: No breaks are postulated in this room as noted above.
2. Criteria: N/A
3. Pipe whip: N/A
4. Jet impingement: N/A
5. Room pressurization: The subcompartment pressurization analysis for the auxiliary feedwater valve compartments and pipe chases is based on a maximum break size of 3/4-inch nozzle on a back-up gas accumulator tank pressurized with N2 gas.

The results of the analysis indicate that the existing vent area is adequate to limit the room pressure to the design value of 1.5 psig.

6. Temperature and humidity: No extreme temperature or humidity environments are experienced as a result of the back-up gas accumulator tank nozzle break.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 26) Room No. 1306 Elev. 2000'-0" Filter Valve Compart-ments - (5) I. Sheets of Figure 3.6-1 showing high-energy 21 (H-E) piping in this room II. Effects Analysis A. Room No. 1306; non-LOCA Breaks.

1. General: Break BG09-39 is a Callaway only break and has a charging pump source only since check valves "1" and "I" BB-V118, V148, V178, and V208 are located between the break and the downstream source. No restrictions are considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: All equipment in each compartment is uniquely associated with the seal water filters and a redundant path through CCW is available to the seals.

Whip restraints are, therefore, not required.

4. Jet impingement: No jet targets are required to ensure post-accident safe shutdown.
5. Room pressurization: Cold water breaks only, P/T analysis not applicable.
6. Temperature and humidity: See 5 above.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 27) Room No. 1322 (No Break Zone) - Elev. 2000'-0" Pipe Penetration Room B I. Sheets of Figure 3.6-1 showing high-energy 21 (H-E) piping in this room II. Effects Analysis A. Room No. 1322; No Postulated Breaks.

1. General: This area is a designated no break zone.

(See Section 3.6.2.1.le)

2. Criteria: NA
3. Pipe whip: None, no postulated breaks.
4. Jet impingement: Analysis is not applicable in "no break zone."
5. Room pressurization: No breaks, cold water cracks only, therefore P/T analysis is not applicable.
6. Temperature and humidity: See Section 5 above.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 28) Room No. 1323 (No Break Zone) Elev. 2000'-0" Pipe Penetration Room A I. Sheets of Figure 3.6-1 showing high-energy 18 (H-E) piping in this room II. Effects Analysis A. Room No. 1323; No Postulated Breaks.

1. General: This area is a designated no break zone.

(See Section 3.6.2.1.le)

2. Criteria: NA
3. Pipe whip: None, no postulated breaks.
4. Jet impingement: Analysis is not applicable in "no break zone."
5. Room pressurization: No breaks, cold water cracks, P/T analysis is not applicable.
6. Temperature and humidity: See 5 above.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 29) Room No. 1324 Elev. 2000'-0" Auxiliary Feedwater Pumps Valve Compartment No. 1 I. Sheets of Figure 3.6-1 showing high-energy NA (H-E) piping in this room II. Effects Analysis Safety-related piping in this room is associated with portions of the auxiliary feedwater system which experience high-energy conditions less than 1 percent of the plant operation time. This piping is therefore considered moderate energy per Section 3.6.1.la, and high-energy line breaks are not applicable.

1. General: No breaks are postulated in this room as noted above.
2. Criteria: N/A
3. Pipe whip: N/A
4. Jet impingement: N/A
5. Room pressurization: The subcompartment pressurization analysis for the auxiliary feedwater valve compartments and pipe chases is based on a maximum break size of 3/4-inch nozzle on a back-up gas accumulator tank pressurized with N2 gas.

The results of the analysis indicate that the existing vent area is adequate to limit the room pressure to the design value of 1.5 psig.

6. Temperature and humidity: No extreme temperature or humidity environments are experienced as a result of the back-up gas accumulator tank nozzle break.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 30) Room No. 1325 Elev. 2000'-0" Auxiliary Feedwater Pump Room B I. Sheets of Figure 3.6-1 showing high-energy NA (H-E) piping in this room II. Safety-related piping in this room is associated with portions of the auxiliary feedwater system which experience high-energy conditions less than 1 percent of the plant operation time. This piping is therefore considered moderate energy per Section 3.6.1.la, and high-energy line breaks are not applicable. Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 31) Room No. 1326 Elev. 2000'-0" Auxiliary Feedwater Pump Room A I. Sheets of Figure 3.6-1 showing high-energy NA (H-E) piping in this room II. Safety-related piping in this room is associated with portions of the auxiliary feedwater system which experience high-energy conditions less than 1 percent of the plant operation time. This piping is therefore considered moderate energy per Section 3.6.1.1a, and high-energy line breaks are not applicable. Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 32) Room No. 1327 Elev. 2000'-0" Auxiliary Feedwater Pump Valve Component No. 2 I. Sheets of Figure 3.6-1 showing high-energy NA (H-E) piping in this room II. Effects Analysis Safety-related piping in this room is associated with portions of the auxiliary feedwater system which experience high-energy conditions less than 1 percent of the plant operation time. This piping is therefore considered moderate energy per Section 3.6.1.1a, and high-energy line breaks are not applicable.

1. General: No breaks are postulated in this room as noted above.
2. Criteria: N/A
3. Pipe whip: N/A
4. Jet impingement: N/A
5. Room pressurization: The subcompartment pressurization analysis for the auxiliary feedwater valve compartments and pipe chases is based on a maximum break size of 3/4-inch nozzle on a back-up gas accumulator tank pressurized with N2 gas.

The results of the analysis indicate that the existing vent area is adequate to limit the room pressure to the design value of 1.5 psig.

6. Temperature and humidity: No extreme temperature or humidity environments are experienced as a result of the back-up gas accumulator tank nozzle break.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 33) Room No. 1328 Elev. 2000'-0" Auxiliary Feedwater Pump Valve Compartment No. 3 I. Sheets of Figure 3.6-1 showing high-energy NA (H-E) piping in this room II. Effects Analysis Safety-related piping in this room is associated with portions of the auxiliary feedwater system which experience high-energy conditions less than 1 percent of the plant operation time. This piping is therefore considered moderate energy per Section 3.6.1.la, and high-energy line breaks are not applicable.

1. General: No breaks are postulated in this room as noted above.
2. Criteria: N/A
3. Pipe whip: N/A
4. Jet impingement: N/A
5. Room pressurization: The subcompartment pressurization analysis for the auxiliary feedwater valve compartments and pipe chases is based on a maximum break size of 3/4-inch nozzle on a back-up gas accumulator tank pressurized with N2 gas.

The results of the analysis indicate that the existing vent area is adequate to limit the room pressure to the design value of 1.5 psig.

6. Temperature and humidity: No extreme temperature or humidity environments are experienced as a result of the back-up gas accumulator tank nozzle break.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 34) Room No. 1329 Elev. 2000'-0" Vestibule I. Sheets of Figure 3.6-1 showing high-energy 43 (H-E) piping in this room II. Effects Analysis A. Room 1329; Non-LOCA Breaks.

1. General: FB-001-HBD-4" has auxiliary steam deaerator feed pump source. No restrictions are considered in the calculation of thrust forces.

Arbitrary intermediate breaks in this line are not postulated as the line has been seismically analyzed and the stresses meet subsection 3.6.2.1.1.b.2 criteria. Non-LOCA terminal end break, FB04-01, is postulated in a specially designed penetration such that no steam will enter Room 1329 in the event of a terminal end break.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: No essential equipment is impacted. Pipe support FB04-H034 is designed to take pipe break load.
4. Jet impingement: No jet targets are required to assure post-accident safe shutdown.
5. Room pressurization: Penetration and anchor are designed such that no steam will enter Room 1329 in the event of a postulated terminal end break.
6. Temperature and humidity: See 5 above.

Rev. 31

WOLF CREEK TABLE 3.6-4 (Sheet 35) Room No. 1330 Elev. 2000'-0" Auxiliary Feedwater Pump Valve Compartment No. 4 I. Sheets of Figure 3.6-1 showing high-energy NA (H-E) piping in this room II. Effects Analysis Safety-related piping in this room is associated with portions of the auxiliary feedwater system which experience high-energy conditions less than 1 percent of the plant operation time. This piping is therefore considered moderate energy per Section 3.6.1.la, and high-energy line breaks are not applicable.

1. General: No breaks are postulated in this room as noted above.
2. Criteria: N/A
3. Pipe whip: N/A
4. Jet impingement: N/A
5. Room pressurization: The subcompartment pressurization analysis for the auxiliary feedwater valve compartments and pipe chases is based on a maximum break size of 3/4-inch nozzle on a back-up gas accumulator tank pressurized with N2 gas.

The results of the analysis indicate that the existing vent area is adequate to limit the room pressure to the design value of 1.5 psig.

6. Temperature and humidity: No extreme temperature or humidity environments are experienced as a result of the back-up gas accumulator tank nozzle break.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 36) Room No. 1331 Elev. 2000'-0" Auxiliary Feedwater Pump Room C I. Sheets of Figure 3.6-1 showing high-energy 49, 46 (H-E) piping in this room II. Effects Analysis A. Room 1331; Non-LOCA Breaks.

1. General: Breaks FC01-01, 02, 09, and 10 have main steam supply to turbine AFP source; source from auxiliary steam supply is considered moderate energy.

Breaks at all intermediate fittings (e.g., elbows, tees, reducers, welded attachments, and valves) in FB-078-HBD-4" have condensate return source. No restrictions are considered in the calculation of thrust forces for the FC breaks. No thrust force calculations are required on the FB line since it is at atmospheric pressure.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Breaks FC01-02, 09 and the upstream break on FC01-10 are restrained per Figure 3.6-1, Sheet 49, such that whipping is prevented.
4. Jet impingement: No jet targets are required to ensure post-accident safe shutdown.
5. Room pressurization: See Appendix 3B, Section 3.B.4.1.
6. Temperature and humidity: See Appendix 3B, Section 3.B.4.1.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 37) Room No. 1407 Elev. 2026'-0" Boric Acid Batching Tank I. Sheets of Figure 3.6-1 showing high-energy 45 (H-E) piping in this room II. Effects Analysis A. Room 1407; Non-LOCA Breaks.

1. General: Breaks at all intermediate fittings (e.g.,

elbows, tees, reducers, welded attachments, and valves) in FB-082-HBD-2" with condensate return source. No restrictions are considered in the calculation of thrust forces.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: No essential equipment is impacted. Whip restraints are, therefore, not required.
4. Jet impingement: No jet targets are required for safe shutdown.
5. Room pressurization: Cold water breaks only, P/T analysis not applicable.
6. Temperature and humidity: See 5 above.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 38) Room No. 1411 (No Break Zone) - Elevation 2026'-0" Main Steam/Main Feedwater Isolation Valve Compartment I. Sheets of Figure 3.6-1 showing high-energy l, 2, 3, 29, 30 (H-E) piping in this room II. Effects Analysis A. Room 1411; No Postulated Breaks.

1. General: This area is a designated no break zone.

(See Section 3.6.2.1.le)

2. Criteria: NA
3. Pipe whip: There is no pipe whip because there are no postulated breaks in the no break zone.
4. Jet impingement: See 3 above.
5. Room pressurization: See Appendix 3B, Section 3B.4.2.
6. Temperature and humidity: See Appendix 3B, Section 3B.4.2.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 39) Room No. 1412 (No Break Zone) - Elev. 2026'-0" Main Steam/Main Feedwater Isolation Valve Compartment I. Sheets of Figure 3.6-1 showing high-energy 1, 2, 3, 29, 30, 49 (H-E) piping in this room II. Effects Analysis A. Room 1412; No Postulated Breaks.

1. General: This area is a designated no break zone.

(See Section 3.6.2.1.le)

2. Criteria: NA
3. Pipe whip: There is no pipe whip because there are no postulated breaks in the no break zone.
4. Jet impingement: See 3 above.
5. Room pressurization: See Appendix 3B, Section 3B.4.2.
6. Temperature and humidity: See Appendix 3B, Section 3B.4.2.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 40) Room No. 2000 Main Steam I. Sheets of Figure 3.6-1 showing high-energy 1 (H-E) piping in this room II. Effects Analysis A. Problem No. 001, Steam Generator A, Secondary Systems Breaks.

1. General: Breaks AB01-01, 02, and 03 have sources from steam generator A and turbine building. No restrictions were considered in the calculation of thrust forces. Breaks AB01-02 and 03 can be deleted per arbitrary break elimination.
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 1, such that no whipping occurs.
4. Jet impingement: The jets from these breaks do not impact any essential systems.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a
7. Flooding: See Section 6.3.2.2 B. Problem No. 001A, Steam Generator B, Secondary Systems Breaks.
1. General: Breaks AB01-05, 06, and 07 have sources from steam generator B and turbine building. No restrictions were considered in the calculation of thrust forces. Breaks AB01-06 and 07 can be deleted per arbitrary break elimination.
2. Criteria: The secondary systems break criteria has been met. (See Note D)

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 40 - cont)

3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 1, such that no whipping occurs.
4. Jet impingement: The jets from these breaks do not impact any essential systems.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a C. Problem No. 002, Steam Generator D, Secondary Systems Breaks.
1. General: Breaks AB01-13, 14, and 15 have sources from steam generator D and turbine building. No restrictions were considered in the calculation of thrust forces. Breaks AB01-14 and 15 can be deleted per arbitrary break elimination.
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 1, such that no whipping occurs.
4. Jet impingement: The jets from these breaks do not impact any essential systems.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a D. Problem No. 002A, Steam Generator C, Secondary Systems Breaks.
1. General: Breaks AB01-09, 10, and 11 have sources from steam generator C and turbine building. No restrictions were considered in the calculation of thrust forces. Breaks AB01-10 and 11 can be deleted per arbitrary break elimination.
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 1, such that no whipping occurs.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 40 cont)

4. Jet impingement: The only target essential to mitigating the consequences of the break is a 10-inch component cooling water return line from the reactor coolant pumps (RCPs). Function of this essential system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 41) Room No. 2000 Main Feedwater I. Sheets of Figure 3.6-1 showing high-energy 2 (H-E) piping in this room II. Effects Analysis A. Problem No. 003, Steam Generator A, Secondary Systems Breaks.

1. General: Breaks AE04-01, 02, and 03 have sources from steam generator A and feedwater heaters. No restrictions were considered in the calculation of thrust forces. Breaks AE04-02 and 03 can be deleted per arbitrary break elimination.
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 2, such that no whipping occurs.
4. Jet impingement: The targets essential to mitigating the consequences of the breaks are the containment cooler C supply and return lines. Function of these essential systems is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a B. Problem No. 003A, Steam Generator B, Secondary Systems Breaks.
1. General: Breaks AE04-04, 05, and 06 have sources from steam generator B and feedwater heaters. No restrictions were considered in the calculation of thrust forces. Breaks AE04-05 and 06 can be deleted per arbitrary break elimination.
2. Criteria: The secondary systems break criteria has been met. (See Note D)

Rev. 31

WOLF CREEK TABLE 3.6-4 (Sheet 41 - cont)

3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 2, such that no whipping occurs.
4. Jet impingement: The targets essential to mitigating the consequences of the breaks are containment cooler C supply and return lines. Function of these essential systems is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a C. Problem No. 004A, Steam Generator C, Secondary Systems Breaks.
1. General: Breaks AE05-01, 02, and 03 have sources from steam generator C and feedwater heaters. No restrictions were considered in the calculation of thrust forces. Breaks AE05-02 and 03 can be deleted per arbitrary break elimination.
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 2, such that no whipping occurs.
4. Jet impingement: The targets essential to mitigating the consequences of the breaks are containment cooler A and C essential service water supply and return lines, RCP-B thermal barrier cooling coil inlet and outlet lines, and component cooling water supply and return header to RCP-B and C. Function of these essential systems is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 42) Room No. 2000 Main Feedwater I. Sheets of Figure 3.6-1 showing high-energy 3 (H-E) piping in this room II. Effects Analysis A. Problem No. 004, Steam Generator D, Secondary Systems Breaks

1. General: Breaks AE05-04, 05, and 06 have sources from steam generator D and feedwater heaters. No restrictions were considered in the calculation of thrust forces. Breaks AE05-05 and 06 can be deleted per arbitrary break elimination.
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 3, such that no whipping occurs.
4. Jet impingement: The targets essential to mitigating the consequences of the breaks are component cooling water supply and return header to RCP-A and D, and RCP-A thermal barrier cooling coil inlet and outlet lines.

Function of these essential systems is ensured.

5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 43) Room No. 2000 Reactor Coolant System -Pressurizer Relief I. Sheets of Figure 3.6-1 showing high-energy 8 (H-E) piping in this room II. Effects Analysis A. Problem No. 234A, Pressurizer-LOCA Breaks.

1. General: Breaks BB02-01, 02, 03, 04, 05, 06, 07, 08, 09, 10, 11, 12, 13, 14, 15, 16, 17, 18, 19, 20, 21, 22, 25, 27, 29, 30, and 31 are LOCA breaks having an H-E source from the pressurizer. The downstream source is moderate energy. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The large-LOCA break criteria has been met.

(See Note A)

3. Pipe whip: Whipping occurs. However, no essential systems are impacted. Whip restraints are not required.
4. Jet impingement: The jets from these breaks do not impact any essential systems.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 44) Room No. 2000 Pressurizer Spray I. Sheets of Figure 3.6-1 showing high-energy 9 (H-E) piping in this room II. Effects Analysis A. Problem No. 242, Loops No. 1 and 2, LOCA Breaks.

1. General: Break BB04-05 is a large LOCA break having sources from the RCS cold leg, Loops No. 1 and 2, the pressurizer, and the regenerative HX. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The large LOCA break criteria has been met.

(See Note A)

3. Pipe whip: The break is restrained per Figure 3.6-1, Sheet 9, such that no whipping occurs.
4. Jet impingement: The jet from this break does not impact any essential systems.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a
7. Flooding: See Section 6.3.2.2 B. Problem No. 242, Loops No. 1 and 2, LOCA Breaks.
1. General: Breaks BB04-01, 02, 07, 08, 09, 10, 11, 12, and 13 are LOCA breaks having sources from RCS cold leg, Loops No. l and 2, the pressurizer and the regenerative HX. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The small LOCA break criteria has been met.

(See Note B)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 9. Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: The only target essential to mitigating the consequences of the break is a 2-inch-high head safety-injection line to RCS hot leg Loop No.
1. Function of this essential system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 45) THIS PAGE HAS BEEN DELETED Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 46) THIS PAGE HAS BEEN DELETED Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 47) Room No. 2000 RCP-D Seal Injection I. Sheets of Figure 3.6-1 showing high-energy 12 (H-E) piping in this room II. Effects Analysis A. Problem No. 249, Loop No. 4 - LOCA Breaks.

1. General: Breaks BB07-01, 03, and 05 are LOCA breaks having sources from the RCS, Loop No. 4, and charging pumps. The thrust force calculation takes into account the fact that the charging pump source is restricted by a throttle valve in the injection line. No other restrictions are considered in the calculation of thrust forces. Breaks BB07-03 and 05 can be deleted per arbitrary break elimination. (Ref. Sect. 3.6.2.1)
2. Criteria: The small LOCA break criteria has been met.

(See Note B)

3. Pipe whip: whipping occurs; however, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 48) Room No. 2000 RCP-A Seal Injection I. Sheets of Figure 3.6-1 showing high-energy 13 (H-E) piping in this room II. Effects Analysis A. Problem No. 250, Loop No. 1 - LOCA Breaks.

1. General: Breaks BB08-09, 10, and 11 are LOCA breaks having sources from the RCS Loop No. 1 and the charging pumps. The thrust force calculation takes into account the fact that the charging pump source is restricted by a throttle valve in the injection line. No other restrictions are considered in the calculation of thrust forces. Breaks BB08-10 and 11 can be deleted per arbitrary break elimination (Ref. Sect. 3.6.2.1)
2. Criteria: The small LOCA break criteria has been met.

(See Note B)

3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 13. Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: The jets from these breaks do not impact any essential systems.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a B. Problem No. 250, Loop No. l - Non-LOCA Breaks.
1. General: Break BB08-04 has a source from the charging pumps only. No source available from RCP A due to double check valves BB-V120 and V121 located between the break and RCP A. The thrust force calculation takes into account the fact that the charging pump source is restricted by a throttle valve in the injection line.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 48 - cont)

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Pipe not capable of whipping due to low thrust force.
4. Jet impingement: The jet from this break does not impact any essential systems.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a C. Problem No. 276, Loop No. 1 - Non-LOCA Breaks.
1. General: Breaks BB08-03, 12, and 13 are non-LOCA breaks having source from the charging pumps only. No source available from RCP A due to double check valves BB-V120 and V121 located between the breaks and RCP A.

The thrust force calculation takes into account the fact that the charging pump source is restricted by a throttle valve in the injection line.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Pipe not capable of whipping due to low thrust force.
4. Jet impingement: The jets from these breaks do not impact any essential systems.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 49) Room No. 2000 RCP-C Seal Injection I. Sheets of Figure 3.6-1 showing high-energy 14 (H-E) piping in this room II. Effects Analysis A. Problem No. 251, Loop No. 3 - LOCA Breaks.

1. General: Breaks BB09-09, 10, and 11 are LOCA breaks having sources from the RCS Loop No. 3 and the charging pumps. The thrust force calculation takes into account that the charging pump source is restricted by a throttle valve in the injection line. No other restrictions are considered in the calculation of thrust forces. Breaks BB09-10, and 11 can be deleted per arbitrary break elimination. (Ref. Sect. 3.6.2.1).
2. Criteria: The small LOCA break criteria has been met.

(See Note B)

3. Pipe whip: Pipe geometry prevents whipping.
4. Jet impingement: The jets from these breaks do not impact any essential systems.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a B. Problem No. 251, Loop No. 3 - Non-LOCA Breaks.
1. General: Break BB09-04 has source from the charging pump only. No source available from RCP C due to double check valves BB-V180 and V181 located between break and RCP C. The thrust force calculation takes into account that the charging pump source is restricted by a throttle valve in the injection line.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 49 - cont)

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Pipe not capable of whipping due to low thrust force.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a C. Problem No. 277, Loop No. 3 - Non-LOCA Breaks.
1. General: Breaks BB09-03, 12, and 13 are non-LOCA breaks having source from the charging pumps only. No source available from RCP C due to double check valves BB-V180 and V181 located between breaks and RCP C. The thrust force calculation takes into account that the charging pump source is restricted by a throttle valve in the injection line.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Pipe not capable of whipping due to low thrust force.
4. Jet impingement: The targets essential to mitigating the consequences of the accident are seal injection to RCP-D and component cooling water injection (CCW) from the excess letdown heat exchanger. Function of these systems is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 50) Room No. 2000 RCP-B Seal Injection I. Sheets of Figure 3.6-1 showing high-energy 15 (H-E) piping in this room II. Effects Analysis A. Problem No. 252, Loop No. 2 - LOCA Breaks.

1. General: Breaks BBll-10, 11, and 12 are LOCA breaks having sources from the RCS Loop No. 2 and the charging pumps. The thrust force calculation takes into account that the charging pump source is restricted by a throttle valve in the injection line. No other restrictions are considered in the calculation of thrust forces. Breaks BB11-10 and 12 can be deleted per arbitrary break elimination.
2. Criteria: The small LOCA break criteria has been met.

(See Note B)

3. Pipe whip: Whipping occurs; however, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a B. Problem No. 252 Loop No. 2 - Non-LOCA Breaks.
1. General: Break BBll-05 has source from the charging pump only. No source available from RCP B due to double check valves BB-V150 and V151 located between break and RCP B. The thrust force calculation takes into account that the charging pump source is restricted by a throttle valve in the injection line.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 50 - cont)

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Pipe not capable of whipping due to low thrust force.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a C. Problem No. 278, Loop No. 2 - Non-LOCA Breaks.
1. General: Breaks BBll-02, 03, 04, 13 (Wolf Creek only),

and 14 (Callaway only) are non-LOCA breaks having source from the charging pumps only. No source available from RCP B due to double check valves BB-V150 and V151 located between breaks and RCP B. The thrust force calculation takes into account that the charging pump source is restricted by a throttle valve in the injection line.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Pipe not capable of whipping due to low thrust force.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 51) THIS PAGE HAS BEEN DELETED Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 52) THIS PAGE HAS BEEN DELETED Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 53) Room No. 2000 CVCS - Normal and Alternate Charging 

                           - Loops No. 1 and 4 I. Sheets of Figure 3.6-1 showing high-energy            24 (H-E) piping in this room II. Effects Analysis A. Problem No. 254, Loop No. 1-LOCA Breaks.
1. General: Breaks BG21-18, 22, and 23 are LOCA breaks having sources from the RCS cold leg Loop No. 1 and regenerative heat exchanger. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The small-LOCA break criteria has been met.

(See Note B)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 24. Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a B. Problem No. 254, Loop No. 1 - Non-LOCA Breaks.
1. General: Breaks BG21-24 and 25 have a source from the regenerative heat exchanger only. No source available from RCS Cold Leg Loop No. 1 due to double check valves BB-8378A and 8378B located between the breaks and Loop No. 1. No restrictions were considered in the calculation of thrust forces.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 53 - cont)

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 24. Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a C. Problem No. 254A, Loop 1 - Non-LOCA Breaks.
1. General: Breaks BG21-08, 09, 10, and 11 have a source from the regenerative heat exchanger only. No source available from RCS cold leg Loop No. 1 due to double check valves BB-8378A and 8378B located between the breaks and Loop No. 1. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 24, such that no whipping occurs.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a D. Problem No. 253, Loop No. 4 - LOCA Breaks.
1. General: Breaks BG21-12, 14, and 15 have sources from the RCS cold leg Loop No. 4 and regenerative heat exchanger. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The small-LOCA break criteria has been met.

(See Note B) Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 53 - cont)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 24. Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: The only target essential to mitigating the consequences of the break is the hot leg safety-injection line. Function of this essential system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a E. Problem No. 253, Loop No. 4 - Non-LOCA Breaks.
1. General: Breaks BG21-16 and 17 have a source from the regenerative heat exchanger only. No source is available from RCS cold leg Loop No. 4 due to double check valves BB-8379A and 8379B located between the breaks and Loop No. 4. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 24. Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: The only target essential to mitigating the consequences of the breaks is the 12-inch RHR pump suction, Loop No. 4. Function of this essential system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a F. Problem No. 139, Loops 1 and 4 - Non-LOCA Breaks.
1. General: Breaks BG21-01, 02, 03, 04, 05, 06, and 07 have a source from the regenerative heat exchanger only. No source available from RCS cold leg Loops No.

1 and 4 due to double check valves BB-8379A and TABLE Rev. 30

WOLF CREEK 3.6-4 (Sheet 53 - cont) 8379B located between the breaks and the loops. No restrictions were considered in the calculation of thrust forces.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 24, such that no whipping occurs.
4. Jet impingement: The only target essential to mitigating the consequences of the breaks is the CCW line from RCP-C thermal barrier cooling coil. Function of this essential system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 54) Room No. 2000 CVCS - Letdown I. Sheets of Figure 3.6-1 showing high-energy 25 (H-E) piping in this room II. Effects Analysis A. Problem No. 245, Loop No. 3 - LOCA Breaks.

1. General: Breaks BG22-19, 24, 25, 26, 27, and 28 are LOCA breaks having sources from RCS crossover leg Loop No. 3 and the regenerative heat exchanger. No restrictions were considered in the calculation of thrust forces. Breaks BB22-25, 24 and 19 can be deleted per arbitrary break elimination.
2. Criteria: The small LOCA break criteria has been met.

(See Note B)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 25. Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: The only target essential to mitigating the consequences of the breaks is seal injection to RCP-B. Function of this essential system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a B. Problem No. 245, Loop No. 3 - Non-LOCA Breaks.
1. General: Break BG22-18 is a non-LOCA break having source from regenerative heat exchanger. No source available from RCS crossover leg due to closure of one of two isolation valves, BG-LCV 459 and LCV 460. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C) Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 54 - cont)

3. Pipe whip: Break is restrained per Figure 3.6-1, Sheet 25, such that no whipping occurs.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a C. Problem No. 145, Loop No. 3 - Non-LOCA Breaks.
1. General: Breaks BG22-01, 02, 03, and 04 are non-LOCA breaks having source from regenerative heat exchanger.

No source available from RCS crossover leg due to closure of one of two isolation valves, BG-LCV459 and LCV460. No restrictions were considered in the calculation of thrust forces.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 25, such that no whipping occurs.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a D. Problem No. 146, Loop No. 3 - Non-LOCA Breaks.
1. General: Breaks BG22-05, 06*, 07*, 08, 09, and 13* are non-LOCA breaks having sources from regenerative heat exchanger and letdown heat exchanger. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

  • Intermediate Break BG22-06, BG22-07 and BG22-13 are deleted.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 54 - cont)

3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 25, such that no whipping occurs.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a E. Problem No. 119, Loop No. 3 - Non-LOCA Breaks.
1. General: Breaks BG22-10, 12, and 14 are non-LOCA breaks having sources from regenerative heat exchanger and letdown heat exchanger. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 25. Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: The targets essential to mitigating the consequences of the breaks are seal injection to RCP-B and C, and CCW from the excess letdown HX and the excess letdown line. Function of these essential systems is ensured. A shield, attached to whip restraint BG22-16, is provided to protect valve BGHV8153A from breaks BG-22-12 and 14.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 55) Room No. 2000 CVCS Charging and Excess Letdown I. Sheets of Figure 3.6-1 showing high-energy 26 (H-E) piping in this room II. Effects Analysis A. Problem No. 244, Loop No. 4 - LOCA Breaks.

1. General: Breaks BG23-11 are LOCA breaks having source from RCS crossover leg. The downstream sources are moderate energy. No restrictions were considered in the calculation of thrust forces. Breaks BB23-10 and 08 can be deleted per arbitrary break elimination.
2. Criteria: The small LOCA break criteria has been met.

(See Note B)

3. Pipe whip: Whipping occurs; however, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c B. Problem No. 147, Non-LOCA Breaks.
1. General: BG23-01, 02, and 03 are non-LOCA breaks having sources from the regenerative heat exchanger and the charging pumps. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 26. Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: The only target essential to mitigating the consequences of the breaks is a CCW line from RCP-B thermal barrier. Function of the essential system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 56) Room No. 2000 CVCS Auxiliary Spray I. Sheets of Figure 3.6-1 showing high-energy 27 (H-E) piping in this room II. Effects Analysis A. Problem No. 242, LOCA Break.

1. General: Break BG24-20 has sources from the RCS Cold Leg Loops No. 1 and 2 and the regenerative heat exchanger. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The small LOCA break criteria has been met.

(See Note B)

3. Pipe whip: Break is restrained per Figure 3.6-1, Sheet
27. Whipping occurs; however, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a B. Problem No. 242, Non-LOCA Breaks.
1. General: Breaks BG24-08 and 19 have a regenerative heat exchanger source only since the breaks are located between the regenerative heat exchanger and check valve BBV084. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C) Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 56 - cont)

3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 27, such that no whipping occurs.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a C. Problem No. 140, Non-LOCA Breaks.
1. General: Breaks BG24-03, 07, 15 and 17 (Callaway only), 16 and 18 (Wolf Creek only) have a regenerative heat exchanger source only, since the breaks are located between the regenerative heat exchanger and check valve BBV084. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 27. Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: The targets essential to mitigating the consequences of the breaks are seal injection to RCP-D, CCW from RCP-D thermal barrier cooling coil, and CCW line to the excess letdown HX and the excess letdown line. Function of these essential systems is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a D. Problem No. 139, Non-LOCA Breaks.
1. General: Breaks BG24-01 and 02 have a regenerative heat exchange source only, since the breaks are located between the regenerative heat exchanger and check valve BBV084. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C) Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 56 - cont)

3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 27, such that there are no whip targets.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 57) Room No. 2000 Steam Generator A&D Blowdown I. Sheets of Figure 3.6-1 showing high-energy 29 (H-E) piping in this room II. Effects Analysis A. Problem No. 219, Secondary Systems Breaks.

1. General: Break BM01-04 has a steam generator D source and a turbine building source. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Whipping occurs; however, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a B. Problem No. 220, Secondary Systems Breaks
1. General: Breaks BM01-01 and 02 have a steam generator A source and a turbine building source. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 29, such that no whipping occurs.
4. Jet impingement: The targets essential to mitigating the consequences of the breaks are seal injection to RCP-A and the CCW return header from RCP-A, B, C, and D thermal barrier cooling coils. Function of these essential systems is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 58) Room No. 2000 Steam Generator B&C Blowdown I. Sheets of Figure 3.6-1 showing high-energy 30 (H-E) piping in this room II. Effects Analysis A. Problem No. 221, Secondary Systems Breaks.

1. General: Break BM02-04 has a steam generator B source and a turbine building source. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Break is restrained per Figure 3.6-1, Sheet 30 such that no whipping occurs.
4. Jet impingement: The only target essential to mitigating the consequences of the breaks is the CCW line from the thermal barrier cooling coil RCP-C.

Function of this essential system is ensured.

5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a B. Problem No. 222, Secondary Systems Breaks.
1. General: Break BM02-01 has a steam generator C source and a turbine building source. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Break is restrained per Figure 3.6-1, Sheet 30 such that no whipping occurs.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 59) Room No. 2000 Steam Generator A, B, C, D Blowdown I. Sheets of Figure 3.6-1 showing high-energy 31 (H-E) piping in this room II. Effects Analysis A. Problem No. 220, Secondary Systems Breaks.

1. General: Breaks BM03-06 and 07 have a H-E source from steam generator A. Downstream source is moderate energy. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 31, such that no whipping occurs.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a B. Problem No. 221, Secondary Systems Breaks.
1. General: Break BM03-01 has a H-E source from steam generator B. Downstream source is moderate energy. No restrictions were considered in calculation of thrust forces.
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Break is restrained per Figure 3.6-1, Sheet 31, such that no whipping occurs.
4. Jet impingement: No essential systems are impacted.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 59 - cont)

5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a C. Problem No. 222, Secondary Systems Breaks.
1. General: Breaks BM03-02 and 03* have a H-E source from steam generator C. Downstream source is moderate energy. No restrictions were considered in calculation of thrust forces.
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 31; such that no whipping occurs.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a D. Problem No. 219, Loop No. 4, Secondary Systems Breaks.
1. General: Breaks BM03-04 and 05* have a H-E source from steam generator D. Downstream source is moderate energy. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 31, such that no whipping occurs.
4. Jet impingement: No essential targets are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a
  • Intermediate Break BM03-03 and BM03-05 are deleted.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 60) Room No. 2000 Steam Generator A Sample and Tube Sheet Drain I. Sheets of Figure 3.6-1 showing high-energy 32 (H-E) piping in this room II. Effects Analysis A. Problem No. 220, Secondary Systems Breaks.

1. General: Breaks BM17-02*, 03, 04, 05*, 06, and 07 (Callaway only) have sources from steam generator A. No restrictions were considered in the calculation of thrust forces:
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 32, such that no whipping occurs.
4. Jet impingement: The targets essential to mitigating the consequences of the breaks are the seal injection to RCP-A and CCW from thermal barrier cooling coil, RCP-A. Function of the essential systems is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a
  • Intermediate Breaks BM17-02 and BM17-05 are deleted.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 61) Room No. 2000 Steam Generator B Sample and Tube Sheet Drain I. Sheets of Figure 3.6-1 showing high-energy 33 (H-E) piping in this room II. Effects Analysis A. Problem No. 221, Secondary Systems Breaks.

1. General: Breaks BM18-01, 04, 05*, 06*, and 07* have sources from steam generator B. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 33. Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: The only target essential to mitigating the consequences of the breaks is the feedwater line to steam generator C. Function of this essential system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c
  • Intermediate Breaks BM18-05, BM18-06 and BM18-07 are deleted.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 62) Room No. 2000 Steam Generator C Sample and Tube Sheet Drain I. Sheets of Figure 3.6-1 showing high-energy 34 (H-E) piping in this room II. Effects Analysis A. Problem No. 222, Secondary Systems Breaks.

1. General: Breaks BM19-01, 02*, 03*, and 04 have sources from steam generator C. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 34. Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a
  • Intermediate Breaks BM19-02, BM19-03 are deleted.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 63) Room No. 2000 Steam Generator D Sample and Tube Sheet Drain I. Sheets of Figure 3.6-1 showing high-energy 35 (H-E) piping in this room II. Effects Analysis A. Problem No. 219, Secondary Systems Breaks.

1. General: Breaks BM20-01,
  • and 04 have sources from steam generator D. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The secondary systems break criteria has been met. (See Note D)
3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 35. Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: The targets essential to mitigating the consequences of the breaks are seal injection to RCP-A, CCW from the thermal barrier cooling coil, RCP-A and D, and the CCW return header, RCP-A, B, C, D.

Function of these essential systems is ensured.

5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a
  • Intermediate Breaks BM20-03, BM20-03 are deleted.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 64) Room No. 2000 Residual Heat Removal, Loops No. 1 and 4 I. Sheets of Figure 3.6-1 showing high-energy 36 (H-E) piping in this room II. Effects Analysis A. Problem No. 255, Loop No. 1, LOCA Breaks.

1. General: Breaks EJ04-06, 07, 08, 09, and 10 have a H-E source from the RCS Hot Leg, Loop No. 1. Sources from the RHR and S.I. pumps are moderate energy. Breaks EJ04-07 and 08 can be deleted per arbitrary break elimination.
2. Criteria: The large LOCA break criteria has been met.

(See Note A)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 36. Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a
7. Flooding: See Section 6.3.2.2 B. Problem No. 256, Loop No. 4, LOCA Breaks.
1. General: Breaks EJ04-01, 02, 03, 04, and 05 have a H-E source from the RCS hot leg, Loop No. 4. Sources from the RHR and S.I. pumps are moderate energy. Breaks EJ04-03 and 04 can be deleted per arbitrary break elimination.
2. Criteria: The large LOCA break criteria has been met.

(See Note A)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 36. Whipping occurs for some breaks. However, no essential systems are impacted.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 64 - cont)

4. Jet impingement: The only target essential to mitigating the consequences of the breaks is seal injection to RCP-B. Function of this essential system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a
7. Flooding: See Section 6.3.2.2 Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 65) Room No. 2000 High Pressure Coolant Injection - Loops 2 and 3 I. Sheets of Figure 3.6-1 showing high-energy 38 (H-E) piping in this room II. Effects Analysis A. Problem No. 248A, Loop No. 2 - LOCA Breaks.

1. General: EM03-08, 28, and 29 are LOCA breaks having a H-E source from the RCS hot leg Loop No. 2. Sources from the hot leg recirculation line and S.I. pumps are moderate energy. No restrictions were considered in the calculation of thrust forces. Breaks EM03-28 can be deleted per arbitrary break elimination.
2. Criteria: The large LOCA break criteria has been met.

(See Note A)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 38, such that there are no whip targets.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a B. Problem No. 248A, Loop No. 3, LOCA Breaks.
1. General: Breaks EM03-05, 26, and 27 are LOCA breaks having a H-E source from the RCS hot leg Loop No. 3.

Sources from the hot leg recirculation line and S.I. pumps are moderate energy. No restrictions were considered in the calculation of thrust forces. Breaks EM03-26 can be deleted per arbitrary break elimination. Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 65 - cont)

2. Criteria: The large LOCA break criteria has been met.

(See Note A)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 38, such that there are no whip targets.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a C. Problem No. 247A, Loop No. 1, LOCA Breaks.
1. General: Breaks EM03-15, 16, 17, and 18 are LOCA breaks having a H-E source from the RCS cold leg Loop No. 1. Source from the boron injection tank is moderate energy. No restrictions were considered in the calculation of thrust forces. Breaks EM03-16 and 17 can be deleted per arbitrary break elimination.
2. Criteria: The small LOCA break criteria has been met.

(See Note A)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 38, such that no whipping occurs.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a D. Problem No. 247A, Loop No. 2, LOCA Breaks.
1. General: Breaks EM03-09, 10, 11, and 12 are LOCA breaks having a H-E source from the RCS cold leg Loop No. 2. Source from the boron injection line is moderate energy. No restrictions were considered in the calculation of thrust forces. Breaks EM03-10 and 11 can be deleted per arbitrary break elimination.
2. Criteria: The small LOCA break criteria has been met.

(See Note A) Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 65 - cont)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 38, such that no whipping occurs.
4. Jet impingement: The only target essential to mitigating the consequences of the breaks is BIT to RCS cold leg Loop No. 3. Function of this essential system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a E. Problem No. 247A, Loop No. 3, LOCA Breaks.
1. General: Breaks EM03-01, 02, 03, and 04 are LOCA breaks having a H-E source from the RCS cold leg Loop No. 3. Source from the boron injection tank is moderate energy. No restrictions were considered in the calculation of thrust forces. Breaks EM03-02 and 03 can be deleted per arbitrary break elimination.
2. Criteria: The small LOCA break criteria has been met.

(See Note A)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 38. Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a F. Problem No. 247A, Loop No. 4, LOCA Breaks.
1. General: Breaks EM03-19, 20, 21, and 22 are LOCA breaks having a H-E source from the RCS cold leg loop No. 4. Source from the boron injection tank is moderate energy. No restrictions were considered in the calculation of thrust forces. Breaks EM03-20 and 21 can be deleted per arbitrary break elimination.
2. Criteria: The small LOCA break criteria has been met.

(See Note A)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 38. Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 66) Room No. 2000 High Pressure Coolant Injection - Loops 1 & 4 I. Sheets of Figure 3.6-1 showing high-energy 39 (H-E) piping in this room II. Effects Analysis A. Problem No. 255, Loop No. l, LOCA Breaks

1. General: Breaks EM05-03 and 04 have a H-E source from the RCS hot leg, Loop No. 1. Source from the S.I. pump is moderate energy. No restrictions were considered in the calculation of thrust forces. Breaks EM05-03 can be deleted per arbitrary break elimination.
2. Criteria: The large-LOCA break criteria has been met.

(See Note A)

3. Pipe whip: Whipping occurs; however, no essential systems are impacted.
4. Jet impingement: No essential targets are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a B. Problem No. 256, Loop No. 4, LOCA Breaks.
1. General: Breaks EM05-01 and 02 have a H-E source from the RCS hot leg, Loop No. 4. Source from the S.I. pump is moderate energy. No restrictions were considered in the calculation of thrust forces. Breaks EM05-02 can be deleted per arbitrary break elimination.
2. Criteria: The large-LOCA break criteria has been met.

(See Note A)

3. Pipe whip: Whipping occurs; however, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 67) Room No. 2000 Accumulator Injection, Loops 1 & 4 I. Sheets of Figure 3.6-1 showing high-energy 40 (H-E) piping in this room II. Effects Analysis A. Problem No. 234, Loop No. 1 - LOCA Breaks.

1. General: Breaks EP01-01 and 04 are LOCA breaks having sources from the RCS cold leg Loop No. 1 and accumulator tank A. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The large LOCA break criteria has been met.

(See Note A)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 40, such that no whipping occurs.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a B. Problem No. 234, Loop No. 1 - Non-LOCA Break.
1. General: Breaks EP01-05, 07, 18, 19, 20, 22, and 27 have a H-E source from accumulator tank A only. No source available from Loop No. 1 due to check valve BB-8948B located between the breaks and Loop No. 1.

Sources from the RHR and SI pumps are moderate energy. No restrictions were considered in the calculation of thrust forces. Break EP01-27 can be deleted per arbitrary break elimination (Ref. Sect. 3.6.2.1)

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 40. Whipping occurs for some breaks. However, no essential systems are impacted.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 67 - cont)

4. Jet impingement: The only target essential to mitigating the consequences of the breaks is the cold leg Loop No. 2 safety-injection line. Function of this essential system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a C. Problem No. 235, Loop 4 - LOCA Breaks.
1. General: Breaks EP01-10 and 13 are LOCA breaks having sources from the RCS cold leg Loop No. 4 and accumulator tank D. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The large LOCA break criteria has been met.

(See Note A)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 40, such that no whipping occurs.
4. Jet impingement: The targets essential to mitigating the consequences of the breaks are seal injection to RCP-C and RCP-B. Function of this essential system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a D. Problem No. 235, Loop 4 - Non-LOCA Breaks.
1. General: Breaks EP01-08, 14, 15, 16, 17, 26, and 28 have a H-E source from accumulator tank D only. No source available from Loop No. 4 due to check valve BB-8948D located between the breaks and Loop No. 4.

Sources from the RHR and SI pumps are moderate energy. No restrictions were considered in the calculation of thrust forces. Break EP01-26 can be deleted per arbitrary break elimination (Ref. Sect. 3.6.2.1)

2. Criteria: The non-LOCA break criteria has been met.

(See Note C) Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 67 - cont)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 40, such that the whip targets are not required for post accident safe shutdown or to mitigate the consequences of the accident.
4. Jet impingement: The targets essential to mitigating the consequences of the breaks are seal injection to RCP-A, B, and C, component cooling water to excess letdown HX, and to the thermal barrier cooling coil, RCP-D, and the excess letdown heat exchanger discharge line. Function of these essential targets is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 68) Room No. 2000 Accumulator Injection - Loops 2 and 3 I. Sheets of Figure 3.6-1 showing high-energy 41 (H-E) piping in this room II. Effects Analysis A. Problem No. 237, Loop No. 2 - LOCA Breaks.

1. General: Breaks EP02-01 and 04 are LOCA breaks having sources from the RCS cold leg, Loop No. 2 and accumulator tank B. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The large LOCA break criteria has been met.

(See Note A)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 41, such that no whipping occurs.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a B. Problem No. 237, Loop No. 2 - Non-LOCA Breaks.
1. General: Breaks EP02-05, 06, 16, 17, 18, 19, and 20 have a H-E source from accumulator tank B only. No source available from Loop No. 2 due to check valve BB-8948B located between the breaks and Loop No. 2.

Sources from the RHR & S.I. pumps are moderate energy. No restrictions were considered in the calculation of thrust forces. Breaks EP02-19 can be deleted per arbitrary break elimination.

2. Criteria: The non-LOCA break criteria has been met.

(See Note C) Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 68 - cont)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 41, such that no whipping occurs.
4. Jet impingement: The only target essential to mitigating the consequences of the breaks is component cooling water from the thermal barrier cooling coil, RCP-B. Function of this essential system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a C. Problem No. 236, Loop No. 3 - LOCA Breaks.
1. General: Breaks EP02-08 and 11 are LOCA breaks having sources from the RCS cold leg Loop No. 3 and accumulator tank C. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The large LOCA break criteria has been met.

(See Note A)

3. Pipe whip: The breaks are restrained per Figure 3.6-1, Sheet 41, such that no whipping occurs.
4. Jet impingement: The only target essential to mitigating the consequences of the breaks is seal injection to RCP-B. Function of this essential system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a D. Problem No. 236, Loop 3 - Non-LOCA Breaks.
1. General: Breaks EP02-07, 12, 13, 14, 15, 22, and 23 have a H-E source from accumulator tank C only. No source available from Loop No. 3 due to check valve BB-8948C located between the breaks and Loop No. 3.

Sources from the RHR and S.I. pumps are moderate energy. No restrictions were considered in the calculation of thrust forces. Break EP02-23 can be deleted per arbitrary break elimination. (Ref. Sect. 3.6.2.1)

2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Breaks are restrained per Figure 3.6-1, Sheet 41. Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 69) Room No. 2000 Loop Drains I. Sheets of Figure 3.6-1 showing high-energy 50 (H-E) piping in this room II. Effects Analysis A. Problem No. 245, Loop No. 2 - LOCA Breaks.

1. General: Breaks HB24-03, 04, and 07 have a H-E source from the crossover leg, Loop No. 2. The downstream source is moderate energy. No restrictions were considered in the calculation of thrust forces. Breaks HB24-07 can be deleted per arbitrary break elimination.
2. Criteria: The small-LOCA break criteria has been met.

(See Note B)

3. Pipe whip: Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a B. Problem No. 245, Loop No. 3 - LOCA Breaks.
1. General: Break HB24-05 has a H-E source from the crossover leg, Loop No. 3. The downstream source is moderate energy. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The small-LOCA break criteria has been met.

(See Note B)

3. Pipe whip: Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 69 - cont)

5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a C. Problem No. 244, Loop No. 1 - LOCA Breaks.
1. General: Breaks HB24-01, 02, and 06 have a H-E source from the crossover leg, Loop No. 1. The downstream source is moderate energy. No restrictions were considered in the calculation of thrust forces. Breaks HB24-06 can be deleted per arbitrary break elimination.
2. Criteria: The small-LOCA break criteria has been met.

(See Note B)

3. Pipe whip: Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: The only target essential to mitigating the consequences of the breaks is the safety injection to RCS hot leg Loop No. 1. Function of this essential system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 70) Room No. 2000 Liquid Radwaste I. Sheets of Figure 3.6-1 showing high-energy 51 (H-E) piping in this room II. Effects Analysis A. Problem No. 234, Non-LOCA Breaks.

1. General: Breaks HB27-01, 02, and 09 have a H-E source from the accumulator tank A. The reactor coolant drain tank pump source is moderate energy. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c B. Problem No. 235, Non-LOCA Breaks.
1. General: Breaks HB27-07, 08, 10, and 11 have a H-E source from the accumulator tank D. The reactor coolant drain tank pump source is moderate energy. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.

Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 70 - cont)

5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a C. Problem No. 236, Non-LOCA Breaks.
1. General: Breaks HB27-05, 06, 12, and 13 have a H-E source from the accumulator tank C. The reactor coolant drain tank pump source is moderate energy. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c D. Problem No. 237, Non-LOCA Breaks.
1. General: Breaks HB27-03, 04, 14, and 15 have a H-E source from the accumulator tank B. The reactor coolant drain tank pump source is moderate energy. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The non-LOCA break criteria has been met.

(See Note C)

3. Pipe whip: Whipping occurs for some breaks. However, no essential systems are impacted.
4. Jet impingement: No essential systems are impacted.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 71) Room 2000 Pressurizer Surge Line I. Pressurizer surge line pipe breaks are shown on Figure 3.6-3 II. Effects Analysis A. Pressurizer Surge Line, Loop 4 - LOCA Breaks. (P-257)

1. General: Breaks 12, and 15 are LOCA breaks having sources from the RCS hot leg Loop 4 and the pressurizer. No restrictions were considered in the calculation of thrust forces.
2. Criteria: The large-LOCA break criteria has been met.

(See Note A)

3. Pipe whip: Breaks are restrained per Figure 3.6-3. The whip targets are not required for post-accident safe shutdown or to mitigate the consequences of the accident.
4. Jet impingement: The only target essential to mitigating the consequences of the breaks is accumulator safety injection, Loop No. 1. Function of this essential system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 72) Room No. 2000 Reactor Coolant Loop 1 I. Reactor coolant loop pipe breaks are shown on Figure 3.6-3 II. Effects Analysis A. Reactor Coolant Loop 1 - LOCA Breaks.

1. General: Breaks BB01-01, 03, 04, 05, 06, 07, and 08 are limited-area circumferential LOCA breaks. Sources are from both ends of each leg which produce radial disc jets. Break BB01-02 is a longitudinal slot break with source from RCS hot leg which produces a simple conical jet. No restrictions were considered in the calculation of thrust forces. Breaks BB01-02 and 05 can be deleted per arbitrary break elimination.
2. Criteria: The large LOCA break criteria has been met.

(See Note A)

3. Pipe whip: Whip restraints designed by Westinghouse limit displacement, such that no whipping occurs.
4. Jet impingement: Function of jet targets is not required for post-accident safe shutdown or to mitigate the consequences of the accident.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c B. Reactor Coolant Loop 2 - LOCA Breaks.
1. General: Breaks BB01-01, 03, 04, 05, 06, 07, and 08 are limited-area circumferential LOCA breaks. Sources are from both ends of each leg which produce radial disc jets. Break BB01-02 is a longitudinal slot break Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 72 - cont) with source from RCS hot leg which produces a simple conical jet. No restrictions were considered in the calculation of thrust forces. Breaks BB01-02 and 05 can be deleted per application of LLB technology.

2. Criteria: The large LOCA break criteria has been met.

(See Note A)

3. Pipe whip: Whip restraints designed by Westinghouse limit displacement, such that no whipping occurs.
4. Jet impingement: The targets essential to mitigating the consequences of the breaks are CCW from RCP-C thermal barrier and RHR injection to intact coolant Loop No. 3. Function of these systems is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c C. Reactor Coolant Loop 3 - LOCA Breaks.
1. General: Breaks BB01-01, 03, 04, 05, 06, 07, and 08 are limited-area circumferential LOCA breaks. Sources are from both ends of each leg which produce radial disc jets. Break BB01-02 is a longitudinal slot break with source from RCS hot leg which produces a simple conical jet. No restrictions were considered in the calculation of thrust forces. Breaks BB01-02 and 05 can be deleted per application of LLB technology.
2. Criteria: The large LOCA break criteria has been met.

(See Note A)

3. Pipe whip: Whip restraints designed by Westinghouse limit displacement, such that no whipping occurs.
4. Jet impingement: The only target essential to mitigating the consequences of the breaks is seal injection to RCP-B. Function of this system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 72 - cont) D. Reactor Coolant Loop 4 - LOCA Breaks.

1. General: Breaks BB01-01, 03, 04, 05, 06, 07, and 08 are limited-area circumferential LOCA breaks. Sources are from both ends of each leg which produce radial disc jets. Break BB0l-02 is a longitudinal slot break with source from RCS hot leg which produces a simple conical jet. No restrictions were considered in the calculation of thrust forces. Breaks BB01-02 and 05 can be deleted per application of LLB technology.
2. Criteria: The large LOCA break criteria has been met.

(See Note A)

3. Pipe whip: Whip restraints designed by Westinghouse limit displacement, such that no whipping occurs.
4. Jet impingement: The targets essential to mitigating the consequences of the breaks are seal injection to RCP-B and RCP-C. Function of this system is ensured.
5. Room pressurization: See Section 6.2.1.1.3a
6. Temperature and humidity: See Section 6.2.1.1.3a,c Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 73) NOTES: A. LARGE LOCA BREAK CRITERIA

1. The effects of large LOCA breaks must be limited to the following:
a. Containment integrity must be maintained.
b. Propagation to the secondary system is not allowed.
c. No break propagation to the three remaining intact LOOPS is allowed.
d. For branch line breaks, break propagation in the affected LOOP must be limited to an increase of 20 percent of the initial break area.
e. For main coolant loop pipe breaks, break propogation limits are stated in PIP Vol. 1-3, Tab 10.
2. The following "ESSENTIAL" functions are required for mitigation of the pipe break via the ECCS systems.
a. Accumulator safety injection to the three intact loops.
b. Low head (RHR) safety injection to the three intact loops.
c. Reactor coolant system equipment supports must maintain their functions.
3. The following other systems located inside the containment must maintain their design redundancy:
a. Containment Spray (EN)
b. Containment Cooling (GN)
c. Containment Hydrogen Control (GS)
d. Containment Isolation
4. The following safety actuation signals must be capable of being generated from instrumentation within the containment.
a. Reactor Trip
b. Safety Injection Signal Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 73 - cont) NOTES:

c. Containment Isolation Phase A and Phase B
d. Containment Spray Actuation
5. All safety-related equipment located outside of the containment is operable and subject to single failure criteria.
6. No non-safety-related equipment either inside or outside the containment is required for mitigation of the effects of this LOCA.

B. SMALL LOCA BREAK CRITERIA

1. The effects of small LOCA breaks must be limited to the following:
a. Containment integrity must be maintained.
b. Rupture of steam-feedwater lines must be prevented.
c. Break propagation must be limited to the affected leg.
d. Break propagation in the affected leg must be limited to 12.5 square inches (4 inches ID).
e. Damage to the high head safety injection lines connected to the other leg of the affected loop or to the other loops must be prevented.
f. Propagation of the break to the high head safety injection line connected to the affected leg must be prevented if the line break results in a loss of core cooling capability due to a spilling injection line.
2. The following ESSENTIAL functions are required for mitigation of the pipe break:
a. High head safety injection via the ECCS systems.
b. Boration via one of the following paths:
1. Boration via the BIT path to the four loops.
2. Boration via the RCP seals for all four loops
3. Boration via normal charging Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 73 - cont) NOTES:

c. Reactor coolant system equipment supports and restraints must maintain their functions.
3. The following other systems located inside the containment must maintain their design redundancy:
a. Containment Spray (EN)
b. Containment Cooling (GN)
c. Containment Hydrogen Control (GS)
d. Containment Isolation
4. The following safety actuation signals must be capable of being generated from instrumentation within the containment:
a. Reactor Trip
b. Safety Injection Signal
c. Containment Isolation Phase A and Phase B
d. Containment Spray Actuation
5. All safety-related equipment located outside the containment is operable and subject to single failure criteria.
6. No non-safety related equipment either inside or outside containment is required for mitigation of the effects of this LOCA.

C. NON-LOCA BREAK CRITERIA

1. The effects of non-LOCA breaks must be limited to the following:
a. Containment integrity must be maintained.
b. A non-LOCA break must not cause a loss of coolant or secondary systems line break.
c. The essential functions required for post-accident safe shutdown due to a non-LOCA break must be maintained.

(See Section 7.4) Rev. 30

WOLF CREEK TABLE 3.6-4 (Sheet 73 - cont) NOTES:

2. The following other systems located inside containment must maintain their design redundancy:
a. Containment Cooling (GN)
b. Containment Isolation
3. The following safety actuation signals must be capable of being generated from instrumentation within the containment:
a. Reactor Trip
b. Containment Isolation
c. Safety Injection Signal
4. No non-safety-related equipment either inside or outside containment is required for post-accident safe shutdown due to a non-LOCA pipe break.

D. SECONDARY SYSTEMS BREAK CRITERIA

1. The effects of secondary systems breaks must be limited to the following:
a. Containment integrity must be maintained.
b. Propagation to the primary system is not allowed.
c. The essential functions required for post-accident safe shutdown due to a secondary systems break must be maintained.

(see Sections 15.1.5 and 15.2.8 and Section 7.4)

2. The following other systems located inside the containment must maintain their design redundancy:
a. Containment Spray (EN)
b. Containment Cooling (GN)
c. Containment Isolation
3. The following safety actuation signals must be capable of being generated from instrumentation within containment:
a. Reactor Trip NOTES:
b. Safety Injection Signal
c. Containment Isolation
d. Containment Spray
4. No non-safety-related equipment either inside or outside containment is required for post-accident safe shutdown due to a secondary systems pipe break.

Rev. 30

WOLF CREEK TABLE 3.6-5 STRESS INTENSITY RANGES AND CUMULATIVE USAGE FACTORS AT DESIGN BREAK LOCATIONS IN THE REACTOR COOLANT LOOP 4 Pipe Break Isometric No.: Figure 3.9(N)-1a (a) Node Equation 12 Equation 13 Cum. Usage Allowable No. Stress (ksi) Stress (ksi) Factor Stress (ksi) 404 40.4 54.8 0.98 56.7 415 40.4 54.8 0.98 56.7 438 20.0 50.6 0.70 56.7 459 20.0 50.6 0.70 56.7 468 31.3 48.1 0.70 56.7 484 31.3 48.1 0.70 56.7 Notes: (a) Node numbers for loop 1 are defined in FIG. 3.9(N)-1a Rev. 13

WOLF CREEK TABLE 3.6-6

SUMMARY

OF FLOOD LEVELS IN ALL SAFETY-RELATED ROOMS AUXILIARY BUILDING AUXILIARY BUILDING (Cont.) Flood Level Flood Level Room Above Floor Room Above Floor No. Elevation No. Elevation 1101 3' 6" 1310 3'1" 1102 3' 6" 1311 10" 1103 3' 6" 1312 10" 1104 3' 6" 1313 0" 1105 3' 6" 1314 2'10" 1106 3' 6" 1315 2'10" 1107 0" 1316 0' 0" 1108 0" 1317 0' 0" 1109 62" 1318 0" 1110 62" 1320 2'10" 1111 62"" 1321 0' 0" 1112 62" 1322 0" 1113 0" 1323 0" 1114 0" 1324 0" 1115 3' 6" 1325 0" 1116 3' 6" 1326 0" 1117 3' 6" 1327 0" 1119 0' 0" 1328 0" 1120 3' 6" 1329 0' 0" 1121 10' 6" 1330 0" 1122 3' 6" 1331 1' 11" 1123 3' 6" 1401 0' 7" 1124 3' 6" 1402 0' 7" 1125 3' 6" 1403 0' 8" 1126 6'9" 1405 0' 0" 1127 15'7" 1406 0' 7" 1128 3' 6" 1407 0' 0" 1129 3' 6" 1408 0' 7" 1130 3' 6" 1409 0' 0" 1201 0' 0" 1410 0' 2" 1202 0' 0" 1411 1' 4" 1203 4' 1" 1412 l' 4" 1204 0' 0" 1413 0' 0" 1205 0' 0" 1501 0" 1206 0' 0" 1502 0' 2" 1207 0' 0" 1503 0' 2" 1301 2'10" 1504 0' 2" 1302 0' 0" 1505 0' 2" 1304 0' 0" 1506 0' 2" 1305 0' 0" 1507 0' 2" 1306 0' 0" 1508 0' 0" 1307 0' 0" 1509 0' 0" 1308 0' 0" 1512 0' 0" 1309 2' 4" 1513 0' 2" Rev. 32

WOLF CREEK TABLE 3.6-6 (Sheet 2) REACTOR BUILDING FUEL BUILDING Flood Level Room Flood Level Room Above Floor No. Elevation No. Elevation 2000 6102 1' 6" LOCA <2004'-8" 6104 1' 6" MSLB <2004'-6" 6105 1' 6" 6203 1' 6" 6303 2' 5" 6304 0' 0" CONTROL BUILDING DIESEL BUILDING Flood Level Flood Level Room Above Floor Room Above Floor No. Elevation No. Elevation 3101 210' 5201 0' 4" 3301 0' 0" 5203 0' 4" 3302 0' 2" 3403 2' 5" 3404 0' 0" 3405 0' 0" 3407 0' 0" 3408 0' 0" 3409 2' 9" 3410 0' 0" 3411 0' 0" 3413 0' 0" 3414 0' 0" 3415 0' 0" 3416 0' 0" 3501 0' 1" 3605 0' 0" 3801 0' 7" Rev. 32

8 7 8 5 3 2 REAc:TOR BUILDING REACTOR AUXILIARY BUILDING BUILDING J+1' *Y H l(' *Z DETAIL (PLAN) E lfi£A CF 

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                                        @*~~~                                                   FIGURE 3.6-1
                                       ~    H      GUIDE RESTRAINt HIGH ENERGY PIPE BREAK ISOMETRIC
                                        ., BUMPER RESTRAINT WITH                         M~N STEAM SYSTEM                                                  A 1
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8 7 6 5 4 3 2 

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                                                                                    '    *     - ...:ailS **u . .uun **u               ,..... ( NQ       TE J    0      -IIIOC.P1111*f*IIDOI AUXILIARY BUILDING                                                                   I   ~      -  :::==~IIC~II'I.III .. TtR.. *111,.111.1111 IJIII.... ..._11 (1111 AOCJM 1322                                                                      !. t;i;) * :::~~YUIIIIIIM.                ..U: 111 IIIII'I.IIIDIII1(

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B REV. 19 WOLF CRIEEK UIPDATED SAFETY ANALYSIS REPORT FIGURE 3.6-1 HIGH ENERGY PIPE BREAK ISOMETRIC REACTOR COOLANT SYSTEM SEAL WATER INJECTION INSIDE CONTAINMENT (8807) A CSHEET 12)

8 7 6 5 4 3 2 H G AUXILIARY BUILDING AREA OF REACT&f!,) BUILDING 

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                                                                                      <NOTE                11)

II BUMPER RESTRAINT WITH ATTACMMENT TO PIP!i: c 1 {FOR FURTHER DETAILS 01' AE5TAAINT TYPES SEE SECTION 3.8.2.3.3) A ANC:IIOiit 

---li*   I!INUBMR
;;    RIC,ID HANeEJI B

REV. 19 WOLIF CREEK UPDATIED SAFETY ANAILYSIS RIEPORT FIGURE 3.6-1 HIGH ENERGY PIPE BREAK ISOMETRIC REACTOR COOLANT PUMP A SEAL WATER INJECTION INSIDE CONTAINMENT A CBB08) 

                                                                                                                                      <SHEET 13)

8 7 6 5 4 3 2 t?l!f H l("* +Z G c1 RESTRAINT LEGEND DIRECTIONAL RESTRAINT 

                                                                                  ~            BUMPER: RESTRAINT
                                                                              ~*~*~*-m H H                          G~IDE RE9TRAIMT F
                                                                               .1 BUMPER AESTR ..II\IT WIT~

ATTACi'tMENT TD PIPE 1 I FOR FURTHER DETAilS OF RESTRAINT TYPES SE.E SECTION u;,a.3.3) l:::,. ANCHOR E NOTES: L 3. 4 i 

                                                                     ,. 0 -llmiCAT&TI!IIIIINA.LENDI"EA.. fOINT ~---~

0 - ...........,.**,,........... POINl ( N0 T E 1111 

                                                                                      -INDICATEIITIIIIINCDi

_ ,IIDICATEICIIICLIIIIFiiiiiE.NTIAL. IIIIAII: oiiTIMTIPIIIIIIIliiATI!' *lli'AI&fDINT, 

                                                                                       -  IIIDICATII LOIIiiTUCIIIIAL. DREAM I<< 1NTPtln.-rJE Dflt'IC,IIIGilif.
                                                                      *               -INDICATYI:IRCIJMFEJUNTIA.L.IIIIAKATTEIIIMINALIND
                                                                      'P. 0          -INDICATIIIIIIIEAI,.OINTIIIJ-ER I@) -INDII:AUirlrt!IIIEAI' PIEITRAINT
                                                                     ! 5TRESS I.E.SULTS WHICI! ARE. GillEN IH "DoBLE.._ **~                                 SHE.E~41*60 C.OIRE!!IPOND TO THE. NU~£R.ICAL- IJCOwU,. PCIAI.TS SHCW/N HERE.
10. ST£Al>Y*STATE TliRUST FOIU:Eli FOR EACH !!IlEAl< ,...._

PROVIDED BELOW. THP.UitFORc;E,I.AI 1109*0.IJ lC.VC*) I!I~CJiHl41CVC.5) 1!1809-GflLOOP) 

                                                                                  &S09 *01 fC.VC5~

111109*10 ILOC>I'l Elliii>'*IO(C'ICS) CNOTE 12) { B&O 9*11 CLOOI'l aa.o'il'-11 tc.vcs.> BBIOSI*IZ CC.IICS} 8&09-oa!eVoOs\.

11. P'!'OC:"r.i AIIILftl*"'--'ftFIIRWMIIIICl!UIEI'mllllBPON1'10t.

IHCIC,.-T*D '&MI!I!'fti CF 'TW!II.R ~6!*iilo 11..0CM.... 

                                                                                                  *~tt I  Tl'aU. !-.**4 SloiE[T 1.000                        !10
12. THESE INTERMEDIATE CAN BE DELETED.

AUXILIARY BUILDING (Jill) B ROOM 1:322 REV. 19 WOLF CRIEEK ( CNOTE 12) CNOTE 12) UIPDATIED SAFETY ANAILYSIS REPORT FIGURE 3.6-1 HIGH ENERGY PIPE BREAK ISOMETRIC REACTOR COOLANT PUMP C SEAL WATER INJECTION DETAIL @ (c.-'I) INSIDE CONTAINMENT CBB09) A CSHEET 14)

AREA OF DllliCTIONAL IIUTAAINT CHANGE IUMP~I IIUTIIAINT 

~*--*~

H H GUIDI! JEST.IIAINT

8 7 6 5 4 3 2 H NOTES : 

                ~DI CATES TERMINALENOBREAK POI N:OINT 2    *       - INDICATES INTER MEDIAT E END BREAK J    0 -       INDICATES STRESS NODE                        AT IN TERMEDIATE 4    [I] _~~~~~TE~OI~~C UMFERENTIAL               BREAK 5    ~       - 1 ~2~C~T E~O I~~NGI TUD I N AL BREAK AT IN TERMED IATE
6. GlJ INDICATES CIR CUMFERENTIAL BREAK AT TERM INAL EN D
7. Q INOICATESBREAKPO:N T NUMBER P:;RE~S TRSRE~~~2~o ~6 ~~~fNN~t.tE1R~~ aot~.:

IN DICATES PIPE BREAK RES TR/IJ NT G

RESUL SHEE TS
                        ~~~~h"' s~o~S       HERE STEAD Y STATE THRUS T FORCES FOR EACH BREAK ARE PROV IDED BELOW                         THRUST FORCE, LBS BREAK POI NT                                   ,000 37 BGOl-01                                    DELE TED BGOI - 02                                  DELE TED BGOI - 03                                  37 ,000 8G01 - 04C3"J                              15,597 BGOl - 0 4<2 "1                            37 ,000 BGOl- 06                                   33,000 BGOl- 07                                   33,000 BGOI - 08                                  33,000 BGO I- 09                                  33,00 0 BG OI-11! 3")                              14 ,000 BG OI-11(2" )                              DEL ET ED BGOI -13                                   33,000 BG01 - 14(J "l 8G01 - 14 12"l 14, 000                 F c:

RESTRAIN T LEGE ND DIREC TI ONAL RES TRAINT 

                                        ,.,        BUMPER REST RAIN T
                                       $ "OcAna< "'"""'                                        E HH                   GUIDERESTR AIN T I*~

BU MPER RE ST RAI NT WITH ATTACHMENT TO PIP E 

                          ~E~RS~~~T~~RJ_~~I.~ ~~ OF RE STAIN T TYPE S
                          .&. ANCHOR l1. - EFFECTS ANAL~~           1 ~6~~~t1 FS~~E~~~ ~~~E Af_~ - 4 PROV IDED ON                      TABL E 3 .6 - 4 SHEE T               D 20
                                              *r      /      *X
                                                ~.z                                            c AREA OF CHANGE B

REV . 32 WOLF CREEK U PDATED SAFETY ANA LYS IS REPORT FIGU RE 3 .6 - 1 HIGH ENERGY PIPE BRE AKH~SOME T R i e NeP TO REGEN eves - OU TSIDE eO NT AINM EN T A CBG0 1) CS HEET 18)

8 7 6 5 4 3 2 H NOTES: G 1 0 INDICATES TERM IN AL END BREAK POIN T D INDICATES INTERMEDIATE END BREAK PO IN T 3 0 INDICATES STR ES SNOD E 4 ~ 01 l~giEC~TEJ ~1fCUMF ERENT I AL BREAK AT INTERMEDI ATE 

                                                                              ;[kj          INDICATES LONGITUDINAL BREAK AT IN TERMED IATE BREAK POINT 6[]'J         INDICATES CIRCUMFERENTIAL BREAK AT TERM INAL END oQ            INDICATES BREAKPOINT NUMBER 6.~           INDICATESP IPEBREAK RESTRAINT 1
                                                                                   *,r."£Y'5:1?'Ro~'1cc)ND~;Q ~~~fNN0~E R~k~ ~0~-~ SHEETS F

10 -E FFECT S ANAL Y SIS RES ULTS FOR EACH ROOM ME PROVIDED ON THE INDIC ATED SHEET OF TABLE 3.6-4 ROOM* TABLE 3_6- 4 SHEET 1107 5 1114 6 1115 7 

                                                                                       ~~g~                                19o 11   STEMJY STATE THRUST FORC ES FOREACHBREA'< ARE PROVIDED BE LO W o9
                                                                ~i~,".

8~ t; ~ ~ 

                                                                '(~ ~ ~

E 

                                                            ~',~~'-~~~~
                                                        ~    .      l -~                                     AREA OF
                                                    -~    ~~         I      ~~~,.                              CHANGE
                                                          . " -~~

0 12 INTERMEDIATE BREAKS BG 02-16!. BG02-17 ARE 

                                                        ~                           DELETEDPERMEB3-1,REV.2 13    IN TE RMED IATE BREAKS BG02-12 tOP 885) BG02-13 WP 95FI
                                                                                    & BG 02-IB <DP 75l ARE DELETED PER ME'B 3-1, REV. 2 D

c a RESTRAINT LEGEND DIRECTIONALRESTRAINT 

              ~BUMPERRESTRAINT B

REV. 32 WOLF CREEK U PD A T ED SA FET Y ANALYSIS REPORT FIGURE 3.6-1 HIGH ENERGY PIPE BREAK ISOMETRIC illi BUMPER RESTRAINT WIT H CCP A & B DISCHARGE A 

           ~    ATTACHMENT TO PIPE I                                                                 eves             OUTSIDE eO NT AINMENT (FOR FURTHER DETAILS OF RESTAINT T YP ES SEESECTION3.6.Z.:'U)                                                                      (8G02l D      ANCH OR                                                                                                   (SHEET 19l

8 7 6 5 4 3 2 H l<' *z 

                                ~
                               '0 G

F 4,026 4 , :l26 4 ,02G(2'"EI<EIIK}

~~: {3"8REA~~x~~~!i~:

8,lH2 (:l"BHEAKl.ETDOW*' 8,972 1,0 88 2,705 J,:)a8 

                                                         ~:~%i t~ss~~\~o~:ro~~'fE 5
                                                         ~:~i~ ~~~~~oH6w t?~cE a  RESTRAINT LEGEND DIRECTIOHAL RESTRAI NT BI.JMPER HESlHAIN )

E AREA OF CHANGE D 

       ~

E 5 INTERMEDIATE BR 390l, BG03-03 (OEAKS BG03-02 (OP 507, 5151, & BG03p1:?95l, BG03-12 (Op DELETED PER MEB 3-i.D~E00~) ARE 13 INTERMEDIATE 1801, BG03-13 BREAKS BG03-09 (Op 2251 ARE DELEI~:o 1631, BG03-16 lOP REV. 2 PER MEB 3-1, B REV. 32 WOLF CREEK U PD A T ED SAFET Y ANALYSIS REPORT FIGURE 3.6-1 HIGH ENERGY PIPE BREAK CVCS LETDOWN O ISOMETRIC CONTAINMENT UTSIDE IBG03l A 

                                                                    !SHEET 201

8 7 5 4 H 

                                                                                        *V l(**     .z G
                                                                     ., BUMPER RES1RAINT WITH                   F
                                                                   ,:m 1 AT1ACHMENT 'fO ,.IPE FURTHER DETAILS OF RESTRAIN1 IYP£5
                                                                    -SE£ s,ECfiON 3.&.2.3.31
                                                                   .... ANCHOR E

D II OJ l,fOJ litO .. 

                                      *.,2. ".,.. t,O~~~,            :r~
                                       ~*~",

rut' !tf.lloo 

                                                 "~-" f II-COIUINLIUU~**I
                                            ~" l~-&1                                            BG09-~~B &      c
                                       -                  12
                                                             *  ~66~-~~        DIATE BREAKS ARE DELETED PER

~ 3-1 REV. 2. 1 (DP 

  =~~a:*:a::'-=:: PQII UCMS                                     INTERMEDDIEALTEET~~E:~R 63~t 3-1,
13. A92) IS 3

REV. 2. AREA OF 

      *oe         .. ,,.

CHANGE B

ii REV. 19
      ~~**

WOILF CREEK

=.~
                   .,                    UP[)ATED SAIFETY ANALYSIS REPOR_T
l--~~-~ ...... FIGURE 3.6 1 HIGH ENER~~A('0~ TER INJECTION BREAK ISOMETRIC cvcgUTSIDE CONTAINMENT CBG09) A CSHEET 21l
                                       ~----

8 7 6 5 4 3 2 

                                                              ~                                      H l                    +X
                                                                                            +Z G
                             ~
1. 0 -INDICAT!ITI!RMINALiNDBREAJCPQINT
2. D -INDICATESINTI!RMIDIATIIREAIICPOINT
                             ~- 0 -INDICATEIITREI&NODE
4. fil - INDI~ATEI CIRCUMFEI'IE'NTIAL BREAK AT INTERM. BREAK. POINT
5. ~ -INDIC~TES LONGITOOINAL BFIEAKATINTERM.BAEAIF. POINT
6. ~ -IND&CATESCIRCUNFERENTIALBREAKATTERMINALEND 1'. 0 -INDH:ATIIBR!AKIIOIN'fNUMBER F
8. CJ_ -INDICATE! PIP~ BREAK RI!STRAINT 9.'-ITREIIA'YULTIWHICHARI!OIVININTABi.EJ.H, SHEI!!'qi J.I.HA CORRE~ND '!d !HI NUM~RICAL NODAL. POif4TI atCIWN'I11!~E.

10.- EFFICTIANALYIIB AEIULTSFOR EACH AOOII ARE PRO-VIDED ON THE INDI.CATEDIHEET OF TABLE 1M. ROOM TABLEI .... iHEET 1107 . a. 1114

  • 1201 **

1204 10 CNOTE 12) DGIG-01 IGID-01 5,2?2 5,1:7'2 E 

                    ,..------..DAID-04                                    5,272 13      IGIG-01                         a.z.n
12) 1010*06 ll71.

AREA OF 12. INTERMEDIATE BREAKS BG10-01 & CHANGE BG10-06 ARE DELETED PER MEB 3-1, REV. 2.

13. INTERMEDIATE BREAK BG10-05 <DP 625) IS DELETED PER MEB 3-1, REV. 2.

D c B REV. 19 WOILF CREIEK UP[)ATE[) SAFETY ANALYSIS REIPORT FIGURE 3.6-1 HIGH ENERGY PIPE BREAK ISOMETRIC CCP A - B MINIFLOW CVCS - OUTSIDE CONTAINMENT CBG10) A CSHEET 22l

8 7 6 5 4 2 

                                                                                   '!+!   *Y H

l:" *Z NOTES: G

1. 0 INDICATES TERMINAL END BREAK PO INT 2 D INDICATES INTERr.!EDIATE END BREAK POIN T 3 0 INDICATESS TRESSNODE 4 [I) ~~~~E~O I~rCUt.I FERE N TI AL BR EAK AT INTER MEDIATE
5. CkJ - ~~~~~IT~O I~~NGIT U DI NAL BREAK AT INTERMEDIATE 6 GlJ INDICATES CIRCUMFERENT IAL BREAK AT TER MINAL END 7 0 INDICATESBREN<POINTNUMBER
8. <::::::> - INDICATES PIPE BREAK RESTR AI NT 9 ~lQ~r~:& :~~~~TR~~~~~2~o ~6 ~~~fNN~~ETR~~ ~Q~-~ SHEETS F 10 - EFFECTS ANALYS IS RESUL TS FOR EAC H ROOM ME PROVIDE D ON THE INDICATED SHEE T OF TAB LE 3.6 --'1 D

c RES T R AIN T LEG END ~ DI RECTI ONALRESTRAJNT ~ BUt.IPERRESTRAINT B REV. 32 WOLF CREEK U PDATED SAFETY ANALYSIS REPORT F IGUR E 3.6 - 1 HIGH ENERGY PIPE BREAK ISOME T RIC Iili Ej BUMPER RES TRAINT WITH ATTACH MEN T TO PIPE L ETDOWN T O REHEAT HX CVCS - OUTS IDE CO NTAINMEN T A (FOR FU RTHER DETAIL S OF RES TRAI NT TYPES SEESECTION3.6.2.3.31 CBG 11 l 6ANCHOR (SHEE T 23l

I a u .. I

  • .~
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          ~
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          ~

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           ..... IMl'M ICIWIICISI
           =
                          }<NOTE 13)
           ~*
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          =

Ul4 l.l& z l'l,li:lo IIIIXIUNn' IUILDING CIN'I nu REV. 19 OLIF CREEK W ANALYSIS REPORT FIGURE 3.6-j HIGH ENERGY PIPELETDOWN BREAK ISOMETRIC IN~YoCES C-ONTAINMENT CBG22) CSHEET 25)

8 7 6 5 4 3 2 RESTRAINT LEGENQ c:: H r DIIIECTIONM. MSTIIAINT 

                                                               -PI:IIMS1UIN1 Hr-1 tir*----        GUIDEII'ItiTIU.INT G
                                                     *~-NIIIIUTIIAINT WITH 1
  • ATTM::HMENT 'Ia PIPI:

I'OIII"UIITHEII DI:TAIU r:W' IIESTIWNT TYPO Ill IECI'ION 3 1.1 , 31 A MCHOII 

                                                        ~
                                                         ;...~* :::=::-..:

I . . . . .....,IITI-.a*IIIA&~ F 

                                                                    --'It*---*----
                                                                                                          .n... - -

t.-----* 

                                                                    -~                           .
                                                          * *1'1'-......T I -. . IIIWIIIMI'hiUU&           - I T DID AREA OF                   .... :n*a:w**n. .._,.._.,.. ..............

CHANGE 

                                                                   =r~;:--**
                                                                          -01
                                                                          *G4 E

lillti*IO 

                                                                   -.clll*ll a,i,Fil;'ft -._'OSIS III.SU*."N IICIII&III~'Titlt ..1 . . .11110

_!~ 11W ZOCO (MU) I IINTWI- SM*T(III'IIel& 810*4 0 

                                                                                         'IJo.M.I. ...........,
                                                                                                   !fo
12. THESE INTERMEDIATE BREAKS CAN BE DELETED.

D AUXILIARY BUILDING REACTOR BUILDING 1100111 I:UI c B REV. 19 WOLF CREIEK UPDATE[) SAFETY ANALYSIS REPORT FIGURE 3.6-1 CHARGING & EXCESS LETDOWN CVCS - INSIDE CONTAINMENT CBG23) A CSHEET 26) 

              *
  • R£STR/&INT L£GE:M?
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c Rev.6 WOLP CREEK B UPDATED SAPETY ARALYSIS REPOR'l' FIGURE 3.6-1 HIGH ENERGY PIPE BREAK iSOMETRIC CVCS - AUXILIARY SPRAY* INSIDE CONTAINMENT (BG24) A 8 SHEET 27 L-.-.-.-.~.~::::::::::::::7 :::::::::::::::*:::::.:.~-~--~--.--- .......... __-ss-**au-~~~-cwwe~~;:~~J~CAU~~~-~~41*_*_*_*-~~-.T.-.-.-.-.-.-.-~3~---.~-.-.-.~.-.-.-.-.-.-.-.zz*_*_*~---.~-~~---.~-.-.-.~.-.-.-.--~ . ..... ..J

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II ATTACHMENT TO PIPE I (~~~ ~~~wg~ ~~:~~~~ 3 ?F RESTAINT TYPES s: RI::V. 113 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.6-1 HIGH ENERGY PIPE BREAK ISOMETRIC STM GEN A,B,C,D SLOWDOWN INSIDE CONTAINMENT A,' 

                                                                                   <BM03l

_-r___________---zr-______________r_I__________ ____L=-=-=-=-=-=-=-=-~;=-=-=-=-=-=-~=C~S=H~=EE~T_231l 1 

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BM17-05 ARE DELETED. ~ RESTRAINT lEGEND 

                                                                          ~        Dlflt.C'TfOHAl *£STJtNNT
                                                                          ,        ttw.Pf",   lti:,Tif&L'IT
                                                                         ~               IIOl.nOHOESFOAINT
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  • ATfACH'~!NT' TON'£ ttoll f'Ulllr)f£'1 DETAilS t1F IIHTWNT TYPES SEE: lrCTION 3 6 r ~ 31 REV. 1a WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.6-1 HIGH ENERGY PIPE BREAK ISOMETRIC STM GEN A SAMPLE & TUBE SHT DRI1IN INSIDE fsO~j~INMENT A:

L __ _ _ _ _ _ _ _ _ _ <-=-cSHEET 3:Zl L--------------,,----------------,----------------~~~~~-

8 --~------~?~________L______ 16 _l_ - 5 ___jz__ 3 2 DIRECTIONAL RESTRAINT BUMPER RESTRAINT 

                                                                           ~~
                                                                            ~

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                                                                         / FURTI-!

WlfH C~OR TO PIPE SECTro~R A EE DETAILS O 

                                                                                         '***'*' 3) '  'E""""' ">eE> C 18                                                                      1\NCHOR.

B' REV. 113 WOLF CREEK SAFETY ANALYSIS REPO FIGURE ___RT 3 6 1 HIGH * - GEN B SAMPLBt~AKT ST M ENERGY PIPE ISC~METRIC INSIDE DRAIN UBEc SHT. f8°~1 1~1NMENT __ l__

5 _ _ _ ___jz_ .-----=8-------'-------'-7-------'--------"6_ _ _ _l_ ____ 4 3 2 

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E' ILIPNC.'B-'*.VS.'Iat.lt.TtHII.ULM~1'RIC: .U.ll 1000(1-Mft) I.,..., **.. 

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13. INTERMEDIATE BREAKS BM19-~
                                                                                            & BM19-03 ARE DELETED.                                   __)

a RESTRAINT lEGEND r IMI!tCY">IIAL OUT* .,NT 111\.IMPEit IIESTRA.'Hf 

                                                                                      ~lsou-*nr*"""'
                                                                                  ~         J---1       *nor RE!fltA.Itn C'
                                                                                      .18\IMI'Ca ltnntAIN1 W.TH 1* AnAcHMrNT TO ""E C""' F'UIUHU ~TAJ\.S 01' "ISlltAINT JYPI:S
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                                                                                    ..A. AMtMOR REV. 18 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.6-1 HIGH ENERGY PIPE BREAK ISOMETRIC STM GEN C SAMPLE & TUBE SHT DRAIN INSIDE CONTAINMENT                                           A:
                                                                                                              <BM19)
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c:,..c ~\,, DIAiCTIONAI. RUlAAINf 1Ur.4PEII RiSfRAINf t--J H GUtD£ IIESTIIAINr AREA OF CHANGE

8 7 6 5 4 3 2 H NOTES:

1. INOIC ATE S TERM IN ALE NDBREAKP OIN T 2
  • INDIC ATE S IN TERMED IATE END BREAK POI NT 3 0 INDIC ATE S STRESS NODE 4 ~ ~~~~:KTE~OI~rCUMFERE NTI AL BREAK AT INTERMED IATE o [;l";] INDIC ATE S LONGITUDINAL BREAK AT IN TERMED IATE BREAKPOINT G

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                '0 8.c=_)- IND IC ATES PIPE BREAK RESTRAINT 10     STEADY STATE THR US T FORC ES FOR EACH BREAK ARE PR OVIDED BELOW F

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                  !FOR FURTHER DETAILS OF RESTR AI NT TYPES SEESECTION3.6.2.3.3J
12. EFFE CTS ANALYSIS RES UL TS FOR EACH ROOM AAE PROVIDED ON THE INDIC ATED SHEET OF TABLE 3.6-4 1323 1122 1126 1204 1114 1203 1107 B

REV. 32 WOLF CREEK U P DATED SAFET Y ANA L YSIS REP ORT FIGURE 3.6-1 HIGH ENERGY PIPE BREAK ISOMETRIC BORON INJECTION TANK INLET SIS OUTSIDE CONT AINMENT (E M02l A (SHEET 37l

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B REV. 19 WOLF CREEK UPDATED SAFIETY ANALYSIS REPORT FIGURE 3.6-1 HIGH ENERGY PIPE BREAK ISOMETRIC BIT AND Sl & RHR RECIRC. SIS INSIDE CONTAINMENT CEM03) AI CSHEET 38l

l:** tZ I8BIIUMPU lEI ~~:rER tEsT RAINT WITH ATTACHMENT IFIIR' 'Ill PIPE OHJ.I.I,J.a:>' DETAILS RESTRAINT 1YPES AREA OF CHANGE REV. 19

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REV. 19 WOLF CREIEK 

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DETAIL ffi~ (F*4) UPDATED SAFIETY ANALYSIS REPORT AREA OF CNOTE 13) FIGURE 3.6-1 CHANGE HIGH ENERGY PIPE BREAK ISOMETRIC ACCUMULATOR INJECTION - LOOPS 2 & 3 A INSIDE CONTAINMENT CEP02) CSHEET 41)

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o1 lfU_Di'IIAA.TI-10-o"M.QI ... EIIIIIIIJAIUU:J. . . lfl'lli4.Z~ Golll ... -r.to'l'..._-f.oe.u_IIIICDilll~o.I'OI-Nn t ff'ncTtMII .. oi.niSIIUUI.TSFOIIE.oloO-I~*IIt~..U. 111\'HiitJOf!ITH!i;II'IDIIC,II,f~D""u;TII;Jif 'f,..,t ~- Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 6-1 HIGH ENERGY PIPE BREAK ISOMETRIC MAIN STM SUPPLY TO TURB AFP OUTSIDE CONTAINMENT ( FC01) SHEET 49

AREA OF CHANGE Lf"~ IIOL4r&ON AUT H w H GUIDI liESTII4111T 

                .,   .UMPiil
  • AlTN;KM':NriAJNT WITH FOil fI FURTWIII 10 PIPE lEI SECTION ~~~AILI 0~ REITIWN A AliCIIO" *J.al ' TTPEI AREA OF CHANGE REV. 19 UPDATED
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Y.~OLE' CHEEK CASE I OUTGOING LINES WITH NORMALLY CLOSED VALVE ( REACT~~:OC~LANT PIPING -{ 

              )               fot...RY                   NOTE:  PRESSURIZER SAFETY VALVES ARE INCLUDED UNDER THIS CASE.

CASE II OUTGOING LINES WITH NORMALLY OPEN VALVES 

                /
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NOTE: THE REACTOR COOLI\NT f'uMP NO. I SEAL IS AS SUM ED TO BE EQUIVALENT TO FIRST FAIL CLOSED OR VALVE FAIL-AS-IS VALVES RESTRAINT CASE ill INCOMING LINES NORMALLY WITH FLOW f -----~----{ NO. I L f NO. 2 -7 t _j_ TT BOUNDARY 

                                -~TEST
                               ,.-           CONNECT! ON CASE rl  INCOMING LINES NORMALLY WITHOUT FLOW f       ----~                   -{

t _j_ BOUNDARY 

                              .       TEST CONNECTION (MEANS OF VERIFYING THAT CHECK VALVE IS CLOSED)

CASEY ALL INSTRUMENTATION TUBING AND INSTRUMENTS CONNECTED DIRECTLY TO THE REACTOR COOLANT SYSTEM IS CONSIDERED AS A BOUNDARY. HOWEVER, A BREAK WITHIN THIS BOUNDARY RESULTS IN A RELATIVELY SMALL FLOW WHICH CAN NORMALLY BE MADE UP WITH THE CHARGING SYSTEM. Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.6*-2 LOSS OF REACTOR COOLANT ACCIDENT BOUNDARY LIMITS

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                                                                                   """'-~**a         ~-*--t LOCATION OF POSTULATED BREAKS IN REACTOR COOLANT
                                                                                                                                                                                        !INCLUDING PRESSURIZER SURGE LIND

GUIDE~i 

           ~REAK PLAN SLAB Rev. 0 WOLF CREEK UPI>ATED SAFETY ANALYSIS REPORT FIGURE 3.6-4 TYPICAL PIPING GUIDE INSTALLATION

WOLJ>? CB:E:EK a.. 

                                                 ~

0 l I (kPIPE~ t---4+--;t-- ~ PI P E ffi~~~~w4 PLAN EAHM (SEE FIG. 3.6-6 SHT. 2 OF 2 SECTIQN 

   ~E~TION@

Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.6-5 TYPICAL ISOLATION RESTRA[NT

WOLF CREEK CONC. WALL EMBEDDED PLATES IN WALL GAP EAHM (SEE SHT. 2) Rev. 0 WOLF CREEK UPDATED SAPETY ANALYSIS REPORT FIGURE 3.6-6 ENERGY ABSORBING HONEYCOMB MATERIAL - LARGE GAP RESTRAINT _SHEET 1

WOLE~ CHEEK NOTE A: 2 WELDS EACH E~J D ~~T OPF'OSITE SIDES. (FIELD HAS OPTION OF. WELDING ANY .:2 OPPOS *- 1NC3 SIDES) (TYP) I 2 I 2 I-I I- <D 

                                               <! <D I     z z         (./)

w w _jw _j SUBSTRUCTURE FRONT PLATE MAY rf&~~- Et~ EN BE CURVED OR FLAT BEND RADIUS FOR SEE NOTE.A CURVED PLATES =0.75 1/8" THK BACK Ft x NOM. DIA. OF PIPE_-------~ 

             ..QJBECTION OF
  • APPLIED LOAD
                                             .01   II     THK COVEF~       fE_

1/16" THK (4 SIDES) PERFORATED FRONT f t - - - - Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.6-6 TYPICAL PREFABRICATED ENERGY ABSORBING HONEYCOMB MATERIAL INSTALLATION suu:r 

                                       "-----------*-                            -;::~

WOLF CREEK GAP CONC. WALL UPSET RODS 

                    --WT
                -----lr EMBEDDED PLATES IN WALL Rev. 0 WOLl!' CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.6-7 TYPICAL UPSET Roo lARGE GAP RESTRAINT

WOLF CREEK SLAB3 *---**--*.:..l-..r--,....,..._,_'-------- SUPPORT*---**-- WHERE CONC. / REO'D SHIM PACK ---,r- WALL___.- 

        ~~====W=====l~-
         -lAPS/ZED TO LIMIT                       ..;.

PIPE MOTION RESTRAINT CUT TO SUIT IN FIELD Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3 . 6-8 TYPICAL CLOSE GAP RESTRAINT

WOLF CREEK L c0 Effective damping of pipe (effective translational damping) Effective damping of substructure Distance from elbow to restraint Distaoce from elbow to restraint to effective mass of rotating pipe Distance from restraint to point of fixity Initial gap between pipe and restraint 

                                                                                  =  Degree of freedom representing rotation of the elbow at the restraint Degree of freedom representing translational motion of pipe at the restraint Degree of freedom representing the motion of the restraint substructure Note:   Tne coupied second order differential equations developed using Lagrangian dynamics representing the model, are solved by a time step integration procedure via computer to yield the time history acceieration velocities and displacement~

of the defined masses. Fj =Jet thrust reaction force (See Reference 5) x1 = Effective mass of rotating pipe (between broken end of pipe and restraint) M2 = Effective mass of rotating portion of pipe between fixity and restraint 

   = Effective mass of restraint substructure Elastic-plastic clock spring representing stiffness of pipe at a restraint due to deflection/rotation caused by Fj (plastic hinge determined by equations in Section 3.6.2.3.4(a))

K p Elastic spring representing stiffness of pipe between point of fixity and restraint

  • Ke = Elastic-plastic spring representing energy dissipating device.

(Active only in compression - provides rebound capability.) K s Elastic-piastic spring representing stiffness of restraint substructure. Rev; 0 WOL!' CUBit DPDA'I'ZD SAFZ'I'T AJIALYSIS UPOR'I' FIGURE 3.6-9 . LUMPED-PARAMETER MODEL ~IPE RESTRAINT SYSTEM

WOLF CREEK 3.7(B) SEISMIC DESIGN In addition to the steady state loads imposed on the system under normal operating conditions, the design of equipment and equipment supports requires that consideration also be given to abnormal loading conditions, such as earthquakes. Seismic loadings are considered for earthquakes of two magnitudes: Safe Shutdown Earthquake (SSE) and Operating Basis Earthquake (OBE). The SSE is defined as the maximum vibratory ground motion at the plant site that can be reasonably predicted from geologic and seismic evi-dence. The OBE is that earthquake which, considering the local geology and seismology, can be reasonably expected to occur during the plant life. For Westinghouse-supplied items, refer to Section 3.7(N). The following material is in addition to Section 3.7(N) and ap-plies to structures, systems, and components not supplied by Westinghouse. This section describes the techniques and discusses the parameters used to develop seismic loadings and criteria for seismic Category I structures, systems, and components. The seismic responses of the major seismic Category 1 structures (containment, auxiliary/control, diesel generator, and fuel build-ing) were originally generated for four sites (Callaway, Wolf Creek, Sterling, and Tyrone). Seismic design envelopes were developed by use of the most restrictive site conditions imposed by any one of the four original sites or by generic design cri-teria which are conservative for each of the sites. With the cancellation of the Tyrone plant, however, the four site envelop-ing approach was modified, for work not yet completed, to include only the remaining three sites. The seismic design envelopes were not revised to reflect the cancellation of the Sterling plant; therefore, since the design of all powerblock structures, sys-tems, and components is based on the responses for three or four sites, the powerblock design is conservative for the remaining two sites. A further discussion of the multiple site enveloping criteria, as applied to the seismic design of the WCGS powerblock, is contained in Section 3.7(B).2.2. The seismic response of the seismic Category 1 ESW vertical loop chase structure was generated by using the design envelope developed for the auxiliary/control building. 3.7(B).1 SEISMIC INPUT 3.7(B).1.1 Design Response Spectra The site design response spectra in compliance with Regulatory Guide 1.60 are illustrated in Figures 3.7(B)-1 and 3.7(B)-2, in both the horizontal and vertical directions for the SSE. For the OBE, the design response spectra values were taken as 60 percent 3.7(B)-1 Rev. 29

WOLF CREEK of the SSE. The values shown are for the site with maximum ampli-fication. Section 2.5.2 and Section 2.5 of BC-TOP-4-A (Ref. 3) discuss the effects of focal and epicentral distances from the site, depths between the focus of the seismic disturbances and the site, existing earthquake records, and the associated amplifica-tion of the response spectra. Earthquake duration influences only the number of loading cycles on equipment because the equipment is designed for the elastic range in accordance with the analytical procedures outlined in BC-TOP-4-A. A 20.48-second duration is considered to be adequate for the time-history type of analysis used for the structures and equipment. The design response spectra and earthquake time-histories are applied in the free field at finished grade for all sites. 3.7(B).1.1.1 Bases for Site Dependent Analysis Section 2.5.2 and BC-TOP-4-A, Sections 2.4 and 2.5, describe the bases for specifying the vibratory ground motion for design use. 3.7(B).1.2 Design Time History Synthetic earthquake time-histories were generated because the response spectra of recorded earthquake motions do not necessarily envelope the site's design spectra. Figures 3.7(B)-3 and 3.7(B)-4 show the synthetic earthquake time-history motions in the horizon-tal and vertical directions, respectively. The time-histories shown were truncated to 20.48 seconds for use in the FLUSH finite element analyses discussed in Section 3.7(B).2.4.2. Figures 2-13, 2-14, 2-17, and 2-18 of BC-TOP-4-A show that the response spectra of the synthetic time-histories for the horizontal and vertical directions envelope the corresponding design spectra for 1 per-cent, 2 percent, 5 percent, 7 percent, and 10 percent damping. Section 2.5.1 of BC-TOP-4-A describes the generation of a typical synthetic earthquake time-history. Typical foundation-level, free-field acceleration response spectra for each of the three sites are presented in Figures 3.7(B)-9A through D. Their envelope is presented in Figure 3.7(B)-10. All curves overlay the WCGS 60-percent design response spectra. Due to site amplification of the seismic input, deconvolution of the SNUPPS control motion applied at grade will inevitably show an attenuation of the foundation level response relative to grade level input motion. Attenuation is maximized at frequencies 3.7(B)-2 Rev. 0

WOLF CREEK corresponding to the soil deposit fundamental frequencies. Hence, at particular frequencies, the computed foundation-level, free-field response spectra for WCGS can be expected to and does fall below the WCGS 60-percent design spectra at some frequencies, similar to the ground spectrum and as shown by the Humboldt Bay results (Ref. 1). 3.7(B).1.3 Critical Damping Values For seismic Category I structures, systems, and components not supplied by Westinghouse, the range of damping values (in percent of critical) is shown in Table 3.7(B)-1, is discussed in Sections 2.2 and 3.2.1 of BC-TOP-4-A, and is in compliance with Regulatory Guide 1.61 as discussed in Appendix 3A. The applicable allowable stress values are given in Section 3.8 for the various loading combinations, which include seismic loadings. The testing of cable tray systems, as discussed in Section 3.10(B).3, clearly demonstrates that a substantial amount of energy is absorbed by friction between the adjacent moving cables and through friction between cables and the cable tray. This phenomenon was also observed to be amplitude dependent. That is, the greater the input level the more pronounced were these losses. Equating these losses during the test program resulted in predicted equivalent viscous damping of up to 50 percent in some cases. After tabulating the results of the several hundred earthquake-type vibration tests and cable tray systems, the allowable damping as a function of the level of seismic input motion was determined. A maximum value of 15 percent of critical was used for cable tray damping. Damping of supports for conduit is 7 percent of critical, regardless of input level. 3.7(B).1.4 Supporting Media for Seismic Category I Structures In the FLUSH finite element analyses, the containment building was supported on stabilized backfill down to a depth of 25 feet below grade. Also in the analyses, the auxiliary/control building was founded directly on in-situ material. The diesel generator and fuel buildings for Wolf Creek analyses were supported on crushed rock. The crushed rock extended from the bottom of the base mats down to a depth below grade of 13 feet in the Wolf Creek analyses. Descriptions of the supporting media at the Wolf Creek site are provided in Section 2.5. A list of the major seismic Category I structures and the depth of the soil and/or backfill deposits over the bedrock for each structure is given in Table 3.7(B)-2. 3.7(B)-3 Rev. 6

WOLF CREEK The foundation embedment depth and minimum base dimension for each seismic Category I structure are provided in Table 3.7(B)-3, along with the method of seismic analysis utilized for each structure. 3.7(B).2 SEISMIC SYSTEM ANALYSIS 3.7(B).2.1 Seismic Analysis Methods Seismic Category I structures, systems, and components were classified in accordance with NRC Regulatory Guide 1.29, as shown in Section 3.2. These structures, systems, and components were analyzed for two earthquake conditions, the SSE and the OBE. The analytical methods utilized for the analysis of the different seismic Category I structures are summarized in Table 3.7(B)-3. Lumped-mass models were developed for the containment, fuel auxiliary/control, and diesel generator buildings, following the techniques discussed in Section 3.2 of BC-TOP-4-A. Figures 3.7(B)-17 through 3.7(B)-20 present the models developed for these structures. Mass and cross-sectional properties were calculated for the two principal normal horizontal directions and the vertical direction. The lumped-mass models of the major seismic Category I structures were incorporated, along with models of the significant non-Category I structures, into finite element models, of which Figure 3.7(B)-13 is typical. Time history analyses were performed using these finite element models, following procedures described in Section 3.7(B).2.4.2. The results obtained from these analyses included maximum accelerations, inertia forces, shears, axial forces, moments, and floor response spectra. It was not possible to obtain displacements directly from the finite element analyses. Consequently, the procedure outline in Section 3.7(B).2.4.2 was used to determine building displacements. The other seismic Category I structures (refueling water storage tank and valve house, emergency fuel oil storage tanks, and associated access vaults) are small compared to the major structures and are not directly adjacent to the major structures. Consequently, structure-to-structure interaction between the major seismic Category I Structures and these remaining seismic Category I structures is considered to be minimal. Therefore, the remaining structures were not included in the main finite element models. The ESW Vertical Loop Chase seismic Category I structure is small in size compared to the major structures and is directly adjacent to the control building. Consequently, the structure-to-structure interaction between the auxiliary/control building and ESW Vertical Loop Chase was analyzed using the original design envelope for the auxiliary/control building. The interaction with the major structure and the ESW Vertical Loop Chase is considered minimal. 3.7(B).2.2 Natural Frequencies and Response Loads A summary of significant natural frequencies for the major seismic Category I structures is provided in Table 3.7(B)-4. The seismic responses generated for these structures, including accelerations, inertia forces, shears, axial forces, moments, and displacements are provided in Table 3.7(B)-5 through 3.7(B)-8. Typical floor response spectra are presented in Figures 3.7(B)-14 and 3.7(B)-15 for the polar crane and upper steam generator support locations, respectively. 3.7(B)-4 Rev. 32

WOLF CREEK All seismic responses were originally generated for the four SNUPPS sites, using the average soil properties for each site. As discussed previously, the responses, from either three or four sites, were enveloped and used in the design of all structures. Likewise, all subsystems and components were designed using either the three or four site envelopes of the floor response spectra of the site specific spectra. The effects of soil property variation on seismic responses were accounted for by the multiple site enveloping procedures detailed above. 3.7(B).2.3 Procedure Used for Modeling 3.7(B).2.3.1 Lump Mass Modeling A description of the procedure used to locate lumped masses for the seismic system analyses for seismic Category I structures and equipment is provided in Section 3.2 of BC-TOP-4-A. A similar discussion for piping systems is provided in Section 3.2 of BP-TOP-1 (Ref. 4). 3.7(B).2.3.2 Finite Element Modeling Procedures used for finite element analysis modeling in seismic system analyses of seismic Category I structures is in accordance with the FLUSH computer program criteria, Reference 2. 3.7(B).2.4 Soil/Structure Interaction Foundation embedment depth below grade, minimum base dimension, and method of analysis are given in Table 3.7(B)-3. The effect of soil-structure interaction was taken into account by coupling the structural model with the foundation medium. 3.7(B).2.4.1 Lumped Parameter Representation A seismic analysis utilizing a lumped mass model on an elastic half space with strain independent soil properties was performed for comparison with the FLUSH finite element results. The purpose of this comparison was to provide a check on the FLUSH analysis. Figure 3.7(B)-12 shows the soil-structure model developed for the containment building. The response spectrum curves obtained by utilizing elastic half space analytical techniques compared favorably with the envelope curves developed for design use on WCGS. 3.7(B).2.4.2 Finite Element Representation The finite element method of analysis was used to determine the seismic responses of the four major seismic Category I structures and the emergency fuel oil storage tanks. Additionally, displacement of the four major Category I structures was determined by using the DISCOM computer program (see Section 3.8(A).1.24) along with time histories from the finite element analysis. 3.7(B)-5 Rev. 10

WOLF CREEK Figure 3.7(B)-13 shows a finite element model typical of the ones used to analyze to major power structures. The analytical model is provided with transmitting boundaries on both the left and right sides. The model also consists of two types of elements--displacement-compatible isoparametric quadrilateral elements (solid elements) and linear bending elements (beam elements). Usage of transmitting boundaries, elements, and analytical techniques are described in Reference 2. The computer program FLUSH, of the same reference, was used to perform the analysis. Models, typically shown in Figure 3.7(B)-13, were used to perform soil-structure interaction analyses. The site dependent soil properties were used. The vertical dimension of each soil element is equal to or less than Cs/5f, where Cs is the lowest soil element shear wave velocity reached during iterations and f is the highest frequency of interest to be transmitted through the soil profile. The highest frequency used was 25 Hz. In the analyses for the same buildings with site dependent soil parameters, the structural elements remained unchanged. The site dependent soil properties consisted of strain dependent damping and modulus relationships for each material. In general, the soil properties are nonlinear in character. An iterative process was used to obtain equivalent linear properties which are strain dependent. The methods generally used for such an analysis are included in the computer program FLUSH. 3.7(B).2.5 Development of Floor Response Spectra Acceleration time-histories obtained from the FLUSH finite element analyses were used in computing the floor response spectra for the major seismic Category I structures. The spectra were generated following the procedures outlined in Section 5.2 of BC-TOP-4-A, using the SPECTRA computer program (see subparagraph 3.8A.12). 3.7(B).2.6 Three Components of Earthquake Motion Procedures for considering the three components of earthquake motion in determining the seismic response of structures, systems, and components follow the recommendations of Regulatory Guide 1.92 and are described in Section 4.3 of BC-TOP-4-A and Section 5.1 of BP-TOP-1. 3.7(B).2.7 Combination of Modal Responses Combination is done according to the criterion of "the square-root-of-the-sum-of-the-squares" (SRSS). Section 4.2.1 of BC-TOP-4-A describes the techniques used to combine modal responses for structures and equipment. For piping systems, closely spaced modes were determined per NRC Regulatory Guide 1.92, Equation 4. 3.7(B)-6 Rev. 10

WOLF CREEK 3.7(B).2.7.1 Significant Dynamic Response Modes The static load equivalent or static analysis method involves the multiplication of the total weight of the equipment or component member by the specified seismic acceleration. Multiple degree-of-freedom systems which may have had frequencies in the resonance region of the amplified response spectra curves were analyzed by using a static load of 1.5 times the peak acceleration or the applicable floor response spectra to account for the contribution of higher modes. Multiplication factors less than 1.5 were not used. Multiplication factors were not used in the equivalent static load method of analysis of conduit and cable tray supports which were multiple-degree-of-freedom, simple span, or cantilever beams. In these cases, other conservatisms such as lumping of masses (i.e., at the center of the simple beam span or at the end of the cantilever beam), consideration of mode shapes, and/or verification by dynamic analysis precludes the need for the use of multiplication factors. Components which can adequately be characterized as a single-degree-of-freedom system were analyzed by using directly the seismic acceleration from the applicable floor response spectra. For piping, refer to BP-TOP-1, Section 2.3.2, and Appendix D. 3.7(B).2.8 Interaction of Nonseismic Category I Structures With Seismic Category I Structures With the use of the computer program FLUSH (see Table 3.7(B)-3), seismic analyses of all seismic Category I structures included the effects of adjacent, significant nonseismic Category I structures. In addition, neither structural failure nor interference causing displacements during an SSE were permitted. Elastic analyses have been performed to assure that the nonseismic Category I structures will not collapse onto seismic Category I structures when subjected to an SSE and will be allowed to reach 0.9 fy or 0.9 of any failure mode. Section 3.4 of BP-TOP-1 describes the techniques used to consider the interaction of seismic Category I piping with nonseismic Category I piping. 3.7(B).2.9 Effects of Parameter Variations on Floor Response Spectra Section 5.2 of BC-TOP-4-A describes the effects on floor response spectra due to expected variations of structural properties, dampings, soil properties, foundation-structure interaction, etc. 3.7(B).2.10 Use of Constant Vertical Static Factors Constant vertical load factors were not used for the analysis of seismic Category I structures, systems, and components. The methodology for vertical seismic analysis of structures is discussed in Sections 3.0, 4.0, and 5.0 of BC-TOP-4-A. The methodology for vertical seismic considerations for equipment is in accordance with IEEE 344, as amended in Section 3.10(B). 3.7(B)-7 Rev. 10

WOLF CREEK 3.7(B).2.11 Method Used to Account for Torsional Effects Torsional effects, if significant, were included in the horizontal models at locations of major mass and/or structure eccentricity. Section 3.2 and Appendix C of BC-TOP-4-A show the techniques used to account for torsional effects. 3.7(B).2.12 Comparison of Responses Not applicable, since only the time-history method of analysis is used on major seismic Category I structures. 3.7(B).2.13 Methods for Seismic Analysis of Dams Refer to Section 2.5.6. 3.7(B).2.14 Determination of Seismic Category I Structure Overturning Moments The effects of overturning moments were evaluated by the simplified, conservative static application of forces caused by the SSE. The more sophisticated energy methods shown in Section 4.4 of BC-TOP-4-A were used when the static method indicated unrealistic results. This section also includes a description of the methods used to compute foundation reactions and to account for vertical earthquake effects. 3.7(B).2.15 Analysis Procedure for Damping The analysis procedure employed to account for damping in different elements of the model of a coupled system is described in Sections 3.2 and 3.3 of BC-TOP A. The criteria used to account for composite damping in the coupled system with different elements are included. The analysis is based on the use of seismic Category I structural models which include a simplified version of the NSSS model provided by the NSSS supplier. 3.7(B).3 SEISMIC SUBSYSTEM ANALYSIS 3.7(B).3.1 Seismic Analysis Methods Also see Section 3.7(B).2.1. Section 2.0 and Appendix D of BP-TOP-1 describe the basis for the simplified dynamic analysis technique used in lieu of response spectrum analyses for piping. Simplified dynamic analysis was not used for seismic Category I structures, systems, and components other than piping. 3.7(B).3.2 Determination of Number of Earthquake Cycles Fatigue analysis, where required by the codes, was performed by the supplier as part of the stress report. The earthquake transients are a part of the mechanical loading conditions specified in the equipment specifications. The origin of their determination was separate and distinct from those transients resulting from fluid pressure and temperature. The fluid pressure and temperature transients are given in Section 3.9(N).1.1. A description of the procedures followed in fatigue evaluations is given in Section 3.7(N).3.2. 3.7(B)-8 Rev. 10

WOLF CREEK The procedures used to determine the number of earthquake cycles for piping during one seismic event are discussed in Section 6.2 of BP-TOP-1. Equipment was designed on the basis of analytical results. The design criteria for equipment assumed elastic behavior. Therefore, the number of loading cycles need not be considered in the design. Fatigue was not considered in the design of seismic Category I structures, because the occurrence of full design earthquake loads is too infrequent to warrant consideration of fatigue design, and the calculated stresses and strains are below yield. 3.7(B).3.3 Procedure Used for Modeling See Section 3.7(B).2.3. 3.7(B).3.4 Basis for Selection of Frequencies Fundamental frequencies of subsystems and components were calculated in accordance with the procedures outlined in Section 4.2.1 of BC-TOP-4-A. To avoid resonance, the fundamental frequencies of subsystems and components were, where possible, selected in such a way as to avoid excessive load amplifications. If the subsystem's or component's frequencies fell within the amplified region of the forcing functions, the subsystems or components were adequately designed for the applicable loads. 3.7(B).3.5 Use of Equivalent Static Load Method of Analysis See Section 3.7(B).2.7.1. 3.7(B).3.6 Three Components of Earthquake Motion See Section 3.7(B).2.6. 3.7(B).3.7 Combination of Modal Responses The seismic design of the piping and equipment included the effect of the seismic response of the supports, equipment, structures, and components. The system and equipment response was determined, using three earthquake components--two horizontal and one vertical. The design ground response spectra specified in Section 3.7(B).1 were the bases for generating these three input components. The input may be the floor time-history motions or floor response spectra. These floor time-history motions and/or floor response spectra are generated for two perpendicular horizontal directions (i.e., N-S and E-W), and the vertical direction. System and equipment analysis was performed with these input components applied in the N-S, E-W, and vertical directions. The damping values used in the analysis were those given in Table 3.7(B)-1. 3.7(B)-9 Rev. 10

WOLF CREEK In computing the system and equipment response by modal analysis, the square root of the sum of the squares of the modal contributions was used to combine all significant modal responses in each direction (see Section 3.7(B).2.7). The combined total response was calculated, also using the SRSS formula applied to the resultant unidirectional responses. For instance, for each item of interest, such as displacement, force, stresses, etc., the total response is obtained by applying the above-described method. This method can be written in equation form. The resultant response at a given node point for the item of interest, for example, , is 3 1/2 2 s = si 3.7(B)-1 i=1 where i is the response in the i-th direction defined as N 1/2 2 i = sij 3.7(B)-2 j=1 with subscripts i and j in Equations 3.7(B)-1 and 3.7(B)-2 representing the i-th direction of input and the j-th mode (for a total of N significant modes). The term ij is the maximum response in the j-th mode for input in the i-th direction, as determined by response spectrum modal analysis. The system and equipment response can also be determined, using time-history analyses. 3.7(B).3.8 Analytical Procedures for Piping Section 2 of BP-TOP-1 describes the analytical techniques applicable to piping systems outside of the Westinghouse scope. Section 4 of BP-TOP-1 discusses the effect of differential building movement on piping. 3.7(B).3.9 Multiple Supported Equipment and Components With Distinct Inputs See Section 3.7(B).3.8. 3.7(B)-10 Rev. 10

WOLF CREEK 3.7(B).3.10 Use of Constant Vertical Static Factors See Section 3.7(B).2.10. 3.7(B).3.11 Torsional Effects of Eccentric Masses The significant torsional effects of valves and other eccentric masses are taken into account in the seismic piping analyses by the techniques discussed in Section 3.2 of BP-TOP-1. 3.7(B).3.12 Buried Seismic Category I Piping Systems and Tunnels Procedures are defined in Section 6.0 of BC-TOP-4-A. All buried components are designed to remain functional after a seismic event by limiting the calculated stresses under all loading combinations, including earthquakes. 3.7(B).3.13 Interaction of Other Piping With Seismic Category I Piping Section 3.4 of BP-TOP-1 describes the techniques used to consider the interaction of seismic Category I piping with nonseismic Category I piping. 3.7(B).3.14 Seismic Analyses for Reactor Internals See Section 3.7(N).3.14. 3.7(B).3.15 Analysis Procedure for Damping See Section 3.7(B).2.15. 3.7(B).3.16 Seismic Analysis for Cable Trays The scope of the cable tray and conduit raceway test program included the evaluation of a large number of variable in the design of cable trays. Included in the test report are discussions of the following variables: o Type of tray o Type and length of hanger o Location of splices o Number of tiers o Trapeze and cantilever support 3.7(B)-11 Rev. 10

WOLF CREEK o Connection details, such as single clip angle double clip angle guesseted clip tray to strut type hanger o Type and location of bracing o Amount of cable fill o Size and distribution of cables o Cable ties o Combined conduit and tray systems o Sprayed fire protection material In order to evaluate the effects of these and other variables, over 2,000 individual dynamic vibration tests were performed over a period of 11 months of testing. As a result of these tests, over 50 volumes of raw data were generated and evaluated. The results of the evaluation of these data form the basis for the conclusion contained in the test report and the design recommendations implemented in the WCGS design. In addition to the wide range of variables that were evaluated, tests were performed on tray and strut systems similar to the WCGS design. As a result of the evaluation of the variables described above and the testing of hardware and support configurations similar to the WCGS design, a set of design recommendations was formulated. These recommendations were developed to be generally applicable to a wide variety of hardware and specifically applicable to the support configurations used by this project and the other test program participants. For example, the recommended damping in intermittently braced strut supported trapeze hanger systems was determined from the data of over 100 dynamic tests on these type of systems. Figure 3.7(B)-21 shows the recommended damping as a function of floor acceleration in the form of a bilinear curve. As can be seen from this curve, the recommended damping, for the most part, represents a lower bound of all the data obtained from the test program. Similar conservative recommendations were formulated from the results of the test program for other aspects of design. Consequently, it is concluded that the design recom-3.7(B)-12 Rev. 10

WOLF CREEK mendations formulated as a result of the cable tray and conduit raceway test program are broadly applicable to the design of strut supported raceway systems and were conservatively applied in the design of the raceway supports. The test fixture used to test cable trays was specifically designed for this test program. Its inverted pendulum design permitted seismic input to suspended tray support systems. Additionally, the fixture was designed to accommodate a 40-foot-long tray system segment of up to five tiers and a hanger of up to 13 feet in length. Sufficient width was provided in the test bay to accommodate two parallel runs, including cross connections and attached conduit. This facility allowed for testing of long, multitiered tray systems with various bracing arrangements. The test program included tests of a large number of varied tray types and support types in various configurations. These test configurations were used during the testing program in order to simulate the actual field installed conditions. Supports with or without bracing and with multitier cable trays were tested. In addition, a combined system configuration comprised of various tray fittings such as tees, elbow, vertical bend, and multitiers of straight cable tray runs was tested. In view of the scope of testing and the various test setups, it was concluded that these tests simulate conditions encountered in the field and, therefore, the results of the testing would be applicable to the design of cable trays on the WCGS project. In a linear dynamic analysis velocity dependent forces (i.e. viscous damping) were introduced to account for various mechanisms of energy dissipation. These mechanisms include such things as: friction and slip-in bolted connections, hysteresis, fluids, and no doubt other mechanisms as well. Since these various mechanisms cannot be accounted for explicitly in a linear analysis, their effect is lumped in a single viscous damping. Dynamic testing is used to determine an effective viscous damping, appropriate for seismic response. This procedure is common to all structural dynamic analysis. During the cable tray and conduit raceway test program, the random vibration of cables was identified as one of the significant energy dissipating mechanisms. This occurred because the cables represent most of the mass of the system, are able to move relative to each other, and were not rigidly attached to the supporting tray. 3.7(B)-13 Rev. 10

WOLF CREEK During the tests, this phenomenon manifested itself as a noticeable relative movement and impact of the cables within the tray. As is the case with other energy dissipating mechanisms, this effect was quantified in terms of an equivalent viscous damping based upon the relationship between the recorded response and the applied input to each test specimen. The test report entitled "Cable Tray and Conduit Raceway Seismic Test Program" provides a detailed discussion of the methods used to compute an equivalent viscous damping from the recorded results of the dynamic tests. This discussion can be found in Section 5 with supplementary information Appendices G, H, and I. The cable tray and conduit raceway test input loading was applied at 45-degree (vector biaxial) because the shake table used was limited to vector biaxial motion. In choosing the 45-degree relationship (i.e, horizontal equals vertical), the floor response spectra of many containments and auxiliary buildings were reviewed, and this equality of horizontal and vertical motion was deemed most appropriate. IEEE-344, and NRC regulatory guides recommend, but do not require, independent biaxial input. In the case of raceways, the modes of vibration are symmetrical and are dominantly either horizontal or vertical and so would be adequately excited by vector biaxial motion. As the different modes of a given raceway generally have quite distinct resonant frequencies, there is no problem introduced by the zero phase between horizontal and vertical loading (i.e., vertical and horizontal responses will be randomly varying in and out of phase even though the vertical and horizontal inputs are in phase). Independent biaxial input is preferred in nonsymmetrical cases and in the possible but unusual case of testing a structure with a mode whose axis of sensitivity would be at 90 degrees to the vector biaxial input, and hence not excited. The raceways are simple structure systems with distinct vertical, transverse, and longitudinal modes; this was confirmed during testing. Therefore, the test results are not affected by the use of vector biaxial input. As described above, widely spaced modes of vibration with little cross coupling were observed during the testing. For example, longitudinal swaying modes were quite low (1.8 Hz), transverse modes followed (3.2 Hz) with tray modes following at 6.1 and 15 Hz for a typical 4'6" single tier unbraced raceway. This data is illustrated in Figures 7.8 and 7.13 of Volume 1 (of test report "Cable Tray and Conduit Raceway Seismic Test Program") for a 100-percent cable loaded raceway of 0.10 g peak raceways are illustrated in relevant data. 3.7(B)-14 Rev. 10

WOLF CREEK The purpose of the cable tray test program was essentially to verify the mathematical model used in the analysis, not to seismically qualify the raceway systems by testing only. Damping of the cable tray system is dependent on the amount of cable in the trays and the input amplitude of vibration. Figure 3.7B-22 presents the lower bound values of equivalent viscous damping as a function of input floor response spectrum ZPA and amount of cable in tray. To be able to use the maximum value of damping, 20 percent, the instructure response spectra must have at least a ZPA value of 0.35 g and the tray must be at least 509 percent full by weight of cable. During the cable tray and conduit raceway seismic test program, various tests were performed on conduit runs on a trapeze raceway to determine their dynamic characteristics. A large number of variables were considered in this test program. The description and results of conduit raceway testing can be found in Section 8 of the test report. The critical damping value computed from test data is 7 percent at 0.1 g input acceleration. High damping value trend was observed for input acceleration higher than 0.1 g. But at present time, for design of conduit raceway system it is recommended to use 7 percent critical damping for all levels of input acceleration at and above 0.1 g. For lower input acceleration, it is recommended to use linear interpolation from 7 percent to 0 percent damping for 0.1 g input to zero input acceleration. Seismic Qualification of Category 1 instrumentation and Electrical Equipment is discussed in Section 3.10(B). The computed damping values from the various tests are tabulated in Appendix K of the test report. Data was taken from these tables and plotted as shown in Figure 3.7(B)-21. On this figure, the data points of computed equivalent viscous damping are plotted as a function of input acceleration (floor spectrum ZPA) for over 100 tests of various braced strut hanger tray systems. These results represented all the data from simulated earthquake inputs. Low level sinusoidal and snap back test data are not included, since they are not directly applicable. Since these tests represented a wide variety of tray type, connection details, struts, and cable configuration, there is a broad scatter in the data. These data, however, do clearly show that the recorded responses of the tested tray systems are best described by a dynamic system with an equivalent viscous damping. It should be noted that the data realistically can be utilized with accepted curve fitting techniques to obtain a "best-fit" curve which reflects the statistical average of the test data. Such an approach would result in a maximum damping value far in excess of the 3.7(B)-15 Rev. 10

WOLF CREEK conservative 20 percent value. However, in the interest of conservatism, a bilinear curve, which effectively bounds the lower end of nearly all the points, was utilized. This curve is given in Figure 3.7(B)-21. This curve represents the recommended design values of equivalent viscous damping. In addition to the determination of equivalent viscous damping, as described in the test report, linear analysis was performed on finite element models of several of the tray system test setups. These analyses confirmed that a very high viscous damping was required in order to predict responses similar to those recorded during the dynamic testing. These analyses confirmed that the application of the damping values recommended for design in a linear analysis was consistent with the results of the test program and, therefore, would result in a conservative design of support systems. Stainless steel 600 volt fire-resistive control and power cables are routed independent of raceways. Fire-resistive cable will be supported by stainless steel unistrut attached to the concrete walls at intervals governed by span loading. Generally, the supports will be standard design used for small conduit, except for the use of stainless steel unistrut and clamps. Seismic testing and analysis verify the adequacy of the supports for fire-resistive cable. 3.7(B).4 SEISMIC INSTRUMENTATION 3.7(B).4.1 Comparison with Regulatory Guide 1.12, Rev. 2 (March 1997) The seismic instrumentation program complies with Regulatory Guide 1.12, Revision 2, except for the location of one (1) instrument. The regulatory guide position, as stated in TABLE 3.7(B)-9, paragraph 1.2 item 3 is that accelerographs should be located at two elevations (except the foundation) on a structure inside containment. The Wolf Creek design has two (2) accelerographs, one (1) located inside containment and the other located at a different elevation on the outside containment wall. The design complies with the intent of the regulatory guide in that two (2) accelerographs measure the containment structure response at two (2) different locations of the containment. And as stated in the regulatory guide, neither of these two (2) accelerographs are located on the containment foundation. 3.7(B)-16 Rev. 25

WOLF CREEK 3.7(B).4.2 Location and Description of Instrumentation A seismic instrumentation program is provided to monitor the effect of earthquakes at the plant site and to collect data necessary to evaluate the safety impact of an earthquake on seismic Category I structures, systems, and components. Detailed location for accelerographs is chosen to coincide with significant points in the seismic model. All seismic instrumentation is designed to seismic Category I requirements, including the battery emergency power supplies for each individual accelerograph and for the instrumentation located in the main control room. Power for normal operation, and power for maintaining the charge on the emergency power supply batteries is provided from the non-Class IE 120V ac instrument bus. 3.7(B).4.2.1 Strong Motion Accelerometer/Recorder Triaxial accelerographs are installed at appropriate locations throughout the plant to provide data on the frequency, amplitude, and phase relationship of the seismic response of the containment structure and the seismic input to other seismic Category I structures, systems, and components. A triaxial accelerograph consists of a self contained, battery backed up, powered triaxial accelerometer and recorder in a single enclosure. Each accelerograph functions to detect, measure and store data from a seismic event. One accelerograph is located in the free field EL. 2000 - 0 (SGAR0001), such that it will measure the input vibration, or ground motion, at a location outside of a structure of the plant Accelerographs are also provided on the containment base, EL. 2000 - 0 (SGAE0001); on the containment building at the operating floor level EL. 2056 - 6 (SGAE0002) above and axially aligned with the accelerograph on the base slab and in the auxiliary building, near the control room air filters (El. 2047'-6") (SGAE0005); on the auxiliary/ control building base slab EL 1974 - 0 (SGAE0004) and on the Floor, Area 1, reactor building, EL.2026-0 (SGAE0003). 3.7(B).4.2.2 Seismic Trigger Each triaxial accelerograph has an adjustable seismic trigger. This hardware or hard trigger is set in each triaxial accelerograph to be above the ground motion level of the location to avoid false triggers. When triaxial accelerograph (SGAR0001) located in the free field and triaxial accelerograph (SGAE0001) located on the containment base both trigger, the AND combination is sensed by the Network Control Center (NCC) (SGAR0009), located in the main control room. A software trigger or soft trigger is sent to all triaxial accelerographs to capture and process the event data. 3.7(B)-17 Rev. 25

WOLF CREEK 3.7(B).4.2.3 Network Control Center (SGAR009)/Supporting Equipment An NCC and additional supporting equipment is mounted in cabinet SG058 in the main control room. The NCC and supporting equipment provide the following functions:

  • The NCC monitors the status of each triaxial accelerograph and provides messages of off normal status.
  • The NCC polls all triaxial accelerographs and provides a signal (soft trigger) to all triaxial accelerographs if a seismic event is detected.
  • The NCC provides polls all triaxial accelerographs and automatically down loads seismic events.
  • The NCC provides self test functions.
  • The NCC provides outputs to alarms in the main control room.
  • DC power supplies: Two (2) power supplies are provided. One (1) supplies power and maintains the charge in batteries for triaxial accelerographs located in the free field and throughout the plant. The second power supply provides power to the NCC. This supply is battery backed up, so that the NCC will be operational in the event of a loss in cabinet 120V ac power.
  • A Laptop PC performs the data retrieval system function.

The function of the system can be described as the remotely located accelographs store the event data, the NCC polls the accelerographs, and downloads the data, and the PC monitors the NCC and automatically downloads and analyzes the data for a seismic event. The PC is also used for parameter setting in the remotely located accelerographs and in the NCC. 3.7(B).4.3 Control Room Operator Notification An annunciator in the main control room is actuated whenever the seismic monitoring system has been triggered, calling the operator's attention to the fact that an event has occurred. Additional annunciation in the main control room is actuated if OBE and SSE levels are triggered. 3.7(B)-18 Rev. 25

WOLF CREEK Following a seismic event, all accessible data will be processed for an initial determination of the earthquake level. Where the site-related safety items (ultimate heat sink, etc.) are designed to an OBE less than the power block OBE, the unit will be shut down and site related items examined when an event of site OBE magnitude or greater occurs. If no evidence of damage is detected, the unit will be returned to service and the NRC notified. 3.7(B).4.4 Comparison of Measured and Predicted Responses If the OBE has been exceeded, a response spectrum will be automatically calculated and displayed on a PC located in the main control room for the instrument location. This spectrum will be automatically compared to the design seismic spectrum. 3.7(B).5 REFERENCES

1. "Seismic Soil-Structure Interaction Effects at Humboldt Bay Power Plant," Journal of the Geotechnical Engineering Division, Vol. 103, No. GT10, October 1977.
2. Lysmer, J., et al., "Efficient Finite Element Analysis of Seismic Structure-Soil-Structure Interaction," Earthquake Engineering Research Center, University of California, Berkeley, Cal., Report No. EERC 75-34, November 1975.
3. Seismic Analyses of Structures and Equipment for Nuclear Power Plants, BC-TOP-4-A, Revision 3, Bechtel Power Corporation, San Francisco, California, November 1974.
4. Seismic Analysis of Piping Systems, BP-TOP-1, Revision 3, Bechtel Power Corporation, San Francisco, California, January 1976.
5. "Nuclear Reactors and Earthquakes", TID-7024, U.S. Atomic Energy Commission, Division of Technical Information, August 1963.

3.7(B)-19 Rev. 25

WOLF CREEK TABLE 3.7(B)-1 DAMPING VALUES FOR SEISMIC CATEGORY I STRUCTURES, SYSTEMS, AND COMPONENTS (Percent of Critical Damping) Operating Basis1 Safe Shutdown Structure or Component Earthquake Earthquake Equipment and large-diameter piping systems2, pipe diameter greater than 12 in.3 2 3 Small-diameter piping systems, diameter equal to or less than 12 in.3 1 2 Welded steel structures 2 4 Bolted steel structures 4 7 Prestressed concrete structures 2 5 Reinforced concrete structures 4 7

1. In the dynamic analysis of active components, as defined in Regulatory Guide 1.48, these values should also be used for the SSE.
2. Includes both material and structural damping. If the piping system consists of only one or two spans with little structural damping, then use the values for small diameter piping.
3. The damping values provided in ASME Code Case N-411 may be utilized for piping systems as an alternative to those identifed above subject to the conditions listed in Regulatory Guide 1.84.
4. The damping values discussed in section 4153.8 of ASME NOG-1-2004 may be utilzied for containment polar crane analysis as an alternative.

Rev. 23

WOLF CREEK TABLE 3.7(B)-2 DEPTH OF SOIL DEPOSITED OVER BEDROCK MAJOR SEISMIC CATEGORY I STRUCTURES Average Elev. of Elev. of Depth of Soil Bottom of Top of Over Rock Structure Base Mat Rock (feet) Reactor building 1088-6" 1065-0" 23.5 Control building 1068-0" 1065-0" 3.0 Fuel building 1093-6" 1063-0" 30.5 Auxiliary building 1068-0" 1065-0" 3.0 Diesel Generators building 1089-6" 1065-0" 24.5 Rev. 0

WOLF CREEK TABLE 3.7(B)-3 FOUNDATION DEPTH BELOW GRADE, MINIMUM BASE DIMENSION AND METHOD OF ANALYSIS FOR SEISMIC CATEGORY I STRUCTURES Ratio of Foundation Minimum Embedment Embedment Base Depth to Method Depth Below Dimension Minimum Base of Structure Grade (feet) (feet) Dimension Analysis (1) Reactor building 11 154 0.071 a Control and aux-iliary building 31.5 222 0.142 a Fuel building 6 91 0.066 a Diesel generators building 10 66.3 0.151 a Refueling water storage tank and 4.5 42.7 0.105 e foundation RWST valve house 13 13.1 0.992 b Emergency fuel oil storage tanks (EFOST) - - - d Vaults for EFOST 6 13.7 0.438 c ESW Vertical Loop Chase 29.5 16.33 1.806 b (1) Method of analysis a Finite-element method, FLUSH computer program b Response spectrum modal analysis technique c Single lumped mass-spring method - vaults are buried below grade with top at grade. d Finite element method in conjunction with the techniques for buried structures outlined in Section 6.0 of Reference 3. e Combination BSAP computer program and conventional hand techniques performed per the requirements of reference 5. Rev. 29

WOLF CREEK TABLE 3.7(B)-4

SUMMARY

FIRST MODE NATURAL FREQUENCIES (Hertz) SSE OBE Building N-S E-W Vert. N-S E-W Vert. Containment 4.4 4.4 13.0 4.4 4.4 13.0 Aux./Control 9.0 9.0 3.6 9.0 8.0 6.8 Fuel 7.0 5.0 9.5 6.7 5.0 9.5 Rev. 0

WOLF CREEK TABLE 3.7 (B)- SA RESPONSE ACCELERATIONS (G's) CONTAINMENT BUILDING SSE Rev. 0 NORTH*SOUTH DIRECTION REF FIGURE 3.7(B)- 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 0.423 0.727 0.707 4~ 0.727 I 2206'-6" 2170'*9" 0.376 0.650 0.626 ~ ~ 0.650 2135'-0" 0.322 0.558 0.528 4~ 0.558 2119'-0" 0.298 0.516 0.484 4~ 0.516 2100'-0" 2080'*0" 0.270 0.242 0.466 0.425 0.429 0.376 

                                                         ...14 u:: 4*
t:
  • 0.466 0.425 2056'*6" 0.208 0.373 0.338 Cl) 4* 0.373 2051'*2" 0.201 0.361 0.329 4. 0.361 2039'-0" 0.185 0.335 0.310 4. 0.335 2028'-0" 0.169 0.313 0.293 4. 0.313 2013'-5" 0.157 0.287 0.270 <:)!6 ~ 0.287 2000'*0" 0.160 0.267 0.250 ' .____.

0.267 2090'*4" 0.202 0.324 0.283 4. 0.324 2060'-0" 0.175 0.301 0.270 4. 0.301 2047'-6" 0.169 0.294 0.265 4-~z 0.294 2034'*0" 0.234 0.286 0.261 4* ffi 0.286 2022'-6" 0.155 0.279 0.258 4 t~ 0.279 2012'-0" 0.155 0.274 0.254 4 ,.1'!,0 -1 J 0.274 2000'-0" 0.160 0.267 0.250 0.267

WOLF CREEK TABLE 3.7 (B)- 56 RESPONSE ACCELERATIONS (G's) CONTAINMENT BUILDING SSE Rev. 0 EAST-WEST DIRECTION _ _ _Rc_:_:E=F. FIGURE 3.7(B)- 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2206'*6" 0.403 0.891 0.814 0.891 2170'*9" 0.357 0.781 0.706 0.781 2135'*0" 0.310 0.651 0.576 0.651 2119'*0" 0.291 0.593 0.521 ------0.593 2100'*0" 0.265 0.528 0.452 0.528 2080'-0" 0.238 0.457 0.394 0.457 2056'-6" 0.205 0.374 0.355 0.374 2051'*2" 0.198 0.362 0.347 0.362 2039'*0" 0.182 0.335 0.327 - - - - - - - 0.335 2028'*0" 0.176 0.311 0.304 0.311 2013'-5" 0.175 0.283 0.281 (!l!6 0.283 2000'-0" 0.173 0.263 0.256 ' 0.263

~------+-------1------+-----------lf----------------------

2090'-4" 0.364 0.432 0.373 0.432 2060'-0" 0.248 0.312 0.292 0.312 2047'-6" 0.227 0.301 0.283 0.301 2034'-0" 0.204 0.288 0.276 ------0.288 2022'-6" 0.186 0.277 0.271 0.277 2012'-0" 0.175 0.270 0.264 0.270 2000'-0" 0.173 0.263 0.256 0.263 _ _ _ _ _ _ _ _ _ _i __ _ _ _ _ _L __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~

WOLF CREEK TABLE 3.7 (B}- SC RESPONSE ACCELERATIONS (G's) CONTAINMENT BUILDING SSE Rev. 0 VERTICAL DIRECTION REF. FIGURE 3.7(B)- 17 MASS SITE POINT ENVELOPE CALLAWAY STERLING WOLFCREEK EL. I 2206'-6" 0.330 0.466 0.388 0.466 

                                                            ~~

2170'-9" 0.323 0.456 0.378 ~ t 0.456 2135'-0" 0.303 0.428 0.349 ~~ 0.428 2119'-0" 0.293 0.414 0.335 4~ 0.414 2100'-0" 0.278 0.392 0.314 _.4 t 0.392 

                                                         ~- ~

2080'-0" 0.266 0.365 0.289 0.365 l: 2056'-6" 0.264 0.328 0.282 C/)4 ~* 0.328 2051'-2" 0.264 0.319 0.280 ~ 0.319 2039'-0" 0.262 0.302 0.275 4~ 0.302 2028'-0" 0.261 0.294 0.271 It 0.294 2013'-5" 0.259 0.282 0.265 <:)!6 )t 0.282 2000'-0" 0.257 0.273 0.259 '*"" 0.273 2090'-4" 0.262 0.280 0.267 4t 0.280 2060'-0" 0.261 0.279 0.265 4t 0.279 

                                                -I 2047'-6"   0.261    0.278        0.264         ~<t                                         0.278 0.277        0.263
                                            *~ffiz                                        0.277 2034'-0" 2022'-6" 0.260 0.259    0.276        0.262
                                            *4~~                                          0.276 1!£:)

2012'-0" 0.258 0.275 0.261 

                                             -~                                           0.275 2000'-0"   0.257    0.273        0.259             '..- ;

0.273

WOLF CREEK.. TABLE 3.7 (B)- 50 RESPONSE INERTIA FORCES (KIPS) CONTAINMENT BUILDING Rev. 0 SSE NORTH-SOUTH DIRECTION REF FIGURE 3 7(B) - 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2206'-6" 1590 2700 2630 4. .. .. 2700 2170'-9" 2830 4840 4660 4* . . 4840 2135' -0" 1930 3330 3150 4. .. .. 3330 2119'-0" 1780 3050 2860 4~ . .. 3050 2100'-0" 1490 2530 2370 ..J 4t . . 2530 2080'-0" 1470 2480 2260 ~4 ~ .. ... 2480 2056'-6" 830 1360 1220 J: Cl) 4* .. . 1360 2051'-2" 480 800 700 4* . ... 800 2039'-0" 580 940* 830 4~ .. .. 940

4. .. ...

2028'-0" 590 960 800 960 2013'-5" 590 1180 770 (!).!6 

                                                                *              ..          1180 2000'-0"      -       -            -                  '

2090'-4" 270 440 390 4~ .. ... 440 2060'-0" 270 460 420 4 

                                                  *                    ..      ..           460 4~~                     ..      ...

2047'-6" 1290 2300 1930 2300 z 2034'-0" 530 1000 850 4*ffi .. 1000 2022'-6" 630 1200 1160 4-~ . ... 1200 2012'-0" 1190 2100 1640 4  ; 1'!.£:) ... .. 2100 2000'-0" - - -

WOLF CREEK TABLE 3.7 (B)- 5E RESPONSE INERTIA FORCES (KIPS) CONTAINMENT BUILDING SSE Rev. 0 EAST*WEST DIRECTION REF. FIGURE 3.7(B)- 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2206'-6" 1520 3310 3040 3310 4~ 2170'-9" 2630 5820 5270 4~ ..

  • 5820 2135'-0" 1860 3880 3460 4~ ...
  • 3880 2119'-0" 1720 3500 3080 4~ ..
  • 3500 2100'-0" 1450 2850 2480 ..J4 ~ .
  • 2850 2080'-0" 1450 2730 2330
                                                         ~-4 ~~
  • 2730
t 2056'-6" 840 1450 1210 Cl) ...
  • 1450 2051'-2" 490 880 710 4. . 880 2039'-0" 590 1020 780 4~ .. 1020 2028'-0" 610 960 770 4. ..
  • 960 2013'-5" 620 940 700 e!6
                                                              *      ..
  • 940 2000'-0" - - - '

4 ...

  • 470 2090'-4" 2060'-0" 410 380 470 480 430 450 4~ ..J ..
  • 480 2047'-6" 1590 2430 1910 4~<z .
  • 2430 2034'-0" 720 1020 750 4~ ffi . ... 1020 4 t~ .. .. 1440 2022'-6" 870 1210 1440 2012'-0" 1620 1930 1470 4 ,1'!.£:)
  • 1930 2000'-0" - - -

WOLF CREEK TABLE 3.7 (B)- SF RESPONSE INERTIA FORCES (KIPS) CONTAINMENT BUILDING SSE Rev. 0 VERTICAL DIRECTION REF. FIGURE 3.7(B)- 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2206'-6" 1220 1760 1400 

                                                               **  ~                       1760 2170'-9"     2380    3440          2740                         4
  • t 3440 2135'-0" 1790 2590 2020 4
                                                                 * ~                       2590 2119'-0"     1710    2480          1920                         4
  • t ~ 2480 2100'-0" 1500 2170 1650 ..J4 2170
                                                          ~4
  • 4*

2080'-0" 1560 2250 1700 2250

1:

2056'-6 930 1350 980

  • t 1350 2051 '-2" 550 790 580 Cl) 4.
                                                                        *t                  790 2039'-0"      680     990           700                         4. t                        990 2028'-0"      710    1010           710
                                              <!).!6           **  ~
                                                                         ~                 1010 2013'-5" 2000'-0" 740 1040 710
                                   -                '
  • 1040 2090'-4" 360 390 370 4. ~ 370 2060'-0" 400 420 410 4~ t 420 2047'-6" 2034'-0" 1250 900 1340 950 1280 910
                                                -~

4 z 4* ffi t 1340 950 2022'-6" 2012'-0" 1630 1690 1740 1800 1650 1720 4 

                                               --~      ,1!£:)
  • t 1740 1800 2000'-0" - - -

WOLF CREEK TABLE 3.7 (B)- 5G RESPONSE SHEAR FORCES (KIPS) CONTAINMENT BUILDING Rev. :J SSE NORTH*SOUTH DIRECTION REF. FIGURE 3.7(B)- 17 MASS SITE POINT ENVELOPE CALLAWAY STERLING WOLFCREEK EL. 2206'-6" ~. 2700 1590 2700 2630 2170'-9" ~~ 7540 2135'-0" 4420 7540 10,870 7290 10,440

4. 10,870 6350 2119'-0" 4~ 13,920 13,300 2100'-0" 8130 13,920

_,4 ~ 16,450 15,670 2080'-0" 9620 16,450 

                                                            ~~
                                                            ~
                                                                 ~                        18,930 11,090   18,930        17,930                     en~
                                                                 ~                        20,290 2056'-6" 11,920   20,290        19,150                        4.

2051'-2" 12,400 21,090 19,850 4. t 21,090 

                                                                          \

2039'-0" 22,030 12,980 22,030 20,680 ~~ 2028'-0" 22,990 13,570 22,990 21,480 (!)!6 ..... 24,170 2013'-5" 2000'-0" 14,160 24,170 22,250 2090'-4" 270 440 390 ~* 4. 440 2060'-0" 900 2047'-6" 540 900 810 4.~z 3200 2034'-0" 1830 3200 2740 4* ffi 4200 2022'-6" 2360 2990 4200 5400 3590 4750 4 

                                                  -~

4 .__ 1"!.£:) 5400 2012'-0" 7500 2000'-0" 4180 7500 6390 '...-"

WOLF CREEK TABLE 3.7 (B)- SH RESPONSE SHEAR FORCES (KIPS) CONTAINMENT BUILDING SSE Rev. 0 EAST.WEST DIRECTION REF FIGURE 3.7(B)- 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2206'*6" 4t 3310 1520 3310 3040 2170'-9" 4~ 9130 4150 9130 8310 2135'-0" 4~ 13,010 6010 13,010 11,770 4~ 16,510 2119'-0" 14,850 . 1 7730 16,510 2100'-0" -J 4~ 19,360 2080'-0" 9180 19,360 17,330 u:! J: 4~ 22,090 10,630 22,090 19,660 C/)4 ~ 2056'-6 23,540 11,470 23,540 20,870 j ~ 2051'-2" 24,420 11,960 24,420 21,580 I ~ 2039'-0" 25,440 12,550 25,440 22,360 4~ 2028'-0" 26,400 13,160 26,400 23,130 (!)!6 )t 2013'-5" 13,780 27,340 23,830 '*;!' 27,340 2000'-0 2090'-4" 4. 470 410 470 430 4~ 

                                                   .<z 2060'-0                                                                                      950
                                                     -J 790      950          880         4 2047'-6"                                                                                     3380 2790 2380     3380                      4 ~ ffi                                    4400 2034'-0" 2022'-6" 3100     4400         3540         4~~                                        5610 3970     5610        4980          4.__       1 '!.8 7540 2012'-0" 2000'-0" 5590     7540        6450                '
                                                         * ;I'

WOLF CREEK TABLE 3.7 (B)- 51 RESPONSE AXIAL FORCES (KIPS) CONTAINMENT BUILDING SSE Rev. 0 VERTICAL DIRECTION REF. FIGURE 3.7(B)- 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2206'-6" 4. 1760 1220 1760 1400 2170'-9" 4~ 5200 3600 5200 4140 2135'-0" 4. 7790 5390 7790 6160 2119'-0" 4. 10,270 7100 10,270 8080 2100'-0" ..J 4. 12,440 2080'-0" 8600 12,440 9730 ~4.

z::: 14,690 10,160 14,690 11,430 Cl) 4~

2056'-6" 16,040 2051 '-2" 11,090 16,040 12,410 4~ 16,830 11,640 16,830 12,990 4~ 

                                                                                 \

2039'-0" 17,820 2028'-0" 12,320 17,820 13,690 4~ 18,830 13,030 18,830 14,400 (!)!6 J~ 2013'-5" 2000'-0" 13,770 19,870 15,110 '.,.-.""' 19,870 2090'-4" 4. 390 360 390 370 4~ ..J 2060'-0" 810 2047'-6" 760 810 780 4~ <( z 2150 2034'-0" 2010 2150 2060 4~ ffi 3100 2022'-6" 2910 3100 2970 4 

                                                   -~                                           4840 4540     4840        4620 4t--.'       ,1!(:)

2012'-0" 2000'-0" 6230 6640 6340

  • 6640

WOLF CREEK TABLE 3.7 (B)- 5J RESPONSE BENDING MOMENTS (MILLIONS OF KIP*FEET) CONTAINMENT BUILDING SSE Rev. 0 NORTH*SOUTH DIRECTION REF. FIGURE 3.7(B)- 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2206'-6" 0/0.0084 0/0.0159 0/0.0107 0/0.0159 2170'-9" 0.0622/0.0802 0.1051/0.1389 0.1030/0.1299 0.1051/0.1389 2135'-0" 0.2362/0.2534 0.4000/0.4267 0.3906/0.4189 0.4000/0.4267 2119'-0" 0.3552/0.3712 0.6005/0.6252 0.5860/0.6121 0.6005/0.6252 2100'-0" 0.5259/0.5414 0.8896/0.9132 0.8651/0.8898 0.8896/0.9132 2080'-0" 0. 7336/0.7503 1.2421/1.2644 1.2032/1.2289 1.2421/1.2644 2056'-6" 1.0108/1.0210 1.7111/1.7252 1.6503/1.6652 1.7111/1.7252 2051 '-2" 1.0845/1.0903 1.8334/1.8416 1. 7673/1.7760 1.8334/1.8416 2039'-0" 1.2413/1.2488 2.0972/2.1070 2.0168/2.0286 2.0972/2.1070 2028'-0" 1.3916/1.3992 2.3500/2.3598 2.2560/2.2658 2.3500/2.3598 2013'-5" 1.5972/1.6052 2.6950/2.7028 2.5774/2.5872 (!)!6 2.6950/2.7028 2000'-0" 1.7954 3.0223 2.8851 '*" 3.0223 2090'-4" 010.0020 0/0.0031 0/0.0027 0/0.0031 2060'-0" 0.0043/0.0048 0.0069/0.0072 0.0061/0.0063 0.0069/0.0072 2047'-6" 0.0113/0.0137 0.0185/0.0200 0.0164/0.0176 0.0185/0.0200 2034'-0" 0.0357/0.0382 0.0622/0.0636 0.0534/0.54 70 0.0622/0.0636 2022'-6" 0.0635/0.0678 0.1117/0.1151 0.0958/0.0989 0.1117/0.1151 2012'-0" 0.0980/0.1018 0.1718/0.1737 0.1483/0.1502 ,1!,0 0.1718/0.1737 2000'-0" 0.1482 0.2636 0.2268 

                                                                  '*                             0.2636

WOLF CREEK TABLE 3.7 (B)- 5K RESPONSE BENDING MOMENTS (MILLIONS OF KIP-FEET) CONTAINMENT BUILDING SSE Rev. 0 EAST-WEST DIRECTION REF. FIGURE 3.7(B) - 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2206'-6" 0/0.0103 0/0.0147 0/0.0118 0/0.0147 2170'-9" 0.0614/0.0821 0.1304/0.1663 0.1206/0.1559 0.1304/0.1663 2135'-0" 0.2293/0.2513 0.4925/0.5302 0.4532/0.4898 0.4925/0.5302 2119'-0" 0.3438/0.3642 0.7381/0.7730 0.6782/0.7121 0.7381/0.7730 2100'-0" 0.5018/0.5212 1.0868/1.1196 0.9943/1.0261 1.0868/1.1196 2080'-0" 0.6887/0.7089 1.5068/1.5410 1.3728/1.4055 1.5068/1.5410 2056'-6 0.9526/0.9614 2.0600/2.0796 1.8677/1.8865 2.0600/2.0796 2051 '-2" 1.0225/1.0276 2.2050/2.2168 1.9972/2.0090 2.2050/2.2168 2039'-0" 1.1731/1.1795 2.5127/2.5264 2.2716/2.2834 2028'-0" 1.3175/1.3242 2.8048/2.8185 2.5304/2.5421 2.8048/2.8185 2013'-5" 1.5161/1.5231 3.2026/3.2144 2.8792/2.8910 8!6 3.2026/3.2144 2000'-0" 1.7079 3.5790 3.2124 '*"' 3.5790 2090'-4" 0/0.0033 0/0.0039 0/0.0033 0/0.0039 2060'-0" 0.0069/0.0080 0.0079/0.0095 0.0068/0.0087 0.0079/0.0095 2047'-6" 0.0179/0.0191 0.0203/0.0223 0.0184/0.0213 0.0203/0.0223 2034'-0" 0.0512/0.0518 0.0665/0.0677 0.0545/0.0556 0.0665/0.0677 2022'-6" 0.0875/0.0891 0.1182/0.1216 0.0964/0.0988 0.1182/0.1216 2012'-0" 0.1308/0.1308 0.1804/0.1819 0.1503/0.1517 ,11£:) 0.1804/0.1819 2000'-0" 0.1978 0.2722 0.2285 '...- 0.2722

WOLF CREEK TABLE 3.7 (B)- SL RESPONSE DISPLACEMENTS (INCHES) CONTAINMENT BUILDING Rev. 0 SSE NORTH-SOUTH DIRECTION REF. FIGURE 3.7(B)- 17 MASS SITE POINT ENVELOPE CALLAWAY STERLING WOLFCREEK EL. 2206'*6" 0.423 0.371 0.370 0.423 2170'*9" 0.382 0.322 0.321 0.382 2135'-0" 0.335 0.262 0.262 0.335 2119'-0" 0.314 0.235 0.236 0.314 2100'-0" 0.288 0.201 0.203 0.288 2080'-0" 0.260 0.165 0.167 0.260 2056'-6" 0.227 0.122 0.125 0.227 2051 '-2" 0.220 0.112 0.115 0.220 2039'-0" 0.203 0.090 0.094 0.203 2028'-0" 0.188 0.071 0.076 0.188 2013'-5" 0.169 0.046 0.052 (!)!6 0.169 2000'-0" 0.152 0.023 0.031 ' 0.152 2090'-4" 0.225 0.072 0.080 0.225 2060'-0" 0.204 0.055 0.063 0.204 2047'-6" 0.196 0.049 0.057 0.196 2034'-0" 0.181 0.042 0.050 0.181 2022'-6" 0.167 0.036 0.044 0.167 2012'-0" 0.160 0.030 0.038 ,11<::) 0.160 2000'-0" 0.152 0.023 0.031 0.152

WOLF CREEK TABLE 3.7 (B)- SM RESPONSE DISPLACEMENTS (INCHES) CONTAINMENT BUILDING SSE Rev. 0 EAST*WEST DIRECTION REF. FIGURE 3.7(B!_- 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 

            - - -l------      +------------1---~--------~---------------1 2206'*6"    0.440     0.471          0.439                                           0.471 2170'*9"    0.397     0.406          0.379                                           0.406 2135'*0"    0.349     0.328          0.308                                           0.349 0.327     0.293          0.277                          _ _ _ _ _ _ 0.327 2119'*0" 2100'*0"    0.301     0.250          0.238                                           0.301 2080'*0"    0.273     0.203          0.196                                           0.273 2056'*6"    0.240     0.148          0.146                                           0.240 2051 '*2"   0.233     0.136          0.135                                           0.233 2039'-0"    0.216     0.108          0.110                         - - - - - - - 0.216 2028'*0"    0.201     0.083          0.088                                           0.201 2013'*5"    0.181     0.052          0.061         e:!,6                             0.181 2000'*0"    0.164     0.025          0.036             '                             0.164 2090'-4"    0.234     0.098          0.104                                           0.234 2060'-0"    0.209     0.073          0.081                                           0.209 2047'-6"    0.200     0.064          0.072                                           0.200 2034'-0"    0.190     0.053          0.062                          ------0.190 2022'*6"    0.181     0.043          0.053                                           0.181 2012'-0"    0.173     0.035          0.045                                           0.173 2000'-0"    0.164     0.025          0.036                                           0.164

WOLF CREEK TABLE 3.7 (B)- SN RESPONSE DISPLACEMENTS (INCHES) CONTAINMENT BUILDING SSE Rev. 0 VERTICAL DIRECTION REF. FIGURE 3.7(B)- 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2206'-6" 0.0207 0.0260 0.0213 0.0260 I 2170'-9" 0.0202 0.0253 0.0206 0.0253 2135'-0" 0.0186 0.0230 0.0186 0.0230 2119'-0" 0.0178 0.0219 0.0176 0.0218 2100'-0" 0.0166 0.0201 0.0160 0.0201 2080'-0" 0.0149 0.0178 0.0140 0.0178 2056'-6" 0.0126 0.0147 0.0112 0.0147 2051'-2" 0.0120 0.0139 0.0105 0.0139 2039'-0" 0.0106 0.0119 0.0089 0.0119 2028'-0" 0.0092 0.0101 0.0073 0.0101 2013'-5" 0.0072 0.0075 0.0052 (!l!6 . 0.0075 2000'-0" 0.0053 0.0049 0.0030 ' - 0.0053 2090'-4" 0.0075 0.0073 0.0053 0.0075 2060'-0" 0.0071 0.0068 0.0049 0.0071 2047'-6" 0.0069 0.0066 0.0047 0.0069 2034'-0" 0.0067 0.0063 0.0044 0.0067 2022'-6" 0.0063 0.0060 0.0041 0.0063 2012'-0" 0.0059 0.0056 0.0037 1!(:) 0.0059 2000'-0" 0.0053 0.0049 0.0030 0.0053

WOLF CREEK TABLE 3.7 (B)- 50 RESPONSE ACCELERATIONS (G's) CONTAINMENT BUILDING OBE Rev. 0 NORTH-SOUTH DIRECTION REF. FIGURE 3.7(B) - 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2206'-6" 0.232 0.509 0.411 0.509 2170'-9" 0.211 0.439 0.365 0.439 2135'-0" 0.187 0.356 0.310 0.356 2119'-0" 0.176 0.321 0.285 0.321 2100'-0" 0.163 0.284 0.253 0.284 2080'-0" 0.148 0.253 0.220 0.253 2056'-6" 0.129 0.216 0.181 0.216 2051 '-2" 0.125 0.207 0.174 0.207 2039'-0" 0.116 0.186 0.162 0.186 2028'-0" 0.110 0.171 0.154 0.171 2013'-5" 0.107 0.152 0.142 e!6 0.152 2000'-0" 0.103 0.139 0.132 ' 0.139 2090'-4" 0.125 0.177 0.146 0.177 2060'-0" 0.113 0.157 0.141 0.157 2047'-6" 0.112 0.154 0.139 0.154 2034'-0" 0.108 0.149 0.137 0.149 2022'-6" 0.104 0.145 0.135 0.145 2012' -0" 0.104 0.142 0.133 .,1!(:) 0.142 2000'-0" 0.103 0.139 0.132 0.139

WOLF CREEK TABLE 3.7 (B)- 5P RESPONSE ACCELERATIONS (G's) CONTAINMENT BUILDING Rev. 0 OBE EAST-WEST DIRECTION REF. FIGURE 3.7(B) - 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2206'-6" 0.240 0.524 0.546 0.546 2170'-9" 0.217 0.462 0.474 0.474 2135'*0" 0.187 0.394 0.389 0.394 2119'-0" 0.178 0.363 0.351 0.363 2100'*0" 0.164 0.324 0.305 0.324 2080'-0" 0.148 0.281 0.258 0.281 2056'-6" 0.128 0.227 0.204 0.227 2051 '*2" 0.124 0.215 0.192 0.215 2039'-0" 0.113 0.187 0.166 0.187 2028'-0" 0.106 0.163 0.156 0.163 2013'-5" 0.101 0.142 0.143 <:)!6 0.143 2000'-0" 0.098 0.134 0.131 ' 0.134 2090'-4" 0.177 0.251 0.223 0.251 2060'-0" 0.131 0.159 0.151 0.159 2047'-6" 0.122 0.154 0.147 0.154 2034'-0" 0.111 0.146 0.144 0.146 2022'-6" 0.102 0.139 0.141 0.141 2012'-0" 0.100 0.135 0.136 1!(:) 0.136 2000'-0" 0.098 0.134 0.131 ""' 0.134

WOLF CREEK TABLE 3.7 (B)- 5Q RESPONSE ACCELERATIONS (G's) CONTAINMENT BUILDING OBE Rev. 0 VERTICAL DIRECTION REF. FIGURE 3.7(B) - 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2206'-6" 0.219 0.269 0.223 0.269 2170'-9" 0.214 0.264 0.218 ,.I 0.264 2135'-0" 0.200 0.248 0.202 0.248 2119'-0" 0.193 0.240 0.194 0.240 2100'-0" 0.182 0.228 0.182 0.228 2080'-0" 0.169 0.213 0.171 0.213 2056'-6" 0.151 0.192 0.157 0.192 2051'-2" 0.146 0.187 0.156 0.187 2039'-0" I 0.140 0.175 0.152 0.175 2028'-0" 0.139 0.164 0.148 0.164 2013'-5" 0.137 0.150 0.143 (!)!6 0.150 2000'-0" 0.136 0.144 0.139 ' - 0.144 2090'-4" 0.139 0.150 0.144 0.150 2060'-0" 0.139 0.149 0.143 0.149 2047'-6" 0.138 0.149 0.142 0.149 2034'-0" 0.138 0.148 0.142 0.148 2022'-6" 0.137 0.147 0.141 0.147 2012'-0" 0.137 0.145 0.140 ,1!£:) 0.145 2000'-0" 0.136 0.144 0.139 0.144

WOLF CREEK TABLE 3.7 (B)- 5R RESPONSE INERTIA FORCES (KIPS) CONTAINMENT BUILDING OBE Rev. 0 NORTH-SOUTH DIRECTION . REF . FIGURE 3 7(B)- 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 

                                                                                ..                  1920 2206'-6"      850    1920           1510                               4.
  • 2170'-9" 1530 3320 2700 4. .. 3320 2135'-0" 1110 2140 1830 4 .. .... 2140 2119'-0" 1030 1910 1680 4 *
                                                                        *      .
  • 1910 2100'-0" 890 1510 1390 ....14 ..
  • 1510 2080'-0" 910 1380 1350 ~4
                                                                        *      .
  • 1380
                                                                        ~

J: (1)4 2056'-6 540 700 740 ~ 740 2051 '-2" 340 390 430 4~

  • 430 2039'-0" 380 440 500 4~ ..
  • 500 4.

2028'-0" 410 380 500 500 

                                                 <:)!6                                               570 2013'-5" 2000'-0" 440 570 470 2090'-4"      160     220            200          4.                            .
  • 220
4. ..

2060'-0" 170 240 220

  • 240
                                                     ...I 2047'-6"      890    1140           1050          4. <{z                                ....         1140 2034'-0"      330      390           520          4 ~ffi                        .       ....          520 4 ~~                          ..      ....

2022'-6" 390 690 580 690 2012'-0" 850 1100 940 4 1!Q .. 1100 2000'-0" - - -

WOLF CREEK TABLE 3.7 (B)- 55 RESPONSE INERTIA FORCES (KIPS) CONTAINMENT BUILDING OBE Rev. 0 EAST*WEST DIRECTION REF. FIGURE 3.7(B) - 17 MASS SITE POINT ENVELOPE CALLAWAY STERLING WOLFCREEK EL. 

        --~

2020 2206'-6" 890 1940 2020 4t 2170'-9" 1620 3390 3530 4t . .. 3530 2135'-0" 1140 2350 2320 4t .. .. 2350 2140 2070 4. .. .. 2140 2119'-0" 1060 2100'-0" 900 1780 1680 .) ~ .. 1780 1570 ~4 ~ .. . 1730 2080'-0" 910 1730

z::

cn4 ~ . 930 2056'-6" 530 930 810 4~ .. 550 2051'-2" 310 550 470

4. ..

630 2039'-0" 380 630 540 4. 2028'-0" 380 610 520 610 

                                                           <:)!6                      ...                  570 2013'-5"                  400    570          470 2000'-0"                  -       -           -

2090'-4" 210 250 240 4~ .. .. 250 2060'-0" 190 240 230 4~ 240 2047'-6" 1180 1470 1050 4.~z 

                                                                                              ...         1470 4* ffi 2034'-0"                  370    350          360                                                           370 2022'-6"                  350    730          820           4t~                        ..                   820 2012'-0"                  900   3860          860           4         ,1!£:)           ..      ...         3860 2000'-0"                  -      -            -

WOLF CREEK TABLE 3.7 (B)- ST RESPONSE INERTIA FORCES (KIPS) CONTAINMENT BUILDING Rev. 0 OBE VERTICAL DIRECTION REF. FIGURE 3.7(B) - 17 

     -   ~-

MASS SITE POINT f----~ ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2206'*6" 820 1010 830 1010 4. 

                                                                           *t t 2170'-9"              1600    1990         1610                        4.                           1990 2135'-0"              1190    1500         1190                        4.                           1500 2119'-0"              1140    1430         1140                        4         t                  1430 2100'-0"               990    1260          990                    _,4 ~*    t                      1260 2080'-0"              1030    1310         1020                    ~4
z: ~ ~ 1310 2056'-6" 610 780 590 (I) 4*
                                                                             ~                      780 2051'-2"               360     460          350                        4~        ~                   460 2039'-0"               440     580          430                        4~     ~                      580 2028'-0"               450     590          500                        4~        ~                   590 2013'-5"               470     610          530         e!6          .J. t                       610 2000'-0"               -       -            -                '*"

2090'-4" 200 200 200 4. t 200 2060'-0" 210 230 210 

                                                       **~~                                         230 2047'-6" 2034'-0" 670 490 700 510 680 490
                                                       ~

z 4 ~ffi 

                                                                            *~ t                    700 510 2022'-6"               870     910          880         4~~                      t                   910 2012'-0"               910     950          910         4.__      1  '!..0   t                       950 2000'-0"               -       -            -                '*"

WOLF CREEK TABLE 3.7 (B)- SU RESPONSE SHEAR FORCES (KIPS) CONTAINMENT BUILDING OBE Rev. 0 NORTH-SOUTH DIRECTION REF. FIGURE 3.7(B) - 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2206'*6" 1920 850 1920 1510 2170'-9" 5240 2380 5240 4210 2135'-0" 7380 3490 7380 6040 2119'-0" 9290 4520 9290 7720 2100'-0" 10,800 5410 10,800 9110 2080'-0" 12,180 6320 12,180 10,460 2056'-6" 12,880 6860 12,880 11,200 2051'-2" 13,270 7200 13,270 11,630 2039'-0" 13,710 7580 13,710 12,130 2028'-0" 14,090 2013'-5" 7990 14,090 12,630 (!)!6 14,660 2000'-0" 8430 14,660 13,100 2090'-4" 220 160 220 200 2060'-0" 460 330 460 420 2047'-6" 1600 1220 1600 1470 2034'-0" 1990 1550 1990 1990 2022'-6" 2680 2012'-0" 1940 2680 2570 

                                                       ,1!(:)                          3780 2000'-0" 2790     3780         3510

WOLF CREEK TABLE 3.7 (B)- 5V RESPONSE SHEAR FORCES (KIPS) CONTAINMENT BUILDING OBE Rev. 0 EAST*WEST DIRECTION REF. FIGURE 3.7(B)- 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2206'-6" 2020 890 1940 2020 2170'*9" 5550 2510 5330 5550 2135'-0" 7870 3650 7680 7870 2119'-0" 9940 4710 9820 9940 2100'-0" 11,620 5610 11,600 11,620 2080'-0" 13,330 6520 13,330 13,190 2056'-6" 14,260 7050 14,260 14,000 2051 '-2" - - - - - 14,810 7360 14,810 14,470 2039'-0" 15,440 7740 15,440 15,010 2028'-0" 16,050 8120 16,050 15,530 2013'-5" 16,620 8520 16,620 16,000 2000'-0" 2090'-4" 250 210 250 240 2060'-0" 490 400 490 470 2047'-6" 1960 1580 1960 1520 2034'-0" ------2310 1950 2310 1880 2022'-6" 3040 2300 3040 2700 2012'-0" 6900 3200 6900 3560 2000'-0"

WOLF CREEK TABLE 3.7 (B)- SW RESPONSE AXIAL FORCES (KIPS) CONTAINMENT BUILDING Rev. 0 OBE VERTICAL DIRECTIO:N~----------'-R._..E._._F_:_.:_:FI=-=G=-=U~R:.::Ec._:3:_:_:.7___,_(B--')_-__1_7 

                     ----~-------------,

MASS SITE POINT ENVELOPE CALLAWAY STERLING WOLFCREEK EL. 12206'-6" 1010 I 820 1010 830 2170'-9" 3000 2420 3000 2440 2135'-0" 4500 3610 4500 3630 2119'-0" 5930 4750 5930 4770 2100'-0" 7190 5740 7190 5760 2080'-0" 8500 6770 8500 6780 2056'-6 9280 7380 9280 7370 2051 '-2" 9740 7740 9740 7720 2039'-0" 10,320 8180 10,320 8150 2028'-0" 10,910 8630 10,910 8650 2013'-5 11,520 9100 11,520 9180 2000'-0" 2090'-4" 200 200 200 200 2060'-0" 430 410 430 410 2047'-6" 1130 1080 1130 1090 2034'-0" - - - - - - - 2550 1570 1640 1580 2022'-6" 3500 2440 2550 2460 2012'-0" 3350 3500 3370 2000'-0"

WOLF CREEK TABLE 3.7 (B) - SX RESPONSE BENDING MOMENTS (MILLIONS OF KIP*FEET) CONTAINMENT BUILDING OBE Rev. 0 NORTH*SOUTH DIRECTION REF. FIGURE 3.7(B)- 17 MASS SITE ENVELOPE POINT EL. CALLAWAY STERLING WOLFCREEK 2206'-6" 0/0.0039 0/0.0109 0/0.0063 0/0.0109 2170'-9" 0.0323/0.0390 0.0766/0.0999 0.0590/0.0758 0.0766/0.0999 2135' -0" 0.1241/0.1310 0.2873/0.3114 0.2240/0.2395 0.2873/0.3114 2119'-0" 0.1868/0.1932 0.4296/0.4520 0.3361/0.3504 - - - - 0.4296/0.4520 2100'-0" 0.2791/0.2852 0.6284/0.6492 0.4971/0.5106 0.6284/0.6492 2080'-0" 0.3936/0.4000 0.8651/0.8865 0.6929/0.7070 0.8651/0.8865 2056'-6" 0.5486/0.5525 1.1729/1.1850 0.9528/0.9610 1.1729/1.1850 2051 '-2" 0.5892/0.5913 1.2538/1.2609 1.0206/1.0263 1.2538/1.2609 2039'-0" 0.6787/0.6815 1.4224/1.4306 1.1668/1.1725 - - - 1.4224/1.4306 2028'-0" 0. 7650/0.7680 1.5813/1.5894 1.3058/1.3114 1.5813/1.5894 2013'-5" 0.8844/0.8875 1. 7950/1.8020 1.4955/1.5008 ~6 1. 7950/1.8020 2000'-0" 1.0004 1.9953 1.6766 '*"' 1.9953 2090'-4" 0/0.0012 0/0.0017 0/0.0014 0/0.0017 2060'-0" 0.0026/0.0029 0.0035/0.0037 0.0031/0.0033 0.0035/0.0037 2047'-6" 0.0070/0.0080 0.0093/0.0102 0.0084/0.0093 0.0093/0.01 02 2034'-0" 0.0234/0.0243 0.0293/0.0301 0.0286/0.0291 0.0293/0.0301 2022'-6" 0.0416/0.0432 0.0525/0.0540 0.0515/0.0526 0.0525/0.0540 2012'-0" 0.0630/0.0652 0.082110.0830 0.0789/0.0790 11<:) 0.0821/0.0830 2000'-0" 0.0951 0.1261 0.1212 0.1261

WOLF CREEK TABLE 3.7 (B)- SY RESPONSE BENDING MOMENTS (MILLIONS OF KIP*FEET) CONTAINMENT BUILDING OBE Rev. 0 EAST-WEST DIRECTION REF. FIGURE 3.7(B) - 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2206'-6" 0/0.0048 0/0.0092 0/0.0078 0/0.0092 2170'-9" 0.0345/0.0422 0.0766/0.0987 0.0802/0.1034 0.0802/0.1034 2135'-0" 0.1319/0.1402 0.2885/0.3116 0.3016/0.3259 0.3016/0.3259 2119' -0" 0.1985/0.2064 0.4322/0.4535 0.4518/0.47 41 0.4518/0.4 7 41 2100'-0" 0.2958/0.3032 0.6345/0.6546 0.6631/0.6840 0.6631/0.6840 2080'-0" 0.4155/0.4234 0.8834/0.9024 0.9165/0.9383 0.9165/0.9383 2056'-6" 0.5766/0.5815 1.2156/1.2270 1.2481/1.2609 1.2481/1.2609 2051'-2" 0.6192/0.6219 1.3030/1.3095 1.3355/1.3428 1.3355/1.3428 2039'-0" 0.7115/0.7148 1.4898/1.4976 1.5190/1.5276 1.5190/1.5276 2028'-0" 0. 7999/0.8034 1.6676/1.6754 1.6927/1.7011 1.6927/1.7011 (!)!6 2013'-5" 2000'-0" 0.9220/0.9257 1.0402 1.9094/1.9171 2.1403 1.9277/1.9357 2.1501 

                                                                ,..,..;                        1.9277/1.9357 2.1501 2090'-4"         0/0.0016       0/0.0021       0/0.0020                                               0/0.0021 2060'-0"  0.0034/0.0040   0.0043/0.0054  0.0040/0.0053                                          0. 0043/0.0054 2047'-6"  0.0090/0.0100   0.0111/0.0124  0.0107/0.0128                                          0.0111/0.0124 2034'-0"  0.0313/0.0318   0.0362/0.0384  0.0297/0.0307                                          0.0362/0.0384 2022'-6"  0.0542/0.0561   0.0609/0.0641  0.0517/0.0535                                          0.0609/0.0641 2012'-0"  0.0786/0.0791   0.0930/0.0951  0.0789/0.0795                   1!(:)                  0.0930/0.0951 2000'-0"      0.1175          0.1549        0.1220                                                   0.1549

WOLF CREEK TABLE 3.7 (B)- SZ RESPONSE DISPLACEMENTS (INCHES) CONTAINMENT BUILDING OBE Rev. 0 NORTH-SOUTH DIRECTION REF. FIGURE 3.7(B) - 17 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2206'-6" 0.190 0.247 0.213 0.247 2170'-9" 0.170 0.212 0.185 0.212 2135'-0" 0.147 0.170 0.151 0.170 2119'-0" 0.137 0.152 0.136 0.152 2100'-0" 0.124 0.129 0.117 0.129 2080'-0" 0.110 0.104 0.096 0.110 2056'-6" 0.093 0.075 0.071 0.093 2051 '-2" 0.089 0.069 0.065 0.089 2039'-0" 0.080 0.054 0.052 0.080 2028'-0" 0.072 0.041 0.041 0.072 2013'-5" 0.062 0.025 0.027 (!).!6 0.062 2000'-0" 0.053 0.012 0.014 0.053 2090'-4" 0.080 0.042 0.044 0.080 2060'-0" 0.071 0.031 0.033 0.071 2047'-6" 0.067 0.027 0.029 0.067 2034'-0" 0.063 0.023 0.025 0.063 2022'-6" 0.060 0.019 0.021 0.060 2012'-0" 0.057 0.016 0.018 11<:) 0.057 

                                                            ~

2000'-0" 0.053 0.012 0.014 0.053

2090'-4" 0.093 0.057 0.059 0.093 2060'-0" 0.081 0.042 0.045 0.081 2047'-6" 0.077 0.037 0.039 0.077 2034'-0" 0.072 0.031 0.033 0.072 2022'-6 0.068 0.025 0.028 0.068 2012'-0" 0.064 0.020 0.023 ,110 0.064 2000'-0" 0.060 0.015 0.017 0.060

WOLF CREEK TABLE 3.7 (B)- SAB RESPONSE DISPLACEMENTS (INCHES) CONTAINMENT BUILDING Rev. 0 OBE VERTICAL DIRECTION REF. FIGURE 3.7(B)- 17 MASS POINT EL. CALLAWAY SITE STERLING WOLFCREEK 0.0125 ENVELOPE I 0.0163 2206'-6" 0.0115 0.0163 2170'-9" 0.0112 0.0158 0.0120 .,.J 0.0158 2135'-0" 0.0103 0.0144 0.0108 ... 0.0144 2119'-0" 0.0098 0.0136 0.0102 ... 0.0136 2100'-0" 0.0090 0.0125 0.0093 ... 0.0125 2080'-0" 0.0081 0.0110 0.0081 . 0.0110 2056'-6" 0.0067 0.0090 0.0065 0.0090 2051 '-2" 0.0064 0.0085 0.0061 0.0085 2039'-0" 0.0056 0.0073 0.0051 0.0073 2028'-0" 0.0048 0.0062 0.0042 0.0062 2013' -5" 0.0037 0.0045 0.0029 (!).!6 0.0045 2000'-0" 0.0028 0.0030 0.0016 ' 0.0030 2090'-4" 0.0039 0.0052 0.0027 0.0052 2060'-0" 0.0036 0.0043 0.0025 0.0043 2047'-6" 0.0035 0.0041 0.0024 0.0041 2034'-0" 0.0034 0.0038 0.0023 0.0038 2022'*6" 0.0032 0.0036 0.0021 0.0036 2012'-0" 0.0030 0.0033 0.0019 ,1'!J::) 0.0033 2000'-0" 0.0026 0.0030 0.0016 0.0030

WOLF CREEK TABLE 3.7 (B)- 6A RESPONSE ACCELERATIONS (G's) FUEL BUILDING SSE NORTH-SOUTH DIRECTION Rev. 0 _ REF. FIGURE 3.7(B)- 18 M ASS SITE p OINT ENVELO PE EL. CALLAWAY STERLING WOLFCREEK 21 06'-6" 0.606 0.635 0.461 0.635 4~ 20 83'-6" 0.504 0.521 0.362 4~ 0.521 20 47'-6" 0.331 0.330 0.270 4. 0.331 20 26' -0" 0.260 0.287 0.255 4. 0.287 20 00' -0" 0.202 0.271 0.252 4. 0.271 

        --- ---  - ~------------- -

WOLF CREEK TABLE 3.7 (B)- 6B RESPONSE ACCELERATIONS (G's) FUEL BUILDING SSE EAST-WEST DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 21 06'-6" 0.418 0.793 0.586 4~ 0.793 2083'-6" 0.371 0.644 0.529 4~ 0.644 2047'-6" 0.307 0.367 0.426 4~ 0.426 2026'-0" 0.272 0.282 0.375 4~ 0.375 2000'-0" 0.242 0.265 0.318 4~ 0.318

WOLF CREEK TABLE 3.7 (B)- 6C RESPONSE ACCELERATIONS (G's) FUEL BUILDING SSE VERTICAL DIRECTION Rev. 0 REF FIGURE 3 7(B)- 18 MASS SITE POINT ENVELOPE EL. - - CALLAWAY STERLING WOLFCREEK 

  -*~----

1--- ---- 2106'-6" 0.328 0.330 0.389 ~ t 0.389 2083'-6 0.327 0.321 0.382 4~ 0.382 2047'-6" 0.322 0.295 0.362 4~ 0.362 2026'-0" 0.320 0.284 0.353 4~ 0.353 2000'-0" 0.316 0.276 0.337 4~ 0.337

WOLF CREEK TABLE 3.7 (B)- 6D RESPONSE INERTIA FORCES (KIPS) FUEL BUILDING SSE NORTH-SOUTH DIRECTION Rev. 0 

    --*-~-

REF. FIGURE 3.7(B)- 18 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2106'-6" 3100 3150 2340 4.

  • 3150 2083'-6" 2300 2340 1640 2340 4.*

2047'-6" 3300 3400 2150 4 3400 2026'-0" 4100 4030 3910 ..

  • 4100 2000'-0" - - - 4.

WOLF CREEK TABLE 3.7 (B)- 6E RESPONSE INERTIA FORCES (KIPS) FUEL BUILDING SSE EAST*WEST DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2106'-6" 2090 3950 2950 4. . .. 3950 4 2083'-6 1650 2890 2380 2890 4.* . 2047'-6" 3010 3820 4400 4 4400 2026'-0" 4430 4100 6180 .. 6180 2000'-0" - - - 4.

WOLF CREEK TABLE 3.7 (B)- 6F RESPONSE INERTIA FORCES (KIPS) FUEL BUILDING SSE VERTICAL DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2106'-6" 1630 1670 2000 4. t 2000 2083'-6" 1450 1460 1770 4. ~ 1770 2047'-6" 3280 3070 3860 4. t 3860 2026'-0" 5170 4700 5980 4. t 5980 2000'-0" - - - 4.

WOLF CREEK TABLE 3.7 (B)- 6G RESPONSE SHEAR FORCES (KIPS) FUEL BUILDING SSE NORTH-SOUTH DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 MASS SITE POINT -** --~-*------ ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2106'*6" 4~ 3150 3150 2083'-6" 3100 2350 4~ 5490 5400 5490 3980 4 2047'-6" 2026'-0" 8700 8890 6130

  • 4.

8890 12,920 2000'-0" 12,800 12,920 10,040 4.

WOLF CREEK TABLE 3.7 (B)- 6H RESPONSE SHEAR FORCES (KIPS) FUEL BUILDING SSE EAST*WEST DIRECTION Rev. 0 REF FIGURE 3 7(B)- 18 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2106'-6" 4. 3950 2090 3950 2950 4 2083'-6" 3740 6840 5330 4~ 6840 2047'-6" 10,660 2026'-0" 6750 10,660 9730 4. 15,910 2000'-0" 11,180 14,760 15,910 4.

WOLF CREEK TABLE 3.7 (B)- 61 RESPONSE AXIAL FORCES (KIPS) FUEL BUILDING SSE VERTICAL DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 

                                                       ~~

2106'*6" 1630 1670 2000 4. 2000 2083'-6" 2047'-6" 3080 3130 3770 4. 3770 7630 6360 6200 7630 4. 13,610 2026'-0" 11,530 10,900 13,610 4~ 2000'-0"

WOLF CREEK TABLE 3.7 (B)- 6J RESPONSE BENDING MOMENTS (MILLIONS OF KIP-FEET) FUEL BUILDING SSE NORTH-SOUTH DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK I 2106'*6" 0 /0.0269 0/0.0240 0/0.0230 4~ 0/0.0269 2083'-6" 0.0981/0.1317 0.0965/0.1257 0.0767/0.1050 4~ 0.0981/0.1317 2047'-6" 0.3247/0.3776 0.3234/0.3661 0.2483/0.2916 4~ 0.324 7/0.3776 2026'-0" 0.5635/0.6161 0.557 4/0.5972 0.4212/0.4631 4~ 0.5635/0.6161 2000'-0" 0.9383 0.9330 0.6718 

                                                                -~                                0.9383

WOLF CREEK TABLE 3.7 (B)- 6K RESPONSE BENDING MOMENTS (MILLIONS OF KIP-FEET) FUEL BUILDING SSE EAST-WEST DIRECTION Rev. 0 

     ----      ~---  ~-  ----------

REF. FIGURE 3.7(B)- 18 MASS SITE 

               - --                                                                      ENVELOPE POINT               ----

EL. CALLAWAY STERLING WOLFCREEK 2106'-6" 0/0.0149 0/0.0215 0/0.0240

4. 0/0.0240 2083'-6" 0.0576/0.0762 0.1124/0.1436 0.0855/0.1204 4. 0.1124/0.1436 2047'-6" 0.2051/0.2239 0.3896/0.4279 0.2899/0.3285 4. 0.3896/0.4279 2026'-0" 0.3640/0.3828 0.6571/0.6938 0.5099/0.5279 4. 0.6571/0.6938 2000'-0" 0.6601 1.0776 0.9415 4. 1.0776

WOLF CREEK TABLE 3.7 (B)- 6L RESPONSE DISPLACEMENTS (INCHES) FUEL BUILDING SSE NORTH*SOUTH DIRECTION Rev. 0 REF. FIGURE 3 7(B)- 18 MASS SITE 

                                   ----                                             ENVELOPE POINT            --------

EL. CALLAWAY STERLING WOLFCREEK 

     ----- --- -  **-   --~----

2106'-6" 0.311 0.124 0.115 4. 0.311 2083'-6" 0.275 0.101 0.092 4. 0.275 2047'-6" 0.216 0.060 0.056 4. 0.216 2026'-0" 0.186 0.041 0.039 4 0.186 2000'-0" 0.154 0.021 0.022 4~ 0.154

WOLF CREEK TABLE 3.7 (B)- 6M RESPONSE DISPLACEMENTS (INCHES) FUEL BUILDING SSE EAST*WEST DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2106'-6" 0.473 0.212 0.293 4~ 0.473 2083'*6" 0.393 0.172 0.245 4~ 0.393 2047'-6" 0.260 0.097 0.160 4. 0.260 I 2026'-0" 0.184 0.065 0.117 4. 0.184 2000'-0" 0.095 0.031 0.067 4~ 0.095

WOLF CREEK TABLE 3.7 (B)- 6N RESPONSE DISPLACEMENTS (INCHES) FUEL BUILDING SSE VERTICAL DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK l 2106'-6" 0.050 0.017 0.045 ~ t 0.050 2083'-6" 0.049 0.016 0.044 ~. 0.049 2047'-6" 0.047 0.013 0.041 ~~ 0.047 2026'-0" 0.046 0.012 0.039 4. + 0.046 2000'-0" 0.044 0.010 0.036 4t u 0.044

WOLF CREEK TABLE 3.7 (B)- 60 RESPONSE ACCELERATIONS (G's) FUEL BUILDING OBE NORTH-SOUTH DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2106'-6" 0.313 0.363 0.269 4. 0.363 2083'-6" 0.249 0.297 0.214 4. 0.297 2047'-6" 0.172 0.188 0.144 4. 0.188 2026'-0" 0.133 0.156 0.138 4 0.156 2000'-0" 0.109 0.142 0.134

4. 0.142

WOLF CREEK TABLE 3.7 (B)- 6P RESPONSE ACCELERATIONS (G's) FUEL BUILDING OBE EAST-WEST DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 MASS SITE POINT ENVELOPE CALLAWAY STERLING WOLFCREEK EL. 0.252 0.462 0.384 0.462 2106'*6" ~~ 2083'-6" 0.221 0.372 0.319 u 0.372 2047'-6" 0.173 0.205 0.207 4~ 0.207 2026'-0" 0.149 0.151 0.187 4~ 0.187 2000'*0" 0.132 0.134 0.163 4~ 0.163

WOLF CREEK TABLE 3.7 (B)- 6Q RESPONSE ACCELERATIONS (G's) FUEL BUILDING OBE VERTICAL DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 

                                                                             =;

MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2106'-6" 0.191 0.168 0.214 H 2083'-6" 0.188 0.163 0.210 4. 0.210 2047'-6" 0.180 0.152 0.200 4. 0.200 2026'-0" 0.176 0.149 0.195 4. 0.195 2000'-0" 0.169 0.143 0.186 4. 0.186

WOLF CREEK TABLE 3.7 (B)- 6R RESPONSE INERTIA FORCES (KIPS) FUEL BUILDING OBE NORTH-SOUTH DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2106'-6" 1540 1790 1350 4. .. 1790 2083'-6" 1140 1340 970 4. .. ... 1340 2047'-6" 1670 1930 1250 4. .. ... 1930 2026'-0" 2000'-0" 2010 2290 1890 4 4~ 2290

WOLF CREEK TABLE 3.7 (B)- 6S RESPONSE INERTIA FORCES (KIPS) FUEL BUILDING OBE EAST*WEST DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK

4. .. ..

2106'-6" 1280 2310 1490 2310 2083'-6" 1000 1660 1890 u .. 1890 4t .

  • 2047'-6" 1780 2130 2100 2130 2026'-0" 2350 2240 2490 4. .. 2490 2000'-0" - - - *t

WOLF CREEK TABLE 3.7 (B)- 6T RESPONSE INERTIA FORCES (KIPS) FUEL BUILDING OBE VERTICAL DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 

                                                                                              ~:

MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK t 

                                                               ---~------~------------

2106'-6" 1000 860 1070 4. 1070 2083'-6" BOO 740 950 4.  ; 950 2047'-6" 1800 1560 2060 4. ~ 2060 2026'-0" 2900 2400 3190 ~ 

                                                           *       ~                       3190 2000'-0"     -       -             -                      4~

WOLF CREEK TABLE 3.7 (B)- 6U RESPONSE SHEAR FORCES (KIPS) FUEL BUILDING OBE NORTH-SOUTH DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2106'*6" 4~ 1790 2083'*6" 1540 1790 1350 4. 3130 2680 3130 2320 4. 2047'-6" 4350 5060 3570 4. 5060 2026'*0" 2000'*0" 6360 7350 5460 4. 7350

WOLF CREEK TABLE 3.7 (B)- 6V RESPONSE SHEAR FORCES (KIPS) FUEL BUILDING OBE EAST*WEST DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2106'-6" 4~ 2310 1280 2310 1490 12083'-6" 4. 3970 2280 3970 3380 2047'-6" 4. 6100 4060 6100 5480 4 2026'-0" 2000'-0" 6410 8340 7970

  • 4~

8340

WOLF CREEK TABLE 3.7 (B)- 6W RESPONSE AXIAL FORCES (KIPS) FUEL BUILDING OBE VERTICAL DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 2106'-6"

4. 1070 1000 860 1070 4.

2083'-6" 2047'-6" 1800 1600 2020 4. 2020 4080 3600 3160 4080 4 2026'-0" 2000'-0" 6500 5560 7270

  • 4~

7270

WOLF CREEK TABLE 3.7 (B)- 6X RESPONSE BENDING MOMENTS (MILLIONS OF KIP-FEET) FUEL BUILDING OBE NORTH-SOUTH DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK f- --*------- -- -- ------------~ 2106'*6" 0/0.0144 0/0.0139 0/0.0126

4. 0/0.0144 2083'*6 0.0483/0.0657 0.0550/0.0717 0.0437/0.0593 4. 0.0550/0.0717 2047'-6" 0.1608/0.1860 0.1847/0.2085 0.1426/0.1665 4. 0.1847/0.2085 2026'-0" 0.2795/0.3043 0.3175/0.3394 0.2431/0.2661 4 2000'-0" 0.4662 0.5306 0.3915 4~
  • 0.3175/0.3394 0.5306

WOLF CREEK TABLE 3.7 (B)- 6Y RESPONSE BENDING MOMENTS (MILLIONS OF KIP*FEET) FUEL BUILDING OBE EAST*WEST DIRECTION Rev. 0 REF FIGURE 3 7(B)- 18 MASS SITE POINT ENVELOPE EL. CALLAWAY STERLING WOLFCREEK II 21 06' -6" 0/0.0083 0/0.0127 0/0.0151 4. 0/0.0151 4. I'I 2083'-6" 0.0348/0.0445 0.0657/0.0841 0.0585/0.0805 0.065 7/0.0841 2047'-6" 0.1247/0.0.1352 0.2272/0.2493 0.1931/0.2206 4. 0.2272/0.2493 2026'-0" 0.2225/0.2335 0.3805/0.4015 0.3269/0.3451 4. 0.3805/0.4015 2000'-0" 0.4001 0.6183 0.5522 4. 0.6183

WOLF CREEK TABLE 3.7 (B)- 6Z RESPONSE DISPLACEMENTS (INCHES) FUEL BUILDING OBE NORTH*SOUTH DIRECTION Rev. 0 

               ~--*-                                                    *---~-

REF. FIGURE 3.7(B)- 18 MASS SITE 

            -----                                                ENVELOPE POINT EL. CALLAWAY     STERLING     WOLFCREEK 2106'-6"   0.119        0.069         0.060                                                                 0.119 4.

0.056 0.048 2083'-6" 0.104 4 0.104 2047'-6" 0.079 0.033 0.028 4

  • 0.079 2026'-0" 0.066 0.023 0.019 *
4. 0.066 2000'-0" 0.052 0.011 0.009 4. 0.052

WOLF CREEK TABLE 3.7 (B)- 6AA RESPONSE DISPLACEMENTS (INCHES) FUEL BUILDING OBE EAST-WEST DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 18 II ~~~~-r SITE ENVELOPE EL. CALLAWAY STERLING WOLFCREEK 12106'*6" 0.206 i 0.114 0.154 4~ 0.206 2083'*6" 0.171 0.092 0.127 4. 0.171 4. I I 2047'*6" 0.111 0.050 0.078 0.111 2026'*0" 2000'*0" 0.079 0.040 0.033 0.014 0.054 0.027 4 4 

                                                                 ~

0.079 0.040

WOLF CREEK TABLE 3.7 (B)- 6AB RESPONSE DISPLACEMENTS (INCHES) FUEL BUILDING OBE VERTICAL DIRECTION Rev. 0 REF FIGURE 3.7(B)- 18 MASS POINT EL. 

           ~-

CALLAWAY SITE STERLING WOLFCREEK ENVELOPE I I 2106'-6" 0.026 0.008 0.020 4~ 0.026 2083'-6" 0.025 0.008 0.020 4~ 0.025 2047'-6" 0.024 0.006 0.018 4~ ---0.024 2026'-0" 0.023 0.005 0.017 4~ 0.023 0.022 0.004 0.016 4~ I . 2000'-0" I 0.022 I I

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7A Attachment of ESW Vertical Loop Chase does not RESPONSE ACCELERATIONS (Gs) affect the response accelerations of the AUXILIARY/CONTROL BUILDING SSE Auxiliary/Control Building. NORTH-SOUTH DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7B Attachment of ESW Vertical Loop Chase does not RESPONSE ACCELERATIONS (Gs) affect the response accelerations of the AUXILIARY/CONTROL BUILDING SSE Auxiliary/Control Building. EAST-WEST DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7C Attachment of ESW Vertical Loop Chase does not RESPONSE ACCELERATIONS (Gs) affect the response accelerations of the AUXILIARY/CONTROL BUILDING SSE Auxiliary/Control Building. VERTICAL DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7D Attachment of ESW Vertical Loop Chase does not RESPONSE INERTIA FORCES (KIPS) affect the response accelerations of the AUXILIARY/CONTROL BUILDING SSE Auxiliary/Control Building. NORTH-SOUTH DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7E Attachment of ESW Vertical Loop Chase does not RESPONSE INERTIA FORCES (KIPS) affect the response accelerations of the AUXILIARY/CONTROL BUILDING SSE Auxiliary/Control Building. EAST-WEST DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7F Attachment of ESW Vertical Loop Chase does not RESPONSE INERTIA FORCES (KIPS) affect the response accelerations of the AUXILIARY/CONTROL BUILDING SSE Auxiliary/Control Building. VERTICAL DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7G Attachment of ESW Vertical Loop Chase does not RESPONSE SHEAR FORCES (KIPS) affect the response accelerations of the AUXILIARY/CONTROL BUILDING SSE Auxiliary/Control Building. NORTH-SOUTH DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7H Attachment of ESW Vertical Loop Chase does not RESPONSE SHEAR FORCES (KIPS) affect the response accelerations of the AUXILIARY/CONTROL BUILDING SSE Auxiliary/Control Building. EAST-WEST DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7I Attachment of ESW Vertical Loop Chase does not RESPONSE AXIAL FORCES (KIPS) affect the response accelerations of the AUXILIARY/CONTROL BUILDING SSE Auxiliary/Control Building. VERTICAL DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7J Attachment of ESW Vertical Loop Chase does not RESPONSE BENDING MOMENTS affect the response accelerations of the (MILLIONS OF KIP-FEET) Auxiliary/Control Building. AUXILIARY/CONTROL BUILDING SSE NORTH-SOUTH DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7K Attachment of ESW Vertical Loop Chase does not RESPONSE BENDING MOMENTS affect the response accelerations of the (MILLIONS OF KIP-FEET) Auxiliary/Control Building. AUXILIARY/CONTROL BUILDING SSE EAST-WEST DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7L Attachment of ESW Vertical Loop Chase does not RESPONSE DISPLACEMENTS affect the response accelerations of the (INCHES) AUXILIARY/CONTROL Auxiliary/Control Building. BUILDING SSE NORTH-SOUTH DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7M Attachment of ESW Vertical Loop Chase does not RESPONSE DISPLACEMENTS affect the response accelerations of the (INCHES) AUXILIARY/CONTROL Auxiliary/Control Building. BUILDING SSE EAST-WEST DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7N Attachment of ESW Vertical Loop Chase does not RESPONSE DISPLACEMENTS affect the response accelerations of the (INCHES) AUXILIARY/CONTROL Auxiliary/Control Building. BUILDING SSE VERTICAL DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7O Attachment of ESW Vertical Loop Chase does not RESPONSE ACCELERATIONS (Gs) affect the response accelerations of the AUXILIARY/CONTROL BUILDING Auxiliary/Control Building. OBE NORTH-SOUTH DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7P Attachment of ESW Vertical Loop Chase does not RESPONSE ACCELERATIONS (Gs) affect the response accelerations of the AUXILIARY/CONTROL BUILDING Auxiliary/Control Building. OBE EAST-WEST DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7Q Attachment of ESW Vertical Loop Chase does not RESPONSE ACCELERATIONS (Gs) affect the response accelerations of the AUXILIARY/CONTROL BUILDING Auxiliary/Control Building. OBE VERTICAL DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7R Attachment of ESW Vertical Loop Chase does not RESPONSE INERTIA FORCES (KIPS) affect the response accelerations of the AUXILIARY/CONTROL BUILDING Auxiliary/Control Building. OBE NORTH-SOUTH DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7S Attachment of ESW Vertical Loop Chase does not RESPONSE INERTIA FORCES (KIPS) affect the response accelerations of the AUXILIARY/CONTROL BUILDING Auxiliary/Control Building. OBE EAST-WEST DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7T Attachment of ESW Vertical Loop Chase does not RESPONSE INERTIA FORCES (KIPS) affect the response accelerations of the AUXILIARY/CONTROL BUILDING Auxiliary/Control Building. OBE VERTICAL DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7U Attachment of ESW Vertical Loop Chase does not RESPONSE SHEAR FORCES (KIPS) affect the response accelerations of the AUXILIARY/CONTROL BUILDING Auxiliary/Control Building. OBE NORTH-SOUTH DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7V Attachment of ESW Vertical Loop Chase does not RESPONSE SHEAR FORCES (KIPS) affect the response accelerations of the AUXILIARY/CONTROL BUILDING Auxiliary/Control Building. OBE EAST-WEST DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7W Attachment of ESW Vertical Loop Chase does not RESPONSE AXIAL FORCES (KIPS) affect the response accelerations of the AUXILIARY/CONTROL BUILDING Auxiliary/Control Building. OBE VERTICAL DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7X Attachment of ESW Vertical Loop Chase does not RESPONSE BENDING MOMENTS affect the response accelerations of the (MILLIONS OF KIP-FEET) Auxiliary/Control Building. AUXILIARY/CONTROL BUILDING OBE NORTH-SOUTH DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7Y Attachment of ESW Vertical Loop Chase does not RESPONSE BENDING MOMENTS affect the response accelerations of the (MILLIONS OF KIP-FEET) Auxiliary/Control Building. AUXILIARY/CONTROL BUILDING OBE EAST-WEST DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7Z Attachment of ESW Vertical Loop Chase does not RESPONSE DISPLACEMENTS affect the response accelerations of the (INCHES) AUXILIARY/CONTROL Auxiliary/Control Building. BUILDING OBE NORTH-SOUTH DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7AA Attachment of ESW Vertical Loop Chase does not RESPONSE DISPLACEMENTS affect the response accelerations of the (INCHES) AUXILIARY/CONTROL Auxiliary/Control Building. BUILDING OBE EAST-WEST DIRECTION

REV. 29 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT REV 29 NOTE: TABLE 3.7(B)-7AB Attachment of ESW Vertical Loop Chase does not RESPONSE DISPLACEMENTS affect the response accelerations of the (INCHES) AUXILIARY/CONTROL Auxiliary/Control Building. BUILDING OBE VERTICAL DIRECTION

WOLF CREEK TABLE 3.7 (B)- SA RESPONSE ACCELERATIONS (G's) DIESEL GENERATOR BUILDING SSE NORTH-SOUTH DIRECTION Rev. G 1 MAss 1 - - -- -~~----- --- slr_E__ ~-----------.--* -------- REF. FIGURE 3.7(B)- 20 ~ P~~~TL~A~~AWAY -1 _ST-~~~ING WOLF CREEK l~~*------ ENVELOPE l 12066'-0" 1 0.581 1 0.470 0.375 1 0581 7 2047'-2" I o.477 o.399 o.325 1 I t 0.477 1 I 2o2r-6" 2000'-0" o.374 0.256 o.317 0.246 o.29o 0.255 u 0.374 0.256

WOLF CREEK TABLE 3.7 (B)- 8B RESPONSE ACCELERATIONS (G's) DIESEL GENERATOR BUILDING SSE EAST-WEST DIRECTION H.ev. 'J REF. FIGURE 3.7(B)- 20 

          ~
                                                    ~~-      ~~~-,-------------

MASS 

                                                          ~~- ~

POINT EL CALLA:A: I STE::,:G t~~----r------- -- - - - - - - - - - t - - i I WOLF CREEK 

                                                    -~-  ---

I 

                                                                 ---i---    ~---------------

ENVELOPE i 1 I I l 2066'-0" o.585 1 0.524 

                                                  '       o.516        1                           I                0.585     I j
                                                          ~::;

I 2047'-2" I I u 

          ,     o.468                     0.457                                                                     0.468 I                                                                                              I I                             i I 2027'-6"       0.318 0.184 0.317 0.258 I                     I' 0.318 0.258 I

. 2000'-0" 0.243 - I I

WOLF CREEK TABLE 3.7 (B)- 8C RESPONSE ACCELERATIONS (G's) DIESEL GENERATOR BUILDING SSE VERTICAL DIRECTION Rev. *._) 

                 ------------ --                                                  REF. FIGURE 3.7(B)- 20 r
                                 ---~------

r--~-~~-~ I I SITE POINT ENVELOPE I EL. CALLAWAY I STERLING WOLFCREEK f t-2066'*0" 0.263 0.251 0.266 0.266 I1 tI ------ o.266 2047'*2 2027'*6" 2000'*0" 0.263 0.261 0.258 0.250 0.247 0.244 0.266 0.264 0.260 u 0.264 0.260

WOLF CREEK TABLE 3.7 (B)- 8D RESPONSE INERTIA FORCES (KIPS) DIESEL GENERATOR BUILDING SSE NORTH-SOUTH DIRECTION Rev, 0 REF. FIGURE 3.7(B)- 20 MASS SITE POINT CALLAWAYl--STERLING WOLFCREEK ENVELOPE I 

                                                                  ..                        JI EL.

I I 2066'-0" 670 550 440 670 I 1 2047'-2" I 2o21'-6" 1450 670 1230 580 1010 490 . ____ 1450 670 II I 12000'-0" I

WOLF CR-EEK TABLE 3.7 (B)- BE RESPONSE INERTIA FORCES (KIPS) DIESEL GENERATOR BUILDING SSE EAST-WEST DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 20 I 

           !-cALLAWAY SITE STERLING    WOLFCREEK ENVELOPE                 ~

I 1 2066'-0" 710 620 600 ... 710 1 2047'-2" 1470 1400 1300 ... 1470 II 2o2T-6" 580 580 510 580 12000'-0"

WOLF CREEK TABLE 3.7 (B)- SF RESPONSE INERTIA FORCES (KIPS) DIESEL GENERATOR BUILDING SSE VERTICAL DIRECTION Rev. REF. FIGURE 3.7(8)- 20 MASS I S!TE ~~- p~~~T icAUAWMT__!_~ERLING WOLF CREEK 1 

                       ------~----------~--------

ENVELOPE i 12066'-0" 300 290 310 310 I 2047'-2" 800 750 790 ------800 I i 2027'-6" 460 440 460 460 12000'*0"

WOLF CREEK TABLE 3.7 (B)- 8G RESPONSE SHEAR FORCES (KIPS) DIESEL GENERATOR BUILDING SSE NORTH-SOUTH DIRECTION Re . . ., * -._, REF. FIGURE 3.7(B)- 20 I !':.~~~ ------~- 

             !-              SITE                               ENVELOPE
~~_EL~~--__:_~~LAWA~ __s_T_E_R_L_IN_G_+--_w_o_L_F_C_R_E_EK_--i---------=~--------------1 i 2066'-0"   li           i I I                       1:>"71"\

UfV II ,...... -, 

  'lnA7' '>"

670 2120 550 440 I~ 2120 12027'-6" 1780 1450 u I ~ 2790 12000'-0"j_ -2790__ j __ 2:0__ _L_*-1=0----~-------------------___J

TABLE 3.7 (B)- 8H RESPONSE SHEAR FORCES (KIPS) DIESEL GENERATOR BUILDING SSE EAST*WEST DIRECTION Rev. G REF. FIGURE 3.7(B)- 20 lMASS I S!TE I POINT ~--- ENVELOPE 

~--~L. I CALLA_W_A_Y~-+-~S_T_ER_L_I_N_G~-+---W_O_L_F_C_R_E_E_K~--}

I 1 2066'-o" I 2047'-2" 710 620 600 n I ~------ 710 2180 2180 2020 1900 II 2o2?'-6" 1 1 2760 2760 2600 2410 L__j 12000'-0"

WOLF CREEK TABLE 3.7 (B)- 81 RESPONSE AXIAL FORCES (KIPS) DIESEL GENERATOR BUILDING SSE VERTICAL DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 20 MASS r"VINI 

                ....... I           SITE
                        ~LLAWAY ENVELOPE EL.                   STERLING     WOLFCREEK I 2066'*0"                                                    I         rl II""*~*""

310 ii!U4 ( *II! 300 290 310 I I L, 1100 1100 1040 1100 I I l 1 I 2o2T-6" 1560 1480 1560 II I I 

                                                                        '----J 1560 12000'-0" I

WOLF CREEK TABLE 3.7 (B)- 8J RESPONSE BENDING MOMENTS (MILLIONS OF KIP*FEEn DIESEL GENERATOR BUILDING SSE NORTH-SOUTH DIRECTION _ _ _ _ _ _ _ _ _ _R_E_F_.F_I_GURE 3.7(B)- 20 MASS POINT EL. II r--- 1 CALLAWAY SITE STERLING WOLFCREEK ENVELOPE l! 

                                                                                            --~~---------~---

12066'*0" 0/0.0007 010.0004 1 0/0.0004 , o1o.ooo1 1 12o47'-2" o.o134/o.o212 o.o1o81o.o14s I o.oo81to.o12o I l - - - - - o.o1341o.o212 I II 2027'-6" i I ~\\ l_ 0.0629/0.0698 I 0.0496i0.0525 i 0.0405i0.0434 0.0629i0.0698 II l~-.2_o_o _*-_o_" __]_____ o_.1_4_6~-~.ms __ 0.096_7_ _ _ 

                                                                                                -                      0.1460 .

WOLF CREEK TABLE 3.7 (B)- 8K RESPONSE BENDING MOMENTS (MILLIONS OF KIP-FEET) DIESEL GENERATOR BUILDING SSE EAST-WEST DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 20 II MASS SITE I I POiNT ENVELOPE II EL. CALLAWAY STERLING WOLFCREEK I i 2066'-0" 0/0.0032 0/0.0015 0/0.0023 0/0.0032 I n 12047'-2" 0.0163/0.0235 0.0131/0.0166 I \----- ::::::;:::~:: 0.0138/0.0195 I 2o21'-6" 0.0664/0.0706 0.0564/0.0582 0.0570/0.0601 12000'-0" 0.1464 0.1296 0.1264 I~ 0.1464

WOLF CREEK TABLE 3.7 (B)- 8L RESPONSE DISPLACEMENTS (INCHES) DIESEL GENERATOR BUILDING SSE NORTH*SOUTH DIRECTION Rev. r. v REF. FIGURE 3.7(B)- 20 !MAss S!TE I l I POINT CALLAWAY STERLING WOLF CREEI<j ENVELOPE I EL. I 12066'-0" 0.278 0.043 0.048 0.278 t I I 2047'-2" 0.248 0.034 0.037 j 0.248 I I 12027'*6" 2000'-0" 0.215 0.168 0.022 0.007 0.024 0.005 u 0.215 0.168

TABLE 3.7 (B)- BM RESPONSE DISPLACEMENTS (INCHES) DIESEL GENERATOR BUILDING SSE EAST*WEST DIRECTION Re""-J ~ 0 REF. FIGURE 3.7(B)- 20 MASS SITE POINT ENVELOPE EL. CALLAWAY II STERLING WOLFCREEK 2066'-0" 0.347 0.090 0.124 0.347 2047'-2" 0.304 0.077 0.102 I u 0.304 2027'-6" 0.245 0.049 0.067 I 4~ 0.245 2000'-0" 0.160 0.009 0.015 4t 0.160 I

WOLF CREEK TABLE 3.7 (B)- 8N RESPONSE DISPLACEMENTS (INCHES) DIESEL GENERATOR BUILDING SSE VERTICAL DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 20 MASS SITE I POINT EL. CALLAWAY STERLING WOLF CREEK ENVELOPE I 

                 ----~----------+-----------~         ---------------------------------------!

i 2066'-0 0.010 0.004 0.007 0.010 I u I !2047'-2" 0.010 0.004 0.007 1 f - - - - - - o.o1o I 2o2T-6" 0.008 0.003' 0.006 0.008 !2000'-0" 0.006 0.001 0.004 0.006 I I

WOLF CREEK TABLE 3.7 (B)- 80 RESPONSE ACCELERATIONS (G's) DIESEL GENERATOR BUILDING OBE NORTH-SOUTH DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 20 MASS POiNi SITE ENVELOPE II CALLAWAY STERLING WOLFCREEK II i i i I 2066'*0" 0.261 0.259 0.205 I 0.261 I I I 7 I I 1 1 2047'*2" 0.221 0.218 0.177 0.221 I 2o2?'-6" 0.180 0.168 0.149 I 0.180 I u I I 12000'-Q" 0.130 o.129 I I o.134 _j I 0.134 

                     --~__L------****                        --------                      I

WOLF CREE=<: TABLE 3.7 (B)- 8P RESPONSE ACCELERATIONS (G's) DIESEL GENERATOR BUILDING OBE EAST-WEST DIRECTION Re -,7

  • 0 REF. FIGURE 3.7(B)- 20 I

I MASS SITE l I POINT I I EL. CALLAWAY STERLING WOLFCREEK I ENVELOPE !I I 2066'*0" 0.291 0.320 0.291 

                                                             '! u 0.320 I                                                                     I 1  2047'-2"   0.238 iI 0.282         0.241                    I   I                  0.282 I 2o21'-6"    0.158 I    0.197         0.163                                           n. ..._n""7 v.nH I

/2000'*0" 0.102 0.136 0.130 0.136 i I

WOLF CREEK TABLE 3.7 (B)- 80 RESPONSE ACCELERATIONS (G's) DIESEL GENERATOR BUILDING OBE VERTICAL DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 20 MASS SITE POiNT ENVELOPE I EL. CALLAWAY STERLING WOLFCREEK I I i 

 ! 2066'-0"   0.142    0.131        0.140                                           0.142 I

uI i II 2047'-2" 0.142 0.130 0.139 0.142 

!I 2027'-6" 12000'-0" 0.140 0.137 0.129 0.127 0.136 0.131                  I II                     0.140 0.137

WOLF CREE!\ TABLE 3.7 (B)- 8R RESPONSE INERTIA FORCES (KIPS) DIESEL GENERATOR BUILDING OBE NORTH-SOUTH DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 20 I MASS POINT EL CALLAWAY 

                       ~~~~

STERLING WOLFCREEK ENVELOPE l  ! I 310 ... ... I 2066'-0" 300 240 310 12047'-2" II 670 670 550 ... 670 I 2027'-6" 320 310 310 320 12000'.0"

WOLF CREEK TABLE 3.7 (B)- 8S RESPONSE INERTIA FORCES (KIPS) DIESEL GENERATOR BUILDING OBE EAST-WEST DIRECTION 

. MASS f POINT I  ~---~-
               -----------,~

SITE J REF. FIGURE 3.7(8)- 20 ENVELOPE ,_- ~L. I CALLAWAY I STERLiNG ,. WOLF CREEK I i 2066'-0" 350 i 380 i 350 i 

                                                                           *       ...          380 I      sao            ?5o        I                      ___.,

12047'-2" 730 860 

                                                                           ~----

I 2o21' -6" 2ao I 36o 290 360 I I 2000'-0" I 1

WOLF CREEK TABLE 3.7 (B)- 8T RESPONSE INERTIA FORCES (KIPS) DIESEL GENERATOR BUILDING OBE VERTICAL DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 20 ENVELOPE II I 2066'-0" 170 150 160 

                                               ~                         170 2047'-2" 430 400        420                          ; --------------430 2027'-6" 2000'-0" 250 230        230
                                               ~                         250

VVOLF CREE::< TABLE 3.7 (B)- 8U RESPONSE SHEAR FORCES (KIPS) DIESEL GENERATOR BUILDING OBE NORTH-SOUTH DIRECTION Rev. ,"'"'\ 

                                                                                                  'J REF. FIGURE 3.7(B)- 20 MASS POINT SITE ENVELOPE                              II EL. CALLAWAY    STERLING     WOLFCREEK                                                       I i                     i                                        ,........,                 ,.,~,...

I 2066;-0;; I I I v1v I 300 I 310 240 I L, 2047'-2" 980 I I~ 1 970 980 790 II 2o2?'-6" II 1290 1290 1290 1100 L_J c I 2ooo*-o" J

VJOLF CREEK TABLE 3.7 (B)- 8V RESPONSE SHEAR FORCES (KIPS) DIESEL GENERATOR BUILDING OBE EAST-WEST DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 20 !MASS SITE II ENVELOPE I POINT I EL. CALLAWAY STERLING WOLFCREEK Il nl I 12066'*0" 380 350 380 350 I 2047'*2" 1080 1240 1100 1240 12027'*6" I ~ 1600 1360 1600 1390 LJ 12000'.()"

WOLF CREEK TABLE 3.7 (B)- 8W RESPONSE AXIAL FORCES (KIPS) DIESEL GENERATOR BUILDING OBE VERTICAL DIRECTION Rev. 0 REF. FIGURE 3.7(8)- 20 MASS SITE P"'"'"" 

    . V!!,_!                                                      ENVELOPE EL.        CALLAWAY STERLING     WOLFCREEK IL i 'lnss* -un" J '"'                                                                                      170 11'\1'\A~t          170      150           160 I           1"\tt I  ~---------:::

LU"ti . , 600 550 580 iI 2027'-6" I 850 780 810 L.J l2000'-0"

TABLE 3.7 (B)- 8X RESPONSE BENDING MOMENTS (MILLIONS OF KIP-FEET) DIESEL GENERATOR BUILDING OBE NORTH-SOUTH DiRECTiON Rev. 0 REF. FIGURE 3.7(B)- 20 II I MASS SITE I I POINT ENVELOPE I I EL. CALLAWAY STERLING I WOLFCREEK II I *--~ i i I ! 2066'-0" 010.0003 010.0002 0/0.0002 0/0.0003 I I I n I iI 2047'-2" 0.0059/0.0087 0.0060/0.0082 0.0048/0.0066 I I \ 0.0060/0.0087 I I 2o2?'-6" I 0.0278/0.0302 0.0275!0:0293 0.0221/0.0237 II I I \ I 0.0278/0.0302 I I 12000' -0" 0.0655 0.0647 0.0525  % I \ 

                                                                                                ---- 0.06551 I

                                                 ~AJOLF   CREEK TABLE 3.7 (B)- BY RESPONSE BENDING MOMENTS (MILLIONS OF KIP-FEET)

DIESEL GENERATOR BUILDING OBE EAST-WEST DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 20 MASS SITE I I ENVELOPE CALLAWAY STERLING WOLFCREEK I i i I 2066'*0" I 0/0.0013 0/0.0008 0/0.0012 0/0.0013 I I n j204T-2;; 0.0078/0.0111 0.0079/0.0099 0.0078/0.0107 i 

                                                                                 \            0.0079/0.0111 i

I iI 2027'-6" Ii 

     ~

I 0.0323/0.0341 0.0343/0.0353 0.0323/0.0338 II \ 0.0343/0.0353 12000'.()" 0.0716 0.0793 0.0720 I ~ 0.0793 I

WOLF CREEK TABLE 3.7(B)- 8Z RESPONSE DISPLACEMENTS (INCHES) DIESEL GENERATOR BUILDING OBE NORTH-SOUTH DIRECTION Rev. J REF. FIGURE 3.7(B)- 20 MASS S!TE II ENVELOPE POINT EL" CALLAWAY STERLING WOLFCREEK II 

                                                                                        . l I                                                                                             I I

2066'-0" 0.101 0.022 0.025 0.101 1 I I 2047'-2" 0.090 0.017 0.019 0.090 I u I II iI I 202?'*6" 0.078 0.011 0.013 0.078 I I 12000'-0" 0.061 0.003 0.003 0.061 I 

                                                               ~"JOLF   CREEK TABLE 3.7 (B)- BAA RESPONSE DISPLACEMENTS (INCHES)

DIESEL GENERATOR BUILDING OBE EAST-WEST DIRECTION Rev. 0 REF. FIGURE 3.7(B)- 20 [ -MASS ,1 SITE POINT ENVELOPE II-- _E_L.~--+1!cALLAWAY I I II STERLING WOLFCREEK 

                           ---+--

i 2066;*0" i 0.132 0.054 0.064 0.132 1 I 2047'-2" I I 2o2T-6" I I 12oo~ __o_.o_5__ 0.115 0.090 5 __ 0.046 0.029 0.005 0.053 0.034 0.007 _ L_ _ _ _ _ _ _ _ _ _L __ _ _ _ _ _ _ _ _ __ J_ _ _ __ 

                                                                                    'i  I LJ I

I I 0.115 0.090 0.055

WOLF C:REEK TABLE 3.7 (B)- 8AB RESPONSE DISPLACEMENTS (INCHES) DIESEL GENERATOR BUILDING OBE VERTICAL DIRECTION Rev~ 0 REF. FIGURE 3.7(B)- 20 r MASS SITE I POiNT ENVELOPE r EL CALLAWAY STERLING WOLFCREEK I' 2066'.o.. o.oo5 i I 

                                                                 '! lf 0.002         0.004                                             0.005 1  2047'-2"
  • o.oo5 1 0.002 0.003 0.005 I 2027' -6" I

0.004 II 0.002 0.003 0.004 ~~2-o_o_o*_-o_"~----o_._oo~----------~------------~~ 0.001 0.002 0.003

WOLF CREEK TABLE 3.7(B)-9 DESIGN COMPARISON WITH R.G. 1.12, REVISION 2, DATED MARCH 1997, TITLED NUCLEAR POWER PLANT INSTRUMENTATION FOR EARTHQUAKES Regulatory Guide 1.12 WCGS Instr. Tag No. Position C. REGULATORY POSITION 1.2 A Triaxial time history Complies - See items 1 accelerograph should be through 5 below. provided at the following locations:

1. Free Field Complies. SGAR0001
2. Containment Foundation Complies. SGAE0001
3. Two elevations (excluding Complies with exception SGAE0002, SGAE0003 the foundation) on a structure that one instrument is inside containment. located inside containment and one instrument is located outside containment on the containment wall to support ALARA.
4. An independent Seismic Complies. SGAE0004 Category I structure foundation where the response is different from that of the containment structure.
5. An elevation (excluding the Complies. SGAE0005 foundation) on an independent Seismic Category I structure selected in 4 above.
6. If seismic isolators are N/A - Seismic isolators used, instrumentation should are not used at WCGS.

be placed on both the rigid and isolated portions of the same or an adjacent structure, as appropriate, at approximately the same elevations. 1.3 The specific loctions for Complies - See 1.3.1 instrumentation should be through 1.3.5 below. determined by the nuclear plant designer to obtain the most pertinent information consistent with maintaining occupational radiation exposures ALARA for the location, installation, and maintenance of seismic instrumentation. In general: Rev. 25

WOLF CREEK TABLE 3.7(B)-9 (sheet 2) DESIGN COMPARISON WITH R.G. 1.12, REVISION 2, DATED MARCH 1997, TITLED NUCLEAR POWER PLANT INSTRUMENTATION FOR EARTHQUAKES 1.3.1 The free-field sensors Complies. SGAR0001 should be located and installed so that they record the motion of the ground surface and so that the effects associated with surface features, buildings, and components on the recorded ground motion will be insignificant. 1.3.2 The in-structure Complies. instrumentation should be placed at locations that have been modeled as mass points in the building dynamic analysis so that the measured motion can be directly compared with the design spectra. The instrumentation should not be located on a secondary structural frame member that is not modeled as a mass point in the building dynamic model. 1.3.3 A design review of the Complies. location, installation, and maintenance of proposed instrumentation for maintaining exposures ALARA should be performed by the facility in the planning stage in accordance with Regulatory Guide 8.8, Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable. 1.3.4 Instrumentation should Complies. be placed in a location with as low a dose rate as is practical, consistent with other requirements. 1.3.5 Instruments should be Complies. selected to require minimal maintenance and in-service inspection, as well as minimal time and numbers of personnel to conduct installation and maintenance. Rev. 25

WOLF CREEK TABLE 3.7(B)-9 (sheet 3) DESIGN COMPARISON WITH R.G. 1.12, REVISION 2, DATED MARCH 1997, TITLED NUCLEAR POWER PLANT INSTRUMENTATION FOR EARTHQUAKES

2. INSTRUMENTATION AT MULTI- N/A-WCGS is a single UNIT SITES unit.

Instrumentation in addition to that installed for a single unit will not be required if essentially the same seismic response is expected at the other units based on the seismic analysis used in the seismic design of the plant. However, if there are separate control rooms, annunciation should be provided to both control rooms as specified in Regulatory Position 7.

3. SEISMIC INSTRUMENTATION Complies.

OPERABILITY The seismic instrumentation should operate during all modes of plant operation, including periods of plant shutdown. The maintenance and repair procedures should provide for keeping the maximum number of instruments in service during plant operation and shutdown.

4. INSTRUMENTATION Complies - See 4.1, CHARACTERISTICS through 4.9 below 4.1 The design should include Complies - System provisions for in-service designed with testing. The instruments continuous and periodic should be capable of periodic self test capability.

channel checks during normal plant operation. 4.2 The instruments should Complies - In addition have the capability for in- to continuous self test place functional testing. capability; system has a periodic self test capability. 4.3 Instrumentation that has Complies - Ring buffer sensors located in provided in each inaccessible areas should recorder with a maximum contain provisions for data pre-event recording recording in an accessible time of 17 seconds; location, and the maximum post event time instrumentation should provide of 30 seconds. an external remote alarm to indicate actuation. Rev. 25

WOLF CREEK TABLE 3.7(B)-9 (sheet 4) DESIGN COMPARISON WITH R.G. 1.12, REVISION 2, DATED MARCH 1997, TITLED NUCLEAR POWER PLANT INSTRUMENTATION FOR EARTHQUAKES 4.4 The instrumentation should Complies - Ring buffer record, at a minimum, 3 provided in each seconds of low-amplitude recorder with a maximum motion prior to seismic pre-event recording trigger actuation, continue to time of 17 seconds; record the motion during the maximum post event time period in which the earthquake of 30 seconds. motion exceeds the seismic trigger threshold, and continue to record low-amplitude motion for a minimum of 5 seconds beyond the last exceedance of the seismic trigger threshold. 4.5 The instrumentation should Complies - 2 Mbytes of be capable of recording 25 internal SRAM for each minutes of sensed motion. recorder provides a maximum recording time of approximately 40 minutes in highest resolution (20 bit) setting of recorder. 4.6 The battery should be of Complies - The sufficient capacity to power batteries have the instrumentation to sense sufficient capacity to and record (see Regulatory power each recorder for Position 4.5) 25 minutes of > 40 hours. motion over a period of not less than the channel check test interval (Regulatory Position 8.2). This can be accomplished by providing enough battery capacity for a minimum of 25 minutes of system operation at any time over a 24-hour period, without recharging, in combination with a battery charger whose line power is connected to an uninterruptible power supply or a line source with an alarm that is checked at least every 24 hours. Other combinations of larger battery capacity and alarm intervals may be used. Rev. 25

WOLF CREEK TABLE 3.7(B)-9 (sheet 5) DESIGN COMPARISON WITH R.G. 1.12, REVISION 2, DATED MARCH 1997, TITLED NUCLEAR POWER PLANT INSTRUMENTATION FOR EARTHQUAKES 4.7 Accelertion Sensors Complies - See 4.7.1 and 4.7.2 below. 4.7.1 The dynamic range should Complies - Dynamic range be 1000:1 zero to peak, or of sensors (included in greater; for example, the motion recorder) is 0.001g to 1.0g. >84 dB or >15,800 to 1 or 0.0001 g to 1.0 g which equals 10,000 to 1. 4.7.2 The frequency range Complies - Frequency should be 0.20 Hz to 50 Hz or range of DC to 150 Hz an equivalent demonstrated to (-3dB) be adequate by computational techniques applied to the resultant accelerogram. 4.8 Recorder Complies - See 4.8.1 through 4.8.3 below. 4.8.1 The sample rate should Complies - Motion be at least 200 samples per recorder has a sample second in each of the three rate of 200 sps. directions. 4.8.2 The bandwidth should be Complies - Motion at least from 0.20 Hz to 50 recorders have a Hz. bandwidth of DC to 50 Hz. 4.8.3 The dynamic range should Complies - Motion be 100:1 or greater, and the recorders have a minimum instrumentation should be able dynamic range (for 16 bit to record at least 1.0g zero A/D converter) of 96dB or to peak. > 65,000 to 1. The recorders can record + or 

                               - 2.0 g zero to peak.

4.9 Seismic Trigger. The Complies - Trigger level actuating level should be is adjustable on all adjustable and within the three (3) axes from range of 0.001g to 0.02g. 0.0005 to 0.05 g. The system trigger is set within the range as stated in 4.9

5. INSTRUMENTATION Complies - See 5.1 INSTALLATION through 5.3 below.

Rev. 25

WOLF CREEK TABLE 3.7(B)-9 (sheet 6) DESIGN COMPARISON WITH R.G. 1.12, REVISION 2, DATED MARCH 1997, TITLED NUCLEAR POWER PLANT INSTRUMENTATION FOR EARTHQUAKES 5.1 The instrumentation should Complies. be designed and installed so that the mounting is rigid. 5.2 The instrumentation should Complies. be oriented so that the horizontal components are parallel to the orthogonal horizontal axes assumed in the seismic analysis. 5.3 Protection against Complies. accidental impacts should be provided.

6. INSTRUMENTTION ACTUATION Complies - See 6.1 through 6.3 below.

6.1 Both vertical and Complies - The time horizontal input vibratory history accelerographs ground motion should actuate contain triaxial the same time-history accelerometers. The accelerograph. One or more accelerographs will seismic triggers may be used trigger on the selected to accomplish this. values if present on any axis. 6.2 Spurious triggering should Complies - Coincidence or be avoided. voting logic will reduce or eliminate spurious triggering. 6.3 The seismic trigger Complies - The trigger of mechanisms of the time-history the free field strong accelerograph should be set motion accelerometer for a threshold ground /recorder is set for acceleration of not more than 0.02g. 0.02g.

7. REMOTE INDICATION Complies - Triggering is Triggering of the free-field annunciated in the MCR.

or any foundation-level time-history accelerograph should be annunciated in the control room. If there is more than one control room at the site, annunciation should be provided to each control room. Rev. 25

WOLF CREEK TABLE 3.7(B)-9 (sheet 7) DESIGN COMPARISON WITH R.G. 1.12, REVISION 2, DATED MARCH 1997, TITLED NUCLEAR POWER PLANT INSTRUMENTATION FOR EARTHQUAKES

8. MAINTENANCE Complies - See 8.1 and 8.2 below.

8.1 The purpose of the Complies. maintenance program is to ensure that the equipment will perform as required. As stated in Regulatory Position 3, the maintenance and repair procedures should provide for keeping the maximum number of instruments in service during plant operation and shutdown. 8.2 Systems are to be given Complies. channel checks every 2 weeks for the first 3 months of service after startup. Failures of devices normally occur during initial operation. After the initial 3-month period and 3 consecutive successful checks, monthly channel checks are sufficient. The monthly channel check is to include checking the batteries. The channel functional test should be performed every 6 months. Channel calibration should be performed during each refueling outage at a minimum. Rev. 25

WOLF CRI!!Et< FREQUENCY (cps) 100 15 0 1 5 l *~ or o~ o;, 02 OP.: c* ITT'1rT'T'"~1..,.,.,rTTT'IrrrT~mTITTn'TTTTT,-.-rrr*TtT~IT',....,..., ..T.-r.,"""fT*"T'"'" f"TTT"'I TTT~ r *.,.......... , I "'!"TT"TTf"fT'TT"f"Tf"n"T -* ....,. --, 

                                                                                             ..                 *.                                          .:            ~*JC
                                                                                                   . .                                       .     .      .       .       ~~
                                                 ~~~-~~~ey~~+:~~r--~~-~E+~~~~~~

ti. -~ 2 PERIOD (sees.) WOLF CREEK REV.22 UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 7<8>-1 SSE HORIZONTAL GROUND SPECTRA 0.20g <WCGS>

WOLF CREEK FREQUENCY (cps) 2 1$ Ql ""Tn~~~~~~~~~~~~~~~~~~~ PERIOD (sees.) WOLF CREEK REV.22 UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 7CB>-2 SSE VERTICAL GROUND SPECTRA 0.20g (WCGS>

WOLF CREEK NOTE: USE APPROPRIATE NORMALIZATION SCALE FACTORS FOR OBE & SSE TIME HISTORY ANALYSIS. 0 3 6  !~ 24 T!ME (SECOND) Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-3 SYNTHESIZED TIME HISTORY HORIZONTAL (QBE AND SSE)

WOLF CREEK z 0 i= <l 0:: 11.1 ..J 11.1 0 0 NOTE: <l USE APPROPRIATE NORMALIZATION 

      -1.00                             SCALE FACTORS' FOR OBE & SSE TIME HISTORY ANALYSIS.

0 3 e. i5 i8 2! 24 T i M E ( SECOND ) Rev. 0 WOLF CRBBB: DPDA~BD SAFETY ANALYSIS RBPOR~ FIGURE 3.7{8}-4 SYNTHESIZED TIME HISTORY RTICAL 

                                                                            \OBE AND SSE J

lvOLF CREEK D r~ CJ () CJ aJ Cl (j ~.* z ....D CJ 1-CI 0::: l1l Cl w  ;--60"/o OF _j 1 DESIGN SF'E:CH1A w u u Cl CI

1' Cl D

(\) Cl 

     ~l    0. 10 1r 10.00 r-r-rr-n-n 100.

FREQUENCY CHZl RE?V. 1 WOLJI!' CREEK -**-"-"1 UPDATED SAFETY ANALYSIS REPORT Figure 3.7(B) - 9A 

                             'Typical Free-Field Bc:.tse Ele*vation Spectra CalLaway Si i <:'

liVOLF CREKK 0 N 0 0 0 OJ

i:!:

EJ ..... 0 1- (() a: * - 6 0 °/o OF 0:: 0 IJJ / DESIGN SPECTRt~ ..J IJJ tl tl a: 0

1' 0

0 N 0 0 0 

     ~+-----,---,...---r-1--r-1~I.....I ...,.I.....I....,I_,__T 0.10                                       1.00                        I II
                                                                                     ],0.00 FREQUENCY CHZl Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8}-98 TYPICAL FREE-FIELD BASE ELEVATION SPECTRA STERLING SITE
                                                                        *-----------**-...                 -~~-

IIi\ WOLF CREEK a (\J a a a IXl a u z 

    ....0  a I-a:
    ~

(() a 

              .                                      60 °/o OF llJ                                              DESIGN SPECTRA
    ...J llJ u

u a a:

1' a

a (\J a a a

1. 00 10.00 100.

FREQUENCY CHZl Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(B)-9C TYPICAL FREE-FIELD BASE ELEVATION SPECTRA TYRONE SITE

0 N 0 0 0 QJ l-- ac

  • Q~ 0 llJ

_J llJ u ll 0 ac :I' 0 

                              /

0 N 0 0 

               ~
                    /

0 

   ~~-----~--~-.-.~

0.10 Ill

1. 00 T n---rrn--1o.oo 1*-rrrTTJ 100.

FREQUENCY CHZl Rev. 0 WOLF CREEK UPDA'l'!:D SAFETY ANALYSIS FIGURE 3.7(8)-9[) TYPICAL FREE-FIELD BASE ELEVATION SPECTRA WOLF CREEK SITE

0 C\t 0 0 J~ 0 Cl) d z 0 1-

  • c(

0 60 *t. OF 11:: L.LI 10 DESIGN SPECTRA

  • ...J d L.LI (J

u

  • c(
                                           /~

0 

     ~

d 

                                         /

0 

                                 /    /1
                                 /

C\t d 0 

                 ~

g-+----.....---.----.---.-1'TTnr--*-r I I I I III r-*-r-r**rrrl 0.10 1.00 10.00 100. FREQUENCY (HZ) Rev. 0 WOLF ClEEK UPDATED SAFETY ANALYSIS FIGURE 3.7(8)-10 TYPICAL FREE-FIELD BASE ELEVATION SPECTRA THREE SITE ENVEI..OPE 

                                            --------------*. **---*1111

WOLF CREEK w 0: t-0 0: t-en (!) z 0'- 0 0'- 0 0'- 0". _, 0 IJ) (/) al w 0::: 0: (/)0::: 0 0 

          .....J::J       I-
          <(I-            <(            t-zu              0:::          0
 .....J w        o::::::>        w             <(

w wo::: z IJ) (/) r-1-- z(/) w 0: w <.9 

 >                        ~
                          <(

w I-(/) RIGID BASE MAT Kx Kz Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-12 MATHEMATICAL MODEL FOR REACTOR BUILDING AND INTERNAL STRUCTURES

1-z 11.1 

                 }:

z 0...J 0 

                 ~                                Q:

z 1-0 z u 0 

               @                                  u Cl

)( 

                                      -Y(t) n!    ol
                      ~~~ ~~~ ~~ ~:~I                                   ~~

Ol 

            ~~

I I .nl ii ~~ 

                                               "'"   ol "'I
  ~ ~I ~I             .   *   -1               ~     ~I ~I    ~I ~ ~

Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-13 THE FINITE-ELEMENT MODEL

DAMPING VALUES WOLF CREEK 

    .0300, .0500, .0700 6.00 5.00

( f) II 4.00 -z 0 r* .... 0:: w_J w 2.00 u u r 1-'1 t/: h ~ 

                               ..r: G  I""      ,....
                            ~                          t:l~
     .0000
     .oo                  1.00                             10.00                    tOO.O FREQ!JENCY <CPS>

DESIGN FLOOR RESPONSE SPECTRA WOLF CREEK REV *22 UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7C8)-14A SPECTRA - CONTAINMENT BUILDING SSE. NORTH-SOUTH DIRECTION. POLAR CRANE LOCATION. CALLAWAY SITE

DAMPING VALUES WOLF CREEK 

    .0300, .0500, .0700 6.00 5.00 U) 4.00 z

0 ~ <( ......., 0::: w _J w 2.00 i-,r\ (.) (.) ~ <( IJ 

                                      ~~
                              ~
                                ~ f'                     '

0000 

    .00                     1.00                           10.00                   100.0 FREQUENCY <CPS>

STERLINC REACTOR BLOC. SHELL EL. 1119'-o* NORTH SSE DESIGN FLOOR RESPONSE SPECTRA WOLF CREEK REV *22 UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 7<8>-148 SPECTRA - CONTAINMENT BUILDING SSE. NORTH-SOUTH DIRECTION. POLAR CRANE LOCATION. STERLING SITE

DAMPING VALUES WOLF CREEK 

                         .0300, .0500, .0700 6.00 5.00

( /) 4.00 A -- 0 ~* ... 

                   -z 0*           -

I- 

                    <(

a:: ,~ w __.J w 2.00 \ 

                                                                    '/

0 0 

                    <(                                            I       '
                                                                         \
                                                              ~?
                                                        ~p
                                                       "//'
                                                                            \ ..
                         .. 0000
                         .10                        1.00                         10.00                      100.00 FREQUENCY CCPS>

MILf CREEK REACTOR BLOC. SHELL El. 2119'-G" NORTH SSE DESIGN FLOOR RESPONSE SPRCTRA WOLF CREEK REV. 22 UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 7(8)-140 SPECTRA - CONTAINMENT BUILDING SSE. NORTH-SOUTH DIRECTION. POLAR CRANE LOCATION. WOLF CREEK SITE

WOLF CREEK DAMPING VALUES 

         .0300, .0500, .0700 6.00 5.00

(/) 4.00 

 -z 0

I- ~ 

 <{

n:: e w_J w t.OO u u 

 <{

v- """' ,... l 

                                            ..c_ .c
                                        ~                            ~       ~
          .0000
         .10                         t.OO                               10.00                      100.0 FREQUENCY <CPS>

DESIGN FLOOR RESPONSE SPECTRA CALLAWAY REACTOR BLDG. SHELL EL. 2119'-0" EAST SSE WOLF CREEK REV

  • 22 UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 7CB>-14E SPECTRA - CONTAINMENT BUILDING SSE. EAST :-WEST DIRECTION.

POLAR CRANE LOCATION. CALLAWAY SITE

WOLF CREEk DAMPING VALUES 

         .0300, .0500, .0700 6.00 5.00

( /) 4.00 r---- 

 -- !I   . --*       *--           .. - ....                                              . -

C) z p~ e 0 I- 

    <l:

0::: w '/~ _J w 2.00 i\ u u t\ 

    <l:                                            1/
                                               /i, ~ "'
                                        £
                                    ~::~r rr                 \ .*.        ~
          .0000
          .00                        1.00                              10.00                        100.0 FREQUENCY CCPS>

STERLINC REACTOR BLOC. SHELL El. 2119'-D" EAST SSE DESIGN FLOOR RESPONSE SPECTRA WOLF CREEK REV. 22 UPDATED SAFETY ANALYSIS REPORT 

                                                                            , FIGURE 3. 7C8)-14F SPECTRA - CONTAINMENT BUILDING SSE. EAST:-WEST DIRECTION.

POLAR CRANE LOCATION. STERLING SITE

DAMPING VALUES WOLF CREEK 

      .0300, .0500, .0700 6.00 is.oo

...... 4.00 (/) II z 0 1- 2.00 <( 0::: I~ w __J w 1f () () <( u \ rL ~ ~ 

                                         ~

p ~ 

                                     ~~
         .0000
         .00                         1.00                              10.00                       100.0 FREQUENCY <CPS>

WOLF CREEK REACTOR BLOC:. SHELL El. 2.119'-G" EAST SSE DESIGN FLOOR RESPONSE SPECTRA WOLF CREEK REV *22 UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 7CB>-14H SPECTRA - CONT AlNMENT BUILDING SSE. EAST -WEST DIRECTION. POLAR CRANE LOCATION. WOLF CREEK SITE

DAMPING VALUES WOLF CREEK 

          .0300, .0500, .0700 2.00

( /) 

  <.:)

z 0 .. - .... . .,..... 

 "I-'       -* -*.     - .   . .                       **-
  <(

cr w _J w .... e ~ u u 

                                                                         ...,.. lr ~
  <(

r 

                                                                  ~

1/,.. I 

                                                                                             \

L0V' 

                                                                                                     ~

b~~'- !::;::7 r-- jl;! IIi 

                                 ~~
           .0000   ... ~ ~
          .00                                    1.00                                      10.00                              100.0 FREQUENCY <CPS>

CALLAWAY REACTOR BlOC. SHEll EL. 2119 '-0" \IERT. SSE DESIGN FLOOR RESPONSE SPECTRA WOLF CREEK REV *22 UPDATED SAFETY ANALYSIS REPORT FIGURE .3. 7(8)-141 SPECTRA - CONTAINMENT BUILDING SSE. VERTICAL DIRECTION. POLAR CRANE LOCATION. CALLAWAY SITE

DAMPING VALUES WOLF CREEK 

      .0300, .0500, .0700 2.00

(/) fi

      • z ... . . ... -- ( ... ..
                                                                              ~n      ..       ..

0 

                                                                            ~u 1--
 <(

a::: w If l r _J w lr r-v ir'l' I tv' l1. ~ 

                                                            -- I

() ~ 

                                                                                     ~

() 8 f- 

 <(

r-1- 

                                                         ~

Vl ~ t- 

                             ~ 115 l'i b~ ~
                     ~~
      .0000  ~ 1--"
       .00                               1.00                               10.00                        100. 0 FREQUENCY <CPS)

STERLING REACTOR BlOC. SHELL El. 2119'-0" VERT. SSE DESIGN FLOOR RESPONSE SPECTRA WOLF CREEK REV

  • 22 UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7C8)-14J SPECTRA - CONT AJNMENT BUILDING SSE.. VERTICAL DIRECTION.

POLAR CRANE LOCATION. STERLING SITE.

DAMPING VALUES WOLF CREEK 

      .0300, .0500, .0700 12.00

( /) II z Jl 0 r <( 0::: w t/~ n _J w u I y r -:,.~ II 

                                               ~

u <( 

                                                                               \:

jiiS ~I:;: 

                                        ~      '/                                 ~ t-
                          ~
                        ~
       .0000     ~~
       .00                             1.00                        10-:olf                         100.0 FREQUENCY <CPS>

WOLF CREEK REACTOR BLOC. SHELL EL. 2119'-D" VERT. SSE DESIGN FLOOR RESPONSE SPECTRA WOLF CREEK REV

  • 22 UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 7-14L SPECTRA - CONTAINMENT BUILDING SSE. VERTICAL DIRECTION. POLAR CRANE LOCATION. WOLF CREEK SITE

DAMPING VALUES WOLF CREEK 

       .0300, .0500, .0700 i                                                      I I

6.00 I ( /) 4.00 z:* .. -** . .. .... ~ 0 I- 

  <(

0::: w __J w 2.00 0 

  ~                                                            ~

II, ~v iii h I be ~. ~ t- 

                                               ./                 ~to-
                                        ~ v---                         ~
       .0000
       .00                             1.00                              10.00                          100.0 FREQUENCY <CPS>

C.ALlAWAY REACTOR BLDG. !~HELL El. 1119'-i)" NORTH OBE DESIGN FLOOR RESPONSE SPECTRA WOLF CREEK REV *22 UPDATED SAFETY ANALYSIS REPORT -e FIGURE 3. 7-14M SPECTRA - CONTAINMENT BUILDING OBE. NORTH-SOUTH DIRECTION. POLAR CRANE LOCATION. CALLAWAY SITE

DAMPING VALUES WOLF CREEK 

       .0300 *. 0500 *. 0700 6.00

- 4.00 r 

           ~ ... .. . .     . .   - .         . .  ..                             ..           ..

z 0 I- <( 0:: w 2.00 r-'\ _J w I (.) ~ r~'~ 

                                             ? ~v
                                                                  '=
       !                                                                  ~

I I ~~ 1 

         .0000
         .00                          1.00                            10.00                          100.0 FREQUENCY <CPS>
                       ~iEP.l!NC ~EACiOR   BLDG. ~HELL    El. 21 19' -0"    NORTH    OB~

DESIGN FLOOR RESPONSE SPECTRA 22 WOLF CREEK REV

  • UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 7-14N SPECTRA - CONTAINMENT BUILDING OBE. NORTH-SOUTH DIRECTION.

POLAR CRANE LOCATION. STERLING SITE,

DAMPING VALUES WOLF CREEK 

       .0300, .0500, .0700 6.00

( /) 4.00 1'- 0 z -* _, - . -** . *- - -* -. .. 0 1- 

 <(

0::: w e _J 2.00 ~\ w u u I 

 <(                                                       II (V         \

r? rt:~ y r::~ 0000 

        .00                                1.00                           10.00                     100.0 FREQUENCY <CPS>
                    #tiOl.F CREEK    REACTOR 9LDi;. SHELL         EL. 2119. -0" NORTH  OBE DESIGN FLOOR RESPONSE SPECTRA WOLF CREEK      REV *22 UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 7-14P SPECTRA - CONTAINMENT BUILDING OBE.

NORTH-SOUTH DIRECTION. POLAR CRANE LOCATION. WOLF CREEK SITE

DAMPING VALUES WOLF CREEK 

          .0300, .0500, .0700 6.00 5.00 en        4.00 Z        ~--~~-+~~~~~-----+--~~~~~H-----+-~~~~~HH 0

t- 

 <{
 ~

w 

 .......J 2.00 w

u u 

 <t:      1----~---r~-+~++++-----r--~~~~~+-----+-~--~~~HH

(~ 

          .0000
          .00                       1.00                      10.00                       100.0 FREQUENCY <CPS)

CALLAWAY REACtOR BLOC. SHELL EL. 2119'-0~ EAST OSE DESIGN FLOOR RESPONSE SPECTRA WOLF CREEK REV *22 UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 7(8)-140 SPECTRA - CONTAINMENT BUILDING OBE. EAST -WEST DIRECTION. POLAR CRANE LOCATION. CALLAWAY SITE

DAMPING VALUES WOLF CREEK 

      .0300, .0500, .0700 6.00 5.00 4.00 U1 i\

z ... .. - .. . 

                                  -                ..              1-\                      -*     ...  -~*

0 I- 

  <(

0:: w 2.00' _J w 0 \ 

  ~

t t:?~ n 

                                                                         ~~~
                                            ~~                                   6
                                    ~  ~
       .0000
       .00                            1.00                                   10.00                          100.0 FREQUENCY <CPS>

STERliNG REACTOR BLOC. SHELL EL. 2119'-0" EAST OBE DESIGN FLOOR RESPONSE SPECTRA WOLF CREEK REV. 22 UPDATED SAFETY ANALYSIS REPORT e_ FIGURE 3. 7(8)-14R SPECTRA - CONTAINMENT BUILDING OBE. EAST -WEST DIRECTION. POLAR CRANE LOCATION. STERLING SITE

DAMPING VALUES WOLF CREEK 

   .0300, .0500, .0700 6.00
   . 5.00 Ann t\

(/) II z 

           .~- . --*  ....   - '- .... . -*
                                                                  !\

0 1- <( 0:: w __J 2.00 w 1\ u u t <( n (IJ ~\.,:~ ,......., 

                                                  ~~                           -

J!j 

                                            ~~                                                 '
     .0000 t.OO                             10.00                            0
     .00 FREQUENCY <CPS>

STERL INC REACTOR BLDG. SHELL El. 2119 '-0" EAST OB£ DESIGN FLOOR RESPONSE SPECTRA WOLF CREEK REV *22 UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 7CB>-14 T SPECTRA - CONTAINMENT BUILDING OBE. EAST -WEST DIRECTION. POLAR CRANE LOCATION. WOLF CREEK SITE

DAMPING VALUES WOLF CREEK 

       .0300, .0500, .0700 2.00

( /) z 0 . ,-. .. - - .. - .. +- <( 0:: w 1\ I \ .....J w v lr ..r 

                                                                                 ~

u  !-... I,I u 

                                                   !r r"

<( rv . .-1-1 ~\ c£ 1/ 

                                                                                     ~
                                                    /
       .0000            IllS""'
      .00                                 1.00                                10.00                      100.0 FREQUENCY <CPS>

CALLAWAY REACTOR BLDC. SHELL EL. 2119'-0~ VERT. OBE DESIGN FLOOR RESPONSE SPECTRA WOLF CREEK REV *22 UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 7(8)-14U SPECTRA - CONTAINMENT BUILDING OBE. VERTICAL DIRECTION POLAR CRANE LOCATION. CALLAWAY SITE

DAMPING VALUES WOLF CREEK 

    .0300, .0500, .0700 2.00

( /) r-1 z 0 I- .. n <( .. lr~ a:: w _J w (.) (.) ~---~ 1(11 v- ~ r... 1,..1 <( II ~ V, tr v-' 

                                                      ~
                                                         ~ 1-L- I-'

II II 

                                      ~v                                             ~ r--
      .0000
     .00
                        -          1.00                               10.00                       100.0 FREQUENCY CCPS>

STERLING REACTOR BLDG. SHELL EL 2.119'-o" ¥£Rl. OBE DESIGN FLOOR RESPONSE SPECTRA WOLF CREEK REV *22 UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 7CBl-14V SPECTRA - CONT AlNMENT BUILDING OBE. VERTICAL DIRECTION. POLAR CRANE LOCATION. STERLING SITE

DAMPING VALUES WOLF CREEK 

      .0300, .0500, .0700 I

2.00 ( /) M ~ (.) z 0 ,......, I- 

 <(                               ..  . ..

a::: w _J w rn () 

                                                                   '\.:.-

() r ~ r-... e 

 <(                                                          r-I,..., ~ """' -  ...... ll I

I 

       .0000 kt          ,_,-/                                      \~
       .00                    1.00                                       10.00                              100.0 FREQUENCY CCPS>

WOLF CREEK REACTOR BLDC. SHELL EL. 2119 '-G" \fERT. OBE DESIGN FLOOR RESPONSE SPECTRA WOLF .. CREEK REV *22 UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 7-14X SPECTRA - CONT AJNMENT BUILDING OBE. VERTICAL DIRECTION. POLAR CRANE LOCATION. STERLING SITE

WOLF CHEEK DAMPING VALUES

  • 0300,
  • 0500,
  • 0700, 2 .. - ,__ _ ,. ..
                                                      ..                                         --         i
                                                                                                              ~*"

I I 1--1--- (J) (!) 

                                                                                                  *-1*-***

I z 

                                                                                                 ..-1*---*

0 ...... 1 ,__ -* a::: w ~- ... _j w I r h 

                                  //:

u ~~*~* u 1\.[r h 1' ~ ... ,__., -I\ < [\-- llvy v -~ 

                                                                  ' ~~--:-
                                                                      .,r-___.;,-.,

k; ~ ,. . .Y 

                  ~  ~
                                                                                    ~' -           -f.-
               ~~                                                                        ~
        , ~~

1.0 10.0 100.0 F R E Q U E N CY ~ CP S ) DESIGN FL~JO!!_~ESPOI\ISE SPECTRA Rev. 0 WOLF CREEK UPDA'l'ED SAFE"l"Y ANALYSIS REPORT FIGURE 3.7(8)~15A SPECTRA - CONTAINMENT BUILDING* SSE, NORTH-SOUTH DIRECTION, STEAM GENERATOR UPPER SUPPORT, CALLAWAY SITE

WOLE' CREEK DAMPING VALUES

  • 0300,
  • 0500,
  • 0700, 2
                                                      ..                                     *-I*-  w ** ,
                                                      ..                                     ---.   ~-'"'

(/) (!) ~~ 

                                                                                              . r- -<*
                                                 /                                                         i' z                                                                                              .  ,_ ...I a
                                                   ~'
                                              -~/(

r--, 

                                               ~

a: UJ  ! 

                                    );
  • 1--

..J ,1..., UJ r-'"' lr IJj; u IV 

                                                                                               -  .   -~

u 

                           /

I 

                                                          \ =~ ~1\r
                                                                                  ~            ..           I f-.1
                   ~ ~v
                       ~ VII v                                                        '~ 1--
                                                                                               *I-
                ~~
        ,, ~~
     .1                        1.0                                        10.0                     100.0 F R E Q U E N CY            CCP S J DESIGN FL90!3__RESPONSE SPECTRA_

Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8}-158 SPECTRA - CONTAINMENT BUILDING SSE, NORTH-SOUTH DIRECTION, STEAM GENERATOR UPPER SUPPORT, STERLINr SITE 

                                                     \   .

DAMP1NG VALUES WOLF CREEK I

  • 0300,
  • 0500,
  • 0700, 2.

(I) (!) z ,_..., I 0 u,LJ 1- 1 n 

 <                                                h I 0::

r-r lLJ h v 

                                       )/
 ...1 I~

vw- ' lLJ u 

                                                                        ~

u 1/ 

 <                                                                 lJ 1/                          l\

I\\£ I/ vv 1/ ~ 1/l/ ~ 

                         ~                                                     ~
                      ~ ~~
            ,~~

0 

         .1                        1.0                                to.o                  100,0 F R E Q U E NC Y       CCP S J DESIGN FLOOR RESPONSE SPECTRA                               Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(B)-15D SPECTRA - CONTAINMENT BUILDING SSE, NORTH-SOUTH DIRECTION, STEAM GENERATOR UPPER SUPPORT, WOLF CREEK SITE

DAMPING VALUES WOLF CREEK

  • 0300,
  • 0500,
  • 0700, 2

(f) (!) z a 1- 1 a:: w .J r-- r r w r Vl n p u I"' 

                                            ~ I I~

r'\1\ 

                                                      -                 '~

u r- 

                                                        ...... ~---~
                                            ~ N~

< ['\ 

                                               ~

I ,. ./ '/ 

                                                                               ~
                         ,//
               ~

f2 

                  ~ F-
                       ~
        ,~

0 

     .I                         1.0                                 1!1,0                    100.0 F R E Q U E NC Y         ~ CPS J DESIGN FLOOR RESPONSE SPECTRA                                   Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-15E SPECTRA - CONTAINMENT BUILDING SSE, EAST-WEST DIRECTION, STEAM GENERATOR UPPER SUPPORT, CALLAWAY SITE

DAMPING VALUES WOL.E~ CREEK 

            .. 0300,
  • 0500,
  • 0700, 2
                                                                                                       - . r*

I I [ 

                                           ,I
                                               - --*                                                   . i--i-*
                                       --                                                      ....J L

(f) I (!) i z J 

                                                    'j 0                                                        r-...,

..... 1 r---'\ 0:: _} 

                                                    ~(

I I w ) {I- 

       ~*
                                                                     )

lr~ w If ;-r-kr\ v u V' __...., v u 1-v v 

                         ;~  ~

f../ 

                     ~~

i-- 

                         /
       -*         ~
          ~
            ~

0 1,0 - 'UI,O *- - UJIO,O F R E Q U E NC Y CC P S l DESIGN FLOOR RESPONSE SPECTRA Rev. 0 WOLF CREEK OPDA~ED SAFE~Y ANALYSIS REPOR~ 

                                                     ~----~~~-----~--~------------

FIGURE 3.7(8)-15F SPECTRA - CONTAINMENT BUILDING SSE. EAST-WEST DIRECTION, STEAM GENERATOR UPPER SUPPORT. STERLI~G SITE

DAMPING VALUES WOLF' CREEK 

          ,, 0300,
  • 0500,
  • 0700, 2

II I I, -- __ , I (f) I (!) I I i li 

                                                                                      -- -t' z

r--; 1-* 0 1-1 lf Tl 0:: __ /( h 1- IJ w r- 

                                    /r:
                                                               ~~\

..J .J w 

                                                           /I-Jj; u                                                                                               -*

v \ 

                                                       '~

u 

                            /

I l ~ v// / 

                            /    _/                     ~
                                                                           ~

b¢ ~ 

                      ~
                        ~
        , ~~ ~

0 

     ,1                         1.0                            10.0                        UIO.O F R E Q U E NC Y          CC P S J DESIGN FLOOR RESPONSE SPECTRA Rev.           0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-15H SPECTRA - CONTAINMENT BUILDING SSE,    EA~T-WEST        DIRECTION. STEAM GENERATOR UPPER SUPPORT, WOLF CREEK SITE

WOLl? CREEK DAMPI"'G VALUES

  • 0300,
  • 0500,
  • 0700,
                                           --*                                                    *t-II
                                           --*                                                **I-(fl                                                                                                  I

(!) I i u

2. 0 -*

z 0 ..... j II I I 

                                               !                                      I 1-c:r::

UJ _J r' - UJ u 

                                                                         ~
                                                   /"

u I ! ,__I-. ...... 

                                                   .r            v f.-
                                           /-!.    )
                                                                        \!\,
                                                                         ~
                            ~~
                                 /,
                                     ~~                                      ~
                         ~

0 L-. 

              ... ~ ~                                                                                   '
           .1                         1.0                               10.0                      10!1.0 FRE       Q U E NC Y         CCP S l DESIGN FLOOR      .

RESPONSE SPECTRA- Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7{8)-151 SPECTRA - CONTAINMENT BUILDING 

                                                           .SSE, VERTICAL DIRECTION, STEAM GENERATOR UPPER SUPPORT, CALLAWAY SITE
                                                      '------------------~,......,

DAMPING VALUES WOLF CREEK

  • 0300,
  • 0500,
  • 0700,

(/) (.!) LJ 2.0 z 0 H 1-0:: L1J _j L1J u I "-fF 

                                                             ,r
                                                ~

u r- r 

                                                                 ~~

< v 

                                            /

v~ 

                                                /                             ~
                                  ..?~                                              r-iii""
                          ~
                    ~~

0 

           .,. ~
        .1                              1.0                      10.0                        100.0 F R E Q U E NC Y           CCP S )

DESIGN FLOOR RESPONSE SPECTRA Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-15J SPECTRA - CONTAINMENT BUILDING SSE, VERTICAL DIRECTION, STEAM GENERATOR UPPER SUPPORT, STERLING SITE

WOLF' CRf~EK DAMPING VALUES

  • 0300,
  • 0500,
  • 0700,
                                             --r-                                            ... t-*1-
        *--                                              -                                   .f-.
                                                                                                  ~l-n                                                                                                     I (J)

(!) v 2.0 1----+--+-+-1-t-t-H-t------* - z i 0 1-f 1- 

           --i----t--i--+--1-+-H-t---- *-    -- t-*                                           . <

0::  ! w ..J I w (J v-I\- r-- h (J 1/ lr- t--- 1-- "'" r - - t--- I - r"~o- ~~'-, _./ \,.. 

                                            ..:;)                         ~                       I-LD 1'-

8 

       .1                         1.0       -                           10.0                        too.o F R E Q UE N C Y             CC P S l DESIGN FLOOR RESPONSE               SPECTR~

Re 11. 0 WOLP CREEK UPDATED SAPB'l'Y ANALY:SIS REPOR'J" FIGURE 3.7(8)-15L SPECTRA - CONTAINMENT BUILDING SSE, VERTICAL DIRECTION. STEAM GENERATOR UPPER SUPPORT. WOL~ CREEK: SITE

DAMPING VALUES WOLF CREEK 

          .0100,     .0200,                  .0500, 2

(/) (!) u z 0 1- l ~ IJJ _J ......-- IJJ u h lJ u 1----,_ / I- 

                                                                             ~
                                                ~

lr-v:- 

                                                     ~ 1\
                                                                   ,,- 'r I-
                                                                                ~           lr--

1/ I ~ vv i\ r\ ~~ ~ I-vI- 1--l/ 1/ v 

                    /
               --:; ~

t:;;: ....... ~.-- 1\. 1-- 

        ~ ~

0 

     ,l                                        1,0                              10,0                              100.0 F R E Q U E NC Y                   CCP S J DESIGN FLOOR RESPONSE SPECTRA                                            Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(B)-15M
                                                                  ~PECTRA - CONTAINMENT BUILDING OBE, NORTH-SOUTH DIRECTION, STEAM GENERATOR UPPER SUPPORT, CALLAWAY SITE

WOLF CREEK DAMPINH VALUES

  • 0100,
  • 0200,
  • 0500, 2

j i 

                                                 .                                                             -~

(f) I (!) i u -* ' - 

                                                              !e---.

z 0 h I- 1 . /  ! < I ,..--  : !l: w ,- j _j w u 7 I --/ f-~ I I r----, 

                                                                        --i\
                                                                     \J r-}1- r-.1 ~I                                  '~

u I'- v I l\ 1.-- I- 

                                                                                    ~~/l I

I/ 

                                                                     \_ v*I-
                            .J:::~
                                                  -                                                    w
                       ~ ~ v-Ir                                                                         ~ 1--
           ... ~ ~

0 ~- 

       ,1                           1,0            -                                10.0 100,0 F R E Q U E NC Y                    CCP S l DESIGN FLOOR RESPONSE SPECTRA
                             -                                                           -                     Re.v.      0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT __

FIGURE 3.7(8)-lSN SPECTRA - CONTAINMENT BUILDING OBE. NORTH-SOUTH DIRECTION, STEAM GENERATOR UPPER SUPPORT, STERLING SITE

WOLF' CRI~EK DAMPING VALUES 

                .0100,  .0200,           .0500,
                                                                                                                   -1 2
                                                --*--    -*---*-;                          --r---*            ,..

I I 

        ---                                     ---  *-- -*--- -*                       *~---- !--*-**** ---   ""
                                                                                                                  --*:.T
                                                -- *--   --*-*- --   --                    -- r------           *-

(f) (!) w -- *-- *-- -- -- 1------- "" II z 

                                                --   ---   *-*  --                         -- 1------     -      ""

0 H 1- 1 0:: w _J r 

                                                                --                                                         ~

w .J I rl I v*t r-- u - r-- u 

        -                                  (;

v---- V-= t1 \ f-Ill 

                                                                                         'n
                                         ~V--                                       ~
                                                                     '1\ r-v
                                -:-,....                                       /~
        -                    _L t-*-

1--* 1-- 

                       -#: ~ v                                                                 ~
           ,.,. ~ ~
     .1
         -                                1.0 10.0                                  100.0 FRE Q UENCY                        CCP S J DESIGN FLOOIR BESPONSE SPECTRA Rev.             0
                                                               -*-*---=~~~~~------*-    WOLF CREEK UPDATED SAFETY ANALY:SIS REPORT FIGURE 3.7(8)-15P SPECTRA - CONTAINMENT BUILDING OBE. NORTH-SOUTH DIRECTION. STEAM GENERATOR UPPER SUPPORT, WOLF CREEK SITE                                    ,

1 

                                                                               *-----*---                           ......- ..- . . . . J

WOLF CREEK DAMPING VALUES 

                                                         --T
                .0100,  .0200,               .0500, n

2 

                                                         -j-....i-.

11 

                                                                                                 -           i
                                                                                                           ***-r l

I  ! 

                                                                                                             ,I 1--*-1                                             - *-
      ---+

(/) (!) l I w --*- 1-* z 0 1-1 

                                                      *-                                                 I   l 0::

w -- *- ...J r \ I w u *- -* u Jr lr 

                                                 ~

I/ J v:-1 < J " r\ 1/ *- 

                                                          "'0,          r/   ir/'

lr-,/'1 

                                          /
                                    ;.... _.../
                                    /
                           -::;; ;/ _,

i 1,-1 v ~ 

                                                                                              \..._
                        ./        /
                                                                                                    ~

7" 0 

           ... ~ ~
      "1                                       1.0                                 10.0                         100.0 FRE Q UENCY                         ( CP S J DESIGN Fl~OO_!!_RESPONSE SPECTRA WOLF CRBBK Uli?DA.TBD SA!'BTY ANALYSIS REPORT FIGURE 3.7(8}-150 SPECTRA - CONTAINMENT BUILDING 013E. EAST-WEST DIRECTION, STEA1'1 GENERATOR UPPER SUPPORT, CALLAWA SITE         .........,_,...,.....

WOLF CHEEK DAMPING VALUES

  • 0100,
  • 0200,
  • 05'00, (J) i I

(!) I I! 1--- 

                                                                      -                    --.        *I-* .

z 

                                                  -    I.,,                                 -          .-

0 1- 1 0:: w _-j It j h r ..J w u I v'[:.. If~ v-I u lJ ~ 1-yv r 1/~ iJ 1r If 

                                                                      \-

IJI-II 

                            ~~--
                                                                       'v-       "'1-.r\
                                                                                          ~

v **- 1-* 

                        ~~
                   ~~                                                                      '  t--

8 """~ - lll 

     .1                           l.D                                            10.0                         100.0 F R E Q UE NC Y                    CC P S l DESIGN FLOOfi RESPONSE SPECTRA_                                            Rev.         0 WOLF CREEK UPDATBD SAFETY ANALYSIS REPORT FIGURE 3.7(8)-15R SPECTRA - CONTAINMENT BUILDING OBE, EAST-WEST DIRECTION. STEAM GENERATOR UPPER SUPPORT, STERLIN SITE
                                                                       *---.......--------*-~~--~-

WOLF CHEEK DAMPING VALUES

  • 0100,
  • 0200,
  • 0500, 2
                                            *--*--     -~-*--*-*                          --.--*
                                             -- *--    --   -*'                        -- f--*
                                             -- *-- *--*- -*   -*                         -- 1----

(f) (!) LJ 

                                             *-- *-    *-f-..                                   -

z 0 I I ...... f--. I 1- 1 

                                                 *-                               -              --      I 0::

w -- *- 

                                                          //          I r--

l 

                                                      /I                               v

__J w r\ 

                                                 ....J u                                              I        *-    I r-,

u 

                                           /

r; __ ~;-

1

_) ~ h r---... IV 11 

                                                                                /
                                 ~~

v "( 

                                                                         "    )
                                                                                         "'\

l 

                           ~

v -- 

                     ~ ~ ... ---                                                                '-- r--
        .... ~~

1.0 

                                             *--  *-                          10.0                                 100.0 F R E Q UE NC Y                        (CP S l DESIGN FL=OQR_.R;SPONSE SPECTRA                                         Rev.           0 WOLF CREEK
                                                              --*-,--~~~~~-------*---.

UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-15T SPECTRA - CONTAINMENT BUILDING OBE, EAST-WEST DIRECTION, STEAM GENERATOR UPPER SUPPORT, WOLF CREEK SITE

WOLF CREEK DAMPING VALUES

  • 0100,
  • 0200,
  • 0500, I
                                                                        *-                             r*

I

I t-* . -

(I) (!) u 2.8 r-- I*- z I~ 11 0 i-r-- *- -.f 0::: IJJ .J IJJ t--- . . .. u u v-- 4' 

                                                                                   ~
                                                         /

I I trlr - f -- *- r--. ~~ 

                                           ~
                                                   ./"

6:::: F~ '- '\. 8 

                 -                        t.e to.o toc.o F R E Q U E N CY                       CC P S l DESIGN FLOOR           ~tESPONSE             SPECTRA Rev.           0 WOLF CREEK tJPDA"l'ED SAPE"l'Y ANALYSIS REPORT FIGURE   3.7(8)~15U SPECTRA - CONTAINMENT BUILDING OBE, VERTICAL DIRECTION, STEAM GENERATOR UPPER SUPPORT, CALl.AWt~

SITE

WOLF CREEK DAMPING VALUES "01 00,

  • 0200,
  • 0500,
                                                          --*...-*                                          -           ***r-
                                                          -- I-      -                                              -I-(/)

(.!) i u 2.0 

                                                          --1--                                             -

z 0 1-1 1- 

                                                          -- - -- 1--*                                     --

0:: lLI _J lLI -- - u u ._,.- ~- jl 

                                                                       ,/
                                                                                   ~.

k. II t-...J 

                                                                                                    }-,               I u-- ,.....
                                                          -- i r                                            -         -:-

6E -"' 

                                                                   ./                              7 T                                                                                         .f                                     ~
                              !:::::!! F""

b::: I=~ ~-/ ~ I-- 0 

         .l
            -,.... Ill='"-

1.0 

                                                          *--                                    10.0                       100.0 F R E Q U E NC Y                      CCP S l DESIGN FLPO.R_RESP'ONSE: SPECTRA Rev.         0
                                                                          -*-*--~~~~~=--*-------. WOLF CREEK UJ>DA1'ED SAFE'l'Y ANALYSIS REPOR'l' FIGURE 3.7(8)-15V SPECTRA - CONTAINMENT BUILDING OBE, VERTICAL DIRECTION, STEAM GENERATOR UPPER SUPPORT, STERLINI SITE
                                                                          --~--*---------*-- ......_                          .. _ 1 1

WOLJI~ CHE:I~K DAMPING VALUES

  • 0100,
  • 0200,
  • 0500,
                                                --*   -*-*   -*                                --r----.       **-*- .. I I

r"l en

  • (.!)

u 2.0 

                                                --*                                                  -        --~-

z 0 1-4 1- 

                                                --.                                                             1-a::

UJ ...J w --. ,_ u u ,-.- 

                                                             -~   ....... r- ""\

< I 

                                                                                 '\
                                                      ~.;- 'L.
                                                                                     .r-""\
                                                                  - r-- r--
                                                --.                                  ...,r--1                    f-
                                          ~*

1'- 1-v -""""" 

                                                      ~
                        =

b:::::: F=~ ~ K r-- I"- 0 

         .I                              1.0
                                                -- .                                10.0                           100.0 F R E Q U E N CY                   ( CP S J DESIGN FL908__RESPONSE SPECTR~

Re. v. 0 WOLF CREEK UPDATED SAFETY ANAI,YSIS REPORT FIGURE 3.7(8)-15X SPECTRA - CONTAINMENT BUILDING OBE, VERTICAL DIRECTION. STEAM GENERATOR UPPER SUPPORT. WOLF CREEK SITE _ _ ,_ _ _ _ _ _ _ _ _ _ _ _ _ _ ...... . _ ** _ 1 1 _ 1 1 1 . . . .,.

WOLF CREEK 

                        ~ ,...~ REACTOR BUILDING EL. 2206'-6"
                                                                                               ~,.

EL. 2170'-9" I r I EL. 2135' -0" f POLAR CRANE ~ EL. 2119'-0" fiNTERNAL C: I STRUCTURE 

                                                         ~STEAM                                                     EL. 2100'-0" I

EL. 2090'-4" GEN 

                                        <IIIII.            I
                                             "                                                                      EL. 2080' -0" EL. 2083'-11"                                       ~         ~-             11                         ~SHELL I

EL. 2060'-0" <Ill EL. 2056'-6" EL. 2051 '-2" EL. 2047'-6" ...... 

                                                             ~-~                                                    EL. 2039'-0" EL. 2034'-0"                             [i EL.

EL. 2025'-9" 2022' -6" 

                     ~

11 ~ 

                                                                          ,.        ~

EL. 2028' -0" EL. 2013'-5" EL. 2012'-0" ,, & I.a. 

                                        ~

(5. ~ *~ 

                                                                                        ~
                                          -*~

t...,...o a; EL. 2000'-0" 

                                                                   ~
                                                                                                   ,/)

EL. 1998'-11" '6** 

                                                                                                            ~TENDON GALLER y
                            ~
                    '.I I
                                                                                       .CSJ
                                                                                       ,.,:.,.                      El. 1974'-3"  Re v. 0
                             ***"                                                                              WOLF CREEK UPDATED S AFETY ANALYSIS REP ORT FIGURE 3.7(8)-17 LUMPED-MASS/FLUSH MODEL, CONTAINMENT BUILDING

WOLF CREEK EL. 2106'-6" EL. 2083'-6" EL. 2047'-6" EL. 2026' -0" EL. 2000'-0" Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-18 LUMPED-MASS/FLUSH MODEL, FUEL BUILDING

WOLF CREEK 

>-      _J a:::
  • 0 .

<( (.') a::: (.') 

-     Q 1- Q

_J_J z_j xOJ om <( u El. 21 02'-6" El. 2090'-0" El. 2087'-2" El. 2073'-6" El. 2065'-0" El. 2047'-6" EL 2032'-0" El. 2026'-0 El. 2016'-0" El. 2000'-0" El. 1988'-0" El. 1984'-0" El. 1974'-0" Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8}-19 LUMPED-MASS/FLUSH MODEL, AUXILIARY/CONTROL BUILDING

WOLF CREEK EL. 2066'-0" I I EL. 2027'-6 EL. 2000'-0" Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-20 LUMPED-MASS/FLUSH MODEL, DIESEL GENERATOR BUILDING

WOLF CREEK 

      ~2 0         0 48 c

0 44 0 0 0 c 

                                                                     ....                      c 40 0

(!) 0 c 0 0 z 36 0 0 ii: 0 0 0 0

lE 0 0 0 0

< 8 0 0 ..J 32 - c 0 0 u 0 go 0 0 0 E a: 28 0 0 u 0 0 0 c 00 

                                                                          ~             oc                 0 LL.            00       ,o   0                    0              )

0 0 24 ~ 0 0 c 0 ~u 0 00 0 05 0 I 0 0 0 m 20 0 r,7o v 0 16 oo, (l

/0 12
               ~

8 0 oc ~ 7°/o oo 0 .... 4 00 0 0.1 0.2 o.3 0.4 o.~ o.e 0.1 o.e o.a 1.0 1.1 1.2 1.3 INPUT (g) BRACED HANGERS Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-21 DAMPING VS. INPUT LEVEL FOR BRACED HANGER SYSTEMS

WOLF CREEK 24 I I I I I 50°/o TO FULLY LOADED TR AY D f A 20 v M p I 16 12 

              /_

v ~ 

                                         '-UNU JADED TRA' 8 N    8                                ./
-..L-7°/o ~--

G CONI UlT 4 (O/o) 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 INPUT FLOOR SPECTRUM ZPA Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-22 LOWER BOUND DAMPING AS A FUNCTION OF INPUT ZPA

WOLF CREEK APPENDIX 3.7(B)A IMPEDANCE FUNCTIONS FOR A RIGID CIRCULAR FOUNDATION ON A LAYERED VISCOELASTIC MEDIUM A.1 FORMULATION OF THE PROBLEM A.1.1 Statement of the Problem In what follows, a study is made of the forced harmonic vibrations of a rigid circular footing of radius a placed on the surface of a layered viscoelastic medium. The layered medium consists of N-1 parallel layers resting on a viscoelastic half-space. Both the layers and the elastic half-space are assumed to be homogeneous and isotropic with densities ri, shear moduli Gi, and Poisson's ratios si (i = 1, 2,...N), respectively. In addition, depending on the type of internal friction considered, the relative viscosity coefficient (G1'/Gi) (for Voigt type dissipation) or the hysteretic damping coefficient ei = G1'/2Gi (for hysteretic type dissipation) are assumed to be known for each one of the media forming the soil deposit. The geometry of the model and the coordinate systems used are shown in Figure 3.7(B)A-1. A welded type of contact is assumed to exist between adjacent layers. Thus, the stresses and displacements are continuous across each interface. The contact between the foundation and the surface of the top layer is assumed to be relaxed, i.e., the contact is frictionless for vertical and rocking vibrations and pressureless for horizontal vibrations. The boundary conditions at z = 0 expressed in terms of displacement and stress components in cylindrical coordinates are the following:

a. Vertical Vibrations uz (r, q, 0) = Dveiwt 0 < r < a (A-1.a) szz(r, q, 0) = 0 r > a (A-1.b) szr(r, q, 0) = szq(r,w,0) = 0 0 < r < infin (A-2) 3.7(B)A-1 Rev. 1

WOLF CREEK

b. Rocking Vibrations uz(r, q, 0) = ar cosqeiwt 0 < r < a (A-3.a) szz(r, q, 0) = 0 r > a (A-3.b) szr(r, q, 0) = szq(r, q, 0) = 0 0 < r < infin (A-4)
c. Horizontal Vibrations ur(r, q, 0) = DH cosq eiwt 0 < r < a (A-5) uq(r, q, 0) = - DH sinq eiwt szr(r, q, 0) = szq(r, q, 0) = 0 r > a (A-6) szz(r, q, 0) = 0 0 < r < infin (A-7)

In the equations above, Dv is the amplitude of the vertical displacement of the center of the rigid foundation, is the amplitude of the rocking angle about the y-axis (é = p/2), H is the amplitude of the horizontal displacement of the foundation in the direction of the x-axis (q = 0), and w is the frequency of the steady-state vibrations. The continuity conditions at the interface z = Hi are: i i+1 ur (r, q, Hi) = ur (r, q, Hi) (A-8.a) i i+1 uq (r, q, Hi) = ue (r, q, Hi) (A-8.b) i i+1 uz (r, q, Hi) = uz (r, q, Hi), (i=1, 2, ...N) (A-8.c) i i+1 uzr (r, q, Hi) = sr (r, q, Hi), (A-9.a) 3.7(B)A-2 Rev. 1

WOLF CREEK i i+1 szq (r, q, Hi) = szq (r, q, Hi), (A-9.b) i i+1 szz (r, q, Hi) = szz (r, q, Hi), (i=1, 2, ...N) (A-9.c) where the superscript i indicates the ith layer. In addition, the displacement and stress components in the underlying half-space must tend to zero as (r2 + z2) tends to infinity. A.1.2 Types of Energy Dissipation In this study, two types of energy dissipation are considered, namely the Voigt viscous model and the hysteretic model. The stress-strain relationships for harmonic vibrations of a solid with Voigt type damping are of the form (Ref. A-1) szz = ( l + iwl')q + 2(m + i w m')ezz (A-10.a) szx = 2(m + i w m') exz (A-10.b) where q = exx + eyy + ezz (A-10.c) In equations (A-10.a) and (A-10.b), w is the frequency of the excitation, l and m are Lame's constants, and l ', m' are the viscosities. It is clear from equations (A-10.a) and (A-10.b) that the viscoelastic problem may be solved if the solution of the corresponding purely elastic problem is known by substituting in the elastic solution l and m by the complex moduli l* = (1 + i l'/ l) (A-11.a) m* = m(1 + i m'/m) (A-11.b) In order to simplify the problem, it is assumed that l ' m' (A-12) ____ = ____ l m 3.7(B)A-3 Rev. 1

WOLF CREEK In this case, the remaining complex constants are given by: (3 l * + 2m*)m* E* = *l + m* E(1 + iwm'/m) (A-13.a) 2 k* = l * + 3 m * = k(1 + iwm'/m) (A-13.b) 3 l

  • s* = 2( l * + m*) = s (A-13.c) where E, k, and w are the Young's modulus, the bulk modulus, and Poisson's ratio, respectively. The assumption given by equation (A-12) has the advantage that the Poisson's ratio for the viscoelastic medium is real and equal to the Poisson's ratio of the corresponding elastic medium. One disadvantage, however, is the fact that the bulk modulus is complex, and consequently there are losses associated with changes of volume.

Equation (A-10.b) indicates that for shear deformations the stress-strain relationship could be described by an ellipse. The energy loss per cycle is given by the area of the ellipse and the corresponding 'specific loss' is DW wm' W = 2p m (A-14) where W is the elastic energy stored when the strain is a maximum. Equation (A-14) indicates that for a Voigt solid the "specific loss," or the energy loss per cycle, is proportional to the frequency of the excitation. The elliptical stress-strain loop in this case is a direct result of the viscosity of the medium. Laboratory tests on soils indicate that the "specific loss" D W/W is independent of the frequency of the excitation and that the stress-strain loop is not an ellipse (Ref. A A-6). It appears then that the mechanism of energy loss in soils is not of the viscous type but rather is a direct result of the anelastic behavior of soils. In spite of this anelastic behavior, an approximate approach is to assume that the soil may be treated in a similar way as a viscoelastic medium, except that in this case the complex shear modulus

  • and the "specific loss" are taken to be equal to m* = m(1 + 2ie ) (A-15)

DW W = 4pe (A-16) 3.7(B)A-4 Rev. 1

WOLF CREEK where e is a damping constant independent of frequency. This model of internal damping is also called constant hysteretic type damping. The damping constant e is analogous to the percentage of critical damping under resonant conditions, or during free vibrations (Ref. A-3). The hysteretic damping constant is strain dependent: values for low strain may be less than 0.02, while for high strains e may reach values of 0.15 or 0.20. In what follows, the shear modulus m is designated by G, and the shear viscosity m' is designated by G'. A.1.3 Integral Representation A solution of the equations of motion in cylindrical coordinates satisfying the conditions at the interface between layers, as well as the conditions at infinity, may be obtained by application of the correspondence principle to a representation derived by Sezawa and reported in references A-7 and A-8. The displacement and stress components of interest on z = 0 are given by ur (r, q, 0) = a ur (r')cos(nq) uq (r, q, 0) = a uq (r')sin(nq) (A-17) uz (r, q, 0) = a uz (r')cos(nq) szr(r, q, 0) = Glszr (r')cos(nq) szq(r, q, 0) = Glszq (r')sin(nq) (A-18) dzz(r, q, 0) = Glszz (r')cos(nq) where n = 0 for vertical vibrations, n = 1 for rocking and horizontal vibrations, r' = r/a, and infin ur(r') + uq(r') = + 2 ' § D11(k)C1(k) + D12(k)C2(k) + D33C3(k)* 

                           µ¨©k            DR                 DL   ¸
                                                                   ¹
                           ¶ 0
                  . J n+1 (a kr')dk                          (A-19) o 3.7(B)A-5                  Rev. 1

WOLF CREEK infin

  • D (k)C1(k) + D22(k)C2(k) uz(r') = + 2 'k 21 Jn (aokr')dk (A-20)
              ¶              DR 0
*
  • infin szr(r') + szq(r') = +2ao ' [kC1(k) + C3(k)] Jn+1(aokr')dk (A-21)
                          ¶ o
  • infin szz(r') = 2ao 'kC2(k)Jn(aokr')dk (A-22)
                ¶ o

In equations (A-19) - (A-22), ao = wa/b1 is a dimensionless frequency defined in terms of the shear wave velocity b1 of the top layer. The functions Dij (i,j = 1,2), bR, b33, and bL appearing in equations (A-19) - (A-22) depend on the properties of the soil column, and are given in Appendix 3.7(B).B. The functions C1(k), C2(k), and C3(k) are to be determined by the boundary conditions on z = 0. The term Jn(aokr') is an infinite series known as the Bessel function of the first kind of order n while the term Jn+1(aokr') is of the order n+1. For vertical and rocking vibrations, equations (A-2) and (A-4) together with equation (A-21) imply that C1(k) = C3(k) = 0. (A-23) Similarly, for horizontal vibrations, equations (A-7) and (A-22) imply that C2(k) = 0. (A-24) Before imposing the remaining boundary conditions, it is convenient to introduce the following substitutions (Ref. A-7, A- 9)

a. Vertical Vibrations 2

D vK1 infin C2(k) = -pa(1-s ) ao 'fv(t)cos(aokt)dt (A-25) 1 ¶ 0 3.7(B)A-6 Rev. 1

WOLF CREEK

b. Rocking Vibrations
               § 2aK2 1
  • 1 C2(k) = -¨p(1-a ) ao¸ ' fR(t)sin(aokt)dt (A-26)
               ©      1    ¹   ¶ 0
c. Horizontal Vibrations 2DHK2 1
                           º   1 C1(k) = <<pa(2-s1) ao>>      '{-f1(t)cos(aokt)
             ¬             1/4   ¶ 0

sin(aokt)º 

           - f2(t)<<cos(aokt) -     aokt   >>}  dt             (A-27)
                    ¬                      1/4 2DHK21
                             º    1 C3(k) = - pa(s-s ) aok>>
                 <<                   f (t)cos(aokt)
                 ¬      1    1/4    ¶{ 1 0
             - (1-s1)f2(t)[cos(aokt)
             - sin(aokt)/aokt]} dt                            (A-28) where fV(t), fR(t), and f1(t), f2(t) are functions to be determined by 2

equations (A-1), (A-3), and (A-5), respectively. Also,K1 = (1 + i wG1'/G1)-1 2 for Voigt-type damping and K1 = (1 + 2ie1)-1 for hysteretic-type damping. The substitutions indicated above satisfy directly the stress boundary conditions prescribed in equations (A-1), (A-3), and (A-6). A.2 INTEGRAL EQUATIONS AND IMPEDANCE FUNCTIONS Substitution from equations (A-25) - (A-28), together with equations (A-23) and (A-24), into equations (A-17), (A-19), and (A-20), and imposition of the remaining displacement boundary conditions leads to the following integral equations for the unknown functions fV(t), fR(t), and f2(t)

a. Vertical Vibrations 1

fV(t) + 'K(t,t')fV(t')dt' = 1 (0 < t < 1) (A-29) 

            ¶ 0

where K(t,t') = L1(lt - t'l )+ L1(t + t') (A-30) 3.7(B)A-7 Rev. 1

WOLF CREEK infin ao KD22 L1(t)=- p '§ + 1*cos (aokt)dk (A-31) 

             ¨
           µ (1-s )D K2         ¸
           ¶©       1 R 1       ¹ 0
b. Rocking Vibrations 1

fR(t) + 'K(t,t')fR(t')dt' = t (0 < t < 1) (A-32) 

           ¶ 0

where K(t,t') = L1(lt - t'l )- L1(t + t') (A-33) The function L1(t) in equation (A-33) is defined by equation (A-31).

c. Horizontal Vibrations 1

f1(t) + '[K11(t,t')f1(t') + K12(t,t')f2(t')] dt' = 0 

       ¶ 0

(0 < t < 1) (A-34) 1 (1 - s1)f2(t) + ' [K (t,t')s1(t')+K22(t,t')f2(t')]dt'=0 

                    ¶ 21 0

(0 < t < 1) (A-35) where 2ao 1 infin K11(t,t')= p §2-s * '((1-s1)H1(k)+H2(k)) cos(aokt)cos(aokt')dk 

               ©    1¹ ¶ 0

(A-36) 2ao §1-s1*infin §cos(a kt') - sin(aokt')*dk K12(t,t') = - ¨ ¸ p ©2-s1¹ ¶ ' [H1(k) - H2(k)]cos(a okt) 

                                                          ©     o           aokt ¹ 0

(A-37) 3.7(B)A-8 Rev. 1

WOLF CREEK 2ao 1-s1 infin sin(aokt) K21(t,t')= p §2-s1* '[H1(k)- H2(k)]§¨cos(aokt)- a kt *¸cos(aokt')dk 

            ©    ¹ ¶                ©               o     ¹ 0

(A-38) 2ao 1-s1 infin K22(t,t')= ? §2-s1* '[H1(k)+ (1-s1)H2(k)] 

            ©    ¹ ¶ 0
                    §cos(a kt) - -sin(aokt)*
                    ¨     o         aokt    ¸
                    ©                       ¹
                    §cos(a kt) - -sin(aokt)
  • dk
                    ¨     o         aokt    ¸
                    ©                       ¹ k        D11 H1(k) =   2             DR - 1                (A-40)

K1 (1-s1) kD33 H2(k) = 2 - 1 (A-41) K1DL The integral equations (A-29), (A-32), (A-34), and (A-35) are of the Fredholm type and have a form suitable for numerical solution. Once these integral equations have been solved, the entire displacement and stress field may be evaluated by substitution from equations (A-25) - (A-28) into equations (A-19) - (A-22). In particular, the total vertical load V, the rocking moment about the y-axis M, and the total horizontal load in the x- direction H may be found to be given by 4G1aDveiwt 1 V = 2 ' 

               ¶f(t)dt                                        (A-42)

(1-s1)k1 0 8G1a3aeiwt 1 M = ' tf dt (A-43) 2 (1-s1)k1 ¶ R 0 3.7(B)A-9 Rev. 1

WOLF CREEK 4G1aDveiwt 1 H = 2 ¶ 'f1(t)dt (A-44) (2-s1)k1 0 Equations (A-42), (A-43), and (A-44) constitute the force- displacement relationship for the circular foundation. It should be mentioned that in deriving these equations, the terms coupling the horizontal and rocking vibrations have been neglected. It is convenient to write equations (A-42) - (A-44) in the following form: 4G1a V = 1-s [kVV(ao) + iaocVV(ao)] Dveiwt (A-45) 1 8G1a3 M = 3(1-s ) [kMM(ao) + iaocMM(ao)]aeiwt (A-46) 1 2G1a H = 2-s [kHH(ao) + iaocHH(ao)] DHeiwt (A-47) 1

where, 1

fv(t) kVV(ao) = 'Re§ 2 *dt, 

               µ ¨ K ¸
               ¶ © 1 ¹ 0

1 1 ' §fv(t)* cVV(ao) = a Im dt (A-48) o µ ¨ K2 ¸ 

                  ¶ ©       1  ¹ 0

1 tfR(t)* kMM(ao) = 3 'Re§ 2 ¸dt 

                 µ    ¨    K1
                 ¶ ©           ¹ 0

1 3 ' §tfR(t)* cMM(ao) = a Im dt (A-49) o µ ¨ K2 ¸ 

                  ¶ ©        1   ¹ 0

1 f1(t) kHH(ao) = 'Re§ 2 *dt 

               µ    ¨    K
                             ¸
               ¶ © 1 ¹ 0

3.7(B)A-10 Rev. 1

WOLF CREEK 1 

                ?1(t) cHH(ao) = a 'Im§ 2 *dt 1

o µ ¨ K ¸ 

             ¶ © 1 ¹ 0

(A-50) The terms inside the square brackets in equations (A-45), (A-46), and (A-47) are the normalized impedance functions for vertical, rocking, and horizontal vibrations; the factors outside the parentheses correspond to the static values (ao=0) of the impedance functions for an elastic half-space having the properties of the top layer. The functions kVV(ao), kMM(ao), and kHH(ao), corresponding to the real part, Re, of the impedance functions, will be called here stiffness coefficients, while the functions cVV(ao), cMM(ao), and cHH(ao), proportional to the imaginary part, Im, of the impedance functions, will be designed here as damping coefficients. Both the stiffness and damping coefficients are functions not only of the dimensionless frequency ao but also depend on the properties of the different media forming the soil column. In solving the problem of the horizontal vibrations, a further approximation has been introduced by assuming that 2(t) is sufficiently small so that the integral equations (A-34) and (A- 35) may be reduced to 1 ~ ~ f1(t)+' 

     ¶K11(t,t')f1(t')dt' = 1      (0 < t < 1)  (A-51) 0 where the kernal K11(t,t') is given by equation (A-36). The basis for this approximation is that for the case of a uniform half- space, the function f2(t) is much smaller than f1(t), in particular, for the static case f2(t) = 0. The above approximation is equivalent to the requirement that zy = 0 under the foundation and thus corresponds to a further relaxation of the boundary conditions.

A.3 NUMERICAL SOLUTION The numerical procedure used to solve the integral equations (A- 29), (A-32), and (A-51) consists of reducing these equations to a system of algebraic equations that are solved by standard methods. A key step in this procedure is the evaluation of the kernels K(t, t') given by equations (A-30), (A-33), and (A-36). 3.7(B)A-11 Rev. 1

WOLF CREEK In the case of a medium with no internal friction, the functions R and L have zeros for real values of k and, consequently, the integrands in equations (A-31) and (A-36) are singular at these points. This situation complicates the numerical evaluation of the kernels. However, if there is internal friction then the zeros of R and L are complex, and consequently the numerical evaluation of the kernels is simplified. The kernels are evaluated numerically by use of Filon's method of integration up to a sufficiently large value of k, the rest is evaluated analytically by using the asymptotic forms of the integrands for large k. A.4 REFERENCES A-1 Kolsky, H., Stress Waves in Solids, Dover Publications, Inc., New York, 1963. A-2 Richart, F. R., Hall, J. R., and Woods, R. D., Vibration of Soils and Foundations, Prentice-Hall, Inc., New Jersey, 1970. A-3 Dobry, R., "Damping in Soils: Its Hysteretic Nature and the Linear Approximation," Research Report R70-14, Massachusetts Institute of Technology, Department of Civil Engineering, Cambridge, Mass., 1970. A-4 Krizek, R. J., and Franklin, A. G., "Energy Dissipation in Soft Clay," Proc. Int. Symp. on Wave Propagation and Dynamic Properties of Earth Materials, University of New Mexico Press, Albuquerque, 1967, pp. 797-807. A-5 Seed, H. B., and Idriss, I. M., "Soil Moduli and Damping Factors for Dynamic Response Analysis," Report EERC 70-10, University of California, Berkeley, 1970. A-6 Hardin, B. O., and Drnevich, V. P., "Shear Modulus and Damping in Soils: Measurements and Parameter Effects," Proc. Am. Soc. Civ. Engs., Vol. 98, No. SM6, 1972, pp. 603-624. A-7 Luco, J. E., and Westmann, R. A., "Dynamic Response of Circular Footings," Journal of the Engineering Mechs. Div., ASCE, Vol. 97, No. EM5, 1971, pp. 1381-1395. A-8 Bycroft, G. N., "Forced Vibrations of a Rigid Circular Plate on a Semi-Infinite Elastic Space or on an Elastic Stratum," Phil. Trans., Royal Soc. of London, Vol. 248, 1956, pp. 327-368. A-9 Luco, J. E., "Impendance Functions for a Rigid Foundation on a Layered Medium", Nuclear Engineering and Design, 1974. 3.7(B)A-12 Rev. 22

WOLF CREEK \ z Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)A-1

  • DESCRIPTION OF THE MODEL

WOLF CREEK APPENDIX 3.7(B)B SOIL DEPENDENT DISPLACEMENT FUNCTIONS FOR THE SOLUTION OF THE EQUATIONS OF MOTION The functions ij(k) (i,j = 1, 2) and R(k) entering in equations (A-

19) and (A-20) are defined by
 é 11(k) 12(k)  = T* A + T* B adj  T* A + T* B 
                    ú     11       12 ø        21      22 ø    (B-1) 21(k) 22(k) û
and,
           = det T21A + T22Bø R                                                        (B-2) where the matrices [A] and [B] are given by
                é -k v'

[A] = N ú (B-3) VN -k ú 

                           û
                   *           é      -2vNk       2k2 - K2 Nø

[B] = GN 

                                                            ú    (B-4)

G1 

                                - 2K2 - KNø 2         '     ú 2vNk    û and T*ij (i, j = 1, 2) are the submatrices of the total transfer matrix T* associated with the set of layers overlying the base half-space. The total transfer matrix T*
         éT11 l T12

[T*] = * * ú (B-5) T21 l T22 û may be obtained in terms of the transfer matrices for each layer Tj (j = 1, N-1) by means of the following product: [T*] = [T1][T2].....[Tj]....[TN-1] (B-6) The transfer matrix for the jth layer is in turn given by 

         é j       j T11 l T12

[Tj] = j j ú (B-7) T21 l T22 û 3.7(B)B-1 Rev. 0

WOLF CREEK

where, 2 2 '

j 1 

          é -2k2CHj + (2k2 - Kj)CHPj -k(2k2 - kj)SHj + 2kvjSHPj T11 = -   2                                                                 ú kj 2kv2SH - k(2k2 - K2)SHP                  2 - K2)CH - 2k2CHP j j                     j      j   (2k         j  j         j û
                         '2 1é
            -k2SHj + vj SHPj         k(CHj - CHPj) j j                                              ú T12 = -                                                                       (B-8) 2 k(CHj - CHPj) -vjSHj + K SHPj û    2 2                    2                        2 j       1 i
              é -4vjk2SHj + (2k2 - Kj)SHPj            -2K(2k2 - Kj)(CHj -CHPj)

T21 = - 4ç ÷ ú Kjlø -2k(2k2 - K2)(CH - CHP 2 '2 j j j) -(2k2 - Kj)2SHj + 4vj k2SHPj û 2 2 2 j 1 

           é -2k2CHj + (2k2 - Kj)CHPj              -2kvjSHj - k(2k2 - Kj)SHPj T22 = -   2               2            '2                     2
                                                                                 ú Kj
            -k(2k2 - Kj)SHj + 2vj k SHPj            (2k2 - Kj)CHj - 2k2CHPj     û The different terms entering in equations (B-3) to (B-8) are defined by 2 2 1/2                                    2 1/2 vj = k2 - jKjø                           vj = k2 - kjø 2    (1 - 2j)                               2     G1j j =       2       (1-j)                    kj  =
  • Gj1 i=Gj÷
*      ç
  • Gj = Gj ç1 + G ÷, or, Gj = Gj (1 + 2ij) j ø (aovjj) (aovjj)

SHj = sinh vj SHPj = sinh ' (B-9) vj CHj = cosh (aovjj) CHPj = cosh (aovjj) hj j = a 

    =a ao =

1 3.7(B)B-2 Rev. 0

WOLF CREEK where j, j, Gj, Gj/Gj, and hj, respectively, are the Poissons ratio, density, shear modulus, relative viscosity, and thicknessof the jth layer. In the last two equations of (B-9), a is theradius of the circular foundation, a is the frequency of the steady-state vibrations, and is the shear wave velocity of the top layer. The first form of Gj corresponds to the Voigt-type damping, while the second corresponds to the hysteretic-type damping, j being the hysteretic dampingconstant for the jth layer. The functions 33(k) and L(k)entering in equation (A19) are defined by 

                 *
  • vNGN 33(k) = L11 + L12 G (B-10) 1
               *
  • vNGN L(k) = L21 + L22 G (B-11) 1 where L* (i, j = 1,2) are the elements of the transfer matrix L*. The transfer matrix L
  • L11 + L 12

[L*] = (B-12) L21 + L22 is defined in terms of the transfer matrices for each layer by [L*] = [L1] . [L2]....[Lj]...[LN-1] (B-13) in which, G1 CHPj

  • SHPj

[Lj] =

  • Gj (B-14)

GGj 2 1 Vj SHPj CHPj 3.7(B)B-3 Rev. 0

WOLF CREEK 3.7(N) SEISMIC DESIGN For the OBE loading condition, the nuclear steam supply system is designed to be capable of continued safe operation. The design for the SSE is intended to ensure:

a. That the integrity of the reactor coolant pressure boundary is not compromised;
b. That the capability to shut down the reactor and maintain it in a safe condition is not compromised; and
c. That the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of 10 CFR 50.67 is not compromised.

It is necessary to ensure that required critical structures and components do not lose their capability to perform their safety function. Not all critical components have the same functional safety requirements. For example, a safety injection pump must retain its capability to function normally during the SSE. Therefore, the deformation in the pump must be restricted to appropriate limits in order to ensure its ability to function. On the other hand, many components can experience significant permanent deformation without loss of function. Piping and vessels are examples of the latter where the principal requirement is that they retain their contents and allow fluid flow. The seismic requirements for safety-related instrumentation and electrical equipment are covered in Sections 3.10(N) and (B). The safety class definitions, classification lists, operating condition categories, and the methods used for seismic qualification of mechanical equipment are given in Section 3.2. 3.7(N).1 SEISMIC INPUT 3.7(N).1.1 Design Response Spectra Refer to Section 3.7(B).1.1. 3.7(N).1.2 Design Time History Refer to Section 3.7(B).1.2. 3.7(N)-1 Rev. 34

WOLF CREEK 3.7(N).1.3 Critical Damping Values The damping values given in Table 3.7(N)-1 are used for the systems analysis of Westinghouse equipment. These are consistent with the damping values recommended in Regulatory Guide 1.61, except in the case of the primary coolant loop system components and large piping (excluding reactor pressure vessel internals) for which the damping values of 2 percent and 4 percent are used as established in testing programs reported in Reference 1. The damping values for control rod drive mechanisms (CRDMs) and the fuel assemblies of the nuclear steam supply system, when used in seismic system analysis, are in conformance with the values for welded and/or bolted steel structures (as appropriate) listed in Regulatory Guide 1.61. Tests on fuel assembly bundles justified conservative component damping values of 7 percent for OBE and 10 percent for SSE to be used in the fuel assembly component qualification. Documentation of the fuel assembly tests is found in Reference 2. The damping values used in component analysis of CRDMs and their seismic supports were developed by testing programs performed by Westinghouse. These tests were performed during the design of the CRDM support; the support was designed so that the damping in Table 3.7(N)-1 could be conservatively used in the seismic analysis. The CRDM support system is designed with plates at the top of the mechanism and gaps between mechanisms. These are encircled by a box section frame which is attached by tie rods to the refueling cavity wall. The test conducted was on a full-size CRDM complete with rod position indicator coils, attachment to a simulated vessel head, and variable gap between the top of the pressure housing support plate and a rigid bumper representing the support. The internal pressure of the CRDM was 2,250 psi, and the temperature on the outside of the pressure housing was 400 °F. The program consisted of transient vibration tests in which the CRDM was deflected a specified initial amount and suddenly released. A logarithmic decrement analysis of the decaying transient provides the effective damping of the assembly. The effect on damping of variations in the drive shaft axial position, upper seismic support clearance, and initial deflection amplitude was investigated. The upper support clearance had the largest effect on the CRDM damping, with the damping increasing with increasing clearance. With an upper clearance of 0.06 inch, the measured damping was approximately 8 percent. The clearance in a typical upper seismic CRDM support is a minimum of 0.10 inch. The increasing damping 3.7(N)-2 Rev. 0

WOLF CREEK with increasing clearances trend from the test results indicated that the damping would be greater than 8 percent for both the OBE and the SSE, based on a comparison between typical deflections during these seismic events to the initial deflections of the mechanisms in the test. Component damping values of 5 percent are, therefore, conservative for both the OBE and the SSE. These damping values are used and applied to CRDM component analysis by response spectra techniques. 3.7(N).1.4 Supporting Media for Seismic Category I Structures Refer to Section 3.7(B).1.4. 3.7(N).2 SEISMIC SYSTEM ANALYSIS This section describes the methods of seismic analysis performed for safety-related components and systems within Westinghouses scope. 3.7(N).2.1 Seismic Analysis Methods Those components and systems that must remain functional in the event of the SSE (seismic Category I) are identified by applying the criteria of Section 3.2.1. In general, the dynamic analyses are performed, using a modal analysis plus either the response spectrum analysis or integration of the uncoupled modal equations as described in Sections 3.7(N).2.1.3 and 3.7(N).2.1.4, respectively, or by direct integration of the coupled differential equations of motion described in Section 3.7(N).2.1.5. 3.7(N).2.1.1 Dynamic Analysis - Mathematical Model The first step in any dynamic analysis is to model the structure or component, i.e., convert the real structure or component into a system of masses, springs, and dashpots suitable for mathematical analysis. The essence of this step is to select a model so that the displacements obtained will be a good representation of the motion of the structure or component. Stated differently, the true inertia forces should not be altered so as to appreciably affect the internal stresses in the structure or component. Some typical modeling techniques are presented in Reference 3. 3.7(N)-3 Rev. 10

WOLF CREEK Equations of Motion Consider the multidegree of freedom system shown in Figure 3.7(N)- 1. Making a force balance on each mass point r, the equations of motion can be written in the form: i criui + kriui = 0 

               ..         . i m ry r +                                  [3.7(N)-1]

where: mr = the value of the mass or mass moment of rotational inertia at mass point r y r = absolute translational or angular acceleration of mass point r cri = damping coefficient - external force or moment required at mass point r to produce a unit transla-tional or angular velocity at mass point i, main-taining zero translational or angular velocity at all other mass points. Force or moment is positive in the direction of positive translational or angu-lar velocity ui = translational or angular velocity of mass point i relative to the base kri= stiffness coefficient - the external force (moment) required at mass point r to produce a unit deflec-tion (rotation) at mass point i, maintaining zero displacement (rotation) at all other mass points Force (moment) is positive in the direction of posi-tive displacement (rotation) ui = displacement (rotation) of mass point i relative to the base As an example, note that Figure 3.7(N)-1 does not attempt to show all of the springs (and none of the dashpots) which are represented in Equation 3.7(N)-1. 3.7(N)-4 Rev. 0

WOLF CREEK Since: yr = ur ys [3.7(N)-2] where: ys = absolute translational (angular) acceleration of the base ur = translational (angular) acceleration of mass point r relative to the base Equation 3.7(N)-1 can be written as: i 

 ..            . i             ..

mru r + criui + kriui = -mr y s [3.7(N)-3] For a single degree of freedom system with displacement u, mass m, damping c, and stiffness k, the corresponding equation of motion is: mu + cu + ku = y s [3.7(N)-4] 3.7(N).2.1.2 Modal Analysis Natural Frequencies and Mode Shapes The first step in the modal analysis method is to establish the normal modes, which are determined by eigen solution of Equation 3.7(N)-3. The right hand side and the damping term as set equal to zero for this purpose, as illustrated in Reference 4 (Pages 83 through 111). Thus, Equation 3.7(N)-3 becomes: 

 .. i mru r + kriui = 0                                     [3.7(N)-5]

The equation given for each mass point r in Equation 3.7(N)-5 can be written as a system of equations in matrix form as: [M] { } + [K] {} = 0 [3.7(N)-6] where: [M] = mass and rotational inertia matrix {} = column matrix of the general displacement and rotation at each mass point relative to the base 3.7(N)-5 Rev. 0

WOLF CREEK [K] = square stiffness matrix { } = column matrix of general translational and angular accelerations at each mass point relative to the base, d2 {}/dt2 Harmonic motion is assumed, and the {D} is expressed as: {} = {} sin = t [3.7(N)-7] where: {} = column matrix of the spatial displacement and rotation at each mass point relative to the base 

        =  = natural frequency of harmonic motion in radians per second The displacement function and its second derivative are substi-tuted into Equation 3.7(N)-6 and yield:

[K] {} = =2 [M] {} 3.7(N)-8] The determinant [K] - =2 [M] l is set equal to zero and is then solved for the natural frequencies. The associated mode shapes are then obtained from Equation 3.7(N)-8. This yields n natural frequencies and mode shapes where n equals the number of dynamic degrees of freedom of the system. The mode shapes are all orthogonal to each other and are sometimes referred to as normal mode vibrations. For a single degree of freedom system, the stiffness matrix and mass matrix are single terms and the deter-minant l[K] - =2 [M] l when set equal to zero yields simply: k - =2m = 0 or: [3.7(N)-9] k 

                = =   m where w is the natural angular frequency in radians per second.

The natural frequency in cycles per second is, therefore: 1 k f = m [3.7(N)-10] 2 To find the mode shapes, the natural frequency corresponding to a particular mode, =n can be substituted in Equation 3.7(N)-8. 3.7(N)-6 Rev. 0

WOLF CREEK Modal Equations The response of a structure or component is always some combination of its normal modes. Good accuracy can usually be obtained by using only the first few modes of vibration. In the normal mode method, the mode shapes are used as principal coordinates to reduce the equations of motion to a set of uncoupled differential equations that describe the motion of each mode n. These equations may be written as (Ref. 4, Pages 116 through 125): 2 .. An + 2anpnAn + anAn = -ny s [3.7(N)-11] where the modal displacement or rotation, A n, is related to the displacement or rotation of mass point r in mode n, u rn, by the equation: urn = Anrn [3.7(N)-12] where: an = natural frequency of mode n in radians per second pn = critical damping ratio of mode n n = modal participation factor of mode n given by: n mrrn n = [3.7(N)-13] n mr2rn where: rn = value of rn in the direction of the earthquake The essence of the modal analysis lies in the fact that Equation 3.7(N)-11 is analogous to the equation of motion for a single degree of freedom system that will be developed from Equation 3.7(N)-4. Dividing Equation 3.7(N)-4 by m gives: 

        .. c . k   ..

u + m u + mu = -y s [3.7(N)-14] The critical damping ratio of the single degree of freedom system, p, is defined by the equation: 3.7(N)-7 Rev. 0

WOLF CREEK c p = c [3.7(N)-15] c where the critical damping coefficient is given by the expression: cc = 2 ma [3.7(N)-16] c Substituting Equation 3.7(N)-16 into Equation 3.7(N)-15 and solving for c/m gives: c m = 2ap [3.7(N)-17] Subsituting this expression and the expression for k/m given by Equation 3.7(N)-9 into Equation 3.7(N)-14 gives: u + 2apu + a2u = -y s [3.7(N)-18] Note the similarity of Equations 3.7(N)-11 and 3.7(N)-18. Thus each mode may be analyzed as though it were a single degree of freedom system, and all modes are independent of each other. By this method, a fraction of critical damping, i.e., c/cc , may be assigned to each mode, and it is not necessary to identify or evaluate individual damping coefficients, i.e., c. However, assigning only a single damping ratio to each mode has a drawback. There are three ways used to overcome this limitation when considering a slightly damped structure (e.g., steel) supported by a massive moderately damped structure (e.g., concrete). The first method is to develop and analyze separate mathematical models for both structures, using their respective damping values. The massive, moderately damped support structure is analyzed first. The calculated response at the support points for the slightly damped structures is used as a forcing function for the subsequent detailed analysis. The second method is to inspect the mode shapes to determine which modes correspond to the slightly damped structure and then use the damping associated with the structure having predominant motion. The third method is to use the Rayleigh damping method based on computed modal energy distribution. 3.7(N).2.1.3 Response Spectrum Analysis The response spectrum is a plot showing the variation in the maximum response (Ref. 5, Pages 24 through 51) (displacement, 3.7(N)-8 Rev. 0

WOLF CREEK velocity, and acceleration) of a single degree of freedom system versus its natural frequency of vibration when subjected to a time-history motion of its base. The response spectrum concept can be best explained by outlining the steps involved in developing a spectrum curve. Determination of a single point on the curve requires that the response (displacement, velocity, and acceleration) of a single degree of freedom system with a given damping and natural frequency is calculated for a given base motion. The variations in response are established, and the maximum absolute value of each is plotted as an ordinate with the natural frequency used as the abscissa. The process is repeated for other assumed values of frequency in sufficient detail to establish the complete curve. Other curves corresponding to different fractions of critical damping are obtained in a similar fashion. Thus, the determination of each point of the curve requires a complete dynamic response analysis, and the determination of a complete spectrum may involve hundreds of such analyses. However, once a response spectrum plot is generated for the particular base motion, it may be used to analyze each structure and component with the base motion. The spectral acceleration, velocity, and displacement are related by the equation: San = anSvn = an2Sdn [3.7(N)-19] There are two types of response spectra that must be considered. If a given building is shown to be rigid and to have a hard foundation, the ground response spectrum or ground time-history is used. It is referred to as a ground response spectrum. If the building is flexible and/or has a soft foundation, the ground response spectrum is modified to include these effects. The response spectrum at various support points must be developed. These are called floor response spectra. 3.7(N).2.1.4 Integration of Modal Equations This method can be separated into the following two basic parts:

a. Integration procedure for the uncoupled modal Equation 3.7(N)-11 to obtain the modal displacements and accelerations as a function of time.
b. Using these modal displacements and accelerations to obtain the total displacements, accelerations, forces, and stresses.

3.7(N)-9 Rev. 0

WOLF CREEK Integration Procedure Integration of these uncoupled modal equations is done by step-by-step numerical integration. The step-by-step numerical integra-tion procedure consists of selecting a suitable time interval, t, and calculating modal acceleration, Ã n , modal velocity, n , and modal displacement, A n, at discrete time stations t apart, starting at t = 0 and continuing through the range of interest for a given time-history of base acceleration. Total Displacements, Accelerations, Forces, and Stresses From the modal displacements and accelerations, the total displacements, accelerations, forces, and stresses can be determined as follows:

a. Displacement of mass point r in mode n as a function of time is given by Equation 3.7(N)-12 as:

urn = Anrn [3.7(N)-20] with the corresponding acceleration of mass point r in mode n as: 

            ürn = Ãnrn                               [3.7(N)-21]
b. The displacement and acceleration values obtained for the various modes are superimposed algebraically to give the total displacement and acceleration at each time interval.
c. The total acceleration at each time interval is multi-plied by the mass to give an equivalent static force.

Stresses are calculated by applying these forces to the model or from the deflections at each time interval. 3.7(N).2.1.5 Integration of Coupled Equations of Motion The dynamic transient analysis is a time-history solution of the response of a given structure to known forces and/or displacement forcing functions. The structure may include linear or nonlinear elements, gaps, interfaces, plastic elements, and viscous and Coulomb dampers. Nodal displacements, nodal forces, pressure, and/or temperatures may be considered as forcing functions. Nodal displacements and elemental stresses for the complete structure are calculated as functions of time. 3.7(N)-10 Rev. 0

WOLF CREEK The basic equations for the dynamic analysis are as follows: [M] {x} + [C] {x} + [K] {x} = {F(t)} [3.7(N)-22] where the terms are as defined earlier and {F(t)} may include the effects of applied displacements, forces, pressures, temperatures, or nonlinear effects such as plasticity and dynamic elements with gaps. Options of translational accelerations input to a structural system and the inclusion of static deformation and/or preload may be considered in the nonlinear dynamic transient analysis. The option of translational input such as uniform base motion to a structural system is considered by introducing an intertia force term of -M {z } to the right hand side of the basic Equation 3.7(N)-22, i.e., [M] {x} + [C] {x} + [K] {x} = {F(t)}-[M]{z} [3.7(N)-23] The vector {z} is defined by its components z i where i refers to each degree of freedom of the system. z i is equal to a 1, a2, or a3 if the i-th degree for freedom is aligned with the direction of the system translational acceleration a1, a2, or a3, respectively. z i = 0 if the i-th degree of freedom is not aligned with any direction of the system translational acceleration. Typical application of this option is a structural system subjected to a seismic excitation of a given ground acceleration record. The displacement {x} obtained from the solution of Equation 3.7(N)-23 is the displacement relative to the ground. The option of the inclusion of initial static deformation or preload in a nonlinear transient dynamic structural analysis is considered by solving the static problem prior to the dynamic analysis. At each state of integration in transient analysis, the portion of internal forces due to static deformation is always balanced by the portion of the forces which is statically applied. Hence, only the portion of the forces which deviates from the static loads will produce dynamic effects. The output of this analysis is the total result due to static and dynamic applied loads. One available method for the numerical integration of Equations 3.7(N)-22 and 3.7(N)-23 is the Newmark Beta integration scheme proposed by Chan, Cox, and Benfield (Ref. 6). In this integration scheme, Equations 3.7(N)-22 and 3.7(N)- 23 are replaced by: 3.7(N)-11 Rev. 0

WOLF CREEK 1 1 [M] {xn+2-2xn+1 + xn} + {xn+2 - xn} [C] (t)2 2(t) 

             + [K] {xn+2 + (1-2)xn+1 +xn}
             = {Fn+2 + (1-2)Fn+1 +Fn}                   [3.7(N)-24]

n+2 n+1 B n Where: n, n+1, n+2 = past, present, and future (updated) values of the variables 

                    = parameter to be selected on the basis of numerical stability and accuracy F  = the total right hand side of the equa-tion of motion (Equation 3.7(N)-22 or 3.7(N)-23) t = tn+2 - tn+1 = tn+1 - tn The value of  is chosen equal to 1/3 in order to provide a margin of numerical stability for nonlinear problems. Since the numerical stability of Equation 3.7(N)-24 is mostly determined by the left hand side terms of that equation, the right hand side terms were replaced by F n+2 . Furthermore, since the time increment may vary between two successive time substeps, Equation 3.7(N)-24 may be modified as follows:

2 xn+2 - xn+1 - xn+1-xn t + t1 [M] t t1 1 1 [C] ( xn+2 - xn) + 3 [K] ( xn+2 + xn+1 + xn) = Fn+2 t + t1 [3.7(N)-25] By factoring x n+2, xn+1, and xn, and rearranging terms, Equation 3.7(N)-26 is obtained as follows: {C5 [M] + C3 [C] + (1/3) [K]} {x n+2} = FN+2 

             + {C 7 [M] - (1/3) [K]}      {x n+1}
             + {-C 2 [M] + C3 [C] - (1/3) [K]}     {x n}   [3.7(N)-26]

3.7(N)-12 Rev. 0

WOLF CREEK where: 2 C2 = t1(t + t1) 1 C3 = t + t1 2 C5 = t(t + t1) C 7 = C2 + C5 The above set of simultaneous linear equations is solved to obtain the present values of nodal displacements {xt} in terms of the previous (known) values of the nodal displacements. Since [M], [C], and [K] are included in the equation, they can also be time or displacement dependent. 3.7(N).2.2 Natural Frequencies and Response Loads Refer to Section 3.7(B).2.2. 3.7(N).2.3 Procedures Used for Modeling Procedures used for modeling are discussed in Section 3.7(N).2.1.1. 3.7(N).2.4 Soil/Structure Interaction Refer to Section 3.7(B).2.4. 3.7(N).2.5 Development of Floor Response Spectra Refer to Section 3.7(B).2.5. 3.7(N).2.6 Three Components of Earthquake Motion The seismic design of the piping and equipment includes the effect of the seismic response of the supports, equipment, structures, and components. The system and equipment response is determined, using three earthquake components- -two horizontal and one vertical. The design ground response spectra are the bases for generating these three input components. Floor response spectra are generated for two perpendicular horizontal directions (i.e., N-S, E-W) and the vertical direction. System and equipment analysis is 3.7(N)-13 Rev. 0

WOLF CREEK performed with these input components applied in the N-S, E-W, and vertical direction. The damping values used in the analysis are those given in Table 3.7(N)-1. In computing the system and equipment response-by-response spectrum modal analysis, the methods of Section 3.7(N).2.7 are used to combine all significant modal responses to obtain the combined unidirectional responses. The combined total response is then calculated, using the square root of the sum of the squares formula applied to the resultant unidirectional responses. For instance, for each item of interest such as displacement, force, stresses, etc., the total response is obtained by applying the above-described method. The mathematical expression for this method (with R as the item of interest) is: 3 21/2 RC = RT [3.7(N)-27] T=1 where: N 1/2 RT = RTi2 [3.7(N)-28] i=1 where: RC = total combined response at a point RT = value of combined response of direction T RTi = absolute value of response for direction T, mode i N = total number of modes considered The subscripts can be reversed without changing the results of the combination. Again, for the case of closely spaced modes, R T in Equation 3.7(N)-28 shall be replaced with RT as given by Equation 3.7(N)-29 in Section 3.7(N).2.7. 3.7(N).2.7 Combination of Modal Response The total unidirectional seismic response is obtained by combining the individual modal responses, utilizing the square root of the sum of the squares method. For systems having modes with closely spaced frequencies, this method is modified to include the 3.7(N)-14 Rev. 0

WOLF CREEK possible effect of these modes. The groups of closely spaced modes are chosen so that the difference between the frequencies of the first mode and the last mode in the group does not exceed 10 percent of the lower frequency. Groups are formed, starting from the lowest frequency and working toward successively higher frequencies. No one frequency is in more than one group. Combined total response for systems which have such closely spaced modal frequencies is obtained by adding to the square root of the sum of the squares of all modes the product of the responses of the modes in each group of closely spaced modes and a coupling factor. This can be represented mathematically as: N S Nj-1 Nj RT2 = Ri + 2 2 RK Rl Kl [3.7(N)-29] i-1 j=1 K=Mj l=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 Mj = lowest modal number associated with group j of closely spaced modes Nj = highest modal number associated with group j of closely spaced modes Kl = coupling factors with: 

               é      '      '    
                   =   -  =      ú K      l Kl = {1 +                   ú  2    } -1        [3.7(N)-30]

(K=K + l=l)û and 1/2 

               = =K é1 -      (K2 )û
         =K                                           [3.7(N)-31]
          '                     2 K    = K      +                            [3.7(N)-32]
                              =Kt d 3.7(N)-15                Rev. 0

WOLF CREEK where: K = frequency of closely spaced mode k K = fraction of critical damping in closely spaced mode k td = duration of the earthquake An example of this equation applied to a system can be supplied with the following considerations. Assume that the predominant contributing modes have frequencies as given below: Mode 1 2 3 4 5 6 7 8 Frequency 5.0 8.0 8.3 8.6 11.0 15.5 16.0 20 There are two groups of closely spaced modes, namely with modes {2,3,4} and {6, 7}. Therefore: S = 2 number of groups of closely spaced modes M1 = 2 lowest modal number associated with group 1 N1 = 4 highest modal number associated with group 1 M2 = 6 lowest modal number associated with group 2 N2 = 7 highest modal number associated with group 2 N = 8 total number of modes considered The total response for this system is, as derived from the expansion of Equation 3.7(N)-29: 2 2 2 2 2 RT = R1 + R2 + R3 + .... + R8 

          +   2 R2 R3 23 + 2 R2 R4 24
          +   2 R3R434 + 2 R6R767                     [3.7(N)-33]

3.7(N).2.8 Interaction of Non-Category I Structures With Seismic Category I Structures Refer to Section 3.7(B).2.8. 3.7(N)-16 Rev. 0

WOLF CREEK 3.7(N).2.9 Effects of Parameter Variations on Floor Response Spectra Refer to Section 3.7(B).2.9. 3.7(N).2.10 Use of Constant Vertical Static Factors Constant vertical static factors are not used as the vertical floor response load for the seismic design of safety classed systems and components within Westinghouses scope of responsibility. All such systems and components are analyzed in the vertical direction. 3.7(N).2.11 Methods Used to Account for Torsional Effects Refer to Section 3.7(B).2.11. 3.7(N).2.12 Comparison of Responses Refer to Section 3.7(B).2.12. 3.7(N).2.13 Methods for Seismic Analysis of Dams Refer to Section 3.7(B).2.13. 3.7(N).2.14 Determination of Seismic Category I Structure Overturning Moments Refer to Section 3.7(B).2.14. 3.7(N).2.15 Analysis Procedure for Damping In instances under the standard scope of Westinghouse supply and analysis, either the lowest damping value associated with the elements of the system is used for all modes, or an equivalent modal damping value is determined by testing programs, such as was done for the reactor coolant loop (Ref. 5). 3.7(N).3 SEISMIC SUBSYSTEM ANALYSIS This section describes the seismic analysis performed on subsystems within Westinghouses scope of responsibility. 3.7(N).3.1 Seismic Analysis Methods Seismic analysis methods for subsystems within Westinghouses scope of responsibility are given in Section 3.7(N).2.1. 3.7(N)-17 Rev. 0

WOLF CREEK 3.7(N).3.2 Determination of Number of Earthquake Cycles For each OBE, the system and component will have a maximum response corresponding to the maximum induced stresses. The effect of these maximum stresses for the total number of OBEs must be evaluated to assure resistance to cyclic loading. The OBE is conservatively assumed to occur 20 times over the life of the plant. The number of maximum stress cycles for each occurrence depends on the system and component damping values, complexity of the system and component, and duration and frequency contents of the input earthquake. A precise determination of the number of maximum stress cycles can only be made, using time- history analysis for each item which is not feasible. Instead, a time-history study has been conducted to arrive at a realistic number of maximum stress cycles for all Westinghouse systems and components. To determine the conservative equivalent number of cycles of maximum stress associated with each occurrence, an evaluation was performed, considering both equipment and its supporting building structure as single degree of freedom systems. The natural frequencies of the building and the equipment are conservatively chosen to coincide. The damping in the equipment and building is equivalent to the damping values in Table 3.7(N)-1. The results of this study indicate that the total number of maximum stress cycles in the equipment having peak acceleration above 90 percent of the maximum absolute acceleration did not exceed 8 cycles. If the equipment was assumed to be rigid in a flexible building, the number of cycles exceeding 90 percent of the maximum stress was not greater than 3 cycles. This study was conservative since it was performed with single degree of freedom models which tend to produce a more uniform and unattenuated response than a complex interacting system. The conclusions indicate that 10 maximum stress cycles for flexible equipment (natural frequencies less than 33 Hz) and 5 maximum stress cycles for rigid equipment (natural frequencies greater than 33 Hz) for each of 20 OBE occurrences should be used for fatigue evaluation of Westinghouse systems and components. 3.7(N)-18 Rev. 0

WOLF CREEK 3.7(N).3.3 Procedure Used for Modeling Refer to Section 3.7(N).2.1 for modeling procedures for subsystems in Westinghouses scope of responsibility. 3.7(N).3.4 Basis for Selection of Frequencies The analysis of equipment subjected to seismic loading involves several basic steps, the first of which is the establishment of the intensity of the seismic loading. Considering that the seismic input originates at the point of support, the response of the equipment and its associated supports, based upon the mass and stiffness characteristics of the system will determine the seismic accelerations which the equipment must withstand. Three ranges of equipment/support behavior which affect the magnitude of the seismic acceleration are possible:

a. If the equipment is rigid relative to the structure, the maximum acceleration of the equipment mass approaches that of the structure at the point of equipment support.

The equipment acceleration value in this case corresponds to the low period region of the floor response spectra.

b. If the equipment is very flexible, relative to the structure, the equipment will show very little response.
c. If the periods of the equipment and supporting structure are nearly equal, response occurs and must be taken into account.

In all cases, equipment under earthquake loadings is designed to be within Code allowable stresses. Also, as noted in Section 3.7 (N).3.2, rigid equipment/support systems have natural frequencies greater than 33 Hz. 3.7(N).3.5 Use of Equivalent Static Load Method of Analysis The static load equivalent or static analysis method involves the multiplication of the total weight of the equipment or component number by the specified seismic acceleration coefficient. The magnitude of the seismic acceleration coefficient is established on the basis of the expected dynamic response characteristics of the component. Components which can be adequately characterized as single degree of freedom systems are considered to have a modal participation factor of one. Seismic acceleration coefficients 3.7(N)-19 Rev. 0

WOLF CREEK for multidegree of freedom systems which may be in the resonance region of the amplified response spectra curves are increased by 50 percent to account conservatively for the increased modal participation. 3.7(N).3.6 Three Components of Earthquake Motion Methods used to account for three components of earthquake motion for subsystems in Westinghouses scope of responsibility are given in Section 3.7(N).2.6. 3.7(N).3.7 Combination of Modal Responses Methods used to combine modal responses for subsystems in Westinghouses scope of responsibility are given in Section 3.7(N).2.7. 3.7(N).3.8 Analytical Procedures for Piping The Class 1 piping systems are analyzed to the rules of the ASME Code, Section III, NB-3650. When response spectrum methods are used to evaluate piping systems supported at different elevations, the following procedures are used. The effect of differential seismic movement of piping supports is included in the piping analysis, according to the rules of the ASME Code, Section III, NB-3653. According to ASME definitions, these displacements cause secondary stresses in the piping system. The response quality of interest induced by differential seismic motion of the support is computed statically by considering the building response on a mode-by-mode basis. In the response spectrum dynamic analysis for evaluation of piping systems supported at different elevations, the most severe floor response spectrum corresponding to the support locations is used. Westinghouse does not have in their scope of analysis any piping systems interconnected between buildings. 3.7(N).3.9 Multiple Supported Equipment Components with Distinct Inputs When response spectrum methods are used to evaluate reactor coolant system primary components interconnected between floors, the procedures of the following paragraphs are used. There are no components in the Westinghouse scope of analysis which are connected between buildings. The primary components of the reactor coolant system are supported at no more than two floor elevations. 3.7(N)-20 Rev. 0

WOLF CREEK A dynamic response spectrum analysis is first made, assuming no relative displacement between support points. The response spectra used in this analysis is the most severe floor response spectra. Secondly, the effect of differential seismic movement of components interconnected between floors is considered statically in the integrated system analysis and in the detailed component analysis. The results of the building analysis are reviewed on a mode-by-mode basis to determine the differential motion in each mode. Per ASME Code rules, the stress caused by differential seismic motion is clearly secondary for piping (NB-3650) and component supports (NF-3231). For components, the differential motion will be evaluated as a free end displacement, since, per NB-3213.19, examples of a free end displacement are motions "that would occur because of relative thermal expansion of piping, equipment, and equipment supports, or because of rotations imposed upon the equipment by sources other than the piping." The effect of the differential motion is to impose a rotation on the component from the building. This motion, then, being a free end displacement and being similar to thermal expansion loads, will cause stresses which will be evaluated with ASME Code methods, including the rules of NB-3227.5 used for stresses originating from restrained free end displacements. The results of these two steps, the dynamic inertia analysis and the static differential motion analysis, are combined absolutely with due consideration for the ASME classification of the stresses. 3.7(N).3.10 Use of Constant Vertical Static Factors Constant vertical load factors are not used as the vertical floor response load for the seismic design of safety-related components and equipment within Westinghouses scope of responsibility. 3.7(N).3.11 Torsional Effects of Eccentric Masses The effect of eccentric masses, such as valves and valve operators, is considered in the seismic piping analyses. These eccentric masses are modeled in the system analysis, and the torsional effects caused by them are evaluated and included in the total system response. The total response must meet the limits of the criteria applicable to the safety class of piping. 3.7(N).3.12 Buried Seismic Category I Piping Systems and Tunnels Refer to Section 3.7(B).3.12. 3.7(N)-21 Rev. 0

WOLF CREEK 3.7(N).3.13 Interaction of Other Piping with Seismic Category I Piping Refer to Section 3.7(B).3.13. 3.7(N).3.14 Seismic Analyses for Reactor Internals Fuel assembly component stresses induced by horizontal seismic disturbances are analyzed through the use of finite element computer modeling. The time-history floor response based on a standard seismic time- history normalized to SSE levels is used as the seismic input. The reactor internals and the fuel assemblies are modeled as spring and lumped mass systems or beam elements. The component seismic response of the fuel assemblies is analyzed to determine design adequacy. A detailed discussion of the analyses performed for typical fuel assemblies is contained in Reference 2. Fuel assembly lateral structural damping obtained experimentally is presented in Reference 2 (Figure B-4). The data indicates that no damping values less than 10 percent were obtained for fuel assembly displacements greater than 0.11 inch for the SSE. The distribution of fuel assembly amplitudes decreases as one approaches the center of the core. The average amplitude for the minimum displacement fuel assembly is well above 0.11 inch for the SSE. Fuel assembly displacement time-history for the SSE seismic input is illustrated in Reference 2 (Figure 2-3). The CRDMs are seismically analyzed to confirm that system stresses under the combined loading conditions, as described in Section 3.9(N).1, do not exceed allowable levels, as defined by the ASME Code, Section III for "Upset" and "Faulted" conditions. The CRDM is mathematically modeled as a system of lumped and distributed masses. The model is analyzed under appropriate seismic excitation, and the resultant seismic bending moments along the length of the CRDM are calculated. The corresponding stresses are then combined with the stresses from the other loadings required, and the combination is shown to meet ASME Code, Section III requirements. 3.7(N).3.15 Analysis Procedure for Damping Analysis procedures for damping for subsystems in Westinghouses scope of responsibility are given in Section 3.7(N).2.15. 3.7(N)-22 Rev. 0

WOLF CREEK 3.7 (N).4 SEISMIC INSTRUMENTATION Refer to Section 3.7(B).4. 3.7(N).5 REFERENCES

1. "Damping Values of Nuclear Power Plant Components," WCAP-7921-AR, May, 1974.
2. Gesinski, L. T. and LeBastard, G., "Safety Analysis of the 8-Grid 17 x 17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident," WCAP-8236, Addendum 1 (Proprietary), and WCAP-8288, Addendum 1 (Non-Proprietary), March 1974.
3. Lin, C. W., "How to Lump the Masses - A Guide to the Piping Seismic Analysis," ASME Paper 74-NE-7 presented at the Pressure Vessels and Piping Conference, Miami, Florida, June, 1974.
4. Biggs, J. M., Introduction to Structural Dynamics, McGraw-Hill, New York, 1964.
5. Thomas, T. H., et al., "Nuclear Reactors and Earth quakes,"

TID-7024, U. S. Atomic Energy Commission, Washington, D. C., August, 1963.

6. Chan, S. P., Cox, H. L., and Benfield, W. A., "Transient Analysis of Forced Vibration of Complex Structural-Mechanical Systems," J. Royal Aeronautical Society, July, 1962.

3.7(N)-23 Rev. 0

WOLF CREEK TABLE 3.7(N)-1 DAMPING VALUES USED FOR SEISMIC SYSTEMS ANALYSIS FOR WESTINGHOUSE SUPPLIED EQUIPMENT Damping (Percent of Critical) Upset Faulted Conditions Condition Item (OBE) (SSE, DBA) Primary coolant loop system components and large piping* 2 4 Small piping 1 2 Welded steel structures 2 4 Bolted and/or riveted steel structures 4 7

1. The damping values provided in ASME Code Case N-411 may be utilized for piping systems as an alternative to those identified above subject to the conditions listed in Regulatory Guide 1.84.
  • Applicable to 12-inch or larger diameter piping.

Rev. 12

WOLF CREEK Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FI GU R E3

  • 7 ( N) --1 MULTI-DEGREE-OF-FREEDOM SYSTEM

WOLF CREEK 3.7(S) SEISMIC DESIGN The following material applies to the nonpowerblock site-related, seismic Category I structures, systems, and components. 3.7(S).1 SEISMIC INPUT 3.7(S).1.1 Design Response Spectra The site design response spectra in compliance with Regulatory Guide 1.60 are illustrated in Figures 3.7(S)-1 and 3.7(S)-2, in both the horizontal and vertical directions for the 0.12g safe shutdown earthquake (SSE). For the operating basis earthquake (OBE), the design response spectra values are taken as 50 percent of the SSE. Section 2.5.2 and BC-TOP-4A, Section 2.5, discuss the effects of focal and epicentral distances from the site, depths between the focus of the seismic disturbances and the site, existing earthquake records, and the associated amplification of the response spectra. Appendix 3C contains a seismic evaluation of the Wolf Creek Generating Station structures using the Lawrence Livermore Laboratories spectrum. This spectrum is enveloped by a Regulatory Guide 1.60 spectrum anchored at 0.15g and is illustrated in Figures 3.7(S)-13 and 3.7(S)-14. A 20.48-second duration is considered to be adequate for the time-history type of analysis used for the structures and equipment. The design response spectra and earthquake time-histories are applied in the free field at finished grade. 3.7(S).1.1.1 Bases for Site Dependent Analysis Section 2.5.2 and BC-TOP-4A, Sections 2.4 and 2.5, describe the bases for specifying the vibratory ground motion for design use. 3.7(S).1.2 Design Time-History Synthetic earthquake time-histories were generated because the response spectra of recorded earthquake motions do not necessarily envelop the site's design spectra. USAR Figures 3.7(B)-3 and 3.7(B)-4 show the synthetic earthquake time-history motions for the SSE in the horizontal and vertical directions, respectively. Figures 3.7(S)-3 through 3.7(S)-8 show that the 10-percent, 7-percent, and 5-percent damping response spectra of the site synthetic time-history in the horizontal and vertical directions envelop the corresponding design spectra. Section 2.5.1 of BC-TOP-4A describes the generation of a typical synthetic earthquake time-history. 3.7(S)-1 Rev. 14

WOLF CREEK A typical foundation-level, free-field acceleration response spectrum is presented in Figure 3.7(S)-9. The curve overlies the 60-percent design response spectrum. Conservative design seismic loads and floor response spectra are obtained by use of the computed foundation free-field response spectra and by broadening the floor response spectra by + 10 percent. 3.7(S).1.3 Supporting Media for Seismic Category I Structures A description of the supporting media for site-related seismic Category I structures is provided in Section 2.5.4. Figure 3.7(S)-10 provides the free-field soil profile. Table 3.7(S)-1 presents all non powerblock site-related seismic Category I structures and respective depths of soil or backfill deposits over bedrock. 3.7(S).2 SEISMIC SYSTEM ANALYSIS 3.7(S).2.1 Seismic Analysis Methods Refer to Appendix 3C, USAR Section 3.7(B).2.1 and the following table and figure for information on seismic analysis methods.

1. Table 3.7(S)-2 lists the methods of analysis for the site-related seismic Category I structures.
2. Figure 3.7(S)-11 shows the typical mathematical model for the ESWS pumphouse.

3.7(S).2.2 Natural Frequencies and Response Loads The SSE fundamental mode natural frequencies for the ESWS pumphouse are 7.0 Hz in both the north-south and east-west directions, and 8.1 Hz in the vertical direction. The OBE fundamental mode natural frequencies for the ESWS pumphouse are 7.2 Hz in both the north-south and east-west directions, and 9.5 Hz in the vertical direction. A summary of response parameters determined by seismic analysis is provided in Table 3.7(S)-3A, -3B, identifying their characteristic responses. Typical floor response spectra are presented in Figure 3.7(S)-12 for both SSE and OBE. 3.7(S)-2 Rev. 10

WOLF CREEK 3.7(S).2.3 Soil/Structure Interaction Refer to USAR Section 3.7(B).2.4 and Table 3.7(S)-2 where foundation embedment depth below grade, minimum base dimension, and method of analysis are given. Refer to USAR Section 3.7(B).2.4 for a description of the FLUSH, finite element method of analysis. Structures completely buried below grade (ESWS Caissons and electrical manholes) move with the ground motion as a single, lumped mass. To account for the inertial effects of the walls and slabs due to the ground motion, the mass of the walls and slabs are multiplied by the site SSE and OBE. To account for the effects of soil pressures on the walls due to the ground motion, additional soil pressures as a function of the site SSE and OBE are applied to the walls [refer to Figure 2.5-107a & 107b]. This procedure is conservative in the design of buried structures. Since response spectra are not needed for equipment qualification, finite element analysis is not performed. 3.7(S).2.4 Methods for Seismic Analysis of Dams The procedure for the seismic analysis of seismic Category I dams is provided in Section 2.5.6.5. The assumptions made, the boundary conditions used, and the procedure by which strain-dependent soil properties are incorporated in the analysis are also provided in Section 2.5.6.5. Appendix 3C also contains the results of the seismic evaluation of the UHS dam using the Lawrence Livermore Laboratories spectrum. 3.7(S)-3 Rev. 28

WOLF CREEK TABLE 3.7(S)-1 DEPTH OF SOIL DEPOSITED OVER BEDROCK SITE-RELATED SEISMIC CATEGORY I STRUCTURES Approximate Average Average Elev. of Elev. Of Depth of Soil Bottom of Top of Over Rock Structure Base Mat Rock (feet) ESWS Pumphouse 1952'-10" 1960'-0" 0 ESWS Electrical Manholes 1979'-5" 1962'-0" 17.5 Rev. 28

WOLF CREEK TABLE 3.7(S)-2 FOUNDATION DEPTH BELOW GRADE, MINIMUM BASE DIMENSION AND METHOD OF ANALYSIS FOR SITE-RELATED SEISMIC CATEGORY I STRUCTURES Approximate Approximate Ratio of Foundation Minimum Embedment Embedment Base Depth to Method Depth Below Dimension Minimum Base of Structure Grade (feet) (feet) Dimension Analysis ESWS 47 40 1.175 1 Pumphouse ESWS Electrical 20 11 1.818 2 Manholes 1 Finite-element method, FLUSH computer program. 2 Single lumped mass-spring method - structures are buried below grade. Rev. 28

WOLF CREEK TABLE 3.7(S)-3A SPECTRAL RESPONSE

SUMMARY

ESWS PUMP HOUSE '

0. 12G SSE REF. FIGURE 3.7(S)-ll NORTH-SOUTH DIRECTION EAST-WEST DIRECTION VERTICAL DIRECTION FLUSH FLUSH FLUSH MASS LUMPED MODEL MAX. INERTIA SHEAR BENDING DISPLACE- MODEL MAX. INERTIA SHEAR BENDING DISPLACE- MODEL MAX. INERTIA AXIAL DISPLACE POINT ELEVATION WEIGHT NODE ACCEL'S FORCES FORCES MOMENTS MENTS NODE ACCEL'S FORCES FORCES MOMENTS MENTS NODE ACCEL'S FORCES FORCES MENT NO. POINT POINT POINT (KIPS) NO. IG'SI (KIPS) (KIPS) (KIP-FT) (INCHES) NO. (G'SI (KIPS) (KIPS) (KIP-FT) (INCHES) NO. (G'S) (KIPS) (KIPS) (INCHES) 0 2038'-8" 534 93 .438 234 0 .082 141 116 115 0 .073 141 .197 105 .031 234 115 105 CD 2025'-0" 1,647 94 .336 586 3,200 .075 142 .187 307 1,570 .074 142 .195 321 .030 820 422 426
 -     2012'-6"     905    95   181        254             13,450       -        143   .161       146             6,850        -    143   .190      172                 -

1,074 568 598 2000'-0" 3,792 96 .207 786 26,870 .070 144 .141 534 13,950 .080 144 .178 676 .026 1,860 1,102 1,274 

 -     1985'~"    2,801    99   .130       363             54,770       -       227    .110       308            30,480       -     227   .139      390                 -

2,223 1,410 1,664 

 -     1969'-0"   2,613   102   .103       269             90,340       -       230    .103      269             53,040       -     230  .121       317                 -

2,492 1,679 1,981 CD 1958'-0" 2,758 104 .102 280 117,750 .081 232 .102 281 71,510 .089 232 .116 319 .010 2,772 1,960 2,300 

 -     1953'~"        -   105     --                      131,610       -       233      -         -             81,310       -     233     -         -                 -

Rev. 0

WOLF CREEK TABLE 3.7(S)-3B SPECTRAL RESPONSE

SUMMARY

ESWS PUMPHOUSE 0.06G OBE REF. FIGURE 3.7(8)-11 NORTH-SOUTH DIRECTION EAST-WEST DIRECTION VERTICAL DIRECTION FLUSH FLUSH FLUSH MASS LUMPED MODEL MAX. INERTIA SHEAR BENDING DISPLACE- MODEL MAX. INERTIA SHEAR BENDING DISPLACE- MODEL MAX. INERTIA AXIAL DISPLACE-POINT ELEVATION WEIGHT NODE ACCEL'S FORCES FORCES MOMENTS MENTS NODE ACCEL'S FORCES FORCES MOMENTS MENTS NODE ACCEL'S FORCES FORCES MENT NO. POINT POINT POINT (KIPS) NO. (G'SI (KIPS) (KIPS) (KIP-FTI !INCHES) NO. IG'SI (KIPS) (KIPS) IKIP-FT) {INCHES) NO. IG'SI (KIPS) {KIPS) {INCHES) 0 2038'-8" 534 93 .225 120 120 0 .024 141 .103 55 0 .026 141 .098 52 .015 55 52 0 2025'-0" 1,647 94 .185 305 425 1,640 .022 142 .091 150 750 .027 142 .osi 159 .015 205 211 2012'-6" 905 95 .150 136 6,950 143 .081 73 3,310 143 .094 85 561 278 296 0 2000'.()" 3,792 96 .115 437 13,970 .021 144 .073 275 6,790 .030 144 .088 333 .012 998 553 629 

 -     1985'.()"   2,801    99   .079        222            28,940        -       227   .061       171             15,080            227    .070       197                -

1,220 724 826 1969'.()" 2,613 102 .055 143 48,460 - 230 .053 139 26,670 - 230 .062 161 - 1,363 863 987 CD 1958'.()" 2,758 104 .053 145 63,450 .033 232 .053 145 36,160 .036 232 .059 163 .004 1,508 1,008 1,150 

 -     1953'.()"           105                -             70,990        -       233                -             41,200       -    233      -                           -

Rev. 0

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WOLF CREEK PERI60. Sfi:C. soo2.rO~~--tr.o~~--oT._s~--~-o.*,2~~-o_.rl~~~~----~~0.02 s.o u w (/)

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WOLF CREEK PERieD. SEC. 2.0 1 *0 o.s 0.2 0.1 o.os 500 5.0 u w en 'z 2.0 

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o.s 1 *0 2.0 5.0 10.0 20.0 50.0 FREQUENCY. CPS Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-4 HORIZONTAL DESIGN RESPONSE SPECTRA FOR 0.12G HORIZONTAL GROUND ACCELERATION (7% DAMPING)

WOLF CREEK PERieO. SEC. 2.0 1 *0 o.s 0.2 0.1 o.os 50 0 s.o u w en z 2.0 1-u CD _J w 1 .0 0.2

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o.s 1 *0 2.0 s.o 10.0 20.0 so.o FREQUENCY. CPS Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-5 HORIZONTAL DESIGN RESPONSE SPECTRA FOR 0.12G HORIZONTAL GROUND ACCELERATION (5% DAMPING)

WOLF CREEK PERIBO. SEC. 2.0 t.o o.s 0.2 0.1 o.os 0.02 50 0 s.o u 4J ff) 'Z. 2.0 u 1!:) 1 .o _J w

0. 1 1 *0 2.0 s.o 10.0 20.0 FREQUENCY . CPS Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-6 VERTICAL DESIGN RESPONSE SPECTRA FOR 0.12G HORIZONTAL GROUND ACCELERATION (10%

DAMPING)

WOLF CREEK PERI~O. SEC. z.o 1 *0 o.s 0.2 0.1 o.os 0.02 500 s.o u w (/') z 2.0 I-u <D _J w 1 *0 o.os ~~-L~~--~~~----~~~--L-~~~~~~--~~ o.s t.o 2.0 s.o 10.0 zo.o so.o FREQUENCY. CPS Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-7 VERTICAL DESIGN RESPONSE SPECTRA FOR 0.12G HORIZONTAL GROUND ACCELERATION (7% DAMPING)

WOL.F CREEK PER I eo. SEC*. 2.0 1 *0 o.s . 0.2 0.1 o.os 0.02 500 .-~~--~~r-.-~--~~~~~~--~~~~----~ s.o u w (f) z 2.0 1-u ~ _J w 1 *0

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0.5 1 *0 2.0 5.0 10.0 20.0 50.0 FREQUENCY. CPS Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-8 VERTICAL DESIGN RESPONSE SPECTRA FOR 0.12G HORIZONTAL GROUND ACCELERATION (5% DAMPING)

WOLF CREEK 0 (0 0 0 t.O 0 0 v 0 (!) z 0 i= <(0 O:M ~0 w (.) (.) <( 0 C'\1 0 0 0 1.00 FREQUENCY (HZ) Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-9 TYPICAL FREE FIELD BASE ELEVATION SPECTRA ESWS PUMPHOUSE

SSE==0.12g OBE = 0.06g_ TYPICAL TYPICAL LAYER SHEAR TYPICAL SHEAR TYPICAL UNIT MATERIAL rrHICKNESS MODULUS DAMPING tv10DULUS DAMPING FDISSON's WEIGHT REMARKS (FEET) RATIO G (KSF) FRACTION G (KSF) FRACTION {PC F) 0 

             ~

3 750 .035 930 .031 STRAIN DEPENfXNT SHEAR MODUWS, 3 540 *063 690 .042 

                                                                                               .40      125       AND DAMPING.       Gmax . = 1,300 6.0   COHESIVE          - - -r - - - - - ------ - - - - - - - - - - - - - - - ~~AT                .000~ 2I_RA!._tit__ _ _ _

OVERBURDEN 5 1,060 ,060 1,350 .041 .45 130 STRAIN DEPEt-DENT SHEAR MODULUS, 1G- AND DAMP! NG. G max = 2,500 5 910 .073 1,200 .050 .45 13 0 KSF AT .0001*1. STRAIN. RErx.JCTION CURVES SAME AS ABOVE. t 16 0 t- ' UPPER HEUv1ADE R 2,440TO 3,300TO STRAIN DEPENDENT SHEAR MODULUS, E::===- ~ 20 -~ 2 1. SHALE 5 2,460 

                                                                     .04 31350
                                                                                        .04     .40      135      CONSTANT DAMPING.

LL. 0 ~ G max = 5/JOO K SF AT .0001"/* STRAIN. ....., 1------- 5,660 6,070 f-- LOWER 10 .04 .04 .35 150 STRAIN OCPENDENT SHEAR MODULUS, ~ f---* HEUMADER TO TO CONSTANT DAMPING. 

<{          1------           SHALE tt30 -                                                      5,750             6,260                                G max " 8,000 KSF AT .0001*1. STRAIN.

(!) 31.0 I I PLATTSMOUTH 

         -tc 0                                                  12     270,000     .02    270,000     .02             165       CONSTANT SHEAR MODULUS, I           LIMESTONE
                                                                                                .30

~ AND DAMPING. 40 

         -I z                       43.0 I
            ~

LL. HEEBNER, LEAVEN'M::>RTH, 14 58,000 .04 5A,ooo .04 .30 140 CONSTANT SHEAR MODULUS, ~50 SNYDERVILLE AND DAMPING. _. FORM~TION w Ill 57.0 

         -I I60                             TORONTO            17      270,000     .02   270,000      .02     .30     160      CONSTANT SHEAR rv10DULUS, tl:                            LIMESTONE                                                                           AND DAMPING.

w 0 70 r---- 74.0 r---- INTERBEDDED RIGID BOUNDARY FOR FLUSH f--- SHALES BEDROCK - - - PROGRAM. V5 . ~ 4,000 FF~S. 80 - - f--* r---- r-- Rev. 1 

            ~

90- - WOLF CREEK 

            ~

UPDATED SAFETY ANALYSIS REPORT Figure 3.7(S)-10

r. Free Field Media Typical Subsurface Profile And Soil Properties SSE and OBE

WOLF CREEK SPECTRA MASS POINTS TOP OF PENTHOUSE ROOF El 2038~s 

         ,.....~--------------------------------------------141r-~~~~205----------~~~~~~~~==-----===~~--

co rri 18 9 TOP OF ROOF 

               ;---------------------------------------------142~4-~-+~~--~~--~266--~~~~=---------~==~~---

EL.2025~o 

      '  U'l N
                                                                             § .,. . .___ PUMPHOUSE
      ' U'l,
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24~~~.~~--~--~~~e~E~~~r-~~~-+-+~~~--+--+--.-~~------~~~~~~~ 2000~0"' 

               ---------1--'~,12    (i_~                                                                ..___ 31 g

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           ;--------11 r-~~~-4---;---+--~---+---+--~--~~-r~--~~r--r--+-~---+--~--~--~330 331   332      333                                                       353    354
                                                                                       /       /    /

12 10.5 10.5 12 12 Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.7(8)-11 MATHEMATICAL MODEL FOR ESWS PUMPHOUSE FOR EAST-WEST VERTICAL ANALYSIS

WOLF CREEK DAMPING VALUES

  • 0300,
  • 0500,
  • 0700.
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WOLF CREEK 3.8 DESIGN OF CATEGORY I STRUCTURES This section provides information on the containment structure and its internal structures, other powerblock seismic Category I structures, and their foundations and supports. 3.8.1 CONCRETE CONTAINMENT The containment structure is designed to house the reactor coolant system and is referred to as the reactor building in the following sections. The reactor building is part of the containment system designed to control the release of airborne radioactivity following postulated design basis accidents (DBAs) and to provide shielding for the reactor core and the reactor coolant system. This section describes the structural design considerations for the reactor building. Section 6.2 describes the functional design of the containment to minimize leakage following a LOCA. Bechtel Topical Report BC-TOP-5-A provides additional structural information relative to the design, construction, testing, and surveillance of the prestressed concrete reactor building. 3.8.1.1 Description of the Reactor Building 3.8.1.1.1 General The reactor building consists of a prestressed, reinforced concrete, cylindrical structure with a hemispherical dome and a conventionally reinforced concrete base slab with a central cavity and instrumentation tunnel to house the reactor vessel. A continuous peripheral tendon access gallery below the base slab is provided for the installation and inspection of the vertical post-tensioning system. Figures 3.8-1 through 3.8-7 illustrate this configuration and also show the relationship between the shell and its interior compartment walls and floors. The internal structures are isolated from the shell by means of an isolation gap to minimize interaction. In addition, the connections used to provide for vertical support of the structural steel floor framing at the shell allow for independent horizontal movement. Figure 1.2-1 shows the relationship between the reactor building and the surrounding structures. As shown, the shell is separated from its surrounding structures by a minimum 3-inch isolation gap to avoid interaction. In some instances, the gap is filled with a fireproof compressible material. The base slab, cylinder, and dome are reinforced by bonded reinforcing steel, as required by the design loading conditions. Additional reinforcing is provided at discontinuities in the 3.8-1 Rev. 0

WOLF CREEK structure and at major penetrations in the shell. The main reinforcing patterns for the base slab, cylinder wall, and dome are illustrated in Figures 3.8-8 through 3.8-14. The interior of the reactor building is lined with carbon steel plates welded together to form a barrier which is essentially leak tight. A post-tensioning system is used to prestress the cylindrical shell and dome. Principal nominal dimensions of the reactor building are as follows: Interior diameter 140 ft Interior height 205 ft Height to spring line 135 ft Base slab thickness 10 ft Cylinder wall thickness 4 ft Dome thickness 3 ft Liner plate thickness 0.25 in. Internal free volume 2.5 x 106 cubic ft 3.8.1.1.2 Post-Tensioning System The tendon system employed to post-tension the cylindrical shell and dome of the reactor building is shown in Figure 3.8-15. The system uses unbonded tendons, each consisting of approximately 170 one-quarter-inch-diameter high strength steel wires and anchorage components consisting of stressing washers. The prestressing load is transferred by cold-formed button heads on the ends of the individual wires, through stressing washers, to the steel bearing plates embedded in the structure. The ultimate strength of each tendon is approximately 1,000 tons. The unbonded tendons are installed in tendon ducts (sheathing) and tensioned in a predetermined sequence. The ducts, which form voids through the concrete between the anchorage points, consist of galvanized, spiral-wrapped, semirigid corrugated steel tubing. They are designed to retain their shape and resist the construction loads. The inside diameter of the ducts is sufficiently large to permit the installation of the tendons with minimum difficulty. Trumpets, which are enlarged ducts attached to the bearing plates, allow the wires to spread out at the anchorage to suit washer hole spacing and facilitate field cold formed button heading of the ends of the wires. The tendon duct provides an enclosed space surrounding each tendon. After stressing, a petroleum-based corrosion inhibitor is pumped into the duct. 3.8-2 Rev. 0

WOLF CREEK The vertical tendons consist of 86 inverted U-shaped tendons, which extend through the full height of the cylindrical wall over the dome and are anchored at the bottom of the base slab. The cylinder circumferential (hoop) tendons consist of 135 tendons anchored at three buttresses equally spaced around the outside of the reactor building. There are 30 additional hoop tendons discussed below. Each tendon is anchored at buttresses located 240 degrees apart. Three adjacent tendons, anchored at alternate buttresses, result in two complete hoop tendons. Refer to Figures 3.8-16 through 3.8-18 for tendon and buttress arrangement. Prestressing of the hemispherical dome is achieved by a two-way pattern of the inverted U-shaped tendons and 30 hoop tendons, which start at the springline and continue up to an approximate 45-degree vertical angle from the springline. Figure 3.8-16 illustrates the arrangement of the tendons in the dome. 3.8.1.1.3 Liner Plate System A carbon steel liner plate covers the entire inside surface of the reactor building (excluding penetrations). The liner is 1/4-inch thick but is thickened locally around the penetrations, large brackets, and major attachments. The liner plate, including the thickened plate, is anchored to the concrete structure. The vertical and dome liner plates are also used as forms for concrete placement. Typical details of the liner plate system are shown in Figures 3.8-19 through 3.8-22. Refer to Section 3.8.2.1 for a description of the penetrations, including the equipment and personnel access hatches, piping penetration sleeves, electrical penetration sleeves, fuel transfer tube penetration sleeve, and purge line penetration sleeves. Attachments to the liner plate which transfer loads through the liner plate to the base slab include equipment support anchors and reinforcing steel for the support of the internal structures. Refer to Figures 3.8-23 through 3.8-25 for typical details. Major structural attachments to the wall which penetrate the liner plate include polar crane brackets, floor beam brackets, and pipe support brackets. Refer to Figures 3.8-26 and 3.8-27 for typical details. Major structural attachments to the dome include various pipe support brackets. Refer to Figure 3.8-28 for typical details. Miscellaneous thickened plates, which form a part of the liner plate, are provided and anchored in the concrete to provide supports. Leak chase channels and angles are also attached at seam 3.8-3 Rev. 23

WOLF CREEK welds where the welds are inaccessible to nondestructive examination after construction. Refer to Figure 3.8-29 for typical details for these items. 3.8.1.1.4 Shell Discontinuities The significant discontinuities in the shell structure are at the wall-to-base-slab connection, the buttresses, and the large penetration openings. The shell wall interface at the base slab incorporated a straight wall-to-slab joint. Refer to Figure 3.8-10 for details of the lower wall configuration. Buttresses project out from the exterior surface of the shell wall and dome to provide adequate space for the hoop tendon anchorage and tendon-stressing equipment. The anchorage surfaces of the buttress are normal to the tangent line of the anchored hoop tendons. Details are shown in Figure 3.8-30. The concrete shell around the equipment hatch opening is thickened by the method shown in Figures 3.8-31 and 3.8-32. 3.8.1.1.5 Special Reinforcing Requirements Special reinforcing is required in such areas as the major penetrations. Refer to Figures 3.8-31 through 3.8-35 for typical details in these areas. 3.8.1.2 Applicable Codes, Standards, and Specifications The following codes, regulations, standards, and specifications were utilized in the reactor building design. Subsequent to operation, additional codes have been approved for use and are noted with an asterik. 3.8.1.2.1 Regulations

a. 10 CFR 50, "Licensing of Production and Utilization Facilities"
b. 10 CFR 100, "Reactor Site Criteria"
c. 10CFR 50.67, Accident Source Term 3.8.1.2.2 Codes
a. American Concrete Institute, Building Code Requirements for Reinforced Concrete (ACI-318-71) 3.8-4 Rev. 34

WOLF CREEK

b. American Institute of Steel Construction (AISC),

Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings, 7th Edition, adopted February 12, 1969, and Supplement Numbers 1, 2, and 3

c. ASME Boiler and Pressure Vessel Code - 1974 Edition or later Section II - Material Specifications Section III, Division 1 - Nuclear Power Plant Components Section V - Nondestructive Examination Section VIII - Pressure Vessels Section IX - Welding and Brazing Qualifications
d. American Welding Society, Structural Welding Code (AWS D1.1-75, *AWS D1.1-90, *AWS D1.1-2004)
e. Acceptable ASME Code cases per Regulatory Guides 1.84 and 1.85, as addressed in Appendix 3A
f. Appendix B, Steel Embedments, to the American Concrete Institute, Code Requirements for Nuclear Safety Related Concrete Structures, ACI 349-80 with the 1984 Supplement. The following sections of Appendix B relating specifically with expansion anchors will be applicable; Sections B.7.1 through B.7.5.

3.8.1.2.3 Standards and Specifications Industry standards, such as those published by the American Society for Testing and Materials (ASTM) and the American Association of State Highway and Transportation Officials (AASHTO), are used whenever possible to describe material properties, testing procedures, fabrication, and construction methods. The applicable standards used are listed in Section 3.8.1.6. Structural specifications are prepared to cover the areas related to the design of the reactor building. These specifications are prepared specifically for the WCGS project. These specifications emphasize the important points of the industry standards for the reactor building and reduce the options that would otherwise be permitted by the industry standards. These specifications cover the following areas:

a. Concrete material properties
b. Mixing, placing, and curing of concrete
c. Reinforcing steel and splices
d. Post-tensioning system
e. Liner plate system 3.8-5 Rev. 22

WOLF CREEK 3.8.1.2.4 Design Criteria The following design criteria form the basis for the reactor building design. Specifically, the criteria contained in Appendix C of the Bechtel Power Corporation Topical Report BC-TOP-5-A are used in the design of the reactor building. Appendix C of BC-TOP-5-A presents a detailed description of compliance with Article CC-3000 of the proposed ASME Code, Section III, Division 2.

a. 10 CFR 50, Appendix A - GDC for Nuclear Power Plants (Compliance is discussed in Section 3.1)

GDC Numbers 2, 4, 16, and 50

b. Bechtel Power Corporation topical reports, as referenced in Section 1.6 3.8.1.2.5 NRC Regulatory Guides NRC Regulatory Guides 1.10, 1.15, 1.18, 1.19, 1.55, 1.84, 1.85, 1.94, and 1.103 are applicable to the design and construction of the reactor building.

Specific editions and the extent of compliance with these guides are discussed in Appendix 3A. 3.8.1.3 Loads And Loading Combinations The applicable loads and loading combinations used in the design and analysis of the reactor building structure, components, and localized areas are those listed in BC-TOP-5-A, Appendix C. DESIGN ACCIDENT PRESSURE LOAD - Transients resulting from the DBA and other lesser accidents are presented in Section 6.2.1 and serve as the basis for the reactor building design pressure of 60 psig. PRESTRESSING FORCES - The prestressing forces are related to the design pressure by selection of a level of prestress, as discussed in Section 6.2.1 of BC-TOP-5-A. THERMAL LOADS - The temperature gradients through the reactor building wall are shown in Figure 3.8-36 for the operating condition and for the postulated DBA condition. WIND AND TORNADO LOADS - The wind and tornado loads are in accordance with Section 3.3. EARTHQUAKE LOADS - Earthquake loads are in accordance with Section 3.7. 3.8-6 Rev. 18

WOLF CREEK HYDROSTATIC LOADS - Hydrostatic loads are in accordance with Section 3.4. EXTERNAL PRESSURE LOAD - External pressure loading with a differential of 3 psig from outside to inside is considered. The external design pressure has conservatively been assumed to account for barometric pressure differentials after the reactor building is sealed. The reactor building is designed to be cooled below 50 F from the operating temperature of 120 F. The inadvertent actuation of the containment spray headers, which induce an external pressure load, is discussed in Section 6.2.1. MISSILE AND POSTULATED PIPE RUPTURE EFFECTS - The internal and external missile and postulated pipe rupture loads are in accordance with Sections 3.5 and 3.6, respectively. TEST PRESSURE LOAD - The structure is designed for a Structural Integrity Test pressure load of 69 psig. POST-LOCA FLOODING - The post-LOCA flooding of the reactor building for the purpose of fuel recovery is not a design condition. Although there are no special provisions incorporated in the structural design of the reactor building or its interior structures for the purpose of fuel recovery after a LOCA, there is sufficient time following a LOCA for the plant operators and/or consultants to assess the extent of the damage to the reactor coolant system, the interior structures of the reactor building, and refueling equipment and to make the necessary provisions, including any additional equipment required, for the recovery of the fuel. 3.8.1.4 Design and Analysis Procedures The procedures utilized in the analysis and design of the reactor building are in accordance with Sections 6.0 and 7.0 and Appendices B and C of BC-TOP-5-A. Computer programs were relied upon to perform many of the computations required for the reactor building analysis. However, in many cases, classical methods and manual techniques were used for the analysis of localized areas of the reactor building and for preliminary proportioning. Manual calculations were generally used for (a) the initial proportioning of the dome, wall, and base slab, (b) evaluation of the effects of locally applied loads, such as pipe rupture or crane loads, (c) the preparation of input for the computer analyses, and (d) areas which do not lend themselves to computer applications. Section 7.0 of BC-TOP-5-A describes the analytical methods in more detail. 3.8-7 Rev. 0

WOLF CREEK The design methods incorporated several phases, as described in Section 6.0 of BC-TOP-5-A. They involved the initial proportioning of structures, using the results of preliminary analyses documented in BC-TOP-5-A. Experience based on the completed design or parametric studies of other structures of a similar nature was used as well. The final design phase incorporated and refined information gained in the earlier phases. It also incorporated closer approximations of the equipment and piping and related loads, based on the completion of the detailed engineering design. Improved assumptions regarding material properties, including the effects of creep, shrinkage, and the cracking of concrete, were used. 3.8.1.4.1 Overall Analysis The reactor building is considered to be an axisymmetric structure for the overall analysis. Although there are deviations from this ideal shape, such as penetrations and buttresses, these deviations are sufficiently localized so as not to affect the overall analysis and are addressed by special local analyses. The overall analysis of the reactor building for axisymmetric loads was performed by using the FINEL finite element computer program described in Appendix 3.8A for combinations of the individual loading cases of dead, live, thermal, pressure, and prestress loads. The entire reactor building is modeled with one finite element mesh consisting of the dome, shell, base slab, reactor cavity, and soil. The concrete structure is modeled by continuously interconnected elements. The liner plate is modeled by a layer of elements attached to the interior surfaces of the concrete structure. The finite element mesh is extended into the soil to account for the elastic nature of the foundation material and its effect on the structure. Since the same reactor building design is used at sites with different foundation properties, the analyses were performed taking into account the range of geotechnical parameters of the foundation media at all the sites. The tendon access gallery and instrumentation tunnel were analyzed as separate structures. The finite element model used for the analysis of the reactor building for axisymmetric loads is shown in Figures 3.8-37 through 3.8-39. The overall analysis of the reactor building for nonaxisymmetric loads (i.e., seismic) was performed, using the SAP three-dimensional finite element computer program described in Appendix 3.8A. One-half of the reactor building is modeled, without the dome, about an axis of symmetry of the structure in plan. Appropriate boundary conditions were simulated at the top of the shell 3.8-8 Rev. 0

WOLF CREEK and along the axis of symmetry to provide for strain compatibility. The shell, base slab, and reactor cavity are modeled with one finite element mesh. Soil springs are provided below the structure to account for the effect of the foundation material on the structure. Since the same reactor building design is used at sites with different foundation properties, the analyses were performed, taking into account the range of geotechnical parameters of the foundation media at all the sites. The upper portion of the shell, dome, tendon access gallery, and instrumentation tunnel are analyzed separately. The finite element model used for the analysis of the reactor building for nonaxisymmetric loads is shown in Figure 3.8-40. 3.8.1.4.2 Local Analysis 3.8.1.4 2.1 Large Penetration Openings Large penetrations are defined as those having an inside diameter equal to or greater than 10 feet (2.5 times the reactor building nominal shell wall thickness). The equipment hatch and personnel lock fall into this category. Local analyses of the reactor building shell in the area of large penetrations were performed using the SAP three-dimensional finite element computer program. The analytical models consist of a one-quarter segment mesh that follows the axes of symmetry of the penetration opening. The points defining the outermost boundary of the model at the boundaries is compatible with that of the undisturbed cylindrical shell. Boundary conditions along the axes of symmetry and the boundaries of the model are specified to provide for strain compatibility. The SAP finite element models used for analyses of the equipment hatch and personnel hatch are shown in Figures 3.8-41 through 3.8-43. 3.8.1.4.2.2 Small Penetration Openings Small penetration openings are defined as those having an inside diameter of less than 10 feet (2.5 times the reactor building nominal shell wall thickness). The local analysis of the shell in the area of small penetration openings is discussed in Sections 6.5 and 7.4 of BC-TOP-5-A. 3.8.1.4.2.3 Buttress and Tendon Anchorage Zones Analysis and design of tendon anchorage zones and reinforcement in buttresses are discussed in Section 6.6 of BC-TOP-5-A and in BC-TOP-7 and BC-TOP-8. 3.8-9 Rev. 0

WOLF CREEK 3.8.1.4.3 Creep, Shrinkage, and Cracking of Concrete In the design of the reactor building post-tensioning system, conservative values of creep and shrinkage for the concrete were utilized, based on past experience. The value used was verified by the evaluation of the tests performed on the concrete which was used in the reactor building shell. In establishing this value, the tests were performed on concrete that was used at each of the SNUPPS sites, and consideration was given to the differences in the environment between the test samples and the actual concrete in the structure. The moments, forces, and shears were obtained on the basis of an uncracked section for all load combinations. However, in sizing the reinforcing steel required, the concrete was not relied upon for resisting tension. Thermal moments were modified by a cracked section analysis, using analytical techniques. 3.8.1.4.4 Tangential Shear The design and analysis procedures for tangential shear are in accordance with Appendix C of BC-TOP-5-A. 3.8.1.4.5 Variation in Physical Material Properties In the design and analysis of the reactor building, consideration was given to the effects of possible variations in the physical properties of materials on the analytical results. The variation in physical properties were accounted for by using allowable stress levels, below ultimate strength, for design of the structure under full service and factored load conditions. 3.8.1.4.6 Steel Liner Plate and Anchors The analysis and design procedures utilized for the liner plate system are in accordance with BC-TOP-1 and Sections 6.8, 7.5, and Appendix C of BC-TOP-5-A. 3.8.1.4.7 Computer Programs The computer programs used in the analysis and design of the reactor building are described in Appendix 3.8A. 3.8.1.5 Structural Acceptance Criteria The fundamental acceptance criterion for the completed reactor building was successful completion of the Structural Integrity Test where measured responses were required to be within the limits 3.8-10 Rev. 1

WOLF CREEK predicted by analyses. The limits were based on test load combinations and code values for stress, strain, or gross deformation for the range of material properties and construction tolerances specified, as described in Section 3.8.1.6. The limits for allowable stresses and strains are given in Appendix C of BC-TOP-5-A and are compatible with nationally recognized codes of practice. In this way, the margins of safety associated with the design and construction of the reactor building are, as a minimum, the accepted margins associated with nationally recognized codes of practice. The Structural Integrity Test yielded information on both the overall response of the reactor building and the response of localized areas. This information, together with the test information documented in BC-TOP-7 and BC-TOP-8, provided direct experimental evidence that the containment structure can withstand the design internal pressure. The design and analysis methods, as well as the type of construction and construction materials, were chosen to allow assessment of the capability of the structure throughout its service life. Additionally, surveillance testing provides further assurances of the continuing ability of the structure to meet its design functions. 3.8.1.6 Materials, Quality Control, and Special Construction Techniques This section contains information relating to the materials, quality control program, and special construction techniques used in the fabrication and construction of the reactor building. 3.8.1.6.1 Concrete 3.8.1.6.1.1 Materials The cement was Type II, conforming to the Specification for Portland Cement (ASTM C150). The sum of tricalcium silicate and tricalcium aluminate did not exceed 58 percent. The cement contained no more than 0.60 percent by weight of alkalies calculated as Na2O plus 0.658 K2O. The limitation of the alkali content of the cement may have been waived provided that the aggregates passed required laboratory tests and had no history of alkali aggregate incompatibility. Certified copies of material test reports showing the chemical composition and physical properties were obtained for each load of cement delivered. 3.8-11 Rev. 0

WOLF CREEK 3.8-12 All aggregates conformed to the Specification for Concrete Aggregate (ASTM C33). For concrete with 1-1/2 inch maximum size aggregate, the coarse aggregate was a combination of 1-1/2 inch and 3/4 inch aggregate. The potential reactivity of the aggregate was established in accordance with ASTM C289. A petrographic examination of the aggregate was performed in accordance with ASTM C295. In addition to the specified gradation, the fine aggregate (sand) had a fineness modulus of not less than 2.5 nor more than 3.1. During normal concrete production, at least four of five successive test samples did not vary more than 0.20 from the average. Coarse aggregate was rejected if the loss, when subjected to the Los Angeles Abrasion Test (ASTM C131) using Grading A, exceeds 40 percent by weight at 500 revolutions. The particle shape of the coarse aggregate was generally rounded or cubicle and did not contain thin, flat, and elongated particles in excess of 15 percent by weight in any nominal size group. A thin, flat, and elongated particle is defined as a particle having a maximum dimension in excess of four times the minimum dimension. Water and ice used in mixing concrete were free of injurious amounts of oil, acid, alkali, organic matter, and other deleterious substances and were tested in accordance with AASHTO T-26. When tested according to AASHTO T-26, the water did not cause unsoundness in the autoclave test, and did not change the final setting time by more than 1 hour, and the 7- and 28-day compressive strength of ASTM C109 cubes was not reduced by more than 10 percent when compared with results obtained with distilled water. Water was tested for pH, chlorides, and sulfates and did not contain more than 250 ppm of chlorides as Cl, nor more than 1,000 ppm of sulfates as SO4. The concrete also contained an air-entraining admixture and a water-reducing admixture. The air-entraining admixture was in accordance with the Specification for Air Entraining Admixtures for Concrete (ASTM C260). It was capable of entraining 3 to 6 percent air, was completely water soluble, and was completely dissolved when it enters the batch. The water reducing and retarding admixture conformed to the Specification for Chemical Admixtures for Concrete (ASTM C494), Types A and D. Type A was used when concrete temperature was below 70 F. Type D was used when concrete temperature was 70 F and above, except for floor slabs where its use was optional. Pozzolans, if used, conformed to the Specification for Fly Ash and Raw or Calcined Natural Pozzolans for Use in Portland Cement Concrete (ASTM C618). 3.8-12 Rev. 0

WOLF CREEK 3 8.1.6.1.2 Concrete Mix Design Structural concrete used in the construction of the reactor building shell and dome has a compressive strength, fc', of 6,000 psi at 90 days. Structural concrete used in the construction of the reactor building base slab, reactor cavity, instrumentation tunnel, and tendon access gallery has a compressive strength, fc', of 5,000 psi at 90 days. Structural specifications were prepared to identify the required concrete material properties and tests. Concrete conforms to the Specification for Ready-Mixed Concrete (ASTM C94), as modified herein. In lieu of the maximum water content specified in ASTM C94, the concrete was mixed so as to be placed at the specified slumps. The mix proportions were established in accordance with Paragraph 3.8 of ACI 301, Method 1. The required average strength was in accordance with Paragraph 3.8.2.3 of ACI 301. In lieu of the requirements in Paragraph 18.2 of ASTM C94, conformance to ASTM E329, with the exception of Paragraph 4 as it pertains to concrete, was required. On December 12 and 13, 1977, the WCGS reactor building base mat was placed as a monolithic pour of approximately 6600 cubic yards of concrete. At the end of the 90 day curing period, some of the sixty-six sets of concrete cylinders tested exhibited strengths below the specified concrete strength of 5000 psi. Tests conducted by the licensees (Portland Cement Association, 1979) and by the Corps of Engineers (Brown, 1979) led to the conclusion that the low strength test results were invalid and probably due to one or more failures to follow current standards of good practice in testing. The NRC was unable to conclude that the low 90-day strengths were attributable to testing machine factors or testing conditions. The NRC required the licensees to complete a reanalysis of the base mat based upon the strength of the concrete determined according to ACI Standard 318-71 criteria from the 90-day cylinder test results (4460 psi). The reanalysis indicated that the WCGS base mat met all design criteria at the 4460 psi concrete strength. The NRC, after an evaluation of the test and reanalysis results, concluded (Varga, 1979) that: 3.8-13 Rev. 0

WOLF CREEK 

     "...the base mat concrete strength has not retro-gressed, that the strength of the base mat meets the original design criteria in the Wolf Creek PSAR, and that the mat will withstand the specified design loads and loading combinations without impairment of the structural integrity or its safety function."

3.8.1.6.1.3 Examination During construction, concrete materials were regularly sampled and tested to ensure quality control. Table 3.8-1 shows the procedures used and the frequency of testing for the concrete materials used. 3.8.1.6.1.4 Placement Conveying and placement of concrete were performed in accordance with the following codes and standards to the extent described below:

a. ACI 301 - Specifications for Structural Concrete for Buildings, Chapters 4, 6, 8, 9, 10, 11, 12, 13, 14, and 15 were used, except as noted below:
1. In lieu of the requirements for the removal of forms specified in Paragraph 4.5.4, the following applies:

Forms for columns, walls, sides of beams, slabs, girders, and other parts not supporting the weight of the concrete were removed as soon as practicable in order to avoid delay in curing and repairing surface imperfections. Wood forms or insulated steel forms for members over 3 feet in thickness were stripped within 24 hours or kept in place for a minimum of 7 days. If forms were stripped within 24 hours, the surfaces were cured by moist curing or membrane curing as specified in ACI 301, Chapter 12.

2. In lieu of the requirements for the placing of mass concrete specified in Paragraph 14.4.1, the following applied:

Slump was specified for particular locations and degree of congestion rather than holding a 2-inch maximum. An inadvertency margin for maximum slump above the stated maximum average value was included in the job standards. 3.8-14 Rev. 0

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3. In lieu of the requirements for curing and protection of mass concrete specified in Paragraph 14.5.1, the following applied:

The minimum curing period was 7 days for heavily reinforced massive sections.

b. ACI 304, Recommended Practice for Measuring, Mixing, Transporting, and Placing Concrete, Chapters 5 and 6, were used without exception.
c. ACI 305, Recommended Practice for Hot-Weather Concreting, was used without exception.
d. ACI 306, Recommended Practice for Cold-Weather Concreting, was used without exception.
e. ACI 318, Building Code Requirements for Reinforced Concrete, Chapters 5 and 6 were used, except as noted below:

In place of Paragraph 6.3.2.4, the specific provisions of the applicable codes that govern the system of which the embedded piping is a part applied. Examples of such applicable codes are: for nuclear piping, ASME Boiler and Pressure Vessel Code, Section III; and for nonnuclear piping, ANSI B31.1, Power Piping. The purpose of ACI 318, Paragraph 6.3.2.4, is to avoid the removal of concrete if a leak is developed in the pipe wall or joints. The testing requirements of Paragraph 6.3.2.4 were valid in the case of noncode piping; they were not valid for the piping that is required to conform to the acceptable industry codes such as the ASME B&PV Code for nuclear piping, ANSI B31.1 for nonnuclear power piping, and the applicable state or local plumbing codes. Where no such codes or code cases govern a particular pipe embedded in structural concrete, the requirements of ACI 318, Paragraph 6.3.2.4, were implemented.

f. ACI 347, Recommended Practice for Concrete Formwork, was used without exception.
g. ACI SP2, Manual of Concrete Inspection, applicable provisions relating to conveying and placement were used without exception.
h. ASTM C94, Specification for Ready-Mixed Concrete, applicable provisions relating to conveying and placement were used without exception.

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WOLF CREEK The placement of concrete complied with the requirements of Regulatory Guide 1.55 to the extent described in Appendix 3A. No aluminum pipe or other conveying equipment containing aluminum that would be in contact with fresh concrete was used for conveying concrete to the point of placement. 3.8.1.6.2 Reinforcing Steel and Splices 3.8.1.6.2.1 Materials Reinforcing bars for concrete are deformed bars meeting the requirements of the Specification for Deformed and Plain Billet-Steel Bars for Concrete Reinforcement (ASTM A 615), Grade 60. For each heat or mill shipment, whichever is less, certified copies of the material test reports covering the chemical and mechanical properties of the reinforcing bars were obtained. Mechanical spices, when used, consisted of T-series and B-series Cadweld-type splices. Tubing used for splice sleeves conforms to the Specification for Seamless Carbon and Alloy Mechanical Tubing (ASTM A519), Grades 1018 or 1026. Certified copies of material test reports showing the results of the chemical and mechanical tests of material from each lot of splice sleeves were obtained. In addition, certification was obtained for each lot showing for each lot that the chemical composition of the powdered metal and the chemical and mechanical properties of the resulting filler material conformed to the manufacturer's standards. 3.8.1.6.2.2 Examination During fabrication and construction, reinforcing steel and mechanical splices were regularly sampled and tested to ensure quality control. The examination methods, frequency, and acceptance standards in Regulatory Guide 1.10 for mechanical splices and Regulatory Guide 1.15 for reinforcing steel were used. Refer to Appendix 3A for a description of the extent of compliance with these regulatory guides. 3.8.1.6.2.3 Erection Tolerances The reinforcing steel was placed in accordance with the tolerances specified in Paragraph 7.3.2 of ACI 318, except as noted below:

a. In place reinforcing steel cover tolerances for the containment were within the following limits:

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WOLF CREEK Base slab -0", +1 1/2" Exterior walls -0", +1 1/2" Dome -1", +1" Exterior wall tolerances were maintained, except for local areas adjacent to some recesses on the exterior surface of the containment shell where a gradual sweep of the continuous reinforcing steel to clear the recesses could have resulted in these cover tolerances being exceeded. However, the resulting cover was within the design allowable specified in BC-TOP-5-A, Appendix C, except for the two electrical penetration banks. The electrical penetration banks, centered at azimuth 222 -30', El. 2035'-3" and azimuth 319 -30', El. 2035'-3" have the outside face of concrete recessed 8 inches. The two banks are approximately 15 feet (vertical) by 48 feet (horizontal) and 15 feet by 39 feet, respectively. The transition zone where the continuous reinforcing sweeps gradually inward to clear the recess extends as much as 16 feet-8 inches away from the outside edge of the recess. Although the reinforcing steel in this area was generally within the limits indicated above, there are a few instances where, including placing tolerances, the cover could be as much as 13-3/4 inches.

b. Cadwelds and other connectors were not considered as reinforcing steel.
c. In no case was the cover reduced by more than one-third of the minimum specified design cover.
d. Minimum splice lengths and minimum embedment lengths were maintained to a tolerance of minus 2 inches. These minimum lengths may have been exceeded without limit, provided that the other requirements for cover and clearances were not violated.
e. The variation in spacing was r 2 bar diameters, except that the minimum clear distance specified in Paragraphs 3.3.2 and 7.4 of ACI 318 was maintained. The total number of bars in any nominal 10-foot segment was maintained.

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f. For longitudinal location of bends and ends of bars that were mechanically spliced, a tolerance of minus 2 inches at the discontinuous end of the member in which the splice occurs was acceptable. Conversely, the cover for this situation may have been increased by 2 inches.

3.8.1.6.3 Prestressing System 3.8.1.6.3.1 Materials The prestressing system consists of load carrying and nonload carrying components. The load carrying components include the prestressing wires which make up the tendons, and anchorage components composed of bearing plates, anchor heads, and shims. Nonload carrying components include the tendon sheathing (including trumpet assemblies, couplers, vent and drain nipples, and other appurtenances), and corrosion prevention material. The prestressing wire is cold-drawn, of the intermediate relaxation or stabilized type, and conforms to the Specification for Uncoated Stress-Relieved Wire for Prestressed Concrete (ASTM A421), Type BA. The materials used for the anchorage components are compatible with the tendon system. Tendon sheathing consists of galvanized, spiral-wrapped, semirigid, corrugated tubing conforming to the requirements of the Specification for Steel Sheet, Zinc-Coated (Galvanized) by the Hot-Dip Process, Lock Forming Quality (ASTM A527) or the Specification for Steel Sheet, Zinc-Coated (Galvanized) by the Hot-Dip Process, Drawing Quality (ASTM A528), 22-gauge cold rolled carbon steel. Trumpet material conforms to the Specification for Electric Resistance-Welded Carbon and Alloy Steel Mechanical Tubing (ASTM A513), Grades MT1010 to 1029, or the Specification for Welded and Seamless Steel Pipe (ASTM A53), Grade B. Couplers and mending sections conform to ASTM A527 or ASTM A528. Vent and drain nipples consist of noncorrosive metal galvanized pipe, or equal. After fabrication, a thin film of temporary corrosion protection material was applied to the prestressing steel. This material is compatible with the permanent corrosion prevention material and is removable with the use of a nonchlorinated petroleum solvent to permit the installation of attached anchorages. The permanent corrosion-prevention coating applied to tendons is a petrolatum or microcrystalline wax-base material, containing additives to enhance the corrosion-inhibiting and wetting properties, as well as to form a bond with the tendon steel. The coating has the following properties for the lifetime of the structure and for the anticipated range of the temperature. 3.8-18 Rev. 0

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a. Freedom from cracking and brittleness
b. Continuous self-healing film over the coated surfaces
c. Chemical and physical stability
d. Nonreactivity with the surrounding and adjacent materials, such as concrete, tendons, and ducts
e. Moisture displacing characteristic Each batch of coatings was analyzed for the presence of water soluble chlorides, nitrates, and sulphides.

3.8.1.6.3.2 Examination Prior to construction, a number of tests were performed on the load-carrying components of the prestressing system to ensure that the performance requirements of the system were satisfied and quality control was maintained. In addition to the tests described below, an in-service surveillance program of the prestressing system is carried out, as discussed in Section 3.8.1.7. All load-carrying components were subject to tensile tests. Materials produced to an ASTM specification were sampled and tested as required by that specification. Materials not produced to an ASTM specification were sampled and tested at the rate of one test for every 20 tons, or fraction thereof, produced from each heat of steel. The tensile strength, yield strength, elongation, and other pertinent data were reported on the Certified Materials Test Report. The stress-relaxation properties of the wire, determined in accordance with the Recommended Practice for Stress-Relaxation Tests for Materials and Structures (ASTM E328), were obtained from the manufacturer for a minimum of three relaxation tests of 1,000 hours duration. In addition to those required by ASTM E328, the manufacturer's reports of the test included detailed test method, initial stress, final stress, test time, temperature limits, and mathematical tools used to interpret the test results. Anchorage components were subjected to hardness tests. For anchorhead assemblies, the Method of Tests for Rockwell Hardness and Rockwell Superficial Hardness of Metallic Materials (ASTM E18) and the Method of Test for Brinell Hardness of Metallic Materials (ASTM ElO) were conducted on 10 percent of the parts from each lot 3.8-19 Rev. 0

WOLF CREEK (after heat treatment) on a random basis. If the hardness requirement was not met by any single part relative to acceptance standards set by design documents, then all parts from the lot were tested. Only those parts meeting the requirements were used. The following tests were performed by the tendon manufacturer in order to qualify his system for use in the reactor building:

a. A static tensile test was conducted to destruction to obtain information on yield strength, tensile strength, and compliance with the following performance requirements:

A full capacity tendon complete with anchorages will develop an ultimate strength equal to 100 percent of the minimum specified ultimate tensile strength of the prestressing steel, without exceeding the anticipated set of the anchorage elements. The total elongation under ultimate load of the tendon will not be less than 2 percent, measured in a minimum gauge length of 100 inches.

b. A high-cycle dynamic tensile test was conducted to ensure that the tendon can withstand, without failure, 500,000 cycles of stress variation from 60 to 66 percent of the tendon minimum specified ultimate tensile strength. A load cycle is defined as an increase from the lower load to the higher load and return. This test was performed on specimens having at least 10 percent of the full-sized prestressing steel area of one production tendon.
c. A low-cycle dynamic tensile test was conducted to ensure that the tendon could withstand, without failure, 50 cycles of stress variation from 40 to 80 percent of the tendon minimum specified ultimate tensile strength. This test was performed on specimens having at least 10 percent of the full sized prestressing steel area of a production tendon.

3.8.1.6.3.3 Erection Tolerances The following are the erection tolerances from the theoretical location of the sheathing in the cylindrical wall:

a. Vertical sheathing r2 inches in the circumferential direction 3.8-20 Rev. 0

WOLF CREEK r 1/2 inch in the radial direction when measured from the liner plate or r1 1/2 inches when measured from the reactor building theoretical centerline r6 inches in elevation for points of tangency between the curved and straight sections r2 inches per 10 feet - 0 inches for variation from the plumb, not cumulative

b. Horizontal sheathing r2 inches in elevation r1/2 inch in the radial direction when measured from the liner plate or r1 1/2 inches when measured from the reactor building theoretical centerline r6 inches in the circumferential direction for points of tangency between the curved and straight sections
c. Requirements at penetrations:

The general criterion for placing sheathing in the area of penetrations was to achieve a smooth configuration without sharp bends which would impair the insertion of the tendons or create undesirable loading combinations. The sheathing was also placed to meet the clear distance between any point on the sheathing and a penetration nozzle as well as the minimum distance between sheathing as given on the tendon placement drawings. The following are the erection tolerances from the theoretical location of the sheathing in the dome:

a. Meridional sheathing r2 inches in the circumferential direction r1/2 inch in the radial direction when measured from the liner plate or r1.5 inches when measured from the reactor building theoretical centerline r6 inches in the meridional direction for points of tangency between the curved and straight sections
b. Horizontal sheathing r2 inches in the meridional direction 3.8-21 Rev. 0

WOLF CREEK r1/2 inch in the radial direction when measured from the liner plate or r1.5 inches when measured from the reactor building theoretical centerline r6 inches in the circumferential direction for points of tangency between the curved and straight sections 3.8.1.6.4 Liner Plate System The reactor building is lined with welded steel plates, as outlined below, to ensure low leakage. These materials were chosen on the basis that they have sufficient strength and ductility to resist the expected strains from design criteria loading and, at the same time, preserve the required leaktightness of the reactor building. They are readily weldable by all commercially available arc and gas welding processes. 3.8.1.6.4.1 Materials The 1/4-inch-thick liner plate material conforms to the requirements of the Specification for Low and Intermediate Tensile Strength Carbon Steel Plates for Pressure Vessels (ASME SA 285), Grade A. Thickened liner plates, ranging from 1/2 inch to 2 inches in thickness, were used at penetrations, brackets, and embedded assemblies and conform to the requirements of the Specification for Carbon Steel Plates for Pressure Vessels for Moderate and Lower Temperature Service (ASME SA516), Grade 70. In the event that significant loads are to be transmitted through the thickness dimension of the liner, nondestructive tests were performed to determine the capability of the liner materials used in these locations. Materials for penetration sleeves conform to the requirements of the following specifications and are impact tested in accordance with Paragraph NE-2300 of Section III of the ASME Code at a temperature no greater than 0 F:

a. Seamless penetration sleeves conform to the Specification for Seamless and Welded Steel Pipe for Low-Temperature Service (ASME SA333), Grade 6.
b. Welded penetration sleeves conform to the Specification for Electric-Fusion Welded Steel Pipe for High Pressure Service (ASME SA155), KCF70, or pipe in accordance with ASME Code Class MC Vessels.

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c. Penetration sleeve reinforcing plates conform to the Specification for Carbon Steel Plates for Pressure Vessels for Moderate and Lower Temperature Service (ASME SA516), Grade 70.
d. Penetration rods conform to the Specification for Carbon Steel Forgings for Piping Components (ASME SA105).

Material used for the liner plate anchors and embedments conform to the Specification for Structural Steel (ASTM A36) or the Specification for Pressure Vessel Plates, Carbon Steel, for Moderate- and Lower-Temperature Service (ASTM A516), Grade 70. Materials used for test piping, fittings, plates, and shapes conform to the following:

a. Specification for Welded and Seamless Steel Pipe (ASTM A53)
b. Specification for Forgings, Carbon Steel, for Piping Components (ASTM A105)
c. Specification for Forged or Rolled Steel Pipe Flanges, Forged Fittings, and Valves and Parts for General Service (ASTM A181)
d. Specification for Piping Fittings for Wrought Carbon Steel and Alloy Steel for Moderate and Elevated Temperatures (ASTM A234)
e. Specification for Low and Intermediate Tensile Strength Carbon Steel Plates for Pressure Vessels (ASME SA285)
f. Specification for Pressure Vessel Plates, Carbon Steel, for Moderate- and Lower-Temperature Service (ASTM A516),

Grade 70

g. Specification for Structural Steel (ASTM A36)
h. Specification for Low and Intermediate Tensile Strength Carbon Steel Plates of Structural Quality (ASTM A283)

Materials used for Cadweld sleeves conform to the Specification for Seamless Carbon and Alloy Mechanical Tubing (ASTM A519), Grades 1018 or 1026. Materials used for pipe anchors conform to the Specification for Seamless Carbon Steel Pipe for High Temperature Service (ASTM A106), Grade B. 3.8-23 Rev. 0

WOLF CREEK Welding electrode materials were selected on the basis of the welding process used and the type of materials to be joined and in accordance with the requirements of ASME Boiler and Pressure Vessel Code, Section III. Written control procedures for welding materials were required, which defined the measures used to control the use of the materials throughout all welding operations. Such controls provided for the complete traceability of welding materials used in the liner plate seams to all tests and examinations and to the welder. Materials for machine bolts conform to the Specification for Carbon Steel Externally and Internally Threaded Standard Fasteners (ASTM A307). Materials for high-strength bolts conform to the Specification for High Strength Bolts for Structural Steel Joints, Including Suitable Nuts and Plain Hardened Washers (ASTM A325). Materials for weld studs conform to the Specification for Steel Bars, Carbon, Cold-Finished, Standard Quality (ASTM A108), Grades 1010, 1015, 1016, 1017, 1018, or 1020. Materials used for weld backing strips were compatible with the materials welded. Where ASTM specifications are referenced, equivalent ASME materials may have been used. Certificates of Compliance were obtained from the manufacturer for bolts, weld studs, weld backing strips, and welding fluxes. Certified copies of material test reports were obtained for all other liner plate system materials which include the actual results of all required chemical analyses, physical tests, mechanical tests, and examinations. 3.8.1.6.4.2 Examination Nondestructive examination of the liner plate welds complies with Regulatory Guide 1.19 to the extent described in Appendix 3A. 3.8.1.6.4.3 Erection Tolerances The liner plate and penetration assemblies are erected to the following tolerances requirements:

a. General Liner Plate
1. The radial location at any point on the liner plate shall not vary from the design radius by more than r3 inches. Measurements shall be made at 30-degree spacings for each 10 feet of rise.

3.8-24 Rev. 1

WOLF CREEK The radius of the hemispherical dome for all elevations between the as-built springline and 15 feet above it shall be within r3 inches of the design radius. The radius of the hemispherical dome for all points above a plane parallel to and 15 feet higher than the plane of the as-built springline shall not exceed the design radius plus 8 inches or be less than the design radius minus 12 inches.

2. Plates to be joined by butt welding shall be matched and retained in position during the welding.

Misalignment in completed joints shall not exceed the limits shown in Table 3.8-2.

3. A 15-foot-long template curved to the required radius shall not show deviations of more than 1 inch when placed against the completed surface of the shell within a single plate section and not closer than 12 inches at any point to a welded seam. When the template is placed across one or more welded seams, the deviation shall not exceed 1 1/2 inches. The effect of change in plate thickness or of weld reinforcement shall be disregarded when determining deviations.
4. A 15-inch-long template curved to the required radius shall not show deviations of more than 1/8 inch inward or 3/8 inch outward when placed against the completed surface of the shell within a single plate section and not closer than 12 inches to a weld seam.

A 30-inch-long template, curved to the required radius, shall not show deviations of more than 1/4 inch when placed against the completed surface of the shell within a single plate section.

5. The deviation from the true vertical for any 10-foot plate shall not vary by more than 3/4 inch. Plates of other depths shall be checked for linearly varying tolerances. The overall out-of-plumbness of the shell shall not exceed 3 inches.
6. A 10-foot straightedge held vertically shall not show deviations greater than r3/4 inch in the horizontal direction between seam welds.

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7. Local bends that deviate from the design radius or a vertical straightedge by an offset of more than 1/2 inch in 1 foot shall not be accepted. The template used to measure the local deviations shall be only 1 to 2 feet longer than the area of the deviation itself.
b. Penetration Assemblies
1. Items 1, 3, 5, and 7 in part "a" above also control the tolerance requirements for penetration assemblies.
2. Alignment of the axes of penetrations, as erected, shall not vary from the alignment shown on the design drawings by more than 2 degrees for pipes 12 inches in diameter or less and by more than one degree for pipes over 12 inches in diameter. Individual penetrations and penetration assemblies shall be located within r1 inch of their design elevations and circumferential locations, at the cylindrical shell.

3.8.1.6.5 Quality Control In addition to the quality control measures discussed in Sections 3.8 1.6.1, 3.8 1 6.2, 3.8.1.6.3, and 3.8 1 6.4, the construction quality control program was discussed in the Quality Assurance Design and Construction Manual which was contained in the PSAR. 3.8.1.6.6 Special Construction Techniques The reactor building is constructed of concrete and steel, using proven methods common to heavy industrial construction. No special, new, or unique construction techniques were used. 3.8.1.7 Testing and Inservice Surveillance Requirements 3.8.1.7.1 Structural Integrity Test Following construction, the reactor building was proof-tested at 115 percent of the design pressure. During this test, deflection measurements and concrete crack inspections were made to determine that the actual structural response is within the limits predicted by the design analyses. The test procedure complied with the requirements of Regulatory Guide 1.18 to the extent described in Appendix 3A. The associated leak rate test procedure is described in Section 6.2.6. Section 9.0 of BC-TOP-5-A also describes test results obtained using a typical procedure as well as those obtained from early tests where a substantial amount of strain information was collected. 3.8-26 Rev. 18

WOLF CREEK 3.8.1.7.2 Long-Term Surveillance The long-term surveillance program consists of evaluating the general conditions of the post-tensioning system. Data on wire corrosion levels and tendon lift-off forces are obtained and analyzed. The surveillance tendons are designated as part of the inservice inspection program which conforms with Subsection IWE and IWL of Section XI, Division I of the ASME Boiler and Pressure Vessel Code as limited and modified by 10 CFR 50.55a. DISCUSSION: The post-tensioning system is described in Section 3.8.1.1.2. Details of the Inservice Tendon Surveillance Program are provided in BC-TOP-5-A, Section 9.3, with the exception of sections 9.3.2.2.2 and 9.3.4.9 for which other provisions are made in accordance with 10 CFR 50.55a as modified with the following clarification: There is unobstructed access to all tendons with the exception of two tendons at el. 2073, Azimuth 281 degrees, which are blocked by the Auxiliary Building Roof. Four other tendons (two at approximate el. 2026 and two approximate els. 2047) at Azimuth 281 degrees are accessible for visual inspection only due to close proximity of elevated slabs. In addition, all tendons terminating at the C buttress above elevation 2079 have been deleted from the surveillance population due to lack of special equipment required for access. If it does become necessary to access these tendons in the future, the required equipment can be obtained. See Technical Specification 5.5.6 and Technical Requirement Manual TR 3.6.1. 3.8.2 CONTAINMENT SYSTEM STEEL ITEMS This section describes the major penetrations and portions of penetrations intended to resist pressure which are not backed by structural concrete. 3.8.2.1 Description of Steel Items The steel items that are part of the containment pressure boundary include access openings, such as the equipment hatch and personnel hatches, piping penetration sleeves, fuel transfer tube penetration sleeves, electrical penetration sleeves, and the purge line penetration sleeves. 3 8.2.1.1 Equipment and Personnel Access Hatches and Penetration Sleeves The equipment hatch, shown in Figure 3.8-44, is a welded steel assembly with a double-gasketed, flanged, and bolted cover. Provision is made for leak testing of the flange-gasket combination by pressurizing the space between the gaskets. One personnel hatch and one auxiliary hatch, both of which are welded steel assemblies, are provided as shown in Figures 3.8-45 and 3.8-46. Each hatch has two doors with double gaskets in series. In order to assure leaktightness, the space between the gaskets are normally pressurized. The doors are mechanically interlocked to ensure that one door cannot be opened unless the second door is sealed. The interlock can be deliberately overridden by the use of special tools and procedures. Each door is equipped with quick-acting valves for equalizing the pressure across the doors. The doors are not operable unless the pressure is equalized. Pressure equalization is possible from every point at which the associated door can be operated. The valves for the two doors are properly interlocked so that only one valve can be opened at one time and only when the opposite door is closed and 3.8-27 Rev. 18

WOLF CREEK sealed. Each door is designed so that, with the other door open, it will withstand and seal against design and testing pressure of the containment vessel. There is visual indication outside each door showing whether the opposite door is open or closed. Provision is made outside each door for remotely closing and latching the opposite door so that in the event that one door is accidentally left open it can be closed by remote control. The access hatch barrels have nozzles which permit pressure testing of the hatch at any time. The hatches are protected from tornado missiles by enclosure structures or shields. A moveable radiation and missile shield is provided on the outside of the reactor building to provide additional tornado missile protection for the inner equipment hatch door. The inner equipment hatch door is bolted in place with 20 bolts in Modes 1, 2, 3 and 4. The missile shield is bolted in place in Modes 1, 2, 3 when RCS pressure is greater than 1000 psig. Subsequently the missile shield may be unbolted and moved away from the equipment hatch opening provided it can be returned to its design location, without bolting, during adverse weather conditions. In Modes 5 and 6, the missile shield is no longer required because the inner equipment hatch door with 6 bolts installed provides adequate missile protection to safety related equipment inside the containment building. Administrative controls ensure that the equipment hatch door is in place in Modes 5 and 6 during adverse weather conditions that could result in the generation of tornado driven missiles. The personnel hatch is enclosed within the auxiliary building. The auxiliary hatch is enclosed within an exterior tornado - resistant concrete structure. The personnel & auxiliary access hatch barrels are designated as ASME Section III, Class MC components. The personnel and auxiliary access hatch barrels are designated as ASME Section III, Class MC components. The hatch penetration sleeves project into the reactor building and are used to support the hatches. These items are made from carbon steels and conform to the requirements of ASME Section III, Subsection NE. 3.8.2.1.2 Piping Penetration Sleeves Piping penetrations are divided into three general groups:

a. Type 1: Flued head penetrations used for most high energy piping. Examples of Type 1 penetrations are the main steam and main feedwater lines.
b. Type 2: Closure plate penetrations used for some high energy, all moderate energy, and all low energy general piping. The use of this type of penetration for high energy piping is limited to only those cases where an analysis based on combination of pressure, temperature, and line size has demonstrated the adequacy of the design.
c. Type 3: Spare penetrations reserved for future use or small access penetrations.

Typical details of the three types of piping penetrations are shown in Figure 3.8-47. Type 1 piping penetrations consist of the following major steel items: 3.8-28 Rev. 20

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a. Process Pipe: This pipe, which is made of welded or seamless carbon or stainless steel and is welded to the flued head, conforms to the requirements of ASME Section III, Subsection NC.
b. Flued Head: This item is made from forged carbon or stainless steel and conforms to the requirements of ASME Section III, Subsection NC. It is designed to contain the full pressure of the process fluid and full reactor building pressure in parts adjoining the pipe sleeve.

The connecting process pipes and the flued heads are designed and analyzed to be capable of carrying loads resulting from the failure of the process pipe, as described in Sections 3.6 and 3.9(B).

c. Pipe Sleeve: This steel item consists of the portion which projects into the reactor building and supports the flued head. It conforms to ASME Section III, Subsection NE, except that authorized inspection and stamping are not performed.

Type 2 piping penetrations consist of the following major steel items:

a. Process Pipe: This pipe, which is made of welded or seamless carbon or stainless steel and is welded to the closure plate, conforms to the applicable requirements of ASME Section III, Subsection NC.
b. Closure Plate: This item is made from carbon or stainless steel plate and conforms to the requirements of ASME Section III, Subsection NC.
c. Pipe Sleeve: This steel item consists of the portion which projects into the reactor building and supports the closure plate. It conforms to ASME Section III, Subsection NE, except that authorized inspection and stamping are not performed.

Type 3 spare penetrations consist of the following major items:

a. Solid Closure Plate (mechanical or welded) or Pipe Cap:

This item is made from carbon steel and conforms to the requirements of ASME Section III, Subsection NE or NC.

b. Pipe Sleeve: This steel item consists of the portion which projects into the reactor building. It conforms to ASME Section III, Subsection NE, except that authorized inspection and stamping are not performed.

3.8-29 Rev. 6

WOLF CREEK 3.8.2.1.3 Fuel Transfer Tube Penetration Sleeve The fuel transfer tube penetration is provided to transfer fuel between the refueling canal and the fuel storage pool during refueling operations of the reactor. The penetration consists of a 20-inch-diameter stainless steel pipe installed inside a 26-inch sleeve. The steel sleeve which projects into the reactor building conforms to ASME Section III, Subsection NE, except that authorized inspection and stamping are not performed. The inner pipe acts as the transfer tube. The sleeve is designed to provide integrity of the reactor building, allow for differential movement between structures, and prevent leakage through the fuel transfer tube in the event of an accident. Figure 3.8-48 shows details of the fuel transfer tube penetration. 3.8.2.1.4 Electrical Penetration Sleeves Steel sleeves, which form a portion of the containment pressure boundary, are provided for electrical/fiber optic penetrations. The electrical/fiber optic penetration header plates are designed as discussed in Section 8.1. The sleeve consists of the portion which projects out of the reactor building and supports the electrical/fiber optic assembly. It conforms to ASME Section III, Subsection NE, except that authorized inspection and stamping are not performed. Figure 3.8-49 shows the details of the electrical penetrations. 3.8.2.1.5 Purge Line Penetration Sleeves The steel sleeves, which are embedded in the reactor building wall concrete, are welded to the purge line piping and form a part of the ASME Section III, Class 2 purge line piping system, as shown in Figure 3.8-50. The sleeves conform to ASME Section III, Subsection NC. 3 8.2.2 Applicable Codes, Standards, and Specifications The following codes, regulations, standards, and specifications are utilized in the design of the steel portions of the reactor building that are intended to resist pressure but are not backed by structural concrete. 3.8.2.2.1 Regulations

a. 10 CFR 50, "Licensing of Production and Utilization Facilities" 3.8-30 Rev. 14

WOLF CREEK 3.8.2.2.2 Codes

a. ASME Boiler and Pressure Vessel Code - 1974 Edition and Later Section II - Material Specifications Section III, Division 1 - Nuclear Power Plant Components Section V - Nondestructive Examination Section IX - Welding and Brazing Qualifications
b. Acceptable ASME Code cases per Regulatory Guides 1.84 and 1.85, as addressed in Appendix 3A 3.8.2.2.3 Standards and Specifications Nationally recognized industry standards, such as those published by the ASTM and IEEE, were used whenever possible to define material properties, testing procedures, fabrication, and construction methods. Applicable ASTM standard specifications for materials are those permitted by Article NE-2000 of Section III of the ASME Code. Applicable ASTM standard specifications for nondestructive methods of examination are those referenced in Appendix X, Article X-3000 of Section III of the ASME Code.

Structural specifications were prepared to cover the areas related to the design of steel portions of the containment pressure boundary. These specifications were prepared specifically for the SNUPPS Project (WCGS and Callaway). These specifications emphasized the important points of the industry standards for these items and reduce the options that would otherwise be permitted by the industry standards. These specifications covered the following areas:

a. Equipment and personnel access hatches
b. Piping penetration sleeves
c. Fuel transfer tube penetration sleeve
d. Electrical penetration sleeves
e. Purge line penetration sleeves 3.8.2.2.4 Design Criteria
a. 10 CFR 50, Appendix A - General Design Criteria for Nuclear Power Plants (Compliance is discussed in Section 3.1) 3.8-31 Rev. 0

WOLF CREEK GDC 2, 4, 16, 50, and 53

b. 10 CFR 50, Appendix J - Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors
c. Bechtel Power Corporation topical reports, as referenced in Section 1.6 3.8.2.2.5 NRC Regulatory Guides NRC Regulatory Guides 1.29, 1.57, 1.60, 1.61, 1.63, 1.84, and 1.85 were applicable to the design and construction of the steel portions of the reactor building that are intended to resist pressure but are not backed by structural concrete. Specific editions and the extent of compliance with these guides is discussed in Appendix 3A.

3.8.2.3 Loads and Loading Combinations 3.8 2.3.1 Dead Loads (D) The dead loads consist of the following typical loads:

a. Weight of the steel item
b. Weight of attached items Weight of electrical connections, mechanisms, ladders, and platforms supported by the containment vessel shell 3.8.2.3.2 Live Loads (L)

The live loads consist of the following typical loads:

a. Live load on the personnel access hatch floor of 200 pounds per square foot
b. Operating fluid weight in attached piping
c. Live load on the equipment hatch floor, using an AASHO (American Association of State Highway Officials) HS 44 loading 3.8.2.3.3 Test Pressure Load (Pt)

The structure is designed for a structural integrity test pressure of 69 psig. 3.8-32 Rev. 0

WOLF CREEK 3.8.2.3.4 Test Temperature Thermal Load (Tt) The thermal load associated with a temperature of 100 F is considered as a design basis for the structural integrity test. Testing may proceed at any temperature below this. 3.8.2.3.5 Thermal Loads (To, Te, Ta)

a. Thermal loads produced by the presence of radial and axial temperature gradients during startup, normal, and shutdown conditions (To)
b. Thermal conditions causing external pressure (Te)
c. Thermal conditions generated by the postulated DBA, including To(Ta) 3.8.2.3.6 Pipe Loads (Ro, Re, Ra)

The following pipe loads, determined in accordance with procedures described in Section 3.9, were utilized in the design of steel items:

a. Pipe reactions produced during startup, normal, or shutdown conditions (Ro)
b. Pipe reactions under thermal conditions, causing external pressure (Re)
c. Pipe reactions under thermal conditions generated by the postulated DBA, including Ro(Ra) 3.8.2.3.7 Seismic Loads (E, E')

The seismic loads used in the dynamic analysis of the steel items were developed by the use of either a response spectra or time history. The development of this response spectra and/or time history for the SSE and the OBE is discussed in Section 3.7(B) and (N). 3.8.2.3.8 External Pressure Load (Pe) The design external pressure differential is 3 psig. Refer to Section 3.8.1.3 for a description of this load. 3.8.2.3.9 Pressure Loads (Pa) Pressure equivalent static load generated by the postulated design basis accident. 3.8-33 Rev. 0

WOLF CREEK 3.8.2.3.10 Design Basis Accident (DBA) Loads (Yr, Yj, Ym) In addition to Pa, Ta and Ra, the following loads are considered:

a. Equivalent static load generated by the reaction on the broken pipe during the design basis accident (Yr)
b. Jet impingement equivalent static load generated by the broken pipe during the design basis accident (Yj)
c. Missile impact equivalent static load generated by or during the design basis accident, such as pipe whipping (Ym) 3.8.2.3.11 Loading Combinations The following loading combinations are considered:
a. D + L + Pt + Tt
b. D + L + To + Ro
c. D + L + To + Ro + E
d. D + L + Ta + Ra + Pa + E
e. D + L + Te + Re + Pe + E
f. D + L + Ta + Ra + Pa + E'
g. D + L + Te + Re + Pe + E'
h. D + L + Ta + Ra + Pa + Yr + Yj + Ym + E' The post-LOCA flooding of the reactor building for the purpose of fuel recovery is not a design loading condition. Refer to Section 3.8.1.3 for a further discussion.

3.8.2.4 Design and Analysis Procedure Except for the purge line penetration sleeves, the steel items described in Section 3.8.2.1 are designed and analyzed in accordance with Article NE-3000 of Subsection NE of the ASME Code, Section III, Division I and as augmented by the applicable provisions of Regulatory Guide 1.57. 3.8-34 Rev. 0

WOLF CREEK The purge line penetration sleeves are analyzed and designed in accordance with ASME Section III, Subsection NC. The following paragraphs provide individual descriptions of the design and analysis procedures performed to verify the structural integrity of the steel items. 3.8.2.4.1 Equipment and Personnel Access Hatches The equipment and personnel access hatches described in Section 3.8.2.1.1 are supported entirely by the concrete shell of the reactor building. The barrels of the personnel hatches are welded to sleeves embedded in concrete which, in turn, are welded at the periphery to the liner plate. The liner plate in the vicinity of the penetration is thickened. The additional thickness in both the barrel and liner plate is provided to satisfy the area reinforcement requirements as well as to resist the external moments and shears due to the cantilevered construction. The discontinuity stresses induced by the combination of external dead and live loads, including the effects of seismic loadings, were evaluated. The required analyses and limits for the resulting stress intensities are in accordance with Articles NE-3130 and NE-3200 of Section III of the ASME Code. The doors for both ends of the personnel hatches are of a flat or dished type. The respective analyses are in accordance with Articles NE-3325 and NE-3326 of Section III of the ASME Code. The required analyses and the stress intensity limits are in accordance with Articles NE-3130 and NE-3200 of Section III of the ASME Code. The cover with the bolting flange is designed in accordance with Article NE-3326 of Section III of the ASME Code. 3.8.2.4.2 Piping and Electrical Penetration Sleeves The penetration sleeves are welded to the thickened areas of the liner plate and are anchored to the reactor building concrete shell. Penetration sleeves are subjected to various combinations of mechanical, thermal, and seismic loadings. The resulting forces due to these various combinations of loadings are combined with the effects of external and internal pressures. The areas within discontinuities are evaluated to determine the primary and secondary stress intensities. If the penetration sleeves are subjected to cyclic service, the associated peak stress intensities were also evaluated. The required analysis and associated stress intensity limits are in 3.8-35 Rev. 0

WOLF CREEK accordance with Articles NE-3130 and NE-3200 of Section III of the ASME Code. 3.8.2.4.3 Purge Line Penetration Sleeves The design and analysis of the purge line penetration sleeves are similar to that described in Section 3.8.2.4.2 with stress intensity limits in accordance with ASME Section III, Subsection NC. 3.8.2.4.4 Fuel Transfer Tube Penetration Sleeve The design and analysis of the fuel transfer tube penetration sleeve are as described in Section 3.8.2.4.2. 3.8.2.4.5 Computer Programs The computer programs used in the analysis and design of the steel portions of the reactor building intended to resist pressure but not backed by concrete are described in Appendix 3.8A. 3.8 2.5 Structural Acceptance Criteria The fundamental acceptance criterion for the completed reactor building was successful completion of the structural integrity test. The structural acceptance criteria for steel items include allowable stress values, deformation limits, and factors of safety, and are established in accordance with ASME Section III, Subsection NC and NE, as applicable, and as augmented by the requirements of Regulatory Guide 1.57. No permanent deformations are allowed under any loading condition. The steel items, which are an integral part of the reactor building pressure boundary, are designed to meet minimum leakage rate requirements. The leakage rate shall not exceed the acceptable value indicated in the applicable Technical Specification. The design and analysis methods, as well as the type of construction materials, were chosen to allow assessment of the steel items' capability throughout the plant life. Additionally, surveillance testing provides further assurances of the steel items' continuing ability to meet their design functions. Surveillance requirements are discussed in Section 3.8.2.7. The stress limits used for the design of the purge line penetration sleeves are in accordance with Subsection NC of Section III of the ASME Code. The stress limits used for the design of all other steel items are in accordance with Subsection NE of Section 3.8-36 Rev. 0

WOLF CREEK III of the ASME Code as augmented by Regulatory Guide 1.57 and are shown in Table 3.8-3 for the load combinations stated in Section 3.8.2.3.11. 3.8.2.6 Materials, Quality Control, and Special Construction Techniques The purge line penetration sleeves are fabricated from materials that meet the requirements specified in ASME Section III, Article NC-2000, except as modified by applicable, acceptable ASME Code cases in accordance with Regulatory Guides 1.84 and 1.85 as discussed in appendix 3A. All other steel items are fabricated from materials that meet the requirements specified in Article NE-2000 of Section III of the ASME Code, except as modified by applicable, acceptable ASME Code cases. Specific information relating to materials used for penetration sleeves is discussed in Section 3.8.1.6.4.1. Details of erection tolerances, quality control, and special construction techniques are provided in Sections 3.8.1.6.4, 3.8.1.6.5, and 3.8.1.6.6. 3.8.2.7 Testing and Inservice Surveillance Requirements Testing and inservice surveillance for the steel items consists of leakage testing of the containment. The leakage tests and associated acceptance criteria are discussed in Section 6.2.6. 3.8.3 CONCRETE AND STEEL INTERNAL STRUCTURES OF STEEL OR CONCRETE CONTAINMENTS 3.8.3.1 Description of the Internal Structures The internal structures consist of the following major components:

a. Reactor support system
b. Steam generator support system
c. Reactor coolant pump support system
d. Primary shield wall and reactor cavity
e. Secondary shield walls
f. Pressurizer support system
g. Refueling canal walls
h. Operating floor 3.8-37 Rev. 17

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i. Intermediate floors, platforms, and hatches
j. Simplified Head Assembly with Reactor missile shield
k. Polar crane support system Descriptions of the supports for the reactor pressure vessel, steam generators, reactor coolant pump, pressurizer, and loop piping are further described in Section 5.4.14.

3.8.3.1.1 Reactor Support System The general arrangement and principal features of the reactor support system are provided in Figures 3.8-51 and 3.8-52. The reactor vessel is supported by steel assemblies under alternate nozzles of the vessel. These assemblies are designed, furnished, and fabricated by the NSSS manufacturer (refer to Section 5.4.14). The supporting assemblies interface with structural steel built-up members that are almost entirely embedded in the primary shield wall. The reactor vessel is supported to resist normal-operating loads, seismic loads, and loads induced by postulated pipe rupture, including the loss-of-coolant accident. The support system limits the movement of the reactor vessel to within allowable limits under the applicable combinations of loadings, and is designed to minimize resistance to the thermal movements expected during operation. 3.8.3.1.2 Steam Generator Support System The general arrangement and principal features of the steam generator support system are provided in Figures 3.8-53 through 3.8-55. The four steam generators are located in the loop compartments and are supported by steel assemblies which are designed, furnished, and fabricated by the NSSS manufacturer (refer to Section 5.4.14). Four vertical columns beneath each steam generator transfer vertical loads to the reactor building base slab. Lateral supports are provided at the lower portion of each steam generator to transfer horizontal loads to the primary shield wall (or refueling canal walls) and the secondary shield walls. These lateral supports interface with embedded anchor bolt assemblies in the walls. The upper part of each steam generator is supported by a support ring which is restrained by means of shear keys and compression bumpers. These shear keys and compression bumpers transfer horizontal loads to the refueling canal walls and the secondary shield walls by interfacing with embedded anchor bolt assemblies in the walls. The steam generators are supported and restrained to resist normal operating loads, seismic loads, and loads induced by pipe rupture. The support system prevents the rupture of 3.8-38 Rev. 19

WOLF CREEK the primary coolant pipes due to a postulated rupture in the main steam and feedwater lines and vice versa. The system is designed to minimize resistance to the thermal movements expected during operation. 3.8.3.1.3 Reactor Coolant Pump Support System The general arrangement and principal features of the reactor coolant pump support system are provided in Figures 3.8-56 and 3.8-57. Each of the four reactor coolant pumps is supported by three vertical columns and three tie rods which are designed, furnished, and fabricated by the NSSS manufacturer (refer to Section 5.4.14). The columns transfer vertical loads to the reactor building base slab. The tie rods transfer horizontal loads to the primary shield wall (or refueling canal walls) and the secondary shield walls by interfacing with structural steel built-up members which are embedded in the walls. The reactor coolant pumps are supported to prevent excessive deflections during normal operating, seismic, and pipe rupture conditions. Under LOCA loads, the pumps are prevented from becoming missiles or generating missiles that might damage other safety-related components. The system is designed to minimize resistance to the thermal movements expected during operation. 3.8.3.1.4 Primary Shield Wall and Reactor Cavity The general arrangement and principal features of the primary shield wall are provided in Figures 3.8-58 through 3.8-61. The primary shield wall is a heavily reinforced concrete cylindrical structure extending from the base slab to the seal ring level, with a minimum thickness of 7 feet. The primary shield wall forms the reactor cavity and houses the reactor vessel, provides shielding, and is designed to withstand the pressure of a LOCA. The wall provides support for the reactor vessel, the steam generators, reactor coolant pumps, cross-over legs, and the refueling canal walls above the reactor cavity. Uplift loads arising from lateral forces acting on the wall are transferred to the reactor building base slab by means of the anchorage system. The inside surface of the reactor cavity is lined with welded carbon steel plates. Large penetrations in the primary shield wall are provided for the primary loop piping and the cavity ventilation system. A permanent cavity seal ring (PCSR), to close the annulus between the reactor vessel and the sides of the reactor cavity to allow flooding the cavity for refueling purposes, is mounted between the reactor vessel flanges and the cavity liner. The PCSR is attached to the vessel flange and the liner by welding and is designed to remain in place during all plant operation and refueling activities. Integral to the PCSR is neutron shielding consisting of Reactor Experiments Type 207 borated polyethylene and type 277 refractory material. [Figure 3.8-61A] 3.8-39 Rev. 6

WOLF CREEK Based on an equivalent 12" thick water shield provided by this integral PCSR/Neutron shield design, an average neutron dose rate at the top of the refueling pool has been calculated to be 1.8 rem/hour, using the Morse Monte Carlo code (Ref. 4). Dose rates in other areas of the containment were estimated using Cain's Hypothesis (Ref. 3) along with actual dose rate measurements (Ref. 6) taken at the Farley Nuclear Plant by Lawrence Livermore Laboratories (LLL). The dose rate values obtained using this technique are given below. Location Neutron Dose Rate (mrem/hr) Equipment hatch 8-31 Personnel hatch 56 D. E. Hankins and R. V. Griffith (Ref. 3) of LLL found that the neutron-gamma dose rate ratio in the Farley containment was 7:1. Based on this ratio, the WCGS gamma dose rates are expected to be as follows. Location Gamma Dose Rate (mrem/hr) Top of refueling pool 260 Equipment hatch 1-4 Personnel hatch 8 The neutron dosimetry method complies with Revision 1 of Regulatory Guide 8.14. Exposures are determined by time-dose calculations, using rem meters. There are no specific requirements for personnel entry into the containment during normal operating conditions. The frequency of entries are based on operational needs and indications of abnormal conditions within the containment. Entries into the containment when the reactor is at power are made by at least two persons, one of whom provides health physics surveillance. 3.8.3.1.5 Secondary Shield Walls The general arrangement and principal features of the secondary shield walls are provided in Figures 3.8-62 through 3.8-65. The reinforced concrete secondary shield walls are 3 feet 6 inches thick and are anchored to the reactor building base slab. The walls extend from the base slab to a level above the top of the steam generator tube bundle to provide shielding for the reactor coolant system. 3.8-40 Rev. 6

WOLF CREEK Portions of the secondary shield walls above the operating floor are designed to be removable for steam generator removal. These reinforced concrete wall panels are bolted together at vertical joints to provide for structural continuity and integrity. They are keyed into the slab at the bottom of the panels and are prevented from becoming missiles during a seismic event. The secondary shield walls, in conjunction with the primary shield wall and refueling canal walls, form the loop compartments and provide support for the steam generators, reactor coolant pumps, pressurizer, cross-over legs, piping, various equipment, platforms, and elevated floors. 3.8.3.1.6 Pressurizer Support System The general arrangement and principal features of the pressurizer support system are provided in Figures 3.8-66 and 3.8-67. The pressurizer is located in a compartment formed by the secondary shield walls and the refueling canal walls, and is supported by steel assemblies which are designed, furnished, and fabricated by the NSSS manufacturer (refer to Section 5.4.14). The pressurizer support skirt at the bottom of the pressurizer interfaces with heavy structural steel framing which transfers vertical and lateral loads to the secondary shield walls by means of embeds. The upper portion of the pressurizer is supported laterally by lugs which are restrained by means of structural steel assemblies which interface with the embedded anchor bolt assemblies in the secondary shield walls. Using this system, the pressurizer is supported and restrained to resist normal operating loads, seismic loads, and loads induced by postulated pipe rupture. The upper lateral support system is designed to minimize resistance to the thermal movements expected during operation. 3.8.3.1.7 Refueling Canal Walls The general arrangement and principal features of the refueling canal (pool) walls are provided in Figures 3.8-68 and 3.8-69. The refueling canal is located above and to the south of the reactor cavity on the fuel building side of the reactor. The entire refueling canal is constructed of minimum 4-foot-thick reinforced concrete walls internally lined with a 1/4-inch-thick stainless steel liner plate. The canal is flooded during the reactor refueling operation. The refueling canal walls, in conjunction with the secondary shield walls, form the loop compartments and provide support for the steam generators, reactor coolant pumps, piping, various equipment, platforms, and elevated floors. 3.8-41 Rev. 0

WOLF CREEK 3.8.3.1.8 Operating Floor The general arrangement and principal features of the operating floor are provided in Figure 3.8-70. The operating floor level is divided between El. 2,047 feet 6 inches and El. 2,051 feet and is supported by the walls of the refueling pool, the secondary shield walls, and the reactor building shell. The floor supports at the shell consist of structural steel brackets welded to the shell liner and anchored into concrete. As described in Section 3.8.1.1 and shown in Figure 3.8-71, adequate separation is provided between the floor slab and the shell to allow for differential horizontal movement. The floor is constructed of reinforced concrete or steel grating, supported by structural steel framing. Plugs and removable hatches are provided for equipment removal. They are keyed in to prevent their movement in the horizontal direction. During a seismic event, the vertical components of acceleration will not overcome gravity. Those plugs and removable hatches which are subject to loads during a LOCA are secured from becoming missiles. 3.8.3.1.9 Intermediate Floors and Platforms The general arrangement and principal features of the intermediate floors are provided in Figures 3.8-72 and 3.8-73. The intermediate floor levels are at El. 2,026 and El. 2,068 feet 6 inches (partial floor). The floors, as well as miscellaneous platforms, are constructed and supported in a manner similar to the operating floor. 3.8.3.1.10, Simplified Head Assembly with Reactor Missile Shield The Simplified head Assembly consists of a welded and bolted structure that integrates the reactor missile shield and the CRDM cooling system into the existing head assembly structure (Figure 3.8-74). The modification eliminates the concrete and steel missile shield, CRDM cooling system mounted on the missile shield, and the associated ductwork that was part of the original configuration. Three CRDM cooling fans are included in the modified head assembly and, with the addition of cooling shroud panels and an upper plenum structure, provide an upflow cooling arrangement that supplies the cooling air flow to the CRDM coils. Retractable CRDM and DRPI cable bridges, including cable connector plates, are part of the modification. After being disconnected, the cables remain on the bridges and are raised with the bridges to a vertical position for removal with the head assembly. The reactor missile shield is integrated into the simplified head assembly. The missile shield consists of a two-inch thick steel plate, and is located approximately three feet above the top of the rod travel housings to provide protection against postulated CRDM missiles. 3.8.3.1.11 Polar Crane Support System The general arrangement and principal features of the polar crane support system are provided in Figure 3.8-75. The polar crane is supported by structural steel built-up crane girders mounted on crane brackets evenly spaced around the inside face of the reactor building wall. The crane brackets are welded from steel plates and embedded in the reactor building wall concrete. Further details of these brackets are discussed in Section 3.8.1.1.3. 3.8-42 Rev. 19

WOLF CREEK 3.8.3.2 Applicable Codes, Standards, and Specifications The following codes, regulations, standards, and specifications were utilized in the design of concrete and steel internal structures of the reactor building. Subsequent to operation, additional codes have been approved for use and are noted with an asterik. Applicable codes, standards, and specifications for the reactor coolant component supports are discussed in Section 5.4.14. 3.8.3.2.1 Regulations

a. 10 CFR 50, "Licensing of Production and Utilization Facilities" 3.8.3.2.2 Codes
a. American Concrete Institute, Building Code Requirements for Reinforced Concrete (ACI 318-71)
b. American Institute of Steel Construction (AISC),

Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings, 7th Edition, adopted February 12, 1969, and Supplement Nos. 1, 2, and 3

c. American Institute of Steel Construction (AISC),

Structural Joints Using ASTM A325 or A490 Bolts, May 8, 1974

d. American Institute of Steel Construction (AISC), Code of Standard Practice for Steel Buildings and Bridges, October, 1972
e. American Welding Society, Structural Welding Code (AWS Dl.1-75, *AWS D1.1-90, *AWS D1.1-2004, AWS D1.3-81, and AWS D9.1-80)
f. International Conference of Building Officials, Uniform Building Code, 1973
g. ASME Boiler and Pressure Vessel Code (1974 Edition, including Summer 1975 Addenda)

Section II - Material Specifications Section III, Division 1 - Nuclear Power Plant Components Section V - Nondestructive Examination Section IX - Welding and Brazing Qualifications 3.8-43 Rev. 22

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h. Acceptable ASME Code cases per Regulatory Guides 1.84 and 1.85, as addressed in Appendix 3A.
i. Appendix B, Steel Embedments, to the American Concrete Institute, Code Requirements for Nuclear Safety Related Concrete Structures, ACI 349-80 with the 1984 Supplement. The following sections of Appendix B relating specifically with expansion anchors will be applicable; Sections B.7.1 through B.7.5.

3.8.3.2.3 Standards and Specifications Industry standards, such as those published by the ASTM, were used whenever possible to specify material properties, testing procedures, fabrication, and construction methods. The applicable standards used are discussed in Section 3.8.3.6. Structural specifications were prepared to cover the areas related to the design and construction of the reactor building internal structures. These specifications were prepared specifically for WCGS. These specifications emphasize important points of the industry standards for these structures and reduce options such as would otherwise be permitted by the industry standards. These specifications cover the following areas:

a. Concrete material properties
b. Mixing, placing, and curing of concrete
c. Reinforcing steel and splices
d. Structural steel
e. Stainless steel and carbon steel liner plate and embeds
f. Miscellaneous and embedded steel
g. Anchor bolts
h. Grating
i. RCS support embeds, pipe whip restraints, and embeds 3.8.3.2.4 Design Criteria
a. 10 CFR 50, Appendix A - GDC 2, 3, 4, and 16. (Compliance is discussed in Section 3.1)
b. Bechtel Power Corporation Topical Reports, as referenced in Section 1.6 3.8.3.2.5 NRC Regulatory Guides NRC Regulatory Guides 1.10, 1.15, 1.55, 1.69, 1.84, 1.85, and 1.94 are applicable to the design and construction of the reactor 3.8-44 Rev. 4

WOLF CREEK building internal structures. Specific editions and the extent of compliance with these guides is discussed in Appendix 3A. 3.8.3.3 Loads and Loading Combinations The loads and loading combinations used in the design of these structures are provided in the sections below. Loading combinations and design stress limits for the reactor coolant system component supports are discussed in Sections 3.9(N).1.1 and 3.9(N).1.4.7. 3.8.3.3.1 Definitions The following nomenclature and definition of terms apply to the design of seismic Category I structures. All the major loads to be encountered and/or to be postulated are listed. All the loads listed, however, are not necessarily applicable to all structures and their elements. Loads and the applicable load combinations for which each structure is designed are dependent upon the conditions to which that particular structure is subjected (see Section 3.8.3.3.2). A full description of the loads and the analyses performed for each structure, is given in Section 3.8.4.4.

a. Normal Loads Normal loads are those loads encountered during normal plant operation and shutdown. They include the following:

D = Dead loads or their related internal moments and forces, including any permanent equipment loads and hydrostatic loads L = Live loads or their related internal moments and forces, including any moveable equipment loads and other loads which vary with intensity and occurrence, such as: Floor area loads, moveable equipment loads, lateral earth pressure, (Table 3.8-5 and Section 2.5.4) 100-year recurrence snowpack load (listed in Table 1.2-1), wind-generated wave loads (Table 3.4-3 and Sections 2.4.3 and 2.4.5) and all other live loads during plant operation (Table 3.8-4) To = Thermal effects and loads during normal oper-ating and shutdown conditions, based on the most critical transient or steady state condition 3.8-45 Rev. 30

WOLF CREEK Ro = Pipe reactions during normal operating or shutdown conditions, based on the most critical transient or steady state condition

b. Severe Environmental Loads Severe environmental loads are those loads that could infrequently be encountered during the plant life. They include the following:

E = Loads generated by the operating basis earthquake (OBE) as specified in Section 2.5.2 W = Loads generated by the design wind, as specified in Section 3.3.1

c. Extreme Environmental Loads Extreme environmental loads are those loads which are credible but are highly improbable. They include the following:

E' = Loads generated by the safe shutdown earth-quake (SSE) as specified in Section 2.5.2 Wt = Loads generated by the design basis tornado, as specified in Section 3.3.2. They include loads due to tornado wind pressure, loads due to the tornado-created differential pressures, and loads due to tornado-generated missiles. N = Probable maximum winter precipitation (PMWP) in the form of snow, 129 psf applied to the roofs of safety-related structures, as specified in Table 1.2-1 and Section 2.4.2.

d. Abnormal Loads Abnormal loads are those loads generated by a postulated high-energy pipe break accident within a building and/or compartment thereof. Included in this category are the following:

Pa = Pressure equivalent static load within or across a compartment and/or building, generated by the postulated break, and including an appropriate dynamic load factor to account for the dynamic nature of the load 3.8-46 Rev. 0

WOLF CREEK Ta = Thermal loads under thermal conditions gen-erated by the postulated break and including T o Ra = Pipe reactions under thermal conditions gener-ated by the postulated break and including R o Yr = Equivalent static load on the structure gener-ated by the reaction on the broken high-energy pipe during the postulated break, and including an appropriate dynamic load factor to account for the dynamic nature of the load Yj = Jet impingement equivalent static load on a structure generated by the postulated break, and including an appropriate dynamic load factor to account for the dynamic nature of the load Ym = Missile impact equivalent static load on a structure generated by or during the postulated break, such as pipe whipping, and including an appropriate dynamic load factor to account for the dynamic nature of the load In determining an appropriate equivalent static load for Yr , Yj , and Ym , elasto-plastic behavior may have been assumed with appropriate ductility ratios and as long as excessive deflections would not result in loss of function of any safety-related system

e. Other Definitions S = For concrete structures, S is the required section strength based on the working stress design methods and the allowable stresses defined in Section 8.10 of ACI 318-71.

For structural steel, S is the required section strength based on the elastic design methods and the allowable stresses defined in Part 1 of the AISC "Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings," February 12, 1969. The 33-percent increase in allowable stresses for concrete and steel due to seismic or wind loadings is not permitted. 3.8-47 Rev. 0

WOLF CREEK U = For concrete structures, U is the section strength required to resist design loads based on methods described in ACI 318-71. Y = For structural steel, Y is the section strength required to resist design loads and based on plastic design methods described in Part 2 of the AISC "Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings," February 12, 1969. 3.8.3.3.2 Load Combinations Structures and components except for the ESWS pipe are designed to resist the load combinations given below. Definitions of individual loads are given in Section 3.8.3.3.1. The ESWS pipes are designed to resist the load combinations given in Section 3.9.3

a. Concrete structures and components The load combinations and load factors for each individual load on powerblock structures are given in Table 3.8-4. Wind (W), tornado (W t ), and probable maximum winter precipitation (N) loadings are not applicable for the design of internal structures. Load combination, load factors and required section strength using both the working stress design method and the ultimate strength design method for nonpowerblock structures are given in Table 3.8-6.
b. Steel structures and components The load combinations for powerblock structure are given in Table 3.8-7. Wind (W), tornado (W t ), and probable maximum winter precipitation (N) loadings are not applicable for the design of internal structures. The load combinations, load factors, and required section strength, using both the elastic working stress design method and the plastic design method are given in Table 3.8-8.

3.8.3.3.3 Explanation of Load Combination Cases

a. Loading cases(1), (1a), and (1b) for Table 3.8-3 These cases include all loads which are expected to be applied during the normal plant operation, including the loads from thermal effects and pipe reactions.

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b. Loading cases (1) to (3), (1a) to (3a), for Table 3.8-5, (2), (2a), (2b), (2b1), (3), (3a), (3b), (3b1) for Table 3.8-8 These cases include all loads which are expected to be applied during the normal plant operation, including the loads from the design wind and the OBE, as well as loads from thermal effects and pipe reactions.
c. Loading cases (4), (5), and (9) for Table 3.8-7, (4) to (6) for Table 3.8-3 These cases include events and the resulting loads which are highly improbable, such as the safe shutdown earthquake, tornado and the probable maximum winter precipitation in the form of snow.
d. Loading case (6) for Table 3.8-7 This case includes the pressure loads and temperature effects resulting from a postulated accident together with pipe rupture loading and generated missiles, where applicable. These loads are not applicable to non-powerblock seismic Category I structures.
e. Loading cases (7) and (8) from Table 3.8-7 These cases include a combination of postulated accident loading, together with loads generated by the operating basis earthquake (OBE) or the safe shutdown earthquake (SSE).

3.8.3.3.4 Specific Considerations

a. In cases (6) to (8), shown in Tables 3.8-5 and 3.8-7, the peak loading effects of pipe rupture and pressurization are considered as acting simultaneously unless time histories of the loading are developed to show the time relationship of the various loads.
b. The mass considered in developing earthquake loading is only the mass contributing to dead loads and identifiable live loads.
c. In all loading cases, the live load is considered to vary from zero to the maximum specified value in determining the most critical loading condition.

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d. For load cases including either earthquake or tornado loads, the live load (L) is limited to only that live load expected to be present when the plant is operating.

3.8.3.3.5 Design Allowables The section strengths given below are used to evaluate the capacity of the section under consideration.

a. Concrete structures and components.
1. Section strengths are determined in accordance with ACI 318.
2. When the effects of tornado missile impact or pipe rupture impulsive or impactive loading are combined in loading cases (5), (7), and (8) of Table 3.8-5, yield strain and displacement may be exceeded to the limits given in Section 4.3 of BC-TOP-9-A.
3. Yielding of reinforcement is permitted in loading cases (6) to (8) of Table 3.8-5 when Ta is combined with the other loadings, provided the following is satisfied:

(a) The effects of Ta are self-relieving. (b) The ability of the structure to resist the other loadings is not jeopardized. The stress in concrete in compression is restricted to 0.85 f'c.

b. Steel structures and components
1. Section strengths are determined in accordance with AISC Specification, Part I. The symbol S is defined as the AISC allowable stress. The permissible stress to be used for each loading case is given in Table 3.8-7.
2. When the effects of tornado missile impact or pipe rupture impulsive or impactive loading are combined in loading cases (5), (7), and (8) of Table 3.8-7, yield strain and displacement may be exceeded to the limits given in Section 4.3 of BC-TOP-9-A.
3. Yielding is permitted in loading cases (6) to (8) of Table 3.8-7 when Ta is combined with the other loadings, provided the following is satisfied:

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WOLF CREEK (a) The effects of Ta are self-relieving. (b) The ability of the structure to resist the other loadings is not jeopardized. 3.8.3.4 Design and Analysis Procedures The basic techniques of analyzing the internal structures can be broadly classified into two groups: (1) conventional methods involving simplifying assumptions such as found in beam theory and (2) those based on plate and shell theories of different degrees of approximation. Analytical methods using computer programs, as described in Appendix 3.8A, were also used. Seismic analyses for the internal structures conformed to the procedures outlined in Section 3.7(B). Internal concrete structures are designed using the strength methods defined in ACI-318. The proportioning of reinforcing steel in concrete structures was based upon accepted codes of practice and detailing methods. Internal steel structures, except for the NSSS supports, are designed in accordance with AISC specifications. The selection of structural steel sections and the methods of fabrication and connection were in accordance with engineering codes and accepted industry practices. NSSS supports are designed in accordance with ASME Section III Division 1, Subsection NF. The internal structures are designed to behave within the elastic range under design loads. However, the ability of the structures to perform beyond yield was considered for loads associated with a pipe break as it affects compartment pressurization, jet impingement and pipe whip, and structural loads associated with missile impact. The loads and loading combinations used in the design of internal structures, as well as the design allowables, are presented in Section 3.8.3.3. As described in Section 3.8.3.1, the internal structures are designed to transfer loads to the foundation by means of anchorage systems. The applicable codes, standards, and specifications used are discussed in Section 3.8.3.2. The following sections discuss, in greater detail, the procedures used for analyzing and designing the reactor coolant system supports, the primary shield wall and reactor cavity, the secondary shield walls, and the refueling canal walls. 3.8-51 Rev. 0

WOLF CREEK 3.8.3.4.1 Reactor Coolant System Supports Models and methods of analysis for the reactor coolant system component supports are discussed in Section 3.9(N).1.4.4. 3.8.3.4.2 Primary Shield Wall and Reactor Cavity The primary shield wall is designed to resist all of the applicable loads, including those due to differential pressure and temperature resulting from a LOCA, operating temperatures, OBE and SSE, and those loads transmitted through the reactor vessel supports. During normal plant operation, a thermal loading on the wall is generated by the attenuation heat of gamma and neutron radiation originating from the reactor core. An insulation and cooling system is provided on the inside face of the wall to reduce the severity of this loading by limiting the concrete temperatures to 150qF except for the area directly below the seal ring support which is limited to 220qF. Analysis of the primary shield wall, depending on the loading condition being considered, was performed using classical techniques and the SAP, ASHSD, and FINEL computer programs described in Appendix 3.8A. The boundary conditions simulated actual conditions at the reactor building base slab and intersections with the refueling canal walls. Analyses for LOCA loads applicable to the primary shield wall, such as those for differential pressure and pipe rupture reaction forces, were treated as time-dependent loads by performing a static analysis and utilizing the peak of the forcing function amplified by an appropriately chosen dynamic load factor. The methods used for determining the effective dynamic load factors are in accordance with recognized dynamic analysis methods, such as those described by Reference 1. The analysis considers the nonaxisymmetric application of loads to the structure. The finite element model used for the analysis of the primary shield wall is shown in Figure 3.8-83. Design of the primary shield wall is performed, using the strength design methods described in ACI-318. 3.8.3.4.3 Secondary Shield Walls The secondary shield walls are designed to resist all of the applicable loads, including those due to differential pressure and temperature resulting from a LOCA, RCS component support forces, OBE and SSE, dead and live loads from the operating floor and intermediate platforms and walkways, and those loads resulting from a postulated pipe break. 3.8-52 Rev. 10

WOLF CREEK Analysis of the secondary shield walls were performed, using classical techniques and the SAP computer program described in Appendix 3.8A. Design for the effects of postulated pipe breaks were performed using BN-TOP-2. The finite element model used for analyzing the secondary shield walls consists of a three-dimensional model of one-half of the structure in plan about an axis of symmetry. An additional finite element model was used for analyzing these secondary shield walls at the pressurizer. Appropriate boundary conditions were modeled to simulate actual conditions at the axis of symmetry and at the intersections with the base slab, refueling canal walls, floors, and RCS component supports. The analysis for time-dependent loads, such as those for differential pressure and pipe rupture reaction forces, was performed in a manner similar to that used for the primary shield wall. The finite element models used for the secondary shield walls are shown in Figures 3.8-79 through 3.8-82. Design of the secondary shield walls was performed, using the strength design methods described in ACI-318. 3.8.3.4.4 Refueling Canal Walls The refueling canal walls are designed to resist all of the applicable loads, including those due to differential pressure and temperature resulting from a LOCA, RCS component support forces, OBE and SSE, hydrostatic loading during the refueling operation, dead and live loads from the operating floor and intermediate platforms and walkways, and those loads resulting from a postulated pipe break. Analysis of the refueling canal walls was performed, using classical techniques and the SAP computer program described in Appendix 3.8A. Design for the effects of postulated pipe breaks was performed using BN-TOP-2. The finite element model used for analyzing the refueling canal walls consists of a three-dimensional model of the entire structure. Appropriate boundary conditions are modeled to simulate actual conditions at the intersections with the base slab, secondary shield walls, primary shield wall, floors, and RCS component supports. The analysis for time-dependent loads, such as those for differential pressure and pipe rupture reaction forces, is performed in a manner similar to that used for the primary shield wall. The finite element model used for the refueling canal walls is shown in Figures 3.8-77 and 3.8-78. 3.8-53 Rev. 0

WOLF CREEK Design of the refueling canal walls is performed using the strength-design methods described in ACl-318. 3.8.3.5 Structural Acceptance Criteria The structural acceptance criteria for the concrete and steel internal structures are defined in Section 3.8.3.3. Stress criteria for the reactor coolant system component supports are discussed in Section 3.9(N).1.4.7. 3.8.3.6 Materials, Quality Control, and Special Construction Techniques This section contains information relating to the materials, quality control programs, and special construction techniques used in the fabrication and construction of concrete and steel internal structures of the reactor building. 3.8.3.6.1 Concrete Structural concrete used in the construction of the reactor building internal structures has a compressive strength, f' c ' of 4,000 psi at 28 days. The concrete materials, mix design, examination, and placement are described in Section 3.8.1.6.1. 3.8.3.6.2 Reinforcing Steel and Splices The reinforcing steel and splices used in the construction of the reactor building internal structures, including materials, examination, and erection tolerances, are described in Section 3.8.1.6.2. 3.8.3.6.3 Structural Steel The following sections describe the basic materials, examination, and erection of structural steel items. 3.8.3.6.3.1 Materials Structural steel shapes, plates, and bars conform to the requirements of the Specification for Structural Steel (ASTM A36). High strength bolting materials conform to the requirements of the Specification for High Strength Bolts for Structural Steel Joints, Including Suitable Nuts and Plain Hardened Washers (ASTM A325) or the Specification for Quenched and Tempered Alloy Steel Bolts for Structural Steel Joints (ASTM A490). Other bolting materials conform to the requirements of the Standard Specification for Low-Carbon Steel Fasteners (ASTM A307). 3.8-54 Rev. 0

WOLF CREEK Welding electrode materials were selected on the basis of the welding process used and the type of materials to be joined and in accordance with the requirements of AWS Dl.1. Written welding material control procedures were required which define the measures used to control the use of the materials throughout all welding operations. Certified material test reports were obtained for structural steel shapes, plates, and bars. All other structural steel materials were furnished with certificates of compliance. 3.8.3.6.3.2 Examination Nondestructive examination of structural steel welds were performed in accordance with the requirements of AWS Dl.1 and as augmented by design documents prepared for the SNUPPS projects (WCGS and Callaway). Inspection of high strength bolted joints was performed in accordance with the requirements of the AISC Specification for Structural Joints Using ASTM A325 or A490 Bolts and as augmented by design documents prepared for the SNUPPS projects. 3.8.3.6.3.3 Erection Structural steel was erected to the following codes, to the extent described:

a. AWS Dl.1 Structural Welding Code was used with the following exceptions:
1. For visual weld inspection performed in accordance with AWS D1.1, undercut shall not exceed 1/32 inch.
2. Fillet welds need not satisfy the convexity limitations of AWS D1.1 provided that all other parameters of acceptable weld profile are maintained.
3. Fillet welds deposited on the opposite sides of a common plane of contact between two parts need not be interrupted at the corner common to both welds as specified by AWS D1.1. The connecting weld shall be inspected for defects such as undercut and cracking, but need not be inspected for size.
4. As an alternate to AWS D1.1, visual weld inspection may be performed in accordance with EPRI NP-5380, Volume 1. (Electrical Power Research Institute -

Visual Weld Acceptance Criteria for Structural Welding at Nuclear Power Plants (NCIG-01, Rev. 2). The alternative use of EPRI NP-5380 was implemented at Wolf Creek after February, 1988.

b. AISC Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings, Sections 1.23 and 1.25, are used without exception.

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c. AISC Specification for Structural Joints Using ASTM A325 and A490 Bolts is used without exception.
d. Erection tolerances are in accordance with the AISC Code of Standard Practice for Steel Buildings and Bridges without exception.

3.8.3.6.4 Restraints and Embedded Items The following sections describe the basic materials, examination, and erection of pipe whip restraints, pipe whip restraint embeds, RCS component support embeds, and other miscellaneous embedded carbon and stainless steel items. 3.8.3.6.4.1 Materials Structural steel plates, shapes, and bars conform to the requirements of the Specification for Structural Steel (ASTM A36) or the Specification for Pressure Vessel Plates, Carbon Steel, for Moderate- and Lower-Temperature Service (ASTM A516), Grade 70, or the Specification for Pressure Vessel Plates, Alloy Steel, Quenched and Tempered (ASTM A533), Class 2. Materials for high strength steel bolts conform to the requirements of the AISC Specification for Structural Joints Using ASTM A325 or A490 bolts. Materials for other bolts and upset rods conform to the requirements of the Specification for Carbon Steel Externally and Internally Threaded Standard Fasteners (ASTM A307). Materials for shear connector studs conform to the requirements of the Specification for Steel Bars, Carbon, Cold-Finished, Standard Quality (ASTM A108), Grades 1015 and 1020, cold drawn steel. Materials for upset rods conform to the requirements of the Specification for Stainless and Heat-Resisting Steel Bars and Shapes for Use in Boilers and Other Pressure Vessels (ASTM A479) or to the requirements of the Specification for Carbon Steel Externally and Internally Threaded Standard Fasteners (ASTM A307). Materials for structural pipe conform to the requirements of the Specification for Welded and Seamless Steel Pipe (ASTM A53), Grade B, or the Specification for Seamless Carbon Steel Pipe for High-Temperature Service (ASTM A106), Grade B, or the Specification for Blank and Hot Dipped Zinc Coated (Galvanized) Welded and Seamless Steel Pipe for Ordinary Uses (ASTM A120) or the American Petroleum Institute Specification for High Test Line Pipe (API-5L), Grade B, or the Specification for Cold-Formed Welded and Seamless Carbon Steel Structural Tubing in Rounds and Shapes (ASTM A500), Grade B, or the Specification for Hot-Formed Welded and Seamless Carbon Steel Structural Tubing (ASTM A501). 3.8-56 Rev. 0

WOLF CREEK Materials for shear pins conform to the requirements of the Specification for Alloy-Steel and Stainless Steel Bolting Materials for High-Temperature Services (ASTM A193) Grade B7 or to the requirements of the Specification for Alloy Steel Bolting Materials for Special Applications (ASTM A540), Grade B23. Materials for stainless steel plates conform to the requirements of the Specification for Heat Resisting Chromium and Chromium-Nickel Stainless Steel Plate, Sheet, and Strip for Fusion-Welded Unfired Pressure Vessels (ASTM A240). Embedded anchor bolt materials conform to the applicable requirements of ASTM A36 or ASTM A193 or the Specification for Carbon and Alloy Steel Nuts for Bolts for High Pressure and High Temperature Service (ASTM A194) or ASTM A307 or ASTM A325 or the Specification for Quenched and Tempered Alloy Steel Bolts and Studs With Suitable Nuts (ASTM A354) or the Specification for Quenched and Tempered Steel Bolts and Studs (ASTM A449) or ASTM A490 or the Specification for Alloy Steel Bolting Materials for Special Applications (ASTM A540). Welding electrode materials were selected based on the welding process used and the type of material being joined and in accordance with the requirements of AWS D1.1 or the ASME Code. Written welding material control procedures were required which define the measures used to control the use of the materials throughout all welding operations. All materials used for restraints and embedded items described above were furnished with certified material test reports or certificates of compliance. 3.8.3.6.4.2 Examination One of the following nondestructive examinations were selectively performed prior to operation on pipe whip restraint, pipe whip restraint embed, and RCS component support embed welds:

a. Visual examination of all welds
b. Magnetic particle or liquid penetrant examination of welds, in accordance with AWS Dl.1
c. Radiographic examination of welds in accordance with AWS Dl.1 All other welds are examined in accordance with AWS Dl.1.

3.8-57 Rev. 0

WOLF CREEK High strength bolted joints are examined in accordance with the requirements of the AISC Specification for Structural Joints Using ASTM A325 or A490 Bolts. Examination of embedded anchor bolt materials used for RCS component support embeds meets the requirements of Section NF-2580 of the ASME Code for Class 1 component supports. 3.8.3.6.4.3 Erection Restraints and embedded items were erected in accordance with the following:

a. AWS Dl.1 Structural Welding Code is used, except that the qualification of welders and welding operators may, alternatively, be in accordance with ASME Section IX. In addition, weld procedures for joining structural steel and sleeves used for mechanical splicing of reinforcing steel may be qualified in accordance with ASME Section IX. The following exceptions are allowed for welding between anchor studs and plates embedded in concrete:
1. Vertical leg of weld may be up to 1/16 inch smaller than that specified on drawings.
2. Unequal legs are permitted.
3. Weld profile and convexity requirements for these welds need not be imposed.
4. An undercut of up to 1/16 inch for 10 percent of weld length may be permitted.
b. AISC Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings Sections 1.23 and 1.25 are used without exception.
c. AISC Specification for Structural Joints Using ASTM A325 or A490 Bolts is used without exception.
d. Erection tolerances for pipe whip restraints, pipe whip restraint embeds, and RCS component support embeds are in accordance with the following:
1. AISC Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings, for rolled plates and shapes 3.8-58 Rev. 0

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2. AWS Dl.1 Structural Welding Code for welded assemblies
3. Additional tolerance requirements are specified in design documents for bearing or contact points, clearances, and transverse locations of restraints
e. Erection tolerances for other embedded items described above are the same as those for concrete forms. All embedded items are secured and protected during placement of concrete.

3.8.3.6.5 Reactor Coolant System Supports Materials, quality control, and special construction techniques for the reactor coolant system supports are discussed in Section 5.4.14. 3.8.3.6.6 Quality Control In addition to the quality control procedures discussed in Sections 3.8.3.6.1 through 3.8.3.6.5, the construction quality control program is discussed in the Quality Assurance Programs For Design and Construction Manual which was contained in the PSAR. 3.8.3.6.7 Special Construction Techniques The reactor building internal structures are constructed using proven methods common to heavy industrial construction. No special, new, or unique construction techniques are used. 3.8.3.7 Testing and Inservice Surveillance Requirements Tests and inspections for the reactor coolant system component supports are discussed in Section 5.4.14. 3.8.4 OTHER CATEGORY I STRUCTURES 3.8.4.1 Description of the Structures The general arrangement of seismic Category I structures is shown in Figure 3.8-84. The seismic Category I structures other than the reactor building are:

a. Auxiliary building
b. Fuel building
c. Control building 3.8-59 Rev. 1

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d. Diesel generator building
e. Refueling water storage tank and valve house
f. Emergency fuel oil storage tanks and vault
g. Buried power block duct banks and piping
h. Essential Service Water System Pumphouse, pipes, and electrical ductbanks and manholes
i. Circulating and warming waterpipe encasements and Essential Service Water System Caissons
j. ESW Vertical Loop Chase Seismic Category I structures are physically separated from adjacent structures by isolation joints, with the exception of the auxiliary and control buildings which share a common base slab and wall and the ESW Vertical Loop Chase which is attached to the west wall of the control building. The isolation joints at the roof, base slab, and exterior walls of buildings contain waterstops to provide environmental protection while allowing free rotation and translation between structures. Figure 3.8-85 shows typical isolation joint details.

3.8.4.1.1 Auxiliary Building The auxiliary building is a multistory, structural steel and reinforced concrete structure which houses the safety injection system, residual heat removal system, CVCS monitoring system, auxiliary feedwater pumps, steam and feedwater isolation and relief valves, heat exchangers, other pumps, tanks, filters, and demineralizers, and heating and ventilating equipment. The arrangement of the auxiliary building is shown in Figures 3.8-86 through 3.8-93. The auxiliary building shares a common base mat and wall with the control building. The building's interior is enclosed on one side by the reactor building wall. The foundation for the auxiliary building is a two-way mat foundation with a minimum thickness of 5.0 feet. The lowest floor elevation is 25.5 feet below plant grade, except for the RHR and containment spray pumps pit which is 33.5 feet below grade. The roof is 74.7 feet above plant grade, except for the southwest corner which is 48 feet above grade, two penthouses which are 84 feet above grade, and the roof over the main steam tunnel, which is 103 feet above plant grade. The intermediate floors and the roof are reinforced concrete slabs supported by structural steel beams and girders. The floor and roof framing are supported by exterior reinforced concrete bearing 3.8-60 Rev. 32

WOLF CREEK walls and interior steel columns. The roof slab and exterior walls are designed to prevent penetration by tornado generated missiles. Concrete plugs provided in the roof for equipment removal are designed to resist tornado missiles. These plugs and additional concrete plugs and removable hatches provided for servicing equipment within the building are adequately anchored or keyed into slabs to prevent displacement during a seismic event. Blockouts are provided in the interior walls for equipment removal and servicing. These blockouts are closed with multiwythes of solid concrete blocks, laid such that the vertical and horizontal joints are not continuous. The blocks are seismically restrained on both faces. Concrete block walls are reinforced to withstand seismic loadings. 3.8.4.1.2 Fuel Building The fuel building is a rectangular, structural steel, reinforced concrete structure which houses the spent fuel pool, transfer canal, cask loading pool and cask washdown pit, spent fuel pool bridge crane, cask handling crane, and other miscellaneous equipment. The arrangement of the fuel building is shown in Figures 3.8-94 through 3.8-98. The fuel building is supported on a two-way, reinforced concrete base mat which is founded 6 feet below plant grade. The minimum thickness of the mat is 6.5 feet, and the mat beneath the spent fuel pool is 12 feet thick. The top of the roof slab is 107 feet above plant grade. The elevated floors and the roof are reinforced concrete slabs supported by structural steel beams and girders. The floor and roof framing are supported by reinforced concrete bearing walls. The exterior walls have integral reinforced concrete pilasters to stiffen the walls against lateral loads and to support the cask-handling crane girders. The roof and exterior walls are designed to prevent penetration by tornado generated missiles. The walls and base slab of the fuel storage pool, transfer canal, and cask washdown pits are lined with stainless steel plates for ease of decontamination. A leak chase system provided to check the leaktightness of the liners, although leaktightness is not the primary liner function. The cask handling crane is capable of moving a loaded fuel cask. The crane travel is limited to prevent movement over the spent fuel pool. 3.8-61 Rev. 14

WOLF CREEK 3.8.4.1.3 Control Building The control building is a rectangular structural steel and reinforced concrete structure which houses the access control areas, control room, upper and lower cable spreading rooms, electrical and mechanical equipment rooms, and locker rooms. The arrangement of the control building is shown in Figures 3.8-99 through 3.8-104. The control building shares a common base slab and wall with the auxiliary building. The bottom of the base mat is 31.5 feet below plant grade, and the mat thickness is 6 feet. The top of the roof is 81.7 feet above plant grade. The intermediate floors and roof are reinforced concrete slabs supported by structural steel beams and girders. The floor and roof framing are supported by exterior reinforced concrete bearing walls and interior steel columns. The roof slab and exterior walls are designed to prevent penetration by tornado-generated missiles. Concrete block walls are reinforced to withstand seismic loadings. 3.8.4.1.4 Diesel Generator Building The diesel generator building is a single-story, rectangular, structural steel and reinforced concrete structure which houses the standby diesel generators, fuel oil day tank, exhaust silencers, and exhaust stacks. The diesel generator building arrangement is shown in Figures 3.8-105 through 3.8-109. The foundation for the diesel generator building is a 10.5-foot-thick base mat founded 10 feet below plant grade. The highest portion of the roof is 66.5 feet above plant grade. The roof is a reinforced concrete slab supported by structural steel beams and girders. The roof framing is supported by reinforced concrete bearing walls and steel columns. The roof and exterior walls are designed to prevent penetration by tornado-generated missiles. 3.8.4.1.5 Refueling Water Storage Tank The refueling water storage tank consists of an above-grade cylindrical steel tank founded on a 5-foot-6-inch-thick reinforced concrete base slab and an associated valve house. Although serving a safety-related function and designed as a seismic Category I structure, the refueling water storage tank is not required for safe shutdown of the plant following a tornado event and is, therefore, not designed to resist the effects of the design-basis tornado. The steel tank is described in Section 6.3. Details of the tank foundation and valve house are shown in Figures 3.8-110 and 3.8-111. 3.8-62 Rev. 0

WOLF CREEK 3.8.4.1.6 Emergency Fuel Oil Storage Tanks The emergency fuel oil storage tanks consist of two buried cylindrical steel tanks and associated reinforced concrete access vaults. The steel tanks are described in Section 9.5.4. Details of the access vaults are shown in Figures 3.8-112 and 3.8-113. 3.8.4.1.7 Buried Power Block Duct Banks and Piping Buried, reinforced concrete electrical duct banks and steel piping that serve safety-related functions are classified as seismic Category I and are shown in Figures 3.8-114 and 3.8-115. 3.8.4.1.8 Essential Service Water System Pumphouse The ESWS pumphouse is a tornado-resistant, rectangular (85 x 39 feet), conventionally reinforced-concrete structure. The pumphouse contains two 100-percent-capacity ESWS pumps, valves, two self-cleaning strainers, two traveling water screens, two trash racks, two transformers, two motor control centers, a redundant HVAC system, and piping. The separate redundant operating floors are at elevation 2000 feet (SNUPPS elevation) and separate forebays extend to elevation 1958 feet. The roof slab elevation is 2025 feet. A 15-x 36-foot apron slab is attached to the pumphouse and extends into the UHS intake channel. The pumphouse is of heavy shear wall construction with concrete slabs. Tornado-resistant missile shields protect the pumphouse forebay pits, the entrances and exits of the ventilation system at the roof elevation and the doors at grade. Removable hatch covers are bolted down to prevent their movement in the horizontal and vertical directions. Typical plans and sections are shown on Figures 3.8-131, 3.8-132, and 3.8-133. 3.8.4.1.9 Essential Service Water System Pipes Two (redundant) below-grade, 30-inch-diameter pipes carry cooling water from the ESWS pumphouse to the powerblock. Two (redundant), below grade, 30-inch-diameter pipes in series with two (redundant), below grade, 24-inch-diameter pipes return cooling water from the powerblock to the ESWS discharge point. One (redundant) below grade, 30-inch-diameter pipe in series with 18-inch-diameter pipe and one (redundant) below grade 18-inch-diameter pipe carry warm water from the two 30-inch-diameter ESWS discharge pipes to the ESWS pumphouse forebay to prevent ice accumulation on the trash racks and traveling water screens. All pipes from the powerblock to the last access vault are buried a minimum depth of 4.5 feet to resist the effects of tornado missiles and frost penetration. The ESW vertical loop chase structure is designed to resist the effects of tornado missile penetration and the insulation installed on the ESW vertical loops resists the effects of frost penetration. The discharge piping from the last access vault to the slope of the UHS is encased in unreinforced concrete with 24 inch minimum thickness above the top of the pipe for protection from tornado missiles. The outlets for the 24-inch discharge pipes (discharge point) are below the minimum elevation (1968 feet) of the UHS to prevent their freezing. Typical plans and sections are shown on Figure 3.8-134, 3.8-135, 3.8-135a, and 3.8-136. 3.8-63 Rev. 29

WOLF CREEK Pipes are of carbon and stainless steel with welded joints at connections, except at the insulating flanged connections to the piping within the ESWS pumphouse and connections between carbon steel and stainless steel. Buried pipe exteriors are coated and wrapped, and are cathodically protected. Additionally, the buried pipe is encased in controlled low strength material (CLSM) to provide additional corrosion resistance to the exterior of the ESWS piping. CLSM is not used for the submerged discharge piping and where the ESWS piping crosses above the circulating water piping and electrical duct bank, these cases are discussed further in sections 3.8.4.1.12 and 3.8.4.1.13, respectively. At points where the 30-inch-diameter, 4-inch-diameter and 18-inch-diameter pipes enter structures, provision is made for flexible, waterproof boot seals between the pipes and the structures (see Figure 3.8-138 and Figure 3.8-144) where necessary. 3.8.4.1.10 Deleted 3.8.4.1.11 Essential Service Water System Electrical Duct Banks and Manholes Redundant, below-grade, reinforced-concrete electrical duct banks housing electrical cables are provided which transmit the required power to the ESWS pumphouse from the standard power block. They are buried a minimum depth of 4 feet to resist the effects of tornado missiles and frost penetration. Typical plans and sections are shown on Figures 3.8-134, 3.8-135, and 3.8-136. At points where the electrical duct banks enter structures, provision is made for flexible filler and waterstops between the duct banks and the structures (see Figure 3.8-139). Redundant, reinforced-concrete, tornado-resistant electrical manholes are provided to permit the pulling of electrical cables through the duct bank. Removable manhole covers are bolted down to prevent their movement in the horizontal and vertical directions. Typical plans and sections are shown on Figure 3.8-140. 3.8.4.1.12 Essential Service Water System Caissons The ESWS piping that crosses above the below-grade, non-seismic Category I circulating water pipe and two electrical ductbanks are supported by eight caissons. Two caissons, one on each side of the circulating water piping, are used to support the ESWS supply and return piping for train A and train B. The caissons provide support to the ESWS piping to ensure continued ESWS function if the circulating water piping ruptured during a seismic event and undermined the ESWS piping. 3.8-64 Rev. 28

WOLF CREEK 3.8.4.1.13 Circulating and Warming Water Pipe Encasements The below-grade, non-seismic Category I circulating and warming water pipes (Unit No. 1) and non-seismic Category I circulating water pipe (future Unit No.

2) are surrounded, as shown in Figure 3.8-137, by seismic Category I reinforced concrete encasements, where they pass under the ESWS duct bank. These reinforced concrete encasements consist of a minimum 1.5-foot-thick concrete encasements on all sides. In addition, the concrete encasements are extended a sufficient distance on either side of the ESWS duct bank to prevent their undermining if the non-seismic Category I pipes are ruptured during a seismic event.

3.8.4.1.14 Essential Service Water System Access Vaults The below grade ESW access vaults (six total) are tornado resistant, conventionally reinforced structures (24 x 39.5, 10.5 x 24, 12.25 x 44, 17.75 x 46, 105 x 24 and 27 x 29.5). Reinforced concrete barrier walls are provided between the redundant ESWS pipes where they share the same vault. The access vaults contain ESWS piping and are for the purpose of accessing these pipes. Refer to figure 3.8-143. 3.8.4.1.15 ESW Vertical Loop Chase The ESW Vertical Loop Chase is a rectangular structural steel, metal plates and reinforced concrete structure (27 X 15-10) which houses a vertical loop of the ESWS return pipes that mitigate water hammer effects. The arrangement of the ESW Vertical Loop Chase is shown in Figures 3.8-99 through 3.8-104. The ESW Vertical Loop Chase is attached to the west face of the control building. The foundation of the ESW Vertical Loop Chase is 4 feet thick base mat founded 29.5 feet below plant grade. The top of the roof is 87.8 feet above plant grade. The intermediate floors and roof are steel platforms and grating supported by structural steel beams and girders. The roof and exterior wall plates are designed to work in conjunction with chase structural steel to prevent penetration by tornado generated missiles. 3.8.4.2 Applicable Codes, Standards, and Specifications The codes, regulations, standards, and specifications utilized in the design of the seismic Category I structures other than the reactor building are the same as those listed in Section 3.8.3.2, with the following exceptions:

a. Structural Specification for Maintenance Truss
b. Structural Specification for RCS Support Embeds, Pipe Whip Restraints, and Embeds
c. The applicable standards used are discussed in Section 3.8.4.6.
d. Regulatory Guide 1.46 and BN-Top-2 are not applicable to Essential Service Water System structures and circulating and warming water pipe encasements.

In addition to the documents listed in Section 3.8.3.2, the following documents are also utilized:

a. NRC Regulatory Guide 1.59 - Design Basis Floods for Nuclear Power Plants 3.8-65 Rev. 32

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b. NRC Regulatory Guide 1.76 - Design Basis Tornado for Nuclear Power Plants
c. Bechtel Power Corporation Topical Report BC-TOP-3A, Tornado and Extreme Wind Design Criteria for Nuclear Power Plants, Revision 3, August, 1974.

3.8.4.3 Loads and Load Combinations The loads and load combinations used in the design of the seismic Category I structures other than the reactor building and ESWS pipe are the same as those described in Section 3.8.3.3 with the following exception. In accordance with the discussion in Section 3.8.4.1.1 and Appendix 3B.4 the terms Y j , Y r , and Y m in Tables 3.8-5 and 3.8-7 do not apply to the main steam isolation valve room since no pipe breaks are postulated in that area. The loads and load combinations used in the design of the ESWS pipe are the same as those defined in Section 3.9.3. 3.8.4.4 Design and Analysis Procedures The analysis of standard plant seismic Category I structures other than the reactor building was performed, using conventional analytical methods which are common to standard engineering practice and analytical methods using computer programs. Analytical methods using computer programs are described in Appendix 3.8A. Seismic analysis conformed to the procedures outlined in Section 3.7(B). Concrete structures are designed, using the strength methods defined in ACI-318. The reinforcing steel is proportioned in accordance with accepted engineering formulae and conforms to the applicable codes and standards. The effects of design variables are accounted for by the use of conservative loads and load combinations and the use of load factors and capacity reduction factors. Steel structures and components, except for tanks and piping, are designed in accordance with AISC specifications. The selection of steel sections is in accordance with accepted engineering formulae and conforms to the applicable codes and standards. The effects of design variables are accounted for by the use of conservative loads, load combinations, and allowable stresses. These structures are designed to behave within the elastic range, under normal operating loads. However, the ability of the structures to perform beyond the yield point is considered for loads associated with missile impact, jet impingement, and pipe whip. The loads, load combinations, and design allowables used in the design of these structures are presented in Section 3.8.4.3. The applicable codes, regulations, standards, and specifications used are discussed in Section 3.8.4.2. 3.8-66 Rev. 6

WOLF CREEK The following sections discuss, in greater detail, the procedures used for the analysis and design of the auxiliary and control buildings, fuel building, diesel generator building, essential service water system pumphouse, ESWS pipes, ESWS electrical duct banks and manholes, ESWS caissons and circulating and warming water pipe encasements. 3.8.4.4.1 Auxiliary and Control Building The auxiliary and control buildings are supported on a common base slab. All vertical loads are transferred to the base slab through reinforced concrete bearing walls and structural steel columns. All lateral loads are resisted by diaphragm action of the roof and intermediate floor slabs which transfer these loads to shear walls, which, in turn, transfer the lateral loads to the base slab. All lateral loads are transferred to the subgrade by friction and passive earth pressure. Typical connection details between the walls and slabs are shown in Figures 3.8-116 through 3.8-118. The reinforced concrete roof and intermediate floor slabs are analyzed and designed for vertical loads as one-way or two-way slabs supported by bearing walls and structural steel beams and girders. The reinforced concrete interior and exterior walls are analyzed and designed for lateral loads as one-way or two-way slabs supported by the base slab, intermediate floor slabs, roof slab, and perpendicular walls. Structural steel beams and girders supporting reinforced concrete slabs are analyzed and designed as composite sections. The reinforced concrete base slab is analyzed and designed as a rigid slab on an elastic foundation. The main steam isolation valve room is located in the north-west corner of the auxiliary building as shown in Figure 3B-2. It is designed to withstand the environmental effects, by means of venting, of a main steam or main feedwater line break equivalent to the flow area of a single-ended pipe rupture. Although no specific pipe breaks are postulated in the main steam/main feedwater isolation valve compartment, this consideration provides an additional level of assurance of operability to the building structure and the safety-related equipment in this compartment. 3.8.4.4.2 Fuel Building The fuel building is supported on a base slab. All vertical loads are transferred to the base slab through the exterior walls, interior walls, and fuel storage pool walls. All lateral loads are 3.8-67 Rev. 28

WOLF CREEK transferred to the base slab by diaphragm action of the roof slab and intermediate floor slabs which transfer loads to shear walls. All hydrostatic and hydrodynamic loads due to the presence of water in the fuel storage pool are transferred to the base slab through the fuel storage pool walls. All lateral loads are transferred to the subgrade by friction and passive earth pressure. Typical connection details between exterior, interior, and fuel storage pool walls and the base slab are shown in Figures 3.8-116 and 3.8-118. The reinforced concrete roof and intermediate floor slabs were analyzed and are designed for vertical loads as one-way or two-way slabs supported by bearing walls and structural steel beams and girders. The fuel storage pool is analyzed and designed as an open top, reinforced concrete tank. The reinforced concrete interior and exterior walls were analyzed and are designed for lateral loads as one-way slabs supported by the base slab, intermediate floor slabs, and roof slab. Structural steel beams and girders supporting reinforced concrete slabs are analyzed and designed as composite sections. The reinforced concrete base slab was analyzed and is designed as a rigid slab on an elastic foundation. 3.8.4.4.3 Diesel Generator Building The diesel generator building is supported on a base slab. All vertical loads are transferred to the base slab through exterior walls, interior walls, and columns. All lateral loads are transferred to the base slab by diaphragm action of roof slab and intermediate floor slab, which transfer loads to shear walls and bracing. All lateral loads are transferred to the subgrade by friction and passive earth pressure. Typical connection details between the exterior and interior walls and the base slab are shown in Figures 3.8-116 and 3.8-118. The reinforced concrete roof and intermediate floor slabs are analyzed and designed for vertical loads as one-way or two-way slabs supported by the base slab, intermediate floor slab, roof slab, and intersection walls. Structural steel beams and girders supporting reinforced concrete slabs are analyzed and designed as composite sections. The reinforced concrete base slab was analyzed and is designed as a rigid slab resting on an elastic foundation. 3.8-68 Rev. 14

WOLF CREEK 3.8.4.4.4 Essential Service Water System Pumphouse The ESWS pumphouse is supported on a concrete floor slab at grade, a concrete pipe pit slab approximately 14 feet below grade, grade beams with varying depths below grade, and a forebay and apron slab approximately 47 feet below grade and in the ultimate heat sink (UHS). All vertical loads are transferred to the grade beams and floor, pipe pit, forebay, and apron slabs through exterior walls, interior walls, and columns. All lateral loads are transferred to the grade beams and floor, pipe pit, forebay, and apron slabs by diaphragm action of the roof and floor slabs which transfer loads to shear walls and by beam action for walls not acting as shear walls. All lateral loads are transferred to the subgrade by friction. The reinforced concrete roof and floor slabs are analyzed and designed for vertical loads as one-way or two-way slabs supported by bearing walls, concrete columns, and concrete beams. The reinforced concrete interior and exterior walls were analyzed and designed for lateral loads as cantilevered, one-way, or two-way slabs supported by the grade beams and the floor, pipe pit, forebay, apron, and roof slabs. The forebay compartments within the UHS were analyzed and designed to resist the effects of hydrostatic and hydrodynamic loads. The reinforced concrete floor and forebay and apron slabs were analyzed and designed as rigid slabs resting on an elastic foundation. 3.8.4.4.5 Essential Service Water System Pipes Refer to USAR Section 3.9(B).3. 3.8.4.4.6 Deleted 3.8-69 Rev. 28

WOLF CREEK 3.8.4.4.7 Essential Service Water System Electrical Duct Banks and Manholes The reinforced concrete ESWS electrical duct banks are buried below grade. They were analyzed and designed as beams on elastic foundations for vertical loads. Differential movement between the duct banks and other seismic Category I structures was considered in the analysis and design. Refer to Figures 3.8-134, 3.8-135, and 3.8-136. The ESWS electrical manholes are supported on base slabs. All vertical loads are transferred to the base slabs through exterior and interior walls. Since the manholes are horizontally continuous frames below grade, all lateral loads on the walls are balanced through the walls as reactions from adjacent walls. The roof slab is bolted to the walls and transfers lateral load to the walls through the bolts. Refer to Figure 3.8-140. 3.8.4.4.8 Essential Service Water System Caissons The ESWS caissons are supported on limestone and are able to maintain support of the ESWS piping with or without backfill. All vertical loads are transferred through the caissons to bedrock. The lateral loads on the caissons are balanced due to the symmetrical nature of the caissons. The reinforced concrete and steel outer shell caissons were analyzed as caisson pile pipe supports. Refer to Figure 3.8-137 Sh. 2. 3.8.4.4.9 Circulating and Warming Water Pipe Encasements The two below-grade circulating water pipe encasements and single below-grade warming water pipe encasements are supported on in situ material and lean concrete backfill. External loads on the encasements are balanced by the symmetrical nature of the encasements, and internal loads are contained by hoop action of the encasements. The reinforced concrete barrier walls were analyzed and designed for lateral loads (assuming there are no resisting lateral loads on one side) as cantilever beams. 3.8.4.4.10 Essential Service Water System Access Vaults The buried ESWS access vaults (six total) are designed to be independent of each other. The vaults are supported on concrete base slabs, approximately 10 feet below grade. All vertical loads are transmitted to the base slabs through the walls. The lateral loads on the walls are balanced by the loads from the opposite end walls through the continuous frame action. The roof slabs are designed to span between the walls and the base slabs are designed as rigid slabs resting on elastic foundation. The roof slabs are designed to resist lateral movements with the use of corbels on the underside of the slab. Refer to figure 3.8-143. 3.8.4.4.11 ESW Vertical Loop Chase The ESW Vertical Loop Chase is supported by a base slab. All vertical loads are transferred to the base slab through A36 carbon steel bearing walls and A500 structural steel columns and to the control building through a combination of grouted anchors and bolts. All lateral loads are resisted by diaphragm action of the roof and intermediate floor plates which transfer these loads to shear walls, which, in turn, transfer the lateral loads to the base slab. All lateral loads are transferred to the subgrade by friction and passive earth pressure. Typical connection details between the walls and slabs are shown in Figures 3.8-116 through 3.8-118. The steel plate roof and intermediate floor plate and grating are analyzed and designed for vertical loads as one-way or two-way plates supported by bearing walls and structural steel beams and girders. The plate metal exterior walls are analyzed an designed for lateral loads as one-way or two-way plates supported by the base slab, intermediate 3.8-70 Rev. 32

WOLF CREEK floor plates, roof plate, and perpendicular walls. Structural steel beams and girders supporting plate and grating are analyzed and designed as composite sections. The reinforced concrete base slab is analyzed and designed as a rigid slab on an elastic foundation. 3.8.4.5 Structural Acceptance Criteria The structural acceptance criteria for the seismic Category I structures other than the reactor building and ESWS pipes are the same as those defined in Section 3.8.3.3. The seismic Category I essential service water pipes are designed to the criteria defined in Section 3.9(B).3. 3.8.4.6 Materials, Quality Control, and Special Construction Techniques The materials, quality control programs, and special construction techniques used in the fabrication and construction of seismic Category I structures other than the reactor building are described in the following sections. 3.8.4.6.1 Concrete Structural concrete used in the construction of these structures has a minimum compressive strength, f'c' of 4,000 psi at 28 days. The concrete materials, mix design, examination, and placement are described in Section 3.8.1.6.1. 3.8.4.6.2 Reinforcing Steel and Splices The reinforcing steel and splices used in the construction of these structures, including materials, examination, and erection tolerances, are described in Section 3.8.1.6.2 3.8.4.6.3 Structural Steel The structural steel used in the construction of these structures, including materials, examination, and erection, are described in Section 3.8.3.6.3. 3.8.4.6.4 Embedded Items The embedded carbon steel items used in the construction of these structures, including materials, examination, and erection, are described in Section 3.8.3.6.4. 3.8.4.6.5 Quality Control The quality control measures are discussed in Sections 3.8.4.6.1 through 3.8.4.6.4. The construction quality control program is discussed in the Quality Assurance Programs For Design and Construction Manual which was contained in the PSAR. 3.8.4.6.6 Special Construction Techniques These structures were constructed of concrete and steel, using proven methods common to heavy, industrial construction. No special, new, or unique construction techniques were used. 3.8-71 Rev. 29

WOLF CREEK 3.8.4.7 Testing and Inservice Surveillance Requirements Testing and inservice surveillance are not required for seismic Category I structures other than the reactor building. Hence, no formal program of testing and inservice surveillance is required. The ESWS is tested and inspected in accordance with the codes described in Section 9.2.1.2.5. 3.8.5 FOUNDATIONS 3.8.5.1 Description of the Foundations Seismic Category I structures have reinforced concrete mat foundations resting on existing rock, undisturbed soil, or engineered backfill. All vertical loads are transferred to the subgrade by direct bearing of the base mat on the foundation media. Horizontal shears, such as those produced by winds and earthquakes, are transferred to the subgrade by friction along the bottom of the base mat. There is no waterproofing membrane between the base mats and the subgrade, with the exception of the ESW Vertical Loop Chase. The foundation for each structure is separated by isolation joints from adjacent foundations and structures, with the exception of the auxiliary and control buildings which share a common base mat. All the foundations are adequately designed to prevent overturning due to horizontal loads. The following sections describe the Category I foundations. Figures 3.8-116, 3.8-117, 3.8-118 and 3.8-8 through 3.8-11 shows the general arrangement of these foundations. 3.8.5.1.1 Reactor Building The reactor building foundation is a 10-foot-thick reinforced concrete mat, 154 feet in diameter, founded 11 feet below plant grade. The central reactor cavity and instrumentation tunnel extend below the reactor building foundation, with the bottom of the 5.5-foot-thick foundation slab located 36 feet below grade. The 8-foot-wide tendon access gallery, located beneath the perimeter of the reactor building mat, has a 4.25-foot-thick foundation slab, the bottom of which is 25.25 feet below grade. The plan and details of the reactor building foundation are shown in Figures 3.8-1 and 3.8-8 through 3.8-11. Refer to Section 3.8.3.1 for a description of the anchorage of internal structures and equipment to the foundation. 3.8-72 Rev. 32

WOLF CREEK 3.8.5.1.2 Auxiliary and Control Buildings The auxiliary and control buildings are supported by a common, reinforced concrete mat foundation, with a minimum thickness of 5 feet, founded 31.5 feet below plant grade. The foundation under the RHR and containment spray pumps pit in the auxiliary building is a 6-foot-thick mat, the bottom of which is 38.5 feet below grade. The shape of the base mat in plan conforms to the arrangement of the building it supports, and the base mat is approximately 220 feet wide at its widest section. The plan and details of the foundation for the auxiliary and control buildings are shown in Figures 3.8-119 and 3.8-120. The equipment in these buildings, such as tanks, heat exchangers, switchgear, and control panels, is rigidly attached to the base mat, intermediate floor slabs, or walls, by means of anchor bolts or welding to embedments in the concrete. All loads from equipment and internal structures not directly attached to the base mat are transferred to the base mat through structural steel columns, which are attached to the base mat by anchor bolts, or reinforced concrete bearing, and shear walls, which are anchored to the base mat by reinforcing steel dowels. 3.8.5.1.3 Fuel Building The fuel building foundation is a 6.5-foot-thick reinforced concrete mat extending 6 feet below plant grade. The mat is essentially rectangular with overall dimensions of 137 feet long and 91 feet wide. The thickness of the mat below the spent fuel pool is increased to 12 feet. Figures 3.8-121, and 3.8-122 show the general arrangement and details of the fuel building foundation. The spent fuel racks are supported by the base slab of the fuel storage pool. Other equipment is rigidly attached to the base mat, intermediate floors, or walls by means of anchor bolts or welding to embedments in the concrete. All loads from equipment and internal structures not directly attached to the base mat are transferred to the base mat through reinforced concrete walls and pilasters, which are anchored to the base mat by reinforcing steel dowels. 3.8.5.1.4 Diesel Generator Building The diesel generator building is supported by a 10.5-foot-thick reinforcing concrete mat, the bottom of which is 10 feet below plant grade. The mat is rectangular, and is 88.25 feet long and 66.25 feet wide. Figure 3.8-123 shows the general arrangement and details of the diesel generator building foundation. 3.8-73 Rev. 14

WOLF CREEK The diesel generators are rigidly attached to the base mat by means of anchor bolts. Other equipment is rigidly attached to the base mat, intermediate platforms, walls, or roof by means of anchor bolts or welding to structural steel framing or embedments in the concrete. Loads from equipment and internal structures not directly attached to the base mat are transferred to the base mat through structural steel columns, which are attached to the foundation by anchor bolts, or reinforced concrete walls, which are anchored to the base mat by reinforcing steel dowels. 3.8.5.1.5 Refueling Water Storage Tank The refueling water storage tank is supported by a 5.5-foot-thick reinforced concrete base mat which extends 4.5 feet below plant grade. The base mat is octagonal, with a distance of 43 feet between parallel edges. Figures 3.8-110 and 3.8-111 show the general arrangement and details of the refueling water storage tank foundation. The refueling water storage tank is rigidly attached to the base mat by means of anchor bolts which transfer all loads, including seismic lateral forces, to the foundation. 3.8.5.1.6 Essential Service Water System Pumphouse At grade, the ESWS pumphouse foundation consists of a 1-foot-8-inch-thick reinforced concrete floor slab spanning between 2-foot-thick grade beams and a 13-foot-thick pipe incasement integral with the floor slab and extending approximately 14 feet and 10 feet, respectively, below grade. The floor slab and integral grade beams and incasement are attached to the pipe pit and the forebay walls which extend to the below-grade portions of the foundations. Below grade, the ESWS pumphouse foundation consists of (1) a 2-foot-8-inch-thick, reinforced concrete pipe pit slab located approximately 14 feet below grade and (2) a 3-foot-8-inch-thick reinforced concrete forebay slab located approximately 47 feet below grade, with an apron slab which varies in thickness. The apron slab provides a transition from the forebay slab to the bottom of the ultimate heat sink. In plan, the combined area of the foundations forms a rectangular-shaped foundation approximately 39 feet wide and 100 feet long. The general arrangement and details of the ESWS pumphouse foundation are shown in Figures 3.8-131, 3.8-132, and 3.8-133. Horizontal shears, such as those that are seismically induced, are transferred to the subgrade foundation media by friction along the bottom of the floor slab in areas that are not waterproofed and through the rock and soil below the shear keys attached to the forebay and apron slabs. 3.8-74 Rev. 7

WOLF CREEK Equipment such as the ESWS pumps, ESWS strainers, and ESWS traveling water screens is anchored to the floor slab by means of anchor bolts which transmit the equipment loads, including seismic forces, to the foundation. Other equipment and piping are anchored to walls, roofs, or to platforms anchored to the floor slab. Refer to Section 3.8.4.4.1 for a description of the anchorage of internal structures to the foundation. 3.8.5.1.7 Deleted 3.8.5.1.8 Essential Service Water System Electrical Manholes The ESWS electrical manhole foundations consist of 1-foot-6-inch-thick reinforced concrete slabs below grade. The slabs are rectangular in shape and have varying dimensions. Typical general arrangements and details of the ESWS electrical manholes are shown in Figure 3.8-140. Transfer of horizontal shears, such as those that are seismically induced, is by means of the walls of the ESWS electrical manholes bearing against the soil which completely surrounds the manholes. Electrical conduit within the manholes is anchored to the walls. 3.8.5.1.9 Essential Service Water System Caissons The foundation for the ESWS caissons consists of a 46.5 inch diameter rock socket in the limestone bedrock filled with reinforced concrete. General arrangement of the caissons is shown in Figure 3.8-137 Sh. 2. Transfer of horizontal shears, such as those that are seismically induced, is by means of the reinforced concrete caisson bearing against the limestone rock socket. 3.8-75 Rev. 28

WOLF CREEK 3.8.5.1.10 Circulating and Warming Water Pipe Encasements The foundations for the circulating and warming water pipe encasements consist of the encasements themselves, which are buried below grade. 3.8.5.1.11 Essential Service Water System Access Vaults The foundation for each ESWS access vault is independent of the other. The foundation for each vault consists of a minimum thickness of 3 feet, reinforced concrete slab. The plan dimensions of the vaults are (24 x 39.5, 10.5 x 24, 12.25 x 44, 17.75 x 46, 10.5 x 24 and 27 x 29.5). Transfer of horizontal loads due to lateral earth pressure and due to seismic is by means of the walls bearing against the soil which completely surrounds the vaults. ESWS pipes are anchored at the slab. 3.8.5.1.12 ESW Vertical Loop Chase The ESW Vertical Loop Chase is supported by a reinforced concrete mat foundation, with a minimum thickness of 4 feet, founded 29.5 feet below plant grade. The shape of the base mat in plan conforms to the arrangement of the building it supports, and the base mat is approximately 27 feet wide at its widest section. The plan and details of the foundation for the auxiliary and control buildings are shown in Figures 3.8-119 and 3.8-120. The ESW piping supports in this building are rigidly attached to the base mat, intermediate floor framing, or control building west wall, by means of anchor bolts or attached to the wall mounting plates. All loads from piping and the internal structures not directly attached to the base mat are transferred to the base mat through structural steel columns, which are attached to the base mat by anchor bolts. Vertical piping loads are transferred to the base mat through supports and lateral piping loads are transferred to the Control building west wall through supports. 3.8.5.2 Applicable Codes, Standards, and Specifications Applicable codes, standards, and specifications are discussed in Section 3.8.4.2. 3.8.5.3 Loads and Load Combinations Foundation loads and load combinations are discussed in Section 3.8.4.3. 3.8.5.4 Design and Analysis Procedures The design and analysis procedures for the reactor building foundation are discussed in BC-TOP-5-A. The foundations for other powerblock seismic Category I structures are analyzed as flat slabs on elastic supports. Loads are applied to the slab through structural steel columns and reinforced concrete walls, with the resulting foundation-bearing pressures being determined using well-established principles and methods of engineering mechanics. The foundations for the powerblock seismic Category I structures are designed using the strength design methods defined in ACI 318. The reinforcing steel is proportioned in accordance with accepted engineering formulas and conforms to the applicable codes and standards. The effects of design variables are accounted for by the use of conservative loads and load combinations and the use of load factors and capacity reduction factors. The foundations of these structures were analyzed, using well-established methods based on the general principles of engineering mechanics. Codes, standards, and specifications prescribed in Section 3.8.4.2 are used in the design and analysis of structures and systems. 3.8-76 Rev. 32

WOLF CREEK 3.8.5.5 Structural Acceptance Criteria The foundations for powerblock seismic Category I structures are designed to meet the same structural acceptance criteria as the structures themselves. The criteria are discussed in Sections 3.8.1.5 and 3.8.4.3. Minimum safety factors for seismic Category I foundations, for the load combinations given in Sections 3.8.1.3 and 3.8.4.3, are: Overturning 1.50 Sliding 1.10 Buoyancy 1.25 The limiting conditions for the foundation media are given in Section 2.5.4.10. The foundations of nonpowerblock structures are designed to meet the structural acceptance criteria described in Sections 3.8.4.2 and 3.8.4.3. The limiting conditions for the foundation medium, together with a comparison between actual capacity and structural loads, are found in Section 2.5.4. Nonpowerblock structures meet or exceed the factors of safety shown in Table 3.8-9 for the load combinations for overturning, sliding, and flotation given in Table 3.8-9. Definitions of D, E, W, E', and W t are found in Section 3.8.4.3.1. H is the lateral soil pressure, and F' is the buoyant force of the ground water which is assumed at grade. No live loads are included in these combinations to help resist overturning, sliding, and flotation. 3.8.5.6 Materials, Quality Control, and Special Construction Techniques The foundations for the seismic Category I structures are constructed of reinforced concrete, using proven methods common to heavy industrial construction. For further discussion, refer to Sections 3.8.1.6 and 3.8.4.6. 3.8.5.7 Testing and Inservice Surveillance Requirements Testing and inservice surveillance are not required, nor planned, for the foundations of the seismic Category I structures. 3.8.6 RADWASTE BUILDING AND TUNNEL 3.8.6.1 Description of the Structures 3.8.6.1.1 Radwaste Building The radwaste building is a rectangular, multistory, structural steel and reinforced concrete structure which houses radioactive waste treatment facilities, tanks, filters, and other miscellaneous equipment. Figures 3.8-124 through 3.8-130 show the general arrangement of the building. 3.8-77 Rev. 6

WOLF CREEK The radwaste building is supported on a reinforced concrete mat foundation with a minimum thickness of 4.5 feet. The building extends 33.5 feet below plant grade. Intermediate floors are reinforced concrete slabs with metal decking, supported by structural steel beams and girders, and reinforced concrete bearing walls. The building has a built-up roof, the top of which is 56 feet above grade, supported by structural steel beams and girders. The roof and intermediate floor framing are supported by structural steel columns and reinforced concrete bearing walls. The storage area of the radwaste building is a single story, structural steel building supported on a reinforced concrete mat foundation, the top of which is 6 inches above plant grade. This area is separated from the adjacent portion of the radwaste building by isolation joints and houses the radioactive waste handling facilities, and storage areas. The storage areas are enclosed by reinforced concrete and/or masonry shield walls. The building has a built-up roof with a high point 31 feet above plant grade. 3.8.6.1.2 Radwaste Pipe Tunnel The radwaste pipe tunnel is a below grade, reinforced concrete, two-cell box structure connecting the auxiliary building and the radwaste building. It is separated from both buildings by isolation joints. The bottom of the tunnel is 25.5 feet below plant grade, and the top is 8 feet below grade. The tunnel provides access and carries electrical cable trays and piping between the auxiliary building and the radwaste building. 3.8.6.2 Applicable Codes, Standards and Specifications These structures were designed in accordance with the codes and standards listed in the following sections. Subsequent to operation, additional codes have been approved for use and are noted with an asterik. 3.8.6.2.1 Codes

a. American Concrete Institute, Building Code Requirements for Reinforced Concrete (ACI 318-71).
b. American Institute of Steel Construction (AISC),

Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings, 7th Edition, adopted February 12, 1969, and Supplement Nos. 1, 2, and 3.

c. American Institute of Steel Construction (AISC),

Structural Joints Using ASTM A325 or A490 Bolts, May 8, 1974.

d. American Institute of Steel Construction (AISC), Code of Standard Practice for Steel Buildings and Bridges, October, 1972.

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e. American Welding Society, Structural Welding Code (AWS D-1.1-75, *AWS D1.1-90, *AWS D1.1-2004).
f. International Conference of Building Officials, Uniform Building Code, 1973.
g. Appendix B, Steel Embedments,to the American Concrete Institute, Code Requirements for the Nuclear Safety Related Concrete Structures, ACI 349-80 with 1984 Supplement. The following sections of Appendix B relating specifically with expansion anchors will be applicable; Sections B.7.1 through B.7.5.

3.8.6.2.2 Standards and Specifications Nationally recognized industry standards, such as those published by the ASTM, are used whenever possible to describe material properties, testing procedures, fabrication, and construction methods. The applicable standards used are discussed in Section 3.8.3.6. Structural specifications were prepared to cover the areas related to the design and construction of these structures. The specifications were prepared specifically for the SNUPPS project (WCGS and Callaway). They emphasize important points of the industry standards for these structures and reduce options such as would otherwise be permitted by the industry standards. The specifications covered the following areas:

a. Concrete material properties
b. Mixing, placing, and curing of concrete
c. Reinforcing steel and splices
d. Structural steel
e. Miscellaneous and embedded steel
f. Anchor bolts
g. Grating 3.8.6.3 Loads and Load Combinations The radwaste building and tunnel are designed for the applicable loads and load combinations specified in the codes listed in Section 3.8.6.2.1.

3.8.6.4 Design and Analysis Procedures 3.8.6.4.1 Radwaste Building The intermediate concrete floor slabs are designed for the combination of dead, live, and lateral loads, in accordance with ACI-318. The structural steel beams and girders are designed as composite sections, in accordance with the AISC manual. 3.8-79 Rev. 22

WOLF CREEK The exterior reinforced concrete walls are designed as one-way or two-way slabs supported at the base slab, intermediate floors, roof, and transverse walls, as applicable. The loading combinations are given in ACI-318. The base slab is designed as a slab on an elastic foundation for loads and load combinations given in ACI-318. The seismic loads for the structure are obtained by the following procedures:

a. The input motion at the foundation of the radwaste building is defined by normalizing the Regulatory Guide 1.60 spectra to the OBE maximum ground acceleration of 0.12g, as outlined in Section 3.7(B).1.1. The damping values given in Table 3.7(B)-1 are used. These are consistent with the damping values recommended in Regulatory Guide 1.61.

A simplified analysis was performed to determine appropriate seismic loads and floor response spectra pertinent to the location of the systems. The simplified analysis involved the modeling of the building by a several-degrees-of-freedom mathematical model and time-history analysis to generate the floor response spectra for radwaste systems and the seismic loads for the building. The design time-histories are defined in Section 3.7(B).1.2.

b. The simplified method for determination of seismic loads for the building consists of (1) calculation of modal frequencies and participation factors for the building, (2) determination of modal seismic loads by item a, input spectra, and (3) combination of modal seismic loads by the square-root-of-the-sum-of-the-squares (SRSS) rule.

Only two orthogonal horizontal inputs need to be considered in two separate analyses, and the greater of the two results of the analyses is used for building design.

c. Time-history analysis is performed to generate floor response spectra. Item a, design time-histories, will be used as input.
d. The load factors and load combinations used for the building are those given in ACI-318. The allowable stresses for steel components are those given in the AISC Manual of Steel Construction.

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e. The construction and inspection requirements for the building elements comply with those stipulated in the AISC or ACI Code, as appropriate.
f. The foundation media of the radwaste building does not liquefy during the operating basis earthquake.

3.8.6.4.2 Radwaste Pipe Tunnel The radwaste tunnel was analyzed as a rigid box in the transverse direction. Dynamic soil and hydro pressures were obtained in accordance with Section 2.5.4.10.3. Longitudinally it is designed as a beam on an elastic foundation. The tunnel is isolated from the radwaste and auxiliary buildings by isolation joints. The load factors and the loading combinations are given in ACI-318. 3.8.6.5 Structural Acceptance Criteria These structures are designed for structural acceptance criteria defined in the codes listed in Section 3.8.6.2.1. 3.8.6.6 Materials, Quality Control, and Special Construction Techniques The materials, quality control programs, and special construction techniques used in the fabrication and construction of these structures are described in the following sections. 3.8.6.6.1 Concrete Structural concrete used in the construction of these structures has a minimum compressive strength, f' c, of 4,000 psi at 28 days. The concrete materials, mix design, examination, and placement are described in Section 3.8.1.6.1. 3.8.6.6.2 Reinforcing Steel and Splices The reinforcing steel and splices used in the construction of these structures, including materials, examination, and erection tolerances, are described in Section 3.8.1.6.2. 3.8.6.6.3 Structural Steel The structural steel used in the construction of these structures, including materials, examination, and erection, are described in Section 3.8.3.6.3. 3.8-81 Rev. 0

WOLF CREEK 3.8.6.6.4 Embedded Items The embedded carbon steel items used in the construction of these structures, including materials, examination, and erection, are described in Section 3.8.3.6.4. 3.8.6.6.5 Quality Control The quality control measures are discussed in Section 3.8.5.6. 3.8.6.6.6 Special Construction Techniques These structures are constructed of concrete and steel, using proven methods common to heavy, industrial construction. No special, new, or unique construction techniques are used. 3.8.6.7 Testing and Inservice Surveillance Requirements Testing and inservice surveillance are not required for these structures. No formal program of testing and inservice surveillance is planned. 3.

8.7 REFERENCES

1. Biggs, J. M., Introduction to Structural Dynamics, McGraw Hill, Inc., 1964.
2. Brown, F. R., 1979, Letter of July 5, 1979 to Karl V.

Seyfrit, Director Region IV, Nuclear Regulatory Commission.

3. Hankins D. E. and Griffith R. V., "A Survey of Neutrons Inside the Containment of a Pressurized Water Reactor," ORNL/RSIC-43, Page 114, February, 1979.
4. Hopkins W. C., "Calculations of the Neutron Environment Inside PWR Containments," ORNL/RSIC-43, Page 127, February, 1979.
5. Portland Cement Association, 1979, Wolf Creek Generating Station Reactor Base Mat Concrete Second Testing Program:

Report dated February 27, 1979, Construction Technology Laboratories Division.

6. Straker E. A., Stevena P. N., Irving D. C., and Cain V. R.,
   "The MORSE Code -- A Multigroup Neutron and Gamma-Ray Monte Carlo Transport Code., ORNL-4585, September, 1975.
7. Varga, S. A., 1979, NRC internal memorandum of July 10, 1979 for G. W. Reirnmuth, Assistant Director Division of Reactor Construction Inspection, Office of Inspection and Enforcement.

3.8-82 Rev. 0

WOLF CREEK TABLE 3.8-1 CONTROL TESTS FOR CONCRETE Material Requirements Test Method Minimum Frequency Cement Standard physical and chemical ASTM C150 Each 1,200 tons properties Fly ash and Chemical and physical ASTM C3111 Each 200 tons pozzolans properties in accordance with ASTM C618 Aggregate Gradation ASTM C136 Once per shift during production Moisture content ASTM C566 Once per shift during production Material finer than #200 sieve ASTM C117 Daily during production Organic impurities ASTM C40 Once per shift during production Flat and elongated particles CRD C-119* Twice per month during production Friable particles ASTM C142 Monthly during production Lightweight particles ASTM C123 Monthly during production Soft fragments ASTM C235 Monthly during production Specific gravity and ASTM C127 (coarse) Initially absorption ASTM C128 (fine) Los Angeles abrasion ASTM C131 Every 6 months during production Potential reactivity ASTM C289 Every 6 months during production Soundness ASTM C88 Every 6 months during production Water and Ice Effect on compressive strength AASHTO T-26 Every 6 months. If chemical data Setting time AASHTO T-26 indicates that the water quality Soundness AASHTO T-26 is unchanged, the tests may be Total solids AASHTO T-26 waived by the owner. Chlorides AASHTO T-26 Admixtures Chemical composition Infrared spectro- Composite of each shipment photometry Concrete Mixer uniformity ASTM C94 Initially and every 6 months Sampling method ASTM C172 Compression cylinders ASTM C31 Compressive strength ASTM C39 One set of 2 cylinders from each 100 cubic yards or a minimum of one set per day for each mix design, for each strength test. Slump ASTM C143 First batch mixed each shift and every 50 cubic yards placed. Air content ASTM C231 First batch mixed each shift and every 50 cubic yards placed. Temperature - First batch mixed each shift and every 50 cubic yards placed. Unit weight ASTM C138 Every 100 cubic yards during production.

  • Alternately, the project Technical Specifications provide for a procedure that may be used in lieu of the test method indicated. Rev. 0

WOLF CREEK TABLE 3.8-2 MAXIMUM ALLOWABLE OFFSET IN FINAL WELDED JOINTS OF REACTOR BUILDING LINER PLATE Direction of Joints in Circumferential Section Thickness Shells (in.) Longitudinal Circumferential Up to 1/2 incl. 1/4 t 1/4 t Over 1/2 to 3/4 incl. 1/8 in. 1/4 t Over 3/4 to 1.5 incl. 1/8 in. 3/16 in. Over 1.5 1/8 in. 1/8 t "t" is the nominal thickness of the thinner section at the joint. Rev. 0

WOLF CREEK TABLE 3.8-3 STRESS LIMITS FOR STEEL PORTIONS OF CONCRETE CONTAINMENTS DESIGNED IN ACCORDANCE WITH SUBSECTION NE OF THE ASME CODE Section 3.8.2.3.11 Primary Stresses Primary & Secondary Peak Stresses Buckling Combination Gen. Memb. Local Memb. Bend + Local Stresses Note (3) No. P P Memb. P + P m L B L (1) .9S 1.25S 1.25S 3S Consider for 125% of allow. y y y m fatigue analy- given by sis NE-3133 (2) & (3) S 1.5S 1.5S 3S Consider for Allow. given m m m m fatigue analy- by NE-3133 sis (4) & (5) S 1.5S 1.5S N/A N/A Allow. given m m m by NE-3133 (6) & (7) Not integral S 1.5S 1.5S N/A N/A Allow. given and continuous m m m by NE-3133 Integral and The greater The greater The greater N/A N/A 120% of allow. continuous of 1.2S of 1.8S of 1.8S given by NE-3133 m m m or S or 1.5S or 1.5S y y y (8) Not integral The greater The greater The greater N/A N/A 120% of allow. and continuous of 1.2S of 1.8S of 1.8S given by NE-3133 m m m or S or 1.5S or 1.5S y y y Integral and 85% of stress intensity limits of Appendix F N/A N/A 85% of allow. continuous given by F-1325 of App. F NOTES: (1) Thermal stresses need not be considered in computing P , P , and P m L B (2) Thermal effects are considered in: (a) Specifying stress intensity limits as a function of temperature. (b) Analyzing effects of cyclic operation (NE-3222.4). (3) If a detailed analysis considering ineleastic behavior is performed for checking instability (buckling), such an analysis should demonstrate that the applied stress is less than 50 percent of the critical buckling stress. Designs utilizing vertical stiffeners are permitted. The allowable axial compressive stress may be determined by considering the effects of circumferential stiffner spacing and the effects of water, if present. Rev. 0

WOLF CREEK TABLE 3.8-4 GENERAL DESIGN LIVE LOADS Stairs and walkways 100 psf Grating, floors, and 100 psf (except in areas platforms of heavier loads, which will govern) Surcharge outside and 250 psf vertical load or 8,000-adjacent to subsurface walls pound wheel load converted to lateral equivalent load, whichever is governing, or railroad surcharge per AREA specification, where applicable Railings 25 plf or 200 pounds applied in any direction at top of railing Concentrated load on slabs 5 kips to be so applied as to (to be considered with dead maximize moment or shear. This load only) load is not carried to columns. Concentrated load on beams 5 kips to be so applied as to and girders (in addition to maximize moment or shear. This all other loads) load is not carried to columns. Ground floor 250 psf Rev. 0

WOLF CREEK TABLE 3.8-5 LOAD COMBINATIONS AND LOAD FACTORS FOR CATEGORY I CONCRETE STRUCTURES A. Load Combinations For Service Load Conditions

a. Working Stress Design Method (1) S = D + L (2) S = D + L + E (3) S = D + L + W la) 1.3S = D + L + T + R o o (2a) 1.3S = D + L + T + R + E o o (3a) 1.3S = D + L + T + R + W o o Both cases of L having its full value or being completely absent are checked.
b. Strength Design Method (1) U = 1.4 D + 1.7 L (2) U = 1.4 D + 1.7 L + 1.9 E (3) U = 1.4 D + 1.7 L + 1.7 W (lb) U = (0.75) (1.4 D + 1.7 L + 1.7 T + 1.7 R )

o o (2b) U = (0.75) (1.4 D + 1.7 L + 1.9 E + 1.7 T + 1.7 R ) o o (3b) U = (0.75) (1.4 D + 1.7 L + 1.7 W + 1.7 T + 1.7 R ) o o Both cases of L having its full value or being completely absent are checked against the following combinations: (2b) U = 1.2 D + 1.9 E (3b) U = 1.2 D + 1.7 W Where soil and/or hydrostatic pressures are present, in addition to all the above combinations where they have been included in L and D, respectively, the requirements of Sections 9.3.4 and 9.3.5 of ACI-318 are also satisfied. Rev. 0

WOLF CREEK TABLE 3.8-5 (Sheet 2) B. Load Combinations For Factored Load Conditions For extreme environmental, abnormal, abnormal/severe environmental and abnormal/extreme environmental conditions, respectively, the strength design method should be used, and the following load combinations are satisfied: (4) U = D + L + T + R + E o o (5) U = D + L + T + R + W o o t (6) U = D + L + T + R + 1.5 P a a a (7) U = D + L + T + R + 1.25 P + 1.0 (Y + Y + Y ) 

              + 1.25 E a     a          a      r     j     m (8)    U = D + L + T + R + 1.0 P + 1.0 (Y + Y + Y )
              + 1.0 E a     a        a      r     j     m (9)    U = D + L + T   + R    + N o     o In combinations (6), (7), and (8), the maximum values of P , T , R , Y , Y , and Y , including an appropriate a   a    a    j   r        m dynamic load factor, are used unless a time-history analysis is performed to justify otherwise. Combinations (5), (7), and (8) are satisfied first without the tornado missile load in (5) and without Y , Y , and Y in (7) r   j      m and (8). When considering these loads, however, local section strength capacities may be exceeded under the effect of these concentrated loads, provided there will be no loss of function of any safety-related system.

Both cases of L having its full value or being completely absent are checked. Rev. 0

WOLF CREEK TABLE 3.8-6 LOAD COMBINATIONS AND LOAD FACTORS FOR SEISMIC CATEGORY I CONCRETE STRUCTURES Working Stress Design Method (1) S = D + L (2) S = D + L + E (3) S = D + L + W (la) 1.3S = D + L + T + R o o (2a) 1.3S = D + L + T + R + E o o (3a) 1.3S = D + L + T + R + W o o Both cases of "L" having its full value or being completely absent should be checked. Strength Design Method (1) U = 1.4D + 1.7L (2) U = 1.4D + 1.7L + 1.9E (3) U = 1.4D + 1.7L + 1.7W (1b) U = 0.75 (1.4D + 1.7L + 1.7T + 1.7R ) o o (2b) U = 0.75 (1.4D + 1.7L + 1.7T + 1.7R + 1.9E) o o (3b) U = 0.75 (1.4D + 1.7L + 1.7T + 1.7R + 1.7W) o o Both cases of "L" having its full value or being completely absent should be checked with the following combinations: (2b) U = 1.2D + 1.9E (3b) U = 1.2D + 1.7W Where soil and/or hydrostatic pressures are present, in addition to all the above combinations where they have been included in L and D, respectively, the requirements of Section 9.3.4 and 9.3.5 of ACI 318-71 should also be satisfied. For the following combinations, which represent extreme environmental conditions, the strength design method should be used, and the following load combinations should be satisfied: (4) U = D + L + T + R + E o o (5) U = D + L + T + R + W o o t (6) U = D + L + T + R + N o o Rev. 0

WOLF CREEK TABLE 3.8-7 LOAD COMBINATIONS AND LOAD FACTORS FOR CATEGORY I STEEL STRUCTURES A. Load Combinations for Service Load Conditions

a. Working and Stress Design Method (1) S = D + L (2) S = D + L + E (3) S = D + L + W (la) 1.5 S = D + L + T + R o o (2a) 1.5 S = D + L + T + R + E o o (3a) 1.5 S = D + L + T + R + W o o Both cases of L having its full value or being completely absent are checked.
b. Plastic Design Method (1) Y + 1.7 D + 1.7 L (2) Y = 1.7 D + 1.7 L + 1.7 E (3) Y = 1.7 D + 1.7 L + 1.7 W (1b) Y = 1.3 (D + L + T + R )

o o (2b) Y = 1.3 (D + L + E + T + R ) o o (3) Y = 1.3 (D + L + W + T + R ) o o Both cases of L having its full value or being completely absent are checked. B. Load Combinations for Factored Load Conditions

a. Working Stress Design Method (4) 1.6 S = D + L + T + R + E o o (5) 1.6 S + D + L + T + R + W o o t (6) 1.6 S = D + L + T + R + P a a a Rev. 0

WOLF CREEK TABLE 3.8-7 (Sheet 2) (7) 1.6S* = D + L + T + R + P + 1.0 (Y +Y +Y ) + E a a a j r m (8) 1.7s* = D + L + T + R + P + 1.0 (Y +Y +Y ) + E a a a j r m (9) 1.6S = D + L + T + R + N o o

  • For these two combinations, (7) and (8), in computing the required section strength, S, the plastic section modulus of steel shapes is used.
b. Plastic Design Method (4) .90 Y = D + L + T + R + E o o (5) .90 Y = D + L + T + R + W o o t (6) .90 Y = D + L + T + R + 1.5 P a a a (7) .90 Y = D + L + T + R + 1.25 P + 1.0 (Y +Y +Y )
                 + 1.25 E a       a            a        j r m (8)       .90 Y = D + L + T    + R + 1.0 P a      a         a
                 + 1.0 (Y +Y +Y ) + E j r m (9)       .90 Y = D + L + T    + R    + N o      o In combination B (a) and (b) above, thermal loads are neglected when it is shown that they are secondary and self-limiting in nature and where the material is ductile.

In combinations (6), (7) and (8), the maximum values of P , a T , R , Y , Y and Y , including an appropriate dynamic load a a j r m factor, are used unless a time-history analysis is performed to justify otherwise. Combination (5), (7), and (8) are first satisfied without the tornado missile load in (5) and without Y , Y , and Y in (7) r j m and (8). When considering these loads, however, local section strengths may be exceeded under the effect of these concentrated loads, provided there will be no loss of function of any safety-related system. Rev. 0

WOLF CREEK TABLE 3.8-8 LOAD COMBINATIONS AND LOAD FACTORS FOR SEISMIC CATEGORY I STEEL STRUCTURES Elastic Working Stress Design Method (1) S = D + L (2) S = D + L + E (3) S = D + L + W (la) 1.5 S = D + L + T + R o o (2a) 1.5 S = D + L + T + R + E o o (3a) 1.5 S = D + L + T + R + W o o Both cases of "L" having its full value or being completely absent should be checked in the above combinations. (4) 1.6 S = D + L + T + R + E o o (5) 1.6 S = D + L + T + R + W o o t (6) 1.6 S = D + L + T + R + N o o Plastic Design Method (1) Y = 1.7 D + 1.7 L (2) Y = 1.7 D + 1.7 L + 1.7 E (3) Y = 1.7 D + 1.7 L + 1.7 W (1b) Y = 1.3 (D + L + T + R ) o o (2b) Y = 1.3 (D + L + E + T + R ) o o (3b) Y = 1.3 (D + L + W + T + R ) o o Both cases of "L" having its full value or being completely absent should be checked in the above combinations. (4) .90Y = D + L + T + R + E o o (5) .90Y = D + L + T + R + W o o t (6) .90Y = D + L + T + R + N o o Rev. 0

WOLF CREEK TABLE 3.8-9 ADDITIONAL LOAD COMBINATIONS FOR SLIDING, OVERTURNING, AND FLOTATION Loading Combination Minimum Factor of Safety Overturning Sliding Flotation

a. D + H + E 1.50 1.10 ---

b D + H + W 1.50 1.10 ---

c. D + H + E 1.50 1.10 ---
d. D + H + W 1.50 1.10 ---
e. D + F t --- --- 1.25 Rev. 0

WOLF c==RE::::::EK Rev. 0 f="TCURF ';_R- 1 AND [LEV.A,T lUN oF D !J l L U l !~~J

WOLF" CREEK .REV.22 UPDATED SAFETY ANALYSIS REPORT fiGURE 3.8-2 REACTOR BUILDING GROUND F"LOOR PLAN - ELEV. 2000'-0" AND 2001'-4"

WOf. E' CREEK Rev.30 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.8-3 REACTOR BUILDING INTERMEDIATE FLOOR PLAN - ELEV. 2026'-0"

WOLF CREEK REV .19 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3 .8-4 REACTOR BUILDING OPERATING FLOOR pLAN - ELEv . 2 0 4 7' - 6 AND II 2051'-0"

WOLF CREEK Rev, 0 WOLF CREEK npn~TZD SAFETY ANALYSIS R!PORT f="IG!JRF ~- R-5

WOLF CREEK REV.30 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.8-6 REACTOR BUILDING EAST -WEST CROSS SECTION

WOLF CREEK SECTION B LOOKING WEST REV.19 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.8-7 REACTOR BUILDING NORTH - SOUTH CROSS SECTION

WOLF CREEK SYMMETRICAL ABOUT ~ 

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i I Rev. 0 II WOLF CREEK

III OPDATED SAFETY ANALYSIS REPORT FIGURE 3.8-8 II II i REACTOR BUILDING BASE MAT I
                          ";c.@      9                             REINFORCING - BOTTOM LAYERS LAI't.f<   J

i<<>LF CREEK 

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FIGURE 3.8-9 REACTOR BUilDING BAS£ MAT R£INfORCING - TOP LAY£RS

WOLF CREEK fI~ RE-'CTOR REACTOR BLDG 

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MCT!QN Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.8-10 o~S~~!~~.~~ILDJNG BA§E MAT 

                                              " :.~ "   :~ .,_.,. . -  '. '- .

WOLF CREEK iI REACTOR BUILDING 3 _4+=11@ 24 EW -----, .---2 -*11@ 24 EW----. 2-*9 & i -*11 @ 24 (BUNDLED) (BUNDLED) ~~; (BUNDLED) 1-*11@ 24 EW . 

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WOLF CREEK L! NE 11\ ~- EL 20i9*-o** r-f)ETAIL 8 1 1!' >I 6 ~* 24 - 

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                         ---~-~                     EL 1988 :,:1:'4_ __ 1--------.!---,J(L-.!.1-:~-:.1~

FIGURE 3.8-12 w ~J ReACTOR BUILDING SHELL 

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REIN FQ RCI i~ G L..LL.... YMI 1'-"1'-'

WOLJ:t, CREEK f-_ -- ____ _"'~ 12EW IBOTT) ICA~WELDEDJ SQUARE PATTERN 

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SOUARE PATTERN NOTE AOJUST*6 OR 0 8HOOPS 0 NOTE 9 SOUARf Fl<ITTERN BARS TO AVOID HOIST lNG TO BE CUT AND HCX)K£0 ,JP"""RD WOLF CREEK HOOK EMBEDS AT HOISTING HOOK EMBEDS PlAN UPDATED SAFETY ANALYSIS REPORT 

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WOLF CREEK UPDATED SAFETY ANALYSIS REPOR.T FIGURE ~5. 8-1 S REACTOR BUILDING TENDON ANCHOF1A(7E SYSTEI~ 

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WOLF CREEK 11* LiHfR It I~ ___71'c~R"-Oill'; 9£____!!00P __ TENDON?_ (LO_C_A.TION OF LO- " __EST 

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SECTION© REACTOR BLDG. Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.8-17 _!l__LOC,. SECTION© j REACTOR BUJ~~I~G FOR CONT!NU ..O..T!ON i Tr~ooNs - ~tLllUNS

WOLF HOOP 1 BUTTRESS 12'-o" CONSTANT BELOW EL 2135'-o" ilvfRIES ABOvEEL -iiJs~ TEI<OON'5 II 

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101~ TYPICAL BUTTRESS DETAIL Rev. U iiOLY l:l<EEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.8-18 r""\~-----,.....- --**- ..,.. _______ _ 

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WOLF CREEK o* Wtx.i Rev

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FIGURE 3.8-20 lNG LINER PlATE - t '1*1**11I;_.IIOWJ REACTg~"~U~~~FFENER PlAN PLAN L

WOLF CREEK Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 8-21

WOLF CREEK 

                          *---+-----L 5"x ~x 1.(4"
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DETAIL @) Kev. U WOLF CREEK 2 DETAIL G) DETAIL ED UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 8-22 REACTOR DUILDING LINER j},

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WOLF CREEK 

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1 ......., * ' vL-'-' 1 - '4 1 I HERE SEE \\ II 2~-)-------L..I~'--+- Y4 LINt::.R Jj;> 

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W / CADWELD SLEEVES FOR PRIMARY SHIELD WALL. Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT ANCHUHAGE

WOLF CREEK

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II o I V4 Ll NER It WOLF CREEK UPDATED SAFETY ANALYSIS REPORT 

                                                       ~IGURE         3.R-?4 ANCHORAGE AT REACTOR CAVITY -

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WOLF CREEK 

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SLAB;~>* I 11 FLOOR V4 LINER it 

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HOLE (FILLED wj CONC ) BOTTOM OF BASE SLAB~ WOLP CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 8-25 TYPICAL ANCHORAGC T!!RQUGH RAsF MAT fOR NSSS EQUIPM[~T n. - - - - -! -;. - 

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(TYP) J J f'tloVENT_~ HOLES~ PLAN -'EXPANSION END" BRACKET( E) END" BRACKET (F) TOP OF BRACKET 

                                                    ~INSPECTION PORT l}4~ STIFFENER __

2" If_ BEYOND- 

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AIIALYSIS 

                                                                                                          -l" tlo VENT I-OL.ES FIGURE 3. 8-26 BUILDING Po BRACKETS LAR SECTION

WOLF CREEK 

                                                                                                                             ; 1 ' ¢ VENT HOLE (TYP)

BETWEEN SLOTS PLAN I TOP OF BRACKET-----. I I ~BRACKET 

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WOLF CREEK REACTOR BLDG. (SYMM. AB T. Cf. ) 

                                                                         ~...-----,-

2-WAY SPRAY PIPING SUPPORT r I 70-0 1 v.; II TO O.S. FACE OF LINER I(. DETAIL ([) ELEVATION Rev. 0 WOLF CREEK

  • UPDATED SAFETY ANALYSIS REPORT ~~

FIGURE 3. 8-28 REACTOR t3UILfiTNG -TYPICAL r--!P~ SUPPORT !jRACKETS IN noME-*- I I i i

WOLF CREEK 70'-d' TO t REACTOR BLDG. 

                                                                          -y4' l

LINER PLATE L 3x2 1 C 3x4.1 w/}_4 CAP 1!.@ VERTICAL 

                                                         .1>---1.;__----1 JOINTS IN LINER 1!. AND TESi POINT LOCATIONS TYP. AT ALL TEST CHANNEL TYPICAL LEAK      CHASE        ADJACENT              TO      WALL WOLF' CREEK UPDATED SAFETY ANALYSIS REFORT FTGIJRE 3.8-29 HE LE t..K, RE l

i

WOLF CREEK 1.1l 

                                                                     ~j
                                                                     ~

II I 1 II I TRUMPLATES 

                               -*6 @ 18" HAIRPIN 6'o"____ +. . _fi.~_Q_~

SYM.ABTj BUTTRESS PLAN OF BUTTRESS @I SHELL WALL Kev. v WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGI_IRF <;_x-30 rtf'!CAL BU! TRESS ELE\/ATlON ReACTOR BulLUlN~ BuTTRESS DETAILi j

WOLF CREEK r--128° 

          ~                 ...............
    /                                       "--..
 /                                                '-....

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                                             ---+i--JL~t...----EL 2056'-6" I

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              ----r-----
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      '-...._        I      _,/                                                      WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.8-31 ELEVATION                                                  REACfCR

WOLF CREEK TYP WALL* 

                  -"9 (TYP WALL)
                                             ~Y:!
                                                  /

0! R j/ /~41'18 (TYP WALL) 1 1 EL 2066 6' 1 s'-o" J SECTION Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.8-32 REACTOR BUILDING EQUIPMENT HATCH 0PE~~INC - TYPICAL ~rc**.-.. . ~

WOLF CREEK I I I I I I r- -- ---- ,_ 

                                                                                           -~

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WOLF CREEK UPDATED SAFETY ANALYSIS REPORT r - - ...... .- 7 n 77 

                                                                                                                                                                             ; lt;UKt        J~ c---~-_1 UPENING - UUISlUl                 M~L

WOLF CREEK WOLF CREEK __ t-.-1AJN STEA~-..1, f'..1AJN FEED\-VATER 

                                 ~BLOW DOWN LINES - - - - - -

(PENETRATIONS ARE PARALLEL TO AZIMUTH o* EL.20<9-ri EL2028'-0" 

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                                                                                                                                                   "----~~~~l

WOLI? CREEK 305-285- 

       -                        TIME (SECJ 265-195                 0 245-                     530                0 1540               6 225-                     6540            -   +

11600 \} 200 STEADY STATE x lL e.. 185-lJJ 165- ~ ~ a: ~ 2: lJJ 1-65 25-1 I I I I I I I I 0 .5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 DISTANCE FROM INNER SURFACE (FT.) Rev. o WOLF CREEK DPDATED SAFETY ANALYSIS REPORT FIGURE 3.8-36

WOLF CREEK 

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1> Ci 1;' Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.8-37 FINITE ELEMENT MODEL FOR AXISYMMETRIC LOADS - STR.

WOLF CREEK Rev. () WOLF CREEK _ __,E,-l.,_.._.2=-c1~3~S_'-~0_*_ _ _ _ ___.So'!PRIHI# L/1'1£ UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 8 FINITE ELEMENT MODEL FOR AXISYMMETRIC LOADS - DOME

WOLF CREEK 207~ 6" I

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Rev. 0 Lil WOLP CREEK 

     ~REACTOR BUILDING                                                                                                                                                         UPDATED SAPETY ANALYSIS REPORT FIGURE 3.8;-39 FINITE ELEMENT MODEL FOR AXISYMMETRIC LOADS - FOUNDA.

1'1E DIU M

WOLF CREEK Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.8-40 FINITE ELEMENT MODEL FOR NONAXISYMMETRIC LOADS

WOLF' CREEK t-----+---+----1-----4---t----1---*-- -*----;-----+----+-- t-----+---+---4--~--+-----*t---- t-----+---+----+---1--+-----r--- -*---*-+----+---+-- t-----+--+----1------4---t------r--- t----+---+----+-~--+----r------ t----+---+----+-~--+----1---- Rev. 0 

                                     -------,--~-~~--------

WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.8-41 FINITE ELEMENT MODEL FOR EQUIPMENT HATCH - ELEVATION

WOLF CRE:E:K Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.8-42 FINITE ELEMENT MODEL FOR EQUIPMENT HATCH - PL.AN

WOLF CRt:t:K 0 0 0 0 () 0 ("") 00 0 (() f".... (j) 0 0 (j) co 1..(") 

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Rev.. I) WOLF CREEK UPDATED SAFETY ANALYS..!.§_.!:EPO !!:!:._. FIGURE 3. 8-43 FINITE ELEMENT MODEL FOR PERSONNEL HATCH

WOLF CREEK HOIST ACCESS PLATFORMS EL 2056!.. 6" I ti EL 2047-6 1------EXTER lOR FACE 1/4" L!NER !t 1.**~ *:. Rev. 0 

                                         -~S~E~C~T_I_O_N____~~_A_ 1-~~~~=w=o~L~F~C~R=E=E~K~~==~~
                                                               \~)         UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 8-44 ELEVATION (FROM INSIDE)

REACTOR BUILDING EQUIPMENT HATCH

WOLF CREEK

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REACTOR BUIL~~~~-p~~~ON~EL HATCH' I

WOLF CREEK 14'-o" 12'-e" 

           't_ AUX. ACCESS
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REACTOR BUILDING Rev. 0 WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.8-46 REACTOR BUILDING AUXILIARY ACCESS HATCH

WOLF CREEK All ~:.J;;,ND 

                          \2 INSERT t  (U T1

______ t,g~ \2 INSERT f (U Tl OR \j INSERT f (U T l 011 \j INSERT f (U T J NOUl£ NruZL~-------===J!~~~~~~;~~ [L -_-o 

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TYP PIPE PENETRATION-TYPE 3 PENETRATION SPARE BLANK! NG TYPICAL PIPE PENETRATION- TYPE TYPICAL PIPE PENETRATION* TYPE 2 FLUED HEAD PENETRATION CLOSURE PLATE PENETRATION Rev. 0 WOLP CIIBBK OPOAYID BAPBYI ANALIBIB RBPORY FIGURE 3.8-47 REACTOR BUILOING TYPICAL PIPE PENETRATION

WOLF CREEK LEAD IMPREGNATED Sl.ICOHE F0Uot 

   '~ ~
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(All SIDES) 

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    *'                                        RECESS COMPAESSIBLF MAfl l*(JYPJ
                                                                         -~".., ""'*            PLAN- NOZZLE FOR FUEL TRANSFER TUBE
   *-~-~-!_                                                                ~    .

LEAD MPREGNATED -R>R CARRON SlFEl SOLID SILICONE ELASTOMER Nf'VZLE 

                               ...                                                                                       tiOLF  CREEK Rev. 0
                               '*"*** -*""                                                                      UPDATED SAFBTI ARALISIS REPORT FIGURE 3. 8-48 REACTOR BUilDING FUEl TRANSFER FUEL TRANSFER TUBE                  PLAN                                                      PENETRATION

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WOLF CREEK REACTOR BLDG.: REACTOR I 01 1*0 REACTOR 180. --fi~r-DETAIL ED STEAM--~ GENERATOR (TYP.) DETAIL Rev. 0 PLAN* STEAM GENERATOR lfOLP CREEK UPPER LATERAL SUPPORTS DPDAYII:D SAPEYI AMALISIS REPORY fiGURE 3.8-53 STEAH GENERATOR SUPPORT SYSTEH - UPPER SUPPORTS

WOLF CREEK REACTOR 

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1 ' ( TYP.) :l.":_:;;- W I:" I DETAIL@ Rev. 0 TYP. SUPPORTS WOLF CREEK REACTOR UPDATED SAFETY ANALYSIS REPORT BLDG. FIGURE 3.8-54 PLAN -STEAM GENERATOR STEAM GENERATOR SUPPORT SYSTEM - LOWER SUPPORTS LOWER LATERAL SUPPORTS

WOLF CREEK __ ff\_~TEAM GEN. COLUMNS F--'---1 (TYP. OF 4 ) I I 

                                                   ~    TRUNNION
                    --18--  -----+-...---EL. 2050'-11 1/16"                    DETAIL@
                                                    \---- ~ SUPPORT     STRUTS I II
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                                                          <[ SUPPORT RING
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   . : :. ~-.                       *~ ,"

Cf. LOWER LATERAL t If SUPPORTS EL. 2022-5 VERTICAL COLUMNS ~ HOT LEG EL. 2014-6 STEAM GENERATOR ELEVATION VIEW Rev. 0 WOLF CREEK UPDATBD SAFETY ANALYSIS REPORT FIGURE 3.8-55 STEAM GENERATOR SUPPORT SYSTEM - ELEVATION

WOLF CREEK I I I OPOA7BO WOLP CRKBI Rev. 0 BAPftf F ARALf!!!..._!!POin' 

                                                     - ,I IGURE 3.S-SG          ~

REACTOR COOLANT SUPPORT [~~~~ 8 lATERAL L

WOLF CREEK LEG 

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II COLUMNS Rev. 0 

                                                        -WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.8-57 REACTOR COOLANT PUMP SUPPORT DETAILS

WOLF CREEK 

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WOLF CREEK llev. 0 WOLP c*s11 UPDAYBD BAI'BYT AHALIBIB RIPDIY FIGURE 3. 8-59 Jozr- r PLAN AJ lb JOlT-!' JO [L JOir- &* P'LAN Af [L 21117'!1"_ TO [L REACTOR CAVITY PLAN - ELEVATION 2011'-6* TO 2021'-7*

WOLF CREEK u *M!'IO'o- 

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            ~    *--* ~ IlEAC 'lOll II                                                      Rev. 0 WOLP CRIIIIlt UPDATID SAPIITI ARALISIS RIIPORT FIGURE 3. 8-63 SECONDARY SHIELD WALLS -

ELEVATION 2025'-0" TO 2047'-0"

I WOLF CREEK rr- C.J. EL. 20lt0'-o" C.J,  !;:UL ::tOlll'-0" ~ i 

 . . 2020'*0" 1&:

C.J.I:L 2012'*0" ..,: 

                                       ~WIELS--._

"' 2000'- Q' EL l!H'* f' Re". 0 WOLF CRBB.tt UPDA'f'ltl SAJ!'BTY UALYSIS UPOR'I' SECTION SECTION FIGURE 3. 8-64 SECONDARY SHIELD WAllS - SECTIONS

WOLF CREEK EL 2065'-9" i '. -;.: . CJ. EL2060'-0" 

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                                                                                                                #11 DWLS El 20od- 0' t#11 OWC!*..                                                           ,
                '--*11 OWLS LcJ.                I                          r------lf!___U:,~::~4+-C-J-,../

SICTIQH@ Rev. 0 

                                                                                                          *WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3.8-65 SECONDARY SHIELD WALLS -

ADDITIONAL SECTIONS

WOLF CREEK 1 

                      , PRES SUR JZER                                 PRESSURIZER 36'-3" TO <i REACTOR BLDG.
            /. "'
    /
         ,.~'

( .. 

                                      <t_ RING GIRDER AND ANCHOR BOLTS PRESSURIZER UPPER LATERAL SUPPORTS                                    Rev. U PRESSURIZER SUPPORT
~"Effi1v~F,~~~:,~*~~51 ~y=SI:S~d:~:;H;]

AT EL 2029~6" UPDATED SAFE'l'Y FIGURE 3.8-66 PRESSURIZER SUPPORTS

PRESSURIZER 

      ~---+---t------t-'r SHIELD WALL 1" RING lr. BY ~

SHIM PACK 

                                  't_ UPPER LATERAL SUPPORT EL 2055'-4 3Vs"
                           *--v
  • EL.2029'-6"7 t:ONN.
                                                          '~

EL 2029'-6" Rev, 0 PRESSURIZER WOLF CREEK UPDATED SAFETY ANALYSIS REPORT FIGURE 3. 8-67 CONNECTION PRESSURIZER SUPPORT DETAILS

WOLF CREEK r.£TAo;TO:lDG ....... t + I.* 1:11' f 

                                                        .,., C!) 12                    t   0
                                                        ,,..k       IIUCTOII IVIIO)
                              -II£FV[L!NO    POOL- --**
                                    *** ~ 12 L__ _ -**~
            . 0 IZ FIGURE 3. 8-68 REFUELING CANAl - TYPICAl PLAN PLAN (5)        EL. 2011'- 6 TO EL. 20t7~5"

WOLF CREEK ( CRANE RAIL ACTOR & CRANE RAIL EACTOR BLDG 

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        ~l204!1'=§"

I:{ 1\: ,.,I .6@12-f c.J EL 204!1'*1!" C.J. EL 2044'*2" ~i. .~ [-*s @12 *s@12-e @ 

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                                   "iHOl12-c.....,

J 

                                                                                                                                                       +

t CJEL2030'-l' I~ +1 

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S.S. LINER l i '"*~'---* 1----"" 

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1 1  !--"~ 11~0* 

                            ~

12 *3" 

                                                  *----l 0

18@12-* ....---- *le(Q) *~ --. 4'-o~ 1/4" 1 1/4' 4!.0 c 

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