ML20235N495
| ML20235N495 | |
| Person / Time | |
|---|---|
| Site: | 05000601 |
| Issue date: | 01/31/1989 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20011C562 | List: |
| References | |
| RESAR-SP-90, NUDOCS 8903010247 | |
| Download: ML20235N495 (196) | |
Text
-__ -_______ _______________ ________ _
l I
l WESTINGHOUSE CLASS 3 O AMENDMENT 2 TO RESAR-SP/90 PDA MODULE 7 STRUCTURAL / EQUIPMENT DESIGN O
l I
O O
WEC P.O. Box 355 Pittsburgh, PA 15230 O WAPWR-S/E AMEN 0 MENT 2 JANUARY, 1989 7106e:1d 8903010247 890214 PDR ADOCK 05000601 K PDC
7.
AMENDMENT 2 TO RESAR-SP/90 PDA MODULE 7 STRUCTURAL / EQUIPMENT DESIGN O INSTRUCTION SHEET Place pages A2-1 through A2-47 (Questions / Responses) af ter page 210-14 of Amendment 1 in the Questions / Answers Section to Module 7. I Replace current pages 11 through xvii with revised pages ii through xvii.
Replace current page 1.6-1/l.6-2 with revised page 1.6-1/1.6-2.
Replace current page (Sheet 41 of 41) f rom Table 3.2-1 with' revised page (Sheet 41 of 41) f rom Table 3.2-1.
Replace current pages 3.5-13 through 3.5-16 with revised pages 3.5-13 through 3.5-16.
Replace current page 3.6-7 through 3.6-30 with revised pages 3 6-7 through 3.6-33.
Replace current Figure 3.6-2 with revised Figure 3.6-2.
Replace current page 3.7-3/3.7-4 with revised page 3.7-3/3.7-4.
Replace current pages 3.7-9 through 3.7-32 with revised pages 3.7-9 through 3.7-32. l Replace current Figure 3.7-8 with revised Figure 3.7-8.
O WAPWR-S/E AMENDMENT 2 2106e:1d JANUARY, 1989
s O . AMENDMENT 2 TO RESAR-SP/90 PDA MODULE 7 (cont'd)-
STRUCTURAL / EQUIPMENT DESIGN INSTRUCTION SHEET Add paget 3A.7 and 3A.7-2 af ter Figure 3.7-20.
Replace current page 3.8-21/3.8-22 with' revised page 3.8-21/3.8-22.
Add pages 3A.8-1 through 3A.8-3 af ter page 3.8-22.
Replace current page 3.9-5/3.9-6 with revisad page 3.9-5/3.9-6.
Replace current pages 3.9-37 through 3.9-48 with revised pages 3.9-37 through 3.9-48.
Replace current pages 3.9-55 through 3.9-78 with revised pages 3.9-55 through 3.9-78.
Replace current pages 3.9-83 through 3.9-92 with revised pages 3.9-83 through L 9-93.
Replace current pages 3.11-3 through 3.11-6 with revised pages 3.11-3 through 3.11-7.
O Replace current page 17.0-1 with revised page 17.0-1.
O O ElAPWR-S/E AMENDMENT 2 JANUARY, 1989 2106e:1d
TABLE OF CONTENTS Reference SAR, Section Section Title Pue Status 1.0 -INTRODUCTION AND GENERAL DESCRIPTION OF 1.1 -1 II.
PLANT
1.1 INTRODUCTION
1.1 -1 11 1.2 GENERAL PLANT DESCRIPTION 1.2-1 11 l.2.2 Principal Design Criteria 1. 2 II 1.6 MATERIAL INCORPORATED BY REFERENCE 1. 6-1 11 1.8 CONFORMANCE WITH THE STANDARD REVIEW PLAN 1. 8-1 II 2.0 SITE CHARACTERISTICS 2. 0-1 NA 3.0 DESIGN OF STRUCTURES, COMPONENTS, EQUIP- 3.1 -1 II MENT AND SYSTEMS 3.1 CONFORMANCE WITH NUCLEAR' REGULATORY COMMISSION (NRC) GENERAL DESIGN CRITERIA .3.12 II (GDC) 3.1.1 Overall Requirements 3.1 -1 1 3.1. 2 Protection by Multiple Fission Product 3.1 -6 I Barriers 3.1. 3 Protection and Reactivity Control Systems 3.1-18 I 3.1. 4 Fluid Systems 3.1-27 I 3 .1. 5 Reactor Containment 3.1-44 I 3.1.6 Fuel and Reactivity Control 3.1-51 1 O- 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, 3.2-1 II EQUIPMENT AND SYSTEMS 3.2.1 Seismic Classification 3.2-1 I 3.2.1.1 Definitions 3.2 1 I O' 3. 2.1. 2 Classifications 3.2-2
- I 3.2.2 Classification Systems 3.2-3 I 3.2.P.1 Nuclear Safety Classification 3.2-3 I O
WAPWR-5/E DECEMBER, 1984 2106e:1d ii
I TABLE OF CONTENTS (Cont.)
Reference ]
SAR )
Section I Section Title Page Status 3.2.2.2 Seismic Classification 3.2-3 I 3.2.2.3 Codes and Standards 3.2-4 I 3.3 WIND AND TORNADO LOADINGS 3. 3-1 I j l
3.3.1 Wind Loadings 3. 3-1 1 3.3.1.1 Design Wind Velocity 3.3-1 1 3.3.1.2 Determination of Applied Forces 3.3-1 I 3.3.2 Tornado Loadings 3.3-2 I 3.3.2.1 Applicable Design Parameters 3.3-2 I 3.3.2.2 Determination of Forces on Structures 3.3-3 1 3.3.2.3 Ability of Category I Structures to 3.3-4 I Perform Despite Failure of Structures Not Designed for Tornado Loads 3.3.3 Design and Analysis Procedures 3.3-4 1 3.3.4 References 3.3-4 1 3.4 WATER LEVEL (FLOOD) DESIGN 3.4-1 I 3.4.1 Flood Protection 3. 4-1 1 3.4.1.1 External Flood Protection 3.4-2 1 3.4.1.1.1 Structural Flood Protection 3.4-2 1 3.4.1.1.2 Surface Drainage System 3.4-3 1 3.4.1.2 Flood Protection for Flooding from 3.4-3 I Component Failures 3 . 4 .1. 3 Permanent Dewatering Systems 3.4-4 I 3.4.2 Analysis Procedures 3.4-4 I 3.4.2.1 Analysis Procedures for External Flooding 3.4-4 I 3.4.2.2 Analysis Procedures for Flooding f rom 3.4-5 I Component O
WAPWR-S/E DECEMBEP, 1984 2*06e:ld iii
D\
Q TABLE OF CONTENTS (Cont.)
Reference SAR Section Section Title Pace Status 3.5 MISSILE PROTECTION 3.5-1 I
- 3. 5 '.1 Missile Selection and Description 3.5-1 I 3.5.1.1 Internally Generated Missiles (Outside 3.5-2 I Containment) 3.5.1.1.1 Rotating Component Failure Missiles 3.5-2 1 3.5.1.1.2 Pressurized Component Failure Missiles 3.5 I 3.5.1.2 Internally Generated Missiles (Inside 3.5-5 I Containment) 3.5.1.2.1 Control Rod Drive Mechanisms 3.5-6 I 3.5.1.2.2 Valves 3.5-7 I 3.5-7 I
~
3.5.1.2.3 Temperature and Pressure Sensors 3.5.1.2.4 Other Missiles 3.5-8 I 3.5.1.3 Turbine Missiles 3.5-9 I 3.5.1.4 Missiles Generated by Natural Phenomena 3.5-9 I 3.5.1.5 Missiles Generated by Events Near the Site 3.5-10 I 3.5.1.6 Aircraft Hazards 3.5-10 1 3.5.1.7 Gravity-Generated Missiles 3.5-10 1 3.5.2 Structures, Systems, and Components to be 3.5-10 I Protected from Externally Generated O Missiles V 3.5.2.1 General 3.5-10 I 3.5.2.2 Missile Barriers Within Containment 3.5-11 1 3.5.2.3 Barriers for Missiles Generated Outside 3.5-12 I of Plant Structures Os 3.5.2.4 Missile Barriers Within Plant Structures 3.5-12 I Other than Containment O )fAPWR-S/E DECEMBER, 1984 2106e:1d iv
l.
! 1 TABLE OF CONTENTS (Cont.)
Reference SAR Section Section Title Page Status 3.5.3 Barrier Design Procedures 3.5-13 I 3.5.4 Missile Protection Interface Requirement 3.5-16 I 3.6 PROTECTION AGAINST THE DYNAMIC EFFECTS 3. 6-1 I ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING 3.6.1 Postulated Piping Failures in Fluid 3.6-1 1 Systems Inside and Outside Containment 3.6.1.1 Design Bases 3. 6-1 I I 3.6.1.2 Description 3.6-5 I 3.6.1.3 Safety Evaluation 3.6-5 I 3.6.2 Determination of Break Locations and 3.6-5 I Dynamic Effects Associated with the Postulated Rupture of Piping 3.6.2.1 Criteria Used to Define High/ Moderate- 3.6-6 I Energy Break / Crack Locations and Configurations 3.6.2.1.1 High-Energy Break Locations 3.6-6 I l
- 3. 6. 2.1. 2 Types of Breaks / Cracks Postulated 3.6-14 I 3.6.2.1.2.1 ASME Section III Piping Other than RCL 3.6-14 I Piping - High Energy
- 3. 6. 2.1. 2. 2 Nonnuclear Piping - High Energy 3.6-14 1 3.6.2.1.2.3 ASME Section III and Nonnuclear Piping 3.6-15 I Moderate Energy 3.6.2.1.2.4 Cracks in High Energy Piping 3.6.16 I 3.6.2.1.3 Break / Crack Configuration 3.6-18 I 3.6.2.1.3.1 High-Energy Break Configuration 3.6-18 I 3.6.2.1.3.2 Moderate-Energy and High Energy 3.6-20 I Crack Configuration WAPWR-S/E AMENDMENT 2 2106e:ld v JANUARY, 1989
[ TABLE OF CONTENTS (Cont.)
Reference
.SAR Section Section Title Pa,qe. Status
~
3.'6.2.2 Analytical Methods to Define Forcing 3.6-20 1 l Functions and Response Models O 3.6.2.2.1 Forcing Functions for Jet Thrust 3.6-20 I 3.6.2.2.1.1 Time Functions of Jet Thrust Force on 3.6-20 I Intact Reactor Coolant Loop (RCL) Piping
, 3.6.2.2.2 Dynamic Analysis of the Reactor Coolant 3.6-23 I Loop Piping and Equipment Supports l 3.6.2.3 Dyaamic Analysis Methods'to Verify 3.6-23 I !
Integrity and Operability 3.6.2.3.1 Dynamic Analysis Methods.to Verify 3.6-23 I Integrity and Operability for Other O than RCL 3.6-23 I I2 3.6.2.3.2 Dyna:ic Analysis Methods to Verify Integrity and Operability for the RCL 3.6.2.3.2.1 General 3.6-23 I 3.6.2.3.2.2 Large.RCS Piping 3.6-25 I 3.6.2.3.2.3 Small Branch Lines 3.6-26 I l 3.6.2.3.3 Types of Pipe Whip Restraints 3.6-26 I 3.6.2.3.3.1 Pipe Whip Restraint: 3.6-26 I 3.6.2.3.4 Analytical Methods 3.6-27 I 3.6.2.3.4.1 Pipe Whip Restraints 3.6-27 I 3.6.2.4 Guard Pipe Assembly Design Criteria 3.6-30 1 3.6.2.5 Material to be Submitted for the 3.6-30 I Operating License Review 3.6.2.6 References 3.6-30 I 3.7 SEISMIC DESIGN 3.7-1 I 3.7.1 Seismic Input 3.7-1 1 3.7.1.1 Design Response Spectra 3.7-2 I WAPWR-S/E AMENDMENT 2 2106e:ld vi JANUARY, 1989 1
TABLE OF CONTENTS (Cont.)
Reference SAR Section Section Title Page Status 3.7.1.2 Design Time-History 3.7-2 1 3.7.1.3 Critical Damping Valves 3.7-2 I )
3.7.1.4 Supporting Media for Seismic Category I 3.7-3 I Structures 1 3.7.2 Seismic System Analysis 3.7-3 I 3.7.2.1 Seismic Analysis Methods 3.7-4 I 3.7.2.2 Natural Frequencies and Response Loads 3.7-4 I 3.7.2.3 Procedure Used for Modeling 3.7-5 1 3.7.2.4 Soil-Structure Interaction 3.7-5 I 3.7.2.5 Development for. Floor Response Spectra 3.7-7 I 3.7.2.6 Three Components of Earthquake Motion 3.7-8 I i 3.7.2.7 Combination of Hodal Responses 3.7-8 I 3.7.2.8 Interaction of Non-Category I Structures 3.7-8 I j With Seismic Category I Structures I 3.7.2.9 Ef fects of Parameter Variations on Floor 3.7-9 I Response Spectra 3.7.2.10 Use of Constant Vertical Static Factors 3.7-9 I 3.7.2.11 Method Used to Account for Torsional 3.7-9 I Effects 3.7.2.12 Comparison of Responses 3.7-9 I l 3.7.2.13 Methods for Seismic Analysis of Dams 3.7-9 I 3.7.2.14 Determination of Seismic Category I 3.7-10 I I Structure Overturning Moment l 3.7.2.15 Analysis Procedure for Damping 3.7-10 I 3.7.3 Seismic Subsystem Analysis 3.7-10 1 Seismic Analysis Methods 3.7-11 I 2l3.7.3.1 3.7.3.2 Determination of Number of Earthquake 3.7-11 I Cycles WAPWR-S/E AMENDMENT 2 2106e:ld vii JANUARY, 1989
i, l-TABLE OF CONTENTS.(Cont.)_
Reference
-SAR
~ Section Section Title Pigg Status 3.7.3.3 Procedure Used for the Modeling :. 3. 7-12 I 3.7.3.4 Basis for Selection'of Frequencies 3.7-14 IL 3.7.3.5 Use of Equipment Static Load Method 3.7-14 I=
of Analysis
+ 3.7.3.6 Three Components of Earthquake Motion 3.7-15 I 3.7.3.7 Combination of Modal Responses 3.7-15 I 3.7.3.8 Analytical Procedures for Piping 3.7-19 1 3.7.3.9 Multiple Supported Equipment Components 3.7-19 I with Distinct Inputs 3.7.3.10 Use of Constant Vertical Static Factors 3.7-21 1 3.7.3.11 Torsional Effects of Ecc'entric Masses 3.7-21 1 3.7.3.12 Buried Seismic Category I Pipirig Systems 3.7-22 I and Tunnels 3.7.3.13 Interaction of Other Piping with Seismic 3.7-22 I Category I Piping 3.7.3.14 Seismic Analyses for Reactor Internals 3.7-22 I (Core, Core Supports, Mechanisms) 3.7.3.15 Analysis Procedure for Damping 3.7-23 I 3.7.4 Seismic Instrumentation 3.7-23 I 3.7.4.1 Comparison with Regulatory Guide 1.12 3.7-24 I 3.7.4.2 Location and Description of Instrument 3.7-24 1 3.7.4.2.1 Time-History Accelerograph 3.7-24 I 3.7.4.2.2 Seismic Switch 3.7-26 1 3.7.4.2.3 Triaxial Spectrum Recorder 3.7-26 I 3.7.4.2.4 Triaxial Peak Accelerograph 3.7-27 1 3.7.4.2.5 Criteria for Instrument Location 3.7-28 I 3.7.4.2.6 Seismic Instrumentation Control Panel 3.7-28 I WAPWR-S/E DECEMBER, 1984 2106e:1d viii
r-TABLE OF CONTENTS (Cont.)
i l Reference SAR Section Section Title Page Status 3.7.4.3 Control Room Operator Modifications 3.7-28 I 3.7.4.4 Comparison of Measured and Predicted 3.7-29 I Responses 3.7.4.5 Inservice Surveillance 3.7-30 I 3.7.5 References 3.7-30 1 2 3.7A Appendix - Computer Code DEBLIN2 3A.7-1 1 3.8 DESIGN OF CATEGORY I STRUCTURES 3.8-1 I
, 3.8.1 Concrete Containment 3.8-1 1 3.8.2 Steel Containment 3. 8-1 1 3.8.2.1 Description of the Containment 3.8-1 1 3.8.2.2 Applicable Codes, Standards, and 3.8-2 I Specifications 3.8.2.3 Loads and Load Combinations 3.8-3 1 3.8.2.4 Design and Analysis Procedures 3.8-3 1 3.8.2.5 Structural Acceptance Criteria 3.8-4 1 3.8.2.6 Materials, Quality Control, and Special 3.8-4 I Construction Techniques 3.8.2.7 Testing and Inservice Inspection 3.8-4 I Requirements 3.8.3 Concrete and Steel Internal Structures 3.8-5 1 3.8.3.1 Description of the Internal Structures 3.8-5 1 3.8.3.2 Applicable Codes, Standards, and 3.8-6 I Specifications 3.8.3.3 Loads and Load Combinations 3.8-7 I 3,8.3.4 Design and Analysis Procedures 3.8-7 I 3.8.3.5 Structural Acceptance Criteria 3.8-7 I 3.8.3.6 Materials 3.8-7 I 3.8.3.7 Testing and Inservice Inspection 3.8-8 I Requirements WAPWR-S/E AMENDMENT 2 2106e:ld ix J ANUARY , 1989 I
TABLE OF CONTENTS (Cont.)_ )
Reference-i i
SAR Section Section Title Page- Status i
3.8.4 Other Seismic Category 1 Structures 3.8-8 I 3.8.4.1 ,
Description of the Structures 3.8-8 I 3.8.4.2 Applicable Codes, Standards and 3.8-9 I Specifications 3.8.4.3 Luads and Load Combinations 3.8-9 I
- 3. 8. 4 . 3.1 Loads 3.8-9 I 3.8.4.3.1.1 Normal Loads 3.8-10 I 3.8.4.3.1.2 Environmental Loads 3.8-11 I 3.8.4.3.1.3 Design Basis Accident Loads 3.8-12 I 3.8.4.3.2 Load Combinations 3.8-13 I
/ 3.8.4.4 Design and Analysis Procedures 3.8-13 I 3.8.4.5 Structural Acceptance Criteria 3.8-14 I 3.8.4.6 Materials 3.8-14 I 3.6.4.7 Testing and Inservice Surveillance 3.8-16 I
' Requirements 3.8.5- Foundations 3.8-16 I 3.8.5.1 Description of the Foundations 3.8-16 I 3.8.5.2 Applicable Codes Standards and 3.8-16 I Specifications 3.8.5.3 Loads and Loading Combinations 3.8-16 I 3.8.5.4 Design and Analysis Procedures 3.8-16 I 3.8.5.5 Structural Acceptance Criteria 3.8-17 I 3.8.5.6 Materials 3.8-18 I
- 3. ,8. 5. 7 Testing and Inservice Inspection 3.8-18 I Requirements 3.8A Appendix - Computer Code WECAN 3 A8-1 1 2 3.9 MECHANICA; SYSTEMS AND COMPONENTS 3.9-1 11 3.9.1 Special Topics for Mechanical Components 3.9-1 I
\ 3.2 1.1 Design Transients 3.9-1 I WAPWR-S/E AMENDMENT 2 2106e:1d x JANUARY, 1989 L _ _ -_---___ -__ _ __-__ _ _____________________-__
TABLE OF CONTENTS (Cont.)
Reference SAR Section Section Title Page Status 3.9.1.1.1 Level A Service Conditions (Normal 3.9-3 I Conditions)
- 3. 9 .1.1. 2 Level B Service Conditions (Upset 3.9-19 I Conditions) 3.9.1.1.3 Level C Service Conditions (Emergency 3.9-29 I I Conditions) 3.9.1.1.4 Level D Servcie Conditions (Faulted 3.9-31 I Conditions 3.9.1.1.5 Test Conditions 3.9-35 1 3.9.1.2 Computer Programs Used in Analysis 3.9-37 I 3.9.1.2.1 NPB Systems and Components 3.9-37 I 3.9.1.3 Experimental Stress Analysis 3.9-38 I l 3.9.1.4 Consideration for the Evaluation of the 3.9-38 I Fauited Condition 3.9.2 Dynamic Testing and Analysis 3.9-38 II 3.9.2.1 Piping Vibration, Thermal Expansion, 3.9-38 I and Dynamic Effects 3.9.2.2 Seismic Cussification Testing of Safety- 3.9-42 I Related Mechanical Equipment 3.9.2.3 Dynamic Response Analysis of Reactor 3.9-45 I Internals Under Operational Flow Transients and Steady-State Conditions 3.9.2.4 Preoperational Flow-Induced Vibration 3.9-45 I Testing of Reactor Internals 3.9.2.5 Dynamic System Analysis of the Reactor 3.9-45 I Internals Under Faulted Conditions 3.9.3 ASME Code Class 1, 2, and 3 Components, 3.9-45 I Component Supports, and Core Support Structures WAPWR-5/E DECEMBER, 1984 2106e:ld xi
i f,h V TABLE OF CONTENTS (Cont.)
Reference SAR Section Section Title Page Status 3.9.3.1 Loading Combinations, Design Transients, 3.9-45 I i and Stress Limits g
3.9.3.1.1 ASME Code Class 1 Components and Supports 3.9-46 I' 3.9.3.1.1.1 Analysis of the Reactor Coolant Loop 3.9-46 I Piping and Supports '
3 . 9 . 3 .1.1. 2 Class 1 Auxiliary Branch Lines 3.9-55 I 3 . 9 . 3.1.1. 3 Loading Combinations and Stress Limits 3.9-58 I 3 . 9 . 3 .1. 2 ASME Code Class 2 and 3 Components 3.9-58 I and Supports 3.9.3.1.3 Analysis of Primary Components and Valves 3.9-58a I l2 3.9.3.2 Pump and Valve Operability Assurance 3.9-60 I 3.9.3.2.1 Pumps 3.9-60 I 3.9.3.3.2 Valves 3.9-62 I 3.9.3.2.3 Pump Motor and Valve Operator 3.9-64 I Qualification 3.9.3.2.4 Active ASME Code Class 2 and 3 Pumps 3.9-65 I l2 3.9.3.3 Design and Installation Details for 3,9-66 I Mounting of Pressure Relief Devices 3.9.3.3.1 Pressure Relief Devices on NPB Components 3.9-66 1 3.9.3.3.2 Other Pressure Relief Devices on 3.9-68 I )
v Components 3.9.3.4 Component and Piping Supports 3.9-69 I 3.9.3.4.1 ASME Code Class 1 Component Supports 3.9-70 I 3.9.3.4.1.1 Primary Component Supports Models 3.9-70 I O and Methods ASME Code Class 2 and 3 Supports 3.9-71 I 3.9.3.4.2 l
3.9.3.4.3 Snubbers Used as Component Supports 3.9-72 1 3.9.3.5 Design of HVAC Ductwork and Supports 3.9-73 I l2 WAPWR-S/E AMENDMENT 2 2106e:1d xii JANUARY, 1989 )
l j
TABLE OF CONTENTS (Cont.)
Reference SAR Section Section Title Page Status 3.9.4 Control Rod Drive Systems 3.9-73 11 2 3.9.5 Reactor Pressure Vessel Internals 3.9-74 II 3.9.5.1 Loading Conditions 3.9-74 1 3.9.5.2 Reactor Vessel and Internals Modeling 3.9-75 I 3.9.5.3 Analytical Methods 3.9-76 I l 3.9.6 Inservice Testing of Pumps and Valves 3.9-77 1 3.9.6.1 Inservice Testing of Pumps 3.9-77 I i 3.9.6.2 Inservice Testing of Valves 3.9-77 I 3.9.6.3 Relief Request 3.9-78 1 3.10 SEISMIC QUALIFICATION OF SEISMIC CATEGORY 3.10-1 I I INSTRUMENTATION AND ELECTRICAL EQUIPMENT 3.10.1 Seismic Qualification Criteria 3.10-1 I 3.10.2 Methods and Procedures for Qualifying 3.10-2 I Electrical Equipment and Instrumentation 3.10.2.1 Seismic Qualification by Type Test 3.10-3 1 3.10.2.2 Seismic Qualification by Analysis 3.10-4 1 3.10.2.3 Combined Analysis and Testing 3.10-5 I 3.10.3 Methods and Procedures of Analysis or 3.10-5 I Testing of Supports of Electrical Equipment and Instrumentation 3.10.4 Operating License Review 3.10-6 I 3.10.5 References 3.10-6 I 3.11 ENVIRONMENTAL DESIGN OF MECHANICAL 3.11-1 I AND ELECTRICAL EQUIPMENT 3.11.1 Equipment Identification and Environmental 3.11-1 I Conditions 3.11.2 Qualification Tests and Analyses 3.11-2 1 3.11.2.1 Environmental Qualification criteria 3.11-2 I WAPWR-S/E AMENDMENT 2 O
2106e:1d xiii JANUARY, 1989
TABLE OF CONTENTS (Cont.)
Reference SAR Section Section Title Page. Status 3.11.2.2 Performance Requirements for Environmental 3.11-2 I n
Q 3.11.2.3 Qualification Methods and Procedures for Environmental 3.11-3 I Qualification 3.11.3 Qualification Test Results. 3.11-3 1 3.11.4 Loss of Ventilation 3.11-3 I 3.11.5 Estimated Chemical and Radiation 3.11-3 I Environment 3.11-4 2 3.11.6 References 1 4.0 REACTOR 4. 0-1 NA 5.0 REACTOR COOLANT SYSTEM AND CONNECTED 5. 0-1 NA SYSTEMS 6.0 ENGINEERED SAFETY FEATURES 6. 0-1 NA 7.0 INSTRUMENTATION CONTROLS 7. 0-1 NA 8.0 ELECTRIC POWER 8.0-1 NA 9.0 AUXILIARY SYSTEMS 9. 0-1 NA 10.0 STEAM AND POWER CONVERSION SYSTEM 10.0-1 NA 11.0 RADI0 ACTIVE WASTE MANAGEMENT 11.0-1 NA 12.0 RADIATION PROTECTION 12.0-1 NA 13.0 CONDUCT OF OPERATIONS 13.0-1 NA 14.0 INITIAL TEST PROGRAM 14.0-1 NA 15.0 ACCIDENT ANALYSES 15.0-1 NA 16.0 TECHNICAL SPECIFICATIONS 16.0-1 NA 17.0-1 O
U 17.0 OVALITY ASSURANCE 17.0-1 II II 17.1 QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION 17.1.1 References 17.0-1 II O
WAPWR-S/E AMENDMENT 2 2106e:1d xiv JANUARY, 1989
\
{ -
l q l V' TA8LE OF CONTENTS (Cont.)
l LIST OF TABLES 1
i m
Page I i Number Title I
- 1. 6-1 Material Incorporated by Reference 1.6-2
- 1. 8-1 Standard Review Plan Deviations 1.8-2
(),/
/
3.2-1 Classification of Structures, Cotponents and -
Systems 3.2-2 Principal Codes and Standards -
3.5-1 Summary of Control Rod Drive Mechanism 3.5-19 Missile Analysis 3.5-2 Valve-Missile Characteristics 3.5-20 3.5-3 Piping Temperature Element Assembly 3.5-21
- Missile Characteristics 3.5-4 Characteristics of Other Missiles Postulated 3.5-22 rm Within Reactor Containment
( )
%/ Essential or Non-essential; High-Energy, or 3.6-29 3.6-1 Moderate - Energy Systems 3.7-1 Regulatory Guide 1.61 Damping Values for 3.7-32 Structures and Components 3.7-2 Seismic Monitoring Instrumentation Requirements 3.7-33 3.8-1 Containment Load Combinations and Load Factors 3.8-19 3.8-2 Stress Intensity Limits for Steel Containment 3.8-20 3.8-3 Load Combinations and Load Factors for 3.8-21 Category 1 Concrete Structures V 3.8-4 Load Combinations and Load Factors for 3.8-25 Category 1 Steel Structures 3.9-1 Summary of Reactor Coolant System Design Transients 3.9-79 3.9-2 Pump Starting / Stopping Conditions 3.9-83
(~
U 3.9-3 Stress Criteria for ASME B&PV Code Section III 3.9-84 Class 1 Components and Supports O WAPWR-S/E AMENDMENT 2 2106e:1d xv J ANUARY , 1989
TABLE OF CONTENTS (Cont.)
LIST OF TABLES (Cont.)
Number Title Page 3.9-4 Stress Criteria for ASME B&PV Code Section III 3.9-86 Class 2 and 3 Components 3.9-5 Loading Combinations for ASME Class 1 3.9-87 Components and Supports 3.9-6 Damping Valves for Auxiliary Piping Systems 3.9-88 3.9-7 Load Combinations and Acceptance Criteria for 3.9-89 Pressurizer Safety and Relief Valve Piping -
2 Upstream of Valves 3.9-8 Load Combinations and Acceptance Criteria for 3.9-90 Pressurizer Safety and Relief Valve Piping Seismically Designed Downstream Portion 3.9-9 Definitions of Load Abbreviations 3.9-91 3.9-10 Load Combinations for Pressurizer Saf ety 3.9-92 and Relief Valve Nozzles and Support Brackets 3.9-11 Loading Combinations for ASME Code Class 2 and 3 3.9-93 Components and Supports for the NPB 3.10-1 Seismic Category I Instrumentation and Electrical 3.10-7 Equipment 3.11-1 Safety-Related Equipment 3.11-5 O
O WAPWR-S/E AMENDMENT 2 2106e:ld xvi J ANUARY , 1989
f^g
.() TABLE OF CONTENTS (Cont'd)
LIST.0F FIGURES
/ Number Title
- 3. 6-1 Analysis Criteria for Mechanistic Pipe Break Approach 3.6-2 Loss of Reactor Coolant Accident Boundary Limits 3.6-3 Typical U-Bar Restraint 3.6-4 Typical EAM Restraint 3.7-1 Safe Shutdown Earthquake Horizontal Response Spectra 3.7-2 Safe Shutdown Earthquake Vertical Response Spectra 3.7-3 Operating Basis Earthquake Horizontal' Response Spectra 3.7-4 Operating Basis Earthquake Vertical Response Spectra 3.7-5 Comparison of Design Response Spectra - Horizon' cal Direction 1 3.7-6 Comparison of Design Response Spectra - Horizontal Direction 2 3.7-7 Comparison of Design Response Spectra - Vertical Direction 3.7-8 Alternative Damping Valves for Piping Systems 2 3.7-9 Analytical Model - Seismic System 3.7-10 Seismic Analysis of Soil-Structure Systems 3.7-11 Enveloped Floor Response Spectra - Operating Deck (N-S) 3.7-12 Enveloped Floor Response Spectra - Operating Deck (E-W).
3.7-13 Enveloped Floor Response Spectra - Operating Deck (Vertical) 3.7-14 Enveloped Floor Response Spectra - RPV Support (N-5) 3.7-15 Enveloped Floor Response Spectra - RPV Support (E-W) 3.7-16 Enveloped Floor Response Spectra - RPV Support (Vertical) 3.7-17 Enveloped Floor Response Spectra - Control Room (N-S)
O- 3.7-18 Enveloped Floor Response Spectra - Control Room (E-W) 3.7-19 Enveloped Floor Response Spectra - Control Room (Vertical) 3.7-20 Post Seismic Event Data Utilization 3.9-1 Reactor Coolant Loop Supports System, Dynamic Structural Model O
O WAPWR-5/E AMENDMENT 2 2106e:1d xvii JANUARY, 1989
A V 1.6 A M.ATERIAl. INCORPORATED BY REFERENCE The WAPWR Structural / Equipment Design Module incorporates, by reference, certain topical reports. The topical reports, listed in Table 1.6-1, have been filed previously in support of other Westinghouse applications.
The legend for the review status code letter follows:
O A -
U.S. Nuclear Regulatory Commission review complete; USNRC acceptance letter issued.
AE - U.S. Nuclear Regulatory Commission accepted as part of the Westinghouse emergency core cooling system (ECCS) evaluation model only; does not constitute acceptance for any purpose other than for ECCS analyses.
B - Submitted to USNRC as background information; no u:.dergoing formal O.i USNRC review.
0 - On file.with USNRC: older generation report with current validity; not actively under formal USNRC review.
U - Actively under formal USNRC review.
O O .
O WAPWR-5/E 1.6-1 DECEMBER, 1984 2106e:1d
TABLE 1.6-1 MATERIAL INCORPORATED BY REFERENCE Westinghouse SAR Topical Revision Section Submitted Revi Report No. Title Number Reference to the NRC Stat _
\
WCAP-7427 Effective Structural Damping Rev. 0 3.7 1/70 0 of the KEP L105 CRDM WCAP-7427 Effective Structural Damping Addendum 1 3.7 12/70 0 of the KEP L105 CRDM WCAP-7558 Seismic Vibration Testing Rev. 0 3.10 10/71 U (Non-Prop) with Sine Beats WCAP-8236(P) Safety Analysis of Eight-Grid Addendum 1 3.7 4/74 A WCAP-8288 17x17 Fuel Assembly for Combined Seismic Loss-of-Coolant Accident WCAP-8252 Documentation of Selected Rev. 1 3.6 5/77 A ,
Westinghouse Structural Analysis Computer Codes 2 WCAP-8370 Westinghouse Energy Systems Rev. 11 17.1 10/88 U Business Unit Quality. Assurance Plan WCAP-8587 Equipment Qualification Data Sup. 1 3.10 2/79 U Packages (Rev. 2) 3.11 WCAP-8587 Methodology for Qualifying Rev. 6 3.10 11/83 U Westinghouse WRD Supplied 3.11 NSSS Safety-Related Electrical Equipment WCAP-3624(P) General Method of Developing Rev. 0 3/10 9/75 U WCAP-8695 Multifrequency-Biaxial Test 8/75 Inputs for Bistables WCAP-8707-P-A MULTIFLEX-FORTRAN-IV Computer Rev. 0 3.6 9/16/77 A ;
(P), Vol I Program for Analyzing Thermal-and II Hydraulic Structure System WCAP-8709-A, Dynamics Vol I and II WCAP-8867 DEBLIN2 - A Computer Code Rev. 0 3/7 11/76 U to Synthesize Earthquake Acceleration Time Historics WC AP-10221 Simplified Pipe Whip Analysis Rev. 0 3.6 12/82 U and Restraint Design Procedures WAPWR-S/E 1.6-2 AMENDMENT 2 2106e:1d JANUARY 1989 i
TABLE 3.2-1 (Sheet 41 of 41)
CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS A)
L p. The quality assurance program to be applied to fire protection systems is described in Branch Technical Position APCSB 9.5-1, Appendix A, attached to Nuclear Regulatory Commission (NRC) Standard Review Plan 9.5.1.
O q. The quality assurance program to be applied to radioactive waste management systems is described in Regulatory Guide 1.143.
- r. Heat tracing and associated heat tracing equipment for the safety injection system is redundant, procured as Class 1E, and seismically and environmentally qualified in accordance with IEEE 323, 344, and 383.
- s. Safety related supports and hangers are to be classified corresponding to the function provided by the supported equipment unless it can be demonstrated by analysis that failure of the support would not jeopardize 2 the function of that equipment. If applicable, Subsection NF of Section III of the Code will be utilized.
O O
I O
O WAPWR-S/E AMENDHENT 2 JANUARY, 1989 4858e:1d
3.5.3 Barrier Design Procedures Missile barriers and protective structures are designed to withstand and absorb missile impact loads in order to prevent damage to safety-related components.
Formulae used for missile penetration calculations (for missiles other than turbine missile) into NPB steel or concrete barriers are based on the report e of the ASCE Committee on Impactive and Impulsive Loads, (ASCE Journal of the Structural' Division, May; 1980):
' Concrete (Modified NDRC Formula) 1.8 0.5 x = [4 KNWd ( ) ] forjs2.0 2
x=KNW(h) +dforj>2.0 where x = penetration depth, inches l W = missile weight, 1bs d = missile diameter, inches N = missile shape factor = 1.0 V = impact velocity (ft/sec)
K = experimentally obtained material coefficient for penetration = S
, e f = concrete compressive strength c
Scabbing thickness,st , and perforation thickness, pt is given by:
=2.12+1.36jfor0.65 1js11.75 2 2
[t =7.91(j)-5.06(j) forjs0.65 .
O l WAPWR-S/E 3.5-13 AMENDMENT 2 2023e:1d JANUARY, 1989
[t =1.32+1.24jfor1.35 5js13.5 2
2
[t =3.19(j)-0.718(j) forj$ 1.35 Steel (Stanford Formula)
(16,000 T2 + 1,500 T) h=
where:
E = critical kinetic energy required for perforation (ft-lb)
D =
missile diameter (in)
S = ultimate tensile strength of the target (steel plate) (psi) ;
T = target plate thickness (in)
W = length of a square side between rigid supports (in)
S2=
length of a standard width (4 in)
The ultimate tensile strength is directly reduced by the amount of bilateral tension stress already in the target. The equation is good within the following ranges:
0.1 < T/D < 0.8, 0.002 < T/L < 0.05, 10 < L/0 < 50, 5 < W/D < 8, 8 < W/T < 100, 70 < Ve < 400 Where:
1 missile length, in, L =
V = vel city, ft/sec, and c
the missile is assumed to be cylindrical.
O i WAPWR-S/E 3.5-14 AMENDMENT 2 !
2023e:1d JANUARY, 1989 ;
I
In using the Modified NDRC and Stanford formulae for missile penetration it is o O assumed that the missile impacts normal to the plane of the wall on a minimum impact area and, in the case of reinforced concrete, does not strike the reinforcing. Due to the conservative nature of these assumptions, the minimum thickness required for missile shields will be taken as the thickness just O' perforated.
O O
O O
O WAPWR-S/E 3.5-15 AMENDMENT 2 2023e:1d JANUARY, 1989
l 4
l Structural members designed to resist missile impact will be designed for flexural, shear, and buckling effects using the equivalent static load obtained from the evaluation of structural response. Stress and strain limits for the equivalent static load will comply with the requirements of applicable codes or specifications except for the area local to the missile impact, where the stress and strain may exceed the allowables provided there will be no loss of function of any safety related system.
In general, Westinghouse-supplied equipment is not designed to withstand the impact of postulated missiles; therefore, the B0P designer considers the effects of V atulated missiles and provides the necessary protection to safety related components as determined by the missile selection bases provided in
. Subsection 3.5.1. l The exception to this is the control rod drive mechanism (CRDM) missile shield, which is supplied by Westinghouse as part of the integrated head. A missile shield structure is provided over the CRDMs to block missiles that might be associated with a fracture of the pressure housing of any mechanism.
This missile shield is a reinforced steel structure attached to the reactor vessel head and located above the CRDHs.
For the case of CRDM housing plug and drive shaft impact, which is the design case, it is assumed that the plug partially perforates the missile shield.
The drive shaft then hits the plug and further penetrates the steel missile shield; the effective thickness of the steel missile shield is more than three times the combined penetration for the design case. The CRDM missile shield is also designed to withstand the dynamic impact loads due to the missile and the water jet.
3.5.4 Missile Protection Interface Requirements The B0P applicant must consider the effects of postulated missiles and provide the necessary protection to safety related components as determined by the bases provided in Section 3.5 of this module. In general Westinghouse O
WAPWR-5/E 3.5-16 DECEMBER, 1984
! 2023e:1d l
l L_________.
(2) Branch intersection points are considered a terminal end O for the branch line unless the following are met: the branch and the main piping systems are modeled in the same static, dynamic, and thermal analyses, and the branch and main run are of comparable size and fixity, i.e., the O nominal size of the branch is at least one-half of that of the main. 5
- b. At all intermediate locations where the following conditions are satisified:
(1) Any intermediate locations where the maximum stress range as calculated by equation (10) and either (12) or (13) exceeds 2.4 Sm (where S,is the design stress intensity) l2 as described in paragraph NB-3653 of the ASME B and PV Code,Section III.
(2) Any intermediate locations where the cumulative usage factor '
exceeds 0.1. l2 B. ASME B and PV Code,Section III - Class 2 and 3 Piping Systems
- 1. Pipe breaks are postulated to occur at terminal ends.
- 2. Pipe breaks are postulated at intermediate locations between terminal ends where the maximum stress value, as calculated by the sum of equations (9) and (10) in subarticle NC-3652 of the ASME B and PV Code,Section III, Reference 3, considering normal and upset plant conditions (i.e., sustained loads, occasional loads, thermal expansion, and an operating basis earthquake 2
(OBE) event) exceeds 0.8 (1.8 Sh + SA)*
O S h
and S A
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 and PV Code,Section III.
WAPWR-S/E 3.6-7 AMENDMENT 2 1999e:1d JANUARY, 1989
C. Nonnuclear Piping (i.e., not ASME Section III Class 1, 2, or 3)
Breaks in nonnuclear piping are postulated at the following locations O !
I in each run:
- 1. For seismically analyzed ANSI B31.1 piping systems, Reference 1:
a) Terminal ends b) At intermediate locations between terminal ends where the maximum stress value, as calculated by the sum of equations 2 (12) and (13) in Section 104.8 of the ANSI /ASME B31.1 Power Piping Code, considering normal and upset plant conditions (i.e., sustained loads, occasional loads, thermal expansion, and an operating basis earthquake (OBE) event) exceeds 0.8 (X+Y). X and Y are the allowable stress values for equations (12) and (13), respectively.
- 2. In the absence of stress analysis, breaks in nonnuclear piping are postulated at the following locations in each run or branch run:
- a. Terminal ends.
- b. Each intermediate fitting, e.g., short and long radius elbows, tees, and reducers; welded attachments; and valves.
D. High-Energy Piping in Containment Penetration Areas Breaks are not postulated in the portions of Class 2 piping between the containment penetration flued-head and five-way restraints (i.e.,
break exclusion zone) provided subject piping meets the following provisions:
WAPWR-S/E 3.6-8 AMENDMENT 2 1999e:1d JANUARY, 1989 l
I
p 1. Stresses do not exceed those specified in Subsection 3.6.2.1.1.B.
V '
- 2. The maximum stress in this piping as. calculated by equation-(9),.
per. paragraph NC-3652'of ASME Section III, when subjected to the g combined loadings of internal pressure, deadweight, and pipe
- h. rupture outside the protective restraints, does not exceed the
. lesser of 2.25 Sh and 1.8 Sy . 2
. 3. The number of circumferential and longitudinal piping welds and branch connections is minimized.
Areas of system piping where no breaks are postulated are as follows: ,
- a. The main steam piping, from the containment penetration flued-head outboard weld, to the upstream weld of the five-way restraint, which is downstream of the main steam isolation valves, including the main steam safety valves and . branch A
Q piping to the main steam saf'ety valves.
- b. The main feedwater piping from the containment penetration to the five-way restraint which is upstream of the isolation valve.
When required for isolation valve operability, structural integrity, or containment integrity, five-way restraints capable of resisting torsional and bending moments produced by a postulated pipe break, either upstream or oownstream of the piping and valves which form the containment isolation boundary, are located reasonably close to the ,
isolation valves or penetration.
The five-way restraints do not prevent the access required to conduct b inservice inspection examinations specified in Section XI of the ASME Code. Inservice examinations completed during each ir.spection interval provide examination of circumferential and longitudinal pipe 1
WAPWR-S/E 3.6-9 AMENDMENT 2 1999e:1d JANUARY, 1989 l
welds within the boundary of these portions of piping during each inspection interval. A 100% volumetric inservice examination is j 2 performed in accordance with IWA-2400, ASME Code,Section XI.
Exemption allowed under Section XI may be applied if justified.
Weleed attachments to these portions of piping for pipe supports or other purposes are avoided. Where welded attachments are necessary, detailed stress analyses are performed to demonstrate compliance with 2 the limits of Subsection 3.6.2.1.1.D-1,2.
The fivo-way restraints outside the containment on the main steam and main feedwater lines are located as close as possible to the containment to accommodate the design for the reactor external ,
building and main steam tunnel and still minimize stresses.
For evaluation of environmental effects (excluding jet impingement) longitudinal breaks, with break flow areas of 1.0 square feet, are postulated in the main steam and feedwater piping. Locations which have the greatest effect on essential equipment are chosen.
E. Piping Within Mechanistic Pipe Break Criteria The criteria below are used to verify that there are no pipe break locations in lines greater than 6 inches nominal diameter in the following high energy systems:
Reactor coolant Emergency core cooling Chemical and volume control Main steam Main feedwater Steam generator blowdown Diesel generator and related systems WAPWR-S/E 3.6-10 AMENDMENT 2 1999e:1d JANUARY, 1989
E rs The mechanistic pipe break approach is used instead of hypothetically located k -
pipe _ ruptures and eliminates the structural analysis associated with these ruptures. Application of this approach is applied to high-energy piping provided:
D (V a. Operating experience, tests, or analyses have indicated no particular susceptibility to failure from effects of intergranular stress corrosion cracking, water hammer, or thermal fatigue.
O V b. Supports of heavy interconnected components (such as reactor vessel, steam generator, and main reactor coolant pump in the reactor coolant system) are designed to withstand normal operation and SSE loads, and loads resulting from any postulated pipe rupture.
Dynamic effects associated with hypothetical full flow area circumferential or longitudinal breaks in the piping need not be considered when application of the mechanistic pipe break approach is technically justifiable in accordance !
with the evaluation criteria described below. The specific dynamic effects excluded are: ,
- a. Pipe whip and reaction forces.
- b. Jet impingement loads.(a)
- c. Subcompartment pressurization such as reactor cavity asymmetric pressurization transients.
- d. Break associated transient loads in unbroken portions of the system such
- as loads on the reactor internals or steam generator internals and pump overspeed.
O The following requirements apply to high-energy piping inside or outside containment:
O
%.)
- a. However, environmental effects, wetting and flooding of surrounding equipment, and spaces due to leakage must be considered.
WAPWR-S/E 3.6-11 AMENDMENT 2 1999e:1d JANUARY, 1989 1
1 l
- a. For purposes of specifying design criteria for emergency core cooling, containment systems, other non-structural engineered safety features, and for the evaluation of environmental effects, loss of coolant (even in the l
piping with applicability of the mechanistic pipe break approach) is !
assumed through an opening equivalent to twice the flow area of the largest diameter pipe in the system, or that pipe which will result in the most limiting accident conditions. {
1
- b. Except as required by (a) above, high-energy piping may be treated with potential of through-wall crack leakage rates equal to: 1) the maximum allowable unidentified leakage conditions associated with the piping, or;
- 2) the leakage for a rectangular crack having a one-half wall thickness j width and a one-half pips diameter length.
The following information is developed in the RESAR-SP/90 FDA for each line for which the mechanistic pipe break approach is applied:
- a. A discussion to support a conclusion that the line is very unlikely to experience stress corrosion cracking, or extreme repetitive loads, or excessive loads such as might occur from thermal or mechanical low and high cycle fatigue or a water hammer.
- b. Identification of types and specifications of all concerned materials; all base metal, forgings and weldments, and safe-ends will be included. The materials properties data and information used in the analysis will be provided, and the sources of all data reported.
- c. Specification of the type and magnitude of the loads applied (forces, ,
bending and torsional moments), their source (s) and method of combination. Identification of the location (s) at which the minimum-margin (e.g., stress-to-strength ratio) occurs for base materials and I weldments and safe-ends. For geometrically complex lines or systems, it may be necessary to analyze several locations to assure that the more i limiting locations are identified.
1 WAPWR-S/E 3.6-12 AMENDMENT 2 1999e:1d JANUARY, 1989 l
1
Step-Wise Analysis Criteria The following analytical steps, illustrated in Figure 3.6-1, assume that circumferentially oriented postulated cracks are limiting. If this is not the q case, then the analysis described in (a) through (c) below will also include C the postulation of axial cracks .and/or elbow cracks. If applied moments (including SSE) are quite low and applied maximum axial forces dominate, relatively long part-through-wall cracks are analyzed to demonstrate that they are stable,
- a. Postulated Fabrication Flaw At the location or locations of (c) above, postulate a fabrication flaw that may be missed during fabrication and preservice inspections or would be permitted by code, whichever is larger. Demonstrate by fatigue analysis that the crack will not grow through the wall or extend significantly in length during plant design life.
- b. Postulated Leakage Crack Even though (a) above demonstrates that a leaking pipe is unlikely, a through-wall crack at the selected location is postulated. The size of the postulated crack should be large enough so that the leakage is assured of detection with adequate margin using 2 times the minimum installed leak detection capability when the pipe (s) is (are) subjected to normal operational loads. If auxiliary leak detection systems are relied on, they will be described.
- c. Stability and Critical Crack Sizes Demonstrate crack size margin by showing that 2 times the postulated leakage crack as defined in (b) above is less than the critical crack size using normal plus SSE loads. In some cases, a limit load analysis may '
suffice for this purpose, however, an elastic plastic fracture mechanics analysis may be used when applicable.
I O WAPWR-S/E 3.6-13 AMENDMENT 2 JANUARY, 1989 1999e:1d l f
E _ _ _ ___
I
! 3.6.2.1.2 Types of Breaks / Cracks Postulated 3.6.2.1.2.1 ASME Section III Piping Other than RCL Piping - High-Energy O
The following types of breaks are postulated to occur at the locations determined in accordance with Subsection 3.6.2.1.1 - A, B and C.
A. In piping of 4 inches nominal diameter or greater, both circumferential and longitudinal breaks are postulated at each selected break location unless eliminated by comparison of longitudinal and axial stresses with the maximum stress as follows. j i
l 1
- 1. If the maximum stress range exceeds the' limits specified in Subsection 3.6.2.1.1.A.2.b or 3.6.2.1.1.B.2 but the circum-ferential stress range is at least 1.5 times the axial stress 2 range, only a longitudinal break is postulated.
- 2. If the maximum stress range exceeds the limits specified in Subsections 3.6.2.1.1.A.2.b or 3.6.2.1.1.B.2 but the axial stress is at least 1.5 times the circumferential stress range, only a 2 circumferential break is postulated.
Longitudinal breaks, however, are not postulated at terminal ends.
B. In piping of nominal diameter greater than 1 inch but less than 4 inches, only circumferential breaks are postulated at each selected break location.
C. No breaks are postulated for piping of nominal diameter 1 inch or less.
3.6.2.1.2.2 Nonnuclear Piping - High-Energy The types of breaks for high energy nonnuclear piping are postulated as discussed in Subsection 3.6.2.1.2.1; the corresponding break locations are determined in accordance with Subsection 3.6.2.1.1.C.
WAPWR-S/E 3.6-14 AMENDMENT 2 1999e:1d JANUARY, 1989
3.6.2.1..'2.3 ASME Section III and Nonnuclear Piping - Moderate-Energy 0p
'Through-wall leakage cracks are postulated in moderate-energy piping including branch runs larger than 1 inch nominal diameter as clarified below: !
()O A. Through-wall leakage cracks are not postulated in those portions of piping between containment isolation valves, provided they meet the requirements of ASME Code,Section III, subarticle NE-1120, and are designed.so that the maximum stress range does not exceed. 0.8 (1.8 S
h
+ S)A for ASME Class 2 and 3 and 0.8 (X+Y) for ANSI B31.1 2 piping, respectively. X and Y are defined in Section 3.6.2.1.1.C.
B. Through-wall leakage cracks are not to be postulated in moderate energy fluid system piping located in an area where a break in the high-energy fluid system is postulated, provided that such cracks do not result in environmental conditions.more limiting than 3 the high-energy pipe break.
C. Through-wall leakage cracks are to be postulated in:
(1) ASME, B and PV Code,Section III, Division 1 - Class 1 piping:
a) At terminal ends when the stresses exceed one half the limit given in Section 3.6.2.1.1A.2.b(1).
b) At intermediate locations defined in 3.6.2.1.1A.2.o(1). ..I2 (2) ASME, B and PV Code,Section III, Division 1 - Class 2 or 3 piping:
a) At terminal ends when the stresses exceed one-half the limit' given in Section 3.6.2.1.1.B.2.
O b) At intermediate locations defined by Section 3.6.2.1.18.2.
WAPWR-S/E 3.6-15 AMENDMENT 2 1999e:Id JANUARY, 1989
l (3) Seismically analyzed ANSI B.31.1 piping: j O
a) At terminal ends when the stresses exceed one-half the limit ,
given in Section 3.6.2.1.10.
I b) At intermediate locations defined by Section 3.6.2.1.1C. g 2
(4) Non nuclear piping, in the absence of stress analysis:
a) At terminal ends, and b) At each intermediate location of potential high stress or fatigue such as pipe fittings, valves, flanges, and welded on attachments.
To simplify analysis, cracks may be postulated to occur everywhere in moderate-energy piping regardless of the stress analysis results to determine the maximum damage from fluid spraying and flooding, with the consequent '
hazards or environmental conditions. Flooding effects are determined on the basis of 30 minutes operator time required to effect corrective actions.
3.6.2.1.2.4 Cracks in High Ener7y Piping Through-wall leakage cracks are postulated in high-energy piping including branch runs larger than 1 inch nominal diameter as clarified below. The flow from this crack is evaluated for wetting, environmental and flooding effects.
A. ASME B and PV Code,Section III - Class 1 Piping Systems 2
- 1) Cracks are postulated at terminal end locations when the stresses exceed one half the limit given~in Section 3.6.2.1.1A.2.b(1).
- 2) Cracks are postulated at intermediate locations defined in Section l 3.6.2.1.1A.2b.(1). l l
WAPWR-S/E 3.6-16 AMENDMENT 2 f T999e:1d JANUARY, 1989 l I
)
B. ASME B and PV' Code'Section III - Class 2 and 3 Piping Systems O
- 1) Cracks are postulated at terminal end locations when the stresses exceed one half the limit given in Section 3.6.2.1.1.B.2. -]
I
- 2) Cracks are postulated at intermediate locations defined by. Section 3.6.2.1.1B.2. ~
1 C. Non nuclear Piping O 1) Cracks are postulated at terminal end locations when the stresses exceed one half the limit given in Section 3.6.2.1.1C.1.
- 2) Cracks are postulated at intermediate locations defined by Section 3.6.2.1.1C.1.
D. Containment Penetration Areas Cracks with break flow area of 1.0 square feet, are postulated 'in accordance with Section 3.6.2.1.10.
O O .
O WAPWR-S/E 1999e:1d 3.6-17 AMENDMENT 2 JANUARY, 1989
3.6.2.1.3 Break / Crack Configuration 3.6.2.1.3.1 High-Energy Break Configuration Following a circumferential break, the two ends of the broken pipe are assumed to move clear of each other unless physically limited by piping restraints, structural members, or piping stiffness. The effective cross-sectional (inside diameter) flow area of the pipe is used in the jet discharge evaluation. Movement is assumed to be in the direction of the jet reaction initially with the total path controlled by the piping geometry.
The orientation of a longitudinal break, exceot when otherwise justified by a detailed stress analysis, is assumed to be at opposing points on a line perpendicular to the plane of a fitting for a nonaxisymmetric fitting and anywhere around the circumference of the fitting for axisymmetric fittings.
The flow area of such a break is equal to the cross-sectional flow area of the pipe. Both circumferential and longitudinal breaks are postulated to occur, but not concurrently, in all high-energy piping systems at the locations specified in Subsection 3.6.2.1.1, except as follows:
- a. Circumferential breaks are not postulated in piping runs of 1 inch nominal diameter or less.
- b. Longitudinal breaks are not postulated in piping runs of a nominal diameter less than four inches.
- c. Longitudinal breaks are not postulated at intermediate locations in piping runs where the stress and cumulative usage factor limits for postulating intermediate rupture locations as specified in Subsection 3.6.2.1.1 for Class 1 piping and for Class 2 and 3 piping are not exceeded.
- d. Longitudinal breaks are not postulated at terminal ends.
O WAPWR-S/E 3.6-18 AMENDMENT 2 1999e:1d JANUARY, 1989
- e. Only one type of break is postulated at locations where, from a detailed stress analysis such as a finite element analysis, the state of stress can identify the most probable type. If the primary plus secondary stress in the axial direction is found to be at least 1.5 times that in the circumferential direction for the most severe loading combination associated with Level A and Level B service limits, then only a circumferential break is postulated. Conversely, if the primary plus secondary stress in the circumferential direction'is found to be at least 1.5 times that in the axial direction for the most severe loading e.mbination associated with Level A and Level B service limits, then only a longitudinal break is postulated,
- f. Where the postulated break location is at a tee or elbow, the locations and types of breaks are determined as follows:
- 1. Without the benefit of a detailed stress analysis, such as a finite element analysis, circumferential breaks are postulated to occur individually at each pipe-to-fitting weld, and longitudinal breaks postulated to occur individually' (except in piping with a nominal diameter less than four inches) on each side of the fitting at its center and oriented perpendicular to the plane of the fitting, or
- 2. Alternatively, if a detailed stress analysis or test is performed, the results may be used to predict the most probable rupture location (s) and type of break.
- g. Where the postulated break location is at a branch run connection, a circumferential break is postulated at the branch run pipe-to-fitting weld.
O
- h. Where the postulated break location is at a welded attachment (lugs, stanchions, etc.) a circumferential break is postulated at the centerline q of the welded attachment.
V l
WAPWR-S/E 3.6-19 AMENDMENT 2 1999e:1d JANUARY, 1989
- i. Where the postulated break location is at a reducer, circumferential breaks are postulated at each pipe-to-fitting weld. Longitudinal breaks are oriented to produce out-of plane bending of the piping configuration 1 on bo n sides of the reducer at each pipe-to-fitting weld.
2 3.6.2.1.3.2 Moderate-Energy and High-Energy Crack Configuration g i Moderate-enorgy and high-energy crack openings are assumed to be a circular orifice with cross-sectional flow area equal to that of a rectangle one-half the pipe inside diameter in length and one-half 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 Jet Thrust 2 To determine the forcing function for breaks identified by Subsection 3.6.2.1.1-A, B and C, the fluid conditions at the upstream source and at the break exit dictate the analytical approach and approximations that are used.
For most applications, one of the following' situations exists:
o Superheated or saturated steam.
o Saturated or subcooled water.
o Cold water (nonflashing).
Analytical methods for calculation of jet thrust for the above-described 2 l situations are based on ANSI /ANS-58.2-1980.
3.6.2.2.1.1 Tin Functions of Jet Thrust Force on Intact Reactor Coolant Loop (RCL) Piping To determine the thrust and reactive force loads to be applied to the RCL during the postulated loss-of-coolant accident (LOCA) in Subsection 3.6.2.1.1-A, .B and C, it is necessary to have a detailed description of the hydraulic transient. Hydraulic forcing functions are calculated for the j intact RCLs as a result of a postulated LOCA in branch runs connecting to the WAPWR-S/E 3.6-20 AMENDMENT 2 1999e:1d JANUARY 1989 )
)
l
primary RCL. These forces result from the transient flow and pressura histories in the reactor coolant system (RCS). The calculation is performed in two steps. The first step is to calculate the transient m essure, mass flowrates, and thermodynamic properties as a function of time. The second step uses the results obtained from the hydraulic analysis, along with input O
of areas and direction coordinates, and calculates the time-history of forces at appropriate locations (e.g., elbows) in the RCLs.
The hydraulic model represents the behavior of the coolant fluid within the entire RCS. Key parameters calculated by the hydraulic model are pressure, mass flowrate, and density. These are supplied to the thrust calculation, tocother 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 the 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 locp forces. This requires the ability to predict the flow, quality, and pressure of the fluid throughout the reactor system. The MULTIFLEX code, Reference 2, was developed with a capability to provide this information.
The MULTIFLEX computer 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 flexibility required to represent the various flow passages within the primary RCS. The system geometry is represented by a network of one-dimensional flow passages.
WAPWR-S/E 3.6-21 AMENDMENT 2 1999e:id JANUARY, 1989
l l
l The THRUST computer program was developed to compute the transient (blowdown) hydraulic forces resulting from a LOCA. The THRUST code calculates forces exactly the same way as the STHRUST code which is described in Reference 3.
The blowdown hydraulic loads on primary loop components are computed from the equation:
- 2
'F = 144A ((P - 14.7) + )
pg A2 x 144 The symbols and units are as follows:
F =
Force (ibf )
-A = Aperture area (ft2)
P = System pressure (psia)
~
ih =
Mass flowrate (Ibm /s)
- = Density (ibm /ft3 )
2 g = Gravitational constant = 32.174 ft-lbm/lb -s A
m
=
Mass flow area (ft )
In the model to compute forcing functions, the RCL 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:
A. Blowdown hydraulic information.
B. 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. !
O WAPWR-S/E 3.6-22 AMENDMENT 2 i 1999e:1d JANUARY, 1989
\
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 compo-nents using the projection coefficients. The force components are then summed over the total number of apertures in any one node to give a total x force, a total y force, and a total z force. These thrust forces serve as input to the piping / restraint dynamic analysis.
3.6.2.2.2 Dynamic Analysis of the Reactor Coolant Loop Piping and Equipment Supports The dynamic analysi, of the RCL for LOCA loadings is described in Section 3.9.
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 RCL The analytical methods of Reference 4 and ANSI /ANS-58.2-1980 are used to determine the jet impingement effects and loading effects applicable t 2 components and systems resulting from postulated pipe breaks. (1987 Draft of ANS 58.2 provides relaxed jet criteria.)
3.6.2.3.2 Dynamic Analysis Methods to Verity Integrity and Operability for the RCL 3.6.2.3.2.1 General The boundary limits for pipe ruptures which result in loss of reactor coolant are summarized in Figure 3.6-2. For normally closed isolation valves or incoming check valves (Cases I and IV, Figure 3.6-2), a LOCA is assumed t 2 occur for pipe breaks on the reactor side of the valve. For outgoing lines O 'with normally open automatic isolation valves (Case II, Figure 3.6-2), a LOCA is assumed to occur for pipe breaks on either side of the valves. For WAPWR-S/E 3.6-23 AMENDMENT 2 1999e:1d JANUARY, 1989
l normally open incoming check valves (Case III, Figure 3.6-2) a LOCA is assumed 2 to occur for pipe breaks on the reactor side of the second check valve, if either of the two valves close.
Branch lines connected to the RCL are defined as large strictly for the purpos,e of pipe break criteria if they have an inside diameter greater than 4 inches. 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 for the purpose of pipe break analysis 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 inches 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 100 guidelines. These safety systems are described in Subaction 6.2.1.2 of RESAR-SP/90 PDA Module 10 " Containment Systems."
To assure the continued integrity of the essential components and the engineered safety systems, consideration is given to the consequential effects of the pipe break itself to the extent that:
A. The minimum performance capabilities of the engineered safety systems are not reduced below that required to protect against the postulated break.
B. The containment leak tightness is not decreased below the design value if the break leads to a LOCA.(a)
O WAPWR-5/E 3.6-24 AMENDMENT 2 1999e:1d JANUARY, 1989
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. Ho' wever, a break which is not a LOCA is
- permitted to propagate to a single 0.375 inch diameter primary side line provided that line is not part of -a post accident monitoring system.
. 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 RCS Piping-
^
Large branch line piping, as defined in Subsection 3.6.2.3.2.1, is restrained to meet the following criteria in addition to items A through C of Subsection 3.6.2.3.2.1 for a pipe break resulting in a LOCA:
A. Propagation of the- break postulated in accordance with Subsection 3.6.2.1.1-A to the unaffected loops is prevented to ensure the delivery capacity of the accumulators and low head pumps.
B. Propagation of the break postulated in accordance with Subsection-3.6.2.1.1-A 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 is voluntarily applied so as not to increase substantially the severity of the LOCA.
O a. The containment is here defined as the containment structure and penetrations, the steam generator shell, the steam generator steam side instrumentation connections, the steam, feedwater, blowdown, and steam generator drain pipes within the containment structure.
O WAPWR-S/E 3.6-25 AMENDMENT 2 JANUARY, 1989 1999e:1d
i 3.6.2.3.2.3 Small Branch Lines j Should one of the small pressurized lines, as defined in Subsection 3.6.2.3.2.1, 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 Subsection 3.6.2.3.2.1: h 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. However, a break is permitted to propagate -to a single 0.375 inch diameter line attached to another leg of the affected loop provided that line is not part of a post accident monitoring system.
Damage to the high-head safety injection lines connected to the other leg of the affected loop or to the other loops is prevented.
B. Propagation of the break in the affected leg is permitted but is limited to a total break area of 12.5 square inches (4-inch 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. 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.
3.6.2.3.3 Types of Pipe Whip Restraints 1
3.6.2.3.3.1 Pipe Whip Restraints To satisfy varying requirements of available space, permissible pipe deflection, and equipment operability, the tsda in tt dre designed as a combination of an energy-absorMag element and a restraining structure suitable for the geometry required to pass the restraint load from the whipping pipe to the main building structure.
WAPWR-S/E 3.6-26 AMENDMENT 2 1999e:Id JANUARY, 1989 I
l l
The restraint structure is typically a structural steel frame or truss and -the energy-absorbing element is usually either. stainless steel U-bars or energy-absorbing material as described below:
A. Stainless Steel U-Bar This. type consists of one or more U-shaped, upset-threaded rods of stainless steel looped around the pipe but not in contact'with_the pipe to allow unimpeded pipe motion during seismic and thermal O\ movement of the. pipe. At rupture, the pipe moves against the U-bars, which absorb the kinetic energy of pipe motion by yielding elastically.
A typical example of a U-bar restraint is shown in Figure 3.6-3.
B. Energy Absorbing Material This type of _ restraint consists of a crushable, stainless steel, internally honeycomb-shaped element designed to _ yield elastically under impact of the whipping pipe. A design hot position gap is provided between the pipe and the energy-absorbing material to allow-unimpeded pipe motion _during seismic and thermal pipe movements. A typical. example of an energy-absorbing material restraint is shown in Figure 3.6-4.
L l- 3.6.2.3.4 Analytical Methods 3.6.2.3.4.1 Pipe Whip Restraints A. Location of Restraints
- 1. For purposes of determining pipe hinge length and thus locating' the pipe whip restraints, the plastic moment of the pipe is determined in the following manner:
M p
= 1.1 z pj S
O WAPWR-S/E 3.6-27 AMENDMENT 2 JANUARY, 1989 1999e:1d E _ _ _ _ ---_-_---_
l I
1 where: I z = Plastic section modulus of pipe p
= Yield stress at pipe operating temperature.
S" 1.1 = 10 percent factor to account for strain hardening.
Oi Pipe whip restraints are located as close to the axis of the reaction thrust force break as practicable. Pipe whip restraints are generally located so that a plastic hinge does not form in the pipe, if, due to physical limitations, pipe whip restraints are located so that a plastic hinge can form, the consequences of the whipping pipe and the jet impingement effect are further investi-gated. Lateral guides are provided where necessary to predict and control pipe motion.
- 2. Generally, restraints are designed and located with sufficient clearances between the pipe and the restraint such that they do ;
not interact and cause additional piping stresses. A design hot position gap is provided that will allow maximum predicted thermal, seismic, and seismic anchor movement displacements to occur without interaction.
Exceptions to this general criterion may occur when a pipe support and restraint are incorporated into the same structural steel frame, or when a zero design gap is required. In these cases the restraint is included in the piping analysis, if required.
- 3. In general, the restraints do not prevent the access required to conduct inservice inspection examination of piping welds. When the location of the restraint makes the piping welds inaccessible ,
for inservice inspection, a portion of the restraint is made removable to provide accessibility. l WAPWR-5/E 3.6-28 AMENDMENT 2 1999e:1d JANUARY, 1989
l B. Analysis and Design Analysis and design of pipe whip restraints for postulated pipe break I effectr are in accorjance with References 5 and 6. Specifically, the 2 following criteria are adopted in analysis and design:
- 1. Pipe whip restraints are designed based on energy absorption principles by considering the ela'stic plastic, strain-hardening behavior of the materials used.
- 2. A rebound factor of 1.1 is applied to the jet thrust force.
- 3. Except in cases where calculations are performed to verify that a plastic hinge is formed, the energy absorbed by the ruptured pipe
( is conservatively assumed to be zero; i.e., the thrust force developed goes directly into moving the broken pipe and is not reduced by the force required to bend the pipe.
- 4. In elastic plastic design, limits for strains are as follows:
c = Allowable strain used in design.
- a. Stainless Steel U-Bars e = 0.5c u
where:
O c u
= ultimate uniform strain of stainless steel (strain at ultimate stress).
- b. Energy-Absorbing Material
- = 0.8c u
O WAPWR-S/E 3.6-29 AMENDMENT 2 JANUARY, 1989
!999e:1d
where:
O c = maximum crushable height at uniform crushable u
strength.
- 5. A dynamic increa" factor is used for steel which is designed to )
remain elastic.
i 3.6.2.4 Guard Pipe Assembly Design Criteria Guard pipe assemblies are designed in accordance with ANSI /ANS 58.2-1980. The recommendation in SRP 3.6.2 to use the design pressure for the pressure test 2
is not followed. Instead, the containment pressure is used for the pressure test.
3.6.2.5 Material to be Submitted for the Operating License Review This will be provided in RESAR-SP/90 FDA version.
3.6.2.6 References O
- 1. American National Standard ASME Code for Pressure Piping, B31, Power 2
Piping, ANSI /ASME B31.1-1983.
- 2. "MULTIFLEX, A FORTRAN-IV Computer Program for Analyzing Thermal-Hydraulic-S'tructure System Dynamics," WCAP-8708 (Westinghouse Proprietary Class 2),
February 1976, and WCAP-8709 (Westinghouse nonproprietary), February 1976.
- 3. " Documentation of Selected Westinghouse Structural Analysis Computer Codes," WCAP-8252, Revision 1, (Westinghouse), May 1977.
- 4. Moody, F. J., Fluid Reaction and Imoincement Loads, Paper Presented at the ASCE Specialty Conference, Chicago, December 1973.
WAPWR-S/E 3.6-30 AMENDMENT 2 1999e:1d JANUARY, 1989 1
I _ _ . _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
I
- 5. '" Simplified . Pipe Analysis and . Restraint Design Procedures,."-
Whip WCAP-10221, December 1982.
- 6. ANSI /AISC N690-1984, Nuclear Facilities - Steel Safety-Related Structures 2
for Design, Fabrication and Erection.
O O
O .
'O O
l WAPWR-S/E 3.6-31 AMENDMENT 2 1999e:1d JANUARY, 1989
-. S TABLE 3.6-1 (SHEET 1 of'2)
' ESSENTIAL AND HIGH-ENERGY SYSTEMS Essential (*) Higb(b)
/m v System Systems Energy Reactor Coolant X X O
V Component cooling water X f
Emarger cy core cooling X Residual heat removal X Containment spray: X Chemical volume and control X X f
Nuclear sampling -
X Spent fuel cooling and - -
purification 1
Main steam X X O
O a. Not' all essential systems are required for all postulated piping failures. Also, not all portions of essential systems are required for a postulated piping failure.
- b. Not all portions of high-energy systems contain high energy fluid.
O WAPWR-S/E 3.6-32 AMENDMENT 2 JANUARY, 1989 1999e:1d i
TkBLE.3.6-1 (SHEET 2 of 2)'
ESSENTIAL AND HIGH ENERGY SYSTEMS Essential (*) High(b)
O System -Systems Energy Main feedwater X- X Emergency Feedwater X X Steam generator blowdown X X ,
Safety-related heating, X ventilating, and air conditioning Essential chilled water X Waste processing -
X Diesel generator and related X X systems Fire protection - -
Instrument and service air. - -
0 0 .
O WAPWR-S/E 3.6-33 AMENDMENT 2 JANUARY, 1989 1999e:1d
A b CASEI OUTG0 LNG UIES WTTH NORMALLY CLOSED VALVE
[ REACTOR COOLANT PIPING (
J U ) g O n V
BOUNDARf CASE ll 0UTGolNG UNES WITH NORMALLY OPEN VALVES O
) -
R
)
NOTE: THE REACTOR COOLANT PUWP
- 2 NO.1 SEAL IS ASSUWED TO BE FML CLOSED OR EOUlVALINT TO FIRST VALVE FAIL-AS-IS VRVES R A BOUNDARf U
CASE 111 INCOWING UNES NORWALLY WITH FLOW p
h NO.1 h 2 h BOUNDARY NO.2 p 4 TEST CONNECTION v
CASE N INCOMING UNES NORMALLY WITHOUT FLOW r
lr I 2
,l u ,i u _/- BOUNDARf g
A+V TEST CONNECTION (WEANS OF VERIFYING THAT CHECK VALVE IS CLOSED)
CASE V ALL INSTRUMENTATION TUBING AND INSTRUMENTS CONNECTED DIRECTLY TO THE REACTOR COOLANT SWTEM IS CONSIDERED AS A BOUNDARf. HOWEVER, A BREAK WITHIN THIS BOUNDARY O RESULTS IN A RELATNELY SMAll FLOW WHICH CAN NORMALLY V BE MADE US WITH THE CHARGING SYSTEM.
Figure 3.6-2. Loss of Reactor Coolant Accident Boundary Limits O WAPWR-S/E AMENDMENT 2 JANUARY, 1989 1999e:1d
l O Consistent with the Westinghouse positions in Reference 1, the NPB design uses the damping values of 4% and 2% of critical for the respective SSE and OBE 2 events for the primary coolant loop systems. For the remaining safety related piping systems, the f requency dependent damping values as established by the Pressure Vessel Research Council Technical Committee on Piping Systems and endorsed by Westinghouse in Reference 1 are considered. Figure 3.7-8 summarizes the damping values used for the piping design / analysis for the NPB.
O Also consistent with Regulatory Guide 1.61, damping values higher than those cited above may be used if justified by test results. Using the tests reported in Ref erences 2, 3 and 4, the OBE and SSE damping ratios of 7 percent and 10 percent, respectively, will be considered for the fuel assembly and a damping ratio of 5 percent will be specified conservatively for both the OBE and SSE seismic response analyses for the control rod drive mechanism (CROM).
3.7.1.4 Supporting Media for Seismic Category I Structures O Refer to Subsection 2A.S.3 of RESAR-SP/90 PDA Module 3, " Introduction and Site".
3.7.2 Seismic System Analysis This subsection describes the seismic analyses of the Category I structures, systems and components. Seismic systems are defined herein as the Category I structures, systems or component which, for analysis purposes, are considered in conjunction with foundation medium in forming a soil-structure or 9 foundation-structuro interaction model. All Category I structures, systems and components not designed as seismic systems and all Category I distributive systems such as heating, ventilation and air-conditioning systems, electrical cable trays, conduits, and piping are considered as seismic subsystems and O their analyses are described in Subsection 3.7.3 of this module. -
O WAPWR-S/E 3.7-3 AMENDMENT 2
. 1656e:1d JANUARY, 1989
i 3.7.2.1 Seismic Analysis Methods Seismic systems are analyzed by direct integration to determine the ef fects of ;
input ground motions on the Nuclear Power Block to obtain structural design loads for the seismic systems and to define seismic environment for the subsequent seismic analysis of structures, systems and components which are not supported directly on soil or foundation medium.
The analyses of the seismic systeins are performed on the baseline configura-tion of the Nuclear Power Block to resist the ground motions of 0.1 g and 0.3 g ZPA for the respective OBE and SSE as defined in Subsection 3.7.1. The potential variability of any site specific soil condition as defined in Section 2A-5 of RESAR-SP/90 PDA of Module 3, " Introduction ar.d Site" ir covered by this baseline design as a result of employing the envelope seismir requirements which are resulting f rom using bounding soil properties in the soil-structure interaction analyses. As discussed in Subsection 3.7.2.4, the envelope seismic requirements are derived by considering three analysis case.
characterized by the soil shear wave velocity of 1000 f t/sec, 2500 f t/sec and infinite each in the half-space impedance function modeling method of the soil-structure interaction models. By performing seismic analyses as described in the following paragraphs, the seismic performance of the baseline configuration is expected to be more than sufficient when the site specific data are incorporated into the final design verification by the impedance function method and the finite boundaries modeling methods to demonstrate the design adequacy as required by the Standard Review Plan (NUREG-0800, Rev.1, July 1981).
Natural Frequencies and Response Loads O
3.7.2.2 The seismic system analyses of the building structure of the NPB are performed using a time-history direct integration method. The floor response spectra generated from this method is indicative of the frequency content of the soil-structure system. Natural frequencies, mode shapes or modal responses are not obtained in this method as in the response spectrum analysis method.
O WAPWR-5/E 3.7-4 DECEMBER, 1984 T656e:1d
_______-a
O 3.7.2.9 Effects of Parameter Variations on Floor Response Spectra The uncertainties associated with the analytical models for deriving the i-in-structure response spectra are accounted for by broadening the spectrum f peaks' by 115% across the frequency band in accordance with Regulatory Guide 1.122. As described in Subsection 3.7.2.5, each' generic spectrum curve l
envelopes the broadened spectrum peaks for the three soil-structure analytical models. The resulting broad-band envelope floor response spectra are more than suf ficient to address the potential uncertainties in modeling the seismic systems.
3.7.2.10 Use of' Constant Vertical Static Factors No ' constant vertical static fcctors are used for Category I structures. The same method of analysis is used for both vertical and horizontal responses of the structures.
O 3.7.2.11 Method Used to Account for Torsional Effects The development of the generic floor response spectra by varying the underlying soil properties contains sufficient margin to account for the accidental torsional ef fects on a site specific configuration. The generic floor response spectra are judged to be suf ficient for analysis and design of subsystems anchored on floors.
3.7.2.12 Comparison of Responses O This section does not apply as response spectrum analysis will not be utilized for the systems of the NPB. Response spectra comparison, however, will be used to demonstrate the conservatism of the generic baseline design as applied O to site specific configuration.
j 3.7.2.13 Methods for Seismic Analysis of Dams Not applicable to the NPB scope.
3.7-9 DECEMBER, 1984 WAPWR-S/E 1656e:ld 1
1 l
I 3.7.2.14 Determination of Seismic Category I Structure Overturning Moment f i The Category 1 NPB structure is designed to resist overturning due to the com-bined ef fects of the vertical and two horizontal components of seismic ground t motion. The moment equilibrium method is considered in which the maximum seismic overturning moment is obtained from the analyses described in Subsection 3.7.2.2. The gravity force reduced by the hydrostatic buoyance force provides stability of the structure in resisting overturning moment.
The minimum safety factor against overturning moment is 1.1 for an SSE combined with the other applicable loading conditions.
3.7.2.15 Analysis Procedure for Damping The damping property of the NPB soil-structure system is af fected by the type of soil medium and the details of the structural concrete and steel used for constructing the plant. For the building-soil analysis stick model shown in Figure 3.7-9, the structural damping of the concrete 'and steel is represented using Rayleigh type viscous damping. For this portion of the model, the damping is a linear combination of system mass and stiffness damping.
Mathematically, this can be expressed as:
[C] = A[M] + B[K]
where "A" and "B" are scalar quantities equivalent to the system mass and stiffness damping coefficients, respectively. [C], [M], and [K] are system damping, mass, and stiffness matrices, respectively.
For the soil medium, damping and stif fness are modeled using discrete damper and spring elements in accordance with the impedance function approach of Reference 5. The resulting nonuniformly damped soil-structure model is analyzed using the direct integration routine of the WECAN computer program (Ref erence 13).
3.7.3 Seismic Subsystem Analysis This section describes the seismic analysis performed on subsystems.
WAPWR-S/E 3.7-10 AMENDMENT 2 1656e:Id JANUARY, 1989
i O 3.7.3.1 Seismic Analysis Methods Both the time-history solution and the response spectrum analysis technique O
V are used for analyzing the subsystems of the NPB. In general, analyses follow the ASCE Seismic Analysis Standard Committee approaches of Reference 5,
" Standard for the Seismic Analysis of Safety-Related Nuclear Structures."
The generic floor response spectra of Subsection 3.7.2.5 serve as design input
'- for the subsystems.
When the time-history solution is considered, synthesized time histories of 10-second total duration are generated for each of the three components - two horizontal and one vertical, of the floor response spectra. The Westinghouse program DEBLIN2, Reference 6, is utilized to synthesize the spectrum-compatible time histories. The program modifies earthquake motions by a frequency suppressing and raising technique in an iterative scheme to assure that the response spectra of the resulting time histories will properly O envelope the corresponding floor response spectra. Statistical independence among the time history components is assurcd by requiring the cross correlation coefficients among different inputs to be less than 0.3 (References 5 and 7). The resulting three components of the acceleration time l 2 histories will be simultaneously input to subsystems for either a direct-integration or a modal superposition time-history solution.
For subsystems modeled with linear elastic response, the response spectrum analysis of Ref erence 5 is performed. The generic floor response spectra are applied to subsystems with consideration of the three components of earthquake motion as per Subsection 3.7.3.6 and the combination of modal responses as per Subsection 3.7.3.7.
3.7.3.2 Determination of Number of Earthquake Cycles component will For each OBE the system and have a maximum response corresponding to the maximum induced stresses. The effect of these maximum stresses for the total number of OBE's must be evaluated to assure resistance to cyclic loading.
WAPWR-S/E 3.7-11 AMENDMENT 2 1656e:ld JANUARY, 1989
_ - - . - _ _ - _ - - ._-___________._-._--__.-.-_-_._--a
The OBE is conservatively assumed to occur five 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, duration and frequency contents of the input earthquake. A precise determination of the maximum number of 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 muximum stress O
associated with each occurrence, an evaluation was performed considering both equipment and its supporting building structure as single degree-of-f reedom systems. The natural frequencies of the building and the equipment are conservatively chosen to coincide.
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 eight cycles. If the equipment was assumed to be rigid in a flexible building, the number of cycles exceeding 90% of the maximum stress was not greater than three cycles.
This study was conservative since it was performed with single degree-of-f reedom models which tends to produce a more uniform and unattenuated response that a complex interacted 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 f requencies greater than 33 Hz) for each of 5 OBE occurrences should be used for fatigue evaluation of WAPWR systems and components. Ten magimum stress cycles are 2
used for all ASME Class 1 systems and components.
3.7.3.3 Procedure Used for the Modeling A. Modeling of Piping Systems for Dynamic Analysis The piping systems are modeled utilizing a three-dimensional structural representation composed of concentrated lumped masses connected by WAPWR-S/E 3.7-12 AMENDMENT T 1656e:ld JANUARY, 1989
Q v appropriate piping system. The model accounts for the interaction ef fect between piping, equipment and supports. Supports are modeled as flexible' members with the appropriate stiffness to represent the support compliance. The piping model is terminated at equipment nozzles which are modeled as rigid anchors with consideration given to the seismic amplification of equipment, as follows:
/' 1. For rigid equipment in which the fundamental f requency is equal - or higher than 33 Hz, the amplified response spectra of the structure is used.
2 .. For equipment in which the fundamental f requency is lower than 33 Hz, the amplified response spectra and the seismic anchor displacement of the equipment at the pipe / nozzle interface point is used.
Alternatively, a simplified model of the equipment to account for dynamic interaction and amplification is coupled with the piping n model, and the amplified structure response spectra are used to excite U the coupled model.
All in-line components are included in the model. The concentrated mass of in-line components such as valves, flanges, and strainers are represented as lumped masses. Valve operators are modeled as an of f set lumped mass to account for the torsional and in-plane bending ef fects on the piping.
The following criteria are used for the decoupling of piping subsystems:
O 1. When piping is decoupled f rom the equipment, the nozzle is modeled as I a full, six-degree-of-freedom restraint.
- 2. For the analysis of main runs, branch connections are decoupled f rom the main runs when the ratio of the branch to run moment of inertia is 2 1/25.
- 3. Piping subsystems (main or branch runs) which are decoupled into separate analytical models satisfy one of the following criteria- i WAPWR-S/E 3.7-13 AMENDMENT 2 1656e:ld J ANUARY , 1989
_ _ _ _ _ l
(a) The decoupling point is a full anchor for the piping of both separate models.
(b) The boundary of each decoupled model contains a suf ficiently long 2
region of overlap to other models which effectively provides restraint (s) in each of the three orthogonal directions in order to justify decoupling.
B. Modeling of Equipment Seismic Category I equipment is modeled as lumped systems which consist of a series of discrete mass points connected by elastic members. All significant concentrated weights are represented as lumped masses.
Typical examples of concentrated weights are weights of motor rotor and pump impeller in the analysis of shafts. The number of dynamic degrees of f reedom is at least twice the number of modes having f requencies less than 33 Hz.
Basis for Selection of Frequencies O
3.7.3.4 There are no specific design criteria that attempt to cause the fundamental f requencies of NPB equipment to be dif ferent f rom the forcing f requencies of the supporting structures. The effect of the equipment fundamental frequen-cies relative to the supporting structure forcing frequencies is, however, considered in the analysis of the NPB equipment.
3.7.3.5 Use of Equivalent Static Load Method of Analysis The static load equivalent or static analysis method involves the multiplica-tion of the total weight of the equipment or component member by the specified seismic acceleration coef ficient. The magnitude of the seismic acceleration coefficient is established on the basis of the expected dynamic response characteristics of the component. Components that can be adequately charac-l terized as single-degree-of-freedom systems are considered to have a modal 1
O l WAPWR-S/E 3.7-14 AMENDMENT 2 l
1656e:1d JANUARY, 1989 l
l
O participation factor of 1. Seismic acceleration coef ficients for multidegree-of-freedom systems -may be determined :as 1.5 times the peak- spectral acceleration of the applicable response spectrum. Smaller values may be used,
( if justified.
3.7.3.6 Three Components of Earthquake Motion Seismic Category I subsystems and components are analyzed by considering the combined effects of seismi'c loads occurring in three mutually perpendicular directions, two in 'the horizontal direction and one in the' vertical direc-tion. The total combined response (displacer.cnts, stresses, and forces) due to the three components 'of earthquake motion is obtained by using the
, square-root-sum-of-the-squares (SRSS) formula applied to the resultant codirectional responses. For instance, for each item of interest, such as displacement, force, stress, etc., the total response is obtained by applying the SRSS method. The
- thematical expression for this method (with R as the item of. interest) is:
3 1/2 U}
R C" T=1 T) where:
= total combined response at a point.
~
R C
R = value of combined response of direction T.
T f~
The system and equipment response can also be determined using time-history analyses. When a time-history analysis is performed, the two horizontal and the vertical time-history components are applied simultaneously, f
3.7.3.7 Combination of Modal Responses The total codirectional seismic response is obtained by combining the indi-vidual modal responses utilizing the SRSS method. An optional method is the O algebraic combination of modes with closely spaced f requencies (Reference 8). 2 l 3.7-15 AMENDMENT 2 WAPWR-S/E T656e:ld JANUARY, 1989 l
The groups of closely spaced modes are chosen such that the dif ference between the f requency of the first mode and the last mode in the group does not er.ceed 10 percent of the lower f requency. Groups are formed starting from the lowest frequency and working toward successively higher frequencies. No one <
frequency is in more than one group. The combined total response is obtained as follows:
N S "j" "j R2=I R2 +'2 I I I RKgR 'Kt (2)
T i=1 j=1 K=M) t=K+1 where:
R = total codirectional response.
T R = response of mode i.
N = total number of lower frequency, flexible modes.
S = number of groups of closely spaced modes.
= lowest modal number associated with group j of closely spaced M) modes.
N) = highest modal number associated with group j of closely spaced j
modes. 1 factor with j c
g = coupli .1
,K
'_ ' -1 c gg = {l + ( , )2} and (3)
(B wgg+Sw) gg w =w g (1 - (Bg ) ) (4)
Ol O
K "0K+wg t d
O WAPWR-5/E 3.7-16 AMENDMENT 2 T656e:1d JANUARY, 1989 l
1 - - - - - - _ _ _ -
hG: ;n where:
w = frequency of mode K.
0 = fraction of critical damping in mode K.
1 t = duration of the earthquake. .
d The options used to account for high--f requency (>33.0 HZ) modes are described
_q() below:
A. The Residual Load Method (RLM) with Uniform Response Spectrum Analysis is based on the following' equations (Reference 9): 2
{Xc }= -[K] (6)
[M]([J]-[*d3 E*d] [M] [J]) {Xg}
(7)
{Xc }= ([J] - [$d3 E*d] [M] [J]) {Xg }
where
[J] = influence matrix
[K] = stiffness matrix
[M] = mass matrix 1 {X } = residual displacement vector from truncated higher modes '
{X } = residual acceleration vector
{X } = ground acceleration vector
[$ ] = flexible mode shape matrix d
O The combination of shock directions for these truncated higher modes is obtained from equation (1). The total response from flexible and truncated higher modes is given by:
R CTOTAL=(Rfp+RfT) (8)
L where R are the combined flexible and truncated mode responses f rom l
0, CF' equation (1).
N CT WAPWR-S/E 3.7-17 AMENDMENT 2 1656e:ld JANUARY, 1989 1
B. The full zero period acceleration method (FZPA) with Uniform Response 2l Spectrum Analysis is based on the following equations (Reference 9):
{Xc } = -[K] [M] [J] {Xg} (9)
{Xc } " E 3 I g} ( 0) 1 The combination of shock directions for the FZPA response is obtained f rom equation (1). The total response from flexible modes and the FZPA response is obtained by SRSS combination similar to equation (8).
C. The RLM with Multiple Response Spectrum Analysis is based on the following 2 equations (Reference 10):
( I}
~
{Xc }= -[K] l [M]([Y]-[$d3 E*d3 E"3E 3) Ig I (12)
{Xc }= ([Y] - [$d l E*d] [M] [Y]) {Xg }
where
[Y] = -[K]~ [Kg ] (13) i and Kg is system-support coupling stiffness matrix, The combination of shock directions along with flexible and truncated modes is performed in the same manner as in item A above. (See subsection 3.7.3.9 for f urther details).
D. The FZPA method with Multiple Response Spectra Analysis is based on the 2l following equations (Ref erence 10):
"~
O (X g c
{X c
"E I gI WAPWR-S/E 3.7-18 AMENDMENT 2 T656e:ld JANUARY, 1989
i
. f%
d The combination of shock directions along with flexible and truncated
- modes is performed in the same manner as in item B above. (See Subsection 3.7.3.9 for f urther details.)
(O u 3.7.3.8 Analytical Procedures for Piping The Seismic Category I piping systems are analyzed and evaluated according to the rules of the American Society of Mechanical Engineers (ASME) code, Section U
III. When modal Seismic response spectrum analysis methods are used to evaluate piping seismic response due to inertial loading arising from excitations at different supports within one or more buildings, the procedures described in Section 3.7.3.9 are used. The effect of' dif ferential seismic anchor motions at dif fcrent supports are included in the pipirq analysis according to the rules of the ASME code,Section III. The piping stresses due
. to seismic anchor motions are combined with stresses from other applicable loads including seismic inertial loading and then evaluated as required by the ASME code, Section III. For analf sis of seismic anchor motions, the procedures described in Section 3.7.3.9 are used.
3.7.3.9 Multiple Supported Equipment Components With Distinct Inputs A. To evaluate piping and equipment components seismic response due to inertial loading arising f rom excitations at dif ferent supports within one or more buildings, either modal envelope seismi,c response spectra analysis method or modal non-uniform seismic response spectra analysis method is used.
.O A.1 The modal envelope seismic response spectra analysis method is the same as the. standard model seismic- response spectra analysis method for a singly- or multiply-supported system subject to uniform V transnational seismic excitation except that it utilizes, for each direction of excitation the single envelope spectrum or the worst single spectrum. The single envelope or worst spectrum is assumed by this procedure to account for the influence of phasing and interdependence characteristics of non-uniform excitation represented by transnational spectra at various supports.
WAPWR-5/E 3.7-19 AMENDMENT 2 1656e:ld JANUARY,1989
I 1
A.2 In the modal non-uniform seismic response spectra analysis method (References 10,11 and 12), for each direction of excitation, multiple i 2l input response spectra representing the non-uniform seismic excitation at all support (boundary) points of the structural system (model) are explicitly used without being approximated, consolidated or replaced 1
as in the case of the modal superposition envelope response spectra method. Further, for each direction of excitation, the phasing and interdependence characteristics of the multiple input response spectra representing non-uniform seismic excitation are identified and properly accounted for by this method as outlined below.
Proportional Input - For this type of ' input, the support motion at a given point can be obtained simply through multiplication of a reference excitation by a real number. This, therefore, includes the uniform excitation as a special case. Support motions that are 180*
out-of-phase are also included here since they can be obtained through multiplication of a reference excitation by a negative real number.
Support point motions associated with a single-mode response of a supporting structure or with a rigid supporting structure are examples of this type of input. For this type of input, the representative maximum modal response is obtained by algebraic combination of contributions of individual support point inputs.
Independent Input - For this typq of input, the support motions are treated to be statistically independent and are therefore essentially uncorrelated. Support point motions associated with supporting structures of widely differing dynamic characteristics can be considered as practical examples of this type of input. For this type of input, the representative maximum modal response is obtained by the square-root-sum-of-squares (SRSS) combination of contributions of individual support point inputs.
Mixed Input - This type of non-uniform excitation consists of a combination, of the two types described above. For this type of input, O
WAPWR-5/E 3.7-20 AMEN 0 MENT 2 1656e:ld JANUARY, 1989
)
l l
l-1 O the representative maximum modal response is obtained by the SRSS, combination of contributions of the representative maximum modal responses obtained for each of the two types described above.
v,O Af ter maximum possible use of algebraic and SRSS combinations, as described above, absolute sum combination is used only .as a last resort in absence of another more realistic combination..
O V B. The response due to differential seismic . anchor motions is calculated using static analysis (without including dynamic load f actor). In this analysis, the static model is identical to the static portion of the dynamic model used to compute the seismic response due to inertial loading. In particular, the structural system supports in the static model are identical to those in the dynamic model. The effect of relative seismic anchor displacements are obtained either by using the worst combination of the peak displacements or by proper representation of the relative phasing characteristics associated with different support inputs.
C. The results of modal seismic spectra analysis in Item A above and the results of seismic anchor motion analysis in Item B above are combined by the SRSS when required by consideration for the ASME classification of the stresses.
3.7.3.10 Use of Constant Vertical Static Factors Constant vertical load f actors are not used as the vertical floor response Q load for tne seismic design of safety-related components and equipment within the Westinghouse scope of responsibility.
3.7.3.11 Torsional Effects of Eccentric Masses O .
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 ef fects caused by them are evaluated 3.7-21 AMENDMENT 2 WAPWR-5/E J ANUARY, 1989 1656e:U l
and' included in the total system response. The total response must meet the limits of the criteria applicable to the safety class of the piping.
3.7.3.12 Buried Seismic Category I Piping Systems and Tunnels O
There are no buried seismic Category I piping systems and tunnels in the NPB.
3.7.3.13 Interaction of Other Piping With Seismic Category I Piping Where seismic Category I piping systems are in close proximity to non-seismic piping, the non-seismic pipes are restrained so that no f ailure of the seismic Category I system can occur. For large diameter non-seismic pipes (greater than 2" nominal pipe size) this is accomplished by performing an SSE seismic analysis for the non-seismic pipes in accordance with ASME Code requirements.
2 For the small diameter non-seismic pipes (nominal pipe size 2" and smaller) this is accomplished by establishing limits on support and piping displacements. These limits assure that no f ailure of the adjacent seismic piping can occur for the SSE event.
Where seismic Category I piping is directly connected to non-seismic Category I piping, the seismic ef fects of the latter are prevented from being transferred to the seismic Category I piping by use of anchors or a combination of restraints; or when this is not practical, the interactive ef f ects of the unrestrained portion of the non-seismic Category I piping are included in the analyses, and evaluated for acceptability.
i
(
3.7.3.14 Seismic Analyses for Reactor Internals (Core, Core Supports, Mechanisms) {
l Fuel assembly, core support structure, and control mechanism component stresses induced by seismic disturbances, are analyzed by finite element computer techniques. The time-history response of the building is used to j generate the input to the system model of the above components. These components are modeled as spring and lumped mass systems or beam elements.
l O
WAPWR-S/E 3.7-22 AMENDMENT 2 1656e:ld JANUARY, 1989 l
1 l
7 NJ The component scismic response of the fuel assemblies is analyzed to determine design adequacy. The response of the core structures and mechanisms is used in the ASME B&PV code evaluations. Fuel assembly damping and grid strength (o) capability are determined experimentally.
v The mechanisms, both the control rod drive mechanism (CRDMs) and the displacer rod drive mechanisms (DRDMs), are seismically analyzed to confirm that stresses under the combined loading conditions, do not exceed allowable levels
"' as defined by the ASME Code, subsection III, for condition B and condition D events. The mechanisms are modeled as a system of lumped and distributed I masses, and the resultant seismic bending moments and shear loadings along the length of the mechanisms are calculated. The corresponding stresses are then combined with the stresses f rom other loadings and the combination is shown to meet the requirements of the ASME Code,Section III.
3.7.3.15 Analysis Procedure for Damping
() Where the equipment or component consists of subcomponents with the same damping characteristics, the same critical damping value is used for the entire equipment or component. The corresponding critical damping value is chosen from Table 3.7-1. For seismic Category I equipment or component 2 consisting of subcomponents with different damping characteristics, two approaches are considered: 1) the lowest critical damping value associa- ted with the subcomponents in the equipment or component is used in the analysis for all modes, 2) the composite damping values or nonproportional damping models as proposed by the ASCE Seismic Analysis Standard Committee of
) Reference 5 are used. For piping systems, the damping values f roc Figure 2 3.7-8 may also be used, as discussed in Section 3.9.3.
3.7.4 Seismic Instrumentation n
Seismic instrumentation is provided to the NPB to gather information on the input ground motion and the output vibratory responses of the representative Category I structures and equipment so that an evaluation can be made as to:
(j o Whether input design response spectra were exceeded, WAPWR-S/E 3.7-23 AMENDMENT 2 1656e:1d JANUARY, 1989
O, o Whether the vibratory responses of the representative Category I structures and equipment were exceeded, o The need for shutdown of the plant, and o The degree of conservatism of the mathematical models used in the seismic analysis of the building and equipment.
The design (. consideration of the seismic instrumentation is based on a Safe O
Shutdown Earthquake (SSE) of 0.3g ZPA and an Operating Basis Earthquake (OBE) of 0.19 ZPA.
3.7.4.1 Comparison With Regulatory Guide 1.12 The seismic instrumentation described below consists of time-history accelerographs, seismic switches, response spectrum recorders and peak accelerographs meeting the USNRC Regulatory Guide 1.12, Revision 1 (April 1974), as required for a severe earthquake with ZPA of 0.3g or higher.
3.7.4.2 Location and Description of Instrument The instrumentation described below is employed to measure and record the seismic inputs and the plant structural and equipment responses and to provide displays and alarms to operators to act and engineers to evaluate the plant seismic capability after an earthquake.
When external power supply is needed for operating the instrument during and af ter earthquakes, the Class lE 120 V uninterruptable power supply will be provided.
3.7.4.2.1 Time-History Accelerograph 1
Three triaxial time-history accelerographs will be provided, one each at the l
following locations:
WAPWR-S/E 3.7-24 AMENDMENT 2 1656e:ld J ANU ARY, 1989
O 1. A free field at approximately 30 ft from the edge of the reactor external i
building
- 2. The top of the foundation base mat (Reference Elevation = 72m) and inside the reactor external building l
- 3. The concrete operating floor (Reference Elevation = 100m) and inside the l D
i reactor containment.
A fourth time-history accelerograph will be installed in the main control room concrete floor if the design site ground motion is 0.3g ZPA or higher.
.Each time-history accelerograph package consists of a triaxial sensor with triaxial starter unit and a recorder unit. The triaxial sensor unit is responsive in the .f requency range of 0.1 Hz to 30 Hz in the three orthogonal
. axes. The starter unit also has corresponding acceleration sensors set to energize the triaxial sensor unit whenever the threshold acceleration is exceeded in any of the three orthogonal axes. The threshold accelerations are set between 0.005g and 0.02g, depending on locations, to avoid actuation due to insignificant motion, but to record a seismic disturbance which may have a ground acceleration magnitude significantly lower than that of the Operating Basis Earthquake of 0.1 g ZPA.
The recorder units and a common playback unit will be housed in a control panel which in turn will be located in the main control room (Subsection 3.7.4.2.6). The three starter units installed in the main control room, the
(./ operating floor and the basemat of the reactor containment will be oriented such that their axes and the axes of the sensor units are pointing in the same direction and aligned to the principal axes of the reactor external building.
[ The time-history accelerograph is fully operational within 0.1 second of seismic starter actuation. Once actuated, an amber light, one for each accelerograph package, remains on in the control room. The accelerograph will operate continuously during that period in which the acceleration exceeds the O starter threshold plus at least five (5) seconds.
WAPWR-S/E 3.7-25 AMEN 0 MENT 2 T656e:ld JANUARY, 1989
The recorder unit will be capable of a minimum 25 minutes total recording O
time. The common playback system allows immediate graphical time-history accelerogram playback capability.
The starter unit and seismic switch as described in the next paragraph can be O
tested from the main control room.
3.7.4.2.2 Seismic Switch One seismic switch will be located at the top of the foundation mat inside the reactor external building and in the general vicinity where the time-history accelerograph of Subsection 3.7.4.2.1 is installed. If the design site specific ground motion is 0.3g ZPA or higher, a second seismic switch will be installed on the Class 1 piping connected to reactor coolant loop. The seismic switch will be responsive to f requencies f rom 0.1 Hz to 30 Hz. The switch on the basemat will be set at 0.1g corresponding to an OBE. The seismic switch is a triaxial low f requency acceleration sensor with adjustable threshold accelerations in three orthogonal directions. It operates with an internal rechargeable power supply. The minimum duration of the switch actuation is adjustable (6-20 seconds), and remains actuated as long as the setpoint is exceeded. Audio alarm will result once the seismic switch is actuated.
3.7.4.2.3 Triaxial Spectrum Recorder The triaxial response spectrum recorder provides a permanent record of spectral accelerations at 12 discrete frequencies on all three axes. The recorded values in the main control room provides a basis to see whether the spectral acceleration levels at individual discrete f requencies are within, or above, the OBE response spectrum levels.
Tne response spectrum recorders will be responsive to a frequency range of 1 Hz to 30 Hz with appropriate damping value to facilitate comparison of spectral acceleration values associated with the OBE response spectra. They will be employed to provide more information on the seismic input and the WAPWR-5/E 3.7-26 AMENDMENT 2 1656e:1d JANUARY, 1989
l i- j O potential plant seismic response property with no need to wait for detailed l processing of the time-history accelerograph records.
A total of four response spectrum recorders are provided, one each at the following locations: j
- 1. The top of the foundation base mat inside the reactor external building -
and near the vicinity of the time-history accelerograph' of Subsection 3.7.4.2.1.
- 2. The class 1 piping connected to reactor coolant loop.
- 3. The concrete operating floor of the main control room.
- 4. The support to a Category I piping system.
f In case the desig'n site specific ground motion is 0.39 ZPA or higher, a fif th-Y response spectrum recorder will be installed on the supporting pad of a Category I equipment structure.
3.7.4.2.4 Triaxial Peak Accelerograph The peak accelerograph is a self contained passive device capable of permanently recording peak acceleration. It detects peak acceleration in a f requency range f rom 0.1 Hz to 20 Hz. Data f rom the peak acceleregraph will be manually retrieved following an earthquake and will be used in the detailed O evaluation of seismic performance of the plant structures, systems and components.
A total of three triaxial peak accelerographs are provided, one each at the following locations:
- 1. The reactor coolant pump motor
- 2. The Class 1 piping connected to reactor coolant loop.
- 3. The Category I piping outside the containment.
WAPWR-S/E 3.7-27 AMEN 0 MENT 2 1656e:ld JANUARY, 1989
}
In case the design site specific ground motion is 0.3g ZPA or higher, a fourth O
peak accelerograph will be installed on the supporting pad of a Category I equipment structure.
3.7.4.2.5 Criteria for Instrument Location The selection of the above locations for installing seismic instrument is based on the guidance provided in the USNRC Regulatory Guide 1.12, Revision 1, for an SSE acceleration of 0.3g cr higher, unless as noted.
All instruments are accessible for inspection, test and service except for the instruments on the reactor coolant pump motor and the reactor coolant Class 1 piping which are accessible only during reactor shutdown.
Table 3.7-2 summarizes the locations of the seismic instruments. >
3.7.4.2.6 Seismic Instrumentation Control Panel An instrumentation panel located in the main control room will be provided to house the recording, playback and calibration units which are used in conjunction with the time-history accelerographs. It also contains the audio alarms and visual displays in association with the operation of the seismic switches and the response spectrum recorders.
3.7.4.3 Control Room Operator Notifications Operator notification consists of alarms, indicating lights and graphical displays.
Audio and visual alarms will be provided in the main control room f or the following parameters:
o Containment foundation ZPA input in excess of 0.1g (OBE) o Actuation of any time-history accelerographs WAPWR-5/E 3.7-28 AMENDMENT 2 1656e:ld JANUARY, 1989
l i
O V
o Response spectral values in any frequency and any axis in excess of design OBE spectral accelerations as recorded by the response spectrum recorders. j The time-history accelerograph records will be played back to provide visual displays as needed after earthquakes. 1 3.7.4.4 Comparison of Measured and Predicted Responses r
The plan for utilization of the seismic data includes both the function of the operator and engineering to evaluate the effects of an earthquake on the plant. For a detail description of the data flow, refer to Figure 3.7-20.
Initial determination of the earthquake ef fect is performed immediately af ter the earthquake by comparing the measured response spectra from the containment base mat with the OBE and SSE design response spectra for the corresponding location. .
If the measured spectra exceed the OBE response spectra, the plant will be shutdown and a detailed analysis of the earthquake motion will be undertaken.
After an earthquake, the data f rom the seismic recording instruments are reviewed. See Figure 3.7-20. The data from these instruments will be analyzed to obtain the seismic accelerations experienced at the location of major Category I structures and equipment. The measured responses f rom the instruments will t;e used to evaluate seismic Category I structures and systems in which the spectra are compared with those used in the design to determine O whether the OBE design level has been exceeded or not.
During shutdown as a result of OBE earthquake, the equipment mounted triaxial peak accelerographs will be used to determine if the design limitation of specific equipment to which it is fastened has been exceeded. If the measured responses are less than the values used in the design and qualification of the Seismic Category I structures, systems, and equipment and a visual inspection of the systems and components reveals no damage, the structure, system, or WAPWR-S/E 3.7-29 'sMENDMENT 2 1656e:1d Ji JARY, 1989
l equipment is considered adequate f or future operation. Otherwise, damage is corrected, and a new analysis is made to assure the adequacy of those items for future use.
3.7.4.5 Inservice Surveillant.e Calibration and alignment on three orthogonal axes will be performed prior to fuel loading in crder to assure proper operation. Periodic testing and calibration will be performed in accordance with technical specification.
3.7.5 References
- 1. "Use of Frequency Dependent Damping Values," Westinghouse letter NS-EPR-2955, Rahe, E.P. to J.P. Knight, dated August 16, 1984.
- 2. Gesinki, T.L., and Chiang, D., " Safety Analysis of the 17x17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident," WCAP-8236 (Proprietary), December 1973, and WCAP-8288 (Nonproprietary), January 1974.
- 3. Obermeyer, F.D., " Effective Structural Damping of the KEP L105 CRDM,"
WCAp-7427 (Proprietary), January 1970.
- 4. Obermeyer, F.D., WCAP-7427, Addendum 1 (Proprietary) December 1970.
- 5. " Seismic Analysis of Safety-Related Nuclear Structures and Commentary on 2 Standard for Seismic Analysis of Safety Related Nuclear Structures," ASCE Standard Number ASCE 4-86, September,1986.
- 6. Lin, C.W., "DEBLIN2 -
A Computer Code to Synthesize Earthquake Acceleration Time Histories," WCAP-8867, November,1976.
- 7. Lin, C.W., " Time History Input Development for the Seismic Analysis of Piping Systems," Journal of Pressure Vessel Technology, Vol.102, No. 2, May 1980.
O o WAPWR-5/E 3.7-30 AMENDMENT 2 1656e:ld JANUARY, 1989
- 8. " Evaluation of Other Dynamic Loads and Load Combinations," Report of thel 2 U.S. Nuclear Regulatory Commission Piping Review Committee, NUREG-1061, Vol. 4, December 1984.
- 9. Park, I.B., and Johnson, E.R., "Computationally Efficient Methods ~ forl 2 Seismic Response f rom ' Flexible and Rigid Modes," Seismic and Dynamic Analysis Methods (C.W. Lin, ed.), ASME-BVP-81, New York,1984.
- 10. Vashi, K.M., " Seismic Spectral Analysis for Structures Subjected Tol2 l Non-Uniform Excitation," ASME Paper No., 83-PVP-69. )
- 11. Lin, C.W., and tocef f, F., "A New Approach to Compute System Response with l 2 Multiple Support Response Spectra Input," Nuclear Engineering and Design 1 60, North-Holland Publishing Company,1980, pp. 347-352.
- 12. Lin, C.W., and Guilinger, W.H., " System Response to Multiple Support l 2 Response Spectra Input," Proceedings of Sixth International SMIRT O Conf erence, Paper K10/4, Paris,1981.
- 13. " Benchmark Problem Solutions Employed for Verification of the WECAN 2 Computer Program," WCAP-8929 (Westinghouse Non-Proprietary), April,1977.
O O
O WAPWR-5/E 3.7-31 AMENDMENT 2 1656e:1d JANUARY, 1989 i
TABLE 3.7-1
]
REGULATORY GUIDE 1.61 DAMPING val.UES FOR STRUCTURES OR COMPONENTS (a)
O Percent of Critical Damping Per Mode Structure of Component OBE SSE Welded steel structures 2 4 Bolted steel structures 4 7 i l
Prestressed concrete structures 2 5 l
1 Reinforced concrete structures 4 7 !
O\
l l
l l
O 1 1
- a. Damping values for foundation material, used in foundation-structure '
interaction analysis, are not included in this table.
O1 WAPWR-S/E 3.7-32 DECEMBER, 1984 l 1656e:1d L-_____-______-
e i O
~
10 '
Percent of O Critical Domping
- (%)
9 s
O '
6 J
5 q .
3 2
2 1
i i 10 20 Frequency (H2)
O O
Figure 3.7-8 Alternative Damping Values For Piping Systems O
WAPWR-S/E AMENDMENT 2 JANUARY, 1989 T656e:ld i
i e
il e /g4 APPENDIX 3.7 i
'% J j i
j COMPUTER CODE DEBLIN2 )
.f i
In a time history analysis of systems and components where the input is ;
defined either in the form of design ground response spectrum or design floor j response spectrumm, it is necessary to construct a synthesized time history.
motion whose response spectrum essentially envelopes the input response !
spectrum. The generation of such a motion can be accomplished by the repeated !
modification of an actual earthquake motion using the spectrum suppression and spectrum raising techniques (Ref. 1). The first of these techniques is used when the motion's response spectrum lies above the design response spectrum more than a specified margin. The second technique is used when the motion's response spectrum lies below the design response spectrum. The computer code DEBLIN2 was developed using these techniques.
The design response spectrum may be the ground response spectrum contained in
- s. - the Safety Analysis Report for a specific damping value, or a floor response 2 spectrum generated for the purpose of analyzing systems and components supported by a concrete structure. The former is required for the plant analysis using input at the ground level, and the latter is used for systems and components supported at elevations above the foundation mat. Since all of the NSSS systems and components are generally located above ground, the floor (in structure) respense spectra are normally dependent upcn the building design and which becomes plant dependent. It may be desirable to conduct seismic analysis based on a response spectrum which is constructed to envelope V the floor (in structure) response spectra for different plants. For the time history analysis purpose, DEBLIN2 code can also be used to generate a synthesized time history motion which has the response spectrum resembling' closely the envelope response spectrum.
A major portion of the code consists of subroutines which compute the ground-acceleration response spectrum of a vibratory motion chosen as the input. The code then compares the response spectrum with the design response WAPWR-S/E 3A.7-1 AMENDMENT 2 7139e:1d JANUARY, 1989
APPENDIX 3.7 (con't) spectrum and selects frequency points for which time history has to be modified. Subroutines using the techniques of spectrum suppression and raising then makes the appropriate changes in the time history motion. This process is repeated until a time history motion with an acceptable response spectrum is determined.
O A complete description of the Westinghouse computer program DEBLIN2 is included in Westinghouse Nuclear Energy Systems Report WCAP-8867 "DEBLIN2: A Computer 2
Code to Synthesize Earthquake Acceleration Time Histories". This report also include: several verification examples. As shown in Table 1.6-1 of RESAR' SP/90-Hodule 7 (Docket No. 50-601), this report was submitted to the NRC in November, 1976.
O Reference 1.: N. C. Tsai, " Spectrum Compatible Motions for Design Purposes,"
b Journal of Engineering Mechanics Division, ASCE, Vol. 93, April, 1972.
O ,
i I
O WAPWR-S/E 3A.7-2 AMENDMENT 2 7139e:1d JANUARY, 1989
L .
TABLE'3.8-3~
~
LOAD COMBINATIONS'AND LOAD FACTORS FOR CATEGORY I CONCRETE STRUCTURES-
~
Load Combinations and Factors Combination # 1 '2 3~ 4 5 6 7 8 9- 10' 11 Load Description Dead D 1.4. 1.4- 1.4' 1.0 l'. 0 - 1. 0 ' 1.0- 1.0 1.05' 1.05 1. 05 --
Liquid -1.4 1.4 1.411.0 1.0. 1.0 1. 0 ' 1.0 .1.05 1.05 1.05' O Live F
L 1.7 1.7 1.7 l'. 0 1.0 1.0 1.0 1.0 1.3 1.3 1.3 Earth H 1.7. 1.7 1.7 1.0. 1.0 1.0 1.0 1.0 1.3 1.3 1.3 Normal reaction Ro 1.7 1.7 1.7 1.0 1.0 1.3 1.3 '1.3 Normal-thermal To 1.0 1.0' 1.4 1.4- 1.4.
OBE Eo .1.9 1.25(3) 1.3 Wind W 1.7 1.3 SSE. Es 1.0 1.0(3)
Tornado Wt- 1.0 Accident thermal- Ta -1.0 1.0 1.0 O Accident thermal reactions Ra 1.0 1.0 1.0 o
1 2
Ace' dent pressure Pa 1.5 1.25 1.0 Accidentjet, Missile reactions Y 1.0 1.0 1
i Notes: 1) Design per'ACI-349 Strength Design Method for all load combinations
- 2) Where any load reduces the effects of other loads, the correspond- 1 ing coefficient for that load shall be taken as 0.9 if it can be-demonstrated that the load is always present or occurs simultane-ously with the other loads. Otherwise the coefficient for the load shall be taken as zero.
- 3) Seismic loads will only be combined with ruptures of pipes that are not seismically supported.
O WAPWR-S/E 3.8-21 AMENDMENT 2 2047e:1d J ANUARY,1989 i 1
TABLE 3.8-4 g LOAD COMBINATIONS AND LOAD FACTORS FOR CATEGORY I STEEL STRUCTURES T Load Combinations and Factors Combination # 1 2 3 4 5 6 7 8 9 10 11 Load Description -
Dead D 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Liquid F 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Live' L 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Earth H 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Normal reaction Ro 1.0 1.0 1.0 ) - 1.0 1.0 1.0 1.0 Normal thermal To 1.0 1.0 1.0 1.0 1.0 l
OBE Eo 1.0 1.0(3) 1.0 Wind W 1.0 1.0 SSE Es 1.0 1.0(3)
Tornado Wt 1.0 Accident thermal Ta 1.0 1.0 1.0 Accident thermal reactions Ra 1.0 1.0 1.0 Accident pressure Pa 1.0 1.0 1.0 Accident jet, Missile reactions Y 1.0 1.0 Allowable Stress 1.0S 1.65 1.5S ,
Notes: 1) S denotes allowable stresses per AISC Specification, Part I
- 2) Where any load reduces the effects of other loads the correspond-ing coefficient for that load shall be taken as zero unless it can be demonstrated that the load is always present or occurs simul-taneously with the other loads.
- 3) Seismic loads will only be combined with ruptures of pipes that are not seismically supported.
O WAPWR-5/E 3.8-22 DECEMBER, 1984 2047e:1d
b v APPENDIX 3.8 COMPUTER CODE WECAN O The WECAN computer program is used to solve a large variety of structural l analysis problems. These problems can be one, two, or 'three-dimensional in L nature. It has the capability to do static elastic and inelastic analysis, steady state and transient heat conduction, steady state hydraulic analysis, standard and reduced modal analysis, harmonic response analysis, transient f
dynamic analysis, and linear buckling analysis, i The WECAN program is based on the finite element method of analysis. The j structure is modeled in terms uf discrete elements and appropriate loadings j and boundary conditions are applied. The stiffness (or conductivity) matrix l for each element is assembled into a system of simultaneous linear equations ;
for the entire structure. This set of equations is then solved by a variation !
p.
,)
of the Gaussian elimination method known as.the wave front technique, f2 Verification of portions of WECAN capability has been performed since the first version of WECAN was released in January, 1973. In October, 1973, a set l of 72 verification problems was proposed to be the first set of problems to l verify portions of the code. These were assigned to individuals in 11 !
divisions to perform the analyses and document the results. Twenty-two of ,
these problems were written up and placed in the Verification Manual (WECAN, 1974b) which was issued in August, 1974. i l
1 1
(~'N Q Currently, the input for these verification problems for WECAN is stored on an j update tape maintained at the Westinghouse Power Systems Computer Center by J the WECAN Working Group. The file name of the tape is VERWEC. ;
WECAN is continually reverified. The problems included in VERWEC are run i before sending changes to Configuration Control. About once a year, each problem on VERWEC is rerun in its entirety to ensure that it still produces good results.
O V
WAPWR-S/E 3A.8-1 AMENDMENT 2 7139e:1d JANUARY, 1989
l
()
V' APPENDIX 3.8 (con't) l ,
A detailed desc'ription of the verification of computer code 'WECAN' is (3
\"/
included in Westinghouse Nuclear Energy Systems Report, WCAP-8929, " Benchmark Problem Solutions Employed for Verification of the WECAN Computer Program."
As stated in NRC Docket No. 99900404/84-03, dated January 7, 1985, an l inspection to evaluate the technical justification and verification of WECAN l
,/ was conducted in October 1984, by the NRC, Division of Quality Assurance, j O] Safeguards and Inspection Programs, Office of Inspection and Enforcement.
Detailed verification of various Wt:CAN capabilities is also provided in the following references:
i
- 1. WECAN Verification Manual. Stephen E. Gabrielse, Editor. Pittsburgh:
WECAN Working Group.
- 2. Filstrup, Alvin W., " Verification of a General Purpose Finite Element 2 b(x Computer Code - A Case History" pp-1-8 in Pressure Vessels and Pioing Comouter Procram Evaluation and Qualification. New York: American Society of Mechanical Engineers, September,1977, PVP-PB-024.
- 3. Riggio, Mark D. and Vikram N. Shah, " Verification of Modal Superposition Method in Structural Computer Code." pp. 103-111 in Pressure Vessels and Piping Computer Program Evaluation and Qualification. New York: American Society of Mechanical Engineers,
, Special Publication PVP-PB-024, 1977.
- 4. Chan, Siu-Kee, Yung Fan, Alvin W. Filstrup and Stephen E. Gabrielse, j
" Verifying the Plastic Capabilities in a General Purpose Computer !
Code." pp. 51-66 in Pressure Vessels and Piping Computer Program
/"_N
(
v) Evaluation and Qualification. New York:
Mechanical Engineers, September, 1977, PVP-PB-024.
American Society of I
l Ov WAPWR-S/E 3A.8-2 AMENDMENT 2 7139e:1d JANUARY, 1989
APPENDIX 3.8 (con't)
- 5. Vachi, Kiran M., " Verification of the WECAN Non-Linear Elastic Dynamic L Analysis Capability." Pittsburgh: Westinghouse Electric Corporation-
]V Nuclear Energy Systems, WCAP-8281, 1974.
- 6. Bennett, Stephen B., " Structural Analysis of Turbine Components with i
l (' the WECAN Code: A Status Report." Lester, PA: Westinghouse Steam '
Turbine Division, Engineering Report EM-1481, 1974.
L l- Computer Code 'ASHSD' l This program employs the finite element method for the dynamic analysis of l -' complex axisymmetric structures subjected to any arbitrary static or dynamic i
loading or base acceleration. The three-dimensional axisymmetric continuum is represented either as an axisymmetric thin shell or as a solid of revolution or as a combination of both. The axisymmetric shell is discretized as a series of frustrums of cones and the solid of revolution as triangular or 2 quadrilateral " toroids" connected at their nodal point circles.
Hamilton's variational principle is used to derive the equations of motion, and the diagonal mass matrix formulations are adopted to economize the computer storage and time. These equations of motion are solved numerically through the time domain either by direct integration or by mode superposition. In both cases, the numerical scheme adopts to the step-by-step )
integration procedure. For an earthquake analysis, the method of modai l analysis by response spectrum can be called for.
a The computer code was developed in 1969 by S. Ghosh and E. Wilson at the' Earthquake Engineering Research Center, University of California, Berkeley
-(Report No. EERC-69-10). It has been used in the analysis and design of l several nuclear plant containments. Verification examples, for the version used by Westinghouse, are documented in Westinghouse files.
O 1 WAPWR-S/E 3A.8-3 AMENDMENT 2 7139e:1d JANUARY, 1989
Case 2.- Last Pump Start-up (First Pump Shutdown) q NJ This case conservatively . represents the variations in reactor coolant loop -flow accompanying start-up of the second 'and third pumps.-
Initially flow exists through the second and' third loops in the reverse direction as the result of starting the first pump. The remaining pumps in these loops are then started in sequence and a new-equilibrium flow is established. The magnitude of flow reversal is the largest in. the loop containing the last pump to be started. For O the first pump shutdown case, the transient is the reverse of the last Q
pump start-up transient.
Table 3.9-2 includes the RCP start-ups and shutdowns associated with 2 RCS heatup and cooldown.
The values shown in this table represent the design conditions for the pump starting and stopping operations. The processes by which these conditions are attained are parts of other operations and are not defined here. For example, the RCS venting operation involves pressurizing the. system to approximately 400 psig with a charging pump, starting and stopping one RCP to purge out air during the venting, then depressurizing back to essentially. atmospheric pressure. For design purposes this process is assumed to be repeated four times per loop for each of 200 venting operations during the plant lifetime. This establishes the design value of 800 starts / stops for each RCP associated with the venting operation.
Another consideration is that the loop flow change associated with O pump start-up develops a pressure differential in the normal (forward) direction across the divider plates of the steam generator in that loop. In the loops undergoing reverse flow, the direction of the l divider plate AP is reversed. The magnitudes of the dynamic pressure drops depend on the volumetric flow rate through the loop and on the density and viscosity of the reactor coolant.
O WAPWR-S/E 3.9-5 AMENDMENT 2
, 2045e:1d JANUARY, 1989 l
l L__ _ _ .____ _ _____________ _
i l
l B. Plant Heatup and Cooldown The plant heatup and cooldown operations are conservatively represented by !
uniform temperature ramps of 100'F per hour when the system temperature is above 350*F. This rate bounds both potential nuclear heatup operations and cooldown using the steam dump system.
Below 350'F, only reactor coolant pump heat and small amounts of decay heat are available to heat the RCS, Cooldown between 350*F and the shutdown temperature of 120*F(1) is accomplished by the Residual Heat Removal System (RHRS). In this range, a uniform ramp rate of 50*F per hour bounds the temperature rate change resulting from these operations.
Rates in excess of the above values will not be attained in actual practice because of other limitations such as:
- a. material ductility considerations which establish maximum permissible temperature rates of change, as functions of RCS pressure and temperature.
- b. Reduction in heatup rates on pump energy only because of increasing losses and decreasing pump power as system temperature increases,
- c. Reductio.' in cooldown rates as steam dump and residual heat removal approach tncir respective temperature endpoints.
- d. Interruptions in the heatup and cooldown cycles due to such factors as protection against RCS cold overpressure, pressurizer steam bubble formation, control rod withdrawal, sampling, water chemistry control and gas adjustments.
- 1. Reactor Coolant System temperature can be as low as 70 F during the shut-down period. Between 70*F and 120*F the temperature is assumed to change very slowly, without causing any significant thermal transient effects.
WAPWR-S/E 3.9-6 DECEMBER, 1984 O
2045e:1d
(~) Both the primary and secondary sides of the steam generator are be at ambient temperature during these tests.
l The total number of tube leakage test cycles is defined as 800. during the
/'N 40 year life of the plant. Following is a breakdown of the anticipated number of occurrences at each secondary side test pressure:
Number of
(~3 Test Pressure, psig Occurrences l V I 200 400 400 200 600 120 1
840 80 Neither the primary side nor secondary side design pressures are exceeded during the Tube Leakage Test. This test is included under Test Conditions j since the expected secondary-to primary pressure differentici exceeds the V design value of 670 psi for some of the test cycles.
3.9.1.2 Computer Programs Used in Analysis i
3.9.1.2.1 NPB Systems and Components The following computer programs will be used in dynamic and static analyses to determine mechanical loads, stresses, and deformations of Seismic Category I components and equipment.
O\
O A. WESTDYN - static and dynamic analysis of redundant piping systems.
B. FATCON - fatigue analysis of piping systems f'
O) C. WESAN - reactor coolant loop equipment support structures analysis and evaluation, f3 l U l WAPWR-S/E 3.9-37 DECEMBER, 1984 2045e:1d
D. WECAN -
finite-element structural analysis and nonlinear time history seismic analysis.
3.9.1.3 Experimental Stress Analysis No experimental stress analysis methods are used for Category I systems or I components. However, Westinghouse makes extensive use of measured results l from prototype plants and various scale model tests as discussed in Subsection 3.9.2.
3.9.1.4 Consideration for the Evaluation of the Faulted Condition O
The analytical methods used to evaluate the faulted conditions for seismic 2 l Category I ASME Code and non-code items are described in the Subsection 3.9.3 l of this module.
3.9.2 Dynamic Testing and Analysis 3.9.2.1 Piping Vibration, Thermal Expansion, and Dynamic Effects A pre-operational piping test program will be implemented on selected piping systems to verify the following:
A. The steady-state and transient vibration levels in the piping system are within allowable limits.
2 B. The dynamic factors due to fluid transients are within allowable limits.
C. The thermal expansion of the piping systems is as predicted by design analysis. ,
O' l
WAPWR-S/E 3.9-38 AMENDMENT 2 Ol 2045e:1d J ANUARY, 1989
_ _ _ _ _ _ _ _ . J
O The detailed test plan covering piping pre-operational testing is submitted to the NRC at least 60 days prior to the initiation of the test program.
.A. Steady-State and Transient Vibration Test l V.O The assessment of piping system vibration will be based on testing procedures defined in the draft ASME/ ANSI OM-1987, Part 3,
" Requirements for Pre-operational and Initial Start up Vibration
,, Testing of Nuclear Power Plant Piping Systems." The standard will serve as a guide for the evaluation of vibration levels of applicable
, )
piping systems. When required, additional design or restraint modifications will be provided to reduce the effects of vibration to acceptable levels. Essential safety related instrumentation lines will be included in the vibration monitoring program.
l I
B. Piping System Dynamic Testing 2 Anticipated operational fluid transients will be assessed in the
() pre-operational test program for the following systems:
- 1. Main steam turbine step valve trip
- 2. Main steam atmospheric dump valves opening
- 3. Main steam condenser dump valves opening
- 4. Steam generator power-operated relief valve opening
- 5. Main steam isolation valve closure
- 6. Main feedwater line check valve closure
- 7. Pressurizer power-operated relief valve opening O The piping system dynamic response to the fluid transients will be recorded and compared to the design basis analysis.
O O WAPWR-S/E 3.9-38a AMENDMENT 2 2045e:1d JANUARY, 1989 l
1 1
l
_ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ . _ . _ . _ . _ _ . ..__ . _ . _ _ . _ . . ._ ___ m ._. !
C. Thermal Expansion Testing O
During pre-operational testing, piping system movements will be measured or observed at various locations on selected portions of the following systems:
O
- 1. Reactor cooling system
- 2. Main steam system
- 3. Main feedwater system 2 4. Chemical and volume control system
- 5. Residual heat removal (RHR) system
- 6. Containment spray system
- 8. Secondary side safeguards system
- 9. Steam generator blowdown system
- 10. Component cooling water system This thermal expansion test will be implemented for one complete thermal cycle (from cold position to hot condition and back to the original cold position). The purpose of. this test will be to confirm that the pipe moves as predicted by design analysis.
O O
I 3.9-38b AMENDMENT 2 O
WAPWR-S/E 2045e:1d JANUARY, 1989 l
O O
O 2
.I O
i O
O WAPWR-5/E 3.9-39 AMENDMENT 2 7045e:1d JANUARY, 1989
O O
O' 2
0 G
O l
3.9-40 AMENDMENT 2 O
WAPWR-S/E 7045e:1d JANUARY, 1989 l
1
.y 0
1 i
^ l l
2
(
l i
f i
l WAPWR-S/E 3.9-41 AMENDMENT 2 2045e:1d JANUARY, 1989
l 3.9.2.2 Seismic Qualification Testing of Safety-Related Mechanical Equipment O
Westinghouse utilizes analysis, test, or a combination of test and analysis for seismic qualification of equipment. Testing is the preferred method; however, analysis is utilized when one of the following conditions is satisfied:
A. The equipment is too large or the external loads, connecting elements, or appurtenances cannot be simulated with a shaker table test.
B. The only requirement that must be satisfied relative to the safety of the O
plant is the maintenance of structural integrity (mechanical equipment only).
C. The component represents a simple linear system or nonlinearities can be conservative; accounted for in the analysis.
The operability of Seismic Category I mechanical equipment must be demonstrated if the equipment is active; i.e., mechanical operation is relied on to perform a safety function. The operability of active Safety Class 2 and 3 pumps, active Safety Class 1, 2, or 3 valves and their respective drives, operators and vital auxiliary equipment is shown by satisfying the criteria given in Subsection 3.9.3.2.
Inactive Seismic Category I equipment such as heat exchangers, racks, and consoles are shown to have structural integrity during a seismic event by analysis satisfying the stress criteria applicable to the particular piece of equipment.
A list of Seismic Category I equipment is provided in Table 3.2-1.,
The criteria used to decide whether dynamic testing or analysis should be used to qualify Seismic Category I mechanical equipment are as follows:
l WAPWR-5/E 3.9-42 DECEMBER, 1984 O
2045e:1d
_ _ - - _ - _ l
O A. Analysis Without Testing d
- 1. Structural analysis without testing is used if structural integrity-alone can assure the design-intended function. Equipment which falls
'into this category includes:
- a. Ductwork
- b. Tanks and vessels
- c. Heat exchangers (3
() d. Filters
- e. Inactive valves
- 2. Rotational analysis without testing is used to qualify rotating machinery items where it must be verified that deformations due to seismic loadings will not cause binding of the rotating element to the extent that the component cannot perform its design-intended function.
The seismic qualification of p. umps is discussed more fully in Subsection 3.9.3.2.1. The procedure discussed therein applies, with some variations, to other items in this category.
- 3. Dynamic analysis without testing is used to qualify heavy machinery too large to be tested. It is verified that deformations due to seismic loadings will not cause binding of the moving parts to the extent that the component cannot perform its required safety function. Components which fall into this category include:
I o Pumps o Turbines o Generators o Fans o Diesel engines O 3.9-43 DECEMBER, 1984 WAPWR-S/E 2045e:1d
1 l
l B. Dynamic Testing <
Dynamic testing is used for components with mechanisms that must change j position in order to perform their required safety ' unction. Such 1
components include:
Electric motor valve operators O
o Valve limit switches o
o Similar appurtenances for other active mechanical equipment e
The seismic qualification of Seismic Category 1 electrical equipment is discussed in Section 3.10.
C. Combinations of Analysis with Testing A combination of analysis, static testing, and dynamic testing is used for seismic qualification of complex equipment. Such equipment includes:
- 1. Standby diesel generators
- 2. Turbine-driven emergency feedwater pumps
- 3. Main steam and main feedwater isolation valves
- 4. Other active valves The seismic qualification of active valves is discussed more fully in Subsection 3.9.3.2.
The acceptance criteria which are used are as follows:
- 1. Tests, when used, demonstrate that the component is not prevented from performing its required safety function during and after the test.
- 2. Analysis, when used for qualification of vessels, pumps, or valves, verifies that stresses do not exceed the allowables specified in 2 Tables 3.9-3 and 3.9-4 and that deformations do not exceed those which !
I WAPWR-S/E 3.9-44 AMENDMENT 2 7045e:1d JANUARY, 1989
will permit the component to perform its design-intended function.
(n "}_ The results of tests and analyses of safety-related mechanical.
equipment are available for inspection.
t 3.9.2.3 Dynamic Response Analysis of Reactor Internals Under Operational Flow Transients and Steady-State Conditions For a discussion of the dynamic response analysis of the reactor internals, see Subsection 3.9.2.3 of RESAR-SP/90 PDA Module 5, " Reactor System."
l 3.9.2.4 Preoperational Flow-Induced Vibration Testing of Reactor Internals For a discussion of the preoperational flow-induced vibration testing of the l reactor internals, see Subsection 3.9.2.4 of RESAR-SP/90 PDA Module 5,
" Reactor System."
3.9.2.5 Dynamic System Analysis of the Reactor Internals Under Faulted p Conditions V
For a discussion of the dynamic system analysis of the reactor internals, see Subsection 3.9.2.5 of RESAR-SP/90 PDA Module 5, " Reactor System."
3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures J
3.9.3.1 Loading Combinations, Design Transients, and Stress Limits American Society of Mechanical Engineers (ASME) Class 1, 2, and 3 components I and component supports are designed to an appropriate combination of plant conditions and design loadings. The plant conditions are design, normal, upset, emergency, and faulted conditions. The design loadings ars pressure,
' temperature and deadweight loads.
The ASME Code Class components are constructed in accordance with the ASME B&PV Code, Section 111 requirements. For Code Class 1 components, very i stringent requirements are imposed and are met. For Code Class 2 and 3 WAPWR-S/E 3.9-45 DECEMBER, 1984 2045e:1d
i l
components, the requirements are less stringent but adequate, in accordance with the lower classification.
3.9.3.1.1 ASME Code Class 1 Components and Supports Loading combinations for ASME Class 1 components and component supports are 2 presented in Table 3.9-5 and stress limits for these components are given in Table 3.9-3. A detailed discussion of design transients for the NPB components is provided in Subsection 3.9.1.
O The structural stress analyses performed on the ASME Class 1 components and 2 supports consider the loadings specified as shown in Table 3.9-5. These loads result from thermal expansion, pressure, weight, operating basis earthquake (OBE), safe shutdown earthquake (SSE), design basis loss-of-coolant accident, and plant operational thermal and pressure transients.
3.9.3.1.1.1 Analysis of the Reactor Coolant Loop Piping and Supports l
The loads used in the analysis of the reactor coolant loop piping are described in detail below.
A. Pressure Pressure loading is identified as either membrane design pressure or general operating pressure, depending upon its application. The membrane design pressure is used in connection with the longitudinal pressure stress and minimum wall thickness calculations in accordance with the ASME 1 Code as stated in Table 3.2-1.
1 The term operating pressure is used in connection with determination of )
the system deflections and support forces. The steady-state operating hydraulic forces based on the system initial pressure are applied as I general operating pressure loads to the reactor coolant loop model at change in direction or flow area.
O WAPWR-S/E 3.9-46 AMENDMENT 2 2045e:1d JANUARY, 1989
B. Weight A deadweight analysis is performed to meet Code requirements by applying a 1.0 g load downward on the complete piping system. The piping is assigned a- distributed mass or weight as a function of its properties. This' method O provides a distributed loading to the piping system as a function of the weight of the pipe and contained fluid during normal operating conditions.
C. Seismic The input for the reactor coolant loop piping seismic analysis is in the form of three statistically independent orthogonal time-history accelerations. The earthquake accelerations for the horizontal directions are applied to the containment basemat simultaneously with the vertical acceleration in the vertical direction For' the OBE and SSE seismic analyses, 2 and 4 percent critical damping, respectively, are used in the reactor coolant loop / supports system analysis. The values are based on 2 test results reported in " Damping Values of Nuclear Power Plant Components," WCAP-7921-AR, May 1974.
Optional seismic analysis methods which may be used for low seismic plants include the uniform response spectra method and the multiple response spectra method. The damping values used in these methods are the same as those in the time history method. Alternatively, when the uniform response spectra method is used, the damping values in Figure 3.7-8 may be 2 used. Energy absorbing supports are not used in the reactor coolant loop supports.
O D. Loss-of-Coolant Accident Dlowdown loads are developed in the reactor coolant loops as a result of transient flow and pressure fluctuations following a postulated pipe break ,
in one of the reactor coolant loop auxiliary connections. Structural consideration of dynamic effects of postulated pipe breaks requires postulation of a finite number of break locations. Postulated pipe break locations are given in Section 3.6.
WAPWR-S/E 3.9-47 AMENDMENT 2 2045e:1d JANUARY, 1989 l
Time-history dynamic analysis is performed for these postulated break g cases. Hydraulic models are used to generate time dependent hydraulic W forcing functions used in the analysis of the reactor coolant loop for each break case.
E. Transients The ASME Code,Section III requires satisfaction of certain requirements relative to operating transient conditions. Operating transients are summarized in Subsection 3.9.1.1.
To provide the necessary high degree of integrity for the RCS, the transient conditions selected for fatigue evaluation are based on conservative estimates of the magnitude and anticipated frequency of occurrence of the temperature and pressure transients resulting from various plant operation conditions.
The analytical methods used in obtaining the solution consist of the transfer matrix method and stiffness matrix formulation for the static structural analysis, the time-history integration, or response spectra methods, for seismic dynamic analysis, and time history integration analysis methods for effects of high-energy line pipe breaks.
l The integrated reactor coolant loop / supports system model is the basic system I
model used to compute loadings on components, component supports, and piping.
The system model includes the stiffness and mass characteristics of the reactor coolant loop piping and components, the stiffness of supports, the l stiffnesses of auxiliary line piping which affect the system. The deflection solution of the entire system is obtained for the various loading cases from l which the internal member forces and piping stresses are calculated. I l
A. Static ,
l The reactor coolant loop / supports system model, constructed for the WESTDYN computer program, is represented by an ordered set of data which O
WAPWR-S/E 3.9-48 DECEMBER, 1984 2045e:1d
3.9.3.1.1.2 Class 1 Auxiliary Branch Lines The allowable stresses for ASME Code Class 1 components and supports are given 4 in Table 3.9-3. All Class 1 components and supports are designed and analyzed I for Levels A, B, and C Service Conditions, and corresponding service level requirements to the rules and requirements of the ASME Code,Section III. The analysis or test methods, and associated stress or load allowable limits that are used in evaluation of Level D Service Conditions are those that are defined in Appendix F of the ASME Code,Section III with the following supplementary option The analytical methods used to obtain the solution consist of the transfer matrix method and stiffness matrix formulation for the static structural analysis, and the response spectrum method for seismic dynamic analysis.
The integrated Class 1 piping and supports system model is the basic system model used to compute loadings on components, component and piping supports, and piping. The systeu models include the stiffness of supports that affect 2 q(/ the system response. The deflection solution of the entire system is obtained for the various loading cases from which the internal member forces and piping stresses are calculated.
A. Static The Class 1 piping system models are constructed for the WESTDYN computer program, which numerically describes the physical system. A network model is made up of a number of sections, each having an overall transfer
(
) relationship formed from its group of elements. The linear elastic properties of the section are used to define the characteristic stiffness matrix for the section. Using the transfer relationship for a section, the loads required to suppress all deflections at the ends of the section f%
Q arising from the thermal and boundary forces for the section are obtained.
O WAPWR-S/E 3.9-55 AMENDMENT 2 2045e:1d JANUARY, 1989
After all the sections have been defined in this manner, the overall stiffness matrix and associated load vector to suppress the deflection of all the network points is determined. By inverting the stiffness matrix the flexibility matrix is determined. The flexibility matrix is multiplied by the negative of the load vector to determine the network point deflections due to the thermal and boundary force effects. Using 1 the general transfer relationship, the deflections and internal forces are then determined at all node points in the system.
The support loads are also computed by multiplying tha stiffness matrix by i the displacement vector at the support point.
B. Seismic The models used in the static analyses are modified for use in the dynamic analyses by including the mass characteristics of the piping and equipment.
The lumping of the distributed mass of the piping systems is accomplished by locating the total mass at points in the system that will appropriately represent the response of the distributed system. Effects of the primary equipment motion, that is, reactor vessel, steam generator, reactor coolant pump, and pressurizer, on the Class 1 piping system are obtained by modeling the mass and the stiffness characteristics of the primary ;
equipment and loop piping in the overall system model. Alternately, the effects of the primary equipment and loop motion are represented by resonse spectra and anchor motions calculated at the connection points to the RCS loop.
The supports are represented by stiffness matrices in the system model for the dynamic analysis. Shock suppressors that resist rapid motions are also included in the analysis. The solu + ion for the seismic disturbance l
employs the response spectra method. This method employs the lumped macs technique, linear elastic properties, and the principle of modal I
superposition.
DECEMBER, 1984 O
WAPWR-S/E 3.9-56 2045e:1d l
\ - - - - - - - - - ---
The damping values for auxiliary piping systems are shown in Table 3.9-6.
Piping systems with different nominal diameters and different damping characteristics are evaluated using the methods of Subsection 3.7.3.15.
Alternatively, when the uniform response spectra method is used, the O damping values in Figure 3.7-8 may be used. Energy absorbing supports are Figure 3.7-8 is applicable to not used for auxiliary piping systems.
2 i piping system models which include flexible components such as valves or tanks. The use of composite damping values for these systems is not
,C'\ justified based on the following:
V a) The test data base for the Code Case N-411 damping including systems with in-line components are building mounted components.
I b) Sample analytical calculations for composite damping show an insignificant contribution from the flexible components.
i The total response obtained from the seismi,c analysis consists of two p parts: the inertia response of the piping system, and the response from
\ differential anchor motions. The stresses resulting from the anchor ,
motions are considered to be secondary and, therefore, are included in the fatigue evaluation.
C. Loss-of-Coolant Accident The mathematical models used in the seismic analyses of the Class 1 lines are also used for RCL auxiliary pipe break effect analysis. To obtain the !
dynamic solution for auxiliary lines 6 inches and larger, and certain small-bore lines required for ECCS consideration, the time history deflections from the analysis of the reactor coolant loop are applied at branch nozzle connections. For other small-bore lines that must maintain structural integrity, the motion of the RCL is applied statically.
O WAPWR-S/E 3.9-57 AMENDMENT 2 2045e:1d JANUARY, 1989
)
D. Fatigue A thermal transient heat transfer analysis is performed for each different piping component on all the Class 1 branch lines. The normal, upset, and test condition transients identified in Section 3.9.1.1 are considered in g the fatigue evaluation. W For each thermal transient, two load sets are defined representing the maximum and minimum stress states for that transient.
O The FATCON computer program is used to calculate the primary plus-secondary and peak stress intensity ranges, fatigue reduction factors, and cumulative usage factors for all possible load set combinations. It is conservatively assumed that the transients can occur in any sequence, thus resulting in the most conservative and restrictive combinations of transients.
O O
O l
I WAPWR-S/E 3.9-57a AMENDMENT 2 7045e:1d JANUARY, 1989
____________________O
The combination 'of load sets yielding the highest alternating stress intensity. range is determined and the incremental usage. factor calculated. Likewise, the next most severe combination is then determined and the incremental usage factor calculated. This procedure is repeated 6
p- until all combinations having an allowable cycle of <10 are formed.
U The total cumulative usage factor at a point is the summation of the incremental usage factors.
3.9.3.1.1.3 Loading Combinations and Stress Limits Loading combinations and stress limits for Class 1 components and supports are 2 given in Tables 3.9-5 and 3.9-3. Detail load combinations and stress limits for. the pressurizer and safety and relief valve piping are described in Subsection 3.9.3.
3.9.3.1.2 ASME Code Class 2-and 3 Components and Supports i The loading combinations for ASME Code Class 2 and 3 components and supports 2 l furnished with the NPB are given in Table 3.9-4.
The allowable stress limits established for the components are sufficiently low to assure that violation of the pressure retaining boundary will not occur. These limits, for each of the loading combinations, are component oriented and are presented in Table 3.9-4. Active (a) pumps and valves are 2
further discussed in Subsection 3.9.3.2. The component supports are designed in accordance with ASME B&PV Code,Section III, Subsection NF.
The seismic analysis methods are described in Subsection 3.7.3.8. The damping values are described in Subsection 3.9.3.1.1.2B.
( a. Active components are those whose operability is relied upon to perform a safety function (as well as reactor shut down function) during the j transients or events corsidered in the respective operating condition I categories. i i
1 1 O WAPWR-S/E 3.9-58 AMENDMENT 2
'2045e:1d JANUARY, 1989
3.9.3.1.3 Analysis of Primary Components and Valves Primary components that serve as part of the pressure boundary in the reactor coolant loop include the steam generators, reactor coolant pumps, pressurizer, O
e l
O e
e i
WAPWR-S/E 3.9-58a AMENDMENT 2 Ol!
2045e:1d JANUARY, 1989 l l
I
! piping, and reactor vessel. This equipment is American Nuclear Society (ANS)-
l Safety Class l'and the pressure boundary meets the requirements of ASME Code, Section III. This equipment is evaluated for the loading combinations outlined in Table 3.9-5. The equipment is. analyzed for (1) the normal loads 2 of weight, pressure, and . temperature, (2) mechanical transients of OBE, SSE, and auxiliary line pipe. ruptures, and (3) pressure and temperature transients outlined in Section 3.9.1.1.
The results of the reactor coolant loop analysis are used to determine the
. loads acting on the equipment nozzles and the support / component interface locations. These loa'ds are supplied for all loading conditions on an
" umbrella" load basis. That is, on the basis of previous plant analyses, a set of loads is determined which should be larger than those'seen in any single plant analysis. The umbrella loads represent a conservative means of allowing detailed component analysis prior to the completion of the system analysis. Upon completion of'the system analysis, conformance is demonstrated between the actual plant loads and the loads used in the analyses of the components. Any actual loads larger than the umbrella loads are evaluated by l O individualized analysis.
Seismic analyses are performed individually for the RCP, the pressurizer, and the steam generator. Detailed and complex dynamic models are used for the dynamic analyses. The response spectra corresponding to the building elevation at the highest component / building attachment elevation is used for the component analysis. Seismic analyses for the steam generator, RCP, and 2
pressurizer are, performed using 2 percent damping for the OBE and 4 percent damping for the SSE. The reactor pressure vessel is seismically qualified in accordance with ASME III by the reactor vessel vendor. The loadings. used in the analysis are supplied by Westinghouse and are based on loads generated by a dynamic system analysis.
Auxiliary equipment that serves as part of the reactor coolant system pressure boundary include Class 1 valves and Class 1 auxiliary piping. Class 1 valves in the RCS are designed and analyzed according to the requirements of Subsection NB-3500 of ASME Code,Section III. This equipment is evaluated for O
WAPWR-S/E 3.9-59 AMENDMENT 2 2045e:1d JANUARY, 1989
2 the loading combinations and stress limits in Tables 3.9-5 and 3.9-3. The operability criteria for these valves are described in Section 3.9.3.2.
Valves in sample lines connected to the RCS are not considered to be ANS Safety Class 1 nor ASME Class 1. This is because the nozzles where the lines connect to the primary system piping are orificed to a 3/8-inch hole. This hole restricts the flow so that loss through a severance of one of these lines can be made up by normal charging flow.
3.9.3.2 Pump and Valve Operability Assurance 3.9.3.2.1 Pumps Safety related active pumps are subjected to in-shop tests which include hydrostatic tests of casing to 150 percent of the design pressure, and performance tests to determine the following:
o Total developed head, o Minimum and maximum head.
o Net positive suction head (NPSH) requirements except as noted below, o Other pump / motor characteristics Where applicable, bearing temperature and vibration are monitored during the performance tests, After the pump is installed, it undergoes cold hydrostatic testing, hot functional testing, and applicable periodic inservice inspection and testing to verify and assure the functional ability and reliability of the pump for the design life of the plant.
In addition to the required testing, the pumps are designed and supplied in accordance witn the following specified criteria:
O WAPWR-S/E 3.9-60 AMENDMENT 2 2045e:1d JANUARY, 1989
_ ____-__0
A. In order to ensure that the active pump will not be damaged during the seismic event, the pump manufacturer must demonstrate by test or analysis that the lowest natural frequency of the pump is greater than 33 Hz. The pump, when having a natural frequency above 33 Hz, will be considered essentially rigid. This frequency is considered sufficiently high to avoid problems with amplification between the component and structure for all seismic areas. A static shaft deflection analysis is performed. The b natural frequency of the support is determined and used in conjunction with the project seismic response spectra. The deflection determined from the static shaf t analysis is compared with the applicable clearances.
If the natural frequency is found to be below 33 Hz, a dynamic analysis is performed using a finite element model to determine the amplified input -
accelerations necessary to perform the shaft analysis. The shaft deflectica analyses are performed using the adjusted accelerations and the deflections compared with allowable shaft clearances. Assumptions used for generating the analytical model are verified by test.
B. The maximum seismic nozzle loads are also considered in an analysis of the pump supports to ensure that unacceptable system misalignment cannot occur.
C. To complete the pump qualification, the pump motor and all appurtenances vital to the operation of the pump are independently qualified for operation within their specified environment, as well as during the maximum seismic event in accordance with Institute of Electrical and Electronics Engineers (IEEE) Standard 344-1975 and the requirements of Regulatory Guide 1.100, " Seismic Qualification of Electric and Mechanical 2 Equipment far Nuclear Power Plants."
From this, it is concluded that the safety-related pump / motor assemblies will not be damaged, will continue operating under safe shutdown earthquake (SSE) loadings, and will perform their intended functions.
These proposed requirements take into account the complex characteristics of the pump and are sufficient to demonstrate and assure the seismic operability of the active pumps.
WAPWR-S/E 3.9-61 AMENDMENT 2 2045e:1d JANUARY, 1989
1 3.9.3.2.2 Valves Safety-related, active valves are subjected to a series of stringest tests prior to service and during the plant life. Prior to installation, the following tests are performed: shell hydrostatic test, backseat and main seat leakage tests, disc hydrostatic tests, and operational tests to verify that the valve opens and closes. For the operability qualification of motor operators for the environmental conditions over the installed life, refer to Section 3.11 and Subsection 3.1.3, Regulatory Guide 1.73. Cold hydro tests, hot functional tests, periodic inservice inspections, and periodic inservice operations are performed in situ to verify and assure the functional ability of the valve. These tests guarantee reliability of the valve for the design life of the plant. The valves are constructed in accordance with the ASME B&PV Code, Section III. On active valves, an analysis of the extended structure is performed for static equivalent seismic SSE loads applied at the center of gravity of the extended structure. The stress limits used for 2 active Class 2 and Class 3 valves are shown in Table 3.9-4. In addition to these tests and analyses, representative valves of each design type are tested for verification of operability during a simulated plant faulted condition event by demonstrating operational capabilities within the specified limits.
The testing procedures are described below.
The valve is mounted in a manner that will conservatively represent typical valve installations. The valve includes the operator, accessory solenoid valves, and limit switches when attached to the valve in service. The operability of the valve during a faulted condition is demonstrated by satisfying the following criteria:
A. Active valves shall have a first natural frequency that is not less than O
33 Hz.
B. A static load or loads equivalent to those resulting from the faulted condition accelerations is applied to the extended structure center of gravity so that the resulting deflection is in the nearest direction of 9
WAPWR-S/E 3.9-62 AMENDMENT 2 2045e:1d JANUARY, 1989
() the extended structure. The design pressure of the valve is applied to the valve during the static deflection tests.
,. , C. The valve is cycled while in the deflected position. The valve must
) function within the specified operating time limits while subject to design pressure.
D. Electrical motor operators, limit switches, and pilot solenoid valves i ) necessary for operation are qualified in accordance with IEEE Seismic v
Qualification Standards. IEEE Standard 344 and Regulatory Guide 1.100 2 are used for this qualification.
The above testing program applies to valves with extended structures. The testing is conducted on a representative number of valves. Valves from each of the primary safety related design types are tested. Valve sizes that cover the range of sizes in service are tested.
[ l s
Valves that are safety related, but can be classified as not having an
' '/ extended structure such as check valves and safety valves, are considered separately.
Check valves are characteristically simple in design, and their operation is not affected by seismic accelerations or the maximum applied nozzle loads.
The check valve design is compact, and there are no extended structures or masses whose motion could cause distortions that could restrict operation of the valve. The design of these valves is such that once the structural p integrity of the valve is assured, using standard design or analysis methods, U' the ability of the valve to operate is assured by the design features. In addition to these design considerations, the valve also undergoes the following: (1) in-shop hydrostatic test, (2) in-shop seat leakage test, and n (3) periodic in situ valve exercising and inspection.
0
(~h I
%J WAPWR-S/E 3.9-63 AMENDMENT 2 2045e:1d JANUARY, 1989 )
l Pressurizer safety valves are qualified for operability in the same manner as valves with extended structures, as described above. The qualification methods include analysis of the bonnet for static equivalent SSE loads, in i shop hydrostatic and seat leakage tests, and periodic in situ valve inspection. Additionally, representative pressurizer safety valves are tested to verify analysis methods. This test is described as follows: l A. The safety valve is mounted to represent the specified installation.
B. The valve body is pressurized to its normal system pressure.
C. A static load representing the faulted condition seismic load is applied to the top of the valve bonnet in the weakest direction of the extended structure.
D. The pressure is increased until the valve actuates.
E. Actuation of the valve at its setpoin.t ensures its operability during the faulted condition.
Using these methods, all the safety related valves in the system are qualified for operability during a faulted event. These methods conservatively simulate the seismic event, and assure that the active valves perform their safety related function when necessary.
3.9.3.2.3 Pump Motor and Valve Operator' Qualification Active pump motors and active valve motor operators, limit switches, and solenoid valves are seismically qualified in accordance with IEEE Standard 344-1975 and meet the requirements of Regulatory Guide 1.100, " Seismic 2
Qualification of Electric and Mechanical Equipment for Nuclear Power Plants."
O WAPWR-S/E 3.9-64 AMENDMENT 2 O1 7045e:1d JANUARY, 1989 l l
l
-_ 1
3.9.3.2.4 Active ASME Code Class 2 and 3 Pumps Safety related active pumps are subjected to in-shop tests that include hydrostatic tests of casing to 150 percent of the design pressure,, and performance tests to determine total developed head, minimum and maximum head, net positive suction head (NPSH) requirements, and other pump / motor characteristics. Vibration is monitored during the performance tests.
In addition to the required testing, the pumps are designed and supplied in accordance with the following specified criteria:
A. In order to ensure that the active pump will not be damaged during the seismic event, the pump manufacturer is required to demonstrate by test or analysis that the lowest natural frequency of the pump is greater than 33 Hz. The pump, when having a natural frequency above 33 Hz, will be considered essentially rigid. This frequency is considered sufficiently high to avoid problems with amplification between the component and structure for all seismic areas. A static shaft deflection analysis of the rotor is performed. The natural frequency of the support is determined and used in conjunction with the project seismic response spectra. The deflection determined from the static shaft analysis is compared to the allowable rotor clearances.
If the natural frequency is found to be below 33 Hz, an analysis is performed to determine the emplified input accelerations necessary to perform the static analysis. The static deflection analyses are performed using the adjusted accelerations.
B. The maximum seismic nozzle loads are also considered in an analysis of the pump supports to ensure that unacceptable system misalignment cannot occur.
'C. To complete the seismic qualification procedures, the pump motor and all appurtenances vital to the operation of the pump are independently qualified for operation during the maximum seismic event in accordance O '
WAPWR-5/E 3.9-65 AMENDMENT 2 7045e:1d JANUARY, 1989
with IEEE Standard 344-1975, and meet the requirements of Regulatory i 2 Guide 1.100, " Seismic Qualification of Electric and Mechanical 1 Equipment for Nuclear Power Plants."
From this, it is concluded that the safety related pump / motor assemblies will not be damaged and will continue operating under SSE loadings and will perform O
l O
O O
WAPWR-S/E 3.9-65a AMENDMENT 2 O
2045e:1d JANUARY, 1989
i j
i their intended functions. These proposed requirements take'into account the
. U) complex characteristics of the pump and are sufficient to demonstrate and assure the seismic operability of the active pumps.
3.9.3.3 Design and Installation Details for Mounting of Pressure Relief
' Devices 3.9.3.3.1 Pressure Relief Devices on NPS Components 3
(d
\
The pressurizer safety and relief valve (PSARV) discharge piping systems provide overpressure protection for the RCS. The spring-loaded safety valves located on top of the pressurizer are designed to prevent system pressure from exceeding design pressure by more than ten percent. The power operated relief l valves, also located on top of the pressurizer, are designed to prevent system pressure from exceeding the normal operating pressure by more than 100 psi. A water seal is maintained upstream of each valve to minimize leakage.
Condensate accumulation on the inlet side of each valve prevents any leakage p of hydrogen gas or steam through the valves. The valve outlet side is sloped d to prevent the formulation of additional water pockets.
If the pressure exceeds the setpoint and the valve opens, the water slug from the loop seal discharges. The water slug, driven by high system pressure, generates transient thrust forces at each location where a change in flow !
direction or area occurs. The valve discharge conditions considered in the thrust analysis of the PSARV piping systems are as follows: 1) the safety valves are assumed to open simultaneously while the relief valves remain j closed, and 2) the relief valves open simultaneously while the safety valves !
) are closed.
In addition ~ to these two cases, which consider water seal discharge (water' l
slug) followed by steam, solid water from the pressurizer (cold overpressure) i is also analyzed.
O DECEMBER, 1984 WAPWR-S/E 3.9-66 2045e:1d
l 1
f) For each pressurizer safety and relief piping system, an analytical hydraulic J model is developed. The piping from the pressurizer nozzle to the relief tank l nozzle is modeled as a series of single pipes. The pressurizer is modeled as ]
a reservoir which contains steam at constant pressure (2500 psia for safety
/3 system and 2350 psia for relief system) and at constant temperature of 680*F.
The pressurizer relief tank is modeled as a sink which contains steam and I
water mixture.
Fluid acceleration inside the pipe generates reaction forces on all segments D)
L' of the line which are bounded at either end by an elbow or bend. Reaction forces resulting from fluid pressure and momentum variations are calculated.
These forces are defined in terms of the fluid properties for the transient
' hydraulic analysis.
Unbalanced forces are calculated for each straight segment of pipe from the pressurizer to the relief tank. The hydraulic analysis includes the effect of water slag discharge. The time histories of these forces are used for the n subsequent structural analysis of the pressurizer safety and relief lines.
The structural model used in the seismic analysis of the safety and relief lines is modified for the valve thrust analysis to represent the safety and relief valve discharge. The time-history hydraulic forces are applied to the piping system lump mass points. The dynamic solution for the valve thrust is obtained by using a modified predictor-corrector-integration technique and normal mode theory.
The time-history solution is performed in subprogram FIXFM. The input to this subprogram consists of the natural frequencies and normal modes, applied forces, and nonlinear elements. The natural frequencies and normal modes for the modified pressurizer safety and relief line dynamic model are determined with the WESTDYN program. The support loads are computed by multiplying the (n) support stiffness matrix and the displacement vector at each support point.
v The time-history displacements of the FIXFM subprogram are used as input to the WESDYN2 subprogram to determine the internal forces, deflections, and stresses at each end of the piping elements.
G WAPWR-S/E 3.9-67 DECEMBER, 1984 7045e:1d
The loading combinations considered in the analysis of the PSARV piping are g ,
2 given.in Tables 3.9-7 through 3.9-10. W '
3.9.3.3.2 Other Pressure Relief Devices on Components The design of pressure-relieving devices can be generally grouped in two categories: open discharge and c4 sed discharge.
A. Open Discharge:
O An open discharge is characterized by a relief or safety valve discharging to the atmosphere or to a vent stack open to the atmosphere.
The design of open discharge valve stations includes the following considerations:
- 1. Stresses in the valve header, the valve inlet piping, and local stresses in the header-to-valve inlet piping junction due to thermal effects, internal pressure, seismic loads, and thrust loads are considered. These stresses are calculated in accordance with the applicable subsections of Section III of the ASME B&PV Code.
- 2. Thrust forces include both pressure and momentum effects.
- 3. Where more than one safety or relief valve is installed on the same pipe run, valve spacing is as specified in ASME Code.
- 4. Where more than one safety or relief valve is installed on the same pipe run, the sequence of openings that induces the maximum stresses is considered as recommended by Regulatory Guide 1.67.
- 5. The minimum moments to be used in stress calculations are those specified in ASME Code.
)
O l WAPWR-S/E 3.9-68 AMENDMENT 2 2045e:1d JANUARY, 1989 l
i L- -- __--_-______________________9
- 6. The' effects of the valve discharge on piping connected to the valve (O header are considered.
- 7. The reaction forces and moments used in stress calculations include p the effects of a dynamic load factor (DLF), or are the maximum instantaneous values obtained from a time-history structural analysis. A dynamic load factor of 2.0 is used, if a dynamic structural analysis is not performed, to determine the dynamic load factor as recommended by Regulatory Guide 1.67.
L B. Closed Discharge A closed discharge system is characterized by piping between the valve and a tank or some other terminal end. Under steady-state conditions, there are no net unbalanced forces. The initial transient response and resulting stresses are determined using either a time-history computer solution or a conservative equivalent static solution. In calculating initial transient forces, pressure and momentum terms are included. Water O slug effects are also included.
3.9.3.4 Component and Piping Supports For statically applied loads, the stress allowables of Appendix F of ASME Code,Section III are used for Code ccmponents.
Dynamic loads for components loaded in the elastic range are calculated using dynamic load factors, time-history analysis, or any other method that assumes A component is assumed to be in the Q elastic behavior of the component.
elastic range if yielding across a section does not occur. The limits of the elastic range are defined in Paragraph F-1323 of Appendix F for Code components. Local yielding due to stress concentration is assumed not to
' affect the validity of the assumptions of elastic behavior. The stress allowables of Appendix F for elastically analyzed components are used for Code components.
i
! 3.9-69 DECEMBER, 1984 WAPWR-S/E ,
2045e:1d
For non-Code components, allowables are based on tests or accepted industry standards comparable to those in Appendix F of ASME Code,Section III.
3.9.3.4.1 ASME Code Class 1 Component Supports The load combinations and allowable stresses for ASME Code Class 1 components l 2 and component supports are given in Tables 3.9-5 and 3.9-3, respectively.
3.9.3.4.1.1 Primary Component Supports Models and Methods The static and dynamic structural analyses employed the matrix method and O
normal mode theory for the solution of lumped" parameter, multimass structural l models. The equipment support structure models are dual purpose, since they are required to represent quantitatively the elastic restraints that the supports impose upon the loop, and to evaluate the individual support member stresses due to the forces imposed upon the supports by the loop.
A description of the supports is found in Subsection 5.4.14 of RESAR-SP/90 PDA Module 4, " Reactor Coolant System." Detailed models of the supports are developed using beam elements and plate elements, where applicable. The reactor vessel supports are modeled using the WECAN computer program. Steam generator and RCP supports are normally modeled as linear or nonlinear springs.
For each operating condition, the loads (obtained from the reactor coolant loop analysis) acting on the support structures are appropriately combined.
i The adequacy of each member of the steam generator supports, RCP supports, and piping restraints for auxiliary connections is verified by solving the stress 2 and interaction equations of ASME Code,Section III, Subsection NF and Appendix F, 1986 or the code of record. The adequacy of the reactor pressure vessel support structure is verified using the WECAN computer program and comparing the resultant stresses to the criteria given in ASME Code,Section III, Subsection NF and Appendix F.
3.9-70 AMENDMENT 2 O
WAPWR-S/E 2045e:1d JANUARY, 1989
t.
(( .
r i
LO l
2
~
tQ v
v
- < 3.9.3.4.2 ASME Code Class 2 and 3 Supports Class 2 and 3 component supports are designed and analyzed for design, normal, upset, emergency, and faulted conditions to the rules and requirements of ASME 2 Section III, Subsection NF, and Appendix F, 1986 or the code of record. The stress limits for Class 2 and 3 component supports for all loading- conditions are defined in Table 3.9-4. The -analyses or test methods and associated l 2 stress or load allowable limits that are used in the evaluation of linear supports for faulted conditions are those defined in Appendix F of the ASME Code. Plate and shell type supports satisfy the faulted condition limits provided i t. Appendix F of the ASME Code, Section 111. Supplementary require- 2 ments are presented in Subsection 3.9.3.2.1 which include stress analysis and evaluation of pump / motor support alignment. Thus, the operability of active O pumps is not compromised by the supports during faulted conditions. The allowable stresses and loading combinations for ASME Code Class 2 and 3 G component and component supports are given in Tables 3.9-4 and 3.9-11. l2 O WAPWR-S/E 3.9-71 AMENDMENT 2 7045e
- 1d JANUARY, 1989
i 3.9.3.4.,3 Snubbers Used as Component Supports The location and size of the snubbers are determined by stress analysis. The location and line of action of a snubber are selected based on the necessity of limiting seismic stresses in the piping and nozzle loads on equipment.
Snubbers are chosen in lieu of rigid supports where restricting thermal growth would induce excessive thermal stresses in the piping or nozzle loads or equipment. The snubbers are constructed to ASME Boiler and Pressure Vessel Code,Section III, Subsection NF standards.
Two types of tests are performed on the snubber.
O A. Production tests are made on every unit:
B. Qualification tests are performed on randomly selected production models to demonstrate the required load performance (load rating).
In the piping system seismic stress analysis, the mechanical snubbers are modeled as stops. Where necessary, the snubber spring rates are incorporated into the analysis.
The recommendations of Regulatory Guide 1.124 applicable to the service limits and loading combinations for Class i linear supports are met as discussed in 2 Table 3.9-4.
A tabulation of snubbers utilized as supports for safety-related systems and componen':s is provided in the Technical Specifications.
Supports for active pumps and valves are included in the overall design and qualification of the component.
Design specifications for snubbers include:
o Seismic requirements.
o Normal environmental parameters.
WAPWR-S/E 3.9-72 AMENDMENT 2 2045e:1d JANUARY, 1989
Accident / post-accident environmental parameters, f o o Full-scale performance test to measure pertinent performance requirements.
(
o Instructions for periodic maintenance (in technical manuals).
3.9.3.5 Design of HVAC Ductwork and Supports O
V Safety Related HVAC ductwork and supports are designed to the following codes:
- 1) High Pressure Duct Construction Standards, Sheet Metal and . Air Conditioning Contractor's National Association (SMACNA, 1975).
- 2) Specification for the Design of Cold Formed Steel Members, American 2
. Iron and Steel Institute. Allowable stresses for ductwork under SSE loads are 1.6 times the normal allowables of AISI.
- 3) Specification for the Design, Fabrication, and Erection of Steel Safety Related Structures for Nuclear Power Flants, American Institute of Steel Con.<truction (AISC-N690), 1984).
- 4) Seismic Analycic of Safety Related Nuclear Structures, American Society of Civil Engineers (ASCE, 1986).
3.9.4 Control Rod Drive Systems Descriptive information on the control rod drive mechanism (CRDM) and gray rod drive mechanism (GRDM) is provided in Section 3.9.4 of RESAR-SP/90 PDA Module 5, " Reactor System."
'Information relating to design specifications and design stresses for the drive mechanisms is provided in Sections 3.1 and 3,.9.3 of this module, and 4.5 I
of RESAR-SP/90 PDA Module 5, " Reactor System." i i
O 3.9-73 AMENDMENT 2 WAPWR-S/E 2045e:1d JANUARY, 1989 i
l l
l l
The control rod drive mechanisms (CRDMs), the gray rod drive mechanisms (GRDM) and their support structures are evaluated for the loading combinations 2 outlined in Table 3.9-5.
A detailed finite-element model of the drive mechanisms and supports is constructed using the WECAN computer program with beam, pipe, and spring 2 elements. Nonlinearities in the structure are represented; these include RPI plate impact, tie rods, and lifting leg clevis /RPV head interface. The time-history motion of the reactor vessel head, obtained from the RPV analysis, is input to the dynamic model. Maximum forces and moments in the drive mechanisms and support structures are then determined.
2 O
O O
WAPWR-S/E 3.9-73a AMENDMENT 2 O
7045e:1d JANUARY, 1989
p The bending moments calculated for the drive mechanisms for the various-V loading conditions are compared with maximum allowable moments determined from a detailed finite-element stress evaluation of the drive mechanisms. Adequacy-of the drive mechanisms support structure is verified by comparing the calculated stresses to the criteria given in ASME Code,Section III, Subsection NF.
Operational transients are listed in Section 3.9.1 of this module.
(3j f
3.9.5 Reactor Pressure Vessel Internals Information on the design arrangements, loading cor.ditions, and design bases is provided in Section 3.9.5 of RESAR-SP/90 PDA Module 5, " Reactor System."
The structural analysis of reactor vessel and internals consider simultaneous application of the time-history loads resulting from the reactor coolant loop mechanical loads and internal hydraulic pressure transients. The vessel is restrained by reactor vessel supports and by the reactor coolant loops with 2 the primary supports of the steam generators and the RCPs.
3.9.5.1 Loading Conditions Following a postulated auxiliary line pipe rupture, the reactor vessel is excited by time-history forces. As previously mentioned, these forces are the combined effect of phenomena: (1) reactor coolant loop mechanical loads and (2) reactor internals hydraulic forces.
The reactor coolant loop mechanical forces are derived from the blowdown reaction force and the jet impingement force at the break location.
2 O WAPWR-S/E 3.9-74 AMENDMENT 2 2045e:1d JANUARY, 1989
I 1
i g The internals reaction forces develop from asymmetric pressure distributions I (,/ inside the reactor vessel. For an auxiliary line break on the vessel inlet j leg, the depressurization wave path is through the broken loop inlet nozzle and into the region between the core barrel and reactor vessel. This region is called the downcomer annulus. The initial waves propagate up, down and
! / around the downcomer annulus and up through the fuel. In the case of an auxiliary line break on the vessel outlet leg, the wave passes through the reactor pressure vessel outlet nozzle and directly into the upper internals i region, depressurizes the core, and enters the downcomer annulus from the 7
Q bottom of the vessel.
Thus, for an outlet leg auxiliary line break, the downcomer annulus is depressurized with much smaller differences in pressure horizontally across the core barrel than for the inlet leg auxiliary line break. For both breaks, the depressurization waves continue their propagation by reflection and translation through the reactor vessel fluid but the initial depressurization wave has the greatest effect on the loads.
The reactor internals hydraulic pressure transients are calculated including the assumption that the structural motion is coupled with the pressure transients. This phenomena has been referred to as hydro-elastic coupling or fluid-structure interaction. The hydraulic analysis considers the fluid-structure interaction of the core barrel by accounting for the deflections of constraining boundaries which are represented by masses and springs. The dynamic response of the core barrel in its beam bending mode responding to blowdown forces compensates for internal pressure variation by increasing the volume of the more highly pressurized regions.
3.9.5.2 Reactor Vessel and Internals Modeling The mathematical model of the reactor pressure vessel is a three-dimensional nonlinear finite-element model which represents the dynamic characteristics of
(]
V the reactor vessel and its internals in the six geometric degrees of freedom.
The model was developed using the WECAN computer :cde. The model consists of WAPWR-S/E 3.9-75 DECEMBER, 1984 2045e:1d
three concentric structural submodels connected by nonlinear impact elements g and stiffness matrices. The first submodel represents the reactor vessel W shell and associated components. The reactor vessel is restrained by the reactor vessel supports and by the attached primary coolant piping. Each reactor vessel support is modeled by a linear horizontal stiffness and a vertical impact element. The attached piping is represented by a stiffness matrix.
The second submodel represents the reactor core barrel, neutron panels, lower support plate, tie plates, and secondary core support components. This submodel is physically located inside the first, and is connected to it by a stiffness matrix at the internals support ledge. Core-barrel-to-vessel-shell impact is represented by nonlinear elements at the core barrel flange, core ,,
barrel nozzle, and lower radial support locations.
The third and innermost submodel represents the upper support plate, guide ;
tubes, support columns, upper and lower core plates, and fuel. The third !
submodel is connected to the first and second by linear stiffness and i nonlinear elements. !
3.9.5.3 Analytical Methods i The time-history effects of the internals loads and loop mechanical loads are combined and applied simultaneously to the appropriate nodes of the mathematical model of the reactor vessel and internals. The analysis is performed by numerically integrating the differential equations of motion to obtain the transient response. The output of the analysis includes the displacements of the reactor vessel and the loads in the reactor vessel supports which are combined with other applicable faulted condition loads and subsequently used to calculate the stresses in the supports. Also, the reactor vessel displacements are applied as a time-history input to the dynamic reactor coolant loop blowdown analysis. The resulting loads and stresses in the piping components and supports include both loop blowdown WAPWR-S/E 3.9-76 DECEMBER, 1984 2045e:1d
loads and' reactor vessel displacements.- Thus, the effect of the vessel f
\ displacements upon loop response and the effect of' loop blowdoun upon vessel displacements'are both evaluated.
I p 3.9.6 Inservice Testing of Pumps and Valves j Inservice testing of ASME Code Class 1, 2, and 3 pumps and valves will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda, as required by 10 CFR 50, Section 50.55a(g). (
g U .
A preservice inspection program (nondestructive examination) and a preservice ,
test program (pumps and valves) for each unit will be prepared. The inservice inspection program and inservice test program will be prepared prior to issue 2
of the plant operating license. These programs will comply with applicable inservice inspection provisions of 10 CFR 50.55a(g). The preservice programs will provide details of areas subject to examination, as well. as the. method and extent of preservice examinations. Inservice programs will detail the areas subject to examination and method, extent, and frequency of examinations after start-up. An inservice inspection program will be outlined in the FDA submittal for NRC review and comment.
3.9.6.1 Inservice Testing of Pumps The pump test program will list all safety-related pumps that are provided 2 with an emergency power source and/or are necessary to shut the plant down safely or mitigate the consequences of an accident. The pump test program will be in accordance with Subsection IWP of the ASME Code,Section XI and I will comply with all applicable portions of 10 CFR 50.55a(g). The hydraulic and mechanical test parameters to be measured or observed will be defined in a separate inservice inspection program.
3.9.6.2 Inservice Testing of Valves l
The valve test program will list all safety-related (i.e., those valves j necessary to shut the plant down safely or mitigate the consequences of an l O WAPWR-S/E 3.9-77 AMENDMENT 2 1 JANUARY, 1989 2045e:1d (
l
)
l I
l 2 accident) valves subject to operational readiness testing and will indicate !
the test parameters to be measured or observed. The test program will conform to the requirements of ASME Code,Section XI, Subsection IWV, to the extent practical, and comply with all applicable portions of 10 CFR 50.55a(g). Test parameters to be measured or observed will be defined in a separate inservice inspection program.
3.9.6.3 Relief Requests i
Relief from the testing requirements of Section XI will be requested when full compliance with requirements of the code is not practical. In such cases, specific information will be provided which identifies the applicable code requirements, justification for the relief request, and the testing method to be used as an alternative.
O O
O WAPWR-S/E 3.9-78 AMENDMENT 2 O l 2045e:1d JANUARY, 1989 4
I 2
Table 3.9-2 PUMP STARTING / STOPPING CON 0ITIONS
' Plant RCS SG Secondary Number of O Condition (*F)/(psig}, (*F)/(psig) Starts /Stoos Operation Cold 70/400 70/0 800 RCS venting Cold 70/400 70/0 200 RCS heatup, cooldown Restart 100/400 100/0 500 Hot functionals RCP stops, starts Hot 567/2235 567/1183 1250 Transients and miscellaneous Hot 567/2235 567/1183 1250 Transients and miscellaneous O
l O
O O 3.9-83 AMENDMENT 2 WAPWR-S/E JANUARY, 1989 7045e:1d i
L ---- --- _----__ _ __ . _ _ _ _ _ _
) e , e , e , e h p
d d F d F d F d
( o N n o N n o N n o na
) CI o CI o CI o CI or
_ c I ni I ni I ni I Fi g
[ VI ot VI ot VI ot VI t a ns P ii P ii P ii P xir) et & ntd & ntd & ntd & nida4
_ nr B ocE B ocE B ocE B ode p oo pp i e i e i e in 3 Et s6 Et s6 Et s6 Et e6 e mp
_ M cb8 Mcb8 M cb8 Mcp8 e9 ou S eu9 S eu9 S eu9 S e p9 s
_ CS ASS 1 ASS 1 ASS 1 ASA1(3 e e e d 5 d d o 2 o o
_ s CI 5 CI CI c I 3 I I
_ I /s VI VI VI I si P , P P I eD & n0 & n5 & n6 v B o2 B o2 B o2 N l e i 5 i 5 i 5 O av Et3 Et3 Et3 I S Vl M c- M c- M c- )
TT a S eB S eB S eB b CR V ASN ASN ASN (
EU e ,
e e e
_ DS d 2 d 0 d 0 d 0 O o 2 o 0 o 0 o p 0
_ CD CI 20 CI 4 CI 4 CI m 4
_ ) N I 30 I 3 I 3 I eu 3 2 VA VI 4 VI VI VI vp 3 P P ,3 P ,
P ,
P i ,
- f j & n1 - & n3 & n4 & nt1 5 9 o g B o2B B o2 B o2 B oc 2 g s i 2N i 2 i 2 i as)2 31 ES p Et3 Et3 Et3 Et sd3 MT m M cd M c- M c- M coae-et SN u S eB n S eB S eB S eNl sB l e AE P ASN a ASN ASN AS(C uN be N ah RO TS OP , , , ,
( FM e e e e O d 3 d d d h AC o 5 o o o p I CI 6 CI CI CI a R1 I 3 I I I r E VI VI VI VI g TS P ,
P P P a I S & n2 & n4 & n5 & nr16 RA g B o5 B o5 B o5 B oa 5 CL n i 6 i 6 i 6 i p36 C i Et3 Et3 Et3 Et 3 S p M c - M c- M c- -
S i S eB S eB S eB M c e 9. B S ee E P ASN ASN ASN ass 3N R
T S , , , ,
e e e e s d 2 d d d h k o 2 o o o p n CI 2 CI CI CI a a I 3 I I I r T VI VI VI VI g
/ P ,
P P P a s & n1 & n3 & n4 & nr15 l B o2 B o2 B o2 B oa 2 e i 2 i 2 i 2 i p 3. 23 s Et3 Et3 Et3 Et s M c- M c- M c - M -
e S eB S eB S eB S ceee 9. B V ASN ASN ASN AS s3N A B C D e
c l l l l i e e e e v v v v) v r d e e ey e e nl l l c l )
Sl a n d
/ e e e) ee cg ee nv nc ct ct ge gi i e i r il iL i v vs ve vu s sr rp rm ra e ee eU eE eF D D s S( S( S(
O m a
r g
nt ee t m e
o t e r ah p mt i
t l f O
t s uo t
e (
8 Sn e o h s t e zg i n f l i o i s O l s
i n t u u
se d
I ao et I ti ua l l I et au Dp N o vc O l a
I S s n n TT .i o o 2 sc CR sh i i s EO it t t 5 ee SP s c c 4 rb P yf el e d5 t EU l o sa u s e3 sl DS a bnsS b P - l O nn urP u B a CD aoi S e 7 S fN eh
) N t0 o l s 2 VA t mn50 m f b a u 3 P f a oI o eo
- f j ot r 1 r r u wS 9 o g n fh o fh l s o g e t = t an l ,
um s vo O 32 ES mi m ni lI MT ee P wsS Sw P i aI .
et SN il P Xt I 1 l e AE l o e1 5 0 am fE n N r E t .e4 t b e ah RO i V a5r e 2. 5 5. 013.
a5
- e3 qu S oMo Si TS OP n I l 4 u2 l 411
( FM i o T u5 s u5 = = As O t C c3 s5 c3= = m5 n onv i
AC A l - e l -
aB pst ern 2do ii e I dr N aB rm R1 eo I CNPP CNCPOPfS t R E mi ad TS rr ne ,
I S op A id4 RA f I mu2 CL rf R rl 1 C ef E ec .
S pa T t n1 S t I ei E S R n n d e R e C o o 2 td T bo i i oi S t D t t 5 enu y c c 4 . h G ad L el e d5 s t s me E s a s e3 t ey t V b ns b P - r rr euo O st E urP u B o ti L S e S fN p rt t nm t5 o p eaa eb E mn2 m f u h rl nu C oI o eo s weu os I r 1 r u pg p V E fh fh l s e p
me me R E
V mi t =
ni t
s an vo i seR et ob I p c n c S T P wsm S w P i l C PS Xt nti el E A e1 5 g aa uS n t d O
V . e5 t hi t w L t
a5r l 4 u1 e 2. 5 2. 0 13 a5 l 411 q am e3 i
d s) e nsd A
f0 V u5 s u5 = = u i si o9 c3 s1 c3= = m5 l ev
/ I l e l - 2do c ro tP aB rm aB pst ern n nt r sS S CNPP CNCPQPfS I I sp e S t e A h L At C O a b
c d
S CI o CI o CI o CI I ni I ni I ni I Fi t VI ot VI ot VI ot VI t n P ii P ii P ii P xi e & ntd & ntd & ntd & nid n B ocE B ocE B ocE B ode o i e i e ie i n p Et s6 Et s6 Et s6 Et e6 m Mcb8 Mcb8 Mcb8 M c p8 o S eu9 S eu9 S eu9 S e p9 ASA1 C ASS 1 ASS 1 ASS 1 e e e e d d d d o o o o CI CI CI CI I 0 I 0 I 0 I 0 s VI 1 VI2 VI 2 VI 2 e P 5 P 5 P
& n3 5 P
& n3 5
v & n3 & n3 l B o- B o- B o- B o-a i D i D i D i D V EtN EtN EtN EtN I M c/ M c/ Mc/ M c/
I S eC S eC S eC S eC I ASN ASN ASN ASN N
O I
T C e e e e E d d d d S o o o o S CI CI CI CI ET I 0 I 0 I 0 I 0 DN VI 0 VI 0 VI 0 VI 0 OE s P 4 P 4 P 4 P 4 CN p & n3 & n3 & n3 & n3 O m B o- B o- B o- B o-4 VP u i D i D i D i D
- PM P EtN EtN EtN EtN 9 &O M c/ M c/ M c/ M c/
BC S eC S eC S eC S eC 3 ASN ASN ASN ASN E3 e M l SD b AN 3 a A 5 T R e 6 e e e O2 d 3 d d d F o o o o S CI ,
CI CI CI I 3 I AS I 2 I VI I A g VI 5 VI 5 VI RL n P 6 P 6 P P EC i & n3 & n3 & n4 & n5 T p B o- B o- B o5 B o5 I i i D i D i 6 i 6 R P EtN EtN Et3 Et3 C M c/ M c/ Mc- M c-S eC S eC S eC S eC S ASN ASN ASN ASN S
E R
T 0 0 0 0 S 2 2 2 2 e 3 e 3 e 3 e 3 s d 3 d 3 d 3 d 3 k o o o o n CI ,
CI ,
CI ,
CI ,
a I O I 0 I 0 I 0 T VI I VI 1 VI 1 VI 1
/ P 3 P 3 P 3 P 3 s & n73 & n3 & n3 & n3 l B o1 - B o- B o- B o-e i 2D i D EtN i D EtN i D s Et3N EtN s M c / M c/ M c/ M c/
e S eCC S eC S eC ASN S eC ASN V ASNN ASN A
l B C D e e c v l l ) l i e e ey e v L v vc v) r d e en ed el ne L) l e g
L e t
S e ac t
/v i ee er el cu ne nv cs ce gL gr ip im i a i i e vu vE vF s sS r( r( r(
e e e e e D D S S S l
Table 3.9-5 2 LOADING COMBINATIONS FOR ASME CLASS 1 COMPONENTS AND SUPPORTS Plant Design / Serv. ice Loading O Classification Level Combination Design Design Design pressure, design r
t temperature, deadweight Normal Service level A Normal condition transients, deadweight 4
Upset Service level B Upset condition transients,-
deadweight, OBE Emergency- Service level C Emergency condition transients, '
deadweight Faulted Service level D Faulted condition transients, deadweight, SSE, pipe rupture loads O
O .
O WAPWR-S/E 3.9-87 AMENDMENT 2 2045e:1d JANUARY, 1989
Table 3.9-6 i O '
DAMPLING VALUES FOR AUXILIARY PIPING SYSTEMS l
Upset Faulted Condition Condition (OBE) (SSE or DBA)
Large diameter Class 1 piping (12 inch or larger nominal diameter) 2 4 Small diameter Class 1 piping (less than 12 inch nominal diameter) 1 2 Large diameter non-Class 1 piping (nominal diameter larger than 12 inches) 2 3 Small diameter non-Class 1 piping (12 inch or smaller nominal diameter) 1 2 0
O WAPWR-S/E 3,9-88 AMENDMENT 2 O
7045e:1d JANUARY, 1989
Table 3.9-7 2
, O LOAD COMBINATIONS AND ACCEPTANCE CRITERIA FOR PRESSURIZER SAFETY AND RELIEF VALVE PIPING - UPSTREAM OF VALVES Plant / System Piping Operating Load Allowable Stress Combination Condition Combination Intensity 1 Normal N 1.5 S, 2 Upset N + OBE + SOT U 1.8 S,/1.5 Sy 3 Emergency N + SOT E
2.25 S,/1.8 S y 4 Faulted N + SSE + SOT p 3.0 S,/2.0 S y 2 O
O l 2
- 1. Taole 3.9-9 contains SOT definitions and other load abbreviations.
- 2. SRSS is to be used for combining dynamic load responses.
- 3. This also applies to pressurizer nozzles and valve support brackets.
O WAPWR-S/E 3.9-89 AMENDMENT 2 2045e:1d JANUARY, 1989
1 i
l 2 Table 3.9-8 LOAD COMBINATIONS AND ACCEPTANCE CRITERIA FOR PRESSURIZER SAFETY AND RELIEF VALVE PIPING SEISMICALLY DESIGNED DOWNSTREAM PORTION (1)(2)
Plant / System Piping Operating Load Allowable Stress Combination Condition Combination Intensity 1 Normal N 1.0 S h 2 Upset N + SOT 1.2 S h U
3 Upset N + OBE + SOT g 1.8 S h 4 Emergency N + SOT 1.8 S h E
5 Faulted N + SSE + SOT p 2.4 S h g
I 2 l 1. Table 3.9-9 contains SOT definitions and other load abbreviations.
- 2. SRSS is to be used for combining dynamic load responses.
WAPWR-S/E 3.9-90 AMENDMENT 2 2045e:1d JANUARY, 1989
.j i
Table 3.9-9 2 DEFINITIONS OF LOAD ABBREVIATIONS
.N' =: Sustained loads during normal plant operation .q.
SOT = System operating transient SOT =. Relief valve discharge transient ,
U S0T = Safety valve discharge transient E
/ SOT p = Max (50T , SOTE ), r transition flow I U
OBE = Operating basis earthquake SSE = Safe shutdown earthquake S = Basic material allowable stress at maximum (hot) temperature l h
S, = Allowable design stress intensity S= Yield strength value y
O l
O O
O i' WAPWR-S/E 3.9-91 AMEN 0 MENT 2 2045e:1d JANUARY, 1989 i
l l \
Table 3.9-10 2
l LOAD COMBINATIONS FOR PRESSURIZER SAFETY AND RELIEF VALVE N0ZZLES AND SUPPORT BRACKETS ASME Code External Section III Load Internal Condition Combinations Pressure Design I DW + OBE Design Design II DW + VO Design R
Normal / Upset I DW + T + OBE Transient Normal / Upset II DW + T + VO R
Transient Normal / Upset III DW + T + V0 Transient 3
Emergency I DW + V0s Transient 2
Emergency II DW+(V0f+OBE) Transient 2
1/2 Faulted I Transient DW+(V0f+SSE)
O O
1 O
WAPWR-S/E 3.9-92 AMENDMENT 2 Oi 2045e:1d JANUARY, 1989 1
l L
Table'3.9-11 2 LOADING COMBINATIONS FOR ASME CODE CLASS 2 AND 3 COMPONENTS AND SUPPORTS FOR THE NPB
- (~
5 Plant Design / Service LeadingCombination(a'b) i Condition Level Requirements Design Design Design pressure, Design. tempera-ture, Deadweight Normal- Service Level A Normal. condition pressure, normal condition metal temperature, deadweight Upset Service Level B Upset condition pressure, upset condition metal temperature, deadweight, OBE Emergency Service Level C. Emergency. condition pressure, emergency condition metal temper-ature, deadweight Faulted Service 'evel D-- Faulted condition pressure, faulted. condition metal
! temperature, deadweight, SSE, pipe rupture i
- a. Temperature is used to determine allowable stress only.
- b. $ u sure, and temperatures are those associated with the respective plant et,. fitions (i .e. , normal, upset, emergency, and faulted), as. noted, for-the component under consideration.
O O i WAPWR-S/E 3.9-93 AMENDMENT 2 2045e:1d JANUARY, 1989
l O parameters (e.g., temperature, humidity, pressure, radiation, etc.) to be employed by Westinghouse for generic qualification purposes are also identified in the specification as applicable.
I 3.11.2.3 Methods and Procedures for Environmental Qualification The basic methodology to be employed by Westinghouse for qualification of safety related electrical equipment is described in Reference 1. Each E0DP O' -
(Reference 2) contains a description of the qualification program plan for that piece of equipment. Qualification may be demonstrated by either type test, operating experience, analysis, or a combination of these methods.
Qualification by analysis alone is not employed by Westinghouse. Analysis is employed to supplement testing or to provide verification that the test results are applicable. Where applicable, the assumptions and models utilized 2 will be described and, with the results of the analysis and conclusions, will be documented in Equipment Qualification Data Package (EQDP).
3.11.3 Qualification Test Results Qualification program results will be provided in the RESAR-SP/90 FDA version.
3.11.4 Loss of Ventilation Refer to the plant specific applicant's safety analysis report for a discussion of loss of ventilation.
3.11.5 Estimated Chemical and Radiation Environment Generic estimates of the radiation dose incurred by equipment during normal operation are provided in Reference 1. The estimated doses and chemical O
WAPWR-S/E 3.11-3 AMENDMENT 2 2044e:1d JANUARY, 1989 i
-h-^ - - - - _ - _ - _ _ __.i_____._____.____._________.
conditions following an accident are defined in Reference 1 and specified in Reference 2 as they apply to the individual equipment qualification program plans.
3.11.6 References
- l. Butterworth, G. and Miller, R. 8., " Methodology for Qualifying Westinghouse WRD Supplied NSSS Safety Related Electrical Equipment,"
WCAP-8587, Revision 6, November 1983.
- 2. " Equipment Qualification Data Packages," WCAP-8587, Supplement 1, latest revision, 9
O O
O 2 l WAPWR-S/E 3,11-4 AMEN 0 MENT 2
_ 2044e:ld JANUARY, 1989
b V
+
TABLE 3.11-1 (Page 1 of 3)
SAFETY RELATED EQUIPMENT Equipment Qualification
("
Eauipment Data Package Reference
- Safety related valve electric motor-operators HE-1 and HE-4 -
I Safety related pilot solenoid valves HE-2 and HE-5 Safety related externally mounted limit switches HE-3 and HE-6 l
Medium pump motors (outside containment) AE-1 Large pump motors (outside containment) AE-2' Canned pump motors (outside containment) AE-3 Pressure transmitters ESE-1 and ESE-2 Dif ferential pressure transmitters ESE-3 and ESE-4 Resistance temperature detectors ESE-6 and ESE-7
'Excore neutron detectors ESE-9 Main control board switch modules ESE-12 Indicators (post-accident monitoring) ,ESE-14 Recorders (post-accident monitoring) ESE-15 Containment pressure sensor ESE-21 Four section excore neutron detector ESE-22 Reactor coolant pump speed sensor ESE-24 Main control board Primary control console ESE-25 Secondary control console Safety center Reactor trip switchgear ESE-26 Nitrogen-16 detector ESE-27 O *** Refer to WCAP-8587, Supplement 1 (Reference 2).
Items listed as "Later" will be addressed in plant specific applicant's FDA.
WAPWR-S/E 3.11-5 AMENDMENT 2 l2 2044e:1d JANUARY, 1989
l TABLE 3.11-1 (Page 2 of 3)
SAFETY RELATED EQUIPMENT Equipment Qualification Eauipment Data Package Reference
- Rod position detector ESE-28 Rod position data cabinet ESE-29 Integrated protection cabinet ESE-30 Integrated logic cabinet ESE-31 Field termination cabinet -
ESE-32 Instrument bus distribution panel ESE-33 and ESE-34 Instrument power supply (static invertor) ESE-35 Source range preamplifier ESE-36 Post-accident monitoring system demultiplexes ESE-37 Control board multiplexer ESE-38 Fiber optic cable ESE-39 Emergency diesel generator Later**
Room coolers Later Safety related fans Later Air cleaning devices Later Packaged A/C units Later Dampers - HVAC Later Emergency feedwater pump turbine Later Electric H2 Recombiner Later Main steam and main feedwater isolation valves Later
- Refer to WCAP-8587, Supplement 1 (Reference 2).
- Items listed as "Later" will be addressed in plant specific applicant's FDA.
2 l b!APWR-S/E 3.11-6 AMENDMENT 2 2044e:ld JANUARY, 1989
TABLE 3.11-1 (Page 3 of 3)
SAFETY RELATED EQUIPMENT t' Equipment Qualification
' Eauipment Data Packace Reference
- Small motors Later Containment butterfly valves Later Electrical distribution switchgear Later Electrical penetrations Later Transformers Later**
Prefabricated cable assemblies Later Load shedder and . emergency load sequencer Later Motor control centers later
.)
AC/DC switchboards Later Batteries and battery racks Later Battery chargers Later Local control stations Later Auxiliary relay racks Later Main control boards Later Radiation monitors / airborne radioactivity Later monitors Control board HVAC chlorine monitor Later O
O
- Refer to WCAP-8587, Supplement 1 (Reference 2).
- Items listed as "Later" will be addressed in plant specific applicant's FDA.
WAPWR-S/E 3.11-7 AMENDMENT 2 l2 JANUARY, 1989 2044e:1d l
l 17.0 QUALITY ASSURANCE' 17.1. QUALIT' Y ASSURANCE DURIN'G DESIGN AND CONSTRUCTION The' Westinghouse : Energy. Systems Business Unit / Nuclear Fuel Business Unit
' Quality Assurance Program.is described in Reference 1.
.17.1.1 References
- 1. " Westinghouse:. Energy Systems Business Unit / Nuclear Fuel Business Unit-Quality: Assurance Plan," WCAP-8370, Revisions'.11, October,1988.
O l
l O
O O
WAPWR-S/E 17.0 -1 AMENDMENT 2 7106e:ld JANUARY 1989
_ = _ - _ _ - - - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ . _ _ _ - _ _ _ - _ - _ _ - _ _ _ _ _ _ - _ _ - _ _ _ _ _ .
f Formal responses to Staff questions 210.35 through-210.72 were originally transmitted, along with proposed text revisions to P.ESAR-SP/90 .PDA. Module 7 " Structural / Equipment Design", in Addendum 6, .NS-NRC-88-3354, ' Johnson (W) to- Rubenstein (NRC),
. dated July 6, 1988. The formal text revisions are provided as part of this amendment.
( VI. MODULE 7 - STRUCTURAL /E0VIPMENT DESIGN
'210.35 'The responses'to questions 210.1, 210.3, 210.6, 210.7, 210.8, 210.9, 210.12, 210.13, and 210.16, are not acceptable.. The classifications in Table 3.2-1 of Module 7 in the Quality Group, Safety Class, . Seismic Category and Q-List columns in all of these responses are - not consistent with the guidelines of Regulatory Guides 1.26. and 1.29. As stated in question 210.1, endorsed ANSI /ANS 51.1-1983 and cannot use the this staff has not 'in document determining the acceptability of the classification of structures, systems and components. Provide a revision to Table 3.2-1 in Module 7 to be consistent with the above staff. position, or provide further justification for
/ non-compliance. In addition, the ASME Section III Code Class should correspond to the guidance provided in REG Guide 1.26.
RESPONSE
Please refer to our criginal response to Staff 0210.1.
Westinghouse believes that the initiative taken to design the SP/90 plant to the latest industry codes and standards, includ-ing ANSI /ANS 51.1, provides additional assurance that this plant L design will operate more safely and with better reliability than current nuclear power plant designs. If this issue is not settled prior to final design submittal, Westinghouse will reexamine the manner in which safely classifications are assigned for systems, components, and structures for the SP/90 plant.
I O
WAPWR-S/E A2-1 AMENDMENT 2 8558e:1d JANUARY 1989
210.36 The responses to questions 210.2, 210.18 and 210.22 are not acceptable. In the staffs' opinion, Footnote S to Table 3.2-1 in Module 7 places too much emphasis on the judgement of individuals to make a decision on the classifications of a support. This could result in an inconsistent and potentially unconservative approach to such classifications. As stated in question 210.2, the staff position on this issue is that 9 supports and hangers should be classified to the same Quality Group and Safety Class as the component that is being
) supported. Revise Table 3.2-1 in Module 7 to modify or delete Footnote S to be consistent with the above position, or provide justification for not doing so.
RESPONSE
Note "S" in Table 3.2-1 of Module 7 has been modified to show compliance with the staff position, or to provide supporting analysis for not doing so.
210.37 The Westinghouse Positions 2(B) and 3 in the response to question 210.19 are not acceptable. Plant specific safety-related instrument tubing programs which did not completely conform to Regulatory Guide .1.151 have been accepted by the G staff for several recently licensed plants. The staff's basis for approving these programs was that the plants had been designed long before the issuance of Reg. Guide 1.151. For future plants, the staff requires that safety-related instrument sensing lines be designed in accordance with Reg. Guide 1.151.
Revise the response to 0210.19 to provide a commitment to this staff position, or provide justification for not doing so.
RESPONSE
Our original response to Staff Question 210.19, which was submitted as part of Amendment 1 to Module 7, has been revised O as shown below:
Position 2(b)
Westinghouse agrees with this regulatory position, and provides the following clarification regarding implementation:
O WAPWR-S/E A2-2 AMENDMENT 2
) b_
1 (3 In general for inbrumentation-channels used to actuate or-
!l U monitor safety related functions or systems, the " sensing line" between the instrument sensor and the Safety Class 1 or 2 (ASME .;
Clas: 1 or 2) process piping or vessel consists of two sections: i y
- 1. A " piping" section, consisting of a short length of 3/4 inch .
Schedule 160 pipe meeting ASME Class 2 and Seismic Category ~ !
I requirements. Connection of this pipe h the' process piping or vessel is made through a passive flow restrictor,
(
! which limits flow to within the capacity of the normal makeup system in the event of line rupture. The pipe terminates in a normally open' isolation valve.
- 2. A " tubing" section, consisting of instrument or capillary tubing of small inside diameter, between the isolation valve and sensor. The rules of Section Ill'of the ASME Code (see NCA-1130) do not apply to instruments, or to permanently sealed fluid filled tubing systems furnished with instruments. Therefore this section is desiv ?d and specified to the level of detail necessary to. meet its functional . requirements and for compatibility with the instrument sensor. In addition, seismic supports for this section adequate to meet Seismic Category 1 requirements are specified.
These arrangements. assure proper operation and provide protection against both single failure and common mode failure.
In the event of a single random failure of any portion of the installation, up to and including complete rupture in the piping section, tubing section or sensor, the resulting outflow of reactor coolant will be limited to within the capacity of the normi " keup system either by the flow restrictor in the piping
- sectme or by the inherent high flow resistance of the tubing O
WAPWR-S/E A2-3 AMENDMENT 2 4 855Be:1d JANUARY 1989
section.- The sensing line seismic supports protect against a
(~
- V] common mode failure of all sensing lines that might otherwise result from a seismic event.
r3 Position (3)
Westinghouse agrees with this licensing position. Any instrument sensing line between the Safety Class 3 (ASME Class
- 3) process piping or vessel and an instrument sensor used to V actuate or monitor safety related functions or systems will be fabricated in the same manner as indicated above for compliance with position 2(b). The only difference is that the " piping" sections of the sensing.line would be fabricated to ASME Class 3 requirements instead of to Class 2 requirements; also Seismic Category 1 requirements would apply.
210.38 The response to question 210.21 is not completely acceptable.
All of the plants which have been licensed by NRC so far have D been allowed to request relief from the ASME Section XI
'd inservice testing rules for a limited number of pumps and valves. These pumps and valves are generally installed in systems in which it is impractical to meet the Section XI rules because of limitations in the system design which make the pump or valve difficult to test without additional design changes.
Therefore, the staff granted many of these requests for relief-because imposition of these rules would have resulted in hardships to the licensee without a compensating increase in the level of safety. The underlying reason for the regulation allowing these reliefs from the code was that the detailed piping system des.igns for all of these plants was completed prior to the time that the staff began to implement the ASME Section XI rules.
A plant such as the WAPWR, for which the final design is not complete, has suTficient lead time available to include provisions for this type of testing in the detailed design of applicable piping systems. Therefore, requests for relief from the applicable ASME Section XI testing rules for pumps and valves will not be granted for the WAPWR. Revise the response ,
O WAPWR-S/E A2-4 AMENDMP.* 2 8558e:1d JANUARY 1963
./7 to 0210.21 to . provide a' commitment that WAPWR piping systems -
.V' will be designed to accommodate the applicable code requirements' for inservice testing of' pumps and valves. However, with regard to subsequent or future ~ code revisions to the- applicable ASME Code. for WAPWR. request -for. relief-. from certain updated code requirements may still be submitted .for. staff review in.
accordance with 10CFR50.55a(g).
RESPONSE
The SP/90- fluid . systems have 'been designed to maximize the f capability to test pumps and valves important to safety in accordance with Section XI of the ASME Code and applicable NRC puidelines, including:
o Full flow test capability- of the Integrated Safeguards System HHSI and RHR/CS pumps and Emergency Feedwater pumps, with the reactor at full power or.during shutdown operations.
o In order to verify full actuation of all valves, full flow is provided- through ' pumped ECCS injection flowpaths, including the EWST suction lines, during normal cooldown/ refueling operations.
o Full flow testing through all valves in the containment. ,
spray flowpaths and the entire EFWS flowpath can be accomplished during plant shutdown.
o All remote operated valves in safety systems can be exercised while the plant is at full power.
o Valve seat leakage test connections are provided for all ISS/RCS boundary check valves. Note, test connections.have been added for the RHR suction and hot leg injection 0, isolation gate valves. These connections will be used to perform periodic leak testing of all pressure isolation 1 O
WAPWR-S/E A2-5 AMENDMENT 2 8558e:1d JANUARY 1989
___ __ _ _ _ _ ___ _o
o-D valves i n' accordance with~ the _ requirements in .the Westinghouse Standard Technical Specifications.
The _only safety related valves which cannot be-readily tested'in' accordance with Section XI inservice tests requirements in 'the.
O- current SP/90 design are the accumulator and core reflood tank (CRT) injection line check = valves. Testing of1 these-. valves-would require the blowdown of the accumulators and CRT's_which can- have adverse impact on plant operations and equipment.
However, Westinghouse will establish a method to perform inservice tests on these valves, while minimizing impact :on=
plant operations and equipment. This test method and any required design modifications will be included. in the FDA submittal.
210.39 The information in Sections 3.6.2.1.1.A, 3.6.2.1.1.B and 3.6.2.1.1.C of Module 7 relative to postulation of pipe ruptures-in high energy ASME Class 1, 2, and 3 piping and non-nuclear, piping is not consistent with current staff positions on this e subject. These position are _in Standard Review Plan,. Section L 3.6.2, Branch Technical Position HEB 3-1, Revision 2 dated June, 1987. Revise these three sections in Module 7 to be consistent with MEB 3-1, Revision 2, or provide justification for not doing so. This revision should not only include changes in threshold stress and cumulative usage factor levels, but should add all of the other MEB 3-1 guidelines which are currently not' included in the WAPWR high energy break location criteria.
RESPONSE
Sections 3.6.2.1.1A, B and C have been revised to include all MEB 3-1 criteria, except for the following:
- 1. Break locations in seismically analyzed B31 piping are based on Equations 12 and 13 of the B31 Power Piping Code. MEB 3-1 recommends using ASME Case 2 Equations for B31 piping.
(1987 Draft of ANS 58.2).
O WAPWR-S/E A2-6 AMENDMENT 2
-8558e:1d JANUARY 1989
_ = - -
L 2. Leak' age crack locations in high and moderate energy ASME or V seismically analyzed B31 piping are based -on alternative stress limits to .those recommended in MEB 3-1. .(1987 Draft
'of ANS 58.2.)
r t - 3. High energy line pipe break locations are' not selected .to produce the greatest effect on separating structures as recommended in MEB.3-1, Section B.1.C(4).
- 4. Guard pipes are pressure tested up to containment- pressure.
MEB 3-1 recommends using the design pressure of the process pipe.
210.40 The information in Section 3.6.2.1.1.D in Module 7 relative to-high energy ASME- Class 2 piping in containment penetration (break exclusion) areas is not~ completely acceptable. In addition to the unacceptable stress levels discussed in 0210.39, the following revisions.to this section are required to qualify for break exclusion in penetration areas:
A 100% volumetric inservice examination of all pipe welds O 1.
should be conducted during each inspection interval defined in IWA-2400,-ASME Code,Section XI.
as
- 2. If there is any ASME Class I high energy piping' in containment penetration areas in which breaks are nct postulated, provide the basis for not postulating breaks in these systems.
Revise Section 3.6.2.1.1.0 in Module 7 to include the above information, or provide justification'for not doing so.
RESPONSE
O There is no ASME Class 1 high energy piping in containment l
penetration areas. A requirement to perform 100% volumetric-in-service examination for the main steam and main feedwater piping in the containment penetration area has been added to O
Section 3.6.2.1.1D.
O WAPWR-S/E A2-7 AMENDMENT 2 8558e:1d JANUARY 1989
A '210.41 -The information in Section 3.6.2.1.1.E in Module 7 ' relative to V ,
" Leak-Before-Break" criteria needs to be updated. The current staff position on this issue.is contained in the Draft- Standard Review Plan 3.6.3, " Leak-Before-Break. Evaluation Procedures"-
dated August, 1987. This draft was recently sent out for 'public.
comments' in the-- Federal Register Notice, Vol. 52, No. 167, p.
A 32626-32633, dated August 28, 1987. This . document contains complete and up to date guidelines which the staff.will use to Q determine the acceptability of plant-specific Leak-Before-Break submittals.
Revise Section 3.6.2.1.1.E in Module 7 to be consistent with the guidelines'in the above document.
RESPONSE
Draft Standard Review Plan (SRP), Section 3.6.3 " Leak-Before-Break Evaluation Procedures" is currently being. revised by the NRC Staff in' response to industry comments. 'A revised' Section 3.6.2.1.1.E -for Module 7 of RESAR-SP/90 will be prepared after Westinghouse review of the final version of SRP 3.6.3, ' prior to the final SER if applicable.
O ~210.42 Section 3.6.2.1.2, " Types of Breaks / Cracks Postulated" Module 7 does not discuss criteria for postulating cracks in in high-energy piping. Revise this section to provide this information. Acceptable guidelines can be found in Standard Review Plan 3.6.2, MEB 3-1, Section B.1.e, Revision 2 dated June 1987. Piping covered by Sections 3.6.2.1.1.0 and 3.6.2.1.1.E in Module '7. are excluded from these guidelines.
RESPONSE
Section 3.6.2.1.2 has been revised to include postulated cracks in high energy piping in accordance with the 1987 draft of ANS 58.2.
210.43 In Section 3.6.2.1.2.3 of Module 7, the threshold stress values for postulating cracks in moderate energy piping does not agree O with the current Specifically, in Section staff position which is 3.6.2.1.2.3.A and 3.6.2.1.2.3.C.(2),
in SRP 3.6.2.
0.45 should be 0.40 and in Section 3.6.2.1.2.3.C.(1), 1.4 Sm O
WAPWR-S/E A2-8 AMENDMENT 2 8558e:1d JANUARY 1989
should be 1.2 Sm. -Revise this section to be' consistent with the O staff position, or provide justification for~ deviation :from the-guidelines in SRP 3.6.2.
RESPONSE
Section 3.6.2.1.2.3 has been~ revised tol incorporate the moderate energy crack location criteria of the 1987 draft of ANS 58.2.
.210.44 Section 3.6.2.1.2.1. A.1 and 3.6.2.1.2.1. A.2 in Module 7 contain O an- apparent typographical error. To be consistent ~w ith.the staff position on this issue, "circumferential break" should be
" longitudinal break" in paragraph 1 and " longitudinal break" should be "circumferential break" in paragraph 2.
RESPONSE
Section 3.6.2.1.2.1.A.1 and 2 have been revised to correct the-typographical error.
210.45 Section 3.6.2.2.1 in Module.7 references WCAP 10221,'" Simplified
- Pipe . Whip Analysis and Restraint Design Procedures" for analytical methods used in calculating jet thrust loads subsequent to a pipe rupture. The staff has no record of receiving this report and has not reviewed it. Standard Review Plan 3.6.2 and ANS 58.2, " Design Basis for Protection of Light Water Nuclear Power. Plants Against the Effects of Postulated Pipe Breaks" contain acceptable guidelines for. these analytical methods. Revise Section 3.6.2.2.1 to either reference the above-documents .or provide a description of the methodology which will be used to calculate jet thrust loads.
RESPONSE
O Section 3.6.2.2.1 has been revised to refer to ANS 58.2 as the basis for calculating jet thrust loads.
210.46 In addition to the informationload in Section 3.6.2.3.4.1 of provide the loads, Module 7, combinations, and stress y limits that will be used in the design of pipe rupture restraints. Include a discussion of the design methods O )
WAPWR-S/E A2-9 AMENDMENT 2 8558e:1d JANUARY 1989 1
J
(
1 pg applicable 'to the auxiliary steel .used 'to support the pipe
' V' rupture restraint. Provide . assurance' that the pipe rupture-restraint and supporting structure cannot fail'during a seismic event. i Provide the design criteria which will be used for pipe rupture restraints that also support piping, if this criteria is'dif-A ferent from that discussed in Sections 3.9.3.1.1 and 3.9.3.1.2 yj of Module 7. .
RESPONSE: i p)
( The pipe rupture restraints and any auxiliary steel that the restraints are attached to will be designed in accordance with ANSI /AISC N690, Nucisar. Facilities - Steel Safety-Related Structures for Design, Fabrication and Erection. The. Dynamic Load Factor (DLF) that will be used for the design of the restraints and auxiliary steel will be based on WCAP-10221.
Where supports are attached to restraint structures, the restraint structure will not be designed to ASME III, Subsection NF. The design will' remain based on ANSI /AISC N690. The design of the restraints for pipe rupture will be based on the load combination, Equations 10 and 11 of Table Q 1.5.7.1 of ANSI /AISC N690. The loads considered are as specified by the two equa-tions. Note, the equations do include an earthquake loading.
Therefore, a seismic event will be considered for design.
Subsection 3.6.2.3.4.1 has been revised to include ANSI /AISC N690 as a reference.
Section 3.7.3.1 of Module 7 references the ASCE Seismic Analysis D) 210.47 Standard Committee " Standard for the Seismic Analysis of Safety-Related Nuclear Structures," May 1984 Draft for methodology used in both time-history solutions and response spectrum analyses of subsystems. This ASCE standard (including the September 1986 l
Edition) has not been accepted by the staff because some of the I \ basic assumptions and options in the standard do not agree with V current staff positions. Revise Section 3.7.3.1 in Module 7 to be consistent with applicable guidelines in Standard Review Plan, Sections 3.7.1, 3.7.2, and 3.9.2, or provide justification for deviations from these guidelines.
O WAPWR-5/E A2-10 AMENDMENT 2 8558e:1d JANUARY 1989 l
l
q RESPONSE:
D Section 3.7.3.1 of Module 7 references the ASCE seismic analysis standard, " Seismic Analysis of Safety-Related Nuclear Structures," May 1984 draft for methodology used in both O time-history solutions and response- spectrum analysis of subsystems. Since this standard has not been accepted by the USNRC, the USNRC suggested to revise the section 3.7.3.1 to be consistent with applicable guidelines in Standard Review Plan (SRP) Sections 3.7.1, 3.7.2, and 3.9.2 or provide justification 9
for deviations from these guidelines.
The following discussion is provided to demonstrate that the.
applicable portions of the September 1986 edition of the ASCE Standard are consistent with the above mentioned- applicable SRP sections. !
First of all, the SRP Section 3.7.1 is not relevant because it
- addresses seismic design parameters such as design ground motion, design response spectra, design time history, critical damping values, etc. Section 3.7.3.1 of Module 7 addresses time-history solutions and response spectrum analyses of subsystems for which the SRP Section 3.7.3 provides the USNRC acceptance criteria. This section in subsection II.1.a refers- ' .
the reader to the subsuction 11.1 of SRP Section 3.7.2. The SRP Section 3.9.2 is concerned with Dynamic Testing and Analysis of Systems, Components and Equipment. Thus, the applicable USNRC guidelines on time history solutions and res;:ense spectrum analyses of subsystems are really contained in Subsection II.1.a of SRP Section 3.7.2.
I O The requirements of time-history method and the response U spectrum method of the ASCE standard are contained in Sections 3.2.2 and 3.2.3 of this standard. These requirements are more O
U WAPWR-S/E A2-11 AMEN 0 MENT 2 8558e:1d JANUARY 1989
1 detailed. and represent advancements in technology and k refinements in old procedures'to make them mc,re realistic. In a.
comparative . review of these requirements with those of the SRP Section 3.7.2.II.1.a, it is noted that' effects of soil-structure interaction need to be included explicitly in the system analysis but not explicitly in the subsystem analysis. Thus, this consideration which is not listed in' Sections 3.2.2 and 3.2.3 of the ASCE Standard is not directly relevant. In fact, this review shows that the ASCE standard provides more Q requirements (such as convergence checks for accuracy of eigensolution, solution stability and convergence requirements for integration time-step size, implementation methods for number of modes for solution accuracy, residual mode effects etc.) than the applicable guidelines of the SRP 3.7.2.II.1.a.
Requirements for combination of modal and component response contained in the ASCE Standard Section 3.2.7 are somewhat different from those of the Regulatory Guide (R.G.) 1.92. The basic requirements for combination of component responses which j employ the square root sum-of-squares are the same as those of R.G. 1.92; however, the ASCE Standard provide an alternative of 100-40-40 percent rule which is, in general, more conservative than the SRSS rule. The discunion and justification for somewhat different modal combination rule are provided in response to Question 210.50.
Based on above discussion, it is concluded that the requirements of the ASCE standard for time-history and response spectra analysis are consistent with the corresponding requirements in the SRP Section 3.7.2.II.1.a. Section 3.7.5 of the RESAR will j be revised to change Reference 5 to the September, 1986, Edition of the ASCE Standard.
O !
O l WAPWR-S/E A2-12 AMENDMENT 2 8558e:1d JANUARY 1989 L_________-_. 1
i i
Ov 210.48 Section 3.7.3.2 of Module 7 states that the Operating Basis Earthquake.(OBE) is assumed to occur five times over the life of the plant. A time history study was conducted by Westinghouse to . arrive at a realistic number of maximum stress cycles per OBE' occurrence. As a result of this study, Westinghouse concluded that 10 maximum stress cycles for flexible equipment (natural
(' frequencies less than 33 Hz) and 5 maximum stress cycles for
't rigid equipment (natural frequencies greater than 33 Hz)'for each OBE occurrence should be used for fatigue evaluation of WAPWR ASME Class 1 systems and components. However, Table 3.9-1, " Summary of Reactor Coolant System Design ' Transients" in Module 7 lists 50 cycles for OBE.
It is the staff's understanding that Westinghouse generally uses only the 50 stress cycles criterion for all ASME Class 1 systems and components, which is consistent with the guidelines in Standard Review Plan, Section 3.9.2.II.2.b. Either delete the reference to the use of 5 stress cycles per event for rigid equipment or provide additional justification for using only 25 cycles instead of 50 cycles in the fatigue analysis for this equipment.
RESPONSE
Section 3.7.3.2 has been revised to indicate that 10 maximum stress cycles are used for all ASME Class 1 systems and components for each 5 OBE occurrences.
210.49 Section 3.7.3.3.A of Module 7 states that for the analysis of main piping runs, branch connections are decoupled from the main runs when the ratio of the branch to run section moduli is equal to or less than 1/16, or the ratio of the branch to run moment of inertia is 1/50. It further states that the boundary of each decoupled model contains a sufficiently long region of common overlap to other models. Provide more information relative to the basis and justification for each of these assumptions.
O V RESPONSE:
- 1. Dynamic Analysis of Main Piping Runs: !
O V Branch piping runs cannot always be included in the analysis of main piping runs, due to: a) the extent and number of branch O
WAPWR-S/E A2-13 AMENDMENT 2 8558e:1d JANUARY 1989
~
p runs may be-l'arge and beyond the capacity of. piping analysis computer programs, :and 'b) 'the designi sequence for piping and piping.' support proceeds from large: diameter to small diameter piping,. so that design. information- for.. branches may .not be available when the main run is being ' analyzed. The decoupling
-) g criteria chosen for. RESAR-SP/901 provides a reasonable design
' basis.for the main piping runs. Experience has. shown-- that the small diameter branch ~~does not. significantly affect the main run, when the supports on the branch piping are located to avoid' excessive restraint'of the main run.
- 2. Boundary of Decoupled Model:
The entire piping run from anchor to anchor cannot always be-included:in the dynamic analysis model, due to the limited capacity' of piping computer programs. .The incorporation -of additional anchors to reduce the size of the model is undesirable, since it often results in higher thermal stresses.
In these cases,-the entire piping system is subdivided into two or more portions. The dynamic .model for each portion extends-into the other portions. These. extensions. are called- the overlap regions. When the overlap regions'.are sufficiently long, a reasonable design basis is achieved. The' general criteria is 'that- the overlap region contain three or more pipe supports in each- of the
~
X, Y, and Z coordinate , directions.
Revisions to Section 3.7.3.3.A have been made which are consistent with WRC Bulletin 300 and recent industry practice.
O 210.50 In the first paragraph of Section 3.7.3.7, Module 7, an optional method of algebraic combination of modes with closely spaced 1
frequencies is stated. NUREG-1061, Vol. 4 " Report of the U.S. j NRC Piping Review Committee" is referenced as the basis for this !
option. It should be noted that NUREG recommendations should I not always be interpreted as staff positions unless an explicit I'n this
! statement to this effect is documented in the NUREG.
case, the NUREG-1061 recommendations of algebraic combination j l
O i l I
WAPWR-S/E A2-14 AMENDMENT 2 8558e:1d JANUARY 1989
~ . . . .
p: ,
has' not yet been accepted by the , staff. The current' staff.
V : position'as stated in Regulatory Guide 1.92, " Combining Modal.
Responses and:. Spatial' Components in Seismic Response Analysis" has three options in- combining. closely spaced: modes, but the=
proposed algebraic combination - is not one- of. them. -Revise Section.:3.7.3.7 to delete the proposed optional method,- or provide justification for its use.
RESPONSE
The following provides the required justification ^ for' the
!O. optional method of algebraic combination of modes with closely-spaced frequencies, as noted in the first paragraph of Section 3.7.3.7 of Module 7.
When the modal response spectra method is used to perform.
seismic analysis, effectively three separate analyses - one for each of the three earthquake components -
are performed. In each analysis, mode-by-mode maximum response i s' first calculated. Since the representation of the time history.
() excitations by response spectra does not retain time phasing, character of excitations, a modal combination procedure or rule is required for combination of mode-by-mode maximum response to obtain a' resultant response due to excitation by a given earth-
~
quake component. There -are many modal combination procedures available in the literature. Reference 1 contains interesting information on nine different modal combination procedures. Two modes are defined as closely spaced modes if their frequencies differ from each other by 10 percent or less of the lower frequency. In Reference 1, the phrases, " frequencies within 10 percent of each other" and " frequencies within 10 percent of lowest mode in group" should be interpreted to mean that the.
associated modes are closely spaced as defined above.
O V The square-root-sum-of-squares (SRSS) modal combination is based
)
upon the assumption of random phasing of maximum mode-by-mode )
D WAPWR-S/E A2-15 AMENDMENT 2 8558e:1d JANUARY 1989
___-_-_______-__ - -___-_ a
l O responses at the time of peak combined response. This
\') assumption works well (see Reference 2).fornon-closely-spaced modes except at higher frequencies (usually greater than or equal to 33 cycles /second as per Reference 3) where modes are reasonably in phase. Thus, the SRSS combination is deficient
\ for' closely-spaced modes and modes with higher frequencies. It l- is a common practice to uncouple a light secondary system - from the primary system in the_ seismic design of both systems.
,Q Alternatively, e coupled approach may' be used to avoid U unnecessary conservatism. When the' secondary system is at a near resonance b th the primary system, coupling produces closely spaced <. odes. In this situation, the SRSS method of modal combination may produce conservative results for the secondary system (see Reference 4). Just as the SRSS method sometimes over predicts the seismic loads, it sometimes under predicts the loads when the closely-spaced modes are encountered (see References 2 and 5). This is because of the q particular phasing properties.of mode-by-mode responses.
NJ Numerous investigations (References 1, 2, 3, 5 and 8) have been carried out to study various modal combination procedures. Out of these studies, there are two procedures that emerge.very favorably. Thesa are: Rosenblueth and Elorduy's Double Sum Method (Reference 6) for combination of mode-by-mode responses from lower frequency (usually < 33Hz) modes, and algebraic sum for combination of mode-by-mode responses from higher frequency (usually > 33 Hz) modes. The Rosenblueth and Elorduy's Double Sum Method is theoretically based on random vibration theory.
The U.S. NRC double sum method in Reference 7 and the i Rosenblueth and Elorduy's double sum method are identical in all l respects except that the U.S. NRC double sum method uses absolute signs while the Rosenblueth and Elorduy's double sum method retains algebraic signs. There is no theoretical or empirical justification for the U.S. NRC double sum method O ;
WAPWR-S/E A2-16 AMENDMENT 2 ;
8558e:1d JANUARY 1989
l>
except that it is always more conservative'th'an the Rosenblueth- l and Elorduy's double sum method. The Rosenblueth and Elorduy's' method is therefore selected for use in computation of realistic responses and also for elimination of over-conservatism and.
, under-conservatism associated with SRSS. and other modal-combination methods. LThe only apparent problem with the Rosenblueth and Elorduy's double sum method is the increased.
computational time. associated with all the. cross product terms
.for dynamic models with more than-about 10 modes. To eliminate this problem, a minor approximation is made .by including cross product terms only when 'the modes are closely-spaced.
~
Details of this method are given below.
In order to account for-the effects of any closely spaced-_ modes that may be-' present, the resultant response of-interest for design purposes due to excitation by- a given'- earthquake component is obtained by the following modified square-p root-sum-of-squtres (SRSS) combination of the corresponding mode-by-mode maximum responses' due to the earthquake component under consideration. In equation form, the modified SRSS combination, which degenerates to the regular SRSS combination in absence of closely spaced modes, is represented by:
N S N -1 N. 1/2 3
R
=[I R ik
- 2 .I I I R$ ,Rin'in) k=1 J=1t=M) n=t+1 where R = value of combined response for with' direction excitation component R = response for direction i, mode k ik
= response for direction i, mode 1 R)$
O ,
WAPWR-S/E A2-17 AMENDMENT 2 855Be:1d JANUARY 1989
R = response for direction i, mode n in-N = total number of modes having frequencies lower than the zero period-acceleration (ZPA). frequency -or cut-off.
- frequency (usually <, 33 Hz)
S = number of groups of closely spaced modes. The groups of closely spaced modes are formed such -that the.
difference between the frequencies of the last mode and Q.
'b/ the first' mode in the group does not exceed 10 percent of the lower frequency. Groups are formed starting from the lowest frequency and working towards successively higher frequencies in such a way that no one frequency is to be in more than one group.
M = lowest modal number associated with group j of closely 3
spaced modes
= highest modal number associated with group j of closely N) spaced modes e = coupling factor defined below in w,- , 2 -1 c
an
- Il*(sd+w"n'w)3 e e s n
wg = wg [1 - (s g)2)1/2 2 ,
s,' =s+qt g
d O l U wg = frequency of closely spaced mode t (rad /sec)
O WAPWR-S/E A2-18 AMENDMENT 2 8558e:Id JANUARY 1989
t:
-A B = fraction of critical. damping in closely spaced mode t.
If ASME Code Case 411 damping is employed, Bg should be 5% for 0 to 10 HZ, 2% for 20 HZ and over and should be linearly interpolated between 10 HZ and 20 HZ.
~ td = duration 'of the earthquake (sec). This parameter is z i
plant-specific and its value'will be obtained from the ,
Architect-Engineer supplied or other in-house seismic.
information-for a given plant. In absence of other V specific information, a value of 10 seconds can be used for this parameter.
It should be noted that the definition of parameters R ik'-
R$ ), and R does not require absolute values for these '
in parameters. In other words, these parameters which are not absolutized retain. algebraic signs. It should be noted that the above procedure was provided to the U.S. NRC and is contained in Appendix C of Reference 1. . No changes to RESAR-SP/90 are being made in this area.
REFERENCES
- 1. Report of the U. S. Nuclear Regulatory Commission Piping Review Committee, " Evaluation of Other Dynamic Loads and Load Combinations," NUREG-1061, VOL. 4, December 1984.
- 2. A. K. Singh, S. L. Chu, and S. Singh, " Influence of Closely Spaced Modes in Response Spectrum Method of Analysis," ASCE Specialty Conference on Structural Design of Nuclear Plant Facilities, Vol. II, Chicago, December-1973, pp. 479-498.
O WAPWR-S/E A2-19 AMENDMENT 2 8558e:1d JANUARY 1989
C
- 3. K. M. ~Vashi, " Computation of Seismic Response from Higher Frequency Modes," Transactions of the ASME, Journal of' Pressure Vessel Technology, Vol. 103, February 1981, pp.
16-19.
O 4. N. . C. Tsai, " Spectral . Response Analysis of a Coupled System," PVP-Vol. 65, " Dynamic and Seismic Analysis-of Systems and Components," ASME Pressure Vessels and Piping Conference, Florida, June-July 1982, pp. 159-165..
- 5. J. A. M. Boulet, and T. G. Carley, " Response Spectrum.
Analysis of Coupled Structural Response to a Three Component Seismic Disturbance," Proceedings of SMIRT-4 Conference, Vol. K(a), San Francisco, 1977, Paper K I/5, pp. 1-12.
- 6. E. Rosenblueth, and J. Elorduy, " Responses of Linear Systems to Certain Transient Disturbances," Proceedings of q
b 4th World Conference on Earthquake Engineering, Santiago, Chile, 1969.
- 7. V. S. Nuclear Regulatory Commission, Office of Standards Development, Regulatory Guide 1.92, " Combining Modal Responses and Spatial Components in Seismic Response Analysis," Revision 1, February 1976.
- 8. T. S. Aziz, " Coupling Effects for Secondary Systems in Nuclear Power Plants," PVP Vol. 65, " Dynamic and Seismic Analysis of Systems and Components," ASME Pressure Vessels and Piping Conference, Florida, June-July 1982, pp. 31-44.
210.51 Sections 3.7.3.7.A, 3.7.3.7.8, 3.7.3.7.C, 3.7.3.7.D and 3.7.3.9.A.2 of Module 7 discusses several options which will be available in the seismic subsystem analyses of the WAPWR. The bases for these options are provided in References 10, 11, 12, O
WAPWR-S/E A2-20 AMENDMENT 2 8558e:1d JANUARY 1989
('S and 13 of Section 3.7.of Module 7. If Westinghouse intends to i V use these options, the staff will be required to rev4ew the applicable references. Therefore, submit. the referenced .
documents for staff review.
RESPONSE
0 'The requested references are open literature, available to the i
public, and have been provided to the Staff Project Manager assigned to thic review.
O U 210.52 The last sentence in Section 3.7.3.9.B of Module 7 states that the effect of relative seismic anchor displacements are obtained by either using the worst combination of peak displacements or by proper ' representation of the relative phasing characteristics associated with different support inputs. Provide more details relative to how " proper representation" is obtained. Identify and justify any deviations from the guidelines in Standard-Review Plan, Section 3.9.2.II.2.g.
RESPONSE
f s The methods used. to calculate the system response, due to relative seismic support displacement, comply with SRP Section 3.9.2.11.2.g with one exception. In the case of support loca-tions in the same supporting structure, the proper representa-tion is frequently found to be in phase motion of the support location. This occurs when the seismic displacements of the supporting structure are dominated by modes with in phase displacements, such as, rigid body modes and low frequency p modes. Examples are: large concrete supporting structures, such as the containment shell and the containment interior buildings.
210.53 Combination of inertial responses and seismic anchor movements (9 of multiply-supported items by the SRSS method as stated in
() Section 3.7.3.9.C of Module 7 is not acceptable. Revise this section to provide a commitment to combine these items by absolute sum as recommended in Standard Review Plan, Section i 3.9.2.II.2.g, or provide justification for use of this method. I O 1 i
WAPWR-S/E A2-21 AMENDMENT 2 8558e:1d JANUARY 1989
f Il RESPONSE: J t/ )
The following provides the required justification for. !
combination of inertial responses and responses due to seismic O anchor movements (SAM) of multiply-supported items by the SRSS U method as stated in Section 3.7.3.9.c of Module 7. No changes i to RESAR-SP/90 are being made in this' area.
O Brookhaven National Laboratory (BNL) performed extensive studies b as reported in NUREG/CR 3811 on the subject of computation of response due to SAM and on the combination of inertial responses and responses due to SAM. In the computation of responses due to SAM, the time-history method was used. Because of retention of proper phasing among various motions due to SAM, the time-history method would provide lower bound values of responses due to SAM. Subsequently, the inertial responses were combined with the lower bound responses due to SAM using absolute sum combination and SRSS method as well. These nv combined responses were then compared with combined responses obtained using the time-history method. From this comparison, it is observed that the SRSS combination between the dynamic (inertial) and static components of response due to SAM provides a conservative estimate of total response for all the dynamic cases. It was also found that the absolute sum combination between the inertial responses and the static responses due to SAM, as recommended in the SRP, yields very conservative estimates of total response. It should be further noted that i these studies used lower bound estimates of responses due to SAM and showed that the SRSS combination provides conservative results. In actuality, the computation of responses due to SAM would, in general, produce conservative results due to conserva-( tive treatment of phasing. Thus, the SRSS combination based on such conservative estimates of responses due to SAM would produce even more conservative estimates of combined. responses.
O WAPWR-S/E A2-22 AMENDMENT 2 8558e:1d JANUARY 1989
q In another study, reported in'NUREG 1061, volume 4, it is noted O that the 3AM induced responses and inertial responses are not phase uncorrelated. Therefore, the SRSS combination cannot be justified on theoretical grounds. However, peak inertial responses and peak SAM induced responses would be highly L/ unlikely to occur concurrently. .It has been shown that the SRSS combination of inertial responses and_ SAM induced responses produces results that have a 96.4% non-exceedance probability l which is considerably higher than the non-exceedance probability
. () of mean plus-one-standard deviation design response spectra in Regulatory Guide 1.60. Thus, the SRSS combination is more than adequate.
Based on above evidence, studies and discussion, it is more than technically adequate and acceptable to use SRSS combination of )
inertial responses and SAM induced responses. The SRSS combination produces combined responses that are ct:nservative in comparison to the combined responses from time-his'.ory methods.
Further, the abrolute sum combination which would generally be i excessively conservative is not necessary or warranted.
210.54 The information in Section 3.7.3.13 of Module 7 relative to the interaction of other piping .5ystems with Seismic Category 1 piping is too general to be completely acceptable. Revise this section to be consistent with the guidelines in Standard Review~
Plan, Section 3.9.2.II.2.k. In addition, revise applicable portions of Section 3.2.1.1 of Module 7 to be consistent with the revised Section 3.7.3.13.
RESPONSE
Section 3.7.3.13 has been revised to describe the methodology that will be used to evaluate non-seismic piping in close proximity to seismic piping.
210.55 The information in Figure 3.7-8 of Module 7 relative to damping values is not completely acceptable. Specifically, curves (1)
I O
WAPWR-S/E A2-23 AMENDMENT 2 l 8558e:1d JANUARY 1989 l
l l
l L
i O'- and (2) exceed damping values currently acceptable to the
( staff. Curve (3) is consistent with ASME Code Case N-411 and is acceptable for all.=ASME Class 1, 2, .and 3 piping provided a commitment is made to conform to the conditions specified by the staff for using Code Case N-411. These conditions are outlined in Regulatory Guide 1.84, Revision.24, dated June 1986.- Revise f3- ' Figure 3.7-8 to eliminate curves (1) and (2) and to provide a
- commitment to the conditions of REG Guide 1.84, Rev. 24 for use Q of curve (3). .
In addition, revise the portion of the.
discussion in Sections 3.9.3.1.1.1.C and 3.9.3.1.3 of Module 7 '
i which is applicable to this issue. !
RESPONSE
[V]
Westinghouse application of ASME Code Case N-411 damping values is consistent with all of the conditions in Regulatory Guide 1.84, Revision 24, except one condition. Composite damping for piping system models with flexible inline building mounted ,
equipment- is not used. See revised Section 3.9.3.1.1.2B for justification.
A 210.56 The information in Section 3,9.1.2.1 of Module 7 relative to V computer codes which will be used in dynamic and static analyses of seismic Category 1 components and equipment is not complete.
Provide a discussion in this section of the methods used to verify the FATCON and WESAN programs and any other applicable program which is not listed. This verification should be in accordance with the guidelines in Standard Review Plan, Section 3.9.1, i.e., a comparison of the results from each program with either (1) hand calculations, (2) published analytical results, (3) acceptable experimental results, (4) results from a similar program which has been accepted by the staff, or (5) the bench-mark problems in NUREG/CR-1677, " Piping Benchmark Problems."
/^ The WECAN program, which is also listed in Section 3.9.1.2.1, t ( has received only a partial review by the staff. The informa- i tion in WCAP-8929, " Benchmark Problem Solutions Employed for ,
Verification of the WECAN Computer Program" which is applicable 1 to the dynamic analysis of linear and nonlinear elastic
. beam-type structures is the only portion of WECAN which has been i p accepted by the staff. However, WECAN contains many other features and extensive capabilities such as plate and shell U structures, elastic plastic and creep deformation and heat transfer analysis. None of these features have been l
O V
WAPWR-S/E A2-24 AMENDMENT 2 8558e:1d JANUARY 1989
_.-x_.-----...-1. -
1
.. \
( independently verified, although .their theoretical bases are i consistent with present state-of-the-art. If any. of these A additional- WECAN features will be used in the design of the WAPWR, discuss the application of the feature in -Section 3.9.1.2.1 of Module 7 and the staff will review this.information on a case-by-case basis.
r.
I RESPONSE:
In response to Staff Question 220.7, W provided (in Adde.dum #2, n dated January, 1988) detailed descriptions -
including veri- ;
h fication procedures, for the WECAN, DEBLIN2, programs. RESAR-SP/90 PDA Module 7, and ASHSD computer
" Structural / Equipment Design" has 1 been modified (Appendices 3.7 and 3.8) .i n this amendment to include these detailed descriptions. Similar detailed information for the FATCON and WESAN programs will be provided prior to Staff submittal of a final SER.
210.57 Section' 3.9.1.4, " Consideration for the Evaluation of the Faulted Condition" in Module 7 references Section 3.9.4 of g Module 7. Section 3.9.4 only addresses analyses of Control Rod
't Drive Systems. The intended reference appears to be Section
- 3.9.3.4, " Component and Piping Supports." However, Section 3.9.3.4 does not completely address the subject matter of Section 3.9.1.4 because it only discusses Service Level D analyses of component supports in the elastic range. If an elastic plastic method of analysis will be used to evaluate the design of any safety-related system, component or equipment for which ASME Level D Service Limits have been specified, provide the information requested in' Standard Review Plan, Section 3.9.1.11.4 and 3.9.1.111.4. This information should be in Section 3.9.1.4 of Module 7.
RESPONSE
nV If an elastic plastic method of analysis is used to qualify ASME Code components to Level D service limits, the following will be documented:
- 1. Basis for validity of stress-strain relationship.
1 WAPWR-S/E A2-25 AMENDMENT 2 855Be:1d JANUARY 1989
p ,
2.- Basis Lfor validity' of ul timate strength values'at service temperature. j d
- 3. Provide assurance that the displacements 'and deformations' do~
not' violate the assumption of the corresponding. system j l analysis. l Editorial: changes have been made to sections 3.9.1, ~3.9.3 and a 3.9.5 as well as Tables 3.9-3, 3.9-5 and 3.9-6.
Q 210.58 The information in -Section 3.9.2.1 of Module 7 provides a
'i general, discussion of the proposed piping preoperational- test.
program for- the WAPWR.. The staff's current position on this issae is that for Preliminary Design Approval, a commitment is required to develop a test program for Final Design Approval-which will utilize testing procedures _and acceptance criteria in the- Draft ANSI /ASME OM-3, " Requirements for Preoperational and Initial Start-Up_ Vibration Testing of Nuclear . Power Plant Systems." The staff is participating in the development of this standard. By the time a WAPWR Final. Design Approval is submitted, the staff plans to have OM-3 referenced in Standard Review Plan, Section 3.9.2. -For the past several years, the O- staff has been using the acceptance criteria from the OM-3 Draft-dated May 1985 as a guide in its review of piping preoperational
-test programs for near-term operating license plants. Revise Section 3.9.2.1 to provide a commitment to the staff position j described above, or provide justification for deviation from '
this position.
RESPONSE
Subsection 3.9.2.1 of Module 7 has been modified to include a commitment to implement a pre-operational test program based on for b) draft ASME/ ANSI Preoperational OH-1987, Part 3 " Requirements and Initial Start-up Vibration Testing of Nuclear l Power Plant Systems."
safety-related O 210.59 It is the staff's position that all essential instrumentation lines should be included in the vibration moni-toring program during preoperational or startup testing. We require that either a visual or instrumented inspection (as i O .
WAPWR-S/E A2-26 AMENDMENT 2 8558e:1d JANUARY 1989
appropriate)Ibe conducted to identify 'any excessive _ vibration Q that'could result in fatigue-failure. Revise Section 3.9.2.1 in-Module 7 to include a commitment to test-such piping, or provide
, justification for deviation from this position.
RESPONSE
O Please .see W response to Staff Q 210.58- and revisions to-Subsection 3.9.2.1 of Module 7.
210.60 Provide the following revisions to Tables 3.9-3 and 3.9-5 in Module 7, or provide justification for not doing so:
1.- In the Stress Criteria Column of Class 1 Pumps .for Service
' Levels A, B, C.and D, add NB-3400. (Table 3.9-3 only)
- 2. . Footnote (a) in Table 3.9-3 states that a test of components may be performed in lieu of analysis. Revise this footnote to provide a _ commitment that if this option is used for the-WAPWR, details of.the test program will be submitted to the-staff.for. review prior to implementation of the option.
- 3. Revise the Stress Criteria Column for component supports in both tables to be consistent with the response to. Questions 210.28 and 210.65, i.e., commit to use ASME Section III, Subsection NF, 1986 Edition.
RESPONSE
Tables 3.9-3 and 3.9-5 (renumbered as 3.9-4 in revised Subsection 3.9) of RESAR-SP/90 PDA Module 7 " Structural /
Equipment Design" have been modified to , include the modifications and commitment as requested. .
I 210.61 The staff considers valve discs to be a part of the pressure boundary and as such should have allowable stress limits. If the stress limits in the valve column of Tables 3.9-3 and 3.9-5, Module 7 do not inc.lude valve discs, provide this criteria in Section 3.9.3.1 of Module 7.
O O
WAPWR-S/E A2-27 AMENDMENT 2 1 8558e:1d JANUARY 1989
A RESPONSE:
U Valve ' discs are considered part of the pressure-retaining boundary and use the allowable' stress' limits as shown in Table-
'3.9-3 of Module 7. The table has been modified to include valve s discs along with. valves for the various services levels.
l210.62 In addition to the information i n' Table 3.2-1 of Module 7, revise Section 3.9.3 to include the design basis which will be used to insure the structural integrity of safety-related-L O Heating', Ventilation and Air Conditioning (HVAC) ductwork and supports.
RESPONSE
Codes and Standards which will be used for the design of Safety Related ductwork and its supports have been added in Section 3.9.3.5.
210.63 Section 3.9.3.1.3 of Module.7 states that valves in sample lines O connected to the RCS are not considered to be ASME Class 1 because the loss .of flow through a severance of one of these 3/8-inch lines can be made up by normal charging flow.- The staff agrees 'that such piping does not have to be Class 1, however, it must be classified as Quality Group B and Safety Class 2. On Sheet 19, Table 3.2-1 of Module 7, RCS sample valves and piping meet these guidelines. Verify that these valves and piping in Table 3.2-1 are the only components affected by the statement in Section 3.9.3.1.3. If not, provide the classification of other applicable RCS sample valves and piping.
RESPONSE
In addition to the Safety Class 2 sample system valves and piping referenced in Section 3.9.3.1.3 as not required to be Safety Class 1, there are a number of 3/4 inch instrument O
WAPWR-S/E A2-28 AMENDMENT 2 8558e:1d JANUARY 1989
1 e, .
connections to the loop piping that are similarly categorized.
( These include:
o Two Reactor Vessel Level Instrumentation System (RVLIS) connections, one each to the hot legs of Loops C and D.
o Four sets of five elbow tap flowmeter connections in each of the four crossover legs (20 connections total).
The associated Class 2 piping and normally open isolation valves are all located inside containment. The rationale for this classification is as presented in Section 3.9.3.1.3, i.e., in case of severance of one of these. lines the outflow of reactor coolant can be made up by normal charging flow. .The flow limitation is provide by a 3/8 inch flow restrictor located in the nozzle which connects the 3/4 inch line to the loop piping.
210.64 The information in Section 3.9.3.2 of Module 7 relative to pump and valve operability assurance is not completely acceptable.
d(' Sections 3.9.3.2.1.C. 3.9.3.2.2.D, 3.9.3.2.3 and 3.9.3.2.4.C refer to IEEE Standard 344-1975. To be acceptable, these refer-ences should also include a commitment to meet the additional guidelines in Regulatory Guide'1.100. Revise all of the above sections and any other applicable section in the PDA to include such a commitment, or provide justification for not doing so.
RESPONSE
The Westinghouse operability program meets the requirements of Regulatory Guide 1.100. Subsection 3.9.3.2 of Module 7,
" Structural / Equipment Design" has been modified to indicate conformance with those additional guidelines.
210.65 The information in Sections 3.9.3.4.1 and 3.9.3.4.2 of Module 7 relative to the use of ASME Section III, Subsections NF and O Appendix F in the design of ASME Class 1, 2 and 3 component l
WAPWR-5/E A2-29 AMENDMENT 2 855Be:1d JANUARY 1989 l
supports needs to be updated. The staff has potential questions O on the following subjects relative to Subsection NF:
- 1. The bases for the selection of NF vs AISC jurisdictical boundaries for supports is requested. Describe which part of the support will be constructed as NF and which part will be constructed as building steel.
- 2. Provide buckling criteria used in the design of all component supports.
- 3. Are the stresses in supports which are produced by seismic anchor point motion of the supported piping and the thermal expansion of suppcrted piping treated as primary or O secondary stresses?
The staff's positions on the three issues above have been incorporated into the 1986 Edition of Subsection NF. Therefore, a commitment that ASME Section III, Subsection NF, 1986 Edition will be used in the construction of all ASME Class 1, 2 and 3 supports will suffice. It should be understood that
" Construction" is an all-inclusive term as defined in ASME Section III, Subsection NB/NC/ND 1100. Revise Sections 3.9.3.4.1 and 3.9.3.4.2 of Module 7 and Section 5.2.27 of Module 2 to provide this commitment. If this commitment cannot be made, provide a detailed response to the above three questions.
The information relative to the test load method in Appendix F O should also be revised to be consistent with ASME Section III, Appendix F, 1986 Edition.
RESPONSE
l All ASME Class 1, 2 and 3 component supports will be designed and constructed in accordance with the rules of ASME III, Sub-section NF and Appendix F, 1986 Edition or later Code or Record.
Section 5.2-27 of Module 2, and Sections 3.9.3.4.1 and 3.9.3.4.2 O. of Module 7 have been revised to reflect this commitment.
The portion of Section 3.9.3.4.1.1 dealing with the test load method has been deleted. Any alternative qualification methods O will conform to ASME III, Subsection NF or Appendix F, 1986 Edison or later Code of Record.
O WAPWR-S/E A2-30 AMENDMENT 2 8558e:1d JANUARY 1989
fl '210.66 The' following additional information is required in Section
(./ '
3.9.3.4 of Module 7 relative to the design of bolts for component supports:
- 1. Provide 'the allowable stress limits which are applicable to bolts used in.. equipment anchorage, component supports e'd fN flanged. connections. The staff position is that the stress
'd '
in these bolts should not exceed the yield strength at temperature.
-l
- 2. Provide a discussion of the design methods applicable to expansion anchor bolts used in component supports.
RESPONSE
- 1) Design of bolts in safety related component supports and
. equipment anchorages will be in accordance with the codes and standards identified in RESAR-SP/90. The applicable codes giving allowable stress limits for each class of fastener are listed below:
Bolts in Component Supports - ASME Section NF
'O Bolts in Structural Steel - ASCE N690.
Bolts Embedded in-Concrete - ACI 349, Appendix B Expansion Anchor Bolts - ACI 349, Appendix B Design methods for expansion anchor bolts will comply with Appendix B of ACI 349. Flexibility of the baseplate will be considered in the analysis of bolt loads.
210.67 Section 3.9.4 of Module 7 states that in the analyses of the control rod drive mechanisms and the gray rod drive mechanisms,
( a nonlinear elastic LOCA analysis and a separate linear elastic seismic analysis is performed. Provide the basis for performing both a linear and nonlinear analysis on the same structure.
RESPONSE
The information originally provided in Subsection 3.9.4 is incomplete. The linear seismic analyses referred to were O
WAPWR-5/E A2-31 AMENDMENT 2 8558e:1d JANUARY 1989
f') performed as part of an effort to determine the optimum seismic
" envelope for the APWR in Japan; because of the need to run a large number of cases, the non-linear structural model was linearized in order to maintain computer expenditures e. t a rea-sonable level. However, once the seismic design level for the
[~')
U plant had been determined, final seismic analyses were performed using the same non-linear model used for LOCA type analysis.
O Reference to the linear seismic analyses has been deleted from Subsection 3.9.4 since these results were not used for design purposes.
210.68 Section 3.9.6 of Module 7 states that the inservice testing (IST) program will be prepared within 6 months after the WAPWR operating license issue date. This is not an acceptable schedule. To provide the staff sufficient time to review the applicable information -and prepare a Safety Evaluation Report, the IST program must be submitted as a part of the Final Design Approval. Revise Section 3.9.6 to provide this commitment.
,~
RESPONSE
(s s\;
Subsection 3.9.6 of Module 7 has been revised to include a commitment to provide the inservice inspection program outline in the FDA submittal.
210.69 The information in Sections 3.9.6.1 and 3.9.6.2 of Module 7 infers that only ASME Class 1, 2 and 3 pumps and valves will be included in the inservice testing (IST) program for the WAPWR.
It is the staff's position as stated in Standard Review Plan, Sections 3.9.6.11.1 and 3.9.6 11.2 that all pumps and valves (o) which are considered as safety-related should be included in the IST program even if they are not categorized as ASME Class 1, 2 or 3. Revise Sections 3.9.6.1 and 3.9.6.2 of Hodule 7 to clarify this commitment.
RESPONSE
C)N Subsections 3.9.6.1 and 3.9.6.2 of Module 7 have been revised to include all safety-related pumps and valves in the inservice
(~'N test program.
N.]
WAPWR-S/E A2-32 AMENDMENT 2 8558e:1d JANUARY 1989
l
. e 210.70 Section 3.9.6.3 of Module 7 discusses relief. requests from the f)
V IST requirements of ASME Section XI. Revise this section to be consistent with the response to'0210.38.
RESPCNSE:
1 See the Westinghouse response to staff Q210.38. j i
210.71 Section 6.5.2.5 of Module 2 contains a brief discussion of the Westinghouse response to the licensing . issue relative to O
' periodic leak testing of pressure isolation valves. _ A more detailed commitment on this issue should be provided in Section 3.9.6 of Module 7. At the Preliminary Design Approval stage, the staff requires a commitment to perform periodic leak testing of all pressure isolation valves in accordance with applicable requirements in the Westinghouse Standard Technical Specifica-tions. In addition, a commitment is required to provide. at. the Final Design Approval stage, a list of applicable valves for the staff's review. Revise Section 3.9.6 of Module 7 to add these commitments.
RESPONSE
See the Westinghouse response to Staff 0210.38 O
O l
O l WAPWR-S/E A2-33 AMENDMENT 2 8558e:1d JANUARY 1989
(
Formal responses to Staff questions 220.1 through 220.9 were originally transmitted .in Addendum 2, NS-NRC-88-3304, Johnson-(W) to-- Rubsnstein (NRC), dated January 7, 1988 and Addendum 10, NS-NRC-88-3349,' Johnson (W) to Rubenstein'(NRC), dated June 23, 1989.-- Formal text revisions ~are provided as part ~of: this b amendment.
220.1 The design parameters applicable to the design basis tornado' 'do (3.3.2) not cover the entire area of contiguous United States, according-g to the guidelines stated in Regulatory Guide 1.76. . The maximum wind speed of 320 mph is above. the requirement' of Tornado Intensity Region II (300. mph), but is below that of Tornado Intensity Region I, (360 mph). The atmospheric. pressure drop of' l'.96 psi is less than the requirement of Tornado Intensity Region II (2.25 psi). If the plant site selection is intended to cover the entire contiguous United States, the design parameters should be .in -compliance with the requirements stipulated in Regulatory Guide 1.76. Otherwise, provide a map indicating the areas- and borders within.which the plant sites will be located and the bases of their design parameters.
RESPONSE
The parameters chosen for the design basis tornado are based on ANSI Standard 2.3-1983 " Standard for estimating tornado and extreme wind characteristics at nuclear power sites". This Standard is mcm recent than Regulatory Guide 1.76 which was issued in 1974. It is based on tornado occurrence statistics- and provides figures for the contiguous United States that define "
the magnitude of tornado that can be expected with a given recurrence interval. Westinghouse has conservatively selected the tornadic windspeed corresponding to a probability of 10(-7) per year at the worst location in the United States as the design basis for the SP/90. Westinghouse believes this provides an appropriate design basis and request that NRC consider endorsement of this industry standard.
1 O
WAPWR-S/E A2-34 AMENDMENT 2 8558e:1d JANUARY 1989
Provide bases and references for the missile penetration 3 9 220.2 (3.5.3) formulas describe the used for design of barriers.
condition as well as For each formula, the limitation of applicability and the basic assumptions or theories involved.
Does the heading " Steel (Standard Formula)" mean " Stanford Formula"? Please clarify.
O ,
RESPONSE
The missile penetration formulae (Subsection 3.5.3 of this module) for the design of concrete barriers have been changed from the modified Petry fermula to the modified NDRC formula.
This change follows the recommendations of the ASCE Committee on Impactive and Impulsive Loads (May, 1980) and is consistent with the acceptance criteria of the Standard Review Plan.
220.3 Provide in detail the design time history generated for seismic (3.7.1.2) design and analysis of Category I structures. Descriptions should include time durations of rise, strong motion, and decay stages of the time history. A plot of the design time history (in the free field and at the. foundation level of structure) is 6 required. The assumption of a total duration of 10 seconds is significantly less than that typically assumed (approximately 30 seconds) in post licensing reviews. Justify your assumption.
Also, provide a plot showing the power spectral density of the design time history, so that a verification of the uniformity of the spectral energy through the frequency range of interest can be made.
RESPONSE
The free field SSE seismic time history response is shown in Attachment A, Sheets 1 through 3. The rise time when 9 peak-to peak responses build up occurs for the first 1-2 seconds depending upon excitation direction. Significant peak-to peak responses occur in the 1-8 second time interval. Attenuation of the amplitude is observed beginning at about 8 seconds. A
- 6. duration of seismic time history loading of about 10 seconds has been used at several existing nuclear power plants such as Shearon Harris, Comanche Peak, and South Texas. During the design verification for the Nuclear Power Block site specific application, the site requirements will be evaluated against the current envelope seismic capabilities.
WAPWR-S/E A2-35 AMENDMENT 2 8558e:1d JANUARY 1989
Floo'r response -spectra were determined using an interaction, model (Figure 3.7-9 off RESAR-SP/90 PDA. Module 7, " Structural /.
l- . Equipment ' Design") which l accounts for - various ~ soil-structure interactions Eat the building foundation. A range of. soil characteristics' were considered in the parametric study- to enable development of conservathe response spectra . envelopes (Figures 3.7-11 to 3.7-19 of Module ),.
220.4 Provide- in detail the method used .to account' for torsional pd (3.7.2.11) effects on structures in the seismic analysis. . To account for accidental -torsion, an ' eccentricity of +5% of the maximum building dimension at the level under consideration .shall be assumed. The accidental torsional effect shall be additional to the effect due to variation of underlying soil properties of a site specific configuration.
RESPONSE
The building complex is modeled as a series of cantilevers as'-'
shown in Figure 3.7-9 of RESAR-SP/90 PDA Module- 7. Each stick' consists of an assemblage of equivalent beams and rigid links to simulate 'the stiffness and its distribution along building vertical axis.
The mass properties of the model comprise mainly the structural deadweight of the building. Equipment weighing more than 1 ton is included by lumping their masses at floor elevations 'on which it is supported. An equivalent distributed load is also-included to account for lighter equipment. This includes the i effects of piping systems, cable trays and other heavy permanent non-structural items. The masses are lumped at floor elevations s where the finite element nodal points are defined. One half of' the mass of. the wall and column system above and below a floor level is lumped into that level. Due to the eccentricity between the center of rigidity and the center of mass of a O
WAPWR-S/E A2-36 AMENDMENT 2 8558e:1d JANUARY 1989
floor, torsional floor vibration may occur about the building vertical axis. This, in general, will cause an increase in the transnational component of the floor response spectra. The floor eccentricity is modeled by providing a rigid horizontal link between the floor mass point and the center of rigidity.
The dynamic models of the SP/90 Nuclear Power Block account for torsional effects by including the eccentricity between the center of rigidity and the center of mass of each floor. An E]T
\
' arbitrary additional eccentricity of + 5% of the maximum building dimension is not necessary and would needlessly complicate the analysis without benefiting 'the overall seismic safety margin.
220.5 The staff has not reviewed, and therefore has not approved the (3.7.2.15) ASCE nonproportional damping modeling approach of the draft report " Standard for the Seismic Analysis of Safety-Reiated Nuclear Structures" by ASCE Seismic Analysis Standard Committee,
, May 1984. Provide details of this approach including its basic assumptions, theories, experimental verifications, method of (v; analysis, and limitation of applicability.
RESPONSE
Modifications to Subsections 3.7.2.15 and 3.7.5 (References) of RESAR-SP/90 PDA Module 7, " Structural / Equipment Design" have been made in this amendment. (Please also see our response to Staff Question 220.7 which provides a detailed description of the WECAN Computer Code.)
Q i
220.6 Provide detailed description of the Westinghouse computer pro-(3.7.3.1) gram DEBLIN2. Give details of theories, assumptions, algorithm, and verification of accuracy and limitations of the progran.
w) RESPONSE:
A detailed description of the DEBLIN2 computer code, including the verification procedures, is now included in Appendix 3.7 of I RESAR-SP/90 PDA Module 7, " Structural / Equipment Design."
WAPWR-S/E A2-37 AMENDMENT 2 8558e:1d JANUARY 1989
220.7 Provide detailed description of the verification of Westinghouse (3.8.2.4) computer programs WECAN and ASHSD.
RESPONSE
,, ~s
(.) A detailed description of the WECAN and ASHSD computer codes, including the verification procedures, is now included in Appendix 3.8 of RESAR-SP/90 PDA Module 7, " Structural / Equipment
, Design."
('t,))
220.8 The structural acceptance criteria for buckling (instability)
(3.8.2.5) have not been clearly addressed. The ASME Boiler and Pressure Vessel Code,Section III, Subsection NE and the ASME Code Case N-284 did not contain adequate guidance and explanation on the
' subject. Provide design criteria for various structural members and shells on buckling, including knock down factors and safety factors for different loading conditions. Also provide theoretical and experimental bases and justifications for these factors.
- RESPONSE:
b,m Section 3.8.2 of RESAR-SP/90 PDA Module 7 describes the analysis R
of the metal containment. As an alternative to the ASME Boiler
& Pressure Vessel Code,Section III, Subsection NE requirements for determining allowable compressive stresses for Class MC components, the provisions of Code Case N-284 will be used. The Code Case, Section 1700 provides in great detail the stability (buckling) criteria for determining the structural adequacy against buckling of containment shells with more complex shell
( ) geometries and loading conditions than those covered by Subsection NE. Such effects as symmetrical or unsymmetrical dynamic loading conditions, circumferential and/or meridional' stiffening for heads, as well as cylindrical shells, combined stress field, discontinuity stresses and secondary stresses are considered in the stability analysis. It provides guidance on acceptable stress analysis procedures and methods for determining stress components to be used. The shell buckling t'h V
WAPWR-S/E A2-38 AMEN 0 MENT 2 8558e:1d JANUARY 1989 1
capacity' is based on linear bifurcation (classical). analyses reduced by. capacity reduction factors (knock-down factors) which
. account for the effects of imperfections and non-3inearity in geometry and boundary conditions and. by plasticity reduction factors that- account' for non-linearity in material properties.
The capaci.ty_ reduction factors and the safety factors.to be used for various conditions are specified in Code Sections 1500 and 1400, respectively. Plasticity reduction factors,. if required, are used as specified in Section 1600.
220.9 In Table .3.8-3 " Load Combinations and Load Factors'for Category *
(3.8.4.3) I Concrete Structures," the factor for accident pressure Pa of load combination no. 7 deviates from that of SRP 3.8.4,(1.15 instead of 1.25). Provide basis and . justification for- such deviations or change to comply with SRP.
RESPONSE: -
Table 3.8-3 has been ' corrected to show a 1.25 load factor on O accident pressure in load combination 7.
O
~y O
A2-39 AMENDMENT 2 WAPWR-S/E 8558e:1d JANUARY 1989 l 1 -
4
j ATTACHMENT A (Sheet 1 of 3) l l
100.00 g , , , i i , i i i a 3
80.00 60.00 I R
40.00 .
5 20.00 0.0 M
i
- -20.00
-40.00,
-60.00
~ 80.00
- 100.O OF i i ! ' ' '
O.O 1.00 2.00 3.00 4.00 5.00 6.00 7.00 8.00 9.00 10.00 O VME (SECS)
O WAPWR-S/E A2-40 AMENDMENT 2 8558e:1d JANUARY 1989
ATTACHMENT A (Sheet 2 of 3) 350.00
._0RIGlNAL ACCI:LERATION TIM [ HISTOR1 _
a.C 300.o0 250.00 R
200.00 E
150.00 100.0 W
50.00 0.00
-50.o0 I
-100.00 1 1 1 i R A 7
'F i 1 1 0.0 1.00 2.00 s.co 4.00 s.co s.oo 7.oo s.co s.co 10.00 TWE (SECS)
O O
WAPWR-S/E A2-41 AMENDMENT 2 8558e:1d JANUARY 1989 i
. .. ,i
ATTACHMENT'A O (Sheet 3 of 3)
O ..e.ee ORIGlNAL ACCl:LERATION TIM : H15 TORY __
300.00 a,c 250.00 R
200.00 E j 150.00 g 100.0 W
l 50.00 0.00
- S0.00
- 100.OC i i i t i 1 1 'T F i i O.0 1.00 2.00 3.00 4.00 5.00 6.00 7.00 8.00 9.00 10.00 TIME (SECS)
O .
O WAPWR-S/E A2-42 AMENDMENT 2 8558e:1d JANUARY 1989
_ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ - - _ _ _ _ _ _ _ - _ _ _ _ . _ _ _ __ ._ -. - . _ _ _ . _ _ = - - _ - _ . _ _ _
Formal responses to Staff questions 430.1 through 430.8 were
(]
U originally transmitted in Addendum 4, Johnson (W) to Rubenstein (NRC), dated May 13, 1988 and Addendum 9, Johnson (W) to Rubenstein(NRC),datedJune 14, 1988.
D.
430.1 Which power supplies are used for the alarms on the watertight (3.4.1.2) access doors to rooms used to protect safety-related equipmen:
(Mod. 7) from being adversely affected by flooding? Which power supplies are used on such alarms when they are on doors used to house components for more than one train of a single safety-related O system? Confirm that redundancy in power supplies is provided V as necessary in order to ensure that at least one train of safety-related equipment is protected from flooding.
RESPONSE
- 1) There are separate safety related power supplies associated with each system train.
- 2) There are no watertight access doors that lead to compartments containing more than one train. Separation of trains and redundancy of systems was a major design consideration to protect 'against the effects of flooding, fire, etc. . . .
430.2 Discuss any periodic tests or surveillance performed to assure (3.4.1.2) that the emergency floor drainage system is capable of pre-(Mod. 7) venting unacceptable water accumulation in safetr related equipment areas.
RESPONSE
Subsection 9.3.3 of RESAR-SP/90 PDA Module 13, " Auxiliary Systems" describes the " Equipment and Floor Drainage Systems."
A major safety design basis of the EFDS is to ensure prevention of water accumulation in areas which house safety related equipment. At the FDA inservice inspection and tests will be detailed to assure that the EFDS will function under the full O
WAPWR-S/E A2-43 AMENDMENT 2 8558e:1d JANUARY 1989
Q range- of design transients. The plant specific. Technical Specifications will include surveillance requirements to provide maximum assurance that the system will function to prevent flood damage to safety-related systems and components.
~b V 430.3 In subparagraph E of Section 3.5.1.1.2, you note the " remote" (3.5.1.1.2) likelihood of bonnets for valves rated- at 600 psig and below (Mod. 7) from -becoming missiles because of the low probability ' of-
. simultaneous failure -of the bonnet-to-body bolts. Provide information or data which shows that such failure .is only a remote possibility. Otherwise, bonnets (in valves rated at Os <600 psig) must be considered potential missiles and adequate protection must be provided against them.
RESPONSE
The position with regard to missiles generated by valves rated at 600 psig and below is consistent with that provided in other SAR's (e.g.,- South Texas FSAR Subsection 3.5.1.1.1, Amendment 36). W believes that the probability of those valve bonnets becoming missiles is low; however, specific plant layout to be developed later in the design of SP/90 will minimize the potential effects of these low energy missiles. If additional information is required by the NRC Staff, it will be provided at the final design stage.
430.4 In paragraph G of Section 3.5.1.1.2, you note that nuts, bolts, I (3.5.1.1.2) nut ana bolt combinations, and nut and stud combinations are not (Mod. 7) considered potential missiles because they have only a small amount of stored energy. Provide'information which demonstrates that these potential missiles are not capable of damaging O sensitive safety related equipment such as instrumentation or provide appropriate protection for such equipment.
RESPONSE
The response to 430.3 for valve bonnets is also applicable for nuts, bolts, nut and bolt combinations, and nut and stud combinations.
WAPWR-S/E A2-44 AMENDMENT 2 8558e:1d JANUARY 1989
1 430.5 In Section 3.5.1.1.2, you consider "certain vertical missiles 9 (3.5.1.1.2)
(Mod. 7)
(although not considered credible)" for which you state that pressurizer compartment coverall roof slab provides protection containment, engineered l
against potential damage to the safeguard components outside the pressurizer compartment. It is l difficult to determine how protection is provided for components near the pressurizer when these vertical missiles fall after O striking the slab. Therefore, discuss how protection is provided for safety related components in the vicinity of the pressurizer from such an occurrence.
RESPONSE
The pressurizer compartment will be designed to prevent vertical i missiles generated by valves located at the top of the pressurizer from leaving the compartment. Since no safety related equipment is located at lower elevations in the pressurizer compartment, falling missiles within that compartment are not a safety concern.
430.6 When discussing temperature and pressure sensors as a source of potential missiles in Section 3.5.1.2, you conclude that the G (3.5.1.2)
(Mod. 7) missile characteristics of these assemblies "are not of concern from a containment penetration standpoint." However, containment protection is not the only concern. It must also be shown that potential missiles from these sources are not of concern to safety-related components or protection must be provided for them.
RESPONSE
We agree that a potential missile generated by pressure or temperature sensor should not impair the function of safety 9 grade equipment that may be required to achieve cold shutdown.
This type of analysis will be performed at the FDA stage, when detailed information will be available on the type of missiles that potentially can be generated, as well as the locations of 9 safety related equipment such as valves.
O WAPWR-S/E A2-45 AMENDMENT 2 8558e:1d JANUARY 1989
/N 430.7 In Section 3.5.4, " Missile Protection Interface Requirements,"
() (3.5.4) you state that you have evaluated valvos in high pressure systems (Mod. 7) outside containment within the NPB scope for potential missile sources. You have concluded that there are no credible missile sources from these valves "since there~ is no single failure associated with any potential valve parts that can result in the i A generation of a missile." Provide additional details concerning !
U the design of valves such as the presence of' backseats, or special holddown devices which are capable of preventing missile generation. The discussion of missile generation. should be i revised to assume the potential generation of missiles from valves which do not have features to prevent missiles on failure p of the component. . The discussion should also address appropri-d ~
ate protection of safety related equipment from potential valve missiles. We note that only in Section 3.5.1.2.3 is a potential method for propulsion of missiles mentioned, i.e., jet propelled missiles. We find no mention of potential missile propulsion by means known as piston-type missiles such as those that result from. failures in rotating equipment. Explain the omission of any discussion covering the means by which missiles are propelled especially since you provide formulas showing missile penetration of barriers (Section 3.5.3, " Barrier Design Procedures") but no discussion on the means by which the missile velocities used in those calculations may be determined.
Confirm that all potential sources of missiles have been properly identified in accordance with the criteria of the SRP.
RESPONSE
The SP/90 design includes a high degree of separation between safety related and control grade equipment. The plant is designed such that potential missiles generated by control grade (i.e. non safety related) equipment located in a non-safety related area of the NPB will not affect safety related equipment located in the redundant safety related areas.
f)
Also, the SP/90 design includes a high degree of physical separation between redundant trains of safety related equipment; furthermore, there is significant compartmentalization within
(
each safety-related area. Thus, damage caused by missiles generated by safety-related equipment will be strictly limited and will not extend beyond the compartment in which the missile is generated.
WAPWR-S/E A2-46 AMENDMENT 2 8558e:Id JANUARY 1989 L__________
Valves .in high pressure systems in the NPB will be provided with V special features to prevent missiles from being generated.- In those cases where ,this is not practical, analyses will be . ' ;
performed at the FDA stage to demonstrate that potential missiles will not result in loss of safety related equipment that may'be required to achieve cold shutdown.
1 Failures of rotating equipment in the '
NPB are discussed in 1 Subsection. 3.5.4, in particular with regard to pumps and motor generator (MG) sets. The diesel generators are 'not included in' this discussion because a vendor for this equipment )
has not yet' been selected and thus design d'etails are not I available at this time. The FDA application will expand on Subsection 3.5.4 to include the diesel generators.
l '430.8 In.Section 3.11.2.3, " Methods and Procedures for Environmental (3.11) Justification," it is stated that; qualification may be demon-(Mod.'10)- strated by either type test, operating experience, analyses or a combination of these methods. However, 10CFR 50.49 does not provide for qualification by analysis only. Therefore, it is the staff's position that qualification methods -should be in IO compliance with paragraph (f) of 10 CFR 50.49. Your approach .to equipment qualification should be revised.'accordingly.
RESPONSE
The Westinghouse position on qualification of equipment by analysis is clarified in WCAP-8587, Section 5.3, page 5.5, first paragraph -
" Qualification by analysis alone is not employed by Westinghouse." Current Westinghouse practice does conform with 10CFR50.49(f). A paragraph has been added to Subsection 3.11.2.3 of RESAR-SP/90 PDA Module 7, " Structural / Equipment Design" to clarify this conformance. Note the referenced Section 3.11.2.3 is titled " Methods and Procedures for Environmental Qualification."
oo WAPWR-S/E A2-47 AMENDMENT 2 8558e:1d JANUARY 1989