ML19339A328
| ML19339A328 | |
| Person / Time | |
|---|---|
| Site: | Zimmer |
| Issue date: | 10/31/1980 |
| From: | CINCINNATI GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19339A326 | List: |
| References | |
| NUDOCS 8011030593 | |
| Download: ML19339A328 (650) | |
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' ' , .. _ . ZPS-1-MARK:,II'. DAR -
- AMENDMENT!13
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-INSTRUCTION' SHEET FOR.WM."H.'ZIMMER1 NUCLEAR ~ POWER
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' STATION-UNIT-1. MARK-II-DESIGN ASSESSMENT REPORT .- -
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l g WM. H. ZIMMER NUCLEAR POWER STATION UNIT 1 M ARK ll DESIGN O
ASSESSMENT REPORT THE CINCINN ATI G AS & ELECTRIC CO.
COLUMBUS & SOUTHERN OHIO ELECTRIC CO. .
THE DAYTON POWER AND LIGHT CO.
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ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 FOREWORD TO DESIGN ASSESSMENT REPORT (DAR) AMENDMENT 13 v
Amendment 13 to the DAR incorporeres all pertinent information from the original DAR plus Amendments 1 through 12, the Closure Report ( Appendix I to the FSAR) , and all information generated since submittal of the Closure Report. Since the DAR has been rewritten and reformated, bars have been added in the right hand margin of all text pages in Chapters 1 through 10 to indicate the origin of each paragraph. The key below explains the meaning of the bars:
No bar indicates that the material is from the original DAR as amended through Amend-ment 12. Some editorial or format changes may be included but the technical information is essentially unchanged.
Dashed lines indicate that the material originated in the' Closure Report. Some editorial or format changes may be included but the technical information is essentially unchanged.
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Solid lines indicate either new material or revisions to material formerly from the DAR or Closure Report.
It is hoped that this information will facilitate and expedite the review of Amendment 13.
v
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 e THE WM. H. 2IMMER NUCLEAR POWER STATION - UNIT 1 k )s MARK II DESIGN ASSESSMENT REPORT TABLE OF CONTENTS PAGE
1.0 INTRODUCTION
1.0-1 2.0 2IMMER EMPIRICAL LOADS 2.0-1
2.1 DESCRIPTION
OF THE ZIMMER EMPIRICAL LOADS 2.1-1 2.1.1 Vent Clearing 2.1-1 f 2.1-1 2.1.2 Pool Swell 2.1.3 Condensation Oscillation 2.1-1 2.1.4 Chugging 2.1-2 2.1.4.1 Chegging Lateral Loads 2.1-3 2.1.4.2 Chugging Boundary Loads 2.1-3 2.1.5 SRV (Quencher) Loads 2.1-3 2.1.6' Submerged Structure Loads 2.1-4 2.1.6.1 SRV Submerged Structure Loads 2.1-4 2.1.6.2 LOCA Submerged Structure Loads 2.1-4 2.1.7 Load Combinations 2.1-5 2.1.8 Design Changes 2.1-5 O
kJ 2.2 PIPING ASSESSMENT 2.2-1 2.2.1 Comparison of Rams Head Design-Basis Response Spectra and T-quencher Assessment Response Spectra 2.2-1 2.2.2 T-quencher Assessment - Drywell Piping 2.2-2 2.2.2.1 Assessment of Support Load 2.2-3 2.2.2.2 Assessment of Drywell Piping Stress Increases 2.2-3 2.2.2.3 Summary of Drywell Piping Assessment -2.2-4 2.2.3 Additional Piping Design Margins Obtained Using :Zirmner Empirical Loads 2.2-4 2.2.3.1 Impact of the Empirical Limiting CO Load Definitions on Drywell Piping Support Loads 2.2-4 2.2.3.2 Impact of Empirical Limiting CO Load Definition on Drywell Piping Stresses 2.2-5 2.2.4 Balance-of-Plant Piping (Outside Containment) 2.2-5 2.2.5 Wetwell Piping Assessment 2.2-6 ;
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2.3 BALANCE-OF-PLANT EQUIPMENT 2.3-1 2.3.1 Assessment and Requalification Procedure 2.3-1 2.3.1.1 Procedure for Equipment Originally Qualified by Testing 2.3-1 i
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ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
(~) TABLE OF CONTENTS (Cont'd)
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PAGE 2.3.1.2 Procedure for Equipment Originally Qualified by Analysis 2.3-1 2.3.2 High and Low Frequency Conce, c., 2.3-2 2.3.3 In Situ Testing 2.3-2 2.3.4 Equipment Foundation Loads 2.3-3 2.3.5 Results of Equipment Assessment 2.3-3 2.3.5.1 Valve Qualification Assessment 2.3-3 2.3.5.2 Equipment and Instrumentation Assessment 2.3-3 2.4 STRUCTURAL ASSESSMENT 2.4-1 2.4.1 Method of Assessment 2.4-1 2.4.2 Primary Containment 2.4-1 2.4.3 Drywell Structural Steel 2.4-1 2.4.4 Downcomer Bracing System 2.4-1 2.4.5 Pedestal Straps Supporting Piping 2.4-2 2.5 NSSS EQUIPMENT 2.5-1
2.6 CONCLUSION
S 2.6-1
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3.0 SRV IN-PLANT TEST PROGRAM 3.1-1 3.1 BACzGROUND 3.1-1 .
3.2 PURPOSE 3.2-1 3.3 TEST
SUMMARY
3.3-1 3.4 TEST MATRIX 3.4-1 3.5 DATA ACQUISIT"ON 3.5-1 3.6 TEST SCHEDULE AND REPORTING 3.6-1 4.0 GENERAL DESCRIPTION OF THE PLANT 4.0-1 5.0 LOADS CONSIDERED 5.0-1 5.1 ORIGINAL DESIGN LOADS 5.1-1 5.1.1 Loads on the Structure 5.1-1
- 5. i . 2 Loads on Piping and Equipment 5.1-3 ii
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
,g TABLE OF CONTENTS (Cont'd)
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PAGE 5.2 SAFETY / RELIEF VALVE (SRV) LOADS - PRESENT j DESIGN-LOADS-(T-00ENCHERS) 5.2-1 5.2.1 Design-Basis SRV Loads - Rams' Head 5.2 5.2.1.1 Conservatism in SRV --Rams Head Methods 5.2-3 5.2.1.2 Safety / Relief-Valve Discharge Cases i 5.2-6 5.2.1.2.1 Single Valve Actuation 5.2-6 4
5.2.1.2.2 Asymmetric SRV Actuation 5.2-6 5.2.1.2.3 Automatic Depressurization System (ADS) 5.2-7 5.2.1.2.4 All Valve Discharge Cases 5.2-7 5.2.1.2.5 Second Actuation 5.2-9
, 5.2.1.3 Safety / Relief Valve Boundary Loads 5.2.1.4 Submerged Structure Loads 5.2-10 5.2.2 Assessment for SRV Loads - T-Quencher 5.2-11 5.2.2.1 5.2-11 Conservatism of the T-Quencher Load Definition 5.2-11 5.2.2.2 SRV Quencher Discharge Cases 5.2.2.2.1 Single Valve 5.2-11 5.2.2.2.2 Asymmetric SRV Load 5.2-12 5.2.2.2.3 Automatic Depressurization System (ADS) 5.2-12 5.2.2.2.4 All Valve Discharge 5.2-12 5.2.2.3 Quencher Boundary Loads 5.2-12 O- 5.2.2.4 Quencher Submerged Structure Loads 5.2-12 5.2.3 Assessment of NRC Acceptance Criteria - SRV 5.2-12 5.2.4 References 5.2-13 5.2-14 5.3 LOSS-OF-COOLANT ACCIDENT (LOCA) LOADS 5.3-1 5.3.1 Design-Basis LOCA Loads 5.3-3 5.3.1.1 Load Definitions 5.3-3
' 5.3.1.1.1 LOCA Water Jet Loads 5.3-3 5.3.1.1.2 LOCA Charging Air Bubble Loads 5.3.1.1.3 Pool Swell 5.3-3 5.3.1.1.4 Pool Fallback 5.3-3 5.3-3 4
5.3.1.1.5 Condensation Oscillation 5.3.1.1.6 Chugging 5.3-4 5.3.1.1.7 Lateral Loads on Downcomers 5.3-4 5.3.1.2 LOCA Boundary Loads 5.3-5 5.3-7 5.3.1.2.1 LOCA Water Jet 5.3-7
- 5.3.1.2.2 Charging Air Bubble 5.3.1.2.3 Pool Swell 5.3-8 4
5.3.1.2.4 Pool Fallback 5.3-8 5.3-8
- 5.3.1.2.5 Condensation Oscillation 5.3-8 l 5.3.1.2.6 Chugging 5.3-8 5.3.1.3- LOCA Submerged Structure Loads 5.3-8 5.3.1.5.1 !
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' -LOCA-Water Jet- 5.3-9 '
5.3.1.5.2 Charging-Air Bubble 5.3-9
(]) 5.3.1.5.3 Pool Swell 5.3-9 l
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ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 TABLE OF CONTENTS (Cont'd)
PAGE
- 5. 3.'1. 5. 5.1 Poul Swell Impact Loads '_,,, 5.3-9 5.3.1.5.5.2 Pool Swell Drag Loads 5.3-11 5.3.1.5.4 Pool Fallback 5.3-11 5.3.1.5.5 Condensation Oscillation Drag Loads 5.3-11 5.3.1.5.6 Chugging Drag Loads 5.3-12 5.3.1.4 Annulus Pressurization 5.3-13 5.3.1.4.1 Transient Asymmetric Differential Pressure Events 5.3-13 5.3.1.4.1.1 Acoustic Loading 5.3-14 5.3.1.4.2 Annu]us Pressurization - Design Considerations 5.3-14 5.3.1.4.3 Annulus Pressurization - Design Analysis 5.3-15 5.3.1.4.5.1 Calculation of Mass and Energy Flow Rates 5.3-15 5.3.1.4.5.1.1 Comparison of General Electric Analysis to RELAP 5.3-17 5.3.1.4.5.2 Application of Mass-Energy Release to Compute Force-Time Histories of RPV
, and Shield Wall 5.3-18 5.3.1.4.5.3 Acceleration Time-Histories and Response Spectra Generation 5.3-19 5.3.2 Assessment of NRC Acceptance Criteria - LOCA 5.3-19
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5.3.2.1 LOCA Water Jet Loads 5.3-20 5.3.2.2 Pool Swell 5.3-20 5.3.2.2.1 Pool Swell' Velocity 5.3-21 5.3.2.2.2 Pool Swell Impact 5.3-21 5.3.2.3 Drag Load Calculations 5.3-21 5.3.2.4 Chugging Lateral Loads 5.3-21 5.3.2.5 Condensation Oscillation Loads 5.3-22 5.3.3 References 5.3-23 5.4 ZIMMER POSITION ON NRC LEAD PLANT ACCEPTANCE CRITERIA (NUREG-0487) 5.4-1 6.0 LOAD COMBINATIONS CONSIDERED 6.1-1 6.1 CONTAINMENT AND INTERNAL CONCRETE STRUCTURES 6.1-1 6.2 CONTAINMENT LINER 6.2-1 6.3 OTHER STRUCTURAL COMPONENTS 6.3-1 6.3.1 Load Combinations 6.3-1 6.3.2 Acceptance Criteria 6.3-1
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J ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
(} TABLE OF CONTENTS (Cont'd)
PAGE 6.4 BALANCE-OF-PLANT PIPIN3 AND EQUIPMENT 6.4-1 6.4.1 Wetwell Piping 6.4-1 6.4.2 Non-Wetwell BOP Piping 6.4-2 6.4.2.1 Load Combinations and Acceptance Criteria 6.4-2 l 6.4.2.1.1 Piping Systems Within the Reactor Building 6.4-2 6.4.2.1.2 Piping Systems Outside the Reactor Building 6.4-3 6.4.3 Balance-of-Plant Equipment 6.4-3 6.4.3.1 Loading Combinations 6.4-3 6.4.3.2 Acceptance Criteria 6.4-3 6.5 NSSS PIPING AND EQUIPMENT, VESSEL, AND INTERNALS 6.5-1 6.5.1 References 6.5-2 7.0 REEVALUATION AND DESIGN ASSESSMENT 7.1-1 7.1 CONTAINMENT AND INTERNAL CONCRETE STRUCTURES 7.1-1 7.1.1 Structural Analysis for SRV Loads 7.1-1 7.1.2 Structural Analysis of LOCA Loads 7.1-4 7.1.2.1 Vent Clearing Analysis 7.1-5 7.1.2.2 Pool Swell Analysis 7.1-5 7.1.2.3 Condensation Oscillation Analysis 7.1-6 7.1.2.4 Chugging Analysis 7.1-7 7.1.3 Effects of Downcomers on the Drywell Floor 7.1-8 7.1.4 Design Assessment Margin Factors 7.1-10 7.1.4.1 Critical Design Sections 7.1-10 7.1.4.2 Design Forces and Margin Factors 7.1-11 7.1.5 References 7.1-14
, 7.2 CONTAINMENT WALL AND BASEMAT LINER ANALYSIS 7.2-1 7.2.1 Basemat I.iner 7.2-1 7.2.1.1 Description of Liner and Liner Stiffeners 7.2-1 7.2.1.2 Loads for Analysis 7.2-1
. 2.1.3 Load Combinations 7.2-1 7.2.1.4 Acceptance Criteria 7.2-1 7.2.1.5 Analysis 7.2-2 7.2.2 Containment Wall Liner 7.2-2 4
7.2.2.1 Description of Liner 7.2-2 7.2.2.2 Loads for Analysis 7.2-2 7.2.2.3 Load Combinations 7.2-2 g) 7.2.2.4 Acceptance Criteria 7.2-2
(_/ 7.2.2.5 Analysis 7.2-3 V
2PS-1-MARP II DAR AMEN 9 MENT 13 OCTOBER 1980 TABLE OF CONTENTS (Cont'd)
PAGE 7.3 OTHER STRUCTURAL COMPONENTS 7.3-1 7.3.1 Downcomers and Downcomer Bracing 7.3-1 7.3.1.1 General Description 7.3-1 7.3.1.1.1 Downcomer Properties 7.3-1 7.3.1.' 1 Bracing Properties 7.3-1 7.3.1.s._ Connection Properties 7.3-2 7.3.1.2 Loads for Analysis 7.3-2 7.3.1.3 Design Load Combinations 7.3-3 7.3.1.3.1 SRV Actuation Load Combination 7.3-4 7.3.1.3.2 LOCA Associated Load Combinations 7.3-4 7.3.1.4 Acceptance Criteria 7.3-4 7.3.1.4.1 Acceptance Criteria for Downcomers 7.3-4 7.3.1.4.2 Acceptance Criteria for Downcomer Bracing 7.3-4 7.3.1.4.3 Acceptance Criteria for Connections 7.3-5 7.3.1.4.4 Acceptance Criteria for Welded Joints 7.3-5 7.3.1.5 Analysis 7.3-5 7.3.1.5.1 Static Analysis 7.3-5 7.3.1.5.2 Dynamic Analysis 7.3-5 7.3.2 References 7.3-7 7.4 REACTOR PRESSURE VESSEL HOLDDOWN BOLTS '
y t-1 7.5 BALANCE-OF-PLANT (BOP) PIPING AND EQUIPMENT 7.5 i
- 7. 5.1 - BOP Piping 7.5-1 7.5.1.1 Evaluation of Bounded Load Combinations (Rams Head Definition) 7.5-1 7.5.1.1.1 Loads and Load Combinations Evaluated 7.5-1 7.5.1.1.2 Drywell Piping 7.5-1 7.5.1.1.3 BOP Piping Inside the Reactor Building 7.5-2 7.5.1.1.4 Wetwell Piping 7.5-3 7.5.1.2 Impact of Change to T-Quencher Discharge Device 7.5-3 7.5.1.3 Impact of SRV T-quencher and LOCA on Rams Head Design Basis 7.5-3 7.5.2 Balance-of-Plant Equipment 7.5-5 7.6 NUCLEAR STEAM SUPPLY SYSTEM-(NSSS) EQUIPMENT 7.6-1 7.6.1 Piping and Equipment 7.6-1 7.6.1.1 Reevaluation Procedures for NSSS Piping 7.6-3 7.6.1.2 Reevaluation Procedures for NSSS Equipment 7.6-3 7.6.1.2.1 Reevaluation of Pipe-Mounted / Connected Equipment 7.6-3 7.6.1.2.2 Reevaluation for Floor / Structure-Mounted Equipment 7.6-3 vi 1
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'ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
(}L . TABLE OF CONTENTS (Cont'd)
PAGE 7.6.2 Reevalation for the Reactor Pressure Vessel Supports and Internal Components 7.6-4 8.0 SUPPRESSION POOL WATER TEMPERATURE MONITORING SYSTEM .8.1-1 8.1 SYSTEM DESIGN 8.1-1 8.1.1 Safety Design Basis 8.1-1 8.1.2 General System Description 8.1-1 8.1.3 Normal Plant. Operation 8.1-2 8.1.4 Abnormal Plant. Operation 8.1-2 8.1.4.1 Plant Transients 8.1-3 8.1.4.2 Abnormal Events 8.1-3 8.1.4.3 Primary System Isolation 8.1-3 8.1.4.~4 Stuck-Open Relief Valve 8.1-3 8.1.4.5 Automatic Depressurization System (ADS) 8.1-4 8.1.5 Transients of Concern 8.1-4 8.2 SUPPRESSION POOL' TEMPERATURE REPONSE 8.2-1
() 8.2.1' References 8.2-2 9.0 PLANT MODIFICATIONS AND RESULTANT IMPROVEMENTS 9.1-1 9.1 STRUCTURAL MODIFICATIONS 9.1-1 9.2 BALANCE-OF-PLANT (BOP) PIPING AND EQUIPMENT 9.2-1 9.2.1 BOP Piping 9.2-1 9.2.1.1 Drywell Piping 9.2-1 9.2.1.2 Wetwell Piping 9.2-1 9.2.1.3 BOP Piping ~ 9.2 9.2.2 Equipment 9.2-2 9.3 NSSS PIPING AND EQUIPMENT. 9.3-1 9.4 SRV DISCHARGE QUENCHER 9.4-1 10.0 PL?.T SAFETY MARGINS 10.1-1 10.1 CONSERVATISMS IN PLANT DESIGN 10.1-1 10.1.1 Conservatistis in Pool Dynamic Loads 10.1-1 10.1.2. Structural Conservatisms 10.1-2 10.1.3 Mechanical.Conservatisms 10.1-2 10.1.3.1 Conservatisms in BOP Piping Analysis
.0 -10.1.3.2 Conservatisms in BOP Equipment 10.1-2 10.1-4' vii.
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
( TABLE OF CONTENTS (Cont'd) PAGE 10.1.4 Conservatism in NSSS Design 10.1-4
11.0 CONCLUSION
S 11.0-1 4 APPENDIX I. COMPUTER PROGRAMS A.1-1 APPENDIX B NRC OUESTIONS WITH RESPONSES B.1-1 APPENDIX C SOIL-STRUCTURE INTERACTION MODEL C.0-1 APPENDIX D MASS ENERGY RELEASE METHODOLOGY D.1-1 APPENDIX E MASS OPERABILITY ACCEPTANCE- E.1-1 APPENDIX F LINE INVENTORY MODEL - MAIN STEAM-LINE RUPTURES F.1-1 APPENDIX G SUBMERGED STRUCTURE METHODOLOGY G.1-1 APPENDIX H T-OUENCHER REEVALUATION FOR PIPING SYSTEMS H.1-1 f i O viii
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 THE WM. H. ZIMMER NUCLEAR POWER STATION - UNIT 1 () MARK II DESIGN ASSESSMENT REPORT LIST OF TABLES NUMBER TITLE PAGE 2.1-1 Plant Modifications 2.1-7 2.2-1 Piping Acceptance Criteria 2.2-7 2.2-2 Load Combinations and Acceptance Criteria 2.2-8 2.2-3 Drywell Piping Assessment: Comparison of Piping Support Load Magnitudes (UPSET-B) 2.2-10 2.2-4 Drywell Piping Assessment: Comparison of Piping Support Load Magnitudes (EMERGENCY-C)- 2.2-11 2.2-5 Drywell Piping Stress Assessment 2.2-12 2.2-6 Piping Overstress 2.2-13 2.2-7 Piping Stress Summary 2.2-14 2.2-8 Load Combinations Evaluated for the Wetwell Piping 2.2-15 2.5-1 Summary of Load Cases for Equipment for Study Purposes 2.5-3 2.5-2 Load Case Definitions 2.5-4 2.5-3 Review of Previous Results 2.5-5 s 2.5-4 NSSS Safety-Related Components Assesscd 2.5-6 s . 2.5-5 NSSS Safety-Related Components Assessed 2.5-7 2.5-6 Zimmer Main Steam System Calculated Snubber Loads 2.5-8 2.5-7 Zimmer Recirculation System Calculated Snubber Loads 2.5-9 3.2-1 List of Equipment Being Monitored During In Situ SRV Test 3.2-2 3.3-1 Test Matrix 3.3-3 3.3-2 Test Matrix - Definition of Abbreviations - and Footnotes 3.3-5 4.0-1 Primary Containment Principal Design Parameters and Characteristics 4.0-2 5.2-1 SRV Discharge Line Clearing Transient Parameterization 5.2-15 5.2-2 SRV Bubble Dynamics Parameterization 5.2-16 5.2-3 Trsient Analysis Assumptions 5.2-17 5.2-4 Re.11ef Valve Inputs - Zimmer - Analysis 5.2-18 5.2-5 Zimmer Transients Results 5.2-19 5.3-1 Acoustic Loading on Reactor Pressure Vessel Shroud 5.3-24 I 5.4-1 Zimmer Position on NRC Lead Plant Acceptance Criteria (NUREG-0487) 5.4-2 6.1-1 Design Load Combinations 6.1-4 1 1x ! l h
2PS-1-MARK II DAR AMENDMENT 13 2 OCTOBER 1980 1 i LIST OF TABLES (Cont'd) i $ NUMBER TITLE PAGE l - 6.3-1 Load Definitions and Combinations for Reinforced Concrete (Struc-I ture Other Than Containment) 6.3-2 6.3-2 Load Definitions and Combinations for Structural Steel 6.3-5 ! 6.5-1 Loading Combintions and Acceptance
- Criteria - Operating Condition
- Categories _
6.5-3 1 6.5-2 Load Combinations and Allowable Stress Limits for.Non-Fluid System Equipment 6.5-5 j 6.5-3 Load. Combinations and Allowable Stress
- Limits _for Active Fluid System Equipment 6.5-6 j 7.1-1 Dynamic Soil Properties 7.1-15 7.1-2 Margin Table for Basemat - Resonant Sequential Symmetric Discharge 7.1-16 7.1-3 Margin Table for Basemat - ADS Valve
, Discharge 7.1-17 7.1-4 Margin Table for Basemat - Two Valve i Discharge 7.1-18 7.1-5 - Margin Table for Basemat - LOCA Plus One SRV _ 7.1-19 7.1-6 Margin Table for Containment - Resonant O- Sequential-Symmetric Discharge 7.1-20 7.1-7 Margin Table for Containment - ADS. 1 Valve Discharge 7.1-21 7.1-8 Margin Table for Containment - Two-i Valve Discharge 7.1-22 I
7.1-9 Margin Table for Containment - LOCA-i Plus One Valve Discharge 7.1-23 I 7.1-10 Margin Iaole for Reactor Support - Resonant Sequential Symmetric Discharge 7.1-24
, 7.1-11 Margin Table for Reactor Support -
p ADS Valve Discharge 7.1-25 7.1-12 Margin Table for Reactor Support - Two-Valve Discharge 7.1-26 l 7.1-13 Margin Table for' Reactor Support -
- LOCA Plus One SRV 7.1-27 7.1-14' Margin Table for Drywell Floor -
SRV Only.and LOCA Plus One SRV 7.1-28 7.1-15 Margin Table for Drywell Floor Column - i All-Valve and Jd)S Discharge 7.1-29 7.1-16. Margin Table for Drywell Floor Column - Two-Valve Discharge 7.1-30 7.1-17 Margin Table for Basemat - All-Valve SRV Quencher Discharge. 7.1-31 7.1-18_ Margin Table for Basemat -~ ADS SRV Quencher Discharge 7.1-32 l x g -"--rw- W -+'%3 - +w- ---w,y g - +wmy q w py-eqe --,ygm ne e e-> -eTgg-g 9 wp ry - M 1 -p 9y7 ymWwN *s NT+Tm-g. $3N*-t+9y
2PS-1-MARK 'II DAR AMENDMENT 13 OCTOBER 1980 s LIST OF TABLES (Cont'd) NUMBER TITLE PAGE 7.1-19 Margin Table for Basemat - Asymmetric (Three-Valve) SRV Quencher Discharge 7.1-33 7.1-20 Margin Table for Basemat - Single-Valve SRV Quencher Discharge 7.1-34 7.1-21 Margin Table for Containment - All-Valve SRV Quencher Discharge 7.1-35 7.1-22 Margin Table for Containment - ADS SRV Quencher Discharge 7.1-36 7.1-23 Margin Table for Containment - Asymmetric (Three-Valve) SRV Quencher Discharge 7.1-37 7.1-24 Margin Table for Containment - Single-Valve SRV Quencher Discharge 7.1-38 7.1-25 Margin Table for Basemat - ADS SRV Quencher Discharge 7.1-39 7.1-26 Margin Table for Basemat - Single-Valve SRV Quencher Discharge 7.1-40 7.1-27 Margin Table for Containment - ADS SRV Quencher Discharge 7.1-41 7.1-28 Margin Table for Containment - Single Valve SRV Quencher Discharge 7.1-42 7.2-1 Summary of Containment Wall Liner Plate (-)/
's Stresses / Strains for All SRV Cases (Rams Head) 7.1-4 7.2-2 Summary of Containment Wall Liner Anchorage Load / Displacement for All SRV Cases (Rams Head) 7.2-5 7.2-3 Summary of Containment Wall Liner Plate Stresses / Strains for All SRV Cases (T-quencher) 7.2-6 7.2-4 Summary of Containment Wall Liner Anchorage Load / Displacement for All SRV Cases (T-quencher) 7.2-7 7.3-1 Load Combinations and Acceptance Criteria for Downcomer and Downcomer Bracing 7.3-8 7.4-1 Maximum Forces on RPV Holddown Bolts 7.4-2 7.4-2 Design Forces and Margin Factors of RPV Holddown Bolts 7.4-3 7.5-1 Load Combinations and Acceptance Criteria 7.5-7 7.5-2 Impact on Piping Supports ABSUM 7.5-8 7.5-3 Percent Restraints Bounded for Various Factors 7.5-9 7.6-1 Dynamic Methods for Zimmer NSSS Assessment 7.6-6 7.6-2 Class IE Equipment Qualification j 4 7.6-7 1 i ()
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ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 LIST OF TABLES (Cont'd)
-v NUMBER TITLE PAGE 7.6-3 Class 1E Control Panels and Local Panels and Racks Seismic Qualification Test Summary 7.6-8 8.2-1 Common Assumptions for All Zimmer Suppression Pool Temperature Response Analyses 8.2-3 8.2-2 Zimmer Event 1 - Stuck-Open Relief Valve from Power Operation 8.2-4 8.2-3 Zimmer Event 2 - Stuck-Open Relief Valve from Hot Standby 8.2-5 O
i rii I _- . _ , , , . . - . _ _, _ . . . . . ~ . _ , . _ _ . _ . _ . . _ . . - . . ~ - - .
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 THE WM. H. ZIMMER NUCLEAR POWER STATION - UNIT 1 0' MARK II DESIGN ASSESSMENT REPORT LIST OF FIGURES NUMBER TITLE 2.1-1 Zimmer Empirical Limiting CO vs. NRC Criteria 2.1-2 Asymmetric Chugging Load Distribution 2.1-3 Spatial Load Distribution - SRX g KWU vs. SRV Plant Specific ADS 2.2-1 Response Spectra Comparison - OBE + SRV ALL, OBE + SRV RH, OBE + SRVTOASY TO 2.2-2 Response Spectra Comparison - DBE + SRV RH vs. SSE + CO (2-7 Hz) 2.2-3 Response Spectra Comparison - DBE + SRVgnt RH vs. CO(EL) + SSE + SRV att TO 2.2-4 Response Spectra Comparison - DBE + SRVp3 vs. SSE + CHUG + SRVgtL 2.2-5 Support Load Changes - OBE SRVggnRH vs. OBE + T SRVALL O 2.2-6 Rams Head Support Load Changes - OBE + SRV gtn RH v3. OBE + SRVgtpTO 2.2-7 Drywell Piping Support Load Changes - OBE + SRV 33t RH vs. OBE + SRVAsyTO () 2.2-8 Drywell Piping Support Load Changes - 1.875(OBE) + SRVgtLRH vs. SSE + CO(2-7) 2.2-9 Support Load Changes - 1.875(OBE) + SRVgtt RH vs. SSE + CO(2-7) 2.2-10 Drywell Piping Support Load Changes -1.875(OBE) + SRVgttRH vs. SSE + CHUG + SRV T 2.2-11 Support Load Changes - 1.875(ADS)O OBE + SRVggn RH vs. SSE + CHUG + SRVADSTO 2.2-12 Response Spectra Comparison - DBE + SRV RH vs. SSE + CO(EL) + SRVADSTO 2.2-13 Drywell Piping Support Load Change -1.875(OBE) + SRVgtg RH vs. SSE + CO(EL) +SRVA TO 2.2-14 Support Load Changes - 1.875(OB + SRV RH vs. SSE + CO(EL) + SRV ALL TO 2.2-15 Drywell Supports Ahhklable Design Margin 2.4-1 Drywell Piping Support Load Change - 1.875(OBE) + SRVg RH vs. SSE + CO(EL) + SRV. TO 2.4-2 Suppbht Load Chcnges -1.875(OBI}DS+ SRVggg RH vs. CO(EL) + SRVgn3 TQ +9SE 2.4-3 Downcomer Bracing Layout 2.4-4 Typical Suppression Pool Section 2.4-5 Connection of Bracing To Wall Embedment : 2.4-6 Downcomer Bracing Support to Pedestal l 2.4-7 Connection of MSRV Ring Plate i 2.4-8 Connection of Bracing to Downcomer 1 3.3-1 Accelerometer Locations xiii
/= }
b-.. 2PS-1-MARK-II DAR- . AMENDMENT 13 OCTOBER 1980 l {} LIST OF FIGURES (Cont'd)
. NUMBER j TITLS 3.3-2 -Suppression Pool Prersurc 3.3-3 Suppression Pool Temperature Sensors 3.3-4 Temperature and Pressure Sensor Locations 3.1-5 Suppression Pool Strain Gauge Locations 3.3-6 SRV Discharge Line Sensor Locations 3.3-7 HPCS Suction Line Strain Gauge Locations 3.3-8 Downcomer Strain Gauge and Accelerometer Locations 3.3-9 T-quencher Sensor Locations 3.3-10 0uencher Support Strain Gauge Locations 3.6-1 SRV Test Schedule-4.0-1 Primary Containment Before Modification 5.2-1 Orientation Depressurization Actuation of SRV Rams Head Devices for Automatic 5.2-2 Reactor Vessel Cycling Steam Pressure vs. Time - ATWS Pump Trip Off 5.2-3 Reactor. Vessel Cycling Steam Pressure vs. Time - ATWS Pump Trip On 5.2-4 Cross Section of Suppression Pool and Definition of Suppression Chamber Walls Loading 2ones 5.3-1 Drywell/Wetwell Pressure History for Recirculation
-Line Break (DBA)
() 5.3-2 Drywell/Wetwell' Temperature History for Recirculation Line B. Teak (DBA) 5.3-3 Drywell/Wetwell Pressure History for Main Steamline Break 5.3-4 Drywell/Netwell Temperature History for Main Steamline Break 5.3-5 Drywell/Wetwell Pressure History.for Intermediate Size Break'(IBA) 5.3-6 Drywell/Wetwell Temperature' History for Intermediate Size Break (IBA) 5.3-7 Drywell/Wetwell Pressure History for Small Break 1 (SBA) 5.3-8 Drywell/Wetwell Temperature History for Small Break (SBA) 5.3-9 Vent Clearing Jet Angle 5.3-10 Jet Impingement Load 5.3-11 Chugging Load 5.3-12 Drag Load Due to Vent Clearing Jet 5.3-13 Maximum ~ Impact Pressure on Small Structures 5.3-14 Maximum Impact Force on Pipes 5.3-15 Maximum Impact Force on I-Beams 5.3-16 -Time _ History of Impact Load for Small Structures in Pool. Swell Region 5.3-17 Lateral Loads on Groups of Downcomers at Probability Level 10-* 5.3-18 Time Sequence
..)
xiv - - _ _ - - _ . _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - - - _ - . _ _ _ _ _ _ _ _ _ _ _ _ . _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ - - - - - . _ _ _ - _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - - _ _ - _ - . _ _ - ~ _
ZPS-1-MARK,II DAR : AMENDMENT 13 OCTOBER 1980 LIST OF FIGURES (Cont'd) l (:) i NUMBER TITLE-5.3-19 -Acoustic Load Illustration 5.3-20 Loading Description Restricted Pipe Motion During Breakdown 5.3-21
.5.3-22 Recirculation Line Break 5.3-23 Break Flow vs. Time ^- Feedwater Line Break 5.3-24 Recirculation Line Break Nodalization ,
5.3-25 Feedwater Line Nodalization 5.3-26 Drag Coefficients- ~ 7.1-1 Zimmer FSI Analysis Model 7.1-2 Average Shear Strain Versus K, 7.1-3 Average Snear Stain Versus Critical Damping 7.1-4 Cross Section-of Suppression Pool and Definition of Suppression Chamber Walls Loading Zones 7.1-5 Typical Resonant Sequential Symmetric Discharge Forcing Function - Zone 4 7.1-6 Typical Asymmetric SRV Discharge Forcing Function - i Zone 4 7.1-7 LOCA Vent Clearing Pressure Loads on Basemat
; 7.1-8 LOCA Vent ~ Clearing Pressure Distribution i'
7.1-9 Structural Model Including Soil 7.1-10 Drywell/Wetwell Pressure History for Main Steamline Break () 7.1-11 7.1-12 Pool Swell Symmetric Load Pool Swell Asymmetric Load 7.1-13 LOCA Cyclic Condensation Pressure Load on Basemat Containment and Reactor Support for Rams Head 7.1-14 Spatial Distribution of LOCA Condensation Oscillation Load for T-quenchers 7.1-15 . Chugging Load - Magnitude'and Spatial Distribution 7.1-16 Chugging Load - Time History 7.1-17 Drywell Floor Analytical Model i' 7.1-18 Circumferential Variation of Moment in Drywell Floor Due'to Radial Moment Applied to Radius 22'-3" 7.1-19 Radial Variation of Moment'in Drywell Floor Due to Concentrated Radial Moment Applied at Radius 22'-3" 7.1-20 Circumferential Variation of Moment in trywell Floor Due to Concentrated Circumferential Moment Applied at Radius 22'-3" 7.1-21 Radial Variation of Moment in Drywell Floor Due to Concentrated Circumferential Moment Applied at Radius 22'-3" 7.1-22 Critical Design Section 2 Variation of Radial Moment with Number of Loaded 1Downcomers I
-7.1-23 Critical Design Section 2. Variation of Radial Moment with Number of Downcomers Located per DFFR 7.1-24 Basemat Plan Top Reinforcing-Layout 7.1-25' Basemat Plan Bottom Reinforcing Layout
() . XV
.. _ -- .. _ _ , _ ~ _ - . _ - . . .
~_- - ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (~3 LIST OF FIGURES (Cont'd) x) NUMBER TITLE 7.1-26 Containment Wall Post-Tensioning Layout 7.1-27 Containment Wall Reinforcing Layout 7.1-28 Reactor Support Concrete Plug 7.1-29 Reactor Support - Reinforcing Layout Before Modification 7.1-30 Drywell Floor Reinforcing Layout 7.1-31 Drywell Floor Column Reinforcing Layout 7.1-32 Design Sections - Primary Containment, Reactor Containment, Reactor Support, and Basemat 7.1-33 Design Sections - Drywell Floor 7.1-34 Design Sections Drywell Floor Column 7.1-35 Typical Interaction Diagram for Basemat 7.1-36 Typical Interaction Diagram for Containment 7.2-1 Basemat Liner Detail 7.2-2 Containment Liner Detail 7.3-1 Downcomer in the Suppression Pool 7.3-2 Downcomer Bracing Layout 7.3-3 Connection of Bracing to Downcomer 7.5-1 Embedment Load Change Inside Containment for N + CO (DFFR) + SSE 7.5-2 Embedment Load Change Inside Containment for N + CHUG (')
+ SRV TO SSE 7.5-3 Embedment Load Change Inside Containment For N + CO (Empirical) + SRV SSE 7.5-4 EmbedmentLoadChakge+OutsideContainmentforN+CO (DFFR) + SSE 7.5-5 Embedment Load Change Outside Containment For N + CO(Erapirical) + SRVTO + SSE 7.5-6 Embedment Load Change Outside Containment for N + CO(Empirical) + SRVTQ + SSE 7.6-1 Design and Evaluation Flow 8.2-1 Suppression Pool Temperature vs. Mass Flux - Stuck-Open SRV from Power 8.2-2 Suppression Pool Temperature vs. Mass Flux - Stuck-Open SRV f om Hot Standby 9.4-1 T-quencher Discharge Device 9.4-2 Plan Lccation of T-quenchers 'v_)
xvi
ZPS-1-MARK II DAR AMENDMENT 13 SEPTEMBER 1980 (~) U CHAPTER 1.0 - INTRODUCTION The purpose of this revision of the Design Assessment Report (DAR) is to demonstrate that the Wm. H. Zimmer Nuclear Power Station, Unit 1 (ZPS-1) containment can accommodate all hydrodynamic load phenomena associated with the SRV discharge and LOCA in the BWR Mark Il containment, to provide evidence of conformance with the NRC Lead Plant Acceptance Criteria (NUREG-0487), and to provide a response to the formal questions posed by the Nuclear Regulatory Commission (NRC). In the summer of 1979, the Wm. H. Zimmer Power Station (2PS-1) design and construction status was such that additional load changes requiring plant modifications would seriously impact the construction schedule. To avoid this situation the 2PS-1 "three-pronqed" approach was adopted. The three facets of this approach are:
- a. Evpedite construction based on conservative loads and upgrade immediately containment capability where possible.
- b. Assess the plant for the Zimmer Empirical Load Design Basis which is expected to bound any future changes in pool dynamic loads.
() c. Confirm adequacy of design with results of the Zimmer in-plant SRV test and the long-term Mark II program. The Zimmer empirical loads are described in Section 2.1. This report describes the original design-basis for 2PS-1, subsequent reassessments for revised and newly identified loads, and finally the current reevaluation of the design using the Zimmer empirical loads to ensure the adequacy and conservatism of the containment structures, piping, and equipment. This report also describes the conformance of the Zimmer design to the NRC Lead plant Acceptance Criteria (NUREG-0487). This report provides the NRC staff with all information necessary to continue and complete the licensing of the Wm. H. Zimmer Nuclear Power Station as scheduled. All pertinent information related to loads, load specification, load combinations, acceptance criteria, plant modification, plant margins, and confirmation of loads that apply to ZPS-1 has been compiled in this document. In addition, an in-plant SRV test will be performed to confirm the adequacy of loads used for design a messment. In this report the individual loads and load combinations that are being utilized in the reassessment are identified and described in the first four sections. Reports defining the I individual loads and providing justification for application to e- the 2PS-1 containment are referenced rather than repeated. This (_3) is consistent with the objective of this report. , l 1.0-1
v 1 ZPS-1-MARK IILDAR AMENDMENT 13- ,. SEPTEMBER 1980 (>g x . The methods used-in reevaluating the structures, piping systems,
'and-equipment are described in Chapter.7.0. The plant modification and resultant changes that have-been completed are
- described in. Chapter 9.0. The plant margins end conservatisms are summarized in. Chapter-10'.0.
;The long-term Mark'II program is expected to confirm that the
~
4 plant, as presently designed and' constructed, is completely safe-and adequate. However, additional design modifications and plant changes.are'being implemented to utilize the full containment capability. This' ensures that the maximum possible margins are j i built into the plant, so that if load definitions should change later, they can be accommodated with>ut plant hardware changes. The.2PS-1 plant startup should, therefore, proceed as scheduled. u i i n O f I l i O 1.0-2
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 w CHAPTER 2.0 - ZIMMER EMPIRICAL LOADS The original design of the Zimmer Power Station was based on loads developed in the Mark II Containment Program as documented in the Mark II Containment Dynamic Forcing Function Report (DFFR, j Report NEDO/NEDE 21061-P). Although these-loads were felt to be
- conservative, questions about the adequacy of the loads resulted in-replacement of the rams head SRV discharge devices with quenchers in all Mark II plants and led the Mark II program to perform additional full-scale, single-vent LOCA tests (4TCO tests). As a result of these changes, the potential existed for Mark II pool dynamic loads of higher magnitude or altered frequency _ range.
In the summer of 1979, the status and schedule of construction 4 and design work for the Zimmer station was such that any further changes in the pool dynamic loads would have a serious impact on the cost and the schedule for operation. It was recognized, at this time, that-full results from the-various tests would not be available in rime for incorporation into the design basis. Therefore, Zimmer implemented a three-pronged approach to completion of the plant. This approach, although requiring a significant amount of additional design work and significant plant modifications, was felt to be advisable to minimize the risk of delays in plant operation and to maximize the safety of the design. The three-pronged approach was:
- a. Expedite construction based on conservative loads and
; upgrade immediately to containment capability where possible.
- b. Assess the plant for the Zimmer Empirical' Load Design Basis which is expected to bound any future changes in. pool dynamic loads.
- c. Confirm adequacy of design with results of the Zimmer 1 in-plant SRV test and the long-term Mark II program.
This chapter describes the Zimmer Empirical Load and demonstrates the capability of the Zimmer Power Station to. accommodate these-very conservative loads. This information was discussed with the NRC at a' meeting on December 5, 1979. The remainder of the DAR provides more detail of the design of the Zimmer. Plant including the original design methods and design work done subsequent to
- the December 5,--1979 meeting. Section 5.4 summarizes the
- .conformance of the Zimmer Station to NUREG-0487, the Mark II Lead Plant Acceptance Criteria.
O 2.0-1
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
2.1 DESCRIPTION
OF THE ZIMMER EMPIRICAL LOADS The Zimmer Empirical Loads constitute a complete Mark II hydrodynamic load design basis. This design basis was formulated not only to meet or exceed the DFFR and NUREG-0487 (Lead Plant Acceptance Criteria) but to also contain additional conservatism in those areas where uncertainty remained in the Mark II loads. Since this approach was formulated in the summer of 1979, certain loads haveinbeen justified somebetter cases.defined and load reductions have been These changes have not been incorporated into the Zimmer Empirical Load which ensures the high margin of safety in the design. The following subsections fully define the load and provide documentation of the material presented to the NRC in the December 5, 1979, meeting. 2.1.1 Vent Clearing The vent clearing boundary load used in the Zimmer design is a 33 psi overpressure (above hydrostatic) applied uniformly below the vent exit and attenuated to zero at the pool surface. This exceeds both the Mark II Owners Group load and the NRC requirements Zimmar design. demonstrating an increased safety margin in the 2.1.2 Pool Swell ,s The pool swell methodology used in the Zimmer design meets or \i) exceeds the N3C Acceptance Criteria. In those areas where the Acceptance Criteria were different from the original Zimmer design, the loads have been calculated using both methods and the more conservative the design margin. load used for the design, thereby increasing Zimmer has been modified to remove most piping and structures from the pool swell zone to eliminate pool swell loads. 2.1.3 Condensation Oscillation Prior to implementation of the Zimmer Empirical Load approach, Zimmer had been designed to accommodate the condensation oscillation 2-7 Hz). This (CO) load specified in DFFR, Revision 3 ( 3.75 psi, load was accepted by the NRC in NUREG-0487. Certain questions were raised about the adequacy of this load definition because tne original 4T tests (GE report NEDE-13442-01p5/76) were not entirely prototypical of Mark II containments. To resolve these questions, the Mark II Owners Group performed the 4TCO test (NEDE-24811-P, 5/80) with conservative single-cell representation of the Mark II drywell and appropriate vent length and geometries. Because the schedule for availability of results from the 4TCO tests was not compatible with the design and construction (~} ss schedule of Zimmer, the Zimmer Empirical CO Load was defined very 2.1-1 m
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 conservatively based on existing steam condensation data. () B?cause condensation oscillation occurs over a wide range of blowdown conditions, two CO. loads were defined. The first is a high mass flux CO load which would correspond to the early portion of a large break LOCA. The load is defined as:
- a. Sinusoidal Pressure Fluctuations i 4.6 psi 8 2-7 Hz i 2.2 psi a 11-13 Hz
- b. Random Pressure Fluctuations Steam' Bubble Collaspe: 15-50 Hz The 2 to 7 hertz component specified represents an increase of about 20% over the DFFR/NUREG-0487 load. The 11 to 13 hertz component is an additional load to account for any vent acoustic effects. The higher frequency portion of the load is added to bound random high frequencies which may appear in test data.
At lower mass fluxes there may be a possibility of a higher contribution from the vent acoustic effect with a corresponding decrease in the low frequency component. This load is defined as: () a. Sinusoidal Pressure Fluctuations t 2.2 psi 2-7 Hz i 3.8 psi 11-13 Hz
- b. Random Pressure Fluctuations Steam Bubble Collapse: 15-50 Hz The 2 to 7 hertz component here is 50% of the Icw frequency component used in the high mass flux load while the vent acoustic amplitude has been conservatively assumed to be even higher than the amplitude specified in the lower 2 to 7 hertz range in the DFFR. The higher frequency load is defined as described above.
The Zimmer Empirical Condensation Oscillation Load bounds the requirements of the NRC Lead Plant Acceptance Criteria (NUREG-0487), as demonstrated by Figure 2.1-1. 2.1.4 Chuaging Chugging loads are divided into two areas. The chugging lateral load is the self loading of the downcomer vent during chugging and affects the design of the downcomers, bracing, and drywell floor. The chugging event also generates a hydrodynamic load (" which loads the submerged boundaries of the supprecsion pool. 2.1-2
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 ,c3 2.1.4.1 Chuggino Lateral Loads V Using the Zimmer Empirical Loads Approach, lateral loads are calculated as described in Subsection 5.3.1.1.7. This approach is more conservative than required by the NRC Acceptance Criteria tNUREL-0487). Subsection 7.1.3 describes the additional conservatism added in the method of drywell floor assessment. 2.1.4.2 Chuoging Boundary Loads The chugging load used was the DFFR methodology which meets the NRC Acceptance Criteria (NUREG-0487). The symmetric chugging load is obtained from the full-scale, single-cell 4T data and conservatively applied with all vents in-phase. An amplitude of
+4.8/-4.0 psi and a 20-30 Hz frequency range is applied. The asymmetric load utilizes the same frequency range and a maximum magnitude of +20/-14 psi. Again, all vents were assumed to act in phase. The asymmetric distribution is shown in Figure 2.1-2.
2.1.5 SRV (Ouencher) Loads The Safety / Relief Valve (SRV) actuation loads used in the original design of Zimmer were based on the rams head discharge device. Quencher discharge devices have now been installed to eliminate concerns about discharge into high temperature pools and to reduce the magnitude of the SRV loads. The quencher load k-} definition (Susquehanna DAR) is supported by full-scale, single-cell tests of an actual Mark II quencher. This load is included in the Zimmer Empirical Loads. Because this load has been shown to be conservative by comparison to full-scale tests and because it includes a wider frequency range than the original rams head load, the quencher load definition provides a very conservative basis for plant design assessment. Subsequent to adoption of the Zimmer Empirical Load, information has been provided to the NRC supporting an amplitude reduction of approximately 30% in the quencher load. Consistent with the Zimmer philosophy of retaining the maximum design margin, this load reduction has not been incorporated into the Zimmer Empirical Load. The T-quencher load definition consists of three actual pressure time histories. The amplitude of these data traces are then increased by 50% to ensure conservatism and the frequency range is adjusted to give primary frequencies between 3.4 to 10 Hz. This load definition provides amplitude which bound both first and subsequent actuation loads. Since an all-valve case is used as the design basis, the Zimmer design basis will bound an all-valve subsequent actuation case (with all ';ubbles in-phase) although a maximum of 5 of the 13 valves are predicted to undergo subsequent actuation in the Zimmer plant. The quencher load definitions incorporate a very conservative representation of the (a~') spatial distribution of pressure on the boundary of t'le 2.1-3
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (']' suppression pool. The Zimmer Empirical Load carries this conservatism even further and uses the all-valve case to represent the ADS case. To investigate the conservatism of this approach, a more realistic prediction of the Zimmer ADS load has been formulated utilizing the DFFR methodology to predict the spatial distribution. The conservatism of the Zimmer Empirical ' Load Approach is demonstrated by this comparison (Figure 2.1-3). 2.1.6 Submerged Structure Loads S"bmerged structure loads have been calculated using forcing functions consistent with the boundary loads just described. The submerged structure methodology has also been modified to address the NRC Acceptance Criteria (NUREG-0487). This subsection will
, cover the submerged structure load definitions. The revised i
niethodology is documented in Appendix G. 2.1.6.1 CRV Submerged Structure Loads The actual quencher locations are used to define the position of i the SRV air bubbles. The bubble size is conservatively predicted l by utilizing the actual plant parameters (such as line length). The bubble pressure and typical load time history are calculated using the quencher correlations in the DFFR (NEDO 21061). The time history is then adjusted to give a frequency range of 3.4 to 10 Hz. Other aspects of submerged structure load calculation, () such as, drag coefficients and nodalization of structures, are treated in accordance with NUREG-0487, as explained in Appendix G. 2.1.6.2 LOCA Submerced Structure Loads The water jet, vent clearing, and pool swell submerged structure loads have been reassessed taking into consideration the NRC Lead Plant Acceptance Criteria (NUREG-0487) in both the forcing functions and application methodology. The methodology information in Appendix G is applicable to LOCA submerged structures also. The chugging submarged structure load is derived from the chugging boundary load. This is described in more detail in Subsection 5.3.1.3.6. The only modification to the chugging load is to address the concerns in NUREG-0487. The condensation oscillation submerged structure loads have been recalculated to be consistent with the Zimmer Empirical Loads (as described in Subsection 2.1.3 and NUREG-0487). A forcing function was derived from the original load specification ( 3.75 psi, 2-7 Hz), but additional, lower pressure loads were defined up to 21 Hz to bound uncertainties in the load definition. The frequencies above 21 Hz in the boundary load specification result ( from acoustic pressure waves and were, therefore, not included in (-]- the fluid drag loads. 2.1-4
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 e 2.1.7 Load Combinations Nm) The load combinations applicable to the design of the Zimmer Station are listed in Section 6.0. The Zimmer Emnirical Loads Approach considers all these combinations. However, to expedite the assessment, some of the loads are combined in a more conservative way than is actually required. In additicn, some of the individual load cases are replaced by more conservative loads to minimize the amount of analysis required. As described in Subsection 2.1.5, the SRV loads are defined with a very conservative spatial distribution. Because of this, the ADS (6-valve) load is almost as large as the all-valve (13-valve) load. Additional margin is built into the design by using the all-valve case to represent the ADS case. In addition, a non-mechanistic three-valve case enveloped with the all-valve and ADS cases to bound any symmetric effects. This results in additional margin being built into the design. The largest loads generally result from the combination of an earthquake with the ADS discharge and either chugging or condensation oscillation. This is clearly an event with a very low possibility. In spite of this low probability, the very conservative load definitions described in the section have been combined using the absolute sum method of load combination. The combination of SRV and LOC /. loads is particularly O' conservative in the case of the drag loads on submerged structures. The flow fields established by the quencher air bubble and the downcomer st.ar bubble collapse are superimposed as if they each had the worst possible phasing and direction at the same time. Because of the difference in the position, ' frequency, and shape of the forcing function, it is very unlikely that a significant reinforcement of the flow field will result. The method of combination used with the Zimmer Empirical Loads is the absolute sum method. The square root of the sum of the ' squares (SRSS) method is more appropriate and has been approved by the NRC, but at the time the Zimmer Empirical Load Approach was adopted, the acceptability of SRSS was unclear. Therefore, the conservative absolute sum method was used to ensure adequate margins. 2.1.8 Design Changes In an effort to maximize the design margin, the Zimmer Empirical Loads were defined with sufficient conservatism, that in many cases, the design of the piping equipment and structures approaches the containment and embedment capacity. This approach has required a significant number of changes in the plant. The rest of Chapter 2.0 describes the assessments which were done to redesign or confirm the adequacy of the plant. O 2.1-5
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 e^s A list of plant changes, grouped by the initiating cause, is (_) included in Table 2.1-1. This list shows that a large number of changes have been made in the wetwell and drywell as well as some changes outside containment. In the wetwell area, the addition of-the bracing and quenchers resulted in the relocation and upgrading of virtually all the piping and supports, such that, the capability has been considerably increased. Similarly, the use of the governing building response (containment capability) in the drywell design resulted in a significant upgrading of the structural steel and pipe supports. (~)h (~'N %) 2.1-6
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 TABLE 2.1-1 PLANT MODIFICATIONS WETWELL 4 1 USE T-QUENCHERS IN LIEU OF RAMS HEAD i Add 79 embedments in walls and basemat i Add 6 pedestal bands for MSRV line supports 1 Add 226 supports for MSRV and non-MSRV lines ) Upgrade sections of MSRV piping size and wall thickness Replace rams heads with T-quenchers J Relocate T-quenchers for better distribution
- Redesign support steel under drywell floor REMOVE FROM POOL SWELL l
Remove access hatch grating ] Relocate DW-WW vacuum breakers I ADDITION OF DOWNCOMER BRACING Add 13 wall embedments Add downcomer bracing g Add structural steel beam in pedestal l REMOVE LOAD FROM WW COLUMNS i Reroute all 24 non-MSRV lines Upgrade sections of non-MSRV piping wall thickness Remove all support attachments to columns [ OTHER CAUSES i
- Remove downcomer bottom flange Fill pedestal with concrete to water level DRYWELL s
UPGRADE FOR GOVERNING BUILDING RESPONSE 4 Upgrade approximately 30% of drywell steel Upgrade embedment capacity to_ limit
- Add 15% new snubbers Upgrade 25% of snubbers and rigid struts i
Reinforce HVAC supports (Approximately 440 total snubbers and 180 rigids in drywell) I O i 2.1-7 4
-. ~
, . -- . . . . , , .7., - - . ,
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 TABLE 2.1-1 (Cont'd) O OUTSIDE CONTAINMENT UPGRADE FOR GOVERNING BUILDING RESPONSE Upgrade RBCCW Hx supports Upgrade RHR Ex supports Modify selected HVAC supports Add 10% new snubbers Upgrade 20% of snubbers and rigid struts Upgrade HVAC supports t Upgrade cable tray supports (Approximately 470 total snubbers and 600 rigids in Rx building) O O 2.1-8
i AMENDMENT 13 OCTOBER 1980 20.0 ZIMMER P~~N EMPERICAL I \ LIMITING CO : I- I i J
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SRV RL KWU VS. SRVADS PLANT SPECIFIC i i
2PS-1-MARK 17 DAR AMENDMENT 13 OCTOBER 1980 2.2 PIPING ASSESSMENT '~# The piping analysis for 2PS-1 was originally completed using the rams head response spectra as the design basis. The approprir.te load combinations are defined in Table 2.2-1. As indicated in ; this table, the governing loads for the rams head design basis were reduced to three load cases; cases 2, 3, and 7. The Zimmer Empirical Load Approach uses very conservative T-quencher load definitions. The plant had previously been assessed for rams head loads and has thu capability to accommodate those loads. As the associated load definitions and response spectra became available, it became apparent that there were some differences between the rams head response spectra and the T-quencher response spectra. The following subsections describe two separate evaluations which were performed to compare the original design basis against 1) the T-quencher loads and 2) the Zimmer Empirical Loads which include more conservative LOCA loads. With the modifications as implemented, the Zimmer plant is believed to be more than adequate. A detailed assessment was completed for the loads and load combinations, which meet or exceed those specified in the NRC Lead Plant Acceptance Criteria (NUREG-0487), using the Zimmer Empirical Loads for the KWU T-quencher discharge device. These load combination cases are defined in Table 2.2-2 in the column labeled "T-quencher Assessment." Since several of these load combinations are bounded by other load combinations, a summary is provided in Table 2.2-2. The results of the assessment indicate that all of the piping supports currently designed for rams head loads a;e adequate for the T-quencher load definitions. Additional design margins have been incorporated in the support design to accommodate uncertainties in the LOCA loads, which may be subject to revision, based on future results of test programs. Finally, all safety-related piping will be evaluated for adequacy using the LOCA load definitions from the long-term Mark II program based on the 4TCO test data and SRV load definitions, based on in-plant test results in order to confirm the existing design margins. 2.2.1 Comparison of Rams Head Desian-Basis Response Spectra and T-Ouencher Assessment Response Spectra Figures 2.2-1, 2.2-2, 2.2-3, and 2.2-4, illustrate the typical differences between the rams head (original design basis) response spectra and the T-quencher assessment response spectra. It was found that the T-quencher assessment response spectra were typically less than the rams head design-basis response spectra in all horizontal directions. This is illustrated in Figure 2.2-3. It was also found that the vertical T quencher assessment response spectra was higher in the low frequency range, i.e., below 7 hertz, as illustrated in Figures 2.2-1, 2.2-2, and 2.2-4. Since the majority of the safety-related piping in the Zimmer (} plant were designed to be relatively stiff, i.e., with 2.2-1
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 7, fundamental frequencies greater than 7 hertz, it was not expected (,) that this low frequency content would have significant impact. The assessment deme' trated that this is the case and is explained in detai in later sections. 2.2.2 T-quencher Assessment - Drywell Pipino A detailed a.c' .. dent was made to evaluate the adequacy of the rams head design basis against the very conservative T-quencher load definitions. This assessment was completed by performing analyses of 13 of the 25 major piping subsystems in the drywell. The remaining piping subsystems in the drywell were either symmetric to the subsystems analyzed, or their design basis was governed by operating transients. Because of this, it was expected that the T-quencher load assessment would not have a significant itpact on these subsystems. A few small diameter piping subsyst. ems (nominal diameter less than 2 inches) and all instrumentation lines were not included in this assessment. All these lines will be included in the final design review of the Zimmer plant. Both static and dynamic computer analyses were performed on '.hese piping subsystems using techniques identical to production piping analysis. Representative piping systems were analyzed for all applicable load cases, and the governing load combinations were tabulated for comparison purposes. The results of these load combinations were comp red to the equivalent rams head load (-]/ ss combinations in both the support loads and in the piping stresses. In general the results indicated that:
- a. For load combinations currently required by the NRC, the support loads tend to decrease when rams head design-basis loads are compared to the T-quencher loads.
- b. The loads that did increase were all associated with small diameter (nominal O.D. less than 4 inches)
' piping systems and even these load increases were all within the rating of the snubber load capacity.
- c. The load increases were primarily due to the increases in the lower frequency range of the response spectra.
- d. The impact of the piping stresses was insignificant.
- e. The irpact of the Zimmer Empirical CO Load was significant on the piping systems. This impact was due to the larger amplitude in the higher frequency of the Empirical CO Load response spectra. In
(~l t_/ addition, the load combination with the CO Empirical 2.2-2
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 Load loads. included the effects of the SRV (T-quencher) It was expected that the overstress of the piping due to load combinations, including the Empirical CO Load, could be resolved by a more refined analysis, by using the actual material properties, or by use of the SRSS method of load combination. 2.2.2.1 Assessment of Support Load The impact2.2-5, in Figures of the assessment on the drywell piping is illustrated 2.2-6, and 2.2-7. Figure 2.2-5 illustrates the comparison results between the OBE + SRV ggn rams head case and the OBE + SRVatt T-quencher case. As indicated in the histogram, the majority of the supports in the drywell piping show a decrease in the load magnitude when comparing the T-quencher to the rams head loads. There are some supports which show a load increase; but as illustrated in Figure 2.2-6, all of the load increases are,for support loads less than 2,000 pounds. As commented earlier, these occurred on piping systems with less than a nominal 4-inch diameter. Even more favorable results were obtained when comparing the rams head design basis with the SRV Ann T-quencher responses. Figure 2.2-7. The SRV These results are illustrated in defined as the envelope of SRVanggtn T-quencher and SRVload is conservatively ASYM-T C'/ A more given indetailed comparison showing typical numerical results is Table 2.2-3. As indicated in tnis table for the few cases where the T-quencher assessment loads exceeded the rams head oesign-basis loads, the final capacity of the snubber. loads are still within the Figures 2.2-8 through 2.2-11 illustrate the impact of the T-quencher load definitions and the original LOCA load definitions, as compared to the emergency rams head design-basis load combinations for which the piping was designed. As with other comparison load cases, the majority of the restraintthe supports show a decrease in load magnitede between the rams head design-basis loads and the T-quencher assessment loads. Also, as shown in more detail in Tab'e 2.2-4, the restraints which show a load increase are on the smaller piping systems, and the i magnitude support. of the increase is still within the capacity of the 2.2.2.2 Assessment of Drywell Pipina Stress Increases As illustrated in this in Table 2.2-5 for the 1471 data points evaluated assessment, combinations involving OBE and SRVno overstress conditiens existed for the load ALL (rams head or T-quencher). Only one minor overstress data point was found in the CO comparisons and it was expected that this overstress condition ! (_w) )
~
2.2-3 i
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
,- could be resolved by using slightly more refined analysis
(_j techniques. 2.2.2.3 Summary of Drywell Pipino Assessment As reviewed in Subsections 2.2.2.1 and 2.2.2.2, the rams head design basis results were compared to the load combinations in the NRC acceptance criteria using the KWU T-quencher load definitions. The results of this assessment clearly indicate that the Zimmer design, based on rams head loads, is adequate for the KWU T-quencher load definition. 2.2.3 Additional Pipino Design Margins Obtained Usino Zimmer Empirical Loads. As discussed earlier, the DFFR condensation oscillation load was defined only in the 2 to 7 hertz frequency range. As described earlier, in order to obtain additional design margins, a new empirical limiting steam condensation oscillation load was selected. The impact of this Empirical Limiting Load definition was compared to the original design-basis load definition. Because of this, a criteria has been proposed for the Zimmer Power Station to upgrade the piping support design in order to accommodate the conservative Empirical Condensation Oscillation Loads. This upgrading was accomplished by selecting an Empirical Limiting CO Load with a modified high frequency content. The yg same piping systems that were assessed for the KWU T-quencher (_/ load, as described in Subsection 2.2.2, were also assessed for the Empirical CO Load. In the assessment, the load combination of CO (EL) + SSE + SRV T-quencher was compared to the rams head design-basis rams head. emergency load combination of 1.875 OBE + SRV^LL In this assessment, which is discussed in the following subsection, the impact of the bounding Empirical Limitin7 CO Load was identified on both the support loads and on the pip ng stresses. The impact of this load combination is shown bf. the response spectra comparison in Figure 2.2-12. 2.2.3.1 Impact of the Emoirical Limitina CO Load Definitions on Drywell Piping Support Loads .' As can be seen in Figures 2.2-13 and 2.2-14, the Empirical Limiting CO Load definitions did have an impact on the piping support loads. While not all loads did increase, it was felt that in order to account for the uncertainties in the high frequency range it was necessary to increase all the loads on the drywell supports. These increased loads were evaluated against the existing support design to determine whether they could be accommodated. If required, these supports were upgraded to a larger size. The resulting design margins available in the drywell supports after the loads were upgraded is illustrated il Figure 2.2-15. Because of this upgrading, all drywell supports will accommodate the Zimmer Empirical Load Criteria and have ) (]) t 2.2-4
I I 2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 margin over the requirements of NUREG-0487, the Lead' Plant Acceptance Criteria. ({)- 2.2.3.2 Impact of Empirical Limitino CO Load Definition on Drywell Piping Stresses
, Of the 1471 locations evaluated-in the drywell piping, 17 were l overstressed due to the Empirical Limiting Criteria of CO(EL) +
SSE + SRV These data points are identified in ~ Table 2.2h3b . T-quencher. While the overstress conditions could be corrected by adding additional restraints, this would result in a more rigid pipino system which would not be beneficial when the piping design mtG .. also consider high thermal stresses (especially inside the drywell). Because of this, i t is not intended at this time to add additional restraints to correct the overstress
; condition. As indicated in Table 2.2-7, many of the overstress conditions can be easily resolved by using the SRSS load i combination technique. The remaining three overstress conditions could be resolved in the future by either 1) adding more restraints, 2) more refined analysis, 3) less conservative load definitions, or 4) consideration of the actual material properties. All of these overstressed locations will be resolved j as discussed or they will be reviewed in detail with the NRC.
The piping design will be reevaluated as required using more realistic CO load definitions, based on test results, when they are available through the Mark II Owners Group generic program. () .2.2.4 Balance of Plant P'ipina (Outside Containment) As was described in Subsections 2.2.2 and 2.2.3 for the drywell piping, detailed assessment was also completed for the piping and supports outside containment. The results of this assessment indicate that the piping and supports outside containment are
! adequately designed for the NRC Lead plant Acceptance Criteria using the original design-basis suppression pool loads with the T-quencher load definitions.
I The Zimmer Empirical load has only a localized effect on the
, piping outside containment. In the regions which were affected by the Empirical Load, the affected supports were upgraded which resulted in 95% of the supports outside containment being adequate to accommodate the Empirical Load combination. It is felt that the remaining 5% of supports outside containment can be accommodated by a more detailed evaluation.
For the 3199 stress locations evateated on the piping systems j outside containment for the Empirical Limiting Load Criteria, no overstress conditions occurred, even using the absolute sum i combination technique. 1 i 2.2-5
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EPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (^s 2.2.5 Wetwell Piping Assessment %.) Extensive modifications were made to the piping in the wetwell area including pipe rerouting for the following reasons:
- a. installation of T-quencher,
- b. addition of downcomer bracing, and
- c. reduction of stress of wetwell columns.
The changes involved consisted of:
- a. rerouting of all the wetwell piping,
- b. replacement of the rams head with T-quencher discharge devices,
- c. upgradir. of the piping wall thicknesses to accommodate new loads,
- d. addition of 226 wetwell supports, and
- e. relocation of the T-quenchet for better load distribution.
(~'t 'd The wetwell piping is being evaluated for the load combinations defined in Table 2.2-8. Since the piping is essentially being redesigned for the Zimmer Empirical Loads, including the Empirical CO Limiting Load definition, no problems are expected in this area. 2.2-6
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (J'~ N, inBLE 2.2-1 PIPING ACCEPTANCE CRITEPIA ZIMMER RAMS HEAD DESIGN BASIS (PIPING) N + OBE + SRV 3L RH UPSET (B) N + DBE* + SRVg RH EMERGENCY (C) N + Tf(DBE) + (AP) FAULTED (C OR D) NRC ACCEPTANCE CRITERIA USING T-QUENCHER LOADS (PIPING) N + OBE + SRV j3g7 TQ UPSET (B) N + SSE + CO (2-7) EMERGENCY (C) N + SSE + CHUG + SRV 0 EMERGENCY (C) ADS f N + N(SSE) + (AP) FAULTED (C OR D) (~T (Where DBE = 1.875 x OBE) 'w) 2.2-7
I) V (9 v (v'i TABLE 2.2-2 LOAD COMBINATIONS AND ACCEPTANCE CRITERIA LOAD -NRC LOAD COMBINATIONS RAMS HEAD DESIGN T-QUENCHER ACCEPTANCE CASE FOR MARK II PLANTS BASIS ASSESSMENT CRITFRIA 1 N+SRV Governed by Governed by B (N+SRV +0BE) (N+SRVh*g+0BE) 2 N+SRVX+0BE N+SRV +0BE N+SRV g +0BE B 3 N+SRV +SSE +1.8750BE Governed by C*
$i N+SRV^
." (N+SRVQ+SSE+ CHUG) {$
7 m 4 N+SRVADS+IBA (SBA) Go m ned by Go m ned by C* $ (N+SRVg +1.8750BE) (N+SRVQ+ CHUG +SSE) 5 N+SRV A A o w ned by ADS Go m ned by C* $- (N+SRV +1.8750BE) N+SRV**g+SSE+ CHUG % 6 Governed by N+SRVADS+SSE+IBA(SBA) C* (N+SRV +1.8750BE) N+SRVQ+SSE+ CHUG 7 N+SSE+DBA N+kSSE+AP N+ (SSE) +(AP) ; C* O N+SSE+CO Q 8 N N oz N A
- o tn 9 N+0BE Governed by (N+SRV Governed by B e$
+0BE) (N+SRV g +0BE) gg OW
O O O TABLE 2.2-2 (Cont'd) LOAD NRC LOAD COMBINATIONS RAMS IIEAD DESIGN T-QUENCHER ACCEPTANCE CASE FOR MARK II PLANTS BASIS ASSESSMENT CRITERIA 10 N+SRV +SSE+DBA Cor'cainment only Containment only C X (GE Justification (GE Justification submitted) submitted) N s W Y w e M U s N bb aE 8
- Load combination not bounding.
"s
[8
** SRVggg represented by de envelope of SRV gLL and SRV ASYM
O O O TABLE 2.2-3 DRYWELL PIPING ASSESSMENT: COMPARISON OF PIPING SUPPORT LOAD MAGNITUDES (UPSET-B) PIPE OBE + OBE + OBE + MANUFACTURER'S SIZE TYPE AND SRVALLRH SRVALLTQ SRVAsyTQ SNUBBER RATING SUBSYSTEM (in) LOCATION DIRECTION (lb) (lb) (lb) (lb) WR-02 4 35B RIGID GY 1209 1071 984 5800 WR-02 4 130B SNUBBER SX 269 417 279 3000 WR-02 4 96 RIGID SZ 1089 1426 1010 8000 Y F E N RT-1A 6 200B SNUBBER SX 8685 2321 3542 50000 g RT-1A 6 135 SNUBBER SX 360 601 949 3000 [ RT-1A 6 ll5A SNUBBER GX 1518 921 930 3000 0 m HP-01 10 S21 SNUBBER SZ 2914 2215 1968 20000 HP-01 10 S22 SNUBBER SZ 2606 2022 1733 20000 HP-01 10 S45 SNUBBER SZ 3783 2744 2196 10000 RH-01 18 725A SNUBBER SZ 27771 19636 20093 50000 b$ 89 RH-01 18 739A SNUBBER SX 33661 25025 27215 70000 @@ wm RH-01 18 790A SNUBBER SZ 24751 15240 16170 50000 ~$
$c
O O O TABLE 2.2-4 DRYWELL PIPING ASSESSMENT: COMPARISOr, OF PIPING SUPPORT LOAD MAGNITUDES (EMERGENCY-C) PIPE 1.8750BE+ SSE+ CHUG + MANUFACTURER'S SIZE TYPE AND SRVALLRH SSE+CO(2-7) SRVAggTQ SNUBBEK RATING SUBSYSTEM (in) LOCATION DIRECTION (lb) (lb) (lb) (lb) WR-02 4 35B RIGID GY 1315 944 1440 5800 WR-02 4 130B SNUBBER SX 434 607 430 3000 WR-02 4 96 RIGID SZ 1519 2212 1654 8000
" 7 w RT-1A 6 200B SNUBBER SX 9726 1562 4244 50000 5 x
b H X RT-1A 6 135 SNUBBER SX 2644 431 1172 3000 H RT-lA 6 llSA SNUBBER GX 2040 443 954 3000 y w HP-01 10 S21 SNUBBER SZ 4184 2230 3170 20000 HP-01 10 S22 SNUBBER SZ 3711 1999 2893 20000 i HP-01 10 S45 SNUBBER SZ 4950 1764 3286 10000 RH-01 18 725A SNUBBER SZ 41823 21237 32550 50000 8$ RH-01 18 739A SNUBBER SX 53321 24436 30420 as 70000 gg xm RH-01 18 790A SNUBBER SX 34838 11195 19778 50000 gy 5C i
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 TABLE 2.2-5 DRYWELL PIPING STRESS ASSESSMENT (Piping Stress Summary) NUMBER OF POINTS TOTAL NUMBER LOAD EXCEEDING OF POINTS COMBINATION ALLOWABLES EVALUATED OBE + SRV g RH 0 1471 1.875 OBE + SRVgg RH 0 1471 OBE + SRV TO 0 1471 OBE + SRVggy TO 0 1471 SSE + CO(2-7) 1* 1471 SSE + CHUG + SRV 1* 1471 ADS Q
./
()
- Expect no problem by using more refined analysis 2.2-12
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 TABLE 2.2-6 O PIPING OVERSTP"jS COMPUTED STRESS SUBSYSTEM CO (EL) + SSE ALLOWABLE % NUMBER LOCATION + SRVgggTQ STRESS OVERSTRESS WR-02 100A 40949 27000 51.7 100B 37221 27000 37.9 100A 34114 27000 26.3 100B 33857 27000 25.4 ll1A 32697 27000 21.1 lllB 32259 27000 19.5 130B 33518 27000 24.1 180B 33475 27000 24.0 295 37511 27000 38.9 300 34081 27000 26.2 310 54117 27000 10G.4 315 32813 27000 21.5 320 38361 27000 42.1 340 42463 27000 57.3 350 36273 27000 34.3 WR-01 115 32102 27000 26.3 t RT-OlA 280 53634 38925 98.6 1 2.2-13
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 TABLE 2.2-7 PIPING STRESS
SUMMARY
LOAD NUMBER OF POINTS TOTAL NUMBER OF COMBINATION EXCEEDING ALLOWABLES POINTS EVALUATED ABS SRSS CO (EL) + SSE
. + SRV TQ 17* 3* 1471 I
l l O
- Overstress can be resolved by:
- a. adding restraints,
- b. more refined analysis,
- c. less conservative load, or C' d. actual material properties.
2.2-14
. ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 l
TABLE 2.2-8 i LOAD COMBINATIONS EVALUATED FOR THE WETWELL PIPING N + OBE + SRV UPSET (B) N + DBE + CO EMERGENCY (C) N + DLE + SRV + CHUG EMERGENCY (C) WHERE: SRV - Drag - T-quencher Air Clearing Load Inertia - Rams Head 1 CO - Drag - CO(2-7) + Additional Consideration Is Up To 21 Hz () Inertia - Modified High Frequency CO CHUGGING (20-30 Hz) , - DBE - 1.875 OBE b 2.2-15
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. . - . = . . . .. . . - -
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WM. H. ZIMMER NUCLEAR POWER STATION UNIT 1 , MARK 18 DESIGN ASSESSM ENT REPORT 4 FIGURE 2.2-1 4 Q RESPONSE SPECTRA COMPARISON - OBE + SRV ALL, OBE + SRV ' TQ RH' OBE + SRV ASY- ; TQ i
,-. . - , , - - , - - , - L
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.001 0.004 0.006 0.010 0.020 0.040 0.050 0.*00 0.t00 0.400 0.tr30 .0 0.0 PER!dD IF SECeNO3 201-V WM. H. ZIMMEF. NUCLEAR POWER STATION. UNIT 1 MARK 16 DESIGN ASSESSMENT REPORT FIGURE 2.2-2 o)
(" RESPONSE SPECTRA COMPARISON - DBE + SRV pg VS. N + C0(2-7 Hd
4 HORIZONTAL EXCITATION AMENDMENT 13 r.Exz;cy tu t: 3 OCTOBER 1980
.J.J t'0 . *J lW.3 *
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.( JZ 0.004 0.009 0.010 0 020 C.0+0 0.tM*0 0.100 0.2"JO 0.400 0.*.D0 80 f.0 PE.TIOD IN SEc3ND3 SET-2 ZIti-1 SPECTRR 115-H WM H. ZlMMER NUCLEAR POWER GTATION. UNIT 1 MARK 11 DESfGN ASSESSMENT REPORT FIGURE 2.2-3 k_
RESPONSE SPECTRA COMPARISON - DBE + SRV RH VS. ALL ALL N C0(EL) + SSE + SRV
AMENDMENT 13 rReat;tucv tw cra OCTOBER 1980 ese.e 200.0 50.0 s.0
.00 ... . .....,. 100.0 ... . . . . .
to.o lo.o
' t.e 4..... ,..
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'. .00t 0.*Q4 0 000 0.010 0.0t0 0.040 0.09,1 0.100 0.000 5 0.400 0.000 1.0 f.e rtRISD IN SEC8WD3
- 104-V DBE+SRV-RH VS SSE+ CHUG +SRVALL Whl. H. IIMMER NUCLEAR POWER STATION, UNIT 1 M ARK - il DESIGN ASSESSMENT REPORT FIGURE 2.2-4 p RESPONSE SPECTRA COMPARISON -
V DBE + SRV vb' RH SSE + CHUG +-SRV ALL _ _ - . . . _ . - _ , . . . . _ , . . . , _ _ _ _ , . _ , . _ _ - _ , _ . ~
AMENDMENT 13 NO. OF SUPPORTS OCTOBER 1980 HISTOGRAM SHOVS! 70 HOW MANY SUPPORf> PHANGED THEIP LOAD BY "X" rERCH.T ? 60 _ ALL LOAD INCREASES 50 ARE WITHIN $NUBBER RATING 40 30 29 10 l i O -80 60 40 20 -10 10 20 30 40 50 60 5 LOAD DECREASE ~~ I' LOAD INCREASE DP.YWELL PIPING SUPPORT LOAD CHANGE OBE + SRV RH VS. OBE + SRV TQ ALL ALL t
% CHANGE = 0 LOAD COMB. - RH LOAD COM. x100 RH LOAD COMB.
FOP. EXAMPLE: IF: RH = 1000 # THEN: 7 CHAFGs = 500 - 1000 x 100 1 Te = 503 # ^000
% CHANGE = -50%
WM. H.IIMMER NUCLEAR POWER STATION. UNIT 1 MARK II DESIGN ASSESSMENT REPORT I FIGURE 2.2-5 SUPPORT LOAD CHANGES - ALL E OBE + SRV RH VS. OBE + SRV ALL l i I
r
- d. F ,<x $ A0 g 5 N<xm
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AMENDMENT 13-
'NO. OF SUPPORTS OCTOBER 1980
, A 70 A , V t 60 ALL LOAD INCREASES
! ARE WITHIN SNUBBER f 50 RATING 40
' 30 20 10 O
- 60 40 20 -10 10 20 30 40 50 60
% DECREASE
. % INCREASE DRYWELL PIPING
- ' SUPPORT LOAD CHANGE 1
OBE + SRV RH VS. OBE + SRVASY TQ ALL
% CHANGE = 10 LOAD COMB. - RH LOAD COMB. .
RH LOAD COMB. i WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK ll DEJIGN ASSESSMENT REPORT , + j FIGURE 2.2-7 DRYWELL. PIPING SUPPORT LOAD CHANGES -- OBE + SRV ALL RH VS. OBE + SRV ASY TQ 4
NO. OF SUPPORTS AMENDMENT 13 J OCTOBER 1980
- 70
__ 60 ALL LOAD INCREASES ARE WITHIN SNUBBER
-- 50 RATING
- 40
-- 30 4
_. 20
-- 10 l ! +-
1 ) 80 60 40-30 30 10 20 30 40 50 60 70 80 90
% DECREASE ' *
% INCREASE DRYWELL PIPING SUPPORT LOAD CHANGE 1.875 OBE + SRVALL RH VS. SSE + C0(2-7)
% CHANGE = TO LOAD COMB.-RH LOAD COMB.
4 RH LOAD COMB. WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT Q FIGURE.2.2-8 DRYWELL PIPING SUPPORT LOAD ' '?.dGES - 1.875(0BE) + SRVALL RH VS. SSE + C0(2-7)
- AMENDMENT 13 E OCTOBER 1980 aH
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C 3 O, 0 30 'ON 3SY333NI e-- 3sy33330 WM. H. ZIMMER NUCLEAR POWEP STATION, UNIT I l MARK 11 DESIGN ASSESSMENT REPORT l iO l w) FIGURE 2.2-9 ' i SUPPORT LOAD CHANGES - 1.875(0BE) + SRV ALL RH VS. SSE + C0(2-7)
NO.0F SUPPORTS AMENDMENT 13 OCTOBER 1980
,1 O
60 ALL LOAD INCREASES ARE WITHIN SNUBBER
~
RATING 40 30
--20
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70 50 -40 20 -10 10 20 30 40 50
% DECREASE < ;
% INCREASE DRYWELL PIPING SUPPORT LOAD CHANGE 1.875 OBE + SRVALL RH VS. SSE + CHUG + SRV TQ ADS
% CHANGE = TO LOAD COMB. - RH LOAD COMB.
RH LOAD COMB. WM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT FIGURE 2.2-10 DRYWELL PIPING SUPPORT LOAD CHANGES - 1.875(0BE) + SRVALL RH VS. SSE + CHUG + SRV ADS E
AMENDMENT 13 0 a. 0CTOBER 1980 OH
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-, _ _ _ . _ . . _._ . -- . . ~ . _ . __ . - - . _ . . . . . . . ._.. ,_,
)
AMENDMENT 13 OCTOBER 1980 rnEcuEHcy tw crs seo.o too.o 100 0 50.o ro.o r.o r.o 8 00 ,, , , , , , , , ,,, , , , , , to.o
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-r 1 e m .- >,-e--- --
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NO. OF SUPPORTS AMENDMENT 13 OCTOBER 1980
.A
--70
_ 60 J
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- -40
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- O
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- l ;
-100 80-70-60-50-40-30-20-10 10 20 30 40 50 60 70 80' 9010'0 n0 i
30 150 120 140 j % DECREASE %-INCREASE
, 7-DRYWELL PIPING L
SUPPORT LOAD CHANGE 1.875 OBE + SRV RH vs. SSE + CO( EL ) + SRVADS
% CHANGE = 0 LOAD COMB. - RH LOAD COMB. X100
, RH LOAD COMB.. i l WM. H. ZIMMCR NUCLEAR POWER STATION. UNIT I MARK ll DEStGN ASS ESSM ENT REPORT ^
-FIGURE 2.2-13 O oavwEtt eietNo Sueroar toAo Chance - ..
1.875(OBE) + SRVALL RH VS. l SSE + C0(EL).+ SRVADS E +
,-, ..r.- , ., .
a- ., , . ,, -- . - , , - - - , - , . . , - ,.~e -- , , - - - - - - - - - , -
AMENDMENT 13
; OCTOBER 1980 an SE O- Mo
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l- J l ! 1 l m l lN l l il r - O 1 Ii 1F12 O I ! l $- SIV0ddnS S U 3 , S. 0 $ 30 *0N 2SV3H3NI e* 3SV3133G WM. H. ZIMMER NUCLEAR POWER STATION. UNIT I MARK 18 DESIGN ASSESSMENT REPORT )
. FIGURE 2.2-14 SUPPORT LOAD CHANGES -
1.875(OBE) + SRV ALLRH VS. SSE + C0(EL) + SRVALL TQ
d AMENDMENT 13 OCTOBER 1980 DRYWELL SUPPORTf' DESIGN MARGIN AVAILABLE O Number of. Supports' Upgraded' Design. Load Capacity
- 78
, ll.l.. -l-. ; ,k l !,Ie i
18
- fjl,l, j llj 10%
39 I'l j ,
,j,,
l . 20% I/j, f 'i
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35 j ,ij l j' ./l 'l 40% t i 36 [,' i //'/. u/ / 50% 19 /
., / -60%
! O 3
,/ /
j /, 70% 1 '
/ / / /
TOTAL 269 TQ** Margin Available , Load Over Empirical Load
- I
- Upgraded Design. Load Capacity Includes The Largest Of:
- 1) Upgraded Ramshead Design Basis Load Combination
- 2) CO.(EL) Load Combination
- i i
- **TQ Load = Includes The Largest Of
i
- 1) NRC Acceptance Criteria Using T-Quencher Loads j 2) CO (EL) Load Combiliation l
WM. H. ZIMMER NUCLEAR POWER STATION. UNIT I ' l MARK li DESIGN ASSESSMENT REPORT 4 FIGURE 2.2-15 DRYWELL SUPPORTS AVAILABLE DESI N MARGIN 1
- , . . - - ~ , - . , , . -4,-y ,<ai,.e,..-, - - - , , - ..,...,,-p-. -
, , , , , - ,- ,,,,,-w,-.,...-.,,y~.v,.,w
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2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 2.3 BALANCE OF PLANT EQUIPMENT () g, An assessment of safety-related balance of plant equipm ..t has been performed to evaluate the impact of the Zimmer Empirical Loads. The results of this assessment were presented to the NRC Staff on December 5, 1979, and are summarized in this section. 2.3.1 Assessment and Requalification Procedure The balance of plant equipment was originally qualified by a program of dynamic testing, analysis, and a combination of test and analysis. This assessment was performed by evaluating the new loads against the design-basis loads included in the existing qualification documentation. 2.3.1.1 Procedure for Equipment Originally Oualified by Testing
- a. New required response spectra curves were generated by combining the individual response spectra to obtain cne set of curves for each new loading combination.
- b. The design-basis curves were compared against the new curves.
r~ c. Where the new curves exceed the design-basis curves,
\_T/ requalification will consist of additional analytical work to supplement the testing in order to demonstrate adequacy.
- d. 11 additional analytical work is not possible or fails to satisfy the acceptance criteria, additional testing will be performed. Limited scope testing, to supplement existing tests, will be considered before complete requalification testing,
- e. If qualification cannot be adequately demonstrated, the component will be modified or replaced.
2.3.1.2 Procedure for Equipment Originally Oualified by analysis
- a. .New required response spectra curves were generated by combining the individual response spectra to -
obtain one set of curves for each new loading combination.
- b. Based upon the nature of the new curves, the validity of the model and methodology used in the original qualification was checked.
l n Y) i i 2.3-1
a l 2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
- c. If the model and methodology were-still valid, the new loads were compared against the design-basis loads.
f
- d. Whenever the design-basis loads have been exceeded, requalification will consist of repeating the analysis for the new loads.
- e. If the model or methodology are no longer valid for the new curves, requalification will consist of revising, as required, and reanalyzing for'the new loads. In all cases, the acceptance criteria (described in Subsection 6.4.3.2) must be satisfied.
- f. If qualification cannot.be adequately demonstrated, the component will be modified or replaced.
2.3.2 Hioh and Low Frequency Concerns The impact of the high and low frequencies, associated with the new SRV and LOCA loads, 'ere addressed by surveying the existing qualification reports to find where the egaipment natural frequencies occur. It was found that, with the exception of equipment mounted on vibration isolators, no equipment has natural frequencies below 5 hertz, so the equipment will not be (~) sensitive to the low frequency loads. The exceptions will be handled by additional analytical calculations. 1 The impact of the high frequency loads is presented in Subsection 2.3.4. The basis of the review for evaluating the impact of the-high frequency loads was determined by the extent end method of existing qualification. It was found that some equipment has sufficiently high natural frequencies and consequently is not sensitive to the high frequency loads. Where this was not found to be the case, the acceptance criteria was based upon method of qualification. When qualification was by. analysis, additional moden were considered to cover the extended frequency range and i the basis for accepting static analysis was reviewed and shown to
- still~be valid. When the original qualification was by testing, j it was shown that the tested 'g' values were sufficiently high to take into account pcssible modal participation at higher frequencies. Additional in situ testing planned to address this sub]ect is described in Subsection 2.3.3.
2.3.3 In Situ Testing The in situ testing program will be used to confirm the conclusions of-the high and low frequency concerns of the equipment requalification program. This program consists of the in situ SRV testing and in situ-impedance testing. During the SRV test, selected pieces of equipment will be monitored to s determine the amplitude and frequency content of the input to the equipment and the corresponding equipment response. For the [ 2.3-2
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980'
<- impedance testing, selected pieces of equipment will.be tested as
(_)/ installed to determine their natural frequencies and mode shapes. To accomplish this, the equipment will be excited at as many locations as necessary. The input will be of sufficient intensity as to excite all significant modes in the frequency range of 1 to 100 hertz. The response will be measured at locations deemed necessary to detect natural frequencies and mode shapes. 2.3.4 Equipment Foundation Loads The equipment foundation loads for all balance-of-plant equipment (safety-related and non-safety-related) located in safety-related
-structures are being recalculated and the adequacy of equipment anchor bolts or welds, equipment foundation, and floor slab is being demonstrated.
2.3.5 Results of Equipment Assessment 2.3.5.1 Valve Qualification Assessment Of the 572 safety-related valves affected by the new SRV and LOCA loads, 143 were studied to evaluate the impact of the new loads. The basis of this study was to compare piping accelerations against the accelerations for which the valves were qualified. Cases where the piping accelerations exceed the valve qualified
) accelerations have been identified as requiring further action.
It is important to point out that this does not imply that the valve is inadequate, but rather that the. existing documentation does not demonstrate its adequacy. The results of the study are summarizea relow: NUMBER NUMBER FURTHER VALVE TYPE ACCEPTABLE ACTION REQUIRED Manual Operator 55 16 Motor Operator 20 23 Air Operator 2 8 4 Check 7 5 Relief 6 1 TOTAL 90 53 2.3.5.2 Equipment and Instrumentation Assessment I A total of 130 pieces was studied for the various loading combinations using both the absolute sum method and the square
- root of the sum of the squares method. The basis of this study l
is discussed in Subsection 2.3.1. The results are summarized below: O l 2.3-3 I l l
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 r3 COMBINATION METHOD (/ NUMBER WHICH NUMBER WHICH NUMBER NUMBER MAY REQUIRE MAY REQUIRE IN LOAD COMBINATION A_CCEPTABLE REANALYSIS RETEST PROGRESS ABS SRSS ABS SRSS ABS SRSS N+0BE+SRV 3g 126 128 2 0' 0 0 2 N+DBE+SRV 128 128 0 0 0 0 2 Agy 126 2 2 0 0 2 N+SSE+CO(DFFR def. ) 126 N+SSE+SRVADS+CO (Zimmer empirical) 100 110 21 11 4 4 5 N+SSE+SRVAng+CHG 104 110 17 11 4 4 5 I 130 130 N+ %(SSE)2+(AP)2 0 0 0 0 0 0 O 2.3-4
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 2.4 STRUCTURAL ASSESSMENT The Zimmer Empirical Loads have been used to assess the structure of the William H. Zimmer Power Station. This assessment includes the primary containment, drywell structural steel, the downcomer bracing system and pedestal straps supporting MSRV and non-MSRV piping. 2.4.1 Method of Assessment The assessment was done in accordance with the load combinations listed in' Chapter 6.0. These load combinations were considered conservatively, as explained in Subsection 2.1.7. As noted in this subsection, the SRSS method of load combination was not used. . 2.4.2 Primary Containment As a result of the assessment, the containment wall was found to be adequate for the Zimmer Empirical Load. A smali overstress (approximately 5%) was found in the basemat. The basemat design is adequate to accommodate loads which meet the NRC Lead Plant Acceptance Criteria (NUREG-0487). The pedestal and drywell floor are being evaluated for both the Zimmer Empirical Load and the Lead Plant Acceptance Criteria. 2.4.3 Drywell Structural Steel [} Studies of the piping support loads indicate that use of the Zimmer Empirical Loads will generally not increase the loads more than 33% above the original loads. Assessments of the drywell structural steel r c ce then done with the loads increased by this 1.33 factor. Although a significant portion of the drywell i support steel was found to be not adequate, the capability could be adequately increased by modifications. It is estimated that about 30% of the beams in the drywell will be modified or changed. When complete, these modifications will raise the drywell steel capability almost to that of the primary containment. The adequacy of the 1.33 factor is being verified by application of the Zimmer Empirical Load. Figures 2.4-1 and 2.4-2 show the actual load increases or decreases. The majority of' pipe support loads are decreased or increased less than 33%. Load increases greater thcn 33% are individually checked to verify the adequacy of the design but advantage is not taken of load reduction. This procedure results in increased design margin for the drywell steel. It should be again noted here that the piping reaction loads used were combined by the absolute sum method rather than the SRSS method. 2.4'.4 Downcomer Bracino System The Zimmer Empirical Loads contain increased low frequency loads (]) in both the SRV and condensation oscillation loads. This change 2.4-1
\
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 had a significant effect on the original unbraced downcomers ({} which had a relatively low natural frequency. In order to accommodate the Zimmer Empirical Loads and also to conform to the NRC Lead Plant Acceptance Criteria, a bracing system has been designed and installed near the pool surface. The bracing system is shown in Figures 2.4-3 and 2.4-4. This system required significant_ changes in the suppression pool, including additional embedments in the containment wall (Figure 2.4-5), installation of beams in the pedestal for bracing supports (Figure 2.4-6), and re-routing of wetwell piping. 2.4.5 Pedestal Straps Supporting Piping The Safety Relief Valve (SRV) piping required complete re-routing when the rams heads were replaced by quenchers. The quenchers were rearranged from the original rams head positions to minimize containment loads and maximize pool mixing. Installation of the bracing required additional re-routing of the SRV and other piping in the suppression pool. To support these pipes, straps, as chown in Figure 2.4-7, were installed around the pedestal. O 2.4-2
- 1. _ - _
NO. OF SUPPORTS AMENDMENT 13 OCTOBER 1980 h (^) v
-- 7 0 60
--50
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% DECREASE , 7-. % INCREASE DRYWELL PIPING SUPPORT LOAD CHANGE 1.875 OBE + SRV RH vs. SSE + CO( EL ) + SRVADS
% CHAliGE = TO LOAD COMB. - RH LOAD COM .
RH LOAD COMB. X100 WM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 11 D E SIG N ASSESSMENT REPORT FIGURE 2.4-1 DRYWELL PIPING SUPPORT LOAD CHANGE - 1.875(OBE) + SRV ALL RH VS. SSE + C0(EL) + SRVADSTQ
Om 9Q VW3NOW3N1 iE E3 031083B L680 m _ xr l
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m , .,__a - - -- - - - - - - - , AMENDMENT 13 0CTOBER 1980 O 18 0
- ING MEMBERS 175'26 0 XXS PtPE (AIO6 GR B)(TYR)
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' l o' WM. H. ZlMMER NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASSEGSM ENT REPORT l FIGURE 2.4-3 DOWNCOMER BRACING LAYOUT
AMENDMENT 13 OCTOBER 1980
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AMENDMENT 13 0CTOBER 1980 O U . i.
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WM. H.2IMMER NUCLEAR POWER STATION, UNIT 1 MARK la DESIGN ASSESSMENT REPORT (m FIGURE 2.4-5 CONNECTION OF BRACING TO WALL EMBEDMENT
AMENDMENT 13 OCTOBER 1980 18 0 ' ( 1 TYPlCAL BR ACING
~ ~ ~
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AMENDMENT 13 OCTOBER 1980 824 OR B24V CLASS I W/HEWY HEX NUT A 194 O j 4<Asn.aR.r> . i < A-si7. oR. r> e---
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AMENDMENT 13 8*9 XXS PIPE ( T YP. ) (AIO6 GR. B )
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I
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 2.5 NSSS EQUIPMENT (/ The NSSS equipment was originally designed for pressure loads, thermal loads, and seismic loads. Significantly, after the original design was completed, pool dynamic loads were identified. These loads were associated with SRV and LOCA phenomena. New building response spectra and LOCA response spectra were generated, as well as dynamic loads associated with annulus pressurization. The NSS.. equipment and piping were reassessed for the combined effect of the original and additional new loads. These were presented to tha NRC in November 1978. Data became available in 1979 from domestic and foreign tests for which the applicant made a decision to upgrade the plant design basis as well as update the reassessment to reflect the installation of SRV T-quenchers. The results of the preliminary reassessment for the combined SRV T-quencher loads and Zimmer Empirical CO Load for the NSSS were also presented with the BOP assessment on December 5, 1979. Table 2.5-1 summarizes the three different cases evaluated assuming various acceptance criteria and method of load combination (SRSS and Absolute Sum-ABS). Table 2.5-2 is a summary of load case definitions used in developing Table 2.5-1. Table 2.5-3 briefly summarizes the results of previous assessments for SRV rams head and earlier LOCA defined loads. (') Tables 2.5-4 and 2.5-5 summarize the results for the RPV, RPV service equipment, and NSSS safety-related components, respectively. The preliminary assessments show that the RPV and RPV service equipment can accommodate the most current KWU, SRV T-quencher loads, and the Zimmer Empirical CO Load in both Case A & B combinations. The only overload identified for the RPV internals is the top guide hold-down latch for which a fix in is process. NSSS instrumentation and floor mounted equipment is being evaluated. It is expected that additional dynamic analysis will demonstrate adequacy for the increased loads. In addition, as noted in Table 2.5-5, the ECCS pumps will be " beefed up" to provide additional margin. A preliminary assessment of the reactor recirculation, piping, main ; team piping, and associated pipe mounted equipment is summarized in Table 2.5-5. The conclusion reached is that these components can accommodate the conservative Zimmer design loads. Tables 2.5-6 and 2.5-7 list the main steam and recirculation system snubbers, rating, previous governing load combination, and previous and current margin. In summary, the NSSS systems design adequacy has been updated to reflect the final design and to provide increased margins. The evaluation was made using conservative criteria as applied to load definitions and accentance criteria. Corrective action is being taken to increase design margin for the ECCS pump / motors {) 2,5-1
~
j 2PS-1-MARK II DAR AMENDMENT.13 OCTOBER'1980 and the.RPV top guide hold-down latch. The final design adequacy
- -Q analysis'to formalize the reassessment is in progress and will be completed before fuel loading. -
i l I 4 I i 1 i , i .I i i, lO 4 i i l I s i I l 4 i 2 l s t t l 4-4, O i 2,5-2
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ZPS-1-MARK Il DAR AMENDMENT 13 OCTOBER 1980 TABLE 2.5-1 /~N (_)
SUMMARY
OF LOAD CASES FOR EQUIPMENT FOR STUDY PURPOSES CASE OBE,OT OBE+OT SSE+OT SSE+LOCA SSE+AP AP Design B-SRSS C-SRSS D-SRSS D-SRSS ------ D-SRSS Case A ------ B-SRSS D-SRSS D-SRSS D-SRSS ------ Case B ------ B-ABS D-ABS D-SRSS D-SRSS ------ ACCEPTANCE CRITERIA A - Normal B - Upset C - Emergency D - Faulted C) O i (,/ 2.5-3
s ZPS-1-MARK II-DAR AMENDMENT.13 OCTOBER 1980
- TABLE 2.5-2 1
k-)o LOAD CASE DEFINITIONS 1 OPERATING TRANSIENT (OT) Structural Response To SRV Discharge Acoustic Load Due to SRV Discharge Acoustic Load Due to Turbine Stop Valve Closure i-i LOSS OF COOLANT ACCIDENT (LOCA) Small/Large/ Intermediate Breaks SRV ADS For Small/ Intermediate
- Chugging Condensation Oscillation Vent Clearing ANNULUS PRESSURIZATION (.AP)
Annulus Pressurization Jet Loads i O 2.5-4
1 ZPS-1-MARK II DAR AMENDMENT 13 l OCTOBER 1980 TABLE 2.5-3 /^ ^N, (_ ) REVIEW OF PREVIOUS RESULTS
- Reactor Pressure Vessel OK
*RPV Internals OK
*RPV Equipment OK
- Floor Mounted Equipment OK
' Piping And Pipe Mount'ed Equipment OK (Limited Overloads Justified) f- s Y_Y l
l l l l k u.i 2.5-5 l l
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 TABLE 2.5-4 ([) NSSS SAFETY-RELATED COMPONENTS ASSESSED (Preliminary) COMPONENT CASE A CASE B RPV Skirt OK OK Shroud Support OK OK Stabilizer Bracket OK OK Steam Dryer Bracket OK OK RPV INTERNALS Core Plate OK OK Top Guide OK* OK* Phroud OK OK ('h k/ Fuel Assembly OK OK CRD Guidelines OK OK CRD Housing OK OK Jet Pump Assemblies OK OK Core Spray Lines OK OK RPV EQUIPMENT Fuel Storage Racks OK OK Refueling Platform OK OK Stabilizer OK OK CRD Restraint Beams OK OK /'N
- Holddown Latch Overloaded - Fix In Process (J
2.5-6
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 TABLE 2.5-5 t) NSSS SAFETY-RELATED COMPONENTS ASSESSED (Preliminary) COMPONENT CASE A, CASE B NSSS INSTRUMENTATION A FLOOR MOUNTED EQUIPMENT Evaluation Underway RCIC Pump 1. Dynamic analysis will be performed RCIC Turbine to increase allow-loads. RHR Pump and Motor
- 2. ECCS pumps will be RHR Ex . beefed up to pro-vide additional Standby Liquid Control margin, if required.
Pump and Tank HPCS Pump and Motor ( )) LPCS Pump and Motor Flammability Control Equipment PIPING AND PIPE MOUNTED EQUIPMENT Recirculation Piping System OK OK Recirculation Pump And Valves OK OK Main Steam Piping system OK OK Main Steam Safety Relief Valves OK OK Main Steam Isolation Valves OK OK n 2.5-7
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 TABLE 2.5-6
- ( ZIMMER MAIN STEAM SYSTEM i
CALCULATED ONUBBER LOADS PREVIOUS PREVIOUS CURRENT RATING GOVERNING MARGIN MARGIN SNUBBER (kips) COMBINATION % % SA2 30 SSE + RV 63 60 2 SA3 30 SSE + AP 58 58 SA4 36 SSE + TSVC 45 45 SAS 30 SSE + TSVC 51 51 SA6 30 SSE + TSVC 36 36 SA7 30 SSE + AP 73 73 SA8 45 SSE + AP 21 21 SA10 30 SSE + AP 19 19 () SAll 30 SSE + RV 2 49 44 SA1 54 SSE + TSVC 79 79 SA9 36 SSE + RV 2 56 52 2.5-8
AMENDMENT 13 ZPS-1-MARK II DAR OCTOBER 1980 TABLE 2.5-7 () ZIMMER RECIRCULATION SYSTEM 4 CALCULATED SNUBBER LOADS PREVIOUS CURRENT PREVIOUS MARGIN MARGIN RATING GOVERNING SNUBBER (kips) COMBINATION 58 58 SA24 30 SSE + RV 2 61 55 SA25 30 SSE + RV 2 58 52 SA22 30 SSE + RV 2 63 58 SA23 30 SSE + RV 2 50 43 SA8 75 SSE + RV 2 51 43 SA9 75 SSE + RV 2 79 76 SAll 75 SSE + RV 2 70 70 SA14 105 SSE + AP 60 60 SA2 75 SSE + AP 60 60 sal 90 SSE + AP 65 60 SA17 75 SSE + RV 2 69 69 SA18 75 SSE + RV 2 O (d l i 2.5-9 l
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
2.6 CONCLUSION
S O The Zimmer Empirical Load provides a conservative basis to continue the construction and licensing of the Zimmer Power Station. The approach taken includes adequate conservatism to accommodate any load increases which may be required due to test data or other information which is not now available. As a result of the conservative approach taken, extensive modifications and additions have been made to the wetwell, drywell, and reactor building. In many cases, this has resulted in an upgrading of the plant capability to that of the containment itself. This work was undertaken with the purposes of avoiding costly and time consuming delays in the plant operation and to ensure that the plant design is as safe as possible. This chapter demonstrates that the Zimmer Empirical Load Approach is an adequate basis to allow the continued construction and licensing of the Zimmer Power Station and has sufficient conservatism to account for any uncertainty in the load. O W 1 2.6-1
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 0 CHAPTER 3.0 - SRV IN-PLANT TEST PROGRAM
3.1 BACKGROUND
The William H. Zimmer Station - Unit I was originally designed i i with rams head type safety / relief valve (SRV) discharge devices. i After new pool dynamic loads were identified, the plant designs ' were reevaluated and modifications implemented to enable the DFFR rams head load definitions to be accommodated. ' i After a large portion of the reevaluation effort had been i completed, a devision was made to replace the rams head SRV dis- , charge devices with T-quencher devices. This decision was based upon tests that indicated that the T-qaancher exhibits better steam condensation stability at higher pool water temperatures than the rams head devices. It is also expected that the T-quencher discharge will result in l' loads considerably below the loads used to assess the plant (Zimmer Empirical Loads). The results of the test will serve to quantify this conservatism. () 1 f i 1 0 3.1-1
i 2PS-1-MARK II DAR AMENDMENT 13 ! l OCTOBER 1980 i . 3.2 PURPOSE 1 The objectives of the SRV in-plant tests are (1) to demonstrate that the plant structural response with T-quencher devices is not greater than the structural response defined in Chapters 3.0 and 4.0 of the Design Assessment Report, and (2) to demonstrate that adequate thermal mixing of the suppression pool will occur during quencher discharge. The test has been designed to provide plant , structural response data with additional instrumentation added for hydrodynamic data necessary to extend the plant structural response data from those load cases tested to other design load conditions. A secondary objective of the test.is to provide the j suppression pool temperature distribution during an extended SRV r discharge. Another objective of the test is to verify that high-frequency oscillations in the forcing function will not be amplified in the building structure or the floor-mounted equipment. This portion of the test is provided in response to the request of the NRC seismic qualification review team (SQRT) for in situ testing. Table 3.2-1 is a list of the equipment to i be instrurented for the SORT in situ test. I ($) 4 i 1 i a i 8 3.2-1
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 ( TABLE 3.2-1 LIST OF EQUIPMENT BEING MONITORED DURING IN SITU SRV TEST RADIAL EQUIPMENT ELEVATION AZIMUTli DISTANCE HPCS PUMP 475 ft 6 in. 315' 67 ft (1E22C001A) RCCI INSTRUMENT PACK 1A 475 ft 6 in. 315' 63 ft 4-(lHilP017) 20-inch HPCS GATE VALVE 481 ft 6 in. 220* 47 ft (lE22F015) 480V RX MCC lE 525 ft 7 in. 315 82 ft (LAP 15EE) REACTOR RECIRCULATION 525 ft 7 in. 315* 22 ft PUMP 1A (1833C001B) RX VESSEL LEVEL AND PRESSURE 546 ft 0 in. 220' 63 ft INSTRUMENT RACK D (lH22P027) 3-inch STANDBY LIQUID 571 ft 8 in. 206* 60 ft CONTROL GLOBE VALVE (lC41F001A) 1 0 3.2-2 j i
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (8 \-) 3.3 TEST
SUMMARY
The primary objective of this test program is to demonstrate that the plant structural response with T-quencher devices is not greater than the structural response defined in Chapters 3.0 and 4.0 of the Design Assessment Report. The specific areas of the plant to be instrumented are as follows:
- a. SRV discharge lines,
- b. T-quencher and T-quencher supports, C. downComers,
- d. drywell floor support column,
- e. suppression pool liner,
- f. reactor building,
- g. drywell floor,
- h. reactor shield wall and pedestal,
(~') i. HPCS suction line, and w/
- j. suppression pool for thermal mixing characteristics.
Test instrumentation will be installed to measure structural response and pressure loads generated by SRV discharge. Single, consecutive, multiple and extended SRV actuations will be included in the test. Most of the instrumentation in the suppression pool will be con-centrated near four quencher locations. Temperature sensors in the suppression pool will provide pool temperature distribution data. Strain gauges will be used to measure the effects of pressure loads and/or drag forces on downcomers, quenchers, quencher supports and the suppression peel liner. Pressure transducers and level probes in the SRV discharge lines will measure peak line pressure during and following SRV actuation. Furthermore, the level probes will provide discharge line water reflood information. Accelerometers will be mounted at various locations in the reactor building, and on the suppression pool floor, drywell i floor and reactor shield wall to measure the structural response due to SRV discharge. Additional accelerometers will be installed at locations of selected floor and pipe-mounted equipment. Strain gauges located on an SRV discharge line will (,) measure pipe strain during the water clearing, air clearing and steam discharge phases of the transient. 3.3-1
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 O Figures 3.3-1 through 3.3-10 show the actual sensor locations. Figure 3.3-1 illustrates the accelerometer locations. The suppressior, pool pressure sensors are shown in Figure 3.3-2. Figure 3.3-3 illustrates the suppression pool temperature sensors. Figure 3.3-4 shows some of the SRV discharge line temperature and pressure sensor locations. The suppression pool strain gauge locations appear in Figure 3.3-5. Figure 3.3-6 illustrates the locations of the SRV discharge line level sensors. Figures 3.3-7 through 3.3-10 show sensor locations on various submerged structures in the suppression pool. O 4 O 3.3-2 1
O O O
. TABLE 3. 3-1 l Initial Vessel Pressure: 985+30 Suppression Pool level: 497'9"+psig. 3" i ..
VALVE (s) VALVE TEST TTST TO BE DIS 0t W I CIDSURE PIPE WOLNG TD4E INITRVAL .! NLPSER 1WE AC11JA11" TIK ,SEC TIME "RIOR PRIOR 10 R!iR CIRGILATION 10 NErr TEST CCPHENTS SRV PIPE POOL Rh1R i ' II) WNDITIONS TUFfF)lmyt TEST , .) SE SE SEE OFF SDI SD 193*J013D) O,El(2) m4 M7TE 5 5 N/A M7TE 6 SEE M7TE 7 SLE MJTE 8 SD2 SD 220*(P' t') CP,NWL 5. 1 SD3- SD 257'(F013E) & ,NWL 5 g i o e"n l SD4 SD 270*(F013P) CP,%1 V 5 o V b '-w , !W Vril SVA 198* CP,Nh1 85 100 15 30:00 30:00 MIN. 4a ; MT12 CVA 198* WP.Ah1(3) 85-100 5 N/A 30:00 MIN. SE M7E 8 e SEE , MT21 SVA 220' CP.Nh1 85'00 15 00:10 M7FE 6 10 SEC. % i l MT22 CVA 220' IF,Ahl 85 100 5 N/A 10 SEC. SE NJTE 8 3 - SE j MT31 SVA 257' CF,N!" 85-100 5 N/A M7TE 6 SE M7E 8 i . SE ]. MT41 SVA 270* CP,Nkt 85 100 5 N/A M7E 6 SE M7TE 8 o SE oN '.I MT51 SVA 19B* O ,NWL 85-100 15 00:10 M7E 6 10 SEC. o$ to e my
- j. MTS2 CVA 198* IF,Ahl N 85-100 5 N/A 10 SEC. Y SE M7E 8 %
;. Additional tests shall be performed to pro-' gg I vide a statistical basis for the. limiting ow
- single relief valve actuation case. A pro-2 cedure is used to determine which of the 4
above matrix tests (MTil through !!T52) is j the limiting case. A
-- . - - . - . . . . - ~ , .w. ,
O O O Table 3.3 l (Cont'd) Initial Ve:sel Pressure: 985+30 psig TEST MATRIX Suppression Pool level: 4 97' a"+ 3" t VALVL VALVE (s) CIINTetE FSE COOLING TDtE IhWAL TEST TEST TO BL DISCilARGE TDIE PRIOR PRI0ft 1D PJF CIRCUIATION TD htXT 1T.ST CO MENTS NUllBER TYPE ACTUATED SRV PIPE TOOL IUkIR TIME,SEC TD CVA 11:ST (1) CONDITIONS TU!PfF) h pnN:STO _ SEF 70-85 199-5 A SET OFF PIT 61 )!VA 1980 ,2200 CP,NWL p g '4 .'200-10 NUE '6 SEE NUE 7 SEE NUE 8 MT71 MYA 1980,25"o CP.NWL 70-85 2 0 PtT81 FNA 25F .2700 CP,NWL 70 85 hh0 22@~5 o-N w MT91 FfV.% 220 .2570,270 C CP,NWL 70-85 gCP g r.n t g I !!T92 MVA 2200,2570,276 CP.NWL 70-85 h No h 27(P - 15 Ad t1 X 22(p-5 g 1tysts
!!T9 3 INA 2200 ,2570 ,276 CP,NWL 70-85 Qg Dischar y fir 101 f.ge timedeter MT102 for U
?!T101 ESVA 198 CP.NWL 85-100 See Cm ments mined by criteria h 0 CP NWL 85-100 MT102 ESVA 198 MTlll ESVA 2200 CP.NWL 65-100 See Coments gg O
y pa 1 de r mined by criteria E' MT112 ESVA 220 0 CP,NWL 85-100 y o p-1Til g AIRD (9) CP,NWL 85-100 VARIABLE VARIABLE VARIABLE Q% oz 1T21 AIRD (9) CP,NWL y 85-100 VARIABE y y VARIAB E VARIAB E g ,
*N: One RIR loop shall be operated during these tests in the circulation mode only. Cr7.E test director shall select the appropriate loop. y 8C 3
AMENDMENT 13 ZPS-1-MARK II DAR OCTOBER 1980 O Table 3 3-2 TEST MATRIX - DEFINITION OF ABBREVIATIONS AND FOOTNOTES ABBREVIATIONS: J SD = Shakedown Test 1 MT = Matrix Test TT = Transient Start-up Test SVA = Single Relief Valve Test CVA = Consecutive Relief Valve Test MVA a Multivalve Test ESVA = Extended Single Relief Valve Test CP = Cold Pipe ,
' WL t =
Normal Water Levei AWL = Actual Water Level WP = Warm Pipe 4 {) HP = Hot Pipe j NOTES: (1) Valve Designations shown in parenthesis correspond to the actual control room identification of the valve mode switches. (2) NWL implies that the level of the discharge pipe water leg is coincident with the suppression pool water surface, j (3) ANL implies that the level of the discharge pipe water leg has reached a steady state value which is either above or below NWL. (4) An initial suppression pool temperature shall be main-tained within +5'F for all SRV, CVA and MVA tests. Extended valve discharge tests shall be started "ith the suppression pool at the minimum attainable temper-ature. - 1 , (5) Shakedown tests shall be oerformed when the reactor . power is high enough to support steady state steam flow through an SRV discharge line. 3.3-5
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AMENDMENT 13 OCTOBER 1980 Cow TAIN M E N T WA L L - , t p ,f,p P3 P38 ' HPCS SUCTION [Q
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G 6'Pl'34,P[ RHR SUCTION 83 '. ^ 1 717
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- o. RHR RETURN SUCTION 81 1 PLAN - WETWELL
$_ REACTOR Q .l
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. . P 3 5,P 3i * . P3.P 1 S.P 18 P2ia .
I (scNsons MAve me.zw RoTATEo swTo vizw) ELEVATION l l WM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK ll DESIGN ASSESSMENT REPORT Ch (j FIGURE 3.3-2 SUPPRESSION POOL PRESSURE
~ _ .
AMENDMENT 13 OCTOBER 1980 l18 0* t I w/ CON TA I N M E N T . .
'd WALL _
Tigre ?*
. 14 T30 12,T2 8 ,
4 .g HPCS SUCTION T3 RHR o
.f 11,T27 T
SUCTION # 3 8,T 2 6 T2 T T,T6,T 2 4 V 4 O O ' RHR T 5,T 6,T 2 2 ~, , LPCS RETURN SUCTION T3Tte T3 T21 4
. . . T2 T16 o ' ^
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PLCS.) SUCTION #2 . c RHRd N ION # 1 RHR RETURN O* NOTE: - PLAN -
@ - Designates existing plant temperature sensors WM. H.ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT (l
v FIGURE 3.3-3 SUPPRESSION P00L TEMPERATURE SENSORS (SHEET 1 of 2)
AMENDMENT 13 OCTOBER 1980 O 8 REACTOR -
- \, !
l .o
, -o . . . _
- a ,s 9
SU PPORT ~ COL U M N PEDESTAL WALL , ;
; I
-' . 4 A
EL. 4 96'- 10~ 4 U EL. 4 9(.' - 4 g [. 73,77,
, T28a -
l 7,. ' T 2 6,T 2 7,T 2 9 CL 407 ~
, T2 2,T2 4 e < >T25 d
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g 8 l T 17,T 16,T 19,T 2 0 9 4 a EL~ 478 '- 4 T23 T 10,T 1 1,T 12,T 13 sY 1* T 2,T 3,T 4 T 6'T 6 a T eI d
' 21 T 14,T 15,T 16
' . T35,T3 8
. E.L 474'- 10~ _ .T 3 3.T 3 4 o y ,
e- , 1 (SENSORS HAVE BEEN ROTATED NTO VIEW) ELEVA l ION NOTE: 9 - Designates existing plant temperature sensors ! l l WM. H. IBMMER NUCLEAR POWER STATION, UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT C FIGURE 3.3-3 SUPPRESSION POOL TEMPERATURE SENSORS (SHEET 2 of 2)
AMENDMENT 13 OCTOBER 1980 1 0 _ Sa fe ty Relief Valve SRV Discharge Line El. 568'-8" (app rox . ) T ._ / PXX g El . 5 5 3 '- 0"
\ ,
L TXX - Temperature Sensor (RTD) ffain Stean Line PXX - Pressure Sensor O SRV SRV LINE VALVE AZ POOL AZ SFNSOR PXX SFXSOR TXX F013D ?!S07AD10 310 193 P25,P34* T'7 F013C MS07AC10 269 220 P26,P27,P28* T38'T43**
*P28 and P34 are low range sensors.
**T43 located 1 f t. above vacuum breaker in the SRV line 9
WM. H.ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK ll DESIGN ASSESSM ENT REPORT O ricuae 3.3 4 TEMPERATURE AND PRESSURE SENSOR LOCATIONS
AMENDMENT 13
; neo- OCTOBER 1980 CowTAsNMENT WA66 l HPCS SUCTION f
g RHR SUCTION 83 % O i O . e f RHR RETURN LPCS SUCTION O *
%O SSA6 S3.84 O
O Y1 9C* k - -y . h 1T0* O - O . a u. 90 g O ' g RciC SUCTION 4 O O O agey=_T - crva.is Pwca-) .,.
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-RHR RETURN l 'o*
, PL AM - WETWELL i d REACTOA v .- .
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) pKDESTAL WALL .*
1 S , , f- CoNTAJN MENT } ( WA66
. -4 l .
EL 4% v - = = - 3 EL 4(% 4' ' k ' ct 47s' 4- S3,84 , EL 4 74- to" . STA8 S1.32. ,85,S6 l (SCNSCRs NAvc eEEN ROTATro swTo v cw) ELEVATION WM. H,ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 18 DESIGN ASSESSMENT REPORT
.Q- FIGURE 3.3-5 l SUPPRESSION POOL STRAIN GAUGE LOCATIONS
AMENDMENT 13 p Pipe Restraint 0CTOBER 1980 e
~
'f O 7r i ni. s16'-iot" NOTE: LEVEL PROBES IN PARENTHESES INDICATE Als07AD10 ] THOSE ON SRV DISCHARGE LINE AT THE 198 AZIMUTH.
- Y l g g (1MS07BC12 Mh07BD12 J J
" EL. 506 '-10" - . L4(L8) ! L3(L7) EL. 501 '-10" - .
~ '
1 liigh Water Level El. 497'-4" + _-_-=_-n Low Mater Level EL. 489'-2" El. 496'-10" L2(L6)
'** " "5)
O E1. ~ 490-L. _l_ x NOTE: Diametrically aisl lj l sis." opposed longitudinal ' sia > gauges should be connected to A A 834 measure bending strains. I 815W 'e s i s S12 SEC. A-A T-Quencher
\
Radius: 34 ' - 3" I I I } Az. 198' %/ Sunnression Poot Floor
' . E1. 474'-10" WM. H.ZIMMER NUOLEAR POWER STATION, UNIT 1 MARK 11 DESIGN ASSESSM ENT REPORT 4
FIGURE 3.3-6 SRV DISCHARGE LINE SENSOR LOCATIONS
AMENDMENT 13 hv. /p j OCTOBER 1980
'~ ',*
- - - - - - - - sto O
~~ ~ ~ - - - - -
sts - e'< '
, .1 v
. , , Rs3
.s ~ s .
d Containment Wall
. /
'3 d HPCS Suction 4 ,
j ' ' ms s.t s c-E1. 481'-7" _ .- Rask Asal e sto i
'se' D ! *se
- A SF.C A-A I
t O . . v
\ - El. 474'-10" g AZ.220*
SEC. B-B Required: 3 strain gauges are arranged to fom a rectangular rosette, with 1 gauge parallel to the longitudinal axis of the pipe and cne gauge oriented along the circumference of the pipe. 3 ad<litional gauges are oriented parallel to the pipe's longitudinal axis at 90* intervals. All gauges am located at least 2" away from any ' welds. n i i WM. H.ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT Q FIGURE 3.3-7 HPCS SUCTION LINE STRAIN GAUGE LOCATIONS I
AMENDMENT 13 0CTOBER 1980 A: 204* Az 228' sse Drywell Floor ses ese sie sas El. 520'-4-3/4" i # ' sie l l l T**' f I'-0" 1 *** e Each downcomer has 4 uniaxial strain gauges each 90* apart oriented along the longitudinal axis of downcomer. locate gauges at least 2" from any weld. Downcomer a Containment Nall A40, A 4 s E1. 486'-10" - h AZ 204*
/
O A: 228' 0'-9"R , O 4 M 18'-3"R . 2N' N WM. H. ZlMMER NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT l l { FIGURE 3.3-8 DOWNCOMER STRAIN GAUGE AND ACCELER0 METER LOCATIONS
)
. . . - _ . - . ___ = _ . . . - _ - - - - . _ - _ . - _ _ . - - - - - - - . . .. . _ . _ ~ . .
O O O 1 Required: 4 strain gauges on each arm orientated parallel to the arm's longi-
,m Containment Wall tudinal axis and located 90' apart, with 2 gauges in and 2 out of the plane of the arm and support. Both arms also have 2 diametrically opposed strain gauges orientated s27 along the circumference of the arm. Located as shown are two strain gauges 180* apart on ,
saa ses the sphere surface oriented parallel to the f 1 gd as l sinis _ s22
} lt pa,agitudinal ris S23 6 S34, axisS27 of the 6 S28arms.
Strain gauge and S3S 6 S30 are , t ,, _"[,aso
,, ysts
*1 )
to be connected in an additive bridge. . 1 in 1 824 i 83 { 826 Strain gauges should be located at least
, 2" away fron any welds. AZ PXX TXX P29.P3 T39 T4r M y 19!1
- E % SRV Discharge Line F
z i Quencher Arm S29 _f _St2 f f W c *
% [
- s3o) l s23 ) \ '
i:sy, *D -
*f
- is- 327
$ l92 sr. **i: , t-o 5
,,, t,
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22
,1
- 5 E_
o , ss2 N -# s2s > r. ... ra m $ 2 m j g g > $ Pressure Temperature Sensor Sensor 33 w
$1 m O f ,
Quencher (Inside (inside wall "PP pressure) temperature) r- Y y% [ i @ I ;c
/ '
M z $ Supression Pool Floor' I i o, 5
- D CM q5-z 85 4 o gm a C =
Z gH 1 a
=
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)
AMENDMENT 13 OCTOBER 1980 O sss
.e tRS2 al -t- -
( of Quencher Arm
\, ,e RS1 2 rosettes required with one gauge parallel to the longitudinal axis of the support and one gauge oriented along the cir- '
cumference. The rosettes are 90 apart with one
' a directly below the quencier i arm ( . Uniaxial gauges O assq , i, are tocated 18o aw=v trom each rosette oriented 7[ asyI ilc._ r_ parallel to the longitudinal nst RS2 axis.
Strain gauges should be 15,, 1 c ted at least 2" away from any welds. _L I I Suppression Pool Floor i e WM. H.2IMMER NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT O Ftouat 3.3-10 QUENCHER SUPPORT STP,AIN l GAUGE LOCATIONS j 1
o ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
) 3. 4' TEST MATRIX l
The test matrix is presented in Table 3.3-1. Definitions and footnotes for the test matrix are contained in Table 3.3-2. The test matrix includes single, consecutive, multiple and extended , valve ac;uations. The tests have been selected to provide a data L
-base which can be extrapolated to the design load cases.
Approximately 20% of the data channels will be recorded in real , time during the testing to allow immediate judgments to be exercised regarding comparison to design-basis loads, data scatter, etc. O l 3.4-1 1
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () 3.5 DATA ACQUISITION The Zimmer SRV In-Plant Test will require 174 fast channels of instrumentation with a maximum frequency response capability of 200 hertz. The signals from the fast-channel instrumentation will be processed through appropriate signal-conditioning equipment and stored in digital format on magnetic tape. Each sensor , vill be scanned at 1000 samples /second. In addition, approximately 20% of the fast channel will be recorded on oscillation recorders to provide real-time data. Real-time acceleration data will be processed through a shock spectrum analyzer to provide " quick-look" response spectra. The real-time data will be compared to the appropriate design values. An additional 36 slow-response (0 to 3 hertz) thermocouples will be located in the suppression pool. Thermocouple outputs will be recorded in real time on a Fluke Data Logger. O 3.5-1
ZPS-1-MARK II DAR AMENDM.".NT 13 OCTOBER 1980 , () 3.6 TEST SCHEDULE AND REPORTING Figutt 3.6-1 shows the proposed test schedule and includes p;e-liminary od final reports. In-process adjustments to the test matrix will be made during testing if deemed to be required. The final report, including plots of data in engineering units, tabulations of maximums / minimums and comparisons to design basis, will be completed approximately 5 months following the completion of the tests. O 3.6-1
J WEEKS
^0
-ife 12 s -G 1 O E 4- 6 8 10 IE I4 16 18 20 EE E4 E6 28 30 f
1 75 % tico % FUEL LOAD t ?.5 % SO Y. POWER POWER POWER goog goo g INITI AL H EATUP POWER gg,ow ytow TEST SETUP A A SHAKEDOWN O TESTING SHAKEDOWN DATA A A EVALUATION TEST M ATRI X O TR AMSIENT H , STARTUP TESTS (100 */. POWER) FULL MSiv CLOSURE E - T-6 LOAO g
= i REJEC.TIO N
- I
- m m g 2 PRELIMINARY A A E g$ (Q UIC K LOO K) g - ap aEeoaT 4 5 m
*9 A w $2 FIN AL REPORT 9 F *O l_ L l m m a g - .
8 r--
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- h O '>
"t Elm 88 3sz 9m J
E@ ! 5 E:
1PS-1-MARK II DAR AMENDMENT 13 i OCTOBER 1980 I () CHAPTER 4.0 - GENERAL DESCRIPTION OF THE PLANT The Wm. H. Zimmer Nuclear Power Station, Unit 1, employs a GE-BWR/5 housed in a Mark II type containment structure (see Figure 4.0-1). The unit has a rated core thermal power level of 2436 MWt. The Mark II primary containment is a steel-lined, post-tensioned concrete pressure-suppression system of the over-and-under configuration. Pertinent physical data on the containment is summarized in Table 4.0-1. The pressure-suppression design incorporates a total of 88 downcomers with a submergence of 10 feet below the low water level of the suppression pool. The steam generated in the nuclear boiler is directly used by the Westinghouse main turbine-generator unit. The main turbine is an 1800 rpm, tandem-compound, four-flow nuclear steam unit. The nuclear boiler has 13 safety / relief valves to limit pressure buildup in the system as required by the ASME Boiler and Pressure Vessel Code. The valves are mounted on the four main steamlines upstream of the inboard main steam isolation valves and are located in the drywell portion of the primary containment. Six of the 13 safety / relief valves are part of the automatic depressurization system (ADS) which is designed for pressure relief following an intermediate line break. The discharge lines from all of the safety / relief valves are routed into the suppression pool. Each discharge line terminates with a quencher (]) discharge device. Each quencher is located approximately 3.5 feet above the top of the suppression pool basemat; this is equivalent to a submergence of approximately 18.5 feet below the pool low water level. As a result of the reassessment of the Wm. H. Zimmer Power Station to the bounding pool dynamic loads, many changes have been made to the structure, piping, and equipment. Some of the more significant modifications are:
- a. Installation of quenchers and associated MSRV line rerouting.
- b. ' Addition of downcomer bracing.
! c. Filling of pedestal with concrete to elevation 497 feet 6 inches.
- d. Additional supports and restaints for wetwell and drywell piping.
These modifications are listed and explained more completely in Chapter 9.0.
- O 4.0-1
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (} TABLE 4.0-1 PRIMARY CONTAINMENT PRINCIPAL DESIGN PARAMETERS AND CHARACTERISTICS I. DESIGN PRESSURES i A. Containment Internal Design 45 psig Pressure B. Containment External Design +2 psig Pressure C. Drywell Floor Differential Design Pressure
- 1. downward 25 psi
- 2. upward 9 psi II. VOLUMES A. Maximum Drywell Free Air Volume 180,000 ft 3
() B. Maximum Suppression Chamber Free Air Volume 95,350 ft 3 C. Maximum Suppression Chamber Water Volume 95,380 ft 3 III. DOWNCOMER SUPPRESSION VENTS A. Number of Downcomers 88
^
B. Internal Diameter 2.0 ft C. Wall Thickness (Nominal) 0.5 in. D. Material SA 516 Grade 60 E. Length
- 1. unembedded length 33 ft 6-3/4 in.
- 2. total length 37 ft 3-3/4 in.
- 3. submergence depth 10.0 ft IV. SAFETY / RELIEF VALVE DISCHARGE LINES O' A. Number of Discharge Lines 13
^ 4.0-2
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 4 () B.
. TABLE 4.0-1 (Cont'd)
Size and Schedule
- 1. Wetwell 10 in./ Schedule 40 10 in./ Schedule 120
- 2. Wetwell Risers 12 in./ Schedule XS or 12 in./ Schedule 80
- 3. Drywell Floor 12 in./ Schedule 160
- 4. Drywell 10 in./ Schedule 40 C. Material SA 106 Grade B D. ASME Code Class Section III, Class 3 E. Type of Discharge Device Quencher F. Height of Quencher Centerline Above Basemat 3 ft 6 in.
G. Line Lengths, Location, and Valve Setpoints Valve Set Line Pressure Line Number of Azimuth Number (psig) Length Fittings Location Radius MS11A10 1150 164 ft 9 in. 9-90* elbows 92 19 ft 0 in. MS10AA10 1150 179 ft 5 in. 9-90 elbows 270 19 ft 0 in. MS08AA10 1175 180 ft 3 in. 8-90 elbows 257* 34 ft 3 in. MS08AD10 1175 188 ft 1 in. 8-90 elbows 168* 19 ft 0 in. MS10AB10 1175 140 ft 2 in. 6-90 elbows 21* 19 ft 0 in. MS07AA10 1185 153 ft 11 in. 6-90 elbows 327* 19 ft 0 in. MS07AD10 1185 186 ft 1 in. 9-90* elbows 198 34 ft 3 in.
*MS09AA10 1185 160 ft 6 in. 7-90 elbows 138* 34 ft 3 in.
*MS09AB10 1195 -177 ft 4 in. 7-90* elbows 162 34 ft 3 in.
*MS08AC10 1195 165 ft 9 in. 6-90 elbows 68 34 ft 3 in.
*MS08AB10 1195 185 ft 9'in. 8-90* elbows 306 34 ft 3 in.
*MS07AB10 1205 151 ft 6 in. 5-90 elbows 7 34 ft 3 in.
*MS07AC10 1205 160 ft 5 in. 5-90 elbows 220 19 ft 0 in, b
a
- Indicates an ADS valve discharge.
4.0-3 i
-- -- -- - _ -. . , . . ~ _ _ - . . . , _ _ , . _ . ._ _ _. I
AMENDMENT 13 OCTOBER 1980 or.~ .;:t.' A n. x .. f
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. 8.M WM. H.ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 18 DESIGN ASSESSMENT REPORT q
FIGURE 4.0-1 PRDiARY CONTAINMENT BEFORF MODIFICATION
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
/^) \- CHAPTER 5.0 - LOADS CONSIDERED I i
In this section, it is intended to identify all the loads that , are being considered in the 2PS-1 design reassessment. In the next section, load combinations are identified and discussed. This section includes a brief summary of the original design ' loads, safety / relief valve discharge loads, and loss-of-coolant i accident loads. Loads caused by safety / relief valve discharge i and loss-of-coolant accidents are listed below for reference: ,
- a. Safety / Relief Valve (SRV) Actuation
- 1. Original design loads such as weight, thermal, i pump trip, valve closure, pipe internal pressure,i and reaction loads are always considered (as ,
appropriate) in addition to the other SRV loads. ,
- 2. Bubble-induced loads: 8 1
a) on containment structures; i b) on submerged structures; and l c) on piping, equipment, RPV, and internals. I () 3. Water jet loads: a) on submerged structures, and b) on wetted containment surfaces. 8
- 4. Second actuation loads. l
- 5. '
Quencher SRV support loads.
- b. Loss-of-Coolant Accident (LOCA) Loads i 1
- 1. Original design loads such as weight, thermal, ,
pump trip, valve closure, pressure and , temperature transient in wetwell/drywell, subcompartments, pipe jet loads, and pipe ' reaction loads are always considered (as e appropriate) in addition to the LOCA loads. :
- 2. Water clearing loads:
a) on submerged structures; ' i b' on containment structures; and i i c) on piping, equipment, RPV, and internals. , 5.0-1
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (~T N/ 3. Air clearing loads: , a) on submerged structures; ' I b) on containment structures; and i i c) on piping, equipment, RPV, and internals. ,
- 4. Pool swell loads: '
i a) drag loads, i b) impact loads, and l c) fallback loads. I i
- 5. Condensation oscillation loads: i a) on submerged structures; b) on containment structures; and '
c) on piping, equipment, RPV, and internals.
- 6. Chugging loads:
("3 1 a) on submerged structures; 5 i b) on containment structures; and i c) on piping, equipment, RPV, and internals. I
- 7. Downcomer lateral loads: '
s a) static equivalent load, and , b) dynamic load. I
- 8. Loads on drywell floor:
a) downward differential pressure, , b) upward differential pressure, and ' s c) loads due to forces on downcomers and main i vent deflectors. i
- 9. Annulus pressurization:
, I a) on sacrificial shield, and 8
() b) on piping, equipment, RPV, and internals. 5.0-2 I .,. . - _
2pS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 The original design loads, of course, are always c'onsidered to occur in combination, as appropriate, with the pool dynamic loads. It should be ncted that these pool dynamic loads are 8 relatively small compared to the original containment and reactor i pressure vessel (RPV) design basis. Therefore, the original i design contains adequate margin'to accommodate these pool dynamic , loads. These conservatisms are discussed in Chapter 10.0 of this report.
.These additional pool dynamic. loads are significant, however, when compared to the original design basis for the piping and equipment. Therefore, design modifications are being implemented in these areas which will allow these additional loads to be safely accommodated by meeting all code requirements. These modifications are discussed in Chapter 9.0 of this report.
e This chapter provides the most recent load compilation and ' tend i map" of the pool dynamic loads used for the reassessment. These , loads, identified as the "Zimmer Empirical-Loads," are defined in Section 2.1. In some cases, this reassessment has not been ' completed, and in these cases, this chapter describes the methods i that will be followed in the reassessment evaluation. This i report, together with the DFFR, FSAR, and the referenced reports, i constitutes a complete basis for the design reassessment , evaluation. () i 1 l 1 5.0-3
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 A kJ 5.1 ORIGINAL DESIGN LOADS The design basis of the 2PS-1 structure, piping, and equipment , was established before all pool dynamic loads were identified. Hence, this design basis must be reassessed for these newly identified pool dynamic loads in addition to the original design ' loads. The original design basis loads for the structure, a piping, and equipment are summarized in the following i subsections. , 5.1.1 Loads on the Structure The following loads were used in the original design cf the plant:
- a. Operating-Basis Loads D -
dead loads, F - prestressing loads, L - live loads, To - operating temperature loads, ( )) Rb - operating pipe reactions, Pb - operating pressure loads, P t containment test pressure, E - aperating-basis earthquake, 3 W - wind load, and H - hydrostatic flood load.
- b. Safe Shutdown and Design-Basis Loads E -
safe shutdown or design-basis earthquake, ss W - design-basis tornado, t H' - probable maximum flood load, P a accident pressure loads, Ta - accident temperature loads, Ra - pipe break temperature reaction loads, G)
- v. Rr -
reactions and jet forces due to pipe break, and 5.1-1
.2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 O 'H -
postaccident: flooding of the containment. l Details'of the above loads are given in Table 3.8-7 of the FSAR.
-A brief description of the loads is given below:
D = Dead load of the structure, permanent equipment, soil,
- - or hydrostatic pressure. Construction loading j .. is considered as dead load for the construction combination.
F = Loads resulting from the application of prestress. ! L = Live loads including any movable equipment. loads, l roof loads,.and crane loads. Appropriate impact factors are included for moving loads. T = Transient or steady-state thermal load on the structure at normal operation or shutdown conditions. Ro = Pipe, cable pan, and duct reactions-due to D, T, o and unbalanced pressure. Po = Normal operating pressure differentials. P = Containment test pressure. t 4 Eg = Seismic excitation effects from the operating-basis ' earthquake (see FSAR Section 3.7), dynamic soil prem ' sures, hydrodynamic pressures, and sloshing. , W = Design wind velocity loads (see FSAR Section 3.3). H = Hydrostatic effects from the flood of record (see l FSAR Section 3.4). I E-88 = Seismic excitation effects from the safe shutdown ] earthquake (see FSAR Section 3.7), including dynamic . j soil-pressure,-hydrodynamic pressures, and sloshing. l W t = Design-basis tornado loads, including wind velocity pressure, atmospheric pressure change, and missile impact' effects (see FSAR Section 3.3). j H' = Hydrostatic effects from the probable maximum flood-or precipitation (see FSAR Section 3.4 and Subsection
- 2.4.3.1).
P" = Maximum differential pressure defined by a postulated pipe break, iA 'U T^
= Effects of transient thermal load following a postu-lated pipe break.- This includes To for all other areas not affected by the pipe break.-
5.1 1
,,.,...-,w -- -
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () R a = Effects of Ta on pipe and equipment reactions generated by a postulated pipe break. This includes 80 for all other areas not affected i by the pipe break. ! R
= Local effects on the structure generated by a postulated pipe break. These loads include reactions from pipe supports and whip restraints, jet impingement, and missile impact.
Ha = Postaccident flooding of the containment. 5.1.2 Loads on Piping and Equipment i ' I The following loads on the piping and equipment were considered , in the original design bases. These loads and load combinations are generally consistent with current codes such as ASME B&PV ' Cection III, Division 1, and Regulatory Guide 1.92. All loads I are applied to the piping and analyzed to ensure compliance to i Section III of the ASME Boiler and Pressure Vessel Code, i therefore ensuring the pressure integrity of the piping. , Functional capability of the piping will be evaluated according to the criteria discussed in FSAR Subsection 3.9.3.1. ' l. WEIGHT
' () The sustained load consisting of the weight of pipe, contents, '
and insulation. i j THERMAL i i t A sccondary, self-limiting load resulting from the constraint of , the free end displacement of the piping system imposed by its ' rigid restraints. i SEISMIC t A dynamic load caused by the excitation of the piping by its i restraints during an earthquake event. A more detailed explanation of the methodology used in this analysis can be found 8 in FSAR Subsection 3.7.3. 8 PRESSURE Stress induced in the pipe by the internal pressure on the pipe ' wall. s HYDRAULIC TRANSIENT i 1 A dynaniic load on applicable systems due to appreciable and , sudden changes in the mass flow rate in the piping system caused () by sudden valve opening or closure, pump trip, or pump startup. ' 5.1-3
2PS-1-MARK-II DAR AMENDMENT 13 OCTOBER 1980 POOL SLOSH LOADS 3' The dynamic response of suppression pool water due to rigid body I horizontal motion of the pool boundary resulting from a seismic i event. The water velocities and accelerations induce i hydrodynamic forces on~ submerged structures. These forces are ' lomposed of a pressure' force on the pipe walls, velocity drag loads,'and water inertia loads. The seismic loads used in design ' include all expected pool slosh loads. i i ) 4 i 4 4 h 1, I O 5.1-4
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 ( 5.2 SAFETY / RELIEF VALVE (SRV) LOADS - PRESENT DEbiGN LOADS (T-0UENCHERS) Actuation of safety / relief valves (SRV) produces direct transient loads on components and structures in the suppression chamber region and the associated structural response produces transient 8 loadings on piping systems and equipment in the containment i region and reactor building. These transient SRV loadings are i discussed in the following subsections. , Prior to actuation, the discharge piping of an SRV line contains atmospheric air and a column of water corresponding to the line I submergence. Following SRV actuation, pressure builds up inside i the piping as steam compresses the air in the line. The , resulting high-pressure air bubble that enters the pool , oscillates in the pool as it goes through cycles of overexpansion and recompression. The bubble oscillations resulting from SRV actuation and discharge cause osillating pressures throughout the 8 , pool, resulting in dynamic loads on pool boundaries and submerged i structures. These dynamic loads cause a dynamic structural i response sufficient to affect piping systems and equipment in the , containment and reactor buildings. The assessment of the affected systems for these responses is discussed in Chapter 7.0. ' i Steam condensation vibration phenomena can occur if high- i (]) pressure, high-temperature steam is continuously discharged at high-mass velocity from rams head devices into the pool, when the i pool is at elevated temperatures. and submerged structures. This phenomena is mitigated by maintaining a low pool temperature ' as discussed in Chapter 8.0 and by installing quencher discharge i devices. i The characteristics of the SRV actuation load vary depending on the piping configuration and the discharge device (rams head or quencher) located at the exit of the SRV line. ' quencher device produces lower dynamic loads.Zimmer . Typically, Power the i Station used a bounding load calculated for a rams head device as i an original design basis for structures, equipment, and piping , systems. A bounding quencher load is now used. To provide increased plant safety margins for containment SRV loads and to increase the threshold temperature limit for steam condensation vibration, SRV quencher devices are installed in the plant. Pool temperature transients for several postulated cases involving a stuck-open safety / relief valve are presented in Section 8.2. The calculated maximum pool temperature was calculated to be a few degrees below the threshold temperature limit for steam condensation instability for a rams head discharge device. (~ In order to increase the margin between the calculated maximum i (_) temperature and this threshold temperature limit, it was decided i to install a quencher device having a higher suppression pool , 5.2-1 "g eg. p y--w- w .. rw - yg < vrw
. -_ . . . _ _ - . ._ ~ .__ . _ . .
2PS-1-MARK II DAR AMENDMENT 13 l OCTOBER 1980 temperature limit as reported in NEDE 21078, October 1975, rather than to perform additional testing with the rams head discharge ' device. The quencher device provides an additional benefit, 8 since the peak pressure amplitude of the containment structural i i loads due to the oscillating air bubble are reduced below the i corresponding design-basis values for the rams head device. , Therefore, it was concluded that a quencher discharge device not only provides an increased margin for the threshold pool ' temperature limit, but that the plant will generally experience i lower loads than those used in the rems head design basis. i The quencher device being used is the two-arm "T"-quencher j developed for the Mark II Susquehanna Plant by KWU. This device has been tested in a full-scale, single-cell facility as reported' in Chaoter 8 of the Susquehanna Design Assessment Report. The i test facility is prototypical of the Susquehanna plant. i Parameters were varied to include a range of initial conditions and the longest and shortest linas of Susquehanna. The tests i- wcro renducted to duplicate expected operating conditions 1 including first and subsequent actuations. The geometry and l initial conditions tested closely simulate those for the Zimmer Power Station. These tests showed that the device will condense j steam without significant loads at pool temperatures up to and i even above 2000 F. In addition, the tests showed that the actual quencher loads are conservatively bounded by the design loads I () given in Chapter 4 of the Susquehanna DAR. Since ZPS-1 is being assessed for these design loads in addition to the rams head loads, this demonstrates again the conservatism of the 2PS-1 j design, a 1 Quenchers with four arms (X-quenchers) have been installed and tested at Caorso, a Mark II plant in Italy. This test included l' single valve first and subsequent actuations, multiple valve actuations (up to eight valves), and an extended blowdown thermal mixing test. The results of these tests are reported in NEDE 25100P, " Mark II Containment Supporting Program'Caorso Safety Relief Valve Discharge Tests, Phase I Test Report" (May 1979), and by GE letter MFN-090-79 (L. J. Sobon to J. F. Stolz, March 1979). The measured loads were much less than those predicted by the analytical models in DFFR. The increase in load between single and multiple valve. discharge was less than predicted. The extended blowdown. indicated good mixing with a final bulk to
- local temperature differential of about 100 F.
4 i In the following subsections several current liceasing issues are i discussed and the methods used to predict load.c for the ZPS-1 i ! plant design reassessment are summarized. , 5.2.1 ~Desian-Basis SRV Loads - Rams Head ' i i
, The original design basis for reassessment of the structure, i attached piping systems, RPB, and. equipment was based upon i dynamic loads calculated for a rams head discharge device. The ,
5.2-2 J
- . - + - ---,m--,,,e _, -,_ y. . . , , .-..#. 1- ...%-ym . - - . . . . a -- ,,..,-m,_,,m.------ ..rr
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (D w/ SRV' rams head analytical model is described in NEDO-21061, i Revision 2 (September 197t ) . The method of application of this i model is described in tr.e following subsections. These , analytical models have been compared to results obtained from in-plant tests in the Mark I Monticello plant, and it has been demonstrated that these models conservatively bound the measured I load magnitudes (NRC-GEN-0394, September 1977). This subsection is concerned caiy with the rams head design basis. The 2PS-1 plant has been reassessed using a limiting empirical load as explained in Section 2.1. A description of the T-quencher design method and a discussion of how it has been rpplied is presented in Subsection 5.2.2. 5.2.1.1 Conservatism in SRV nams Head Methods i This subsection contains a summary of additional conservatisms in i the suppression pool dynamic load definition for rams head SRV i discharge. The methodology is explained and summarized in NEDO , 24070, October 1977. I First, :t is important to recognize that a number of different i SRV actuation cases have been investigated in order to select the i conservatiue design cases used for plant reassessment. The SRV i actuation cases investigated are identified and defined in , Subsection 5.2.1.2. It was concluded from this investigation (") that the all-valve case, SRV-ALL, was the governing case. Furthermore, five cases were investigated in' greater detail to I select the maximum SRV-ALL loading condition which provides an i adequate, conservative design-basis load. The all-valve cases , are explained further in Subsection 5.2.1.2.4. Second, since the SRV discharge loadings are of a dynamic nature, I three separate key characteristics were bounded by appropriate i definition of input assumptions for each load case. The three i key characteristics are: ,
- a. load magnitude and its spatial distribution, '
I
- b. frequency content of the forcing function, and i
- c. load duration. l As an aid in selecting the bounding input assumt.tions, '
sensitivity of the several dependent variab1; ~ ; the independent i variables involved in both the SRV vent cleatt., and oscillating i air bubble dynamics aspects of the problem were investigated. i Tables 5.2-1 and 5.2-2 delineate the independent and dependent i variables considered in these investigations. Third, in order to define a conservative forcing function for I (~} each SRV discharge case, four major steps were applied for each a ss case. A bounding, conservative approach was adopted for each of i these steps. Thus, the resulting final forcing functions include , 5.2-3
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 ' t') the compounded conservatisms' utilized in each step of their i definition. The SRV discharge line clearing transient was i determined such that the maximum discharge pressure was obtained. , Maximizing the vent clearing discharge pressure produces bounding air bubble pressures, and hence, maximized wall loads. The ' analysis in each step in the load definition process was bounded I and then used as input to the next step. In this manner the i final load definition contains the multiplicative effects of i several degrees of conservatism. The four major steps and the , conservatisms used in each of the load definitions are listed below. The conservatisms for the structures are discussed in ' Subsaction 10.1.2. and the conservatisms for the piping are i disc.ssed in Subsection 10.1.3. i 1
- a. Discharge Line Clearing Transient l
- 1. '
The SRV mass flow rate was maximized by utiliping
- the spring setpoint instead of the relief setpoint (PSPRING >PRELIEP) even in cases where i this was physically not possible. Further, the i ASME rated capacity was increased to 122.5% based on the most conservative selection of flow characteristic tolerances. '
I
- 2. All friction losses in both the air /tteam and
() water leg portions of the piping were includad. A bounding value of the rams head discharge i coefficient was also included. Thus, the resulting SRV discharge line pressures were ' maximized. I i
- 3. The rams head submerged depth, i.e., water column ,
leg, was maximized by using the suppression pool ' high water level in the ariclysis. i
- 4. In the asymmetric disc.'arge case, the maximum a predicted water column recovery height was i utilized to maximize the water column leg for the ,
valve undergoing the second actuation. The normal maximized water column height was utilized ' for the adjacent valve. 8 i
- b. Oscillating Air Bubble Dynamics i
- 1. Each rams head was assumed to form a pair of single, spherical bubbles. No interaction with '
piping or structures submerged in the pool was ' considered; hence, only bubble escape through the i pool surface terminated the loading condition. i This increased the load duration. , () ' 5.2-4
t 2PS-1-MARK II DAR . AMENDMENT 13 OCTOBER 1980 4 .
- 2. Individual SRV discharge line characteristics I were considered to maximize the load duration and i frequency content. i
- 3. Dissipation of bubble energy was not allowed.
i
- 4. A 15%-20% variation of bubble' frequency due to '
changing hydrostatic pressure during bubble rise i was included in the analyses. ,
- c. Pool Geometry
- 1. The effects of the rigid boundaries on the
} pressure loading in the pool were included. This i produces enhancement of the loads, especially , near the containment wall. . i I 2. The effects of the free surface on the pressure ' 2 distribution were included. i 3 3. The vertical translation and the resulting time- , ! dependent spatial distribution of the wall loads j were included'in the analyses. '
- d. SRV-Actuation Load Case
() 1. Four SRV actuation conditions were used in the i i containment design, and each load condition was investigated separately to maximize loads on ' structures and components. The SRV actuation I conditions (i.e., SRV load combinations) used for i containment design are listed as follows for i convenience: (1) SRV(ALL); (2) SRV(ADS); , (3) SRV(ASY); and (4) SRV(1). Several load definitions were studied for SRV(ALL), and the ' most severe loading case was selected for containment design. i i
- 2. For each load case the load magnitude, frequency ,
content, and load duration were maximized. This ensured that the dynamic nature of the forcing ' function was addressed and that a limiting set of 8
- its characteristics was selected. :
i
- 3. The selected forcing functions were defined to ensure that both the vertical and horizontal load components were maximized. 8 i
- 4. For all multiple SRV actuation cases, the i contribution to the loading from each SRV line ,
was linearly summed. p/ rs 3
, l l
5.2-5
l 2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 O/ Since there are a number of conservatisms in the bounding load definition in addition to the conservatisms in the application of this bounding load, it is concluded that dynamic loads caused by SRV discharge irom a rams head provided a conservative design basis for plant reassessment. Results from the KTG full scale tests (Section 8, Susquehanna DAR) have demonstrated that increased plant safety margins will be achieved if a quencher device is installed in the containment (1) a higher suppression pool temperature limit, as reported in NEDE 21078, will provide increased plant safety margin for postulated transients such as a stuck-open relief valve; and (2) reduced bubble pressure loads on containment structures and submerged piping will also result in increased plant safety margin for SRV actuation during postulated plant transients. Therefore, SRV quencher devices are installed and an in-plant SRV test will be performed to confirm the conservative design basis. This provides sufficient timely resolution of all licensing issues related to SRV actuation. 5.2.1.2 Safetv/ Relief Valve Discharge Cases A large number of potential rams head SRV discharge cases were considered in the process of selecting a set of realistic bounding load cases. These cases and the assumptions and (~) ' conditions associated with them are described here. The corresponding quencher discharge cases are described in Subsection 5.2.2.1. 5.2.1.2.1 Sinole Valve Actuation The single SRV discharge loading case is required to account for the actuation of an SRV. Since it is conservatively assumed that any of the 13 SRV's may discharge, the line which produces the largest structural load was chosen to define the load magnitude. This load magnitude is conservatively assumed to apply to any single SRV. This load condition is provided for use in the required load combinations specified in Chapter 6.0. 5.2.1.2.2 Asymmetric SRV Actuation The asymmetric loading case is described in Subsection 3.2.4.1.3 of the DFFR. The asymmetric SRV discharge loading for 2PS-1 involves the actuation of two lines whose discharge devices are adjacent. The loading situation is constructed by first considering a low-setpoint SRV which is cycling. Then, it is assumed that, coincident with the second actuation of the cycling SRV, an adjacent SRV line is actuated. The specific line characteristics were considered for both discharge lines. A vacuum breaker system for the discharge line was evaluated to em (,) determine the maximum water column recovery height in the line with the cycling SRV. A partial vacuum resulting from steam condensation in that vent line permits the water column to rise 5.2-6
2pS-1-MARK II~DAR AMENDMENT 13 OCTOBER 1980 1 v momentarily in the vent line to a height greater than the initial submergence depth of the vent line. Even though the vacuum breaker system is sized to minimize the recovery height of the water column, this conservative assumption (using the maximum water column recovery height) results in a bounding load for asymmetric SRV actuation. 5.2.1.2.3 Automatic Depressurization System (ADS) The ADS valve loading case is described in Subsection 3.2.4.1.2 of the DFFR. The rams head automatic depressurization system consisted of six valves which discharge at the rams head locations shown in Figure 5.2-1. The six ADS valves are located nearly symmetrically and actuate simultaneously at the same system pressure. A conservative upper bound for the system pressure of 1250 psig was selected. This pressure corresponds to the highest spring setpoint and maximizes the SRV mass flow rate. The specific individual line characteristics (length, friction factor, etc.) were used in the load evaluation. The all-valve case produced a larger containment load than the ADS actuation. The ADS actuation loading is provided for load combinations which include IBA and SBA line breaks. O 5.2.1.2.4 All Valve Discharge Cases i i When multiple SRV vent lines discharge into a suppression pool, , the relative timing among the air bubbles' dynamics depends on individual characteristics of the valves and lines involved. In ' the calculation o'f dynamic loads, the fcllowing factors may be 8
; taken into account for various postulated discharge cases: i
- a. main steam supply pressure transient; ,
- b. SRV pressure setpoint; '
- c. vent line characteristics (length, diameter, t equivalent friction factors, etc.); and i
- d. initial conditions in the line. ,
The supply pressure (including its time rate of increase) and SRV 8 setpoint determine the actuation time for each valve. The line i j characteristics and initial conditions determine line clearing i i time as well as bubble formation times and dynamics (bubble , pressure, radius, and depth versus time). Appropriate sont clearing times are calculated by using the vent clearing : odel ' provided in Reference 1. The line clearing time is accurately 8 pg calculated as demonstrated by a predicted clearing time of 240 1 (/ msec compared to a range of 200-300 msec indicated by test data
~ i i (Reference 2) for the same clearing transient. The SRV flow rate , l 5.2-7 4
. l l
ZPS-1-MARK II D?o. AMENDMENT 13 OCTOBER 1980 O was calculated using a conservative method which gives flow 22.5%i higher _than expected. A conservatively short valve opening time a was also used which will maximize the bubble pressure , The bounding load approach taken in rams head design assessmen* ' calculations was to postulate a number of conceivable discharge i situation, then mechanistically calculate the suppression pool i loading functions for each case, and finally select the bounding i case on the basis of the load function or its structural ' response. The bounding discharge case usually varies depending on the configuration of the loaded structure. That is, major 8 structural loads on the pedesAal, basemat, and containment are i often bounded for a different discharge case than are loads on i submerged structures, such as support columns, downcomers, and i SRV vent lines themselves. The discharge cases must also include, bounding structural loads for forces in the vertical and horizontal directions as well as bounding " rocking" moments. ' Mechanistic calculations include individual vent line transients,I air bubble dynamics, and the load factors which relate bubble i dynamics to pressure or drag forces on specific structures. Each, calculation is unique to each plant, structure, and discharge , case. I The following rams head discharge cases were considered in designi
~
reassessment as reported in this subsection: i
- a. Simultaneous Bubble Discharge l
All 13 bubble pairs are identical and in phase. A ' single-pressure time history corresponding to the SRVi line giving the maximum bubble pressure is used. Thei square root of the sum of the squares method (SRSS) , is used to simultaneously combine the effect of all the bubbles. ' I
- b. Symmetric Discharge I
Simultaneous firing of all 13 valves. The bubble , pairs are all unique and are not in phase. The effect of each bubble pair is combined by the SRSS ' method and the effect of each line is then added i linearly. e
- c. Ganged Sequential Discharge l All 13 lines are discharged in accordance with their '
given pressure relief setpoints for a linear RPV 8 pressure transient. The maximum anticipated RPV i pressure ramp rate of 136.4 psi /sec is used. The i bubble pairs are all unique and out of phase. The , /~) k' effects of the bubbles are combined as in case b above. 5.2-8
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (D
- d. Continuous Sequential Discharge i
All 13 lines are discharged in accordance with 13 ' different relief setpoints which could occur due to i setpoint drift. The " drift" is assumed to cause all i 13 setpoints to be equally spaced (in pressure) over , the duration of SRV discharge. A linear RPV pressure , transient is used. The maximum anticipated RPV pressure ramp rate of 136.4 psi /sec is used. The ' bubble pairs are all unique and out of phase. Their i effects are combined as in case b above. i
- e. Resonant Sequential Symmetric Discharge l All 13 lines are discharged in accordance with their 8 given pressure relief setpoints for a linear RPV i pressure transient. These setpoints are equally i spaced in pressure. The period of oscillation of the ,
first bubble pair in the pool is determined. The RPV pressure ramp rate is then chosen such that the period between actuation of adjacent relief setpoints 8 equals the oscillation period of the bubbles in the i pool. In thic manner, an effort is made to cause the i discharge of subsequent relief valves to be in ,
- " resonance" with the bubbles in the suppression pool.
(_S) Variations of the pressure ramp rate or valve setpoint will generally result in bubbles further out 8 of phase, since these variables have been chosen i within an allowable range to be as closely phased as i possible. The effects of bubble pairs are combined , as in case b above, i As described above, five all-valve discharge cases were I considered before selecting the one (symmetric discharge) that i was judged to produce the most severe structural response. , By considering five SRV(ALL) discharge cases which utilize worst-case mechanistic assumptions and conservative load methodology, 8 it was judged that this procedure results in a conservative and i i appropriate SRV load for design reassessment. SRV(ALL) case b i ; referenced above gives a loading condition that was judged to , l produce the most severe' structural response. , j l 5.2.1.2.5 Second Actuation 8 i i l On October 1.1 1977, the NRC was formally notified of a a ! potentially reportable condition under 10 CFR 21 involving , transient analysis predictions of the sequence and number of , relief valves expected to operate following a reactor isolation event. ' /~T i (-) To investigate the effects of second actuation, a plant unique i analysis was performed. Tables 5.2-3 and 5.2-4 list the , 5.2-9
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
' assumptions and input parameters used in the analysis and Table i 5.2-5 shows the number of valves actuated on each lift. Figure 5.2-2 is a pressure versus time plot without the recirculation ,
pump trip, and Figure 5.2-3 is a similar plot with the pump trip. , For further information, the transient model used in this 8 analysis assumes a single spatially uniform pressure in the i vessel. The fluid in the vessel is divided into five homogeneous i nodes. Four nodes contain liquid or liquid vapor mixture, with , two inside the shroud and two outside. A steam dome of saturated vapor fills the remainder of the vessel. There are two sources ' of heat: (a) the stored and decay heat from the fuel, and (b) 8 the stored heat from the metal within the vessel. The relief a valve modellig included the following for each group of relief i valves: opelning and closing setpoint, opening and closing relief , valve delay time, opening and closing relief valve stroke times, and the flow rate at the opening pressure. The water level within the vessel was tracked, and the auxiliary systems, such as i HPCS and RCIC, were activated when their level setpoints were i reached. , The results of further investigation of the subsequent relief valve phenomenon demonstrated that for ZPS-1, conservative ' predictions show that five valves would lift during the second i actuation of the primary system relief valves following the i s, closure of all main steam isolation valves at 105% of rated steam , flow. This conservative analysis, together with the assumptions and conclusions, were presented to the NRC during a January 5, 1978 meeting between the NRC and CG&E on the 2PS-1 docket. The ' predictions were judged to be conservative by comparison with I operating experience which shows only 1 to 3 subsequent i actuations for isolation events. It was also demonstrated during , that meeting that the ZPS-1 containment was conservatively designed to withstand the second relief valve actuation effects. ' With the incorporation of a quencher device for relief valve 8 discharge, the resulting loads due to relief valve discharge will i be mitigated, and it will be demonstrated in an in-plant test i that the second actuation load will be less than or equal to the , first actuation load. , With the incorporation of SRV quencher devices, the loads are reduced sufficiently to eliminate concern over multiple subsequent actuation. As described in Subsection 5.2.2.1, an assessment is being made using the T-quencher load definition, which includes subsequent actuation loads. This assessment demonstrates that the 2PS-1 design is adequate for even the subsequent actuation of all valves.
)
5.2.1.3 Safety / Relief Valve Boundary Loads The submerged portion of the suppression pool is divided into k/<^} nine zones for analysis purposes. The specific pool geometry and , the definition of the nine zones are shown in Figure 5.2-4. The 5.2-10 l
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 O forcing functions for the containment walls and basemat are specified by determining the time history of the spatially averaged pressure for each zone. For example, the forcing function for Zone 4 (on the basemat) was calculated by determining the point pressures for a grid of several hundred points on the entire submerged boundary. Next, the integral from a radius of 15 feet to a radius of 24.5 feet for a full 2r radians of the pressure distribution is evaluated for a series of points in time. 5.2.1.4 Submerced Structure Loads The oscillating SRV bubble will cause fluid motion in the suppression pool, causing structures in the pool to be subjected
- to drag loads. Standard and acceleration drag are calculated 1 using the analytical methods documented in Peports NEDO/NEDE 21471-P, NEDO/NEDE 21472, and NEDE 21730. Loads are calculated for all structures in the pool by subdividing the structure and evaluating the characteristics of the flow at the center point of each segment.
5.2.2 Assessment for SRV Loads - T-Ouencher Although the loads expected with quenchers are lower in magnitude than those predicted with the rams head, changes in the discharge () device locations and differences in the load definitions required an assessment to confirm adequacy of the design. This has been acuomplished. The T-quencher load has been incorporated into the 7tmmer Empirical Load (Section 2.1). 5.2.2.1 Conservatism of the T-Ouencher Load Definition A T-quencher load definition has been developed by KWU as a result of an extensive reduced-scale and full-scale test program. j This load definition is documented in Chapter 4 of the i Susquehanna DAR. This load definition is based on actual pressure time histories of single valve discharges. These pressure time histories include subsequent actuations. Conservatism is added by increasing the amplitude (by using a 1.5 multiplier) and expanding the frequency range (to include all significant frequencies from 3.4 to 10 Hz). The KTG single cell tests (reported in Chapter 8 of the Susquehanna DAR) confirm the conservatism of this load definition. Additional conservatism is added when the spatial load distribution for the SRV cases (as discussed in Subsection 5.2.2.2) is also applied conservatively. 5.2.2.2 SRV Ouencher Discharoe Cases The T-quencher methodology includes cases corresponding to the rams head cases described in Subsection 5.2.1.2. A specific subsequent actuation case is not defined since the data used to {>S s form the load definition includes subsequent actuation cases. 5.2-11
- - _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ - - _ - _ _ - - _ _ _ _ _ - - _ _ - _ _ - _ _ _ _ _ - ~ _ _ _ _ _ _ _ _ _ - _ - _ _ _ - _ _ _ _ _ _ _ _ - - -
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (3 (_/ 5.2.2.2.1 Single Valve The load distribution on the containment walls for a single val /e actuation is shown in Figure 4-26 of the Susquehanna DAR. This load is better described as a subsequent actuation of a single valve. 5.2.2.2.2 Asymmetric SRV Load The asymmetric quencher load is defined as a three-valve discharge rather than the two-valve discharge used in the rams head asymmetric load. Although this condition is not realistic it gives a maximized asymmetric distribution as depicted in Figure 4-25 of the Susquehanna DAR. 5.2.2.2.3 Automatic Depressurization System (ADS) Figure 4-27 of the Susquehanna DAR shows the ADS pressure distribution. This distribution was constructed by combining single valve discharge loads at typical quencher locations. This would yield the expected distribution of more or less evenly spaced peaks but because of a conservative increase in the azimuthal angle of the single valve load, this results in an almost uniform distribution. For additional conservatism, the all valve distribution is used in most cases. (~)' 5.2.2.2.4 All Valve Discharge The all valve T-quencher discharge case is defined as the single valve discharge load applied uniformly throughout 3600 The physical interpretation of this load would be a subsequent - actuation of all valves with a13 bubbles entering the pool simultaneously and oscillating in phase. 5.2.2.3 Ouencher Boundary Loads The above described quencher load definitions have been applied to the suppression pool wetted boundaries to assess the structure, piping, and equipment. This assessment is documented in Chapter 7.0. 5.2.2.4 Quencher Submerged Structure Loads Submerged structure loads are affected by geometric changes in the pool because these loads are local loads. The change in discharge device location was assessed by using the existing submerged structure methodology with pressure amplitude, frequencies, and bubble locations appropriate to the KWU quenchers. The bubble pressure amplitude is determined for both first and subsequent actuation using the correlation in NEDO r3 21061, Revision 3 (DFFR). The bubble frequency range is reported l ,! in Subsection 5.2.2.1. 5.2-12
l 2PS-1-MARK II DAR AMENDMENT 13 . OCTOBER 1980 i 4 5.2.3 Assessment of NRC Acceptance Criteria - SRV The original design methods and the' design reassessments
- described'in the above subsections address all the NRC concerns in the Lead Plant Acceptance Criteria (NUREG-0487). An itemized list of the Zimmer Power Station response to the NRC Acceptance Criteria is contained in Section 5.4.
i i l 4
- O a
1 4 s i 1 i e i t 0 l l 5.2-13 .
t 2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
.O 5.2.4 References
- 1. .A. J. James, "The General Electric Pressure Suppression Containment Analytical Model," General Electric Report NEDO-10320, April 1971.
i 2. W. J. Bilanin, "The General Electric Mark'III Pressure Suppression Containment System Analytical Model," General Electric Report NEDO-20533, June 1974. t 1 O 4 1 4 l(2) 4 5.2-14 l
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 1 TABLE 5.2-1 SRV DISCHARGE LINE CLEARING TRANSIENT PARAMETERIZATION INDEPENDENT PARAMETERS CONSIDERED
- a. SRV pressure setpoint
- b. SRV opening time
- c. SRV mass flowrate
- d. Discharge line length
- e. Discharge line submergence depth
- f. Discharge line diameter
- g. Equivalent friction factor
- h. Initial line pressure and density DEPENDENT VARIABLES CONSIDERED
- a. Pressure at SRV exit
- b. Pressure at discharge line exit
- c. Density at discharge line exit
- d. Water velocity at discharge line exit
- e. Vent clearing time O
5.2-15
ZPS-1-MARK II DAR AMENDMENT 13 l OCTOBER 1980 i 1 l ()T sm TABLE 5.2-2 SRV BUBBLE DYNAMICS PARAMETERIZATION INDEPENDENT PARAMETERS CONSIDERED
- a. Initial pressure
- b. Initial density
- c. Air inflow rate
- d. Duration of air discharge
- e. Initial bubble radial velocity
- f. Initial bubble radius DEPENDENT PARAMETERS CONSIDERED
- a. Bubble pressure
- b. Bubble radius
- u. Bubble frequency
- d. Bubble duration O
5.2-16
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 O inte 5 2-3 TRANSIENT ANALYSIS ASSUMPTIONS
- 1. Initial conditions - 105% nuclear boiler rated steam flow
- 2. Limiting event - closure of all main steam isolation valves
- 3. Minimum main steamline isolation valve closing times
- 4. End of equilibrium cycle nuclear conditions
- 5. ANS5+ 20% decay heat at infinite exposure
- 6. Nominal relief setpoints 2
- 7. Design specification relief valve dynamic response
- 8. Maximum expected relief capacity O
i O 5.2-17
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 TABLE 5.2-4 RELIEF VALVE INPUTS - ZItefER ANALYSIS 1 NUMBER OF OPEN SETPOINT CLOSE SETPOINT VALVES (psia) (psia) , 2 1091 1041 3 1101 1051 3 1111 1066 3 1121 1071 2 1131 1081 1 I l Opening delay = .4 seconds * (valve + pressure sensor delay) Opening stroke time = .15 seconds
- Closing delay = 1.2 seconds
- Closing stroke time = .3 seconds
- Capacity = 122.5% of ASME rated value
- Design / purchase specification limits O
5.2-18
O O O , TABLE 5.2-5 i ZIMMER TRANSIENT RESULTS FIRST ACTUATION SECOND ACTUATION THIRD ACTUATION MARGIN TO MARGIN TO NEXT NEXT , i NUMBER PEAK PRESSURE NUMBER GROUP NL5BER GROUP CASE VALVES (psia) VALVES (psi) VALVES (psi) ,
- 1. Base case, ATWS RPT off i (Figure 5. 2-1) 13 1143 5 4 5 1
- 2. Base case, ATWS RPT on (Figure 5.2-2) 13 1143 5 3 2 5 y 7
- 3. Repeat of 2 - ASME rated RV cap 13 1147 5 6 2 7 7 1
% i i ,a 4. Repeat of 2 - 75 psi blowdown 13 1143 5 8 2 7 M N
i t M [ 5. Repeat of 1 - 90% power 13 1133 5 6 2 7 i 5
- 6. Repeat of 2 .2 second RV open delay 10 1128 5 9 2 9
- i a
sn hh + a 9i E ! i g5 80 . i t I
AMENDMENT 13 O' OCTOBER 1980 IMS07ABIO O 7 0 i4-1 O 13.5 0 O IMS O 8 AB10 306* 1 pS08ACIO
/ -2 7. 5 '* 66*
IMSO9AA10 O 126* IMSO7ACIO p 34' 226* 40 , ASSUMED POSITION OF l" DISCHARGED BUBBLES - O EET FROM RAM'S HEAD IMSO9ABIO l C k 17 4 *
- N N 13 SECTION A-A NOTES:
S HE D flEORIENTED RADIALLY. MARM li DESIGN ASSESSMENT REPORT (~
\
- 3. AUTOMATIC DEPRESSURIZATION SYSTEM ACTUATION (SIMULTANE0US FIGURE 5.2-1 DISCHARGE OF 6 SRVs).
- 4. THE QUENCHERS N0W USED IN THE ORIENTATION OF SRV RAMS HEAD DEVICES ADS ARE LOCATED DIFFERENTLY. FOR AUTOMATIC DEPRESSURIZATION l SYSTEM ACTUATION
J AMENDMENT 13 8
- OCTOBER 1980 O R V O E
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= 2 9
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= ,-
lI I T iiei liiii - o 8 8 8 8 ' u = o e (VISd) 3HOSS38d WM. H. 2IMMER NUCLEAR POWER STATION. UNIT 1 MARK 18 DESIGN ASSESSMENT REPORT FIGURE 5.2-2 l REACTOR VESSEL CYCLING STEAM PRESSURE VS. TIttE - ATWS PUMP TRIP 0FF l l
g AMEN 0 MENT 13 OCTOBER 1980 O m 8 8 E a. 5 ti 0; i 8 C E s2 -
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s_ : g _ G y, * - 1 l 7 I t I !I I t i o o o o o o o o o U 9 m (VISd) 380SS38d WM. H. IBMMER NUCLEAR POWER STATION, UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT FIGURE 5.2-3 REACTOR VESSEL CYCLING STEAM PRESSURE VS. TIME - ATWS PUMP TRIP ON
I (' ~ ' '; AMENDMENT iJ Y f OCTOBER 1980
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PEDESTAL ,,. b CONTAINMENT
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ISCHARGE LINE g/ u /E /b 7.5, 3 m " 7 ~ ZONE NUMBER t < 1 i 3 5'
. .'. .':- ; 4 5 6 . , ,
T ' BASE M^T O 13.5- - 22.33' + 27.5' > 31.1 7 ' = 40'
- PEDESTAL RADIUS-13.5 FEET CONTAINMENT RADIUS-40.0 FEET POOL DEPTH-22.5 FEET SUBMERGENrE DEPTH-13.5 FEET NOTE: DRAWING NOT TO SCALE.
\VM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK ll DESIGN ASSESSMENT REPORT O FIGURE 5.2-4 CROSS SECTION Of SUPPRESSION P00L AND DEFINITION OF SUPPRESSION CHAMBER WALLS' LOADING ZONES
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 O U 5.3 LOSS-OF-COOLANT ACCIDENT (LOCA) LOADS Depending upon the rate of mass release into the drywell, LOCA events may be classified as design-basis accidents (DBA), in mediate break accidents (IBA), or small break accidents (SBA). In all cases, the wetwell and drywell will become pressurized. Typical temperatures and pressure transients are given in Figures 5.3-1 through 5.3-8. The original containment design adequately withstands loading due to these pressure and temperature transients. In addition to the pressure and temperature transients that were ' considered in the original design bases, certain pool dynamic i loads have also been identified to occur during a postulated i LOCA. These loads are described in the following subsections. , Following a postulated LOCA, the drywell pressure increasas due to blowdown of the reactor system. Pressurization of the drywell I causes the water initially in the vent system to be accelerated i out through the vents. During this water expulsion process, the i resulting water jets cause impingement and drag loads on local , containment structures. l Following vent clearing, an air / steam bubble forms at the vent ' (]) exit which causes a hydrostatic pressure increase in the pool water resulting in a loading condition on the pool boundaries. i The steam condenses in the pool. However, the continued addition and expansion of the drywell air through the vent pipes causes the pool volume above the elevation of the vent exit to rise, ' resulting in a rise of the pool surface. This phenomenon is ' referred to as pool swell. Upward motion of this slug of water i creates a drag load on structures submerged in the pool and ,
~
impact loads on unsubmerged structures located just above the initial pool surface. Before the pool has risen 1.5 times the ' initial submergence of the main vents, the rising slug of water ' breaks apart. Subsequent pool swell involves a two-phase i air / water froth which may produce further structural-impingement i loads near the elevation of the maximum pool swell height. This , entire process affects only those structures between the pool surface and a maximum height of 1.5 times the initial submergence ' of the main vents. A gravity-induced fallback of the pool ' returns the pool surface to the original elevation. At the time i of maximum pool swell height, the drywell floor can be subjected i to an upward lead due to an imbalance in pressure between the , compressed air in the wetwell free air space and the air-purged drywell volume. The capability of the drywell diaphragm floor to ' withstand these pressure transients is discussed in Section 6.1 8 of this report. i l
' l Following the pool swell transient, there will be a period of ,
(_~)3 high steam flow through the main vent system. At these high steam flow conditions, the water / steam condensation interface oscillates due to bubble growth and collapse. These condensation ' I . I 5.3-1
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 O oscillations' result in an oscillator" load on the pool boundary. At low vent flow rates, the water /st am condensation interface ' can oscillate back and forth in the vents, causing " chugging." ' The chugging action results in loads on both the downcomer vents i and the containment boundaries. The major Mark II-related experimental programs which confirm the design-basis load include the 4T full-scale, EPRI 1/13-scale, and ' PSTF 1/3-scale tests as referenced in Table 1.0-1. ' 1 The temporary tall test tank (4T) facility consists of a single i cell representative of a typical Mark II containment. The test , facility utilized a single full-size vertical downcomer in a tank. A total of 46 steam blowdown tests were conducted by GE in ' this full-scale facility during a test program consisting of ' three phases (NEDE 13442-P-I and NEDE 13468-P) . These tests i provided the primary data source from the DFFR LOCA pool dynamic i loads, including those related to pool swell (NEDE 21544-P), , wetwell pressurization, vent flow, pool thermal response, conden-sation oscillation, and chugging (NEDE 23617-P). The test matrix ' included a range of vent submergence, break size, vent size, i blowdown fluid, vent bracing, and initial conditions to reflect i Mark II plant-to-plant differences. , (]) Test results from the General Electric pressure suppression test facility (PSTF) supplied the data base for pool impact loads on ' representative small containment structures, including pipes, 8 I-beams, and grating. Impact pressures verrus pool velocity a correlations were developed from these data, which are used in , combination with calculated pool swell velocities and 4T data to establish the Mark II impact loads. The impact load data were obtained from PSTF test series 5805. ' 1 The Electric Power Research Institute (EPRI) Mark II test i facility consists of a 1/13-scale model of a typical Mark II , containmert system. The facility contains 21 vents and represents a 900 sector of the suppression chamber, including the pedestal region. The EPRI tests consisted of air-charged tests I in contrast to the 4T steam blowdown tests. About 90 tests were i performed by Stanford Reserrch ?nstitute for EPRI to provide data i related to Mark II pool swell phenomena (EPRI NP-441). , Specifically, data from these tests were used to verify the ade-quacy of the 4T unit cell approach to study pool swell phenomena, ' validate the 4T air / steam tests, and validate the DFFR pool swell 8 analytical model, i i In addition to the above test programs, the Mark II owners group , has provided information relating to tests conducted outside the scope of the Mark II program to support some of the loads ' gs specified in the DFFR. This includes data for steam blowdown () 8 tests from the Marviken test facility and data resulting from i tests in GE foreign licensee single and multivent, large-scale i facilities. Data from these tests were used in the Mark II , 5.3-2
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 f V) program to support the conservatism of single-vent tests for vent i lateral loads and pool boundary chugging loads. , 5.3.1 Desion-Basis LOCA Loads The loads described in this subsection are those used to design the containment structure, piping, and equipment. Subsection 5.3.2 describes assessments made to ensure that the design is adequate when the NRC Lead Plant Acceptance Criteria (NUREG-0487) are applied. 5.3.1.1 Load Definitions This subsection describes the load definitions which comprised the original 2PS-1 design basis. 5.3.1.1.1 LOCA Water Jet Loads Structures located near the projected area of the downcomer vent, which is about 2 feet in diameter, will experience a water jet impingement load acting along the jet axis during the vent clearing process. The jet will flow around structures less than 2 feet wide and cause drag load. For larger structures, the significant load will be due to impingement. The maximum (~3 velocity of the jet is 60 ft/sec; its dispersion angle is given
\_/ in Figure 5.3-9 (Reference 1).
5.3.1.1.2 LOCA Charging Air Bubble Load As the drywell air is forced into the wetwell, bubbles will grow at the vent exits. These bubbles will grow until they touch and pool swell begins. Submerged structures and boundary loads calculated during this portion of the event are generally much less than loads during other phases. 5.3.1.1.3 Pool Swell The pool swell transient is conservatively predicted by MKII-SWELL (Sargent & Lundy implementation of PSAM described in Reference 1). During pool swell, structures initially above the vent exit and below the maximum pool swell height (1.5 times vent submergence) will be loaded by drag loads. Structures initially above the pool may experience impact loads. The compression of the wetwell air space will result in an upward force on the drywell floor. 5.3.1.1.4 Pool Fallback l During fallback, the air mixes with the pool water slug and
-s greatly reduces or eliminates drag loads. However, it is l
(') conservatively assumed that the slug remains intact and falls back into the pool under the force of gravity, causing drag loads. 5.3-3 l
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 5.3.1.1.5 Condensatior. Oscillation Following the pool swell transient, steam flows through the main vent system into the suppression pool, where it condenses. Evaluation of the steam-condensation phase of the 4T test results revealed the existence of a dynamic load during high steam mass flux into the suppression pool. This load, called condensation oscillation, is a low-amplitude, symmetric, pseudo-sinusoidal pressure fluctuation occurring over a range of frequencies. The DFFR load definition was used in the original design of the 2PS-1 containment, piping, NSSS, and equipment. This load is defined as a sinusoidal loading with a magnitude of
- 3.75 psi and a frequency range of 2-7 hertz. Subsection 5.3.2.5 describes a reassessment for a more conservative load definition. This load definition, referred to as the Zimmer Empirical Limiting Condensation Oscillation Load, is described in detail in Section 2.1.
3.3.1.1.6 Chuacina Condensation oscillation occurs during the high and medium mass flux phase of a blowdown transient. At later times (or for smaller breaks), the steam mass flux will ba lower. Under these (~ conditions, the water / steam conJensation interface can oscillate in the vents, causing " chugging." The chugging action results in lateral loads on downcomer vents and also in dynamic pressure loads on the wetted containment boundary. These boundary loads produce structural responses in the containment and reactor building. The application of bounding chugging loads is described in the
" Mark II Phase I - 4T Tests Application Memorandum," submitted to the NRC in June 1976. This application memorandum is currently being used for 2PS-1 containment assessment to expedite licensing review.
i A study of chugging impulse was made to demonstrate that containment response to chugging using the load specification in t the 4T Bounding Loads Report (NEDO 23617, July 1977) is greater than the response obtained using a realistic definition for ' chugging loads which does not include the system response of the 4T test tank. i i The present Mark II chugging load specification is based on the application of chug pressure histories measured at the wetted boundary of the 4T test facility. These chug pressure histories 8 were applied in a conservative manner directly to the suppression i pool boundary of the Mark II containment. These 4T boundary i ,,s pressures include both the chug excitation and the system i (,) response of the 4T system. 5.3-4
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 l ($)
! provide a realistic chugging load definition, 4T boundary pressure histories have been reviewed to identify the chug source '
impulse, i.e., the impulse forcing function at the vent exit. I This chug impulse is then confirmed using a coupled fluid-struc- i ture model of the 4T facility. The response (total pressure at i water / tank interface) of the 4T is calculated by applying the , chug source impulse at the vent in the 4T model and comparing this response with actually recorded 4T test data. The results from this study have indicated that the chug impulse and not the . detailed chug pressure history are important in determining tha i 4T wall pressure response. The chug impulse used was a i triangular-shaped underpressure with a duration of 5 to 10 msec. , For example, a triangular pressure pulse at the 4T vent, 10 msee in duration and -28 psi in amplitude, will produce a wall pressure response of 25 psi. This is consistent with the impulse i obtained from integration cf the chug wall pressure traces of i this amplitude. , From the impulse chug source defined at the 4T vent, an incident wall pressure chug source has been obtained. This is the wall 8 pressure which produces the same response at the 4T wall as i obtained when the chug source is applied at the vent. i i The containment structure response is obtained for an axisymmetric application of the improved chugging Icad (incident {' pressure) definition which is consistent with the axisymmetric 8 load specification in the 4T boimding load definition. e i The response of the containment structures for the two chugging , load definitions (present/ bounding and improved) presented to the NRC on May 17, 1978, indicate the conservatism inherent in the ' present/ bounding load specification which includes both the 8 chugging excitation and the 4T response. For the 2PS-1 i containment, the improved chugging load definition produces a i maximum displacement wall response which is less than that , obtained from the present 4T bounding chug load specification. i 5.3.1.1.7 Lateral Loads on Downcomers This subsection describes the application of lateral loads to the original design configuration of the dowcomer vents (no bracing). The revised design (including a lower bracing system) is discussed in Subsection 5.3.2.4. The NRC staff has questioned the direct application of the foreign licensee data reported in NEDO 21018 to the Mark II ' plants. Accordingly, they have requested that 4T test data also i be considered in justifying the load definition. These questions i have been thoroughly reviewed. It is concluded that the design , lateral load of 8800-lbf applied at the tip of each downcomer () represents an upper-bound load. The drywell floor is assessed for all the downcomers simultaneously loaded with 8800 lbf lateral load in any direction (see Section 7.1). The downcomers i 5.3-5
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 ("I in the original design were not braced, and the natural frequency i of the empty downcomer was less than 7 Hz. If the hydrodynamic i mass is considered, the frequency wculd be even lower, thus the , 8800 lbf bounding lateral load was applicable to the original design of 2PS-1. 8 i Based on the structural response data measured on the 4T i downcomer during phases I, II, and III, NEDE ;.3806-P defines the i dynamic lateral load on the downcomer as follows: , The lateral chugging load can be represented by the half-sine ' wave of duration ranging from L to 6 msec for high and low i intensities, respectively. The maximum load amplitude ranges i from 10,000 lbf to 30,000 lbf and is considered tc be uniformly i distributed over a 1- to 4-foot length of the downcomer end. , This transient load is represented by: ' F(t) = A Sin , lateral load (1bf), where: , 10* < A < 3 x 104, maximum amplitude (1bf), and I i 3<r < 6, application period (msec). i The comparison of maximum design forces from a dynamic analysis of the downcomer subjected to the transient load and from the ' application of 8800 lbf (static) is given below: 8 LOAD MAX VENT MOMENT MAX VENT SHEAR i 8800 lbf 301.0 ft-kip < 0.5 ksi (static) ' i 10,000 - 30,000 lbf 64.6 ft-kip < l.0 ksi 6 msec - 3 msec , (dynamic) , The above comparison demonstrates that the design lateral' load of ' 8800 lbf represents an upper bound foc 2PS-1. 8 The NRC also recently questioned lateral loads on the downcomer vent at the instant of vent clearing. It should be noted that in the 4T test series, no significant lateral loads were observed between the start of the test and the onset of chugging. However, in the referenced tests (Table 3-3 of NEDE 21078), static equivalent measurements of lateral loads up to 3.5 kips were observed. These loads are thought to be unique to the test setup (Figures 3-1, 3-1A, 3-2, and 3-3 of NEDE 21078) and not (j, applicable to either the 4T facility or the Mark II containment. 5.3-6 i
, 2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (] v In the test facility where the 3.5-kip static equivalent Ic Js were measured, there is effectively no drywell volume except the air occupying the vent line. In contrast to tne 4T facility or a Mark II containment, in the referenced tests only a very small quantity of noncondensible gas is vented to the pool, after which steam condensation occurs immediately. In these tests without a drywell, the vent pressure typically rose to approximately 1 to 1.5 atm while the small air volume cleared. It then dropped approximately 2 atm from this point as condensation of the steam commenced. This reduction in vent pressure is evidence of the collapse of the bubble at the vent exit and an attendant reentry of water into the vent. The bubble collapse (similar to a chugging event) causes the ' lateral load (during vent clearing), which would not have occurred if a drywell were present as in the 4T facility and the Mark II containment. In this case continuous air flow to the bubble at the vent exit would gradually be diluted with a larger flow of steam (which in itself is capable of maintaining a positive bubble pressure at the vent exit). In the absence of a collapsing vent exit bubble, significant lateral loads would not be expected to occur during the 4T or Mark II vent clearing transients, and this was confirmed in the 4T tests. l Nevertheless, m ince the 5,8-kip load used to evaluate the ' downcomer and drywell floor is greater than the 3.5-kip load ' (]) postulated to occur at the instant of vent clearing, it is I clearly demonstrated that the governing design case for the i containment has been considered. i The drywell floor is loaded by the chugging lateral loads. The procedure used for evaluating the structural adequacy of the downcomer and drywell floor is explained in detail in Subsection 7.1.3. The current downcomer and bracing configuration has been assessed in conformance with NUREG-0487. This assessment is described in Subsection 5.3.2.4. 5.3.1.2 LOCA Boundary Loads ' Boundary loads caused by LOCA are analyzed using the same nodalization technique as for SRV boundary loads (Subsection 5.2.1.3). The boundary loads described here are those used in the 2PS-1 design basis. The load definitions are discussed in greater detail in Subsection 5.3.1.1. Reanalysis to meet the , requirements of NUREG-0487 is discussed in Subsection 5.3.2. ; 5.3.1.2.1 LOCA Water Jet * ! The impingement load due to water jet is 48.13 psi at the vent exit and 33 psi at the basemat (Reference 1). The duration L5 O this load is 0.85 sec starting at the initiation of LOCA J 5.3-7
,_ .__ _ _ _ . . - - _ _ _ _ _ _ ._____.- - _ . . . . _ _ ~ . . - _ _ _ . . .
1 2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (Reference 2). Figure 5.3-10 gives the time history of the jet O impingement load. The total force can be determined by multiplying the pressure given in Figure 5.3-10 by the projected area in square inches perpendicular to the jet. 5.3.1.2.2 CEaraina Air Bubble The boundary loads during vent clearing have been assessed for the time of bubble growth at the vent exit. This load is less than the containment design capability. 5.3.1.2.3 Pool Swell The containment wall, pedestal wall, and the basemat do not experience any boundary load due to pool swell phenomena. How?ver, the drywell floor is subjected to an uplfft force during pocl swell as a result of the compression of the wetwell air ' space. This load is computed by comparing the transient pressure history of the wetwell air space with the drywell pressure historv during pool swell. The maximum load for 2PS-1 is 2.5 psi. This is verified by 4T test results. 5.3.1.2.4 Pool Fallback
.nis phase of the transient causes no significant pool boundary loads.
O 5.3.1.2.5 Condensation Oscillation The condensation oscillation load is applied to the containment boundary walls assuming that the load occurs in phase at all vents. The load is described in Subsection 5.3.1.1.5. 5.3.1.2.6 Chuacina Symmetric and asymmetric loads are applied to the containment boundary loads as shown in Figure 5.3-11. The asymmetric load is defined by assuming that the maximum load occurs on one side of the containr. eat while the load is minimized on the other side. This conservative assumption results in the bounding horizontal loading due to chugging. The load is uniform below the vent exit and attenuates linearly to zero at the pool surface. 5.3.1.3 LOCA Submerced Structure Loads Submerged structure drag loads result when the LOCA-related phenomena cause fluid motion in the suppression pool. All initially submerged structures and those in'the pool swell zone will potentially experience submerged structure loads. The LOCA submerged structure loads described here are those used in the ZPS-1 design basis. The load definitions are discussed in greater detail in Subsection 5.3.1.1. {)g Reanalysis to meet s NUREG-0487 is discu sad in Subsection 5.3.2. j l 5.3-8
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 0 5.3.1.3.1 LOCA Water .Tet The drag load due to water jet is (CD x 24.06) psi based on the jet velocity of 60 ft/sec (Reference 1). The duration of this load is also 0.85 sec (Reference 2). The time history of the drag load due to water jet is giveri in Figure 5.3-12. The total force can be determined by multiplying (Co x 24.06) by the projected area in square inches. This drag lead acts on any structure located between the vent exit and the basemat. The direction of application of this' load will be either in the horizontal or in the vertical directions, i.e., all structures experiencing this load must be designed to withstand the load if applied from these directions (Reference 1). Subsection 5.3.2.1 describes a reassessment performed in response to NUREG-0487. 5.3.1.3.2 Charging Air Bubble i The vent clearing transient results in submerged structu a Acads of relatively low magnitude which are bounded by loads during the steam condensation events. 5.3.1.3.3 Pool Swell (~
\
During pool swell, structures are loaded by impact and by drag loads. Impact loads occur only above the initial pool surface up to tha maximum pool swell elevation. Drag loads affect struicures above the vent exit and below the maximum pool swell elsvation. 5.3.1.3.3.1 Pool Swell Impact Loads The impact load depends upon the size and shape of the structure and the velocity of the pool surface at the elevation of the structure. During design, effort was made to exclude structures from the pool swell ~ zone. Because of this, relatively few structures are affected by impact loads. The impact force due .o pool swell occurs over a time period ty . The typical force versus time profile is such that the force increases to a maximum value ducing the first half of the time period and the decreases to the value of the drag force during the second half. The duration, t, f varies from about 7 msec for sr.all structures to about 100 msec for large structures. These ; impuct loads are based on the assumption that pool surface veJc:ity vector is perpendicular to the longitudinal axis of the ! body. The load in any other direction will be the vector ' component in that direction. (]) For design purposes, small structures would mean pl pes, .I-beams, and other similar structures having any one dimension less than or equal to 20 inches (Reference 1). All structures in the ZPS-1 5.3-9
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 O pool swell zone cre classified as "small structures" for impact load design. The loads on small structures have been determined from tests conducted by the General Electric Company (Reference 1). The data reduced from these tests are plotted as average peak pressure, Pave, experienced by the structure versus elevation. Figure 5.3-13 is the plot of measured impact pressure for pipes and I-beams versus elevation. These results include a design margin of 50% (Reference 1), i.e., the measured impact values have been increased by 50%. The results shown in Figure 5.3-13 can be used to determine the maximum force per unit length, Fmax/L, on pipes of different sizes and I-beam located at different elevations above the initial pool surface. Figure 5.3-14 gives the maximum impact force on pipes, and Figure 5.3-15 gives the force on I-beams. The foregoing results give the maximum pressure and force experienced by a structure during impact. Knowing this maximum value, the transient profile can be constructed by using the normalized profile shown in Figure 5.3-16. The abscissa in Figure 5.3-16 is the duration of the impact load, ty. This duration is the same for all small structures; its value is tr = 7 msec. The ordinate in Figure 5.3-16 is normalized pressure or normalized force. The normalization in the former case is done with respect to Pmax, Figure 5.3-13, and in the O latter case witn respect to F ax The conversion from pressure To /L, Figures 5.3-14 and 5.3-15. force is based on the projected area at the diameter in che case of pipes, and on the area of the bottom surface in the case of I-beams. To summarize, the following procedure is out'ined for determining the impact loads on small structures:
- a. Determine the elevation at which the structure, for which the impact load is to be calculated, is located.
- b. For this elevation, the maximum impact pres ure, Pm ax, is obtained from Figure 5.3-13, or the mazi.um impact force per unit length, Fmax/L, is obtained from Figures 5.3-14 and 5.3-15
- c. The pressure and/or force profiles are then determined using Figure 5.3-16.
The results given for the impact loads are based on pipes and I-beams. These results should also be used along with engineering judgment to determine impact loads on other similar structures. Additional impact load calctlations in response to NUREG-0487 are described in Subsection 5.3.2.2.2. 5.3-10
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 0 5.3.1.3.3.2 Pool Swell Draq Loads Structures above the vent exit elevation and below the maximum pool swell elevation will experience drag loads. The velocity and acceleration history at the elevation of the structures, as predicted by pool swell analytical models, is used to calculate the drag load magnitude. The duration of the load will depend upon the pool swell velocity history and the elevation of the structure. 5.3.1.3.4 Pool Fallback Drag loads are calculated during fallback in a similar manner to the pool swell calculation. The pool slug velocity and acceleration time history are calculated assuming that a slug of water of thickness equal to the submergence depth of the downcomer falls freely under the influence of gravity only. 5.3.1.3.5 Condensation Oscillation Draq Loads i The submerged structure loads due to_ steam condensation are cal-culated making use of the analytical methods developed to predict transient water jet loads and oscillating air bubble loads. I These methods have been documented in reports NEDO/NEDE 21471-P, I 21472-P, and 21730. (]) i The same basic approach and fundamentals that are applied to air bubble-induced loads are applied to steam bubbles (see NEDE-21471 ' and NEDE-21730). The steam bubbles are treated as stationary, ' ficite-sized multiple sources. The resulting potential gradient i dist-ibution within the bounded pool is determined by utilizing i the m?thod of images. The total drag loads due to both standard and acceleration drag are determined for each submerged structure , as follcws: ' s C A UlUl Standard drag, Fs= x (3.3-1) i 2gc i pV b ' Acceleration drag, FA= a (3.3-2) i 9c ' I where: i p = pool water density, ' s A x = area of structure normal to flow direction, i I i CD = standard drag coefficient, , Va = acceleration drag volume, '
)
l l 5.3-11 I
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 ( i U = fluid velocity, and , 0 = fluid acceleration. 8 The source strength for steam condensation oscillation loads is i derived from the 4T data (NEDE-13468P). The maximum pressure , oscillations for this phenomenon were observed on the bottom of the tank and are bounded by a
- 3.75-psi value. The load history is considered to be sinusoidal, with an amplitude of
- 3.75 psi 8 and a frequency range of 2-7 hertz. An equivalent source i strength at the downcomer exit is derived from the maximum i observed load on the bottom of the tank. The derivation ,
considers the finite size of the steam bubble and the 4T tank and ' downcomer vent configuration. This derived source strength is then used directly in the ZPS-1 specific submerged structure load I determinations. Condensation oscillation is treated as occurring i simultaneously and in phase at all the downcomer vents, i 5.3.1.3.6 Chuacino Drac Loads The submerged structure loads due to chugging are computed on the I basis of source strengths derived from 4T data in a manner i analogous to that described in Subsection 5.3.1.3.5. Again, the i maximum observed loads on the tank bottom are used eo establish , bounding source strengths. O The bounding positive load used is 20 psi and the bounding negative load is -14 psi. The equivalent ' source strengths at the downcomer exit for each of these loads is ' derived as in the case of the condensation oscillation loads. The frequency range used for chugging loads is 20 through 30 i hertz. Main vent chugging is a stochastic phenomenon in both , occurrence (timing) and load magnitude. Thus, the number of } possible permutations for multiple main vent chugging ' combinations considering relative timing and load magnitude is I immense. The relative location and orientation of the source and i the submerged structure of interest is a major consideration in , defining the forcing functions on that structure. This consideration is complicated when multiple sources are prasent, since their relative phasing is important. This results tirectly ' from the fact that the load on a submerged structure is produced i by the differential pressure that exists across the structure. i Hence, when multiple sources are considered, the situation where , sources on Opposite sides of the submerged structure are out of phase will proacce a larger load than the situation where the sources are in paase. The following conservative approach is ' being utilized as an interim method for the 2PS-1 plant. The i randomness in multiple source timing and phasing is accommodated i by considering the worst case. The worst case is defined for , each specific application such that the pressure gradient across the given structure,is maximized. ' It is recognized that as the number.of niain vents considered to be participating in the ' (]) chugging event increases, the probability of the assumed bounding i source configuration decreases rapidly and soon becomes incredibly small. Thus, as the number of participating main , 5.3-12
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
/~')
o vents increases, the source strength for each is diminished by a i factor as shown in Figure 5.3-17. For example, if five main vents are being considered, the derived 4T source strength for each vent is reduced by multiplying by the factor 0.57. The curve in Figure 5.3-17 is based on Figure 4-10b ' of NEDE-21061-P, which has been extended to range from 1 to 100 i downcomers and then normalized. The source strengths are thus i specified as a function of the number of main vents considered to , be contributing to the loading on the given structure. This , procedr:re is applied by initially considering only the main vent of closest approach to the given structure to be chugging. The ' load is determined for that configuration. Next, the bounding, i combination of the two closest main vents is used to define the i load. This procedure is repeated and each time one additional , downcomer is considered in the load determination. Each successive downcomer is farther (spatially) from the given submerged structure and, hence, after the third or fourth down- I comer, the increment of load increase is small. At the same i time, the likelihood of each specific configuration decreases as i the number of vents increases so the source strengths also , decrease. Hence, the load on the submerged structure goes through a maximum as the number of main vents is increased. The ' combination which produces the largest load on the given struc- ' ture is used for the design assessment. i
)
The procedure described above is quite conservative and is thus , considered to only be an interim procedure. Nevertheless, in an effe~t to expedite the 2PS-1 licensing schedule, this interim ' procedure is being used to evaluate steam condensation loads on i submerged structures in the ZPS-1 suppression pool, i 5.3.1.4 Annulus Pressurization Annulus pressurization refers to the loading on the shield wall ' and reactor vescel caused by a postulated pipe rupture at the i reactor pressure vessel nozzle safe-end to pipe weld, This i loading is in the form of a transient asymmetric differential , pressure described below. 5.3.1.4.1 Transient Asymmetric Differential Pressure Events ' for Postulated Pipe Ruptures i Mass and energy are released during postulated pipe ruptures l ! causing:
- a. a rapid asymmetric decompression acoustic loading of 8 the annular region between the vessel aad shroud from i a recirculation inlet pipe break at or beyond the i pipe to reactor pressure vessel nozzle safe-end weld, ,
and ())) , 5.3-13
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (O V
- b. a transient asymmetric differential pressure event i within the annular region between the biological i shield wall and the reactor pressure vessel (annulus ,
pressurization) from feedwater and recirculation pipes breaks at the pipe to reactor pressure vessel nozzle safe end weld. Associated with this ' postulated event are: i I
- 1. a jet stream release of the reactor pressure ,
vessel inventory, and ,
- 2. impact of the ruptured pipe against the whip '
restraint attached to the biological shield wall. i i See Figure 5.3-18 for the time sequencing of these events. , 5.3.1.4.1.1 Acoustic Loadinq ' I The acoustic loading results from a postulated recirculation pipe i rupture at any location as shown in Figure 5.3-19. However, i because the boiling water reactor (BWR) system is a two-phase , system that operates at or close to saturation pressure (1000 psi), short-duration differential pressure is developed across ' the reactor shroud and the BWR system is not subjected to a I ("T significant shock-type load with respect to structural supports. (/ This short-duration acoustic load is confined to'a bending moment i and shear force on the reactor pressure vessel reactor shroud , support. Typical results of the integrated force acting on the reactor pressure vessel shroud are given in Table 5.3-1. ' i 5.3.1.4.2 Annulus Pressurization - Design Considerations 1 The annulus pressurization event results from a postulated feed- , water or recirculation pipe rupture at the reactor pressure vessel nozzle safe-end to pipe weld. The oipe is assumed to have ' instantaneous guillotine rupture, allowin) mass-energy release e into the drywell and annular region between the biological shield i wall and reactor pressure vessel, as shown in Figure 5.3-20. , However, the energy absorbing pipe whip restraint restricts the pipe separation to less than one full pipe diameter (see Figure ' 5.3-21). (Jet reaction and pipe whip restraint loads are ' calculated as described in ANSI 276 (draft), " Design Basis for i Protection of Nuclear Power Plants Against Effects of Postulated i Pipe Ruptures," January 1977.) This restricted separation, which , occurs after the assumed instantaneous guillotine rupture, is accounted for as a finite break opening time in the mass / energy release calculation. The mass / energy methodology additionally 8 accounts for the effects of the subcooled inventcry initially i present in the line. The feedwater line is conservatively i assumed to fully separate, and the mass / energy release is , (~T calculated as described in Appendix D. u.) ' 5.3-14
. -. . .= . - - - . - . _ . _
2PS-1-HARK II DAR AMENDMENT 13 OCTOBER 1980 4 () The short-term (first'few seconds) mass / energy release calculational method (Appendix D) is intended for use only during the first few milliseconds of the line break transient and conservatively assumes that no vessel depressurization occurs. 8 For calculation of the remainder of the line break transient, i which involves vessel depressurization, the methods described in Reference 1 are employed. Development of the 0.5 multiplier , utilized on the pipe side discharge flow during the inventory period is described in Appendix B of Reference 3. ' s For the case of reactor shield wall' annulus pressurization, the i short-term mass / energy release methodology has been separately , applied to obtain break flow rates including credit for finite , break opening time in the case of the recirculation line. 5.3.1.4.3 Annulus Pressurization - Desion Analysis 8 The followi.ng desigr; analysis has been performed to assess the i plant for the effectL due to annulus pressurization. , 5.3.1.4.3.1 Calculation of Mass and 2nercy Flow Rates 8 The assumption and methods used in the calculations are presented i in Appendix D. Appendix D describes the short-term mass energy i (] release methodology used for the recirculation line break and v describes the mass / energy release calculations used for the ' recirculation and feedwater line break. ' 1 5.3.1.4.3.1.1 Comparison of General Electric Analysis ! to RELAP , For the annulus pressurization event the NRC has queitioned ' General Electric's method for computing mass and energy flow I rates following a postulated LOCA from long lines containing i subcooled fluid. A program was developed to expedite the i licensing of 2PS-1 to perform RELAP analyses using appropriate , assumptions and to compare the results with those obtained using ' General Electric's method. The assumptions (or ground rules) applied to these analyses are as follows: I
- a. Fee 6 water line:
- 1. 2PS-1 RELAP deck as basis.
- 2. Use Henry-Fauske-Moody flow model. 8 1
- 3. Instant break opening.
- 4. Eliminate mas flux terms between vessel and break (short side). '
()
- b. .tecirculation line i 5.3-15
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 A (_) i
- 1. 2PS-1 RELAP deck as basis. ,
- 2. Allow for finite break opening time. 8 i
- 3. Use Henry-Fauske-Moody flow model. i
- 4. Eliminate momentum flux terms in RELAP between '
vessel and break (short side). The resulting break flows using the GE methods are much more con- i servative than those obtained by the use of RELAP. The results i using both methods are shown for comparison on Figures 5.3-22 and , 5.3-23. i RECIRCULATION LINE BREAK ' s Figure 5.3-24 shows the nodalization scheme adopted for the 2PS-1 recirculation line break. The analysis takes into account the , break opening due to the piping dynamics. Since there is mass flow from both ends of the break (vessel and inventory side), the ' break areas must be distributed between both ends. Since the 8 exact distribution cannot be determined, it was decided to i connect both ends of the break into a common node (hereafter i called the break node). This break node, in turn, connects to ' (~'T the containment drywell through a junction whose area is taken as \> the time-dependent break area. Calculations were carried out up ' to the point where quasi-steady choked flow was established. In i order to calculate the stagnation pressure in the break node, i momentum flux was suppressed in the junction that connects the i vessel to the safe end (Node 30 that is in between the vessel and ' the break node). I The plot of total break flow versus time is shown in Figure i 5.3-22 against the calculations done by Sargent & Lundy based on i the GE methodology. There are two important differences between i the results of the GE methodology and the RELAP analysis. GE methodology yields an initial rate of change of flow much larger than RELAP analysis. This is mainly because of the absence of 8 flow inertia in GE methodology as opposed to RELAP analysis. i Flow inertia always tends to decrease the rate of change of flow. i The other main difference is that the quasi-steady choked flow in the GE method yields a higher value because of the absence of frictional losses. Frictional losses produce pressure drop, ' which in turn reduces flow. Choked flow was calculated by the -i Henry-Fauske-Moody model in the RELAP analysis. i FEEDWATER LINE BdEAK The nodalization for the feedwater line break is shown in Figure 8 (,,'.) 5.3-25. The nodalization was done between the RPV (Node 1) and i the feed pump exit (Nodes 37 and 38). Included in the 2PS-1 i nodalization are feedwater heaters, which are modeled as nodes. , 5.3-16
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 v The double-ended break occurs instantaneously at the junction of i 33 and 14. Check valves were placed so as to prevent back flow i from the RPV into Nodes 3 and 4. The spargers in 2PS-1 were , treated as separate nodes (14, 39, 40, and 41) from the piping to which they connect. Spargers were connected to the RPV (Node 1) ' by the limiting area. The vessel was assumed to be at the i saturated condition. The enthalpy in the feedwater line was i assumed to be 397.6 Btu. This is consistent with the S&L i assumption. Momentum flux in the junctions connecting the vessel , to the spargers was suppressed to calculate the stagnation pressure in those nodes. Choked flow was calculated by the Henry-Fauske-Moody model. 8 i The total break flow (the sum of vessel and inventory sides) is a plotted against the S&L calculations based on GE methodology , (Figure 5.3-23). The S&L calculation is far in excess of the RELAP analysis for two reasons. The enthalpy of the vessel in ' the S&L calculations was 397.6 Btu (highly subcooled), whereas in 8 the RELAP analysis, it is saturated. As is well known, subcooled i choked mass flow rate is much greater than saturated choked flow. The other point of difference is that S&L used as area of 0.722 , for both sides of the break. The break area in the case of the RELAP analysis is the full pipe area (0.6303) for the long side ' of the break. For the vessel (sparger) side of the break, ' (~) limiting sparger flow area is 0.36 fta, much smaller than the S&L i
~ value. ,
RELAP analysis of the feedwater line and recirculation line break ' on 2PS-1 indicate that the S&L calculations of break mass flow ' according to GE methodology in the above cases are very I conservative. 5.3.1.4.3.2 Application of Mass-Energy Release to Compute Force-Time Histories of RPV and Shield Wall ' The mass-energy release for the reactor recirculation and feed- i water line breaks (computed by the GE method) were used in calcu- i lating the force time histories on the shield wall and reactor , pressure vessel. The model used for this analysis and the assumptions were presented to the NRC at a meeting with the Cincinnati Gas & Electric Co., Sargent & Lundy, and General I Electric Co., held in September 1977. I I The pressure responses of the RPV shield wall annulus to a , postulated pipe rupture at the RPV nozzle safe end to pipe weld for a recirculation suction line and a feedwater 1.ine were investigated using the RELAP4 computer code. A symmecric model ' consisting of 36 nodes and 76 flow paths was used in the analysis i of the recirculation line break, while an asymmetric model using 7s 36 nodes and 94 flow paths was developed for the analysis of the , t
! feedwater line break. Further description of these analytical models and detailed discussion of the analyses may be found in '
the answer to Question 041.20 in the 2PS-1 FSAR. ' 5.3-17
i 2PS-1-MARK II DAR AMENDMENT 13
- OCTOBER 1980 i O The pressure histories generated by the RELAP4 Code were in turn used to calculate the loads on the sacrificial shield wall and '
the reactor pressure vessel for each of the line breaks 8 considered. The annulus was divided into seven zones and an i eighth order Fourier fit to the output pressure histories made 5 for each zone to produce the Fourier coefficients required for , the structural analysis of the shield wall. The specific loading data requested by GE for the NSSS adequacy evaluation for both ' postulated line breaks consisted of the time-pressure (psia) and 8 time-force (1bf) histories for each node within the annulus, i These time-force histories represent the resultant loads on the i RPV for each node through its geometric center and were generated , by taking the product of the node pressure and its ' effective' surf ace area, ny , or more formally as: i
+A0/2 i F PER y cos Od0 -
EP g i y{ =
-A0/2 ,
t
= Pf2EgR v sin (a0/2) '
P.E nD 4 = Pn l j 4f 8 ' f , Q where: Fy = nodal resultant force on RPV (lbf), i i Pg = node absolute pressure (pnia), ' gg = node height (inches), , Ry = RPV radius (inches), 8 8 40 = azimuthal width of node (degrees), and i i ' Dj = pipe OD (inches). 5.3.1.4.3.3 Acceleration Time-Histories and Response - 8 Spectra Generation i i The force time histories (Subsection 5.3.1.4.3.2) were then , applied to a composite lumped-mass model of the pedestal, shield wall, and a detailed representation of the reactor pressure ' vessel complex described in Appendix F. This analysis produced I acceleration time histories at all nodes for use in evaluating i the attached piping and reactor pressure vessel for the effect of i annulus pressurization of either a feedwater or reactor , recirculation system postulated rupture. Response spectra were also generated at all nodes. ' I
- O The reector pressure vesset end the attached vigino were evaluated using these acceleration time histories in performing a i
dynamic time-history analysis (see Sections 6.4 and 6.5). , 5.3-18
, g , ,w- , . - -- ,n ..,.n-----r-,~ , - , - - .m, -, y.--- m--.~,e g
l 2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 m) These acceleration time histories were also used in determining i the maximum forces and moments to evaluate the vessel hold down i bolts, as discussed in Section 7.4. i 5.3.2 Assessment of NRC Acceptance Criteria - LOCA Section 5.4 provides the response to NUREG-0487, NRC Lead Plant Acceptance Criteria (September 1978). It contains a detailed description of how the intent of NUREG-0487 has been met. The following subsections describe areas where reanalysis has been done to meet the acceptance criteria. 5.3.2.1 LOCA Water Jet Loads , The NRC Lead Plant Acceptance Criteria required LOCA water jet ' loads to include the effects of a spherical vortex of fluid I traveling with the jet front predicted by the Moody jet model i (Reference 4). This procedure is expected to yield conservative , results because the Moody model predicts jet penetrations much greater than those observed in tests. ' I In response to Criterion III.A.1, the LOCA water jet loads have i been reassessed by several methods. The first is essentially i Acceptance Criterion III.A.1, incorporating a modification to the , f') Moody methodology to overcome mathematical difficulties. The second is an adaptation of the method described by Abramovich and , Solan (Reference 5). This method conforms to the intent of the 8 acceptance criteria, but describes the vortex motion by applying a conservation of momentum rather than using the Moody model. A i final method is the ring vortex model which is proposed by the , Mark II Generic Program. The NRC Acceptance Criteria utilizing the Moody jet model results I in a vortex with a motion described by a locus of points. These i points are found by tracking a number of constant velocity i particles exiting from the downcomer and locating the points , where a particle is overtaken by the one exiting after it. This calculation is easily done for a jet with constant acceleration ' but causes difficulties when applied to a jet of increasing i acceleration. When the Moody method is rigorously applied, i depending upon the coordinate system chosen, the jet is predicted i to reverse and move back to the vent, or time at the jet front , reverses. This result leads to mathematical discontinuities and ' is unacceptable. i An alternate method has been applied which resolves these i problems while conforming *o the intent of the original NRC i Acceptance Criteria. If the jet front position and velocity are , described at any time by the particle having traveled the farthest, the jet motion is well behaved until the jet is ' 7s (',) terminated. High accelerations are experienced near the end of 8 the transient that are overly conservative. i 5.3-19
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 O The vortex motion after vent clearing can be calculated assuming i it continues through the pool. The water jet is in fact . i dissipated in the turbulence caused by flow of air into the pool. Calculations show that, until vent clearing, LOCA water jet loads , on submerged structures such as the quencher arms in the ZPS-1 suppression pool are negligible (less than 10% of design values). 8 Higher loads are calculated on the quencher arms if the vortex is i allowed to continue after the air clearing until it impacts the i quencher arm. However, these loads are also within the design i capability of the quencher. The calculations conservatively used , direct jet impingement on the quencher arms (the arms are offset in the actual plant), and no interchange of mass between the jet ' and pool. The vortex was considered a rigid sphere in i determination of the drag load which retards its motion. i i The second method is similar to that described above but uses a . different method to describe the vortex motion. Following Abramovich and Solan (Reference 5), the motion and size of the ' vortex may be described assuming that momentum and mass are i conserved as the jet forms the vortex. Momentum is lost only i through drag on the fluid sphere. i The resulting motion of the vortex is similar to that calculated ' l previously, but without the unrealistic high accelerations noted 8 Q above. The loads are lower throughout the transient. This result is again conservative because interaction between the a i vortex and pool (other than rigid body drag) has been ignored. , The Mark II Generic Program has proposed a ring vortex model of ' the LOCA water jet. Preliminary results indicate this model ' predicts existing experimental data well (Reference 6) and will i result in lower loads than the methods described above. i Based on the above evaluations,.it is judged that the LOCA water jet loads for ZPS-1 have been evaluated in accordance with the ' intent of the NRC criteria. As indicated, additional evaluations I were done which demonstrate the conservatism of the evaluations. i The results of these evaluations were that the loads on the i quencher were negligible relative to the controlling quencher , design loads. 5.3.2.2 Epol Swell. The acceptance criteria stated in NUREG-0487 suggested several changes in the method of calculating pool swell loads. In the following subsections, the changes are discussed and compared with the original methods. 1 5.3.2.2.1 Pool Swell Velocity * (f The pool swell velocity has been increased by 10%. This change 4 increases-all pool swell drag and impact loads. Reassessment shows that the design is adequate for this load. ; 5.3-20
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 r3 V 5.3.2.2.2 Pool Swell Impact The acceptance criteria furnished an alternative method of impact load calculation based on the premise that the methodology presently used is not necessarily conservative for all structural sizes and ranges of natural frequency. Although the method in NUREG-0487 seems overly conservative for structures in the larger size range and for those with higher natural frequency, it was found that most structures in the 2PS-1 pool swell zone are not in this range of size and frequency. An assessment showed that the original method in the DFFR (Reference 1) gives higher loads than the method in NUREG-0487. Therefore the design is adequate. NUREG-0487 also included impact load criteria for large structures (no dimension less than 20 inches) and gratings. The ZPS-1 pool swell zone contains no large structures and no grating. 5.3.2.3 Draq Load Calculations NUREG-0487 pointed out that under certain flow conditions the standard and acceleration drag coefficients used (Figure 5.3-26) may not give conservative results. This has been investigated and the loads have been reassessed using standard and
/~T acceleration drag coefficients. Relatively few calculations \J showed load increases. A detailed description of the methods used for reanalysis is included in Appendix G.
5.3.2.4 Chuggina Lateral Loads The original design of the 2PS-1 downcomers did not include bracing. The downcomers were rigidly supported by the drywell floor and were otherwise unrestrained. The natural frequency of the downcomers in this configuration is low. As described in Subsection 5.3.1.1.7, a lateral load of 8800 lbf was used in the design. Because of the low natural frequency of the vent (below 7 Hz), this load meets the requirements of NUREG-0487. A bracing system is now installed. This bracing is located slightly below the suppression pool surface. The bracing system design is discussed in Subsection 7.3.1 The natural frequency of the braced downcomers is above 7 Hz but below 14 Hz. These conditions meet the requirements of NUREG-0487 if the 8800 lbf load is increased by the ratio of the vent frequency to 7 Hz. This is the method being used for the reassessment of the chugging lateral loads. Because of the location of the bracing system in the pool, it must be capable of accommodating submerged structure loads as well as lateral loads. The bracing system has been designed for 7s ( ,) LOCA and SRV submerged structure drag loads as described in Subsections 5.2.2.4 and 5.3.1.3, incorporating the comments in NUREG-0487 as described in Appendix G. 5.3-21
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 O V 5.3.2.5 Condensation Oscillation Loads The condensation oscillation (CO) load definition originally used in the 2PS-1 design is the Mark II DFFR load definition. However, to account for uncertainties in the load definition and to expedite licensing, a more conservative load definition has been used for reassessment.. This method, called the Zimmer Empirical Approach, is described in Chapter 1.0. This approach is more conservative than required by NUREG-0487. O 4 () 5.3-22
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 i () 5.3.3 References
- 1. General Electric Company and Sargent & Lundy, " Mark II Containment-Dynamic Forcing Functions Information Report,"
NEDO-21061, September 1976 (Revision 2).
- 2. Final Safety Analysis Report, Wm. H. Zimmer Power Station, Chapter 6.0.
- 3. General Electric Company and Sargent & Lundy Engineers,
" Mark II Containment Dynamic Forcing Functions Information Report," NEDE-21061-P, September 1976 (Revision 2).
- 4. " Analytical Model for Liquid Jet Properties for Predicting Forces on Rigid submerged Structures," NEDE-21472, September 1977.
- 5. S. Abramovich and A. Solan, "The Initial Development of a Submerged Laminar Round Jet," Journal of Fluid Mechanics, Vol. 59, Part 4, pp. 791-801, 1978.
- 6. " Mark I Containment Program 1/4 Scale Test Report Loads on Submerged Structures Due to LOCA Air Bubbles and Water Jets,"
NEDE-23817-P, September 1978. O O - 3 5.3-23 1
- _ _ . . _ .__ .. _ . _ . . _ . _ . . _ , , , - . . . . . _ . . . ~ . . . , . _ . . . , _ , . _ _ , . , . . -
. 2 1
ZPS-1-MARK II DAR - AMENDMENT 13 l OCTOBER 1980 (j~T
\
TABLE 5.3-1 ACOUSTIC LOADING ON REACTOR PRESSURE VESSEL SilROUD TIME ACOUSTIC LOAD (msec) (kips) 0 0 1.2 0 1.6 250 2.0 320 2.5 650 2.8 250 3.0 100 3.2 0 O v 5.3-24
I to AMENDMENT 13 OCTOBER 1980 . O _ ' m x5
?
N m
^
_J - W 0 3 "
=
s _ W W l 3 __ _ _ 2
,1 _ 9 e
J o j _J O W v p - T Q - 1 o b 'a l 1 o o o o W W N
- (6!sd) 380SS38d i
WM. H.ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK II DES!GN ASSESSMENT REPORT O FIGURE 5.3-1 DRYWELL/WETWELL PRESSURE HISTORY FOR RECIRCULATION LINE BREAK (DBA) 1
.. _, .-y- , . , , . - - - , _ .
AMENDMENT 13 .
- OCTOBER 1980 l 1
else N Q S 3
~
O b
~
x Y L
- 4 W
b { f - o d w 3 3 m 3 O o
> g o Z m _ J 3
, o l
1 l l I I I I I l l l l 1 - o o o o o o o
@ t n (d.) 38niv83 din 31 WM. H.ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT O FIGURE 5.3-2 DRYWELL/WETWELL TEMPERATURE HISTORY FCR RECIRCULATION LINE BREAK (DBA)
AMENDMENT 13 M OCTOBER 1980 0 j M e N 1 J J - W O j 3 $ 3 H - W _ W 3 2 6 m O _ O J
~
O U J J - W 3 Cr O - O l' a e M i i i i l I i i i l I i i I - O O O O w t N (6!sd) 380SS3 bd WM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK ll DESIGN ASSESSM ENT REPORT O FIGURE 5.3-3 DRYWELL/WETWELL PRESSURE HISTORY FOR MAIN STEAMLINE BREAK
O O O 600 i E 400 e w - w _ DRYWELL
<t m _
N w Q. 2 - w
& L g 200 %
.I -
$ r 7 WETWELL _
r > ef ~ nP !I - 7 BC = 5 ~
-M sg Em 3
,o, z r o I
o h r Mr g " z y 0 I I I I I I I I I I I I ' ' I I Sm * ?" -l O I 2 3 iE 4 E"
=3 EE mE Y . mm LOG 10 [ TIME AFTER LOCA (sec)]
ea u E
- i M$ z
- > $D nz . e 02 m
H o O .z hk 2 i z $H 5 80
O O O 60 [. BEYOND THIS POINT THE TIME _ HISTORY IS THE SAME AS SHOWN ON FIGURE 3.3-18
- l-40 l
.?
m - c. v w l m _ DRYWE L L l ; m w - l E
.I E 20 / I i
gg r _ l "M $ [ir EP MC E
\ WETWELL l EM Rz - l 92 I e8 EP e SE -
M"o A > $ # I i i
- u, O 1 I I I
$A *
" 3 -l O I 2 3 RiO! Y E mE g "'
EE LOG 10 [ TIME AFTER LOCA (sec)] mm 03 Nbw
=3 85 a- o, e -
h
~ z
-4 8-OW
g AMENDMENT 13 OCTOBER 1980 k 1 N
~
. Q d w -
3 3 - to g
> w _ <
T O o 3 O _ _i
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O l I I I l l l ! I I I i T O O O O O O O w t N (_4o) 3Bfl1V83dW31 WM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT FIGURE 5.3-6 DRYWELL/WETWELL TEMPERATURE HISTORY FOR INTERMEDIATE SIZE BREAK (IBA)
(vaS) w 3ua livWS 80a A801SIH 380SS38d ll3M13M/T13MA80 z-E 9 380913 g 1 cod 38 1N 3 M S S 3 S S Y NDI S 30 11 wuyn l 111NO 'NOl1Y1S H3 mod HV37DnN H3WWlZ *H 'WM DRYWELL/WETWELL FREE SPACE PRESSURE (psig) _ N u A m o o o o o o o b, l l l l l l l l l Il l l l II IIIl IIII
=
=
b
=
=
C - 9 5 -
- U (
E ; 8_ : N NA N"
>e E _ 4 8
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- P I
_ s r 1 g u
/ _
086l 8380130 _ E El IN3 WON 3WV 3
O O O 400 ._.
- 300 - -
DRYWELLk w
\
cr
/
~
?
g 200 _ _ j p WETWELL- ( y
- F y r g l00 -- - - - -
? " i - ,c _ E= -
sm ; z g _ I I I IIIl! , gm 5 EE O I I I Illli I i l Illll I I I lllll I I I lllll I I I lilll I I I 11111 pH @
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10 3 10 4 10 5 106
%Q Ny p
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-- ; y es a vi Ok 5 = -s 39 4 93 Me
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3 80
AMENDMENT 13 OCTOBER 1980 0 N I l l l l n i O -
\\ d .
. \
I l l l t v a b Z Lu l > 1 NOT TO SCALE WM. H.ZIMMER NUCLEAR POWER STATION, UNIT 1 1 MARK ll DESIGN ASSESSMENT REPORT FIGURE 5.3-9 VENT CLEARING JET ANGLE 1 i
AMENDMENT 13 OCTOBER 1980 0
=
b W 60 - cc D Value at vent exit , W a-CL 40 - y Value at base mat, W - 2 O W e L 20 - fL 2 F-W o i i i I I i i i l l 0 0.2 0.4 0.6 0.8 1.0 TIME AFTER LOCA (sec) i l 1 WM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 18 DESIGN ASSESSMENT REPORT FIGURE 5.3-10 JET IMPINGEMENT LOAD t _.. _ ~ ,. -. . . - _ - _ _ _ . - _ . , -
E9oA w Raw;4so ; 0
~ 6
- - - - 3 0 - =
6 a,V 3 9 - -
- i s - -
0 " -
, D 4 -
g A 1 - i O ~ D s L # - A p - O C 2 L 0 I * - R M 4 T M - R - E " O / M M - F 8 M I
- 0 I Y - - 7 N 4 S ^ l U + A M - - 2
- - )
B
- - S A E
- E R
- G
- E
_ \ - - ( D O I A B
- - g N I - - - O I
/
A f - N 0 I T A C I 8 L
- - 1 P
- - P
- A F
y O .( V - - E L
- - G N
- A 0 -
8 1 y - l 0 9 A - - 1 o 0 0 A N- _ 7 2 9 - _A
- - - - 0 0 0 0 0 4 1
2 1 1
+ + - -
0 n*~E"we5mda Sr F eIIg t zc902,o5m2 $>dg cz 4 - - ((s r>" = g 52 >* mmmr9 2!;2-n E!m 5 7aa"- nE 8 =5 ro3o )
AMENDMENT 13 OCTOBER 1980 O g a s W e E e W J F P D z d W O V O P =(C x 24.C 3) PSI D D e x O o i l i lh i I i 1 , I o 0.2 o.4 a.s o.s i.o TIME AFTER LoCA (SEC) NOTE:
- 1. OBTAIN VALUES OF C FROM FIGURE 3.3-6.
D WM. H. ZlMMER NUCLEAR POWER STATION. UNIT 1 O
" * " " " ' ' ' " " ^ " ' ' ' " ' " ' " ' ' ' " '
FIGURE 5.3-12 DRAG LOAD DUE TO VENT CLEARING JET
I AMENDMENT 13 i OCTOBER 1980 O 80 1 -BEAM : 5"< widths 20" II n 50 - II
'm 3
o-l- BEAM: 0"< width s 5" 40 - lI w o-D m m w g 30 - s PIPE : 0" < 0. D. 6 2 0 " Q JI 4 II o-2 20 O 2 3 2 E 2 10 - 0 ' ' '
;j -
0 2 4 e 15 ELEVATION ABOVE INITIAL POOL SURFACE (ff) WM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT (3J FIGURE 5.3-13 MAXIMUM IMPACT PRESSURE ON SMALL STRIlCTURES
I l AMENDMENT 13 OCTOBER 1980 6 - PIPE O.D. , , 20* 5 - F 77 15* E g ii E S4 - 4 4 um a & C 10 3 - ff o lJ F-O 2 1 2 O = 2 0 11 x ei s 1 - II l' II O E . t . I II . l 0 2 4 6 15 ELEVATION ABOVE INITIAL POOL SURFACE (FT) WM. H.ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 18 DESIGN ASSESSMENT REPORT (D l kJ FIGURE 5.3-14 MAXIMUM IMPACT FORCE ON PIPES
8 i s I AMENDMNT 13 OCTOBER i3Pa 14
'O FACE WIDTH 20" ff 12 -
D 10 -
.9- 15" 5 rr Il J
N s e 8 - u. d
< 10 n
o J 2 6 - 17 F O o f _2_ ~ s 4 - a
- 2 X n
< 5 2 "
ll 2 _
,n ll 0 ' ' '
O 2' 4 6 15 ELEVATION ABOVE POOL SURFACE (ft) WM. H.IBMMFA NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT FIGURE 5.3-15 MAXIMUM IMPACT FORCE ON 1-BEAMS
AMENDMENT 13 l OCTOBER 1980 i O l.0 _ u U-u.E ui O u. 0.8 - m O _ x oE 0.6 - us x 3 m u
- a. O,4 _
Q N J - O sm g 0.2 O .0 l l I I O 2 4 6 8 TIME AFTER IMPACT (MILU-SEC) 1 i WM. H.I!MMER NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASSESSM ENT REPORT O rIGuaE s.3-is TIME HIST 0'Y OF IMPACT LOAD FOR SMALL STRUCTURES IN P0OL SWELL REGION
+,-OL 73A31 A11II8V808d IV Sd3WO3NM00 30 Sdn0BD NO SOV07 7VB31V7 Il-C'S 380913 $
1*0d3H 4 IN3 MSS 3SSV NDIS30 Il MWVM L llNn 'NOl1VIS H3 mod HV313ON H3WWlZ 'H 'WM FORCE PER DOWNCOMER (KIPS)
~g o - m u a e m I I I I I I C
E m . _ G 2 O O O - 4 z 8 e m $ - 8 - G 086l B380130 El IN3 HON 3WV _ R
O O O INITIAL INSTANTANEOUS RUPTURE ACOUSTIC LOADING ANNULUS PRESSURIZATION j E F l-z Z JET STREAM RELEASE r - w
> tf >
!j w l
=9 o ,-
z I d m Eg PIPE WHIP RESTRAINT LOADING g i?! e EE m M >$ E m *g 5Y R 5 E9 E i5 8E 1
= 5 -
TIME + 9z NE gc 95
- z :- -
.=
g. . 2 " __________________________d _ _ _ _ _ _ _ _ _
AMENDMENT 13 OCTOBER 1980
.c- SHROUD O SuegoRT P P RECIRC.
2 PIPE NOZZLE SHROUD WALL
< VESSEL WALL i
P -OPERATION PRESSURE ~ 1000 PSI P2 -SATURATION PRESSURE ~ 900 PSI
<--VESSEL WALL Q SHROUD WALL RECIRC. PIPE N0ZZLE
@ .r ,
SHRGUD SUPPORT VESSEL CUTAWAY WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK ll DESIGN ASSESSMENT REPORT , FIGURE 5.3-19
, AC0USTIC LOAD ILLUSTRATION l
AMENDMENT 13 OCTOBER 1980 CALCULATION OF FORCE I 4 l
- x. \
/
,<='.-. . . -
/ SHIELD
.- . g /
/
l~:.', *.- j l RPV
' 7}' ' , j .- ~
l
. /
4 Q *' l SHIEL D-* . . .* w / n Y -* I+-RPV
/
;'p -*/
..u e- /
L '
' . " ..-- I
. -- ; f, V '* " /
-,/ l A. PRESSURE DISTRIBUTION 1 ,;L,, n /
! b@
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,r
_ f/ SHIELD Q
.. .-o./
V [E,.! ^@@, j RPV
.,~ Fsi ; ->. Fyl l
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j;;. ,, ,J.*-PEDESTAL B. RESULTANT FORCES
, e :r;. . .
\ ., ' . ; :.' r. . .
- <...- FORCE DESCRIPTION (ALL FUNCTIONS OF TIME)
M .-
!. PRESSURE LOADS
- 2. PIPE RESTRAINT LOAD I
- 3. JET REACTION FORCE
- 4. JET IMPINGEMENT FORCE i
7//.///n WM. H.ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK la D E S IG N ASSESSMENT REPORT O V FIGURE 5.3-20 LOADING DESCRIPTION
l AMENDMENT 13 OCTOBER 1980 THE PIPE WHIP RESTRAINT RESTRICTS PIPE MOTION AND ABSORBS l BLOWDOWN ENERGY, THE RUPTURED PIPE TRAVELS FREELY THROUGH THE CLEARANCE GAP AND THE RESTRAINT DEFORMS TO ABSORB THE KINETIC ENERGY.
/ /
/ .
/
/ l
/ , -PIPE /
~m / ' \
/
/ l
/ SHIELD /
CLEARANCE GAP
/ YIELDED PIPE WHIP RESTRAINT PIPE WHIP RESTRAINT AND PIPE AND PIPE AT REST AFTER IN NORMAL OPERATION POSITION POSTULATED RUPTURE SECTION VIEW (AA)
O MASS / ENERGY FLOW PATH] [ / mr 1 WELD : LOCATION I
/Q e r
////
A
/ /-
FRONT VIEW WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK 11 DESIGN ASSESSM ENT REPORT O riGuRe s.3-zi RESTRICTED PIPE MOTION DURING BLOWDOWN l
O O O 27 - l~
, 24
't 21 ZIMMER BOUNDING FLOW un O 18 - x o w Q en 15 - N -- m ' (RELAP: GRADUAL BREAK OPENING, J HENRY-FAUSKE-M0ODY CHOKED-FLOW) m o
- E J 12 I h-it Is m
!3
==
=
9 - g a m o z C p 3 QC ae s= 5 z P c s -
=
. :=
i au :8 5 b 1 .
! 3 -
4 m 9:
~
sg "5 t 0 ! ! ! l l I i 1 1 I 8= 20 50 60 80 90 m3 o 3 C
-2 O 10 30 40 70 10 0 mg
-4 z TIME M-SEC g"-4 1 -4 m
. ** OW
O O O 21000 - ZIMMER B0UNDING FLOW 18000 - 15000.-
"r Ls.
ai2OOO - u 3 g 9000 -
? /
, z _
6000 - P (INSTANTANEOUS BREAK; HENRY-m e,
~$
m
/ FAUSKE-MOODY CH0KED-FLOW) hs M
"'~
EP
$ 3000 -
Ed u, m 75
- ~ mm
" I m E 2 O 1
$ z $ 0 0.1 02 03 0.4 05 2 g TIME (SECONDS) gk g
, 4 m o eg g 3z gi5i 5E -5 3 $C
AMENDMENT 13 0CTOBER 1980 O e MSly - X STE AM e , i , M2,43 ( ) BREAK Qe 35
+ +
@ J @ L @
035
- BREAK NODE 4
o ' 4 I
- FEEDWATER 36 I 9 9 l
I 36sDRYWELL @ e, g ' 4s '@ 't 'i'O i 6( ,
- 5 tr, g '
=- i
'04 7 ' g er e 34
't o ' t O 'a r 'ie BREAK O 01 'd d 4 [
8 h ..
-+30 I
3g l l 44 e.
- 1r7 48 10 5
8 j @ SO St 4E Q2
$ _.S*
2: - WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 M ARK li DESIGN ASSESSMENT REPORT FIGURE 5.3-24 RECIRCULATION LINE BREAK N0DALIZATION
kyg 0ESE - l _ p'v V P R 3 J ( 9 3 4 4 4 4
\e E &
4I . 81 < 0 I l 2 l 1
. 71 4 m ;
S S R R E E G R A
@ G R
A P P 7 S S
. l 6
3 O T 9 i 1 8 1 s 3 l l l 8 21 L e
. l) i O s i
f' 41j ' 6 1
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?
3 @ 9 3 (- - 4,
@ 2 4
gE5=!m$1$ $35ZEa
@ @ i ;!z >$E 9 " o" nu s%" 'P'Y:
mG E CM 28Esd9
f AMENDMENT 13 OCTOBER 1980 Body Shape Cn O circuier criinoer gF,trortinn Fio. Elliptical Cylinder 02.i o.6-o.* i 4.1 0 .32
' ~
08' i Square [] 2.o 0 Triangle A 120 2.0 Y 1.72 i"120 U
% 90 2.15 90 1.6o o
p0 2.20 t 40 1 3' r 30 i.. t l k(30 t 0 1.0 Semitubular Q ,,, y 1.12 a i l WM. H.ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK ll DESIGN ASSESSMENT REPORT s.J FIGURE 5.3-26 DRAG C0EFFICICflTS
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 5.4 ZIMMER POSITION ON NRC LEAD PLANT ACCEPTANCE I CRITERIA (NUREG-0487) : The following Table 5.4-1 summarizes the extent to which the ' design of the Wm. H. Zimmer Power Station conforms to the NRC Lead Plant Acceptance Criteria. This table demonstrates that all ' NRC concerns have been addressed in the 2PS-1 design i reassessment., and that the loads used in the reassessment meet or i exceed the NRC Lead Plant Acceptance Criteria. O
/
3.4-1
I b/ - V TABLE $.h-1 ZIMMER POSITION ON NRC 1.EAD FI. ANT ACCEPTANCE CRITFRIA (NUREC-0487) MARK II OWNERS CROUP LOAD OR PHENOMENON LOAD SPECIFICATION NRC REVIEW STATUS ZIMMER POSITION ON ACCEPTANCE CRITERIA s
- 1. 14CA-Related Hydrodynamic Loads A. Subs'erged Boundary Loads 33 psi over-pressure added to local Acceptable Acceptable.
During Vent Clearing hydr ostatic below vent exit (walls and basemat) - linear attenuation . to pool n ar f ace. The Mark II program has provided a realistic assessment of wall loads during vent clearing based on 4T results. Re ference General Electric letter MTN-080-79 from Mr. Sobon to Mr. Stolz, dated March 20, 1979. B. Pool Swell Loads
- 1. Pool Swell Analytical Mode l tn
. a) Air Bubble Pressure Calculated by the Pool Swell Anal- Acceptable Y 4 ytical Model (PSAM) used in cal- 7 culation of submerged boundary loads.
b) Pool Swell Elevation 1.5 x submergence. NRC Criteria 1.A.1 Acceptable es c) Pool Swell Velocit'y Velocity history vs. pool eleva- NRC Criteria 1.A.2 Acceptable tion predicted by the PSAM used to compute impact loading on small The impact of a 10% increase in pool swell velocity will structures and drag on gratings be assessed. Although the sssumptions used in the Pool between initial pool surface and Swell Analytical Model are already very conservative maximum pool elevation and steady- and eliminate the need for any additional factors, the state drag between vent exit and resulting calculated load increase will not require maximum pool elevation. Ana l- design changes. ytical velocity variation used up to maximum velocity. Maximum velocity applies thereafter up to maximum pool swell. g d) Pool Swell Acceleration predicted by the PSAM. Acceptable a, Acceleration Pool acceleration is utilized in 9 c., the calculation of a:celeration drag y$ loads on submerged components $w O" during pool swell. e) Wetwell Air Wetwell air compression is cal- Acceptable Compression culated by the PSAM. Defines the pressure loading on the wetwell boundary above the pool surface during pool swell.
._ _ .- .- _ __ m____ ~ __ ..m._ . _ _ _ _ _ .__ _ -_ _ . . .
. O -O l TABLE 5.I+-l (Cont'd) MARK II OWNERS CROUP LOAD OR P2ENOMENON LOAD SPECIFICATION NRC REVIEW STATUS ZIMMER POSITION ON ATEPTANCE CRITERIA 1 f) Drywell Pressure Plant unique. Utilized to PSAM Acceptable if based Acceptable. History to calculate pool swell loads. on NEDH-10320. Other-vise plant unique
- reviews required.
- 2. Loads on Subserged Maximum bubble pressure predicted Acceptable Boundaries by the PSAM added uniformly to .
4 local hydrostatic below vent exit (wells and basemat) linear attenua-tion to pool surface. Applied to walls up to maximum pool swell elevation.
'3. Impact loads NRC criteria I.A.6 Acceptable. Although the criteria is unnecesarily con-servative investigations indicate that, due to the size a) Small Stnctures 1.5 x Pressure-Veloca,ty correla- and f requency of structures in the Zimmer pool swell tion for pipes,and I beams. zone, the design loads used are conservative with respect 7 Constarit duratton pulee, to the NRC Acceptance Criteria. It should be noted that e.
- f. analytical work performed by Sargent & Lundy utilizing 3 i W the PSTF (Pressure Suppression Test Facility) data for /
circumferential targets indicates that the DFFR spe- ' cification is conservative for the size and frequency of structures in the Zinsser Pool Swell Zone. Tests
; performed by EPRI (EPRI No. NP-798, May 1978) to deter- O t
mine flat pool impact on rigid and flexible cylinders e, are also in good agreement with DFFR. The NRC Acceptance $ Criteria utilized an additional assumption (I-beam impact duration is inversely proportional to velecity) which ta inconsis*ent with theory and experimental evidence. Nevertheless the NRC Criteria have been used to assess structures in the pool swell zone and these structures can withstand the conservative criteria. b) Large Struchree None - Plant unique load where Plant unique review Acceptable. Zinuner has no large structures in the pool j applicable. where applicable swell zone. i c) Grating , No impact load specified. P NRC Criteria I.A.3 Acceptable. Zinuser has no grating in pool swell area. l vs.openareacorrelationanh*8 h' @* velocity vs. elevation history
- i from the PSAM. N}
E 80 M
'~
O O TABLE 5.1+-1 (Cont 'd) MARK II OWNERS CROUP LOAD OR PHENOMENON LOAD SPECIFICATION NRC REVIEW STATUS ZIMMER POSITION ON ACCEPTANCE CRITERIA
- 4. Wetwell Air Compression a) Wall Loads Direct application of the PSAM Acceptable calculated pressure due to wet-well compression.
b) Diaphragm Upward 2.5 psid NRC Criteria I.A.4 Acceptable Loads . 5.
~
Asymetric Load None NRC Criteria I.A.5 Acceptable. .. Zimer was assessed for an asymetric load of the vent clearing pressure (22 psig) applied over a 180' sector of the wetwell wall. This load was applied from the basemat to the wet-well floor. The pool hydrostatic load (12 psig at the basemat with linear decrease to 0 at 4 water surface) was superimposed on the asy m etric load. The Zimer design can accommodate this conservative load. General Electric has pro-vided an analysis showing that this asymetric e. load will actually be less than 10% of the U maximum vent clearing pressure (CE Letter s' MFN-076-79, March 16, 1979). Based on a sub- g' e sequent analysis by Brookhaven National Labora- E
- tory, the NRC revised this criteria to 20% of the maximum vent clearing pressure (NRC/MK II Owners U meeting, July 24-25, 1979, Bethesda, Maryland). o Zim er has been assessed for an asymetric load $
5 times greater than the present criteria. C. Steam Condensation and Chugging Loads
- 1. Downcomer Lateral Loads 1
a) Single Vent Loads 8.8 kip static NRC Criteria I.B.1 Acceptable 8$ b) Multiple Vent Loads Prescribes variation of load NRC Criteria I.B.2 Acceptable 8 per downcomer vs. number of downcomers. N O
- 2. Submerged Boundary 80 Loads a) High Steam Flux Sinusoidal pressure fluctuation Acceptable Loads added to local hydrostatic.
Amplitude unifom below vent exit-linear attenuation to pool surface. 4.4 psi peak-to-peak amplitude. 2, 6, 7 Ez frequencies.
. _ _ _ . _ . _ . . . _._ __ _ __ ~ _ __ . _ _ . _ . __ . _ _ _ _. _
( TABLE 5.1.-l (Cont'd) MARK II OWNERS CROUP LOAD OR PHENOMENON IDAD SPECIPICATION NRC REVIEW STATUS ZIMMER POSITION ON ACCEPTANCE CRITERIA b) Medium Steam Flux Sinusoidal pressure fluctuation Acceptable Loads added to local hydrostatic. An-plitude uniform below vent exit-linear attenuation to pool surface. 7.5 psi peak-to-peak amplitude. 5, 6 Hz frequencies. c) Chugging Loads Representative pressure flue- Acceptable pending tuation taken from 4T test resolution of FSI added to local hydrostatic. concerns.
- uniform loading Maximum amplitude uniform below condition vent exit-linear attenuation to pool surf ace. +4.8 pai maximum overpressure, -4.0 psi maximum under pressure, 20-30 Hz frequency.
e y, - asyurnetric loading Maximum amplitude uniform below condition vent exit-linear attenuation to [ e i pool surface. 20 psi maximum v' E overpressure, -14 psi maximum a derpressure, 20-30 Hz fre-N quency, peripheral variation of
- amplitude follows observed statistical distribution with "
maximum and minimum dia-
- metrically opposed.
i g 33 8t:
.== -. . _ - . . - .-- - - -- - .--- ..-.. --. - - - . _ ..
L O O O t a TABLE 5.h-1 (Cont'd) MAPI II OWNERS GROUP I4AD OR THENOMENON LOAD SPECIFTCATTON NRC REVITd STATUS ZIMMER POSTTION ON ACCEPTANCE CRITERIA II. SRV-Pelated Hydrodynamie Loads A. Pool Temperature Limits fbr KWU None specified NRC Criteria II.1 Acceptable and CE four-am quencher and II.3 Quencher Air Cleaning Icsos The Susquehanna LAR is used as a NRC Criterin II.2 The Zi:ener station is being assessed for the T-quencher basis for the T-quencher loads, loads. These loads are considered to be conservative Mark II plants utilizing the four- and demonstrate the adequacy of the Zimmer design. A arm quencher use quench load meth- presentation on the impact of modifications to the SRV odology described in DFFR. frequency range was given in the Tebmary 13,19T9 meeting. Results of an assessment of the SRV T-quencher frequency range was presented at the July 26, 1979 meeting. A further demonstration of the conservatism of the lead plant approach has been documented by Long Island n Lighting Co. (SNRC-374, Nrch 30,1979, Mr. Novarro @ [LILCO] to Mr. S. A. Varga [NRC) transmitting a report / w entitled " Justification of Mark II Lead Plant SRV Load g y Definition. ") ,g h In-plant tes'.s will be run to demonstrate the adequacy M and conservatism of the design loads. o B. Quenchu Tie-Down Loads . a
- l. Quencher Am Loads (a) Four-Am Quencher Vertical and lateral am loads Acceptable developed on the basis of bound-ing assurptions for air / water discharge fmm the quencher and conservative combinations of maximum / minimum bubble pressure acting on the quencher.
4
- 5 a 80 l,
_ __-__- ____ - g
i O O
)
i TABLE 5.41 (Cont'd) MARK II DWNERS GROUP l LOAD OR PHENOMENON LOAD SPECIFICATION NRC REVIEW STATUS ZIMKER POSITION ON ACCEPTANCE CRITERIA i (b) EWU T-Quencher KWU "T"-quencher not included in Review Cc *inuing Acceptable. These loads have been cal mlated usin the Mar k II O.C. Program. T-quencher methodology and assumptions described 1 ; DFFR for our-arm ann loads not specified at this quenchers, as recommended in the Accepta sce Criteria. KWU time. T-quencher methods were used to verify conservatism.
- 2. Quencher Tie-Down Loads (a) Four-Arm Qvencher Includes vertical and lateral Acceptable arm load transmitted to she base-mat via the tie-downs. See II.C.I.a above plus vertical transient wave and thrust loads.
Thrust load calculated using a-standard momentum balance. Ver-tical and lateral moments for air or water clearing are cal-culated based on conservative l l Ln e. clearing assumptions. } KWU "T". quencher not included Review Continuing Acceptable. These loads have been calculated using the h' l
$ (b) KWU "T". Quencher in Mark II O.C. program. T. methodology and assumptions described in DFFR for four- %
l ! quencher tie-down loads not are quenchers, as recommended in the Acceptance Criteria. KWU p specified at this time. T-quencher methods were used to verify conservatism. m III. LOCA/SRV Submerged Structure Loads g A. LOCA/SRV Jet Loads l
- 1. LOCA/ Rams Head SRV Methodology based on a quasi-one- NRC Criteria III.A.1 Rams head - N/A. For LOCA jet see Subsection 5.3.2.1.
Jet Loads dimensional model. .
- 2. SRV-Quencher Jet Loads No loads specified for lead plants. NRC Criteria III.A.2 This item is resolved fwr Zimmer. The spherical zone of Model under development in long- influence defined in the Acceptance Criteria is not ap-i term program. propriate for the two-ara quencher. A sone of influence l
for each arm will be defined as a cylinder with an axis $ 1 coincidental with the quencher arm. The length of the g D cylinder will be equal to the length of the quencher arm plus 10 end cap hole diameters. The radius of the cylinder, N' N is expected to be quite small. However, because no structures $;N are within 5 feet of the quencher arm, 5 f eet will be assumed. gg ll t Since no structures are loca,ted within 5 f eet of the quencher, the intent of the NRC Criterion III.i 2 is now satisfied. I l l
s s e"
%.s k's J
0 TABLE 5.4-2 (Cont'd) MARK II OWNERS CROLT LOAD OR PHENOMENON LOAD SPECIFICATION NRC REVIEW STA5fS ZIMMER POSITION ON ACCEPTANCE CRITERIA B. LOCA/SRV Air Bubble Drag Loads The NRC Acceptance Criteria required modification to the
- 1. LOCA Air Bubble Ioads The methodology follows the LOCA NRC Criteria III.B.1. present methodology in several areas. The Lead Plants have air carryover phase from bubble addressed these concerns as explained below. Generic docu-charging, bubble contact, pool mentation will be provided in a Mark II Owners Group submittal, rise and pool fallback. The For Zimmer, these items have been addressed as follows:
drag calculations include standard , and acceleration drag components. a) Bubble Asymmetry - The NRC A ceptance Criteria recommends a 10% increase in velocity and acceleration to accommodate potential asynnetries in the LOCA and SRV air bubbles. Zimmer feels that this added factor is unnecessary given the available data and the conserystism already included in the calculations. In support of this position it is best to discuss LOCA and SRV air bubbles separately. The LOCA charging air bubble transient is driven by the pressurization of the drywell. The limiting case is a double-ended guillotine break of a recirculation line. y In spite of the low probability of this type of break, @ I"
- additional conservatisms are included in the calculation $
f of the mass-energy release rate. Af ter the bubble begins ! ca to form, observations from tests indicate no significant E assymetries when the bubble expands (EPRI report NP-441 N April 1977). Drag coef ficients are also conservatively [ determined as outlined in Appendix H. Loads measured in tests show that LOCA air bubble loads are greatly h overpredicted by analytical models. The SRV air bubble loads have been demonstrated to be very conservative by the Caorso in-plant tests (CE, letter MFN-090-79 to Mr. J. F. Stolz March 28, 1979) and the KTG single cell tests in Germany (as reported by Pennsylvania Power & Light). In addition, multiple valve phasing has been assessed for very conservative unrealistic conditions. Finally, as noted for the LOCA air bubble, drag coeffi-cients have been determined conservatively. In view of the conservatisms identified here and the con- h servatism of the current methods when compared to test Si g data, no additional multipliers are being applied to 9m velocities and accelerations to simulate postulated bubble $ assynnetries. jg b) Standard Drag in Accelerating Flows - addressed in Appendix G. c) Noda11 ration of Structures - addressed in Appendix C. d) Interference Effects - addressed in Appendix C.
O O O TABLE 5.1+-1 (Cont'd) MARK II OWNERS CROUP IDAD OR PRENOMEN0N IDAD SPECIFICATION NRC REVIEW STATUS ZIMER POSITION ON ACCEPTANCE CRITERIA e) Addressed in Appendix C.
- 2. SRV-Rams Head Air he methodology is based on an NRC Criteria 111.B.2 Bubble leads analytical sedel of the bubble ,
charging process including bubble a) Neglecting Standard Drag - Standard drag is calculated rise and oscillation. Accelera- and included for all submerged structure load calculations. tion drag alone is considered. b) LOCA Bubble Criteria - addressed in Appendix C.
- 3. SRV-Quencher Air
- No quencher drag model provided for NRC Criteria III.B.3. Open Ites Bubble Loads lead plants. Imad plants propose 2
interim use of rama head model (See ne bubile location and radius will be defined appropriately 111.B.2 above). Model will be for T-quenchers. Bubbles are located near the arms. The developed in long-term program. bubble size la predicted from the line air volume. C. Steam Condensation Drag No generic load methodology Lead plant load spe- Described in Subsections 5.3.1.3.5 and 5.3.1.3.6 N u loads provided. Generic model under cification and NRC 3
- development in long-term program. review will be con- J.
E ducted on a plant ' unique basis with confirmation in long-term program ll using generic model. 8 a. 0:3 8t:
f) O t N 5' N %Y TABLE 5.4-1 (Cont'd) MARK II OWNERS CROUP IDAD OR PHENOMENON IDAD SPECIFICATION NRC REVIEW STATUS ZIMMER POSITION ON ACCEPTANCE CRITERIA IV. Secondary Loads A. Sonic Wave Load Negligible Load - none specified Acceptable B. Compressive Wave Load Negligible Load - none specified Acceptable C. Post Swell Wave load No generic load provided Plant unique load Described in Zimmer Closure Report specification and NRC review. D. Seismic Slosh load No generic load provided Plant unique load Described in Zinumer Cissure Report specification and NRC review. E. Fallback load on Submerged Negligible load - none specified Acceptable Boundary y
- Y a F. Thrust Loads Momentum balance Acceptable ';*
E E o C. Friction Drag Loads Standard friction drag calculations Acceptable g on Vents . H. Vent Clearing Loads Negligible Load - none specified Acceptable al a A st:
(
\ (w/) (/
s
)
TABLE 5.1.-1 (Cent'd) MARK II OWNERS CROUP LOAD OR PHENOMENON IDAD SPECIFICATION NRC REVIEW STATUS ZIMMER POSITION ON ACCEPTANCE CRITERIA FUNCTIONAL Interim technical Acceptable, Rodabaugh criteria may be used in some cases if CAPABILITY position (7/19/78) NRC finds acceptable. 4 MASS-ENERGY Vergfy using REIAP / Acceptable RELEASE FOR M)D ANNULUS PRESS. QUESTIONS 15% peak broadening Acceptable MEB-2, HEB-5 to be used. MEB-3, MEB-5 Closely spaced modes Acceptable. NSSS scope uses modified suunnation combined Per 1.92 per approved CESSAR. MEB-1 Dynamic analysis Acceptable methods acceptable u E
," MEB-2 OBE Damping - Level ,L S Damping - Level ccePtable M C or D Y
MEB-6 Seismic slosh-plant Acceptable e, unique review j', MEB-7s and b load Combinations: Acceptable. See load combination table for Case #2 and #7. AP+SSE OBE+SRV HEB-8 Functional capability See load combination table. and piping acceptance criteria a. 3G Il: E s= s
n
\J O/ (3
%J TAE*.E 5.4 1 (Cont'd)
MARK II OWNERS CROUP LOAD PHENOMENON LOAD SPECIFICATION NRC REVIEW STATUS ZIMMER POSITION ON ACCEPTANCE CRITERIA
- 1. N+SRVx T B Acceptable
- 2. N+SRV +0BE to B Acceptable Approved CESSAR approach used for NSSS.
- 3. N+SRVg +SSE to C Acceptable .
- 4. N+SRV +0BE+IBA to C Acceptable ads
- 5. N+SRV +0BE+IBA to C Acceptable ada
- 6. N+SRV +SSE+1BA to C Acceptable 4
- 7. N+SSE+DBA to C Acceptable
- 8. N to A Acceptable ce
- 9. N+0BE to 8 Acceptable 3 7
>=
- 10. N+SRV +SSE+DBA to C Applied to containment structure only (See M 020.22 and M DFFR 5.2.4.)
, E E.
5 E g$ M
- i. cm g
b 8::
i i 2PS-1-MARK II DAR AMENDMENT 13
- OCTOBER 1980 jO i
CHAPTER 6.0 - LOAD COMBINATIONS CONSIDERED 6.1 CONTAINMENT AND INTERNAL CONCRETE STRUCTURES i-The containment and internal. concrete structures were assessed . for the load combinations presented in Table 6.1-1. These r
- combinations include the' forces due to pool dynamic loads as
- reported in DFFR Table 5-2.
i The loading categories considered in Table 6.1-1 are consistent i with those specified in Article CC-3000 of the ASME B&PV Code, , Section III, Division 2 and include the following categories: ,
- a. normal operating loads with and without thermal '
loads, i l
! I
- b. normal operating loads with severe environmental ,
loads, ,
- c. normal operating loads with extrem? environmental 8 loads, i l 2
i
- d. abnormal loads, ,
- e. abnormal loads with severe environmental loads,'and '
' Os-
- f. abnormal loads with extreme environmental loads. i The time sequence of occurrence of the LOCA and SRV transient j loads presented in the DFFR Figures.5-1 through 5-6 were use in l
determining these design load combinations wich pool dynamic ' 4 loads. I a - 1 All combinations of SRV and LOCA loads with other design loads , required by the NRC have been' considered, including the hypothetical combination of LOCA with one SRV valve. actuation.- ' i I
. As shown in Table 6.1-1, the various modes of SRV actuation I
- considered are
- i ;
SRV-ALL 1. Symmetricactuationofallvalvesl
- 2. Resonant sequential symmetric '
actuation of all valves i i SRV-ADS Automatic depressurization actua- , tion of six valves. ,
- SRV-Asymmetric Actuation of two adjacent valves
(rams head). i Actuation of three adjacent valves i i- (T-quencher). , 6.1-1 i 4 F+ v c- --v - v ,rw,W , ,y> e--, - -yy -r y r- 9-+rn-*--,e>s----4--ra g4 w+x. -.-g a > -e * % e ,3--,,,-w t.ye er , . - m - m at re' e ww--p r
l i
)
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 SRV-Single Actuation of one valve. I The LOCA loads are denoted by PA and Pb in the load combination l table and represent three possible pipe break accidents: ,
- a. DBA - design-basis large break accident '
- b. IBA - intermediate break accident i
- c. SBA - small break accident. ,
Wherever applicable the following loads associated with LOCA are ' included whenever P3 or PB ccur in the load combinations: 1 I
- a. LOCA pressure i
- b. accident temperature
- c. pipe break reactions '
i
- d. vent clearing and pool swell i
- e. condensation-oscillation '
I s f. chugging. i
.] Even though the SRV and LOCA loads used for design are bounding loads as discussed in Subsection 5.2.1.3, additional load factors are applied to these loads (see load combination in Table 6.1-1) '
to assure conservatism. 8 The load factors adopted are based upon the degree of certainty and probability of occurrence for the individual loads as discussed in the DFFR. The relation between the different times of occurrence of various time-dependent loads as presented in the DFFR were combined and accounted for to determine the most critical loading conditions. In any load combination, if the effect of any load other than dead load (such as thermal loads) ' reduces the net design forces, it is deleted from the combination i to maximize the design loads. i i The reversible nature of the structural responses due to the pool , dyanmic loads and seismic loads is accounted for by considering fer each the peak positive and negative magnitudes of the ' response forces and maximizing the total positive and negative i forces and moments governing the design. i i Seismic and pool dynamic load affects are combined by summing the , peak responses of each loao by the ABS method with the exception of AP + SSE case where SRSS method is used. This is ' rx conservative, and the SRSS method is more appropriate, since the 8 (_) peak responses of all loads do not occur simultaneously. i However, the conservative ABS method is used in the design i 6.1-2
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 - O assessment of the containment and internal concrete structure to i ' expedite licensing review. , Reevaluation of the containment structure for these loads is described in Section 7.1. ' 1 Acceptance Criteria i i The acceptance criteria used in the reassessment of the ' containment and internal concrete structures for these additional loads and load combinations are the same as those reported in ' 2PS-1 FSAR Subsection 3.8.1. These criteria meet or exceed the i requirements of Article CC-3000 of the ASME B&PV code, Section i III, Division 2. For example, the ASME Code permits yielding of , the reinforcing steel when thermal loads are present, whereas the ' criteria used in the reassessment do not permit yielding of the reinforcing steel. ' O o)
\_
6.1-3
O O O TABLE 6.1-1 DESIGN LOAD COMBINATIONS
- LOAD ASTWET-EQN COND D L PO Pg Tg {g {g Egg {3 {g Tg {g {g SRV ADS ALL RICAL SINCLE No rmal I w/o Temp 1.4 1.7 1.0 1.0 - - - - - - - - -
1.5 0 X X 2 Normal w/ Temp 1.0 1.3 1.0 1.0 1.0 1.0 - - - - - - - 1.3 0 X X 3 No rmal Sev. Env. 1.0 1.0 1.0 1.0 1.0 1.0 1.25 - - - - - - 1.25 0 X X 4 Abnormal 1.0 1.0 1.0 - - - - - 1.25 - 1.0 1.0 - 1.25 X 0 X 4e 1.0 1.0 1.0 - - - - - - 1.25 1.0 1.0 - 1.0 0 0 0 X 5 Abnormal Sev. Env. 1.0 1.0 1.0 - - - 1.1 - 1.1 - 1.0 1.0 - 1.1 X 0 X Sa 1. 0 1.0 1.0 - - -
.1 - -
1.1 1.0 1.0 - 1.0 0 0 0 I ,, m O 6 Normal d. E. Ext. Env. 1.0 1.0 1.0 1.0 1.0 1.0 - 1.0 - - - - - 1.1 0 X X s
- a. I'~
7 Ab normal
- Ext. Env. 1.0 1.0 1.0 - - - -
1.0 1.0 - 1.0 1.0 1.0 1.0 X 0 X ;* 7a 1.0 1.0 1.0 - - - - 1.0 - 1.0 1.0 1.0 1.0 1.0 0 0 0 X E
=
LOAL DESCRIPTION D = Dead Loads Ess
= Safe Shutdown Earthquake L = Live Loads PB
= SBA and IRA Pressure Load 8i a
o F = Prestressing Loads T = Pipe Break Temperature Lead E 3 m T = Operating Temperature Loads RA = Pipe Break Temperature Reactions $ O
=
RO Operating Pipe Reactions [ PA = DBA Pressure Loads (including all
= Operating Pressure Loads Po pool hydrodynamic loadings)
SRV = Safety / Relief Valve Loads RR = Reactions and Jet Forces Due to E = Operating Basis Earthquake 11p. Break O , SEA = Small Break Accident IBA = Intermediate Break Accident
*In any load combinations, if the ef fect of any load other than D reduces *the design forces, it will be deleted from the combination.
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 19Rn fs b 6.2 CONTAINMENT LINER , The load combinations used for the design of the suppression pool ' liner and its anchorage system are the same as given in Table i 6.1-1 except that all load factors are equal to unity. The use i of unity load factors is in accordance with Article CC-3000 of , the ASME B&PV Code, Section III, Division 2. , Acceptance Criteria 8 i For self-limiting thermal loads, the strains in the liner plate i and the displacements of the liner anchorage system are limited , to the values allowed in Article CC-3000 of the ASME B&PV Code, Section III, Division 2. i i For mechanical loads such as any net negative pressure from pool i dynamic load, the liner stresses are limited to the values i specified in Subsection NE-3211 of the ASME B&PV Code, Section , III, Division 1. I Fatigue evaluation of the liner is based on Subsection NE-3222-4 I of the ASME B&PV Code, Section III, Division 1. i O 4 O l l l 6.2-1 l
. . _ _ , - _. _ _ _ . , _ .-- . . - ~ . , . _ -
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 s (~~'I k- 6.3 OTHER STRUCTURAL COMPONENTS 6.3.1 Load Combinations The load combinations, including pool dyncmic loads considered in the reassessment of concrete structures (other than containment and internal concrete structures) such as shear walls, slabs, beams and block walls are shown in Table 6.3-1. The load combinations, including pool dynamic loads considered in the reassessment of steel structures such as framing, containment galleries, embedments, hangers for cable trays, conduits, and ducts are listed in Table 6.3-2. For concrete structures, the peak effects resulting from seismic and pool dynamic loads were combined by the conservative ABS method, even though the SRSS method is more appcopriate, since the probability of all peak effects occurring at the same time is very small. Likewise, for steel structures, the peak effects resulting from seismic and pool dynamic loads were combined by the ABS method. 6.3.2 Acceptance Criteria g (,) The acceptance criteria used in the reassessment of reinforced concrete structures other than containment and internal concrete structures are the same criteria defined in Subsection 3.8.4.5 of the ZPS-1 FSAR and are identified in Table 6.3-1 for each load combination. The stresses and strains are limited to those specified in ACI 318-1971. As indicated in Table 6.3-1, working stress design is used for load combinations 2 through 6. The ultimate strength design of ACI 318-1971 is used for extreme environmental category load combinations 7, 8, and 9. As stated in the FSAR, when a LOCA occurs outside the containment, as in load combinations 10, 11, and 12, yield line theory is used to design reinforced concrete walls and slabs. Block walls are designed according to the specification for the design and construction of load-bearing concrete masonry by the National Concrete Masonry Association. For steel structures, stress and strains in accordance with the 1969 AISC specifications are used for load combinations 2 through 6 defined in Table 6.3-2. For load combinations involving abnormal or extreme environmental loads, as in load combiantions 7 through 12 of Table 6.3-2, the steel stresses were conservatively limited to 0.95 F y. O \> 6.3-1
.~ - _ _ .... _ - =-. m m-__. - . - - _ . .- .
s- __ - - - _ . -
.{
TABLE 6.3-1 i IAAD DEFINITIONS AND Q)MBINATIONS FOR REINFORCED CONCRETE ' (STRUCTURES OTHER THAN Q)NTAINMENT) l IDADING IAAD FACTORS ANCE CATECORY MURIA
~
R T E R T P P R E W DESCRIPTION NO. D L o o o W H r a a b a se e 'H'- SRV* ADS ALL METRICAL SINCIZ SM ESS s Construction 1 - 1.0 1.0 .- -- - 1.0 . - - - - -- - - - - 0 0 0 0 1.33 ACI ' 318-63 WSD -' Test 2 1.0 1.0 - 1.0 - - - - - - - - - - - - 0 0 0- 0 ACI 318-63 WSD s i Normal. 3 1.0 1.0 1.0 1.0 - - - - - - - - - - - 1.0 0 I I I ACI 318-63 , bSD Severe 4 1.0 - 1.0 1.0 1.0 - - - - - - - - - - 1.0 0 K I I ACI 318-63 E 3, . WSD Y w Environmental 5 1.0 1.0 1.0 1.0 - 1.0 - - - - - - - - - - 0 0 0 0 1.33 ACI Eg-318 43 USD 6 1.0 1.0 1.0 1.0 - - 1.0 - - - - - - - - - 0 0 0 'O ACI 318-63 ' WSD h E 8 Extreme 7 1.0 - 1.0 1.0 - - - - - - - - 1.0 - - 1.1 0 1 1 I ACI 318-71 SIst I Environmental 8 1.0 - 1.0 1.0 - - - - - - - - - 1.0 - - 0 0 0 0 ACI 318-71 i SDM 7 9 1.0 - 1.0 1.0 - - - - - - - - - - 1.0 - 0 0 0 0 ACI 318-71 SIst l , Abnormal 10 1.0 - - - - - - - 1.0 1.5 - 1.0 - - - 1.25 0 0 0 x los 1.0 - - - - - - - 1.0 - 1.5 1.0 - - - 1.25 x 0 x 0 g Abnormal / Severe 11 1.0 - - - 1.25 - - 1.0 1.0 1.25 - 1.0 - - - 1.0 0 0 0 I ta Environmental lla 1.0 - - - 1.25 - - - 1.0 - 1.25 - - - - 1.0 I O I N'Id E i j~ 11b 1.0 - - - 1.25 - - 1.0 1.0 - - 1.0 - - - 1.0 0 0 K O 0 I.me Theory J L ou Abnormal / Extreme 12 1.0 - - - - - - 1.0 '1.0 1.0 - 1.0 1.0 - - 1.0 0 0 0 X I 1 Environmental 12a 1.0 - - - - - - - 1.0 - 1.0 - 1.0 - - 1.0 K 0 0 x ; t 12b 1.0 - - - - - - 1.0 '1.0 - - 1.0 1.0 - - 1.0 0 0 0 I ' 4 .
;
- Only one SRV condition shown with ar. 'I' is considered at a time. I 1
i i
, , - - . - . , - - - - . - ~ - - - - - - - - - , - , , , , ,, , n. ,m ,,, , - . . . , . - - - , _ , - . - - . .
,_...__ _ _ _.-_ _ _ -. . _ . _ _ _ _ - _ . . . ~ _ ~ .. . . _ _ _ - _ - _ _ _ . . _ - __ _ .
O O O '
)
i TABLE _6.3-1 (Cont'd) L NOTES: a. Loads not applicable to a particular structure or system are deleted.
- b. If for any combination, the effect of any load other than D reduces the load, it is deleted from the combination.
- c. For SRV, the resultant effects for both horizontal and vertical components shall be determined by combining the individual effects by the square root of ~the ,
j sum of the squares. .a vs I
- d. For DBA (annulus pressurization), loads are combined by SRSS method. y m'
- M w
O 85 80
.O
~ EE ;
- ~5 e
- t CO H r
- OW l
l I k l
_ _ _ _ _ _ -_ _ .= - _ _ _ _ . _ _ ._ ) ZPS-1-MARK II-DAR AMENDMENT 13 I OCTOBER 1980 i ( TABLE 6.3-1 (Cont 'd) ! LOAD DEFINITION LEGEND 4
; D -
Dead loads. L - Live loads. 4 Rg - Operating hanger loads. ' Tg - Operating thermal loads. E - Operating basis earthquake loads. 4 g W - Design wind velocity loads. l H - Flood of record hydrostatic loads. R - Local effects due to pipe break including pipe whip reactions and jet impingement loads. T - Thermal loads due to pipe break. a P a DBA pressure loads (including all pool hydrodynamic loadings: CO, chugging). R - Thermal effects on hanger loads. a () E Wt ss Design-basis earthquake loads. Tornado loads. H' -
~ Probable maximum flood loads.
SRV - Safety / relief valve discharge loads. P b SBA and IBA pressure loads (including all pool 4 hydrodynamic loadings: CO, chugging). SBA - Small break accident. IBA - Intermediate break accident. DBA - Design-basis accident (large break accident). I () Y 6.3-4
O O O TABIZ 6.3-2 IDAD DEFINITIONS AND GMBINATIONS FOR STRUCTURAL STEEL ANEPTANCE IDADING IDAD FAC1DRS A CATECORY R T E 1 T P P R E W DESCRIPTION N0' . D L o o o U H r a a b a so t H' SRV* ADS ALL METRICAL SINCIE STRESS Construction 1 1.0 1.0 - - -
- 1. 0 - - - - - - - - - -
0 0 0 0 1.33 AISC Test 2 1.0 1.0 - 1.0 - - - - - - - - - - - - 0 0 0 0 AISC Normal 3 1.0 1.0 1.0 1.0 - - - - - - - - - - - 1.0 0 X X X AISC Savere 4 1.0 - 1.0 1.0 1.0 - - - - - - - - - - 1.0 0 X X X AISC m Environmental 5 1.0 1.0 1.0 1.0 - 1.0 - - - - - - - - - - 0 0 0 0 1.33 AISC M E 6 1.0 1.0 1.0 1.0 - - 1.0 - - - - - - - - - 0 0 0 0 AISC E. m ** E I Abnormal 7 1.0 - - - - - - - 1.0 1.0 - 1.0 - - - 1.0 0 0 0 X 1.6 AISC g Y, 7a 1.0 1.0 1.0 1.0 1.0 X 0 0 .95 9'J u E X x y 8 Extreme 8 1.0 - 1.0 1.0 - - - - - - - - 1.0 - - 1.0 0 X X X 1.6 AISC < es Q .95 n - D S Environmental 9 1.0 - 1.0 1.0 - - - - - - - - - 1.0 - - 0 0 0 0 1.6 AISC 1
.95 FY 10 1.0 -
1.0 1.0 - - - - - - - - - - 1.0 - 0 0 0 0 1.6 AISC <
.95 FY Abnormal / Severe 11 1.0 - - -
1.0 - - - 1.0 1.0 - 1.0 - - - 1.0 0 0 0 X 1.6 AISC ~< Environmental lla 1.0 - - - 1.0 - - - 1.0 - 1.0 1.0 - - - 1.0 X 0 X 0 .95 FY 11b 1.0 - - - 1.0 - - 1.0 1.0 1.0 - 1.0 - - - 1.0 0 0 X 0 E$ Abnormal / Extreme 12 1.0 - - - - - - - 1.0 1.0 - 1.0 1.0 - - 1.0 0 0 0 X 38 "t:" is6 AISC <~ Environmental 12a 1.0 - - - - - - - 1.0 - 1.0 1.0 1.0 - - 1.0 X 0 0 X .95 FY 12b 1.0 - - - - - - 1.0 1.0 - - 1.0 1.0 - - 1.0 0 0 0 X E me* ow , s'Only one SRV condition shown with an 'X' is considered at a time.
O Q P kJ (> d TABLE 6.3-2 (Cont'd) NOTES: a. Loads not applicable to a particular structure or system are deleted.
- b. If for any combination, the effect of any 2oad other than D reduces the load, it is deleted from the combination.
- c. For SRV, the resultant effects for both horizontal and vertical components shall be determined by combining the individual effects by the square root of the sum of the squares.
- d. For DBA (annulus pressurization), loads are combined by SRSS method.
U
- e. Plastic section modulus of steel member shapes is to be used for stress y computation for load combinations lib and 12b. 7 3
m f. Conduit hangers, electrical cable tray hangers and HVAC hangers have been g . designed for load combinations 3, 11, lla, llb, 12, 12a and 12b only. x u O es NE
~5 5C
ZPS-1-MARK II DAR AMENDMENT 13 l OCTOBER 1980 ; () TABLE 6.3-2 (Cont 'd) LOAD COMBINATION LEGEND D - Dead Loads. L - Live Loads. { Rg - Operating hanger loads. Tg - Operating thermal loads. Eg - Operating basis earthquake loads, i W - Design wind velocity loads. H - Flood of record hydrostatic loads. R Local effects due to pipe break including pipe whio 1 reactions and jet impingement loads. . T, - Thermal loads due to pipe break. P a DBA pressure loads (including all pool hydrodynamic loadings: CO, chugging). R a Thermal effects on hanger loads. E ss Design-basis earthquake loads. 4 () Wt - Tornado loads. H' - Probable maximum flood loads. i SRV - Safety / relief valve discharge loads. P b SBA and IBA pressure loads (including all pool hydrodynamic loading: CO, chugging). SBA - Small Break Accident. ! IBA - Intermediate Break Accident. DBA - Design-Basis Accident (Large Break Accident). O 6.3-7 l
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () 6.4 BALANCE-OF-PLANT PIPING AND EQUIPMENT 6.4.1 Wetwell Piping The wetwell piping is directly affected by the suppression pool loads as well as response spectrum loading. The loads, load combinations, and acceptance criteria that have been used for the wetwell piping are given below. WETWELL PIPING LOADS AND LOAD COMBINATIONS LOAD COMBINATION ACCEPTANCE CRITERIA 1 N Service Level A N + OBE + SRV Service Level B N + DBE + SRV + CHUG Service Level C N + DBE + SRV + CO Service Level C N + DBE + DBA Service Level C I where:
- a. SRV - Safety / Relief Valve loads include:
- 1. drag loads based on the T-quencher air clearing load, and
- 2. inertia loads based on the T-quencher device,
- b. CO - Condensa*. ion Oscillation loads includes i
! 1. drag loads ~ based on CO (DFFR) plus additional l considerations up to 21 hertz, and
- 2. inertia load based on the empirical limiting CO load referenced in Chapter 1.0.
- c. CHUG - Chugging loads (20 - 30 hertz)
- 1. drag loads based on 4T adjusted in accordance with NUREG-0487 methods, and i
i 2. inertia loads.
- d. DBE - 1.875 OBE.
- e. DBA - drag loads due to vent clearing, pool swell or pool fallout. Inertia loads are negligible.
O 4 j 6.4-1
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1180 i 6.4.2 Non-Wetwell BOP Piping 6.4.2.1 Load combinations and Acceptance criteria , 6.4.2.1.1 Pipina Systems Within the Reactor Buildinc ! The table below includes the loads, load combinations, and i acceptance criteria considered for the safety-related piping l l systems within the reactor building or attached to the reactor , building wall (excluding wetwell and NSSS piping systems). , The piping and support design was based on the load combination that resulted in governing loads and piping stresses. , 4 LOAD COMBINATIONS CONSIDERED 1 LOAD COMBINATIONS ACCEPTANCE CRITERIA I N + OBE + SRV RH Service Level B N + DBE + SRV gghRH AL Service Level C N + /DBE8 + Apr Service Level C N + OBE + SRV TO Service Level B N+SSE+COfbhfkf Service Level C N + SSE + CHUG + SRVggnf TO Service Level C () where: (EMPIRICAL $ ALL/AS N - Normal loads including pressure, weight, thermal, and operating fluid transient loads; OBE - Operating-basis earthquake loads; Safety / relief valve loads for an all SRV^LLRH valve discharge based on a rams head discharge device; DBE -
~ Design-basis earthquake (1.875 OBE);
AP - Annulus pressurization; SRygggjggyTQ The envelope of the safety / relief valve load due to an all valve discharge and an asymmetric 3-valve discharge based on a T-quencher discharge device; SSE - Safe shutdown earthquake; (]) CO (DFFR) - Condensation oscillation load defined in the DFFR; 6.4-2
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 t~
~'
CHUG - Chugging load defined between 20 to 30 hertz; and Condensation oscillation load CO(EMPIRICAL) including higher frequency contributions (up to 50 hertz). Includes the envelope of two independent CO empirical loads (see Chapter 1.0). 6.4.2.1.2 Pipina Systems Outside the Reactor Building All essential and safety-related piping systems outside the reactor building considered the effects of seismic loads. The load combinations considered included the following: LOAD COMBINATION ACCEPTANCE CRITERIA N + OBE Service Level B N + DBE Service Level C where the loads are as defined in Subsection 6.4.2.1.1. 6.4.3 Balance-of-Plant Equipment r~ ( >3 6.4.3.1 Loadina Combinations The table below defines the combinations of the normal, seismic, and pool dynamic loads considered in the equipment qualification:
- a. N + OBE + Envelope (SRVALL TO and SRVggy TO)
- b. N + SSE + Envelope (SRV ADS TQ and SRV TC) + CO (Zimmer empirical medium mass flux)*ggy
- c. N + SSE + Envelope (SRVADS TO and SRVggy TO) + Chugging
- d. N + SSE + SRVggy TO + CO (Zimmer empirical high mass flux)*
- e. N + /( AP ) 2 + (SSE)2
- f. N + SSE + CO (DFFR Def.)
i
*See Chapter 1.0 for explanation of condensation oscillation.
6.4.3.2 Acceptance Criteria l I
- a. Allowable Stress Limits
() 1. ASME Equipment 6.4-3
I ~_ 2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 0 The allowable stress limits for ASME equipment will be as per Section III of the ASME B&PV Code 1974 Edition.
- 2. Non-ASME Equipment The allowable stress limits for non-ASME equipment will be as provided below.
STRESS LIMIT PLANT STRESS ACTIVE NON-ACTIVE CONDITION COMPONENT EQUIPMENT EQUIPMENT Upset m OE 0.54 0.6 S (0, or o"L)g + B 0.7 Fy Y 0.9 Sy Emergency a 0.7 S 0.9 ( m,or Of 0 )
- B 0.95 Ey 1.5 where o, = general membrane stress, og = local membrane stress, and
(:) o,= bending stress.
- b. Demonstration of Operability The operability of all safety-related equipment has been demonstrated by:
- 1. satisfying allowable stress limits and deflection when qualification has been by analysis; and
- 2. monitoring the equipment functional capability both during and af ter dynar c testing when qualification has been by oest.
i ! I 6.4-4
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () '~ 6.5 NSSS PIPING AND EQUIPMENT, VESSEL, AND INTERNALS i The load combinations and acceptance criteria used for the 8 analysis of the BOP and NSSS system piping, reactor pressure 8 supports and internal component analysis are shown in Table i 6.5-1. The information presented in that table has been i discussed with the staff in various Mark II Owners Group NRC , meetings. As can be seen in Table 6.5-1, the evaluation design limits are in agreement with the conservative interpretation of 8 MEB-6 (draft) technical position " Stress Limits for ASME Class 1, a 2, and 3 Components and Component Supports of Safety-Related i Systems and Class CS Core Support Structures Under Specified i Service Loading Combinations." Note that, per the intent of MEB-6, the piping, equipment, and the reactor pressure vessel support and internals are treated in the analysis as " Type 1" ' components. i i Tables 6.5-2 and 6.5-3 shcw the acceptance criteria for nonfluid , system equipment and active fluid system equipment. , In response to questions from the staff, additional technical ' justification has been provided for the following load i combinations in Table 6.5-1: i (3 a. N + OBE + SRV (/
- b. N + LOCA (1-7) 8 i
These additional justifications are presented in Reference 1. i Peak responses due to related dynamic loads postulated to occur in the same time frame but from different events are combined by ' the square root of the sum of the squares (SRSS). A detailed I discussion of this load combination technique is presented in i Reference 2. In addition, Reference 3 provides an overview , summary of the SRSS method with a summary of its historical , precedence and additional technical considerations. I The technical justification for the time load combinations /accap- i tance criteria mentioned above, and the use of SRSS is'curr atly i under review by the NRC. , i l i l 1 (~) v l 6.5-1
l 1 2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 O
%J 6.5.1 References
- 1. Transmittal letter MFN-193-78, L. J. Sobon to NRC (attention J. P. Knight), dated May 5, 1978, " Mark II Containment Program-Responses to NRC Request for Additional Information," (a response to additional information, S. A. Varga (NRC) to Niagara Mohawk Power Corp., February 21, 1978, Docket No. 50-410).
- 2. General Electric report NEDE-24010, " Technical Bases for the Use of the Square Root of the Sum of Squares (SRSS) Method for Combining Dynamic Loads for Mark II Plant, July 1977.
- 3. Transmittal letter R. A. Hill to NRC (attention J. P.
Knight), dated April 24, 1978, " General Electric-Executive Summary Report and Supplemental Techanical Bases for the SRSS Method of Combining Dynamic Responses."
- 4. Transmittal Letter MFN-175-78, E. D. Fuller to NRC (attention J. P. Knight), dated April 24, 1978, " Square Root Sum of the Squares (SRSS) Meeting Viewgraphs," (Presented April 6, 1978).
G kJ 1 l l (~) l i i i 6.5-2
l
, EPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 ,
fm. TABLE 6.5-1 - ?
) ^
LOADING COMBINATIONS AND ACCEPTANCE CRITERIA OPERATING CONDITION CATEGORIES CONSERVATIVE LOAD DESIGN EVALUATION (1) , (2) INTERPRETATIg) COMBINATION BASIS (1) , ( 2) BASIS OF MEB-6 N+SRV Upset Upset Upset ALL N+0BE Upset Upset Upset N+SSE } Faulted Faulted Faulted N+0BE+SRV Emergency Upset Upset N+SSE+SRV Faulted Faulted (4) Faulted N+SBA+SRV Emergency Emergency Emergency 2 N+IBA+SRV Faulted Faulted Faulted 2 N+SBA+SRV ADS Emergency mergency Emergency M+SBA/IBA+SSE+SRVADS Faulted Faulted Faulted N+LOCA(1-6)+0BE Faulted Faulted I4} Faulted N+LOCA(1-6)+SSE Faulted Faulted (4) Faulted N+LOCA g,7 ) Faulted Faulted I4I Faulted _) N+LOCA g7) +SSE Not Applicable Faulted Faulted N+0BE+SRVggy N+0BE+SRV ADS N+0BE+LOCA 7 N+OBE+SRV ADS ^3 N+SSE+SRVggy N+SSE+SRVgg NOTES: (1) Peak dynamic load responses due to related dynamic loads occurring at the same time from different events are combined with square root of the sun of the squares per NEDE 24010. (2) In addition to stress criteria shown, operability of active components will l'e demonstrated for emergency and faulted loading conditions. Alternatively, the loads on active components will be limited to those associated with a more rentrictive acceptance criteria. (3) The acceptance criteria shown may be used for essential components provided that, in addition to the stress limits shown, operability / functional capa- 4 bility of essential components is demonstrated for the associated loads. ) l (4) Faulted condition used to assure pressure boundary integrity. 1 6.s-3 ! (v"') l l
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 TABLE 6.5-1 (Cont'd). LOAD DEFINITION LEGEND Normal (N) - Normal and/or abnormal loads depending on acceptance criteria. 1 OBE.
- Operating basis earthquake loads.
SSE - Loads due to vibratory motion from safe shutdown carthquake. SRV - Safety / Relief valve discharge-induced loads from
- 2) two adjacent valves.
SRVggg - The loads induced by actuation of all safety / relief valves which activate within milliseconds of each other (e.g., turbine trip operational transients). SRV - The loads induced by the actuation of safety / relief ADS valves associated with automatic depressurization system which. actuate within milliseconds of
-each other during the postulated intermediate size pire rupture.
LOCA - The loss-of-coolant accident associated with the postulated pipe rupture of large pipes (e.g., main steam, feedwater, recirculation Piping). LOCA y - Pool swell drag / fallback on piping and components located between the main vent discharge outlet and the suppression pool water upper surface. LOCA 2 - Pool swell impact loads acting piping and components located above the suppression pool water upper surface. LOCA 3 - Oscillating pressure induced loads on submerged piping and components during condensation oscil-lations, chugging. LOCA 4
- Building motion induced loads from condensation oscillation, chugging.
J LOCA S
- Building motion induced loads from main vent air clearing.
LOCA - Vertical and horizontal loads on main vent piping. 6 LOCA - Annulus pressurization . loads. 7 (]) SBA - Small break accident (includes LOCA3 ). IBA - Intermediate break accident (includes LOCA3 '
- LOCA , and LOCA }*
s 4 6 ( 6.5-4
- AMENDMENT 13 ZPS-1-MARK II DAR OCTOBER 1980 ,
. TABLE 6.5-2 LOAD COMBINATIONS AND ALLOWABLE STRESS LIMITS FOR NON-FLUID SYSTEM EQUIPMENT l NON-ACTIVE PLANT- ACTIVE (AND ACTIVE , CONDITION (ELASTIC DEFLECTION) EXACT DEFLECTION)' 4 50.5S y (D.M.) 56Sy 0 (D.M.) 50.3 Su (B.M.) 10.4 S u (B.M.)
' Upset ot 50.7 Sy (D.M.) -
at 10.9 Sy (D.M.). 50.4 S (B.M.) 50.6 S u (B.M.) 50.7 Sy (D.M.) 50.9 Sy (D.M.) 10.4 su.(B.M.) 50.6'Su (B.M.) ! Emergency i at 10.95 Sy (D.M.) $1.5 S y (D.M.) 1 3 at 50.6 Su (B.M.). 50.9 S u (B.M.) , i t i u om = Membrane stress. i ot- = Membrane'+ bending stress. ' S y
= Yield stress at corresponding temperature.
S u
=
. Ultimate stress at corresponding-temperature.
i
-D.M.
~
= Ductile material. -
B. M.- = Brittle material. O 1 6.5-5 , y e , , - - w,7 e- - e, -,,-,m-+r~,s'- .e-r ,, -g---- -ww--, e s +r c eg ,-r n--e r o, e e+--ev r--a. - - - w ~,>rw,,--*w'
1 l 2PS-1-MARK II DAR AMENDMENT 13 l OCTOBER 1980 ( TABLE 6.5-3 LOAD COMBINATIONS AND ALLOWABLE STRESS LIMITS FOR ACTIVE FLUID SYSTEM EQUIPMENT PLANT CONDITION ASME CLASS 1 ASME CLASS 2 AND 3 Per ASME Section III Per ASME Section III UPSET Same as Non-Active Same as Non-Active om < 1. 0 0 S m om -
<1.00 S h
EMERGENCY at < l . 5 S, at < l.65 S h O am = Membrane stress, at = Membrane + bending stress. O Sh' Sm
= As defined by Section III.
S = Yield stress at corresponding temperature. y 6.5-6 1
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
%s' CHAPTER 7.0 - REEVALUATION AND DES 1s : ASSESSMENT 7.1 CONTAINMENT AND INTERNAL CONCRETE STRUCTURES '
I The containment and internal concrete structures were reevaluated i for the pool dynamic loads to insure structural adequacy. i Dynamic structural analyses using finite-element models were , performed for the reevaluation. Details of the analyses and reevaluation are summarized in this section. 1 7.1.1 Structural Analysis for SRV Loads The structural response of the reactor containment to the dynamic safety / relief valve (SRV) discharge loads was determined by a detailed dynamic analysis of the system, including the effects of soil structure interaction. The structure was analyzed by a finite-element model which was subjected to the SRV load time histories described in Chapter 5.0, and the dynamic response was obtained by numerical integration of the governing differential equations. The SRV discharge cases were analyzed separately and the results were used to check the structural integrity in combination with all other simultaneous loads in accordance with the applicable load combinations (Section 6.1). For the purpose of analysis, the containment was modeled as an () 7_ axisymmetric structure by thin-shell finite elements (Figure. 7.1-1). The structural model includes the basemat, primary containment, reactor pedestal, drywell floor, the reactor pressure vessel (RPV), the foundation soil, and the fluid in the pool. The fluid was simulated by a fluid finite element as described in Reference 2. Also included in the model were the suppression chamber columns, RPV stabilizer truss, and refueling bellows, as well as the containment building and spent fuel pool slab. The model used 356 elements and 322 nodes to represent the structure, RPV, and soil, of which 85 are thin-shell elements. Forty different material properties were used to describe the characteristics of the various components. The RPV was represented by seven shell elements with as many different properties. Drywell floor support columns were modeled as orthotropic shell elements'. Containment building walls and spent fuel pool slab were included as axisymmetric shells to account for their mass and stiffness contribution. The soil was modeled by 153 axisymmetric solid finite elements in nine horizontal layers to the bedrock level at elevation 400 feet. The dynamic strain-dependent stiffness and damping characteristics of the soil were used to determine a stable set of material properties for'the soil elements, consistent with the input forcing function. Refer to Table 7.1-1 for the factors to r~ be used on the modulus and damping curves of Figures 7.1-2 and (-)> 7.1-3 respectively. l 7.1-1 l 1
ZPS-1-HARK II DAR AMENDMENT 13 OCTOBER 1980 l (~)
'# The containment structure model was analyzed by the'Sargent &
Lundy version of the finite-element computer program DYNAX (Appendix A, Section A.1). This program was suitable to analyze axisymmetric shells and solids subjected to arbitrary static or dynamic loads. SRV discharge loads were specified by individual time history variations for the pressure Fourier harmonics in nine zones along
,the containment basemat and reactor support. Figure 7.1-4 shows the zones used to define the various pressure time-histories.
These SRV discharge loads depend upon the devices used at the discharge end of the SRV lines. Two devices were used in the SRV analysis, rams heads and T-quenchers. The analysis of SRV loads was divided into two cases, rams head loading and T-quencher loading. Rams Head Loading Typical pressure time history plots for the rams head, which is described in Subsection 5.2.1, are shown in Figures 7.1-5 and 7.1-6 for Zone 4 on the basemat due to resonant sequential discharge of all valves and asymmetric discharge respectively. ' Different pressure time histories for the various zones and the r- various harmonics were, therefore, used to represent the pressure ( )3 fluctuations on the suppression pool walls. The effect of the varying circumferential and meridional pressure distributions was accounted for in this manner. The dynamic response of the structure to the hydrodynamic pressure loads was then determined by direct numerical integration of the governing differential equations. The response time histories were thus established and the time-wise maximum values were obtained at each element or node location. The acceleration response time-histories were then used to determine the response spectra at the desired locations using the computer program RSG (Appendix A Section A.5). Thc resulting structural responses to the various SRV loads were combined with the other appropriate loads as per the load combinations shown in Table 6.1-1. The margin factors from these load combinations are presented in Tables 7.1-2 through 7.1-16. T-Ouencher Loading l' Typical pressure time histories for the T-quencher are described in Subsection 5.2.2. The direct integration method of dynamic analysis cannot be used T for the T-quencher case because the frequency of the dynamic load (^J K- is a variable and can assume any value in a defined range. Therefore, a dynamic analysis was performed in the frequency 7.1-2
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 domain rather than in the time domain. The method of such an analysis is known as the " Fourier Transform Method" or " Frequency Response Method." Essentially this method is analogous to the t influence line method used in static analyses. The dynamic response of a particular component k (acceleration, ' force, displacement) can be expressed as: 8 Rk (u) =Tk-(u) F (u) : in which i R k (u) = structrual response of a node / element, I I Tk (u) = transfer function of the kth respense, and i F (u) = Fourier transform of the external load. It should be noted that all quantities in the above equation are i scalars and are only functions of the harmonic frequency. i i Transfer function, also known as complex frequency response ' function Tk (u) by definition, is the response of the kth component for unit harmonic load of frequency (u). The transfer ' function is dependent upon the structural properties (mass, I ({) stiffness, damping) alone and is thus unique for a given structure. This is analogous to an influence line which is the i i response of a component (moment, shear) due to an applied unit load to the structure. I The external load which is usually expressed in the time domain i can be expressed in the frequency domain also, using " Fast i Fourier Transform" algorithm. Using this algorithm, a given , function can be transformed from time domain to frequency domain and vice versa. ' s The 7nalysis was performed in the following steps: i
- 1. The containment was modeled as an axisymmetric structure by ,
thin-shell finite elements. The containment structural model was analyzed by the Sargent & Lundy version of the finite- ' element program DYNAX which was capable of analyzing 8 axisymmetric shells and solids subjected to arbitrary i symmetric and asymmetric static or dynamic loads. The i symmetric and asymmetric SRV loads were applied as Fourier , sine and/or cosine harmonics for each case. A band-limited white-noise time history was used for the analysis. The ' Fourier transform of such a time history has a constant 8 magnitude at all values within the frequency range (0 to i 45 hertz) of interest. i () 2. The response (force, moment) time histories obtained from t."a above white-noise analysis were stored in electronic files. i 7.1-3
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
- 3. The transfer functions of the response were obtained by.the
(~-) s computer program FAST. I From Equation (1) i i Tk (a) = k (a) F (u) , in which Rk (u) was the Fourier transform of the responses ' saved in step (2) and F (u) is the Fourier transform of the I white noise load used in step (1) of the above. i
- 4. For steady-state solution of the harmonic load, by definition from Equation (1), the transfer function itself was the response.
For SRV loads with variable frequency, the transfer functions ' were scanned in the frequency range of the loading. The ' maximum response could be obtained as the product of the i transfer functions and the Fourier transforms of the load, i using the FAST program. Response acceleration time histories , were also generated to obtain response spectra using the RSG , program. The following pressure time histories were used for the present assessment of the ZPS-1 containment. O MAX. PRESSURE FREQUENCY RANGE TRACE I (psi) (hertz) KWU 35 17.41 2.930 ~ 6.510 KWU 76 13.16 3.906 ~ 8.680 KWU 82 11.42 3.418 ~ 7.595 In order to consider a conservative frequency content, the above three traces were expanded into longer and shorter time history durations by multiplying the time scales by a factor of 2.0 and 0.9, respectively. In addition, the pressure scale were multiplied by a factor of 1.6 for each of the ; three traces. The resulting structural cesponses to the various SRV T-quencher loads were combined with the other appropriate loads as per the load combinations shown in Table 6.1-1. The margin factors from these load combinations are presented in Tables 7.1-17 through 7.1-24. 7.1.2 Structural Analysis of LOCA Loads The analysis of the structure for the LOCA loads was performed as s a set of analyses covering each LOCA related phenomenon 7.1-4 i
ZPS-1-MARK II DAR AMENDMENT 13 , OCTOBER .1980 N separately. The methods used for each analysis are summarized in !
'the following for the LOCA-induced loads of chugging, condensation oscillation, pool swell, and vent clearing.
i 7.1.2.1 Vent Clearing Analysis The description of vent clearing load for analysis is presented in Section 5.3 and in DFFR Section 4.2. The spatial distributions of the LOCA vent clearing load on the wetted i surface of the suppression pool are shown in Figure 7.1-7 for the rams head case and in Figure 7.1-8 for the T-quencher case. The magnitude of the load for T-quencher case is 33 psig below the vent exit attenuated linearly to zero at the pool surface. The model used in the analysis of the vent clearing loads was different from the one described in Subsection 7.1.1. The model used in this analysis is shown in Figure 7.1-9. This model was similar to the one used in Subsection 7.1.1 but excluded nodes and elements for the fluid in suppression pool. The structural model included the basemat, primary containment, reactor pedestal, drywell floor, and reactor pressure vessel (RPV). The soil was modeled by axisymmetric solid finite elements in nine horizontal layers down to the bedrock level. For static equivalent analysis, the model described in Subsection 7.1.1 and the model described in this subsection were virtually () 1 the same. The containment structure was analyzed for the effects of the vent clearing load statically using Sargent & Lundy's axisymmetric finite-element computer program DYNAX. See 4 Appendix A Section-A.1 for a description of the computer program. j The response of the structure to vent clearing loading was determined by direct numerical integration of the governing differential equations. l The resulting structural response to the vent clearing load is j combined with the other loads as per the load combinations shown 4 in Table 6.1-1. 7.1.2.2 Pool Swell Analysis The postulated pool swell phenomena induced loads are described in Subsection 5.3.1.3.3 and in DFFR Subsection 4.2.4.4. Using the model described in Subsection 7.1.2.1, the containment
-structure was analyzed for two load cases for the LOCA pool swell load, the symmetric and the asymmetric loads.
0 For the symmetric load, the loading was applied over the entire
- n U
3600 of the containment wall. The pressure history of the drywell and wetwell air space is given in Figure 7.1-10. Curve A l 7.1-5
ZPS-1-MARK II DAR hMENDMENT 13 OCTOBER 1980 () of this figure applies to the drywell, and Curve B applies to the portion of the wetwell wall which is above the pool water surface. The LOCA-pool swell portion of these curves ends at time 2.97 seconds. The peak wetwell air space pressure during this event was 23 psig, while the peak drywell pressure was 21 psig. For the portton of the wetwell walls which is below the water surface, the load definition is given in Figure 7.1-11. This load was 22 psig at the basemat level which decreased linearly to 16 psig at the elevation of the vent exit, and then increased linearly to 23 poig at the maximum pool swell elevation. For the asymmetric load, the peak drywell pressure was applied uniformly over the entire drywell. Figure 7.1-12 shows the pressure distribution of the pool swell asymmetric load for the wetwell. The asymmetric pool swell load of 23 psig was applied over a sector of 1800, in addition to the hydrostatic load. The containmar.t structure was analyzed for the effects of the pool swell loads statically using Sargent & Lundy's axisymmetric finite-element computer program DYNAX. See Appendix A Section O A.1 for a description of the computer program. The spatial pressure load distributions in the circumferential direction were represented by using Fourier hacmonics. The resulting forces and moments on the structure's design sections were obtained directly from the DYNAX computer output. The resulting structural responses to the pool swell loads were combined with the other appropriate loads as per the load combinations shown in Table 6.1-1. 7.1.2.3 Condensation Oscillation Analysis Following the pool swell transient, steam flows through the main vent system into the suppression pool, where it condenses. Evaluation of the steam-condensation phase of the 4T test results revealed the existence _of a dynamic load during high and medium steam mass flux into the suppression pool. This load, called condensation oscillation (CO), is a low-amplitude, symmetric, sinusoidal pressure fluctuation occurring over a range of frequencies. The 2PS-1 containment was designed for the following load definition: O O 7.1-6 i
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (} Magnitude: t 3.75 psi Frequency: 2 to 7 hertz The spatial distributions of the condensation oscillation loads are shown in Figure 7.1-13 for rams head design basis and in Figure 7.1-14 for T-quencher design basis, as 3.75 psig acting at a frequency 2 to 7 hertz on the basemat, containment, and reactor pedestal. The structural model described in Subsection 7.1.2.1 was used for the rams head design basis, and the one described in Subsection 7.1.1 was used for the T-quencher design basis. The load was assumed to be harmonic in time, and only the steady-state response was considered as being of interest. For this purpose, frequency response variations were determined for all response components of interest using the computer program FAST, Appendix A, which obtained the complex frequency response by calculation of the discrete Fourier transform of both load and response. The frequency range of 2 to 7 hertz on the frequency response was considered relevant in evaluating the structural response. The resulting structural responses to the condensation oscillation loads were combined with the other appropriate loads (7-,gj as per the load combinations shown in Table 6.1-1. The margin factors from these combinations are presented in Tables 7.1-2 through 7.1-24. In addition to the above CO load (2-7 hertz), an empirical limiting CO load was considered in combination with the T-quencher design basis of the ZPS-1 containment. This load was intended to be a best estimate of the conservative load specification which resulted from the full-scale condensation oscillation test to be conducted in the 4T facility. All the details for this load are described in Chapter 1.0. This 2PS-1 empirical CO load was incorporated for the T-quencher design basis. The spatlal distributions of this load are shown in Figure 7.1-14. The resulting structural responses to this empirical CO load were combined with the other apptcpriate loads as per the load combination shown in Table 6.1-1. The margin factors from the load combinations are presented in Tables 7.1-25 through 7.1-28. 7.1.2.4 Chuacina Analysis The chugging loads used in the analysis are described in Section 5.3 and presented in Figure 7.1-15. The finite-element model (} used in the analysis is described in Subsection 7.1.2. 7.1-7
1
.ZPS-1-MARK II DAR AMENDMENT-13 l OCTOBER 1980 ,
i
-() The method used for T-quencher loading described in Subsection 7.1.1 was used for the chugging load-due-to the variation of 1 - frequency of the load.
! The typical-pressure time history is shown in Figure 7.1-16. i j - The resulting structural responses to the chugging. loads combined I with the other appropriate loads as.per the load combinations are
- shown in Table 6.1-1.
!, 7.1.3 Effects'of Downcomers on the Drywell Floor l The downcomer vents are now subjected to a. variety of submerged structure dynamic loads resulting from SRV and LOCA loads. By assuming, conservatively, that the maximum responses from the various dynamic loads occur simultaneously and in the same direction, the magnitude of the.resulting moments and forces l being transmitted to'the drywell floor becomes significant with ) i respect to the known existing loads on the design sections. Even j though the downcomers are braced at elevation 496 feet in order i to reduce loads-on the drywell floor, the analysis that is
; summarized in this subsection proves that the drywell floor has
- maintained its structural adequacy despite the addition of new loads.
(3 The loads on the downcomers resulting from submerged hydrodynamic V forces are described in Subsection 5.3.1.1.7. } In addition to'the pool dynamic loads on the downcomers, the . seismic loads were also considered in the analysis. These i considerations assumed that all of the downcomers were loaded
- equally, simultaneously, and in the same direction by using the
! response spectra generated from the various loads on the drywell { floor and performing a moda1' analysis. ! The drywell floor is modeled as a thin elastic circular plate I with a circular hole in-the middle. The slab is assumed to be fully restrained at the pedestal and containment walls and simply
- supported.at the columns. The model of the drywell' floor is
- shown in Figure-7.1-17.
! The locations of the downcomers lie along four. rings at radii 18 feet 3. inches,'22 feet 3 inches,-30 feet 9 inches, and~35 feet
- 3. inches.
i: A concentrated radial or circumferential moment, in the form of-4 Fourier' harmonics, is applied at a point on each one of the i' downcomer rings.
~
Figure 7 1-18 shows the circumferential distribution of floor moments induced by a concentrated radial moment applied at radius (]) - 22 feet 3 inches. For computational convenience, the ordinates are normalized to make the-induced radial moment equal to unity. 7.1-8 t
..m'.. , , . . --,..,m,------.--,. ,- ,, y . en . , - - . .y --E, - .- m r. --w.w, r- e.,<,m.m. ...--,.-,mo.-ye.,,,,-,,.-, ,-,,,.m ,-,..mm,.m -,..e-
l l 2PS-1-MARK II DAR AMENDMENT 13 l OCTOBER 1980 ' (] Figure'7.1-19 shows the radial distribution of floor moments induced by the concentrated radial moment appled at radius 22 feet 3 inches. Figures 7.1-20 and 7.1-21 respectively show the circumferential moment applied at radius 22-feet 3 inches. Similar sets of-curves are generated for the other rings of downcomers. Using the moment coefficients frca these curves in the following equations, the radial and circumferential design moments at any design section resulting from the downcomer loading described in Section,6.1 can be calculated: 4 360* 360*
$ 0 0 n l_ &n $n*0$$n'e=o $$n0) + On On.6$0n* e=o $0n0) 4~ 360 360
~
"O " n=1 On On*000n* Obo 000n0[ &n $n*00$n* OEo 0 0$nO f where:
O 4
=
rediei moment induced et a design section, n = number of rings of downcomer (n = 1, 2, 3, 4); M = radial moment applied at a point on the n
$n ring; k =
4n m$n/M$ ni m$n
=
radial moment induced in the floor at the point where M is applied;
$n M en =
circumferential moment applied at a point on the nth ring. m en = radial moment induced in the floor at the point where M is applied; en k 0n = m en#M0n'
=
8 44n normalized radial moment along the radius through the point where M is applied; n O 7.1-9
I 1 2PS-1-MARK'II DAR AMENDMENT 13 j OCTOBER 1980 i i () 3 44n
= nurmalized radial moment along the radius through the point where M en is applied; 840n
= normalized radial moment along the n th ring due to_M4n' i 3 = normalized radial moment along the n th ring 40nB due-to Meni i s 00ne
=
nggmalized circumferential moment along the n ring due to Men; and
=
- a 04n0 nogmalizedcircumferentialmomentalongthe n t ring due to M n-4
'The absolute values of the moment coefficients are used to account for the random direction of the downcomer lateral loads and to obtain the absolute maximum values of m and m g for design
-assessment. 4 Figure 7.1-22 shows the variation of radial moment at critical-design Section 2 (see Figure 4-10 of Reference 1) as the number of loaded downcomers is increased from 1 to 88 (all). The maximum design moment of 52 ft.k/ft occurs when all the downcomers are loaded simultaneously with 8.8 kips each.
() The conservatism included in the design assessment of the drywell floor is best illustrated by a comparison of Figures 7.1-22 and 7.1-23. Figure 7.1-23 shows the plot of the design radial moment at Section 2 versus the number of downcomers loaded as per Figure j- 4-10a of DFFR (Reference 1), Proprietary Supplement Revision 2,
- which defines the probable load on multiple downcomers as decreasing with increasing number of loaded downcomers. The i maximum moment thus obtained is orily 29 f t.k/f t, whereas a i conservative value.of 52 ft.k/ft is obtained~by the bounding load definition used in the 2PS-1 drywell floor ~ design assessment.
The assessment of this subsection was based on rams head design basis. 7.1.4 Desian Assessment Marcin Factors 7.1.4.1 -Critical Desian Sections. The primary-containment and internal structures have been checked as to-the~ structural-~ capacity to withstand the dynamic loads due to-SRV discharges and LOCA in addition to the other-appropriate i loads' described in the FSAR. The methods of analysis used have l been' described in the preceding subsections, and the design load combinations are-given in Table 6.1-1. The structural capacity l
- acceptance criteria are the same as in the FSAR, for which all y
(]) design sections have been evaluated using the computer program
'TEMCO (described in Appendix A.7).
i 1 i '7.1-10 I e,- - - - -- . .n . n r m , ,-r,-,,,m---.,--- --,-n, .,n=---m,,-,-,-vw,
2PS-1-MARK II DAR AMENDMENT 13
~ OCTOBER 1980 Figure 3.1-1 shows a cross section of the primary containment and q . internal structures. Figures 7.1-24 through 7.1-31 illustrate the reinforcing steel and prestressing tendon layout. Figure 7.1-29 shows the reinforcing in the pedestal prior to modification. Details of the c'ncrete-filled portion of the pedestal are shown in Figure 7.1-28.
Figures 7.1-32 through 7.1-34 show the design sections in the basemat, containment, reactor support, drywell floor, and drywell floor column considered for structural assessment. Figures 7.1-35 and 7.1-36 give typical design section capacity interaction diagrams of the basemat and containment for the T-quencher design basis. 7.1.4.2 Desion Forces ard Marcin Factors The design forces in the critical design sections were obtained by combining with the ABS method the peak effects of all the loads according to the load combinations defined in Table 6.1-1. The material stresses in the critical design sections were obtained using the computer program TEMCO described in Appendix A. Margin factors, defined as the ratio between the allowable stress and the actual stress in the section, were computed for each g design section. If any of the loads (such as temperature) other than dead load reduced the design forces, it was detected from the load combination to obtain the most conservative margin factor. Margin factors for the basemat and containment wall are reported in the following tables: RAMS HEAD DESIGN BASIS
- a. Basemat Tables 7.1-2 through 7.1-5
- b. Containment wall Tables 7.1-6 through 7.1-9
- c. Reactor support Tables 7.1-10 through 7.1-13
- d. Drywell floor Table 7.1-14
- e. Drywell floor column Tables 7.1-15 and 7.1-16 These tables give the calculated design margin factors for the load combinations, including each of the four modes of SRV discharge for which the structures were analyzed (resonant sequential symmetric discharge, ADS, and two valves) and LOCA hydrodynamic effects combined with the single-valve discharge case. ,
7.1-11
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 The forces of reactor support margin factor were obtained by analysis using the model described in Subsection 7.1.2.1. Margins shown in Table 7.1-14 for loading conditions 4a, Sa, and 7a on the drywell floor are for the LOCA effects, including the lateral loads on the downcomers. As per DFFR Subsection 4.4.6.6, a net upward load of 9 psid acting on the drywell floor has been conridered. Margins shown in Table 7.1-14 for loading conditions 1, 2, 3, and 6 on the drywell floor are for all the valves discharge loading which clearly governs the design of the drywell floor rather than the asymmetric two valve discharge loading. Loading conditions 4, 5, and 7 in Table 7.1-14 include all loads resulting frcm a small pipe break combined with the loads due to the discharge of all 13 SRV'c. This was done for reasons of analytical expediency, since the discharge of all 13 SRV's transmits significantly more energy to the drywell floor than the 6 valve ADS discharge. Since ZPS-1 can take this higher loading case, the actual loading from the ADS valves was not considered. For the drag loads on the downcomer, the maximum load described in Section 5.3 was used fcr all loading combinations which include SRV. loads irrespective of the discharge mode (ALL, ASYMMETRIC, or ADS).
~
T-OUENCHER DESIGN BASIS LOAD COMBINATION WITH NRC CO LOAD (DFFR)
- a. Basemat Tables 7.1-17 through 7.1-20
- b. Containment wall Tables 7.1-21 through 7.1-24 LOAD COMBINATION WITH EMPIRICAL LIMITING CO LOAD
- a. Basemat Tables 7.1-25 and 7.1-26 I
- b. Containment wall Tables 7.1-27 and 7.1-28 The margin factors were calculated as results of the assessment based on the NRC acceptance criteria (modified for the T-quencher).
All the margin factors were greater than 1.0 except the following cases: ("}-
\_/
I i 7.1-12 l _
I 2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 19Rn REINFORCING SHEAR DESIGN O- LOAD LOCATION M.F M.F CASE SECTION
! w/NRC-CO Basemat -
0.96 SRV-Single 2 Containment 0.9 - SRV-ADS 12 Wall 0.9 - SRV-Single 12 w/ Empirical Basemat - 0.97 SRV-Single 3
, Limiting CO -
0.98 SRV-Single 3 l Containment 0.9 m SRV-ADS 12 Wall 0.9 - SRV-Single 12 From the comparison of margin factors, it turned out that two
~
different condensation oscillation loads gave similar structural i responses. The assessments for ;ther structures, such as reactor support, e drywell floor,.and drywell floor column, have not been completed. The material strengths used in computing these margins are the O. minimum specified values. The reinforcing and concrete quality 4 control test results show that material strengths are higher than the minimums specified by the materials specifications and that
- the actual margins are therefore greater than shown in Tables 7.1-2 through~7.1-24.
These safety margins are in addition to the overload factors used
; -in the load' equations'given in Table 6.1-1-and the material understrength factors built into the allowable stress criteria.
Therefore the safety margins between the actual internal moments and forces and the ultimate strength of the structures'is considerably higher than those given in Tables 7.1-2 through 7.1-24. As stated in FSAR Table 3.8-3, if in any load combination, the
- effect of any load (such as temperature) other than dead load reduces the design forces, it will be deleted from the combination. ' Safety margins are '"us calculated with and without '
temperature load, and only the smallest-margins obtained are given in Tables 7.1-2 through 7.1-24. Even though a few of the margins reported in the tables are close to or less than 1.0, it must-be emphasized that conservative
- loads, analysis procedures, and material strengths were used in
- the assessment in. order to expeditiously verify the adequacy of 4
the structure for the pool' dynamic loads. Therefore, the margins [}
. 7.1-13
. . . . - - , , . . , - , , , . . . . , _ - , , - - _ . - - , e., ,.,--,- , e- .-em.,- ,...cyr,,,-v rg.-
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 O reported are the most conservative. higher than reported. The actual margins will be 7.1.5 References
- 1. " Evaluation of Fluid Structure Interaction Effects on BWR Mark II Containment Structures," NEDE-21936-P.
2.- A. J. Kalinowski, " Transmission of Shock Waves into Submerged Fluid Filled Vessels," ASME Conference on FSI Phenomena in Pressure Vessel and Piping Systems, TVP-TB-026, 1977. 1 O i O 7.1-14
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980. TABLE 7.1-1 O- DYNAMIC SOIL PROPERTIES
- I I
OVERBURDEN FACTOR ON ELEVATION SOIL PRESSURE DAMPING (ft) om' (KSF) CURVE 469-466 5.18 1.0 466-463 5.42 1.0 463-460 5.66 1.0 456-460 5.94 1.0 448-460 6.42 1.0 440-448 7.06 1.0 430-440 7.78 1.0 420-430 8.58 1.0 400-420 9.78 1.0-O
-*These values are to be used in conjunction with Figures 7.1-2 and 7.1-3 for the average shear modulus and damping curves.
(a"i 7.1-15 k_ _ -
O O O TABLE 7.1-2 MARGIN TABLE FOR BASEMAT - RESONANT SEQUENTIAL SYMMETRIC DISCHARGE (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) s., STRESS COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION
- FACTOR SECTION FACTOR SECTION FACTOR SECTION 1 1.8 2 4.0 3 1.5 2 2 2.3 1 3.3 3 2.4 3 $
3 1.9 1 3.3 3 1.2 2 b
< i e 4 NA NA NA NA NA NA B $ 4a NA NA NA NA NA NA N 5 NA NA NA NA NA NA o
Sa NA NA NA NA NA NA $ 6 1.9 1 3.3 3 1.3 2 7 NA NA NA NA NA NA 7a NA NA NA NA NA NA Og OM e6 mx
- Refer to Table 6.1-1 *$
~
** Margin Factor = Allowable Stress / Actual Stress (;; 8
*** Refer to Figure 7.1-32 mH NA = Not Applicable e
O O O TABLE 7.1-3 MARGIN TABLE FOR BASEMAT - ADS VALVE DISCHARGE (WITH PLhNT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION
- N FAC'IOR SECTION FACTOR SECTION FACTOR SECTION 1 NA NA NA NA NA NA 2 NA NA NA NA NA NA y m
3 NA NA NA NA NA NA 4 H ' 4 1.3 1 2.2 3 1.2 3 4a NA NA NA NA NA NA y 5 1.3 1 2.4 3 1.3 3 [ , Sa NA NA NA NA NA NA U lc 6 NA NA NA NA NA NA 7 1.3 1 2.3 3 1.3 2 7a NA NA NA NA NA NA Og OM 85 ei5 2
- Refer to Table 6.1-1 y8
** Margin Factor = Allowable Stress / Actual Stress mP
*** Refer to Figure 7.1-32 NA = Not Applicable i
O O O TABLE 7.1-4 MARGIN TABLE FOR BASEMAT - TWO-VALVE DISCHARGE (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS OMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL
; EQUATION
- FACTOR SECTION FACTOR SECTION FACTOR SECTION 1 2.5 2 4.5 3 1.8 2 2 2.3 1 3.6 3 2.6 3 y w
3 1.9 1 3.0 3 1.4 2 O H i 4 1.3 1 2.2 3 1.2 3 i d 4a NA NA NA NA NA NA Q 5 1.2 1 2.3 3 1.4 3 [ , 5a NA NA NA NA NA NA $
- o 6 1.9 1 2.9 3 1.4 2 7 1.3 1 2.3 3 1.3 2
- 7a NA NA NA NA NA NA 88 DE z
*Ref er to _ Table 6 .1-1 $
i ** Margin Factor = Allowable Stress / Actual Stress @d
* *
- Refer to Figure 7.1-32 NA = Not Applicable t
O O O TABLE 7.1-5 MARGIN TABLE FOR BASEMAT - LOCA PLUS ONE SRV (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION
- FACTOR SECTION FACTOR SECTION FACTOR SECTION 1 NA NA NA NA NA NA 2 NA NA NA NA NA NA y m
3 NA NA NA NA NA NA 0 7 4 NA NA NA ' s NA NA NA 4a 1.4 1 2.1 3 1.2 3 5 NA NA NA NA NA NA U Sa 1.3 1 2.2 3 1.2 3 $ e w 6 NA NA NA NA NA NA 7 NA NA NA NA NA NA 7a 1.3 1 2.2 3 1.2 3 Oh
- Refer to Table 6.1-1 W
** Margin Factor = Allowable Stress / Actual Stress
*** Refer to Figure 7.1-32 Hk
$g NA = Not Applicable CW t
O O O TABLE 7.1-6 MARGIN TABLE FOR CONTAINMENT - RESONANT SEQUENTIAL SYMMETRIC DISCHARGE (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) N STRESS COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION
- FAC'IOR SECTION FACTOR SECTION FACTOR SECTION 1 103.9 13 3.4 1 1.7 6 2 8.5 9 2.8 13 1.3 6 y m
3 4.8 11 2.7 1 1.3 6 E 7 w 4 NA NA NA NA NA NA 4a NA NA NA NA NA NA > x 5 NA NA NA NA NA NA U Sa NA NA NA NA NA NA 6 4.6 11 2.7 1 1.3 6 7 NA NA NA NA NA NA 7a NA NA NA NA NA NA og OM Pn 5 si5 2
- Refer to Table 6.1-1
** Margin Factor = Allowable Stress / Actual Stress g8 ms"
*** Refer to Figure 7.1-32 NA = Not Applicable i
O O O TABLE 7 .1-7 MARGIN TABLE FOR CONTAINMENT - ADS VALVE DISCHARGE _ (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS COMPONENT REINFORCING TENSION CONCRETE COMPRESSION LOAD SHEAR COMBINATION MARGIN ** CRITICAL *** MARGIN RITICAL MARGIN CRITICAL
' EQUATION
- FACTOR SECTION FACTOR ESCTION- FACTOR SECTION 1 NA NA NA NA NA NA 2 NA NA NA NA NA NA $
y 3 NA NA NA NA NA
?
. NA H
.7 4 4.0 10 2.9 12 ~1.6 6 U .
4a NA NA NA NA h PA NA .x 5 3.7 11 2.9 13 1.6 6 U Sa NA NA O NA NA NA NA y 6 NA NA NA NA NA NA 7 3.3 11 2.9 13 1.6 6 7a NA NA NA NA NA NA og OR i 85
- Refer to Table 6.1-1 95
** Margin Factor = Allowable Stress / Actual Stress p$
* ** Refer to Figure 7.1-32 $g ow NA = Not Applicable
,C')
f.Du .] {s} \s TABLE 7.1-8 MARGIN TABLE FOR CONTAINMENT - TWO-VALVE DISCHARGE (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION
- FACTOR SECTION FACTOR SECTION FACTOR SECTION 1 NA NA 3.6 1 1.3 6 2 8.6 11 2.8 13 1.3 6 Ej m
3 4.8 11 2.8 1 1.3 6 4 7 4 4.0 10 2.9 ' 13 1.6 6 U $ 4a NA NA NA NA NA NA y 5 3.7 11 2.9 13 1.6 6 [ Sa NA NA NA NA NA NA $
- o 6 4.4 11 2.8 1 1.3 6 7 3.3 11 2.9 13 1.5 6 7a NA NA NA NA NA NA Og OM 85 rn
- Refer to Table 6.1-1 :U ,
** Margin Factor = Allowable Stress / Actual Stress HU
*** Refer to Figure 7.1-32 NA = Not Applicable
$s"
O O O TABLE 7.1-9 MARGIN TABLE FOR CONTAINMENT - LOCA PLUS ONE SRV (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION
- FAC'IOR SECTION FACTOR SECTION FACTOR SECTION 1 NA NA NA NA NA NA 2 NA NA NA NA NA NA y m
3 NA NA NA NA NA NA h H 4 NA NA NA NA NA NA 4a 3.5 10 2.8 13 1.3 6 y 5 NA NA NA NA NA NA U Sa 3.7 11 2.8 13 1.4 6 h 6 NA NA NA NA NA NA 7 NA NA NA NA NA NA 7a 3.2 13 2.8 13 1.4 6 og OM 85 '
- Refer to Table 6.1-1 y$
**Margir ' actor = Allowable Stress / Actual Stress g$
*** Refer 1 Figure 7.1-32 e NA = h Applicable $U ,
O O O-TABLE 7.l-10 MARGIN TABLE FOR REACTOR SUPPORT-RESONANT SEQUENTIAL SYMMETRIC DISCHARGE (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION
- FACIOR SECTION FACTOR SECTION FACTOR SECTION 1 4.4 9 11.9 9 3.0 9 2 2.5 6 6.6 9 1.7 9 y w
.3 1.3 7 4.8 9 1.7 9 E 7 4 NA NA NA NA NA NA 4a NA NA NA NA NA NA @
5 NA NA NA NA NA NA U Sa NA NA NA NA NA NA $ w 6 1.1 7 4.6 9 1.8 9 7 NA NA NA NA NA NA 7a NA NA NA NA NA NA og OM 85
- Refer to Table 6.1-1 to M g 2
** Margin Factor = Allowable Stress / Actual Stress y8
*** Refer to Figure 7.1-32 mr NA = Not Applicable "
O O O TABLE 7.1-11 MARGIN TABLE FOR REAC'IOR SUPPORT - ADS VALVE DISCHARGE (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS COMPONENT REINFORCING TENS'3N CONCRETE COMPRESSION SHEAR LOAD s COMBINATION N MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION * 'N FACTOR SECTION FACTOR SECTION FAC'IOR SECTION 1 NA NA NA NA NA NA 2 NA NA NA NA NA NA y en 3 NA NA NA NA NA
.a NA E H 4 1.3 1 '
4.0 2 1.6 9 b m $ 4a NA NA NA NA NA NA 'd 5 1.2 2 4.4 2 1.7 9 [ Sa NA NA NA NA NA NA $
- o 6 NA NA NA NA NA NA 7 1.01 7 4.4 2 1.7 9 7a NA NA NA NA NA NA Og O td 85 tn g
- Refer to Table 6.1-1 :D ".,
** Margin Factor = Allowable Stress / Actual Stress HO
*** Refer to Figure 7.1-32 $p NA = Not Appl cable C"
O O O TABLE 7.1-12 MARGIN TABLE FOR REACTOR SUPPORT - TWO-VALVE DISCHARGE (WITH PL;49T-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) N N STRESS
'NCOMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION
- N FACTOR SECTION FACTOR SECTION FACTOR SECTION 1 5.3 9 12.1 9 6.5 9 2 2.4 6 6.7 9 2.8 9 y rn 3 1.5 4 5.5 8 2.7 9 4
'y 4 1.3 1 4.1 2 '
2.2 9 u 4a NA NA NA NA NA NA 3 5 1.2 2 4.6 2 2.2 9 y 5a NA NA NA NA NA NA $
- c 6 1.19 7 4.7 9 2.6 9 7 1.0 4 4.5 2 2.2 9 7a NA NA NA NA NA NA se Eli
;
- Refer to Table 6.1-1 g$
** Margin Factor = Allowable Stress / Actual Stress
*** Refer to Figure 7.1-32 gg ow NA = Not Applicable
O O O TABLE 7.1-13 MARGIN TABLE FOR REACTOR SUPPORT - LOCA PLUS ONE SRV (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION
- FAC'IOR SECTION FAC'IOR SECTION FACTOR SECTION 1 NA NA NA NA NA NA 2 NA NA NA NA NA NA s
^ 3 NA NA NA N NA NA NA g
-a r 4 NA NA NA NA NA NA 8 4a 1.3 2 4.1 2 1.8 1 5 NA NA NA NA NA NA H Sa 1.15 2 3.4 5 1.8 1 y
- o 6 NA NA NA NA NA NA 7 NA NA NA NA NA NA
[ 7a 1.06 2 3.2 5 1.8 1 i
'Oy OM 85
- Refer to Table 6.1-1
** Margin Factor = Allowable Stress / Actual Stress
* ** Refer to Figure 7.1-32 g$
e NA = Not Applicable $U
O O O TABLE 7.1-14 MARGIN TABLE FOR DR%2LL FLOOR - SRV ONLY AND LOCA PLUS ONE SRV (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD 4 COMBINATION MARGIN ** CRITICAL **.* MARGIN CRITICAL MARGIN CRITICAL EQUATION
- FACTOR SECTION FACTOR SECTION FACTOR SECTION 1 3.6 3 6.4 3 7.1 1 2 5.7 3 5.2 3 10.8 4 y m
3 4.4 3 4.2 3 10.9 4 4 w e 4 1.8 3 4.0 3 4.1 6 b 4a 1.5 2 1.6 1 3.3 1 g 5 3.3 1 6.4 1 4.7 6 [ Sa 1.7 1 1.4 1 2.7 3 6 9.7 6 7.7 3 11.7 4 7 2.4 3 6.5 3 5.3 6 7a 1.4 2 1.5 1 2.1 2 g
' O 95
- Refer to Table 6.1-1 g$
** Margin Factor = Allowable Stress / Actual Stress
*** Refer to Figure 7.1-33 gg ow l
. _ _ . _ _ _ _ _ __ _ _ _ __ __ _. . . _ - ~
O O O TABLE 7.1-15 MARGIN TABLE FOR DRYWELL FLOOR COLUMN - ALL VALVE AND ADS DISCHARGE
\ road COMPONENT AXIAL COMPRESSION MOMENT SHEAR LOAD i COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION
- FACTOR SECTION FACTOR SECTION FAC'IOR SECTION 1 1.40 1 1.10 1 1.22 1 2 1.94 1 1.25 1 1.41 1 1 1 3 1.84 1 1.10 1.25 m i
" 1 1
. 4**** 2.26 1 1.90 2.09 '
w 4a NA NA NA NA NA NA h 5**** 2.46 1 1.71 1 1.90 1 U Sa NA NA NA NA NA NA n ; 1.78 1 1.28 1 1.46 1 8
- 6 i ;
7**** 1.78 1 1.49 1 1.43 1 g f l 7a NA NA NA NA NA NA Oy OM ~ 85
- Refer to Table 6.1-1 to i
5 g
** Margin Factor = Ultimate Load / Actual Load z
*** Refer to Figure 7.1-34 P8
**** ADS Discharge Case Es OW NA = Not Applicable ;
i 1
l O 0 O i TABLE 7.1-16 MARGIN TABLE FOR DRYWELL FLOOR COLUMN - TWO-VALVE DISCHARGE LOAD COMPONENT AXIAL COMPRESSION MOMENT SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION
- s'N FACTOR SECTION FACTOR SECTION FACTOR SECTION 1 1.98 1 1.87 1 2.09 1
- 2. 2.26 1 2.15 1 2.42 1 1
3 2.14 1.71 1 1.93 1 y
." 4 2.26 1 2.24 1 2.51 1 e
h 4a NA NA NA NA NA NA 1 1 5 2.15 1.94 2.20 1 U 5a NA EA NA NA NA NA e 6 2.08 1 1.93 1 2.20 1
- l 7 2.08 1 1.69 1 1.58 1 1.
l 7a NA NA NA NA NA NA 3R
- Refer to Table 6.1-1 Ob
** Margin Factor = Ultimate Load / Actual Load $$
*** Refer to Figure 7.1-34 g$
NA = Not Applicable g OW i
O O O TABLE 7.1-17 MARGIN TABLE FOR BASEMAT - alt.-VALVE SRV QUENCHER DISCHARGZ (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION
- FAC'IOR SECTION FACTOR SECTION FACTOR SECTION 1 1.18 2 3.53 3 1.44 2 2 2.20 2 2.94 3 2.36 3 y
=
3 1.61 2 2.38 3 1.20 2 4
- i 1
'g 4 NA NA NA NA NA NA r 4a NA NA NA NA NA NA g
., 5 NA NA NA NA NA NA [ Sa NA NA NA NA NA NA $
- o
! 6 1.69 2 2.40 3 1.25 2 7 NA NA NA NA NA NA 4 7a NA NA NA NA NA NA og Otn 85
- Refer to Table 6.1-1 $
** Margin Factor = Allowable Stress / Actual Stress gg
*** Refer to Figure 7.1-32 m NA = Not Applicable $U i
. - ~ - . - --. - . _ - _ - - - __ - - -- - . .
O O O TABLE 7.1-18 MARGIN TABLE FOR BASEMAT - ADS SRV QUENCHER DISCHARGE (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AFD SSI SEISMIC FORCES) STRESS COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL E UATION* FACTOR SECTION FACTOR SECTION FACTOR SECTION 1 NA NA NA NA NA NA 2 NA NA NA NA NA NA y m y 3 NA NA NA NA NA NA 5 4 1.20 2 2.21 2 1.17 3 U 4a NA NA NA NA NA NA f g 5 1.09 2 2.06 2 1.05 3 [ Sa NA NA NA NA NA NA E
- c 6 NA NA NA NA NA NA 7 1.11 2 2.10 2 1.05 2 7a NA NA NA NA NA NA o i O Oe EE
- Refer to Table 6.1-1 @
** Margin Factor = Allowable Stress / Actual Stress ms
*** Refer to Figure 7.1-32 NA = Not Applicable 1
- - . - - . . - , . . . . - - _ , . . . . - . - = _ - - - . . . - . .. .. . . . . ~ . - . . - - .
O O O TABLE 7.1-19 MARGIN TABLE FOR BASEMAT - ASYMMETRIC (THREE-VALVE) SRV QUENCHER DISCHARGE (WITH PLANT-UNIQUE-FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD , COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION
- FACTOR SECTION FACTOR SECTION FACTOR SECTION
]
- 1 1.47 2 3.81 3 1.59 2 2 2.36 3 3.10 3 2.50 3 y I m 3 .y 3 .l.88 3 2.47 3 1.28 2 4 4
Y 4 1.59 2 2.88 2 1.25 3
$ $ r g
4a NA NA NA NA NA NA 5 1.40 2 2.59 2 1.11 3 M Sa NA NA NA NA NA NA S
, :c l 6- 1.87 3 2.48 3 1.32 2 1
j 7- 1.39 2 2.59 2 1.12 3 i O* 7a NA NA NA NA NA NA Qg 85
- yE "
, z Hd
- Refer to Table 6 .1-1 $g 0"
, ** Margin Factor = Allowable Stress / Actual Stress.
*** Refer to . Figure 7.1-32 NA = Not Applicable i
l O O O TABLE 7.1-20 MARGIN TABLE FOR BASEMAT - SINGLE-VALVE SRV QUENCHER DISCHARCG
.(WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES)
- STRESS COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL
-EQUATION
- FAC'IOR SECTION FACTOR SECTION ' FAC'IOR SECTION i 1 NA NA NA NA NA NA
'2 NA NA NA NA NA NA y
. m y 3 NA NA NA NA NA NA E 4 NA NA NA NA NA NA'
$ h
- 4a- 1.37 2 2.49 2 1.15 3 g i' 5 NA NA NA NA NA NA U Sa 1.20 2 2.25 2 1.02 2 $
6 AN NA NA NA- NA NA
- 7 NA NA NA NA NA NA ,
i 7a 1.20 2 2.26 2 0.96 2 2 8E aR 85
- Refer to Table 6.1-1 $%-
** Margin Factor = Allowable Stress / Actual Stiess g$
*** Refer to Figure ' 7.1-32 e i
NA'= Not Applicable $U 4
O O O TABLE 7.1-21 MARGIN TABLE FOR CONTAINMENT - ALL-VALVE SRV QUENCHER DISCHARGE (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS OMPONENT REINFORCING TENSION _ _ _ CONCRETE COMPRESSION SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION
- FAC'IOR SECTION FACTOR SECTION FACTOR SECTION 1 70.82 1 2.66 1 1.15 6 2 6.54 9 2.63 13 1.15 6 y m
y 3 2.43 12 2.63 13 1.13 6 4 4 NA NA NA NA NA NA w 4a NA NA NA NA NA NA 5 NA NA NA NA NA NA U Sa NA NA NA NA NA NA 6 2.49 12 2.64 13 1.13 6 7 NA NA NA NA NA NA 7a NA NA NA NA NA NA og OM 85
- Refer to Table 6.1-1 %Mz
** Margin Factor = Allowable Stress / Actual Stress y8
*** Refer to Figure 7.1-32 ms NA = Not Applicable
O O O i TABLE 7.1-22 MARGIN TABLE FOR CONTAINMENT - ADS SRV QUENCHER DISCHARGE (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION
- x FAC'IOR SECTION FACTOR SECTION FACTOR SECTION 1 NA NA NA NA NA NA 2 NA NA NA NA NA NA y m
3 NA NA NA NA NA NA E
." i y 4 1.23 12 2.65 13 1.37 6 w
4a NA NA NA NA NA NA 2 5 0.90 12 2.65 13 1.35 6 U Sa NA NA NA NA NA NA 6 NA NA NA NA NA NA 7 0.90 12 2.44 12 1.31 6 7a NA NA NA NA NA NA ~ ^
- Refer to Table 6.1-1 is M
** Margin Factor = Allowable Stress / Actual Stress gb - *** Refer to Figure 7.1-32 mp NA = Not Applicable "
O O 0: TABLE 7.1-23 MARGIN TABLE FOR CONTAINMENT - ASYMMETRIC (THREE-VALVE) SRV QUENCHER DISCHARGE (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS- 3 COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD ^ COMBINATION MARGIN ** CRITICAL *** iMARGIN CRITICAL MARGIN CRITICAL EQUATION
- FACTOR SECTION FACTOR SECTION FACTOR SECTION 1 55.37 12 2.93 1 1.16 6 2 6.76 9 2.65 13 1.16 6 y ,
m , 3 3.22 11 2.65 13 1.12 6 $ ; H 4 1.71 10 2.68 13 1.39 6 4a NA NA NA NA NA NA @ 5 1.02 12 2.68 13 1.34 6 H Sa NA NA NA NA NA NA f ! 6 3.11 11 2.66 13 1.12 6 3 7 0.99 12 2.68 13 1.30 6 7a NA NA NA NA NA NA oz lc
- Refer to Table 6.1-1 M j
** Margin ' Factor = Allowable Stress / Actual Stress mH
*** Refer to Figure 7.1-32 NA = Not Applicable l
- ~,._=-.: -- _
O O O TABLE 7.1-24 MARGIN TABLE FOR CONTAINMENT - SINGLE-VALVE SRV QUENCHER DISCHARGE (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD
' COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL
' EQUATION
- FACTOR SECTION FACTOR SECTION FACTOR SECTION 1 NA NA NA NA NA NA 2 NA NA NA NA NA NA @
m 3 NA NA NA NA NA NA b
." i e 4 NA NA NA NA NA NA E h
= 4a 1.42 12 2.65 13 1.09 6 @
5 NA NA NA NA NA NA U e 5a 0.90 12 1.89 12 1.09 6 g 6 NA NA NA NA NA NA 7 NA NA NA NA NA NA 7a 0.90 12 1.04 12 1.08 6
==
- Refer to Table 6.1-1 @@j
, ** Margin Factor = Allowable Stress / Actual Stress g$
*** Refer to Figure 7.1-32 e-NA = Not Applicable "[
t
1 D O O TABLE 7.1-25 MARGIN TABLE FOR BASEMAT - ADS SRV QUENCHER DISCHARGE (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) i STRESS COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION * \ FACTOR SECTION FACTOR SECTION FAC'IOR SECTION 1 NA NA NA NA NA NA 2 NA NA NA NA NA NA y m 3 NA NA NA NA NA NA b f i e 4 1.28 2 2.26 2 1.13 3 2 0 $
$ 4a NA NA NA NA NA NA @
5 1.15 2 2.10 2 1.02 3 U ! c Sa NA NA NA NA NA NA y <
- 6 NA NA NA NA NA NA 7 1.16 2 2.14 2 0.99 2 7a NA NA NA NA NA NA Og
& tn 88 1 .N 4
- Refer to Table 6.1-1 $
** Margin Factor = Allowable Stress / Actual Stress ,g [
*** Refer to Figure 7.1-32 l NA = Not Applicable
fg > (>> d,o {d TABLE 7.1-26 MARGIN TABLE FOR BASEMAT - SINGLE-VALVE SRV QUENCHER DISCHARGE (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES)
's STRESS COMPONENT REINFORCING TENSION CONCPETE COMPRESSION SHEAR LOAD COMBINATION N. s MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL EQUATION
- FACTOR SECTION FACTOR SECTION FACTOR SECTION 1 NA NA NA NA NA NA 2 NA NA NA NA NA NA y m
3 NA NA NA NA NA NA E Y 4 NA NA NA NA NA NA $ h 4a 1.28 2 2.28 2 1.07 3 g 5 NA NA NA NA NA NA [ Sa 1.14 2 2.09 2 0.97 3 h 6 NA NA NA NA NA NA 7 NA NA NA NA NA NA 7a 1.15 2 2.11 2 0.98 3 O Q@tn e5 25
- Refer to Table 6.1-1 $5
** Margin Factor = Allowable Stress / Actual Stress g[
*** Refer to Figure 7.1-32 NA = Not Applicable
- O O O 1
TABLE 7.1-27 MARGIN TABLE FCR CONTAINMENT - ADS SRV QUENCHER DISCHARGE , (WITH PLANT-UNIQUE FSI, A.2TUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS , COMPONENT REINFORCING TENSION CONCRETE COMPRESSION SHEAR LOAD COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL , iEQUATION* FAC'IOR SECTION FACTOR SECTION FACTOR SECTION 1 NA NA NA NA NA NA 2 NA NA NA NA NA NA y , m i
. 3 NA NA NA NA NA NA $
Y 4 1.21 12 2.83 12 1.37 6 a g < ..- 4a NA NA NA NA NA NA g 5 0.90 12 2.83 12 1.36 6 [ 4 Sa NA NA NA NA NA NA E ,
- o ,
6 NA NA NA NA NA NA 7 0.90 12 2.44 12 1.33 6 7a NA NA NA NA NA NA oz in o l i ~5 . w l
- Refer to Table 6.1-1
** Margin Factor = Allowable Stress / Actual Stress
$U l
*** Refer to Figure 7.1-32
] NA = Not Applicable 1
O O O TABLE 7.1-28 i MARGIN TABLE FOR CONTAINMENT - SINGLE-VALVE SRV QUENCHER DISCHARGE - (WITH PLANT-UNIQUE FSI, ACTUAL MINIMUM CONCRETE STRENGTH, AND SSI SEISMIC FORCES) STRESS COMPONENT REINFORCING TEllSION CONCRETE COMPRESSION SHEAR LOAD
- . COMBINATION MARGIN ** CRITICAL *** MARGIN CRITICAL MARGIN CRITICAL j EQUATION
- FAC'IOR SECTION FACTOR SECTION FACTOR SECTION 1 NA NA NA NA NA NA 2 NA NA NA NA NA NA y w
3 NA NA NA NA NA NA E 4 NA NA NA NA NA NA s 4a 1.22 12 2.41 1 1.39 6 x 5 NA NA NA NA NA NA U Sa 0.90 12 1.89 12 1.37 6 6 NA NA NA NA NA NA 7 NA NA NA NA NA NA op 7a 0.90 12 1.04 12 1.34 6 Q@ 86 95 - 5
- Refer to Table 6.1-1 $g CW
** Margin Factor = Allowable Stress / Actual Stress
! *** Refer to Figure 7.1-32 4 NA = Not Applicable
1
)
AMENDMENT 13 l' YSwa- OCTOBER 1980 I _, r)
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AMENDMENT 13 OCTOBER 1980 60 g._ A0-- a 2 30?- 20-O G = 1000 Keg , 10 .-- G'd = EFFECTlvE OVER-MRDEN PRESSURE p r,0-4 blb-5 12 10 l'10" Sne^n STRAIN (%) WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT O r1GuaE 7.i-2 AVERAGE SHEAR STRAIN VERSUS K 2
AENDMENT 13 OCTOBER 1980 O I l l I 22 - - 20 - 18 - - lG - - i4 _ _ 9 s y 12 - - e io - - 2_ O. 2
<C 8 - -
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O z - o I I I I O.000l O. 0 0l 0.08 0.1 1.0 l0,0 SHE AR STR AIN (%) WM. H.ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT FIGURE 7.1-3 AVERAGE SHEAR STRAIN VERSUS CRITICAL DAMPING
/4 AMENDMENT 13 p .ig
- . ,:- OCTOBER 1980 m
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- 7. 5, 3 7 - ZONE NUMBER u
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- b - l- y T bBASE MAT 13.5' 22.33' =
27.5' = 31.17 ' = 40' " PEDESTAL RADIUS-13.5 FEET CONTAINMENT RADIUS-40.0 FEET POOL DEPTH-22.5 FEET SUBMERGENCE DEPTH-13.5 FEET NOTE: DRAWING NOT h, aCALE. WM. H.11MMER NUCLEAR POWER STATION. UNIT 1 MARKil DESIGN ASSESSMENT REPORT FIGURE 7.1-4 CROSS SECTION OF SUPPRESSION P0OL AND DEFINITION OF SUPPRESSION CHAMBER WALLS' LOADING ZONES l
# 3N0Z - NOI13NO3 DNI3B03 398VH3SIG 3IB13WWAS 7VI1N30b3S INVNOS3B 7V3IdA1 S-L'l 380913 g 180d38 J N3 MSS 3 SSV NOIS30 11 Muyn i11NM 'NOl1Y1S H3 Mod HY373nN H3WWlZ *H *WM NORMAllZED AVERAGE PRESSURE (psig)
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L llNO 'NOl1Y1S H3 mod HV3 DON H3WWiZ 'N 'WM AVERAGE PRESSURE (psig) i I A N o N A m m o i i - , , 9 GW d gE I P - _ pg mN , e EE Q GN c H m o g w - x 9 ' E* n M' 5 5E y 5N o - _
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AMENDMENT l'3 OCTOBER 1980 A REACTOR t 0 -b
.- RE A(. tor 90?PogT '
( WTAIMMENT WALL-* i
-WATER EL. +9 7'- 6'
't v
9 '-6 R +' ft P(t) c'-d V V V V V V / I i f O _
, 4 o'- o" R _ l 4Ld' I
e I (PSI-9) P(t.) A 58.O 35.0 -
!.$4 hh WM. H.ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT l O ricuae 7.i-7 l LOCA VENT CLEARING PRESSURE LOADS ON BASE MAT
l l i AMENDMENT 13 l OCTOBER 1980 G. REACTOR O i a *----REACTOR SUPPORT
$ F CONTAINMENT WALL--*
I
-WATER EL. 497'-6" n
9'-6" R 4'-i l//' V EL.487'- 6"
= =
c - -.-
- 2 733.0 PSI :
0 : V
=
EL. 474'-lO" v n u n u ) )
./
1 4 O'- 0" R 4'- 0" e : : : 1 I l WM. H.ZlMMER NUCLEAR POWER STATION, UNIT 1 MARK ll DESIGN ASSESSMLNT REPORT O' FIGURE 7.1-8 LOCA-VENT CLEARING PRESSURE DISTRIBUTION
AMENDMENT 13 OCTOBER 1980
$sm.
I r3 ' V , i i 1 1 o 1 5 g , '2' 6 Haluu R 5CPPORT :' 3 g , 4T C34TANME NTJ $
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WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK ll DESIGN ASSESSMENT REPORT {~)% L-FIGURE 7.1-9 STRUCTURAL MODEL INCLUDING S0ll
O AMENDMENT 13 OCTOBER 1980
~
l O J - 1 i N U w :
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w _ w 3 2 _ _ i-e _ O _J O d a (r O _ o e e M O O O O D # N (6 ! sd) 38nss3gd WM. H. ZlMMER NUCLEAR POWER STATION, UNIT 1 MARK 11 DESIGN ASSESSM ENT REPORT FIGURE 7.1-10 DRYWELL/WETWELL PRESSURE HISTORY FOR MAIN STEAMLINE BREAK
AMENDMENT 13 OCTOBER 1980 O SOTTOM OF DRYWELL FLOOR 2.3 (P61G) MAYlMUM Pool El DIS. 0
~
6WELL ELEVAT10M E L.497, G. i NORMAL POOL sugFACB, ELEVAilON y 3-EL,487'- C' O TOP OF EL.474'- to. e> Ace MA r
! i o 10 22 (P616) (P610)
WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK ll DESIGN ASSESSMENT REPORT FIGURE 7.1-11 POOL SWELL SYMMETRIC LOAD
AMEN 0 MENT 13 OCTOBER 1980 BOTTOM CIS DEYWCLL Floor. OUNIFo2M LOAD op 23 P6tG APPLIED OVER lao
- sectoe MOftMAL EL.497'- 8 POOL SURFA4E
+-
EL.48Ud HYDR 0 STATIC LOAD [ APPLIED OVER 360 0 TOP OF EL.47M io" 6A5stMAT O 10 2% (PG14) (P6lG) l WM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 MARM 11 DESIGN ASSESSM ENT REPORT FIGURE 7.1-12 P0OL SWELL ASYMMETRIC LOAD
AMENDMENT 13 g OCTOBER 1980
+-REACTOR, $UPPORT ' - -
7 O carrsumsst wAut - ; I WATEft EL 4d-6" 4 >
, 9 '- 6"R y-Ik -
0 >
; P(d p 4 >
P(t) 4 > EL 474'-10" v v v v v / O l - 4 o'- o R 4.Ld] P(t) = 3.75 COS 2nf(t) NOTE: CYCLIC CONDENSATION LOAD OF 3.75 PSI AT 2 - 7 HZ ON PEDESTAL, CONTAINMENT AND BASE MAT UP TO l l SUPPRESSION POOL WATER ELEVATION l WM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 18 DESIGN ASSESSM ENT REPORT l O FIGURE 7.1-13 l LOCA CYCLIC CONDENSATION PRESSURE LOAD l ON BASE MAT CONTAINMENT AND REACTOR l SUPFORT FOR RAMS HEAD
AMENDMENT 13 0CTOBER 1980 O ( REACTOR 4 i b
~ REACTOR SUPPORT CONTAINMENT WALL ~
~ WATER EL. 497'-6" v
; 9'-6" R , ;4'-lh" _
EL. 487'-6" l : : p* psi :
- i _
EL. 474'-10" u n u n n j l
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40'-0" R 4 - 0"=
= -
- FOR DETAILS, REFER SUBSECTION 6.1.2.3 WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT FIGURE 7.1-14 O SPATIAL DISTRIBUTION OF LOCA CONDENSATION OSCILLATION LOAD FOR T-QUENCHER
O O . O
-270 A. UNIFORM LOA 0
+4.8/-4.0 psig, 0*-360 o
o B. ASYMMETf!IC [0AD MAXIMUM +20/-14 psig 000 90 - -
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9 AMENDMENT 13 I k OCTOBER 1980 C frequency 2 20 Hz. e D (n m o A Anv a g v \j v ~
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. - l 0 .25 .50 TIME-SEC WM. H.ZIMMER NUCLEAR POWER STATION UNIT 1 MARK 11 DESIGN ASSESSM ENT REPORT FIGURE 7.1-16 1 CHUGGING LOAD-TIME HISTORY
AMENDNENT 13 OCTOBER 1980
& REArTbg, 0
REACTOR, doWTAIMMEMT SUPPORT
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WM. H. IBMMER NUCLEAR POWER STAT'ON. UNIT 1 MARK Il DESIGN ASSESSMENT REPORT FIGURE 7.1-17 DRYWELL FLOOR ANALYTICAL MODEL
AMENDMENT 13 OCTOBER 1980 CD a O i ) I I i . I 8 l l
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WM. H.ZIMMER NUCLEAR POWER STATION, UNIT I MARK ll DESIGN ASSESSM ENT REPORT FIGURE 7.1-18 i th CIRCUMFERENTIAL VARIATION OF MOMENT IN DRYWELL FLOOR DUE TO CONCENTRATED RADIAL MOMENT APPLIED AT RADIUS 22'-3" w . . - , , , -
--+.---p g -
t AMENDMENT 13 a OCTOBER 1980
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.LN3ldlL-130] 1N3 WOW WM. H.ZIMMER NUCLEAR POWER STATI N. UNIT 1 MARK 18 DESIGN ASSESSMENT REPORT FIGURE 7.1-19 RADIAL VARIATION OF MOMENT IN DRYWELL FLOOR DUE TO CONCENTRATED RADIAL MOMENT APPLIED AT RADIUS 22'-3"
i e a AMENDMENT 13
._ . OCTOBER 1980
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4 AMENDMENT 13 OCTOBER 1980 l I C) ' if L d - [ [, I l
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.LN31]IJ.d30] 1N3WWi WM. H.ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT FIGURE 7.1-21 RADIAL VARIATION OF MOMENT IN DRYWELL FLOOR DUE TO CONCENTRATED CIRCUMFERENTIAL M0 MENT APPLIED AT RADIUS 22'-3"
AMENDENT 13 OCTOBER 1980
-52.0 a
50-- 1 I 40-H LOAD ON EACH DO W NC OME R -8.B KIPS Z g LOAD DIREC TION- R ANDOM y 30-Op ,, 21 x
._J Y ko 20-
<C Cr O
IO-i t t , , 2'O 40 $0 8'O 8'8 NUMBER OF LOADED DOWNCOMERS WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 MARh la DESIGN ASSESSM ENT REPORT FIGURE 7.1-22 (g CRITICAL DESIGN SECTION 2 VARIATION U OF RADIAL MOMENT WITH NUMBER OF LOADED DOWNCOMERS i
AMENDMENT 13 OCTOBER 1980 40-- O LOAD ON EACH DOWNCOMER - PER DFF R (FIG.4-loa) LOAD DIRECTION-R ANDOM p 30-- Z w h 1 0 9 2 ' x 20-
-J ?
U O' io_ a O 20 40 60 80 88 NUMBER OF LOADED DOWNCOMERS WM. H.ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK II DESIGN ASSESSMENT REPORT FIGURE 7.1-23 O CRITICAL DESIGN SECTIW1 2 VARIATION OF RADIAL MOMENT WITH NUMBER 0F DOWNCOMERS LOADED PER DFFR f
AMENDMENT 13 OCTOBER 1980 l I 1
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AMENDMENT 13 OCTOBER 1980 t.ys@ .w i .. (m-) .w rs
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ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 7.2 CONTAINMENT WALL AND BASEMAT LINER ANALYSIS 7.2.1 Basemat Liner 7.2.1.1 Description of Liner and Liner Stiffeners The base liner plate consists of a 1/2-inch thick carbon steel plate of SA516 Grade 60 material, with 0.05-inch cladding of SA240, Type 304 stainless steel material. The maximum anchor spacing of the base liner plate is 3 feet 0 inch with the anchorage consisting of WT4 x 10 structural steel sections. Refer to Figure 7.2-1 for the basemat liner detail. 7.2.1.2 Loads for Analysis The loads acting on the liner plate are defined in the FSAR. In addition, loads described in Chapter 5.0 of this report are included. These loads are divided into two main categories, (1) self-limiting loads, and (2) mechanical loads, each of which is discussed below. Self-Limitino Loads Strains in the liner plate and anchorage system are imposed by dead loads, post-tensioning, creep, shrinkage, and thermal effects. Mechanical Loads These loads are due to SRV discharge within the suppression chamber area. Due to the pulsating nature of the load, the liner plate and anchorage system will experience stress reversals. For analysis, it has been assumed that there are 3886 actuations (DFFR) with 10 stress reversals per actuation. 7.2.1.3 Load Combinations The liner and anchorage systems are designed for the load com-binations listed in Table 6.1-1, except that all load factors are taken as unity. 7.2.1.4 Acceptance Criteria The strains in the liner plate due to self-limiting loads are limited to the allowable values specified in Table CC-3720 of the ASME Boiler and Pressure Vessel Code, Section III, Division 2, l and the displacements of the liner anchorage are limited to the displacement values of Table CC-3730-1 of the ASME B&PV Code, - Section III, Division 2. Primary membrane stresses due to mechanical loads in the liner p) plate and anchorage system (weld and anchor) are checked (- according to. Subsection NE-3221.1 of the ASME B&PV Code, Section 7.2-1
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (]) III, Division I. Primary plus secondary membrane plus bending stresses are checked according to Subsection NE-3222.2 of the same code. Fatigue strength design stress is based on Subsection NE-3222.4 of Section III. Allowable design stress intensity values, design fatigue curves, and '.m'erial properties used conform to Subsection NA Appendia I of C'e ASME B&PV Code, Section III,. Division I. Subsection NB-33s s of the ASME B&PV Code, Section III, Division I is used to obtain a fatigue strength reduction factor of 4.0 for the fillet weld attachment of the containment wall liner plate and anchorage system. 7.2.1.5 Analysis The hydrostatic pressure head on the basemat is 9.8 psi and the maximum uplift pressure load due to SRV alone is 3.7 psi. Since the negative load due to SRV discharge is more than balanced by the pressure head or water in the suppression chamber, the base liner plate does not experience any negative or uplift pressure load at any time during SRV actuation. Therefore, there are no flexural stresses induced in the basemat liner. The maximum net uplift pressure that the basemat liner can withstand is 11.5 psi acting upward. Therefore, the basemat liner has the capability to carry an SRV negative pressure of 21.3 psi including the hydrostatic head, which is 245% of the o design SRV pressure load. O 7.2.2 Containment Wall Liner 7.2.2.1 Description of Liner The suppression chamber wall liner consists of a 1/4-inch stainless steel plate of SA240, Type 304 up to elevation 500 feet 0 inch. Above elevation 500 feet 0 inch the liner is of carbon steel SA516, Grade 60 material. A 3 x 2 x 1/4-inch angles are welded to this plate intermittently with a 1/4-inch fillet weld at 4 inches every 12 inches center-to-center spacing. Refer to Figure 7.2-2 for the containment liner detail. 7.2.2.2 Loads for Analysis The loads for analysis are described in Subsection 7.2.1.2. 7.2.2.3 Load Combinations The load combinations are described in Subsection 7.2.1.3. 7.2.2.4 Acceptance Criteria - The acceptance criteria for the containment wall liner are de-scribed in Subsection 7.2.1.4. O v 7.2-2 l
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 7.2.2.5 Analysis [} To study the response to the liner plate due to the SRV blowdown loads, a dynamic analysis using finite element idealization was performed. Since the liner plate experiences bending between anchor supports predominantly in one direction, a two-dimensional representation is used for the dynamic analysis. Several beam elements are used to represent the flexibility of the liner plate between two anchor locations. The ends of the model which represent the anchor supports are assumed to be fixed against both in-plane rotation and displacements. In addition, a non-linear stiffness matrix representation is used to simulate the stiffness of the concrete to resist compressive loads only, with no resistance towards tensile or negative SRV loads. The time-pressure history of the oscillating air bubble, which has approximately 10 negative pulses per actuation, is used as the input forcing function to the finite element model. The results of the dynamic analysis show that the dynamic load factor is approximately equal to 1.0. The liner plate can, therefore, be analyted for SRV blowdown load by using a static solution procedure. The suppression chamber wall liner has the capability to carry a SRV negative pressure of 16.5 psi at the most critical location and highest elevation (no credit for hydrostatic pressure), which is 190% of the design SRV pressure load. The summary of stresses and strains in the containment wall liner plate and anchorage system are shown in Tables 7.2-1 through 7.2-4. It is apparent from the tables that the safety margin for each category of mechanical and self-limiting loads is greater than 1.0. Therefore, the suppression chamber wall liner and basemat liner plate and anchorage system are acceptable. s_- 7.2-3
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 TABLE 7.2-1 0.
SUMMARY
OF CONTAINMENT WALL LINER PLATE STRESSES / STRAINS FOR ALL SRV CASES (RAMS HEAD) I MECHANICAL LOADS (Suction Loads) ACTUAL ALLOWABLE STRESS STRESS OR OR STRESS USAGE USAGE SAFETY CATEGORY FACTOR FACTOR MARGIN Primary (Pb) Bending 3.600 ksi 1.5 S m = 30 kai 8.33 Secondary (0) 49.53 ksi 3.0 S m = 60 ksi 1.21 Peak (F) 0.04 1.0 25.0 II. SELF-LIMITING LOADS () STRAIN ACTUhL STRAIN ALLOWABLE STRAIN SAFETY CATEGORY (in/in) (in/in) MARGIN Self-limiting .001 .002 2.0 l (:) - 1 4 7.2-4
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 TABLE 7.2-2
SUMMARY
OF CONTAINMENT WALL LINER ANCHORAGE LOAD / DISPLACEMENT FOR ALL SRV CASES (RAMS HEAD) I MECHANICAL LOADS (Suction Loads) ACTUAL ALLOWABLE STRESS STRESS OR OR STRESS USAGE USAGE SAFETY CATEGORY FACTOR FACTOR MARGIN Weld Primary Membrane (P ,) 0.340 ksi h S,= 10 ksi 29.41 Peak (F) 0.04 1.0 25.0 Angle Primary Membrane (P ) 0.120 ksi S,= 13.9 ksi 115.83 II MECHANICAL LOADS (Suction Loads) ACTUAL ALLOWABLE LOAD LOAD STRESS OR OR SAFETY CATEGORY STRESS STRESS MARGIN Concrete Diagonal Tension Failure 30.0 lbs/in 860.0 lbs/in 28.67 III SELF-LIMITING LOADS ACTUAL ACTUAL STRESS DISPLACEMENT DISPLACEMENT SAFETY CATEGORY (in) (in) MARGIN Anchorage System .015 0.45 3.0 7.2-5
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 e TABLE 7.2-3 i
SUMMARY
OF CONTAINMENT WALL LINER PLATE STRESSES / STRAINS FOR ALL SRV CASES (T-QUENCHER) I MECilANICAL LOADS (Suction Loads) ACTUAL ALLOWABLE STRESS STRESS OR OR STRESS USAGE USAGE SAFETY CATEGORY FACTOR FACTOR MARGIN Primary (Pb} Bending 4.752 ksi 1.5 S,= 30 ksi 6.31 Secondary (0) 44.403 ksi 3.0 S,= 60 ksi 1.35 Peak (F) 0.004 1.0 25.0 II SELF-LIMITING LOADS () STRAIN ACTUAL STRAIN ALLOWABLE STRAIN SAFETY CATEGORY (in/in) (in/in) MARGIN Self-limiting .001 .002 2.0 I l ('/) 7.2-6
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ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 TABLE 7.2-4
SUMMARY
OF CONTAINMENT WALL LINER ANCHORAGE LOAD / DISPLACEMENT FOR ALL SRV CASES (T-QUENCHER) I MECHANICAL LOADS (Suction Load) ACTUAL ALLOWABLE STRESS STRESS OR OR STRESS USAGE USAGE SAFETY CATEGORY FACTOR FACTOR MARGIN Weld Primary Membrano. (P ,) -- ksi S = 10 ksi ---- Peak (F) 0.04 1.0 25.0 Angle Primary Membrane (P ,) -- ksi S,= 13.9 ksi ---- II MECHANICAL LOADS (Suction Loads) ACTUAL ALLOWABLE LOAD LOAD STRESS OR OR SAFETY CATEGORY STRESS STRESS MARGIN Concrete Diagonal Tension Failure lbs/in 860.0 lbs/in ---- III SELF-LIMITING LOADS ACTUAL ALLOWABLE STRESS DISPLACEMENT DISPLACEMENT SAFETY CATEGORY (in) (in) MARGIN Anchorage System .015 0.45 3.0 l 7.2-7 1 J
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1 l J 2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () 7.3 OTHER STRUCTURAL COMPONENTS A reassessment for the additional effects of pool dynamic loads was made for steel and concrete structures in the reactor building including steel framing and galleries, cable pan hangers, conduit hangers, HVAC duct hangers, and concrete slabs, beams and shear walls. The design and analysis procedure for the reassessment was the same as described in Subsections 3.8.3 and 3.8.4 of the ZPS-1 FSAR for the original design of the plant. Load combinations described in Section 6.3 of this report were used for the assessment. 7.3.1 Downcomers and Downcomer Bracing 7.3.1.1 General Description There are 88 downcomers anchored in the drywell floor. The downcomers are also connected to the containment structure by horizontal bracing at elevation 496 feet. Figure 7.3-1 shows the downcomers in the suppression pool, and Figure 7.3-2 shows the layout of the horizontal bracing. The downtomers and downcomer bracing are subjected to static and dynamic loads due to normal, upset, emergency, and faulted plant operating conditions. The loadings cases were obtained from the (,_s) DFFR and are identified in detail in Subsection 7.3.1.2. The loading combinations are explained in Subsection 7.3.1.3. The design limits are identified in Subsection 7.3.1.4, and the analytical methods are presented in Subsection 7.3.1.5. 7.3.1.1.1 Downcomer Properties The following are the properties of the downcomers:
- a. outside diameter - 25.00 inches;
- b. wall thickness - 0.500 inch;
- c. weight per unit length - 131 lb/ft;
- d. material - SA-516, Grade 60; and
- e. damping coefficient - 3%.
7.3.1.1.2 Bracina Properties The following are the properties of the bracing:
- a. outside diameter - 8.625 inches;
_ b. wall thickness - 0.875 inch; s_/ 7.3-1
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
- c. weight per unit length - 72.42 lb/ft; and
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- d. material - A-106 Grade B.
7.3.1.1.3 Connection Properties The following are the properties of the connections:
- a. Connection of the bracing to the downcomer is accomplished through gusset plates and stiffened pipe sleeves,
- b. The gusset plates are 3/4 inch thick, A-588 Grade A or B steel.
- c. The stiffened pipe sleeves are composed of 3/4-inch tnick, 27-inch OD pipe 3 feet long and two 1.5-inch thick, 39-inch OD stiffened rings, as shown in Figure 7.3-3, 7.3.1.2 Loads for Analysis The individual loads affecting the downcomers and downcomer bracing are identified below:
s a. Dead Weight This is a static force due to the weight of the downcomers, bracing members, connections, and live loads.
- b. Pressure Load The pressure differential between the drywell and suppression chamber atmospheres produce loads on the downcomer wall since it acts as a pressure retaining boundary during a loss-of-coolant accident.
- c. Operating Basis Earthquake (OBE)
The OBE causes vibratory motions of the building structure which induce dynamic forces on the downcomers. The OBE also causes water sloshing inside the suppression chamber. The drag and inertia forces of these oscillations will produce a dynamic loading on the submerged portion of the downcomer.
- d. Safe Shutdown Earthquake (SSE)
The SSE causes the same type of dynamic loads on the downcomer as described in item c. However, the (~ magnitude of the loads caused by the SSE is greater \ >) than those caused by the OBE. 7.3-2
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 { (])' e. Loss-of-Coolant Accident (LOCA) Loads i i The following three cases for LOCA loads were considered for analysis:
- 1. As discussed in Section 5.3, during the initial phases of a LOCA, high-steam flow rates through the downcomer produce condensation oscillation load on the downcomer. During the low-steam flow rates, there is a random dynamic chugging lateral load acting on the submerged portion of the downcomer which can be represented by a load applied at the exit of the downcomer. LOCA thernal load will also apply on the downcomers.
- 2. Following the LOCA, the downcomers will experience a dynamic loading due to its response to:
a) the vertical acceleration produced in the drywell floor by water jet impingement on the containment basemat during the downcomer clearing process, and b) the cyclic chugging load on the containment g structure. V
- f. Safe; ./ Relief Valve (SRV) Discharge Dynamic Load The following two cases of SRV discharge are considered for design purposes:
- 1. resonant sequential symmetric discharge of all 13 valves, and
- 2. subsequent actuation discharge of a single valve.
The air discharge from the SRV vent lines forms high-pressure i bubbles which expand and contract periodically until they rise to the pool surface. The bubble oscillations impose time varying lateral forces on the submerged surface of the downcomer. The downcomer will also experience thermal loads and dynamic loads due to its response to the base excitation produced in the building resulting from the forced vibration of the containment structure. induced by The forced of the effects vibration of the the bubble containment oscillations on structures the is submerged surfaces of the wetwell walls. 7.3.1.3 Desian Load Combinations (~)/
'- The downcomer loads defined in Subsection 7.3.1.2 were combined for upset, emergency, and faulted conditions as described below 7.3-3
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () in accordance with Table 7.3-1 and Subsection NC-3600 of the ASME Boiler and Pressure Vessel Code, Section III. The resulting stresses were combined with the long duration load stresses (loads a and b of Subsection 7.3.1.2) on an absolute sum basis. The combinations shown in the following sections were most critical. 7.3.1.3.1 SRV Actuation Load Combinations SRV actuation load combinations are as follows:
- a. For the upset condition, the stresses due to loads a, b and c were combined with the stresses due to load f.
- b. For the emergency condition, the stresses due to loads a, b and c were combined with the stresses due to loads e and f.
7.3.1.3.2 LOCA Associated Load Combinations For the emergency condition, the stcesses due to a and b were combined with the stresses due to loads d and e. 7.3.1.4 Acceptance Criteria n (_) 7.3.1.4.1 Acceptance Criteria for Downcomers The stresses within the downcomer are considered acceptable if they satisfy the ASME Boiler and Pressure Vessel Code, Section III, Subsection NC-3600. The allowable stress S was obtained from Table 1.7-1, Section III, Appendix I for material SA-516, Grade 60 at a design temperature of not exceeding 4000 F. The primary stress intensity includes the primary membrane stresses plus the primary bending stresses. The limits of these stresses depend upon the loading conditions as follows:
- a. The limit of stresses under normal condition: 1.0S.
- b. The limit of stresses under upset condition: 1.2S.
- c. The limit of stresses under emergency: 1.8S.
- d. The limit of stresses under faulted condition: 2.4S.
7.3.1.4.2 Acceptance Criteria for Downcomer Bracing The stresses within the downcomer bracing are considered acceptable if they satisfy the ASME Boiler and Pressure Vessel Code, Section III, Subsection NF-3300. At design temperature, fs) the allowable stresses in tension or bending depend upon the yield stress Sy as follows: 7.3-4
l l 2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () a.
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The allowable stresses under normal condition: 0.6Sy .
- b. The allowable stresses under upset condition: 0.6S . y
- c. The allowable stresses under emergency condition:
(1.33) (0.6)S y. 7.3.1.4.3 Acceptance Criteria for Connections The stresses within the connections, including pipe sleeves, gusset plate and stiffeners of material A-588 are considered acceptable if they satisfy the ASME Boiler and Pressure. Vessel Code, Section III, Subsection NF-3330. The allowable stresses in tension or bending depend upon yield strength S y as follows:
- a. The allowable stresses under normal condition: 0.6Sy .
- b. The allowable stresses under upset condition: 0.6S .y
- c. The allowable stresses under emergency condition: '
(1.33) (0.6)S y. 7.3.1.4.4 Acceptance Criteria for Welded Joints The allowable stress limits in welds in linear type supports (~/ 3 shall not exceed the values set forth in Table NF-3292-1-1 of the s_ ASME Boiler and Pressure Vessel Code, Section III. Under LOCA loads, the stress limit may be increased by one-third over the values given in Table NF-3292-1-1. 1.3.1.5 Analysis The downcomers and downcomer bracing were analyzed for all loads mentioned in Subsection 7.3.1.2 using the PIPSYS computer program. Static and dynamic analyses were performed to obtain the response of the structural system. 7.3.1.5.1 Static Analysis Static analysis tecnhiques were used to determine the stresses due to loads of a steady nature and/or dynamic loads having equivalent static loads. 7.3.1.5.2 Dynamic Analysis The downcomers and downcomer vent bracing were represented by a lumped mass mathematical model. All the masses were lumped at the node joints, which were interconnected by flexible beam elements. The containment and reactor pedestal were considered to provide rigid support to the bracing since the pool boundaries are extremely stiff compared to the bracing system. L.) 7.3-5
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 n A/ The inertia effects of the water surrounding the submerged portion of the structural elements were simulated by the addition ' of a water mass (Reference 1) equivalent to the displaced volume of the structural elements. The mass of water inside the submerged portion of the downcomers was also considered in the model for.all dynamic loadings except the loads associated with the LOCA. For these loads, the water had been vented from the i downcomer and, therefore, it was not included in the model. l Depending upon the form of the loading function, both response l spectrum and forced vibration methods were used to obtain the
; structural response.
() , I C) 1 ( 7.3-6
l 2PS-1-MARK II DAR 73ENDMENT 13 l , OCTOBER 1980 10 7.3.2 aeferences < , 1. Sir Horace Lamb, " Hydrodynamics," Sixth Edition, Dover Press, New York, 1945. ,
- 2. ASME Boiler and Pressure Vessel Code, Section II, Subsections NC, NF, and appendices of Division 1, 1978 edition, including the Winter Addenda.
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O O O TABLE 7.3-1 LOAD COMBINATIONS AND ACCEPTANCE CRITERIA FOR DOWNCOMER AND DOWNCOMER BRACING LOAD NRC LOAD COMBINATION RAMS HEAD CASE T-QUENCHER ASME STRESS (NUREG-0487) DESIGN-BASIS DESIGN-BASIS CRITERIA 1 N+SRV e e in + X B ZPS FSAR Appendix I 2 N+SRV +0BE X N+SRV +OBE B Sm 3 N+SRV + + y X + C h Y m 4 N+SRVADS+ A( A) + G C E 5 N+SRV E+IBA(SBA) M ADS N+SRV +OBEEHUG C e 6 N+SRVADS+SSE+IBA(SBA) N+SRV +SSE+ CHUG C ] 7 N+SSE+DBA N+SSE+CO C 8 N N A g$ 9 N+OBE 95 N+OBE B pk 10 N+SRV +SSE+DBA - X CONTAINMENT STRUCTURE ONLY JUSTIFICATION PROVIDED BY GE. I .l
AMENDMENT 13 OCTOBER 1980
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ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () 7.4 REACTOR PRESSURE VESSEL HOLDDOWN BOLTS An assessment of the RPV holddown bolts for the annulus pres- I surization load is described in this section. I 1 Table 7.4-1 gives the breakdown of the maximum axial and shear , forces in the holddown bolts resulting from dead load (DL), design-basis earthquake (DBE), annulus pressurization (AP) and ' pipe break jet reaction (R ). I g I The following extremely conservative governing load combinations , were considered in computing the design loads on the holddown , bolts: I
- a. DL + 1.4 (AP + R ). I
- b. DL + (AP + R ) + DBE. ,
AISC Code allowable stresses defined in Table 3.8-9 of the ZPS-1 FSAR were used in calculating the margin factors. Table 7.4-2 gives the design forces for governing load combinations and the 8 corresponding margin factors. I As can be seen from Table 7.4-2, the RPV holddown bolts have (~) ample safety margin, even for the extremely conservative load ' (J combinations concidered in the ass?ssment. ' I 7.4-1
O O O TABLE 7.4-1 MAXIMUM FORCES ON RPV HOLDDOWN BOLTS AXIAL FORCE
- SHEAR FORCE LOAD (kips / BOLT) (kips / BOLT)
Dead Load (DL) - 24.00 ----- Design -Basis Earthquake (DBE) + 25.64 10.87 Annulus t4 m Pressurization (AP) + 15.07 29.34 T' e ." Jet Reaction (RA) - 15.26 25.56 I h w m U B w Og OR oz to o e co s
- Positive value for tension and negative value for compression. "
j ZPS-1-MARK II DAR AMENDMENT 13 1 OCTOBER 1980 1 I TABLE 7.4-2 i DESIGN FORCES AND MARGIN FACTORS j OF RPV HOLDDOWN BOLTS i j GOVERNING TENSION SHEAR MARGIN i LOAD COMBINATION (kip / BOLT) _( _ kip / BOLT) FACTOR DL+1.4 (AP+Rg) 18.47 72.66 1.58 DL+(AP+Rg)+DBE 31.97 62.77 1.66 a d i i O 4 i i j ;
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2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () 7.5 BALANCE-OF-PLANT (BOP) PIPING AND EQUIPMENT 7.5.1 BOP Piping 7.5.1.1 Evaluation of Bounded Load Combination (Rams Head Definition) 7.5.1.1.1 Loads and Load Combinations Evaluated The piping analysis for ZPS-1 has been completed based on the rams head design basis. The rams head design basis loads are shown in Table 7.5-1. As the table shows, several assumptions were made in the rams head design basis such that the rams head design basis is governing over loads and load combinations which included SRV (T-quencher) and LOCA loads. The purpose of this section is to compare the rams head design basis against the T-quencher and LOCA loads that were not con-sidered in the plant design basis and to show that the rams head design basis is still the governing design basis. In Table 7.5-1, the first column, "NRC Load Combinations for Mark II Plants," represents the ten load combinations that were given in Attachment A of the NRC Acceptance Criteria (NUREG-0487). rm The second column, " Rams Head Design Basis," represents the loads
's_) and load combinations to which 2PS-1 is currently designed. The table shows assumptions made when a rams head design basis load combination was assumed to govern over an NRC load combination.
The third column shows the loads and load combinations which included T-quencher and LOCA loads that were assessed in order to determine the impact on the rams head design basis. The last column shows the acceptance criteria used; these criteria are consistent with the NRC Acceptance Criteria. 7.5.1.1.2 Drywell Pipino In the assessment of the drywell piping, a detailed study was made to evaluate the adequacy of the rams head design basis a-gainst the NRC Acceptance Criteria (NUREG-0487) using T-quencher loads. The assessment included the evaluation of 13 out of 25 major piping subsystems inside the drywell. The remaining drywell piping was either symmetrical to those piping subsystems assessed or was governed by operating transients so that the increase in the SRV and LOCA inertia loads was insignificant. A few small-diameter (< 2-inch) piping subsystems inside the dry-(~') well were not included in the assessment, and the assessment did not include the impact on instrumentatio,n lines. 7.5-1
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 198C () The detailed results of the assessment were presented at the December 5, 1979 meeting with the NRC. In general, the results indicated that:
- c. For the load combinations currently committed to the NRC, the support loads tended to decrease when comparing rams head design-basis loads to the NRC Icad combinations using T-quencher loads,
- b. The loads that did increase were all associated with small-diameter (< 4-inch) piping systems. The load increases were all within the rating of the snubber load capacity.
- c. The load increases were primarily due to the increases in the lower frequency range of the response spectra.
- d. The impact on the piping stresses was insignificant.
- e. The impact of the empirical limiting CO load was significant on the piping systems. The reason for the impact was due to the large amplitudes in the higher frequencies of the CO empirical limiting response spectra.
In addition, the load combination with the CO empirical limiting load included the effects of the SRV loads. The overstress of the piping due to the empirical CO load can be resolved by a more refined analysis, using either actual material properties or the SRSS method of load combination. 7.5.3.1.3 BOP Piping Inside the Reactor Buildino The detailed assessment of the piping systems outside containment (within the reactor building) was completed for the effect of the T-quencher and LOCA loads on the rams head design basis. The loads and load combinations that were asses' sed are given in Table 7.5-1. The detailed results were presented to the NRC at the December 5, 1979 meeting. The results are summarized as follows:
- a. Piping and supports outside cont *ainment are adequately designed to the NRC Acceptance Criteria using the currently defined suppression pool loads
(]) with the T-quencher discharge device. 7.5-2 1
2PS-1-MARK II DAR AF3NDMENT 13 OCTOBER 1980 i () b. The impact of the empirical limiting CO load had a localized effect on the piping and supports connected to the outer suppression pool wall.
- c. No piping overstress was found for all loads and load l combinations (including the empirical CO load) using !
the absolute sum method of combination. In all, 3,199 locations were evaluated. 7.5.1.1.4 Wetwell Pipino Due to the direct hydrodynamic loading from SRV discharge and LOCA, all wetwall piping was upgraded. Assessment of the rams head design basis was not performed. The design of the wetwall piping and piping supports is based on the bounding SRV T-quencher and LOCA loads outlined in Chapter 5.0 and the load combinations shown in Chapter 6.0. 7.5.1.2 Impact of Change to T-Ouencher Discharce Device The impact on piping systems of the change from a rams head to a T-quencher discharge device has been shown to be minimal. In general, only those piping subsystems whose fundamental mode frequency is less than 7 hertz were impacted. Those piping systems tended to be small-diameter (< 4-inch) piping whose loads were'relatively small. [) , The increases in loads were still within the capacity of the 3 restraints. l Piping overstress was shown to be almost negligible, and those j locations where an overstress condition did exist could be j qualified with a more refined analysis. J The detailed results of the T-quencher reevaluation report are included in Appendix H of this document. 7.5.1.3 Impact of SRV T-Quencher and LOCA on Rams Head Desion Basis In this section the assessment of the impact of the new suppression pool loads on the rams head design basis is summarized. The results of the assessment showed various degrees of impact on the rams head design for various load combinations. This assessment provided the basis for the use of the 1.33 factor for early release of hardware for procurement prior to completion of analysis for the Zimmer empirical load. The impact of three , bounding load combinations were investigated. The three load combinations are:
- a. N + SSE + CO (DFFR) 7.5-3
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () b. N + SSE + CO (EMPIRICAL) + SRV gggTO and
- c. N + SSE + CHUGGING + SRV ALL TO The purpose of this section is to determine how the above three load combinations affect the rams head design basis of N + DBE* + SRV ALLRH
*DBE = 1.875 x OBE for piping supports both inside and outside the drywell and to determine a bounding value for those supports loads which are not governed by the rams head design in order to release piping supports for early procurement and continue field construction.
Fron. Subsection 6. 4. 2, the following bounding combinations were investigated:
- a. N + SSE + CO (DFFR)
- b. N + SSE + CO (EMPIRICAL) + SRVALL/ASYTO and
- c. N + SSE + CHUGGING + SRVggg7ggyTO
{} The loads were combined using the absolute sum method. The above load combinations were analyzed for 43 sample piping systems throughout the reactor building including subsystems both inside and outside the drywell. The support-loads were tabulated and were compared with the corresponding support load for the rams head design basis. L The results are summarized in six histograms showing the percent change in support loads. The six histograms depict the load change for the following load combinations: Figure 7.5 CO (DFFR) Inside Containment Figure 7.5 CO (EMPIRICAL) Inside Containment
, Figure 7.5 Chugging Inside Containment Figure 7.5 CO (DFFR) Outside Containment Figure 7.5 CO (EMPIRICAL) Outside Containment
- Figure 7.5 Chugging Outside Containment The results are also shown in Table 7.5-2, " Impact on Piping 73 Support - ABSUM." Table 7.5-3 summarizes numerically for the t
;'~j three load combinations the quantity of restraint increases and the percentage change.
i 7.5-4 1 l
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 A (_,! Summary Upon evaluating the six histograms shown in this section, it was concluded that the major impact appeared to be due to the CO (EMPIRICAL) load combination inside containment. A large percentage (40%) of restraints had load increases over the rams head design basis with magnitudes increasing up to 150%. The other two load combinations, CO (DFFR) and Chugging, contain load increases, but most of the changes in magnitudes were small (20% to 30%). The summary of results can be depicted in terms of percentage of restraints that are bounded. Table 7.5-2 shows the percent of restraints bounded when various leve!s of multiplication factors are used to increase the rams head design loads. It was also found that the subsystems outside containment with the largest number of restraint increases were found to be affected by the chugging horizontal load generated at the suppresion pool wall at elevation 497 feet. The CO (EMPIRICAL) also affects the low frequency piping in the reactor piping. However, the impact on restraint design is minimal since the low frequency systems are small diameter and /~ are low magnitude support loads. V} A final confirmatory analysis of all safety-related piping will be completed to verify the adequacy of the assessment results. 7.5.2 Balance-of-Plant Equipment The qualification of balance-of-plant (BOP) safety-related equip-ment has evolved in three stages. Each stage is a different set of load combinations for which equipment adequacy has been evaluated. The original design basis was to use the normal plus seismic loads outlined in Chapter 5.0. After identification of pool dynamic loads, the equipment was requalified to demonstrate and document adequacy for these additional new loads. When it was decided to change the quencher device from the rams head to the T-quencher, an assessment of the impact of this change as well as the Zimmer Empirical CO load was made and the results presented to the NRC Staff on December 5, 1979. This assessment showed that the majority of the BOP safety-related equipment will be adequate as defined in Table 7.5-1. Since the meeting of December 5, 1979, a requalification program has begun to document the results of the assessment and to perform additional work required to demonstrate adequacy. This requalification program will evaluate equipment adequacy using 7s the load combinations and acceptance criteria described in (~) Subsection 6.4.3. The loads will be combined using the absolute sum (ABS) method. However, if for a particular piece of l l 1 7.5-5 l l l
i 2PS-1-MARK.II DAR AMENDMENT 13 OCTOBER 1980
.O equipment the resuite de not s ti fy the ecceptence criterie, the i square root of the sum of the squares (SRSS) method will be used before equipment modifications will be made. Cases where the SRSS method will be used in lieu of the ABS method will be identified.
1 0 8 O 1 1 l O 7.5-6 l i . . . . - . . . . , . . , - . . - . . . ... .- - . - . - .
. _ _ . .. . - ~ . - . .. - - _ _ _ -- ._ _.
TABLE 7.5-1 LOAD COMBINATIONS AND ACCEPTANCE CRITERIA LOAD NRC LOAD COMBINATIONS- RAMS HEAD DESIGN T-QUENCHER ACCEPTANCE CASE FOR MK II PLANTS BASIS ASSESSMENT CRITERIA
- 1. N+SRV Governed by Governed-by B X
(N+SRV +OBE) (N+SRVM /ASY+0BE) 2- N+SRVX+0BE N+SRVgg3+0BE B N+SRVgggjggy+0BE 3 .N+SRVX+SSE- N+SRVgLg+~1.8750BE Governed by 'C* (N+SRVg +SSE+ CHUG) u) 4 Governed by Governed by C* N+SRVADS+IBA (SBA) 8 y (N+SRVg +1.8750BE)- (N+SRVg + CHUG +SSE) Y Y 5 Governed by Governed by C* w N+SRVADS+0BE+IBA(SBA) * ! (N+SRV +1.8750BE) N+SRVgLL+SSE+CWG H 6 Governed by N+SRV C* N+SRVADS+SSE+IBA (SBA) g+SSE+CO ) (N+SRV +1.8750BE) N+Sp g +SSE+CHUGJ -
'7 N+SSE+DBA 2 N+kSSE+AP N+k(SSE) +(AP) ; N+SSE+CO' C*
8 N N N A 9 N+OBE Governed by Governed by B g2
-(N+SRV +OBE) (N+SRVgggjggy+0BQ gg tn '
10 N+SRVX+SSE+DBA Containment only Containment only C (GE Justification (GE Justification w% g w i submitted) submitted) $U
- Service Level D was specified for piping not required to function. I i
TABLE 7.5-2 ! IMPACT ON PIPING SUPPORTS ABSUM (CO (DFFR) CO (EMPIRICAL) CHUGGING g LOAD COMBINATION LOAD COMBINATION LOAD COMBINATION
.% Of % Of % Of-Total No. Restraints Restraints Restraints of No.'of with No. of with No. of with
, Restraints Restraint ' Increase Restraint Increase Restraint Increase
-Evaluated Increases Load Increases Load Increases Load
<- N
- 4. Inside y I-Containment 269 25 9% 108 12 40% 4% $
N Ou'. ide [. Containment 861 3 0.3% 180 21% 214 25% o i b 4" 88 . t ' NE
~5 m
O ta)
- t i
i }
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () TABLE 7.5-3 PERCENT RESTRAINTS BOUNDED FOR VARIOUS FACTORS
- INSIDE CONTAINMENT RH* Multiplication Factor Load Combination 1.0 1.1 1.2 1.3 1.4 1.5 1.6 1.7 1.8 1.9 CO (DFFR) .98 .99 .99 .99 .99 .99 .99 .99 99 .99 CO (EMPIRICAL) .60 .68 .73 .77 .83 .86 .90 .92 .94 .96 Chugging .96 .99 .99 .99 .99 .99 .99 .99 .99 .99
() OUTSIDE CONTAINMENT 1 i RH* Multiplication Factor Load j: Combination 1.1 1.2 1.3 1.4 1.5 1.6 1.7 1.8 1.9 4 CO (DFFR) .99 .99 .99 .99 .99 .99 .99 .99 .99 i CO (EMPIRICAL) .79 .88 .94 .96 .97 .98 .99 .99 .99 i Chugging .83 .89 .93 .96 .97 .97 .98 .99 .99 () *RH = Rams head. 7.5-9 l
AMENDMENT 13 0CTOBER 1980 O HIST 0GF<AM FOR LOAD COMBINATION
- INCREASE =
NEW LOADS-RH DESIGN RH DESIGN O 5 fe 0; u S 5 L m 100 _ o 90 _ 80 _ g 70 _ g
~
60 _ O E so - E 40 _ 30 _
~
20 _ 10 __ 7b I I I
-100 0 100 200 300 t
~-> INCREASE IN LOAD WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK 18 DESIGN ASSESSMENT REPORT FIGURE 7.5-1 ;
Et1BEDt1ENT LOAD CHANGE INSIDE CONTAINMENT FOR N + C0(DFFR) + SSE l
i l AMENDMENT 13 l OCTCBER 1980 O HIST 0 GRAM FOR LOAD COMBINATION
% INCREASE =
NEW LOADS-RH DESIGN RH DESIGN N 5 a W S Bi G w x O = n< 50 _ -
~
N
- n 40 _
~
30 _ 20 _. 10 _ _ E' s t
-100 0 100 200 300
% INCREASE IN LOAD WM. H. ZlMMER NUCLEAR POWER STATION, UNIT 1 MARK 18 DESIGN ASSESSMENT REPORT FIGURE 7.5-2 O EMBEDMENT LOAD CHANGE INSIDE CONTAINf1ENT FOR N + CHUG + SRVTQ
AMENDMENT 13 OCTOBER 1980 O HISTOGRAM FOR LOAD COMBINATION
% INCREASE =
NEW LOADS-RH DESIGN RH DESIGN N 5 5 0; u S 55 G w 5 80 -
@ 70 -
O E So - Q 50 - d g 40 _ 30 -
" ~
20 _ 10 - c -{"l_ i - 1
-100 0 100 200 300
% INCREASE IN LOAD l
l WM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK ll DESIGN ASSESSM ENT REPORT FIGURE 7.5-3 EMBEDMENT LOAD CHANGE INSIDE CONTAINMENT FOR N + C0(EMPIRICAL) + SRVTQ + SSE
AMENDMENT 13 0CTOBER 1980 O HIST 0 GRAM FOR LOAD COMBINATION
% INCREASE =
NEW LOADS-RH DESIGN 160 _ RH DESIGN m E 150 _ g 140 _ 130 _ o 120 _ _ Bi 110 _ a m 100 _ _ E5 90 - E 80 g __ m 70 0 :
> 60
@ 50 _
=i g 40 -
~
30 - 20 _ 10 _
= f I
-100 0 100 200 300 4 INCREASE IN L0fD WM. H. ZIMMER NUCLEAR POWER STAT!ON UNIT 1 MARK ll DESIGN ASS ESSM ENT REPORT ,
i FIGURE 7.5-4 EMBEDMENT LOAD CHANGE OUTSIDE CONTAINMENT FOR N + C0(DFFR) + SSE
AMENDMENT 13 OCTOBER 1980 O i i HIST 0 GRAM FOR LOAD COMBINATI0fl
% INCREASE =
NEW LOADS-RH DESIGN RH DESIGN C 5
- 140 _
y 130 _ u 120 _ _
~
E 110 g _ m 100 _ 90 _ 80 _
$ 70 _
Q [> 60 _ 50 E 40 _ 30 _ _ 20 _ 10 _ _ J W- 1
-100 0 100 200 300
% INCREASE IN LOAD WM. H.ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 18 DESIGN ASSESSMENT REPORT FIGURE 7.5-5 EMBEDMENT LOAD CHANGE OUTSIDE CONTAINMENT FOR N + CHUG + SRVTQ
AMENDMENT 13 0CTOBER 1980 O HISTOGRAM FOR LOAD COMBINATION
% INCREASE =
NEW LOADS-RH DESIGN RH DESIGN O E _
- 140 _ _
y 130 _
~
u 120 _ 5 110 _ C m 100 _ E5 90 _ 80 _ g 70 _ g 60 _ 50 _ E 40 _ - 30 _ 20 _ 10 _ __ __ h 1
-100 0 100 200 300
% INCREASE IN LOAD WM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK II DESIGN ASSESSMENT REPORT FIGURE 7.5-6 EMBEDMENT LOAD CHANGE OUTSIDE CONTAINMENT FOR N + C0(EMPIRICAL) + SRVTQ + SS
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (} 7.6 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) EQUIPMENT 7.7.1 F1pino and Equipment i General-Electric in cooperation with CG&E and its consultants is ' reevaluating the design adequacy of the NSSS equipment located in I the. reactor building for the design-basis loads in. combination i with the suppression pool hydrodynamic and annulus pressurizationi structural system. responses loads. The load combinations and , acceptance criteria used for this reevaluation are described in Section 6.5. I The structural system responses for the suppression pool i hydrodynamic phenomena were generated by Sargent & Lundy using- i the forcing functions defined in the DFFR. In the case of the
, LOCA forcing function an empirical definition has been developed I which envelops the forcing function defined in the DFFR and provides additional margin to the plant design. These structural system responses were transmitted to General Electric in the form i of: 1) broadened response. spectra at all NSSS equipment '
attachment points, and 2) acceleration time histories at the i pedestal-to-RPV-support connection and at the stabilizer i elevation. , The response spectra for piping attachment points on the reactor pressure vessel, shield wall and pedestal complex are regenerated 8 () by General Electric based on the acceleration time histories supplied by Sargent & Lundy, using a more detailed lumped-mass i i model f r the reactor pressure vessel internals, including a ,
! representation of the structure. For the reevaluation of the NSSS primary piping (main steam and recirculation), a combination' of the General Electric and Sargent & Lundy developed response 8 spectra is used to obtain an envelope response spectrum or uniquei time history for all attachment points of each piping system. i For the reevalur. tion.of the floor-mounted NSSS equipment, except ,
the reactor pressure vessel, the response spectra supplied directly by Sargent & Lundy are used. ' I The acceleration time histories with the detailed reactor i pressure vessel and the structure lumped mass model are also used i to generate the forces and moments acting on the reactor pressure, vessel supports and internal components (see Appendix A Section A.8). These forces and moments are used for the adequacy ' evaluation of the reactor pressure vessel supports and internals.5 > (See Subsection 7.7.2 for further explanation of the reevaluationi of the reactor pressure vessel supports and internal components.) The transient asymmetric pressure buildup in the annular region between the biological shield wall and the reactor pressure ' , vessel due to annulus pressurization were based on pressure time 8 histories using.the methodology presented in Subsection 5.3.1.4. i These pressure time histories were converted by means of the O lumped-mass model (see Appendix A~Section A.8)) to: 1) response , i 7.6-1
l 2PS-1-MARK II DAR- AMENDMENT 13 OCTOBER 1980 ' - ({)' - spectra / acceleration time histories at piping attachment _ points on the reactor pressure' vessel, shield wall and pedestal _ complex, i ,
'and.2) fccces and moments acting on the reactor pressure vessel ,
supports and' internal components. (See Subsections 7.6.1.1 and ' j 7.6.2, respectively, for the detailed discussion of the primary . . piping reevaluation procedures and the NSSS' reactor supports and i 1 internal components.) i i In certain cases, the dynamic' analysis methods used by General , Electric for the reevaluation were more conservative than the '
. current design practices presented by the Mark II Owners Group.
t The dynamic analysis methods presented by the Mark II Owners i L Group are believed to-be adequately conservative and represent- i , industry design practice for.the. response loads from the i suppression. pool hydrodynamic forcing functions and annulus ,
- pressurization. Use of these more conservative methods as delineated in Table 7.6-1 was considered only to minimize '
potential licensing delays for this lead Mark II plant. ' i i _The schedule for final resolution of the ABS load combination i issue is not supportive of the 2PS-1 SER schedules. ,
! Consequently, in order to quantify the conservatism in the ZPS-1 analysis,:and to aid the NRC staff in their decision process, '
- additional " provisional analyses" are being performed. These 8 4
additional " provisional analyses" will be based on significantly more restrictive load combination / acceptance criteria than the i present industry _ practice illustrated in Table 7.6-1. 1 , Specifically, it will'be conservatively assumed that: 1) the stresses'for loads from.the annulus pressurization. event combined ' with stresses for loads from the vibratory motion of a safe i shutdown. earthquake will be compared with the code _ faulted i. , allowable stresses, and 2) the stresses for the combined , i vibratory loads from the~ operating-basis earthquake and the ' actuation of al1~ safety / relief valves will be evaluated by
- normal / upset code criteria. The stresses / loads'for peak dynamic '
response load events are conservatively being combined by the : absolute. sum method (ABS) for the additional " provisional i assessment." These " provisional assessments" are being performed , on an unofficial basis and are not intended to be submitted as ' e part the official-licensing documentation.- However, based on preliminary load data,~these provisional methods and results for 8 the. restrictive load combinations were presented to the NRC~ staff i in meetings held on January 5, 1978 and March 21,.1978, on the i 2PS-1-Docket to apprise the NRC of the plant design adequacy in order to avoid a licensing schedule delay. These evaluations
- have most recently been' updated through presentations to the NRC staff on December'5, 1979. As' indicated by the results presented, additional hardware-impacts could be identified,
' including further snubbing'and stiffening of the primary piping '
-system, and are anticipated'if these alternate, more restrictive '
load' combination / acceptance criteria methods are imposed as the i design basis. The preliminary results summarized at these (}J i 7.6-2
- t
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
/s t) meetings will be finalized as the ZPS-1 NSSS analysis proceeds using the final load data. '
7.6.1.1 Reevaluation Procedures for NSSS Pipino i i In cooperation with the Cincinnati Gas & Electric Company and i Sargent & Lundy, General Electric has defined the additional snubber requirements to resist the secondary dynamic responses from: 1) the original design-basis loads including seismic ' vibratory motions, 2) the structural system feedback loads from i suppression pool hydrodynamic events, and 3) the structural i system loads from annulus pressurization for postulated feedwater i and recirculation pipe breaks. , Lumped-mass models were developed by General Electric for the ' NSSS primary piping systems, main steam and recirculation. The i lumped-mass models include the snubbers discussed above and the i pipe-mounted valves and represent the major balance-of-plant i branch piping connected to the main steam and recirculation systems. ' I These detailed models are analyzed independently (see Appendix A,i Section A.9 for explanation of the analysis of piping systems) to i determine the piping system reactor forces (s?3ars and moments) , for: 1) each design-basis load, which includes pressure temperature, weight, seismic events, etc; 2) the bo~unding pool ' (~)% x load; and 3) the annulus pressurization dynamic effects on the i unbroken piping system. Additionally, the end reaction forces i and/or acceleration for pipe-mounted / connected equipment (valves i and nozzles) are simultaneously calculated. , The piping stresses for these reactor forces (shears and moments) ' from each load vent are determined and combined in accordance I with the load combinations delineated in Section 6.5. These i stresses are calculated at geometrical discontinuities and i compared to code allowable stresses for the appropriate .oad , combinations to assure design adequacy. I 7.6.1.2 Reevaluation Procedures for NSSS Equipment i i 7.6.1.2.1 Reevaluation of Pipe-Mounted / Connected Equipment i The reaction forces and/or accelerations acting on the pipe-mounted / connected equipment when combined in accordance with the ' load combinations are compared to the equipment allowables to i assure design adequacy. i 7.6.1.2.2 Reevaluation for Floor / Structure-Mounted Equipment i Two methods are used for conducting the reevaluation of the '
. above-referenced equipment: i V
7.6-3 l l
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () a. Method 1 - A'nalyze Equipment Using a Static Coefficient
- The new coefficient was established by calculating the natural frequency of the equipment. For this reevaluation, the response spectra below the first natural frequency was ignored. The peak response above natural ~ frequency was increased by 1.5 in accordance with IEEE 344-1975.
- b. Method 2 - Dynamic Analysis Method (RHR Heat
< Exchanger) The structural integrity of the equipment was evaluated with a beam and lumped mass finite element model using SAP 4LO3. The model included Sargent & Lundy support structure and mass of attaching pipe up to the first rigid anchor point. In some equipment the hydrodynamic mass of the water was modelled to account for the effect on equipment frequency and equipment response in the response spectra analysis. The dynamic analysis was performed in two steps: (1) model extraction to determine all significant modes up to a cutoff frequency of 60 Hz; and (2) response spectrum analysis to determine the total ' (q J structural response due to components of seismic and pool hy3rodynamic loads. The modal combination used in the analysis is by SRSS on modes that are not closely spaced and by implementation of Regulatory Guide 1.92 on closely spaced' modes. 7.6.2 Reevaluation for the Reactor Pressure Vessel Supports
; and Internal Components i i
The procedure used to_ verify the design adequacy of the reactor , pressure vessel supports and internal components is outlined in Figure 7.6-1. '
- I
} The reactor vessel system is conservatively designed fot 'oneric
' i requirements, which results in a degree of conservatism with ,
regard to the actual 2PS-1 site acceleration level. The conservatism is based on using generic forces and design to
, ensure that the calculated stresses for the generic forces are '
below~ material allowables. The design-was. initially verified as 8 adequate by calculating the plant-unique _ seismic forces on the i ! reactor. pressure vessel, supports and components after the , j
- vessel / structure math-model had been defined and then comparing _ l these calculated forces to the assumed bounding generic forces. ,
i I
- a. The bounding load combinations for seismic, I hydrodynamic, and annulus pressurization forces are i
(]) established-within each acceptance criteria range i
. defined in Section 6.5 (upset, emergency and ,
7.6-4
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER'1980 () faulted). For the first assessment, the loads are conservatively. combined using the maximum vertical i force and the maximum horizontal shear and moments ' for all combinations. I
- b. These conservative maximum combinations are then I compared to the generic bounding forces originally i used to establish the design. If the combined i calculated forces are less than the design forces, ,
then the component is deemed adequate. If the calculated forces are greater than the design forces, ' then the actual stresses are compared to the material i
; allowables. If the actual stresses are below the i material allowables, then t he design is deemed ,
adequate. If the actual stresses are above the , material allowables, then the loads combinations (Step 1) are reconsidered to combine the vertical ' forces with their more appropriate horizontal shears I and moments. A limiting case or cases are chosen for i each category (upset, emergency and faulted) and , Step 2 is repeated. However, if the design remains unacceptable, then: ' I
- c. The stress analysis ~i;s refined by acceptable methods i based on the unique load combination with the highest i stress results. ,
- d. '
j ' If the refined stress ana19 sis does not lead to an acceptable design, then a hardware modification is 8 , initiated to ensure an acceptable design. i I d r O 7.6-5
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
- TABLE 7.6-1 DYNAMIC' METHODS FOR ZIMMER NSSS ASSESSMENT INDUSTRY PRACTICE ZIMMER ANALYSIS Peak Broadening 15% > 115%
Combination of primary SRSS SRSS for Design and secondary loads Documentation ABS from the same load for " Provisional sources Assessment" Damping Normal and 1% Pipe OD < 12 in. Upset = OBE Emergency and 2% Pipe OD > 12 in. Faulted = SSE Combination of SRSS Algebraic double sum closely spaced a modes O
- O i 7.6-6 l
l _, . . . _ . ,. . . . . . . _ . . . - . . . . . , , r., ,. . , . . - - , - - , . . .
~, - . - - . - - - _ . . - - . . ._. .. .. .- .-.
O O O TABLE 7.6-2 CLASS lE EQUIPMENT QUALIFICATION (GE supplied equipment previously reviewed by SQRT on the Hatch Docket) CLASS lE EQUIPMENT DESCRIPTION LOCATION -MANUFACTURER COMMENTS Level indicating transmitter H22-P005* Barton Qualified by Test AP indicator H22-P010* Barton Qualified by Test Flow Transmitter H22-P009* Rosemount Qualified by Test Intermediate Range Monitor H13-P635 GE Qualified by Test Power Range Monitor H13-P636 GE Qualified by Test n Radiation Monitor and Indicator H13-P635 GE Qualified by Test 5 , Pressure Switches H22-P014* Barksdale Qualified by Test Y Level Indication Switch H22-P004* .Barton Qualified by-Test 7 1 Pressure Switches H22-P005* Static-O-Ring Qualified by Test g w Pressure Transmitter H22-P004* Bailey Qualified by Test
. Relay-HMA H13-P618 GE Qualified by Test g
m 4 Switch (CR2940) H13-P618 GE Qualified by Test [ ! O i Og OM 85 sE i *These local racks are "94" (height) instead of "84" (height). They will be qualified $ via the same criterla that the Hatch "84" local racks were Qualifled/ Tested. $U
'Q .(} }
-TABLE 7.6-3 CLASS lE CONTROL PANELS AND LOCAL PANELS AND RACKS SEISMIC QUALIFICATION TEST
SUMMARY
(GE supplied equipment previously reviewed by SQRT on the Hatch Docket) PANEL DESCRIPTION TYPE VENDOR COMMENT H13-P601 RCCS Benchboard GE Qualified by Test H13-P608 PRNM Vertical Board GE Qualified by Test H13-P618 Division 2 RHR Vertical Board GE Qualified by Test N H13-P635 Radiaticn Monitoring (Div. 1) Instrument Rack GE Qualified by Test @ H13-P636 Radiation Monitoring (Div. 2) Instrument Rack GE Qualified by Test ! H22-P004* RPV LVL/ Press. Local Rack GE Qualified by Test 4i i y H22-P021* RHR-B- Local Rack GE Qualified by Test
- H22-P010* Jet Pump A Local Rack GE Qualified by Test h
, M i t2 s:
*These local racks are "94" (height) instead of "84"_(height). They will be qualified :c tt h via the same criteria that the Hatch "84" local racks were Qualified / Tested. ge
- $U
AMENDMENT 13. OCTOBER 1980 O Seismic DES 1GN e BOUNDING GENERIC FORCES e CONSERVATIVE DESIGN e CALCULATE STRESSES VERIFICATION e CALCULATE PLANT UNIQUE SEISMIC FORCES e COMPARE FORCES e ADEQUATE DESIGN SRV PROCEDURES e CALCULATE SRV FORCES e COMBINE SRV AND PLANT UNIQUE SEISMIC EVALUATION LOWER FORCES e COMPARE COMBINATION WITH > c ADEQUATE DESIGN g BOUNLING FORCES v HIGHER FORCES STRESSES BELOW ALLOWABLES e RECALCULATE STRESSES 5 e ADEQUATE DESIGN e COMPARE ALLOWABLES STRESSES AB0VE ALLOWABLES STRESSES BELOW ALLOWABLES e REFINED ANALYSIS # e ADEQUATE DESIGN I STRESSES ABOVE ALLOWABLES MODIFIED HARDWARE e MODIFY HARDWARE e ADEQUATE DESIGN WM. H.ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK 18 D E SIG N ASSESSMENT REPORT
. FIGURE 7.6-1 DESIGN AND EVALUATION FLOW
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 CHAPTER 8.0'- SUPPRESSION POOL WATER TEMPERATURE (~)> \- MONITORING SYSTEM 8.1 SYSTEM DESIGN 8.1.1 Safe _ty Design Basis The safety design basis for setting the temperature limits for the suppression pool temperature monitoring system are based on providing the operator with adequate time to take the necessary action required to assure that the suppression pool temperature will always remain below the 2000 F temperature limit established by the NRC in NUREG-0487. The system design also provides the operator with necessary information regarding localized heatup of the pool water while the reactor vessel is being depressurized. If relief valves are selected for actuation, they may be chosen to ensure mixing and uniformity of heat energy injection to the pool. 8.1.2 General System Description The suppression pool temperature monitoring system monitors the pool temperature in order to prevent the local pool water temperature from exceeding 2000 F during SRV discharge and provides the operator with the information necessary to prevent excessive pool temperatures during a transient or accident. (,,) Temperatures in the pool are recorded and alarmed in the main control rcom. The instrumentation arrangement in the suppression pool consists of two bulk and 18 local temperature sensors mounted on the pool wall. The two bulk temperature sensors are dual-element chromel constantan thermocouples located at elevation 491 feet 6 inches, and at azimuths 200 and 2220, respectively. These sensors provide signals which are used to indicate to the operator the bulk temperature of the suppression pool. The local temperature sensors consist of 18 dual-element, copper constantan thermocouples located 1 foot 3 inches below the low water level. Twelve of the sensors are located on the outer supptession pool wall at azimuths 280, 480, 880, 1180, 1500, 1800, 2180, 2440, 2660, 2840, 3250, and 3440 The other six are located on the pedestal at azimuths 550, 1300, 1940, 2440, 298.50, and 352.50 The sensors and readout devices are assigned to ESS-1 and ESS-2 divisions and local discharge areas are monitored by two sensors, one from each division. This represents a conservative measurement of local pool nater hcatup. All instrumentation will be qualified Seismic Category I. The time constant of the (~3 thermocouple installation will be no greater than 15 seconds. 's/ The difference between measurement reading and actual temperature 8.1-1
2PS-1-MARK II DAR AMENDMENT 13
, .g OCTOBER 1980 -(3 will be within 2 20 F. The sampling technique for monitoring the t U. pool temperature 11s to sequentially record the measurements made i by each of.the 18 thermocouples. When all thermocouple temperatures are below alarm level, each point'is recorded at a rate of 5 seconds / point. If any thermocouple temperature is above alarm level', each point is recorded at a rate of I second/ point. The discharge locations and spacing are such that- ; the number of sensors and their arrangement provides conservative j
monitoring of locclized suppression pool water heatup in addition to bulk pool temperature. The quenching of the steam at the quencher discharge forms jets that heat the water and generate convection currents-in the suppression pool. These currents eventually rise and displace
- cooler water near the pool surface.
3 i During an extended blowdown, a large temperature gradient is ex-pected initially near the quencher. After a short time the pool 1 gradients will stabilize with a bulk to local temperature i difference of about 100 F. (Bulk and local temperature are defined in NUREG-0488.) The adequacy of the temperature
~
monitoring system will be confirmed by the in-plant SRV testing. 8.'l.3 Normal Plant Operation i " The temperature monitoring system is utilized during normal plant (]) operation to ensure that the pool temperature will remain low enough to condense all quantities of steam that may be released in any anticipated . transient or postulated accident. When rams head devices were specified for design, there was an NRC concern that high pool temperature might result in high pool dynamic loads during SRV discharge because of unstable steam i condensation. Installation of T-quenchers has eliminated this l concern. The local pool temperature (temperature measured on the containment wall at the elevation of the T-quencher) limit is specified to be 2000 F in accordance with the NRC Lead Plant Acceptance Criteria (NUREG-0487). During normal plant operation, the system is in continuous operation recording the suppression pool water temperature with a readout in the main control room. If the pool temperature rises above normal operating temperatures, an alarm is actuated in the control room allowing the-operator to take actions as required to maintain pool temperature below-the 2000 F limit. 8.1.4 Abnormal Plant Operation
' BWR plants take advantage of the large thermal. capacity of the suppression pool during plant transients which require relief valve actuation. . The discharge'of each relief valve-is piped to the suppression pool, where the steam is condensed. This results in a pool water. temperature increase but a negligible increase in containment 1 pressure. However, certain events have the potential - O- for substantial.-energy addition to the suppression pool and could 1
8.1-2
,. , , -. ..-,r . , . - , , , - ~ . . . . , ,,,e , ,.n, ..,,,,--.-,,,-.n..,.--...n-,-
2PS-1-MARK II DAR AMENDMENT 13 OCTOBFF 1080 (l result in a high local pool temperature if timely corrective action is not taken. When rams head discharge devices are used, test results and operating experience indicate that high magnitude oscillatory loads may occur when a high steam mass flux is injected into a pool with local temperature above 1700 F. Although analysis demonstrates that the pool temperature will remain below 1500 F, when the steam mass flux is high enough to cause these loads, T-quenchers have been installed instead of the rams heads to provide additional margin to the pool temperature limits. 8.1.4.1 Plant Transients This subsection discusses various plant transients which result in SRV discharges to the suppression pool and which could possibly lead to high pool temperature. 8.1.4.2 Abnorm &l Events Various events that result in energy being discharged to the suppression pool via the safety / relief valves are discussed in the following paragraphs. Most of these transients are of short duration and have little effect on the suppression pool temperature. However, three events have the potential for substantially high energy release to the pool that could result n%, in undesirably high pool temperatures if timely corrective action is not taken. These events are: (1) events that result in the isolation of th'e plant from the main condenser, (2) stuck-open relief valve, and (3) automatic depressurization system (ADS) operation. A brief description of each of these events is given in the following subsections. 8.1.4.3 Primary System Isolation When the primary system is isolated from the main condenser, the reactor is scrammed automatically and the stored energy in the vessel internals, fuel relaxation energy, and decay heat is rejected to the suppression pool. The amount of heat rejected to the pool depends on reactor size, power level, and primary system heat removal capability. This includes condensing type heat exchangers which remove steam directly from the RPV. 8.1.4.4 Stuck-Open Relief Valve The steam flow rate through a safety-relief valve (SRV) is proportional to reactor pressure. One method to terminate energy input to the pool is to scram the reactor and depressurize the RPV in the event the relief valve cannot be closed. During the energy dump, the pool temperature will increase at a rate determined by the RPV pressure, flow capacity of the SRV, primary ('} system heat removal system capability, and suppression pool water heat removal capability. 8.1-3
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (~) v 8.1.4.5 Automatic Depressurization System (ADS) Activation of ADS results in rapid depressurization of the RPV by the opening of a designated number of safety-relief valves. During this transient, the bulk suppression pool temperature rises. In a typical case, the RPV is depressurized below 150 _ psia in about 10 minutes. 8.1.5 Transients of Concern There are seven plant depressurization transients that were considered as limiting events (with rams heads) for energy released to the suppression pool. These events are numbered 1 through 7 for ease of reference and are described in the following paragraphs: Event 1 is a stuck-open relief valve with the reactor at full power. The plant is scrammed and depressurization begun via the stuck-open valve. The initial pool temperature is the maximum pool temperature allowed for continuous operation. This is the only event that is not truly limiting, since all RHR heat exchanger equipment is considered operational. Event 2 is identical to Event 1, except that one RHR heat ex-changer is considered unavailable. The remaining heat exchanger equipment is placed in the suppression pool cooling mode. Event 3 is a stuck-open relief valve with the reactor isolated from the main condenser and maintained at operating pressure and temperature. All RHR heat exchanger equipment is considered operational in this case, because if one heat exchanger is not available, the reactor may not remain isolated and pressurized. The bulk pool temperature assumed in this event is the highest allowed while the reactor is maintaining pressure. Event 4 is a controlled depressurization from the plant condition described in Event 3. This event demonstrates the ability of the plant to safely depressurize through a controlled transient from the limiting conditions. Event 5 is a controlled depressurization several minutes after reactor isolation described in Event 3. One heat exchanger is considered unavailable because power operation in that condition is not prohibited. This event demonstates the ability to safely depressurize the plant through a controlled transient with only partial RHR heat exchanger equipment availability. Event 6 is a rapid depressurization of the reactor resulting from actuation of the autcmatic depressurization system (ADS). ADS actuation takes place at the time the plant is scrammed. Because this event takes place quickly, no heat ex;nangers are brought into operation. (j3 w This event represents the most rapid energy 8.1-4
1
- ZPS-1-MARK II DAR AMENDMENT 13 i OCTOBER 1980 1
O v release to the pool and demonstrates that the plant may be safely depressurized in this manner without use of heat exchangers. a Event 7 is also a rapid depressurization via ADS, but with the reactor isolated and pressurized and with a limiting pool temperature condition. All heat exchanger equipment is placed in the pool cooling mode 0.5 hour after scram butt is considered unavailable when depressurization begins. I 4 I l O J i T T O 8.1 1 _ . . - , _ . . . . _ . - . . _ . . - . ,_._m. ... . . _ . . . . . . . . _ _ . . _ , . . . _ . . . . . . . , , , ~ , , , , _ . . . . . , , _ . - - . _ . . . . , . . . . _ , . . . .
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l l 2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 j 8.2 SUPPRESSION POOL TEMPERATURE RESPONSE The NRC has requested analyses of the 2PS-1 suppression pool temperature response to specific plant transients.(Reference 1). While GE has responded to most of the NRC's concerns on a generic basis (Reference 2), further plant unique evaluations were required. Generic assumptions for use in analyzing Mark II suppression pool transients have been provided in a Mark II report transmitted to the NRC on April 18, 1980 (NFN-073-80, R. H. Buchholz to J. F. Stolz). Plant unique evaluations are currently being conducted for ZPS-1. This evaluation will include the final 2PS-1 design parameters including the T-quenchers. Since the results of this analysis are not yet available, a summary of the results of the rams head analysis of the two expected limiting case events, i.e., stuck-open relief valve from power operatio7 and stuck-open relief valve from hot standby, is included here. The analysis assumptions and preliminary results are shown on the accompanying tables and figures. Table 8.2-1 lists common assumptions used for all 2PS-1 suppression pool temperature response analyses. Event 1, described in Table 8.2-2, is given as a stuck-open relief valve from power operation; Event 2, described in Table 8.2-3, is the stuck-open relief valve from hot ([) standby. As can be seen from Figures 8.2-1 and 8.2-2, the preliminary results show a bulk pool temperature of 155.40 F for Event 1 and 148.20 F for Event 2 at the 40 lbm/sec-ft2 critical mass flux (the bounding mrss flux below which instability is not expected to occur for ramu head discharge). The quencher analysis will be extended to consider the complete blowdown. It should be noted that this analysis assumed the use of a rams head device on the SRV discharge lines in the suppression pool. The results reported herein are conservative for the use of a quencher device on these SRV discharge lines because the quencher device increases the bulk temperature limit to 2000 F, as specified in NUREG-0487, and at the same time provides decreased containment loads due to SRV discharge. 8.2-1
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 8.2.1 References
- 1. John F. Stolz (NRC) to Mr. Earl A. Borgmann (CG&E), letter dated September 8, 1977, " Requests for Additional Information -
2 Suppression Pool Temperature Limit (Wm. H. Zimmer Nuclear Power Station, Unit 1)."
- 2. E. D. Fuller (GE) to Mr. Olon D. Larr (NRC), letter dated September 6, 1977, " Memorandum Report, Rams Head Suppression Pool Temperature Limit."
- 3. " Test Results Employed by GE for BWR Containment and Vertical Vent Loads," NEDE-21078-P (GE Proprietary Report), October 1975.
O 8.2-2 4
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, ./ ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 t
O. s TABLE 8.2-1 COMMON ASSUMPTIONS USED FOR ALL ZIMMER SUPPRESSION POOL TEMPERATURE RESPONSE ANALYSES 1
- a. Decay beat per ANS-5.
i l b. Fully crudded RHR heat exchaagers.
- c. Turbine-driven feedwater pump flow assumed available until MSIV closure. (Pump coastdown of 2.6 seconds. )
- d. RCIC and HPCS pump sunction from condensate storage
; tanks.
l
- e. Limiting (122.5% ASME rated) SRV flow rates.
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ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (* ( TABLE 8.2-2 ZIMMER EVENT 1 - STUCK-OPEN RELIEF VALVE FROM POWER OPERATION INITIAL CONDITIONS
- a. Operation at 105% rated steam flow.
- b. Design maximum service water temperatcre.
- c. Suppression pool temperature at normal technical specifichtion limit - l' F.
- d. Minimum technical specification suppressien pool level.
EVENT SEQUENCE, TIME (min) EVENT DESCRIPTION 0.0 SRV fails to open. t" Pool temperature alarm at operating technical specifi-cation limit. t + 10 Reactor scram and isolaticn. ({} t" + 15 Single RHR loop lined-up for pool cooling. t, + 15 Two additional SRV manually actuated. ASSUMPTIONS
- a. Common assumptions.
- b. Normal automatic operation of RCIC and HPCS.
- c. Single RHR loop available for pool cooling.
- d. Maximum operating condensate storage water temperature.
- e. 122.5% ASME rated SRV f 3 ow rate.
O d 8.2-4 ,s
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ZPS-1-MARK II DAR
- AMENDMENT 13 OCTOBER 1980
() TABLE 8.2-3 ZIMMER EVENT 2 - STUCK-OPEN RELIEF VALVE FROM HOT STANDBY INITIAL CONDITIONS
- a. Operation at 105% rated steam flow.
- b. Design maximum service water temperature
- c. Suppression pool temperature at normal technical specification limit.
- d. Minimum technical specification suppression pool level.
EVENT SEQUENCE TIME (min) EVENT DISCUSSION 0.0 Reactor isolation and scram. 0<t<30 Reactor pressure maintained with SRV. 30 Single SRV sticks open. 40 Two additional SRV's open (operator action).
/('}
60* Two RHR loops lined-up for pool cooling. ASSUMPTIONS
- a. Common assumptions.
- b. Normal automatic operation of RCIC and HPCS.
- c. Two RHR loops available
- d. Maximum operating condensate storage water temperature.
- e. 122.5% rated ASME SRV flow rate.
- Delay in pool cooling lineup due to change from steam condensing mode of RHR to pool cooling mode.
8.2-5 l
AMENDMENT 13 OCTOBER 1980 ISO (V3 l 21.5 MIN, RPV AT 179 PSIA 15 0 o' E o D140 15 MIN , RPV AT 432 PSIA
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AENDMENT 13 0CTOBER 1980 O 160 , 150 - 48.6 MIN. AT 179 PSIA o nd E g140 Y - o.
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ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 {) CHAPTER 9.0 - PLANT MODIFICATIONS AND RESULTANT IMPROVEMENTS 9.1 STRUCTURAL MODIFICATIONS In general, the impact of the addition of the pool dynamic loads ' on a majority of the structures was minimal. The primary reasons I are as follows: i i
- a. Fixed-base seismic loads were used in the original ,
design.
- b. Except in local areas, the design of the containment 8 structure is generally governed by load combinations i involving safe shutdown earthquake and design-basis i accident. Pool dynamic loads are relatively small ,
compared to these governing loads. The following is a summary of the structural modifications 8 necessitated by the addition of pool dynamic loads: i l
- a. The inner core of the reactor support was filled with ,
concrete up to elevation 497 feet 6 inches to reduce the bending stresses induced by the pool dynamic ' loads. Structural integrity of this core fill was i ensured by providing reinforcing bars and concrete i stud anchors welded to the reactor support liner. , O' Figures 4.1-5 and 4.1-6 of the DAR give the details ' of this modification. ' I
- b. The gallery platform in the suppression pool at i elevation 510 feet 6 inches has been removed. ,
- 1. Additional steel framing has been installed in '
the suppression pool at elevations 499 feet I 6 inches and 520 feet 1/4 inch to support MSRV and non-iSRV piping. New embedments (anchored plates) and ring girders have been installed in the suppression pool for MSRV and non-MSRV piping.
- c. The flange at the end of the downcomer vent will be '
removed. ' l
- d. Horizontal bracing of the downcomer at elevation 496 feet.
- e. Embedments and pedestal anchor installed for downcomer bracing and for supporting MSRV and non-MSRV guides in the wetwell.
p,, f. Removed vacuum breakers from downcomers. V 9.1-1
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () The following additional modifications may be required, pending i completion of reassessment: i l
- a. The steel framing in the drywell may require '
additional stiffening cover plates or replacement with stiffer members. ' i
- b. Distribution of drywell framing loads to other i support locations may be required to reduce loads on ,
heavily loaded embedments. ,
- c. Ten percent of the cable tray hangers in the reactor building wall have to be stiffened,
- d. Block wall fixes,
- e. HVAC duct hanger fixes.
- f. Conduit hanger fixes.
- g. Main steam wall fix for higher piping loads.
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ZPS-1-MARK II DAR . AMENDMENT 13 OCTOBER 1980 9.2 BALANCE-OF-PLANT (BOP) PIPING AND EQUIPMENT 9.2.1 BOP Pipina 9.2.1.1 Drywell Piping The piping inside the drywell was found to be impacted by the empirical limiting CO load combination as stated in Subsection 7.5.1.1.2. As a result, all service level C support loads based on the rams head design basis loads inside the drywell were increased by a factor of 1.33 in order to accommodate the large amplitudes in the higher frequencies of the CO load. (See Subsection 7.5.1.3 for a discussion of the 1.33 factor.) The supports were then qualified for the increased load and upgraded to a larger support size wherever the existing support capacity was exceeded. If the results of the assessment indicated that a support load increased more than a 1.33 factor, than the higher factor was i used for the design of the support. Another exception to the 1.33 factor was the feedwater header connected to the RPV. Only a 1.0 factor was used because it was shown that transient load due to a feedwater pump trip governed the design of the supports. ({} The effects of the SRV T-quencher loads on the supports designed to the rams head design basis were minimal. The change from rams head to T-quencher devices did not result in any existing design modifications. The only piping found to be affected by the SRV T-quencher load were the small (<4-inch) piping with low (2-7 Hz) fundamental frequencies. However, the small piping supports were found to be adequate even for the T-quencher load since the load was of low magnitude to begin with. ' In conclusion, all drywell supports can accommodate the empirical limiting CO load combination. 9.2.1.2 Wetwell Piping The wetwell piping was upgraded to accommodate the latest sup-pression p ol changes. As a result, all the wetwell piping was rerouted and resupported due to the installation of_the T-quencher device and the addition of the downcomer bracing system, and to reduce the overstress of the wetwell columns. The changes made in the wetwell included the following: l
- a. rerouting all the wetwell piping, -
() b. replacing che rams head with a T-quencher discharge device, 9.2-1 l
~2PS-1-MARK II DAR AMENDMENT 13
- OCTOBER 1980 i t l'
() c. upgrading piping wall thickness and shear lug sizes where required,
- d. adding 226 piping supports, ,
) e. replacing the elbow and stancion arrangement at the l top of the MSRV line riser with a special fabricated
-tee and strut arrangement, and
) f. installing new suction strainers for the ECCS and l .RCIC pump intakes. 9.2.1.3 BOP Pipino The BOP piping which was designed for the rams head design basis l was found to be impacted locally due to the chugging and i empirical CO load. The local impact affected-only those piping , systems attached to the outer suppression pool wall (at , i approximately elevation ~497 feet). It was found that a factor of 1.33 x rams hcad design-basis i emergency loads would be adequate to-accommodate the chugging and l CO loads for the piping supports. As a result, all the support. loads on piping systems connected to
,q or supported on the outer suppression pool wall at mid-center j v (elevation- 497' feet).were increased by 33%.
1 9.2.2 Equipment The reactor building closed cooling water (RBCCW) expansion tank,
- the residual heat removal (RHR) heat exchanger support bolts, and j the RBCCW heat exchanger support bolts have been. modified to
- accommodate the additional pool dynamic loads. This design j modification consisted of strengthening the saddle supports and replacing or adding additional support bolts. As a result of the design assessment performed for assessing the impact of changing
! the quencher device to-the T-quencher, it is anticipated that , design modifications may be required for the following equipment: i i a. core spray cooling system - RHR equipment room
- cooling coil; b ~. core spray cooling system - LPCS/RHR equipment room cooling coil;
- c. core spray cooling system - HPCS equipment room
- cooling coil L
l d. reactor building closed cooling water heat exchanger l' 1B'- !O 9.2-2
. _ _ , , . . , _ _ . _ , , , , - _ . ~ -
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i i e r i I f 1 I & i i, a i i !O a i i i i l l i i' 1 1 I i i O l i l 9.2-3 i
ZFS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () 9.3 NSSS PIPING AND EQUIPMENT , Modifications have been incorporated (or planned) for HSSS 8 components / equipment to accommodate the structural sy3cem i feedback responses from suppression pool hydrodynamic and annulus i
. pressurization loads. The modifications include the following: ,
- a. Upgrading the snubber requirements for the primary '
i system piping, main steam and recirculation, for I additional and increased sizes. I 1
! b. Upgrading the residual heat removal heat exchanger ,
anchor bolt material. , 1 All components / equipment will be modified to meet the square root I of the sum of the squares (SRSS) load combinations defined in i Chapter 6.0 when design adequacy is not demonstrated by refined i analysis or retest. Retesting of components will be made where , practical to demonstrate additional capacity prior to modification. ' I i If components / equipment which are reevaluated to the i
" provisional" absolute sum evaluation exceed allowable stress i i
levels (even after refined analysis techniques), they will be , i placed in a provisional status before making further modifications, and the NRC will be asked to review the ' O- " provisional" reevaluation methods and results. A rationale will i
; be provided for approval of less than ABS for those items where it is deemed necessary. ,
1 I i i 9.3-1 3
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ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () 9.4 SRV DISCHARGE QUENCHER The discharge lines from the 13 safety / relief valves (SRV) are routed from the drywell down into the suppression pool. Each discharge line terminates with a T-quencher discharge device as shown in Figure 9.4-1. Each T-quencher is attached to a base plate in the containment floor. The center,line of the T-quencher arms is 3 feet 6 inches above the top of the suppression pool basemat. L.is elevation is equivalent to a submergence of 18 feet 6 inches below the pool low water level. The plan location of the T-quencher is shown in Figure 9.4-2. The location and orientation of the quenchers was based on
- several considerations which included the following
- a. physical separation from structures to minimize submerged structure loads (a minimum separation of approximately 5 feet has been provided),
i b. physical separation from suction strainers to prevent an air or two-phase mixture from entering the ECCS or RCIC pumps, and
- c. thermal mixing and utilization.
~% d. The plan location of the quencher incorporates SRV (d
symmetry by setpoint group as follows: j 1. Low setpoint group, two valves at lowest setpoint.
- 2. Multiple valve groups, five valves which are from the two lowest setpoint groups.
- 3. ADS valves.
The T-quencher discharge device is substantially different from the original rams head device. The primary reasons for switching from the rams head to the T-quencher were as follows:
- a. The T-quencher provides wider dispersal of the air inventory in the vent line with lower air clearing loads.
- b. The T-quencher enhances the condensation of steam.
- c. The T-quencher discharges steam without steam condensation instability at higher pool temperatures ,
than t-e rams head device. * ' 1 The changes to systems and structures are described in Sections () 9.1 and 9.3. In most areas of the plant these changes were minimal, since for most frequencies the rams head response 9.4-1
' 2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 O spectro enveioge the 1-2uencher resgense spectro. There re so e frequencies, low frequencies in particular, where the T-quencher response spectra exceed the rams head response spectra. 1 4 f f 4 h a ! O 4 4 i I N i lO 9.4-2 4
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S UP S R ES SIO6J POOL EL _' 4W- d WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 cf MARM 18 DESIGN ASSESSM ENT REPORT FIGURE 9.4-2 ' PLAN LOCATION OF T-QUENCHER ; s A
- h h
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 m( / CHAPTER 10.0 - PLANT SAFETY MARGINS 10.1 CONSERVATISMS IN PLANT DESIGN The purpose of this chapter is to discuss the conservatisms in the pool dynamic loads, the structural methods of analysis, the mechanical methods of analysis, the methods used in the NSSS assessment, and the plant margins during maximum transient con-ditions. Each of these conservatisms will ha discussed separately as they relate to plant safety margins. 10.1.1 Conservatisms in Pool Dynamic Loads Conservatisms in the safety / relief valve (SRV) pool dynamic loads have been discussed in detail in Subsection 5.2.2. The rams head load definition contains the ccnservatisms in each step of the calculation as discussed in that subsection. This procedure has resulted in a design which is very conser-vative for the loads anticipated from the rams head discl: .rge device. In order to eliminate concerns abou* high pool temperature operation, quencher devices have been installed at 2PS-1. An additional significant load reduction is expected to result from this change. (]) Loss-of-coolant accident (LOCA) loads have also been treated in a very conservative manner. The design-basis accident is postulated as a double-ended guillotine break, clearly an unlikely mode of failure for a large pipe. The pool swell transient, as described in Subsection 5.3.1, is predicted by a conservative model that has incorporated additional conservative factors recommended in NUREG-0487. A bounding load approach has been utilized in calculating the steam condensation design loads, ignoring all load reductions'due'to random phasing and to the probability distribution of load magnitudes. The steam condensation loads for 2PS-1 are being treated in a more conservative manner than established in the Dynamic Forcing Functions Report (NEDO-21061) of the Mark II Owners Group. In our judgment, the Owners Group position is appropriate and technically preferred; however, to expedite the NRC staff ap-proval, a more conservative position will be used for the 2PS-1 evaluation. In this case the Zimmer empirical CO load has been used for design reassessment, as discussed in Chapter 1.0. This load clearly bounds the load definition recommended in NUREG-0487. Submerged structure loads for both SRV and LOCA events have also been calculated in a conservative manner. In response to the NRC acceptance criteria (NUREG-0487), geometric and transient flow effects have been considered as described in Appendix H. Submerged structure loads were increased when a potential for
~/ higher loads was found, but no loads were decreased because of the additional considerations recommended in NUREG-0487.
10.1-1
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () 10.1.2 Structural Conservatisms I The margin factors obtained in the original assessment are very conservative for the following reasons: 1
- a. The peak responses of the in6ividual transient loads ,
sie combined by the absolutt sum method for a load ' combination, even though probabilistically all peak I effects will not occur at the same time and it would be more realistic to use SRSS of the peaks. i i
- b. The fact that the instantaneous peak responses ,
induced by loads such as earthquake, SRV discharge, LOCA, etc., do not occur simultaneously at all points ' I along the circumference of the structure was conservatively neglected in the design. i i
- c. The amplified building response spectra for pool ,
dynamic loads were widened by a factor of 20% on 8 either side of a peak rather than the conventional t 15% as per Regulatory Guide 1.22. I I
- d. In load combinations, the effects of individual loads are magnified by a load factor.to account for ,
probable overloads. [D ss e. Current ASME Code for the design of concrete I containment structures (ACI-359) treats thermal i stresses as self-limiting secondary stresses and , permits yielding of the reinforcing steel when '
-thermal loads occur in a load combination. However, the structural design criteria for the ZPS-1 '
containment are very conservative and more stringent i than the current practice and do not permit yielding i of the reinforcing steel even under thermal loads. ,
- f. Material understrength factors (4-factors) built into '
the allowable stress criteria will lead to actual I safety margins larger than those computed. i i 10.1.3 Mechanical Conservatisms 10.1.3.1 Conservatisms in BOP Pipino Analysis Conservatisms incorporated in the BOP piping analysis are out-lined in the following:
- a. AL A TQ was used The envelope of the SRVforallSRVloadsintheI.TOandSRV oadcombina5Yonswherethe :
SRVALLTO load was required. ' () b. The SRV A TQ (all valve discharge) load was used in lieuofhkeSRVADS TQ (ADS valve discharge). l 10.1-2 l I
2PS-I-MARK II DAR AMENDMENT 13 OCTOBER 1980 f~)\ (_ c.s The condensation oscillation (CO) load used the envelope of both the high mass flux and medt.um mass
' lux Zimmer empirical limiting CO load as defined in Chapter 1.0.
- d. The CO load combination which included the envelope of both the high mass and medium mass flux Zimmer empirical limiting CO load also included the envelope of SRV antTO and SRV ASYTQ and the SSE loads. The Zimmer empirical limiting CO load is defined in Chapter.l.0.
- e. The piping stresses and support loads were added by
; the absolute sua method. Exceptions were annulus pressurization (AP) and safe-shutdown earthquake (SSE) loads, which were combined by the square root of the sum of the squares (SRSS) method.
- f. The piping subsystem analyses were performed using the enveloped response spectra method.
- g. The analyses used the maximum (or design) operating pressure and temperature for all load combinations.
The actual pressures and temperatures would be lower if actual plant conditions during a shutdown period "T were used (e.g., actual RPV pressure and temperatures ('d following an SRV discharge).
- h. The minimum valve closure time was used in calculating transient loads,
- i. All reactor building restraint loads and piping attached to the outer suppression pool wall near mid-center (excluding instrumentation lines) were increased by a factor of 1.33 or higher times the rams head design load to account for the uncertainties in T-quencher and LOTA loads that had .
, been completely analyzed at the time of reasscssment.
The majority of restraint loads actually decreased from the rams head load, thus providing a fhetcr greater than 1.33 for those restraints. This is conservative procedure that allows continuation of redesign and reassessment without delaying the project schedule.
- j. All instrumentation lines and small-bore piping using a simplified method of dynamic analysis were designed to the envelope of the rams head and T-quencher loads for all response spectra in a particular area. ;
s For example, all response spectra inside the drywell,
) including the spectra for the RPV, drywell floor, biological shield wall, and containment wall, were 10.1-3 --___--_-_-__________-_______u---______________-.
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () enveloped into a combined response spectrum. The combined response spectrula was us'ed to determine the loads for all small piping inside the drywell.
- k. All large-bore piping, except wetwell piping, was i analyzed for both the original design basis loads and for the T-Ouencher and LOCA loads. The design was based on the greater of the two design bases.
- 1. For wetwell piping only,-drag loads due to 4 hydrodynamic events are applied in a conservative manner. The submerged structure load is applied in
- resonance with all piping modal frequencies within i the specified frequency range of the submerged structure load.
10.1.3.2 Conservatisms in BOP Equipment i The conservatisms incorporated in the BOP equipment qualification j are outlined in the following:
- a. Qualification will be by the response spectrum method rather than by the time history method.
- b. Peaks of the pool dynamics response spectra have been broadened by 20%.
Q
- c. The load combinations defined in Subsection 6.4.3 use the envelope of the SRVALLas maximum in the vertical direction while the SRV ASY is maximum'in the
- horizontal direction.
i a
- d. The response spectra curves generated for the pool
! dynamic loads consist of a resultant (radial)
- horizontal and a vertical curve, however for requalification this resultant horizontal curve will be applied in both horizontal directions simultaneously with the vertical curve.
10.1.4 Conservatism in NSSS Desian There are a number of conservations in the NSSS design as follows: i
- a. The response was combined by both the algebraic sum i and SRSS method. In most cases the equipment has I
- been qualified using algebraic sum. I i b. The faulted load case has been used in most cases and i the calculated stress has been compared with upset allowable.
; 10.1-4 i
e
...-._.,.f,,-,_--,,.m,~_,.,,,m._..--.._-._-em,, .,.y. . , -....~_r.--.~,,7,.w,,,,,y._r..__.,.,.._,_
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 s) ("T c. The dynamic responses of the structures are evaluated as a static, that is, steady load. This results in a dynamic reserve margin for which no credit was taken. The difference is typically substantial.
- d. Comparison of analyses predictions and test results in plants indicates that our methods over predict the structural responses of the hydrodynamic loads significantly. '-
- e. ASME Code criteria is used to measure the acceptability of loads and fatigue damage to them, even for provisional (unlikely) load combinations.
Thus the safety factor for these loads is the same as regularly occurring original design loads. Actually, the concept of a constant (incredible) probability of fiailure, as is approached, by having higher allowable stress for events with lower probability of occurrence would suggest the use of higher allowables for these provisional cases. Higher values were not used.
- f. At each step of the new load, definition and evaluation conservatisms are included and compounded.
Examples of these are:
- 1. The inclusion of the worst load components from any of the several SRV cases is often evaluated to avoid the evaluation of the many different cases which would otherwise have to be evaluated.
- 2. Often loads which result from fault conditions are conservatively evaluated against e.aergency or even upset criteria just to reduce the effort required in evaluating more cases.
\> 10.1-5
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 Q kl CHAPTER 11.0 - CONCLUSIONS
- All suppression pool hydrodynamic forcing functions which will be considered in the final assessment of the Wm. H. Zimmer Nuclear Power Station are identified in this report. The report includes summary descriptions and references appropriate documents for more detailed descriptions of the very conservative forcing functions applied for this final assessment. As described in this report, these forcing functions include the Mark II containment lead plant information and other information which has been used in response to comments from the NRC staff and consultants. With the information included in this report or referenced by this report, the NRC staff will have adequate information to determine that suppression pocl hydrodynamic forcing functions have been satisfactorily identified and described for the final 2PS-1 assessment.
For loss-of-coolant-accident (LOCA) loads, the forcing functions are based primarily on the results of full-scale tests which simulate Mark II containment conditions. In our judgment, the description of the LOCA forcing functions in this report are , conservative and consistent with our understanding of the acceptance requirements of the NRC staff. For loads associated with the operation of the safety / relief ([) valve (SRV), the forcing functions used have been those developed for the rams head discharge device. On the basis of our assessments, it is our judgment that 2PS-1 will satisfactorily carry the loads and load combinations resulting from the rams head forcing functions described in this report. However, to bound any uncertainties in the SRV forcing function and to expedite NRC staff approval, T-quenchers will be installed prior to startup. The final 2PS-1 assessment, including suppression pool hydro-dynamic loads, is well under way. This assessment will be performed as described in this report using conservative load combinations, acceptance criteria, and load methodology. Some of these items are being treated in a more conservative manner for 2PS-1 than established in the dynamic forcinq functions report . (NEDO-21061) of the Mark II Owners Group. In our judgment, the owners group positions are appropriate and technically preferred; however, to expedite NRC approval, a more conservative position wi!! be used for the ZPS-1 evaluation. With the information included in this report or referenced by this report, the NRC staff will have adequate information to determine that suppression pool hydrodynamic forcing functions will be adeq;ately included in the final design assessment for 2PS-1. O 11.0-1
, 2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 l
(]) APPENDIX A - COMPUTER PROGRAMS A.1 DYNAX l DYNAX (Dynamic Analysis of Axisymmetric St'ructures) is a finite 4 element program capable of performing both static and dynamic analyses of axisymmetric structures. Its formulation is based on the small displacement theory.
.Three types of finite elements are available; quadrilateral, triangular, and shell. The geometry of the structure can be general as long as it is axisymmetric. Both the isotropic and orthotropic elastic material properties can be modeled. Discrete i
and distributed springs are available for modeling elastic foimdations, etc. For static analysis, input loads can be the structure weight, nodal forces, nodal displacements, distributed loads, or
, temperatures. Loads can be axisymmetric or nonaxisymmetric. For the solids of revolution, the program outputs nodal
, displacements, and element and nodal point stresses in the global
- system (radial, circumferential, and axial). In the case of j shells of revolution, the output consists of nodal displacements,-
and element and nodal point shell forces in a shell coordinate system (meridional, circumferential, and normal). (). For dynamic analysis, three methoos are available; direct
- integration method, modal superposition method, and response
- spectrum method. In the case of dynamic analysis by direct integration method or modal superposition method, a forcing function can be input as (1) nodal force components versus time t
for any number of nodes, or (2) vertical or horizontal! ground i acceleration versus time. For nonaxisymmetric loads the equivalent Fourier expansion is used. In the case of dynamic analysis by response spectrum method, spectral velocity versus natural frequency for up to four damping constants is input. The output of dynamic analysis is in terms of nodal displacements, element stresses, and resultant forces and moments at specified time steps. When the modal superposition method is used, and in the case of earthquake response analysis, the requested number of frequencies and mode shapes are computed and printed together with the cumulative response of all the specified modes, and computed by the root sum square (RSS) method and the absolute sum method. I DYNAX was originally developed under the acronym ASHAD by S. Ghosh and E. L. Wilson of the University of California, Berkeley, in 1969 (Reference 1). It was acquired by Sargent & Lundy in 1972 and is operating under EXEC 8 on UNIVAC 1100 series hardware. I A.1-1
1 I 2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (~) N' To demonstrate the validity of the major analytical capabilities of DYNAX, documented results and hand calculations for several problems are compared with DYNAX results. The first problem is taken from S. Timoshenko's book, " Theory of Plates and Shells" (Reference 2). A clamped shallow spherical shel), shown in Figure A-1, is analyzed for displacements and stresses produced by a uniform pressure applied on its outside surface. DYNAX and Timoshenko's solutions are compared in Figures A-2 and A-3. The second problem, taken from " Theory of Elasticity" by Timoshenko and Goodier, (Reference 3), is a plane strain analysis of a thick-walled cylinder subjected to external pressure. The finite element idealization and the loading system used for this case are shown in Figure A-4. Results of the DYNAX analysis are compared with the exact solution in Figure A-5. The agreement for both stresses and displacements is excellent. The third problem was presented in an article by Budiansky and Radkowski in an August 1963 issue of the AIAA Journal (Reference 4). The structure, illustrated in Figure A-6, is a short, wide cylinder with a moderate thickness to radius ratio. The applied loads and the output stresses are pure uncoupled harmonics. For this finite element analysis the cylinder is dividad into 50 elements of equal size. This problem checks the harmonic () deflections, element stresses and forces. Figures A-7 and A-8 compare DYNAX results with the results given in the article. The fourth problem is taken from an article by Reismann and Padlog (Reference 5). A ring (line) load of magnitude P (500 lb) is suddenly applied to the center of a freely supported cylindrical shell. The dimensions of the shell and the time history of the load are sh;>wn in Figure A-9. Because of symmetry only one-half of the cylindsr is modeled using 80 elements of equal size. The time history of radial deflection and meridional moments from DYNAX and from Reismann and Padlog are compared and are shown in Figures A-10 and A-ll, respectively. For the fifth problem the method of mode superposition is used to solve a shallow spherical cap with clamped support under the action of suddenly applied uniformly distributed load. The dimensions of the shell and the load time history are shown in Figure A-12. The first 12 modes were considered to formulate the uncoupled equations of motion. Each of these equations was solved by the step-by-step integration method using a time step ! of 0.1 x 10-4 seconds. The results are compared graphically with those obtained by S. Klein (Reference 6) in Figures A-13 and A-14. The sixth problem is a hyperbolic cooling tower, as shown in () k> Figure A-15. The tower is analyzed for horizontal earthquake motion. A response spectrum for 2% damping, as shown in Figure A.1-2 1
l 2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
/"N A-16, was used for this analysis. The RMS values of the ll meridional force are compared with those obtained by Abel, et al.
(Reference 7) in Figure A-17. The seventh problem demonstrates the validity of the tieing rou-tines. A moment of 100 k-ft/ft is applied at the top of a 50-foot cylindrical shell (Figure A-18). Figure A-19 shows the results from DYNAX and an analytical solution (Reference 2). In the eighth problem, a plate, shown in Figure A-20, is analyzed for cracking due to varying temperature and the results from DYNAX are compared with hand calculations. The finite element model and material properties are also shown in Figure A-20. The temperature gradient is 2.40 F per inch thickness. The strain calculated by DYNAX is to = 1.8936E-5 and the strain calculated by hand is t o = 1.81E-5. For the ninth problem a cylinder under constant pressure (Figure A-21) is analyzed by DYNAX and SOR III, a public domain program (Reference 8). The flexibility matrix for the boundary conditions of the top and bottom used in SOR III is
.33294 x 10-3 .55426 x 10-3 Il I uh L = I
.55426 x 10-3 .18453 x 10-3 j al
_ (1 () The inverse of this matrix is then input in DYNAX as
-750.8 0. O. -2255.2 1 H
- 0. .5 x 10* 0. O. 1 V 4 - 4 s
- 0. O. O. O. 1 T
-2255.2 0. O. 1354.7 1 M
- - s> u ,
Table A-1 shows a comparison of results for the two programs, t ( A.1-3
- - - . - - ~ , .- , , , . , - - - , , , - - .---p
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 [ )
' (~ The tenth problem analyzes a cylinder under constant dynamic axial pressure (Figure A-22) for nonreflecting boundaries. The velocity of the waves traveling through the cylinder due to the dynamic load can be calculated as follows:
C= (1-v) E = 100 = 10 ft/sec. (1+v) (1-2v) p 1 After two weeks the waves will reach the bottom of the cylinder. Since the dynamic load is constant, the velocity cf the particles should maintain its value of 10 ft/sec over the entire cylinder for the test of the load duration. Table A-2 shows some of the results obtained using the nonreflecting boundary option at the bottom of the cylinder. The velocity in the 2-direction at 2.2 seconds and at time 4.0 seconds is in good agreement with the actual velocity of 10 ft/sec. In the eleventh problem a cylinder (Figure A-23) is analyzed for frequency, mode shapes and mixed modal damping obtained from the input material damping constants. The results are obtained based on weighing the damping factors according to the stiffness of each element. Results of this problem are compared to hand calculations in Table A-3. () As shown in these problems, DYNAX is capable of producing accurate results for both static and dynamic analyses of shells.
- V l
1 l A.1-4 i
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () A.2 FAST FAST (Frequency Response Analysis of Structures) is used primarily for the dynamic analysis of a linear axisymmetric structure subjected to many independent loading cases. The structural model is given as input either in the form of its re-sponse for a typical dynamic load, usually a band limited white noise load, or in the form of its eigenvalues, eigenvectors, participation factors, and modal damping ratios. The structural model input required for FAST is obtained from the results of finite element programs. Using the given input parameters, FAST computes transfer functions for various response components. A transfer function expresses the relationship in the frequency domain between a given loading and a specific response component. Using these transfer functions, the response to different loading cases may then be obtained using FAST. This approach reduces the computer analysis time for a structural model which is subjected to different time history motions since only one detailed analysis of the original structure using finite element programs is required; the analysis for the different time histories is then performed using the transfer functions obtained by FAST. The results are given in print and plot forms. FAST was developed at Sargent & Lundy in 1975. It is currently maintained on a UNIVAC 1100 series hardware operating under e EXEC 8. (_3) fo validate the code, a cylindrical concrete tank on soil was analyzed and results were compared to results obtained from the DYNAX program (S&L Program No. 09.7.083-7.0). The structural model of the tank is shown in Figure A-24(a) and the concrete and soil properties are shown in Table A-4(a). Results were compared for a time history pressure load symmetrically applied about the 0 = 00 meridian. The meridional distribution of this pressure is shown in Figure A-24(b) and the time history distribution is shown in Figure A-25. The Fourier coefficients for circumferential distribution of pressure are shown in Table A-4(b). As shown in Tables A-5 and A-6, the results obtained from the two programs for nodal accelerations at nodes A and B in Figure A-24(a) and for maximum stress resultants at element C in Figure A-24(a) compare favorably. Results were also compared for the same load distribution applied about the e = 00, e = 1200 and e = 2400 meridians simultaneously. Tables A-7 and A-8 show a favorable comparison of the results. A.2-1
2PS-1-MARK II DAR AMENDMENT 13 OCTCBER 1980 () A.3 LUSH LUSH (Complex Response Analysis of Soil-Structure Interaction) is a finite element program designed for the seismic analysis of soils and structures. Unlike other dynamic soil response programs LUSH has frequency independent damping. This makes it especially appropriate for analyses requiring accuracy in the high frequency range such as generating foundation response spectra. LUSH solves plane strain problems excited by an acceleration time history specified at the rigid base of the model. The proper strain dependent elastic moduli and damping values of each soil element are determined by iteration until compatibility between maximum principal shear strains and properties is obtained. Most of the analysis is performed in the frequency domain using the method of complex response with complex moduli. First the stiffness matrix is formulated using the complex moduli given by the equation G* = G (1-2p2 +2 i a /1 - ,2) where for a given element G* is the complex moduli, G is the shear modulus and a is the damping ratio. This accounts for (~3 frequency independent viscous damping. Next, using the complex k/ stiffness matrix the nodal point ecaations of motion are formulated. The input acceleratica time history is converted to the frequency domain by a Fourier transform. Then the equations of motion are solved for each term of the Fourier transform and the results superimposed. This gives the Fourier transforms of all the nodal point displacement histories. From these displacements the Fourier transforms for stress, strain, and acceleration histories can be determined. Finally the stresses, strains, and accelerations are converted to the time domain. This process is repeated for each iteration on the sell properties. The specific output of the analysis is selected by tPt aer. Output relating to elements includes maximum stresses, maximum principal shear strain, strain consistent shear modulus and strain consistent damping. Nodal point output includes maximum acceleration, acceleration time histories, and response spectra. LUSH was developed at the University of California, Berkeley by Lysmer, Udaka, Seed, and Hwang (Reference 9). Sargent & Lundy modified the program and now maintains it for use on the EXEC 8 processor of the UNIVAC 1100 computer. The Sargent & Lundy version of LUSH was validated by comparison , with a problem presented in Reference 9. This problem is quite ' (s) simple but employs all the capabilities of the program. A.3-1 , l l l
.=-- - -. - -. .- . . .. . . --._ .- . - . - - .
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 - i O rae vreble= is outlined in F19ure A-26. soil properties used are presented in Table A-9. The comparisons The strain dependent
- between the S&L version of LUSH results and the results reported 1 in Reference 9 are given in Tables A-10, A-11, and A-12 and j Figure A-27. The agreement is very good.
r I i i e t 4
- O i
( l I e i O
' A.3-2
. - . ~ , , , , _. - -..--..,--.,.y---.----,,,~..,e,,---.: .- ...-.--,..e,,, ...,,3,-..,, ,w.....,.,y.--- .w..,,w,-.r.'
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () A.4 PIPSYS PIPSYS (Integrated Piping Analysis System) analyzes piping systems of power plants for static and dynamic loadings, and computes the combined stresses. The following analyses are performed:
- a. Static: Analysis of thermal, displacement, distributed, and concentrated weight loadings on piping systems;
- b. Dynamic: Analysis of piping system response to seismic and fluid transient loads;
- c. Stress Combination: Computes the combined stresses in the piping components in accordance with the ASME Boiler and Pressure Vescel Code, Section III (Reference 10).
The static, dynamic, and stress combination analyses can be per-formed independently or in sequence. Results of the static and dynamic analyses can be stored on magnetic tape for use at a later date to perform the stress combination analysis. The piping configuration can be plotted on a CALCOMP plotter. s The input consists of the piping system geometry, material pro-s_) perties, static and dynamic loadings. Various options exist to control the length of the output. The default option generally prints only the summary of input data and final results. PIPSYS was developed at Sargent & Lundy in 1972. It is currently maintained on UNIVAC 1100 series hardware operating under EXEC 8. To demonstrate the validity of the PIPSYS program the following three examples are presented. To illustrate the validity of the static portion of PIPSYS, the problem shown in Figure A-28 was analyzed and the results compared to those given in Reference 11. Table A-13 shows the comparison of member end moments. As shown, the results from PIPSYS and Reference 11 are in good agreement. To illustrate the validity of the stress combination analysis portion of PIPSYS, the problem outlined in Reference 12 was reanalyzed on the PIPSYS program. The layout of the piping system is shown in Figure A-29. The stress analysis is performed at location 19. The summary of loads sets and descriptions are presented in Table A-14. The results of the stress analysis are presented in Tables A-15 and A-16. The notations and equation numbers correspond to the ASME Boiler and Pressure Vessel Code (Reference 10). U,_, A.4-1
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 ; () It is' observed that the PIPSYS results are very close to those presented in Reference 12. To illustrate the validity of the dynamic analysis portion of
! PIPSYS, a problem was analyzed and the results obtained from l PIPSYS were compared with those from two public domain computer ; programs, DYNAL (Reference 13) and NASTRAN (References 14 and ; 15). ; Figure A-30 shows a schematic representation of the piping system
- analyzed. The system is modeled with simple beam elements with a l total of 136 degrees of freedom. Figure A-31 shows the time-j dependent blow-down forces at the relief valves locations.
; Results of PIPSYS are compared with DYNAL and NASTRAN in Table
- - A-17 and Figure A-32. The results from all three programs are i quite close.
i l l I l F O
- 1 l
1 a i rO A.4-2 1
- - . - . . .----,-,n . . . . . . , . . - - - ~ , - . , . -......---,--r.-~,-,.. .----,-,,.-...-,,.--~.,,-------.--w -w<
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
- () A.5 RSG RSG (Response Spectrum Generator) generates dynamic response 1 spectra (displacement, velocity, and acceleration) for a single-1 degree-of-freedom elastic systems with various dampings, subjected to a prescribed time dependent acceleration. The program may also be used to obtain a response spectrum consistent acceleration time history in which the response spectrum of the generated acceleration time history closely envelops the given spectrum. The differential equation of motion is solved by Newmark's p-method of numerical integration (Reference 16).
The program has the capability to apply a baseline correction in an earthquake acceleration time history as well as to obtain and to plot the Fourier transform of the given acceleration time his-tory. Options are available to obtain plots of the givan acceleration time history, the generated response spectra and
- their envelopes. In addition, the response spectra can be combined using the probability method, square-root-of-the-sum-of-the-squares (SRSS) and absolute sum methods. An interpolation option to obtain an acceleration time hist 9ty at equal intervals or at a smaller time interval is available. The program can also be used as a postprocessor for other programs with all its options and capabilities.
Depending upon the option, the program output includes the s/ response spectrum, the Fourier transform of a given acceleration time history, or the response spectrum consistent acceleration time history. RSG was developed by Sargent & Lundy in 1969. Since 1972, the program has been maintained on UNIVAC 1100 series hardware operating under EXEC 8. To illustrate the validity of the program, three sample problems are presented. In the first problem, response spectra of the El Centro north-south earthquake record (53-76 seconds duration) are generated using RSG for 0, 2, 5, 10 and 20 percent damping ratios. A time history plot of the earthquake record from RSG and as published in Reference 17 is shown in Figure A-33. A comparison of response spectrum values obtained from RSG with the response spectrum values published in Reference 18 is shown in Table A-18. Comparisons of response spectra plots at varying dampings are shown in Figure A-34. As shown by the comparison, results obtained from RSG are accurate. The second validation example is a Fourier transform plot of a l given 5 cycles /sec sine wave time history from RSG. The Fourier transform plot shown in Figure A-35 shows a peak only at 5 cycles /sec. For the third validation problem, a spectrum-consistent time O-' history was generated. A comparison of the desired response A.5-1
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 t 4 i O spectro and ene reseense s ,ectrum or tae co=9atiste ti e historv is shown in Figure A-36. As seen from tais figure, a good match is obtained. i i 9 i 1 e O 1 a l 4 O i , I 1 ) A.5-2 l _ _ _ _ . _ _ , _ _ - . _ _ _ _ _ . . , ._,__._._,.._.,,.__,_,,_,,_,_.-_,..m,__..__,,,____.,,,,_..,_,.,,,,,,_,, , , , . _ . _ , . _ , _ _ , _ , , _ . , , . _ _
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 A.6 SHAKE (]) SHAKE (Soil Layer Properties and Response / Earthquake) computes response in a horizontally layered semi-infinite system subjected to vertically traveling shear waves. Strain-compatible soil properties are computed within the program. Earthquake motion can be specified at any level of the soil profile and a resulting motion can be computed anywhere else in the profile. The method is based on the continuous solution of the shear wave equation. For soil liquefaction studies, plots of stress time histories at various levels in a soil profile can also be obtained. The input for the program includes data for the soil profile, curves of strain versus shear moduli and damping ratics, and the input earthquake motion. The output includes the strain-compatible soil properties, response spectra of object and computed motions and printer and CALCOMP plots of time histories, Fourier spectra and response spectra. Stress time history plots are also included. SHAKE was originally developed by P. B. Schnabel and J. Lysmer of the University of California, Berkeley (Reference 19). It was modified and has been maintained by Sargent & Lundy since October 1972 on the UNIVAC 1100 series hardware operating under EXEC 8. () To verify the SHAKE program, the results from SHAKE and the public domain progiam OUAD4 (Reference 20) were compared for a typical proolem. QUAD 4, a finite element program, uses a step-by-step integration technique in the time domain to solve the two-dimensional discrete equations of motion; SHAKE uses a numerical solution in the frequency domain to solve the one-dimensional wave equation. For the comparison, it was necessary to impose suitable boundary conditions on the finite model for the OUAD4 analysis to ensure only one-dimensional wave propagation. The problem solved by SHAKE and compared with the OUAD4 results analyzed the seismic response of a 100-foot layer of dense sand (Figure A-37). The properties of the sand are given as: Total unit weight = 125 pcf (K2) max = 65 K o = 0.5 4 The parameter (K2) max relates the maximum shear modulus, 4 axi and effective mean pressure at any depth, y, below the surface. Gmax = 1000(K2) max 6 m
- where A.6-1
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (] 6,
=
(1 + 2Ko) 6. 3 v Kg = Ceafficient of lateral pressure at rest 6' = Effective vertical pressure at depth y. Damping values and the variation of modulus values with strain were based on published data for sands (Reference 21). i The response of the sand layer was evaluated using the time history of accelerations recorded at Taft, California during the 1952 Kern Courity earthquake as base excitation. The ordinates of this time history were adjusted to provide a maximum acceleration of 0.15g. The maximum shear stresses and accelerations from SHAKE and QUAD 4 are compared in Figure A-38 and the response spectra of the sur-face motions are compared in Figure A-39. As illustrated in these figures the two solutions compare favorably. O i O A.6-2
2PS-1-MARK II DAR AMENDMENT 13 - ! OCTOBER 1980 ' (]) A.7 TEMCO' TEMCO (Reinforced Concrete 'ections Under Eccentric Loads and Thermal Gradients) analyt u reinforced concrete sections - subjected to the separate 3r combined actions of eccentric and thermal loads. The program san be used to analyze sections of reinforced concrete beams and columns, slabs, and containment structures subjected i various combinations of thermal and nonthermal (axial, u..taxial and biaxial bending) loads. The effect of temperature is induced in the section by reactions to the deformation restraint. No thermal loads can be specified when analysis under axial force and biaxial bending is desired. A linear distribution of strains across the section is assumed. i The analysis may be done with either a cracked or an uncracked concrete section. Material properties car; be either linear or nonlinear. The program can handle rectangulor as well as non-rectangular sections. The effect of the thermal expansion of the liner on a concrete section can be determined assuming the liner has no strength. Temperature-dependent material properties, composite section and nonlinear temperature variation across the section are allowed. In the cracked analysis, only the loads input as " thermal" are assumed to be of the relaxing-type. All other loads are assumed to be of the sustained-type and are unaffected by concrete cracking. For sustained-type loads (referred to in the program O as " mechanical" loads), the tiddle surface strain and the curvature change are determined by an iterative procedure such
> that equilibrium is satisfied.
For the general type composed of sustained-and relaxing-type loads, thermal moment is relaxed by cracking so that no further change in curvature occurs. When the thermal force is assumed completely relieved by free thermal expansion, the middle surface i strain is determined by an iterative procedure such that i equilibrium is satisfied. When the thermal force is also assumed relaxed by cracking, the actual middle surface strain and curvature are given and the corresponding stress resultants give the relaxed force and moment. Input consists of section dimensions, the area and location of
- , each layer of reinforcing steel, loads, load combinations, and l material properties.
Output consists of an echo print of the input, combined loads, j final location of neutral axis, final stresses in steel and 1 concrete, and final internal forces. Similar intermediate l results (before thermal load is applied) can also be output if , desired.
- The program was developed in 1972 and is maintained by Sargent &
Lundy on UNIVAC 1100 series hardware operating under EXEC 8. {} A.7 l
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 To validate TEMCO, program results were compared with hand cal-('>T culated results. Six example problems are presented. The section and material properties for each problem, along with the
- applied nonthermal and thermal loads, are given in Tables A-19, A-21, A-23, and A-25.
- The first problem involves a section with two layers of steel subjected to a compressive force applied at the centerline of the section, a bending moment and thermal gradient. A cracked analysis of the section is required assuming nonlinear material properties.
The second and third problems each involve a section with two layers of steel subjected to a tensile force applied at the centerline of the section, a bending moment and a thermal gra-dient. A cracked analysis of the section is required assuming nonlinear material properties for the second problem and linear for the third problem. The fourth problem involves a section with 10 reinforcing steel bars subjected to a tensile force and biaxial bending. A cracked analysis of the section is required assuming nonlinear material properties. The fifth problem involves a section with two liners (one on each side) subjected to nonthermal and thermal loads. A cracked () analysis of the section is required assuming nonlinear material properties. . The sixth problem involves a section subjected to nonlinear tem-j perature variation. A cracked analysis of the section is j required assuming nonlinear material properties. Hand calculated solutions were obtained according to the 1 following outlined procedure: Assume that the location of the neutral axis and the o;ess distribution are identical to tlose given by the program under the given mechanical loading. Compute strain distribution under the given mechanical loading. i
- Compute the stress resultants by integration, using the proper stress-strain relationships.
Check for equilibrium with external mechanical loads. If equilibrium _is satisfied, compute the deformation imposed on the section by the given thermal load. ! Compute the final deformations by subtracting the thermal deformations from the mechanical deformations. [} A.7-2
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () For free thermal expansion, compute the new axial strain such that equilibrium is satisfied, and the curvature is kept constant. Compute the final stress resultants by integration, using the proper stress-strain relationships.
~
Compute the thermal loads. Check for equilibrium and compare program results with hand calculated results. Results obtained using this procedure together with those computed by TEMCO for all six problems are presented in Tables A-20, A-22, A-24, and A-26. It is concluded that results given by the program agree very well with results obtained by hand calculations and that equilibrium between internal and external forces is satisfied for all six problems. ~ (E) O A.7-3
2PS-1-MARK II DAR AMENDMENT 13 nCTOBER 1980 () A.B EYSEA A.8.1 Ijtroduction The suppression pool dynamic loads, annulus pressurization, and seismic events all impart primary loads to the containment struc-tures and secondary accelerations to the reactor building equipment. The calculations of secondary accelerations on the equipment are based on a collective development process between Sargent & Lundy (S&L) and General Electric (GE). Structural system response data are developed by S&L u ing a com-posite soil-s?.ucture interaction model with a representation of the reactor pressure vessel. Resulting acceleration time histories are supplied to GE for a local system analysis. The local system analysis is based on a composite lumped-mass model of the pedestal shield wall and a detailed representation of the reactor pressure vessel complex. (See Figure A-40.) The excitation inputs for this local system analysis are acceleration time histories for the case of suppression pool hydrodynamic loadings and seismic vibratory motions, and pressure time histories for the dynamic analysis of annulus pressurization due to postulated pipe breaks. (~ A.8.2 Development of DYSEA Computer Program V) The local system analysis is conducted using the DYSEA computer ' program, a GE proprietary program modified from a well-known structural analysis program, SAP-IV, for specific application to the dynamic and seismic analysis of reactor pressure vessels and internals coupled with supporting structures. In this modified program, the static analysis option and those elements other than truss and beam elements were deleted. The direct integration method of dynamic analysis was also deleted. The finite ~ element techniques to formulate the stiffness matrix, the eigerva]ue solution techniques, and the mode superposition metheJ and the response spectrum method of dynamic analysis remain the same as those used in the SAP-IV program. The new capabilities added to this modified program are a global spring alement, coupling or off-diagonal mass imput, rotational ground acceleration input, acceleration response output, multiple support excitation, and response combinatioa methods used in the response spectrum method. (The response spectrum method was not used in the analysis of the 2PS-1 reactor pressure vessel and internals.) A.8.3 Program Versions
/~N The first version of the DYSEA program, DYSEA01, was developed, verified, and reviewed in the summer of 1976. It has been \) adopted as one of GE's major structural analysis programs since l
A.8-1 L_________----______________________________________________________________________________________________
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () 1977. time. Several improved versions have been developed since that DYSEA04 and DYSEA05 are the two versions utilized in local system analysis of the ZPS-1 plant. DYSEA05 differs from DYSEA04 primarily in the use of an improved formulation of the multiple support excitation solution to eliminate a numerical sensitivity problem encountered in some applications. Test cases were run to confirm that this problem was not encountered in applications of DYSEA04 to local system analysis of the ZPS-1 plant. A.R.4 Program Verification A large number of verification test cases have been developed to check each of the DYSEA program features. The off-diagonal mass and the global spring features have been shcwn to agree precisely , with exact solutions. A total of eight test cases verify correct
- numerical eigenvalue solutions by comparison with exact solutions. For verification of large scale eigenvalue solutions, an internal eigenvector orthogonality check has been programmed into DYSEA.
Thirteen separate test cases provide verification of the multiple support excitation feature. Results for small size cases were compared with paown analytical solutions with very good agreement. Large scale multiple. support excitat' ion solutions have been checked by independent solution approaches internal to
, w the program. Very good agreement was obtained. These test cases include checks on the translational and rotational ground excitation capabilities of the program.
A.8.5 Test Problems Problem 1: The first test problem involves finding the eigenvalues and eigenvectors from the following characteristic equation: (u2 [M] - [K]) {X} = 0 where u is the circular frequency, x is the eigenvector, and [K] and [M] are the stiffness and the mass matrices respect i"etv and i are given by
~
~/ 4 ) 4 4 Fra (1-72)7 f 4 ) 4
[M} = 1 1- 92 1 72 ( / Symmetric fl - 4h l 25r2 , ! - ( l- l a A.8-2
. _ . w.., , - - - - --wm, m. -r r=~- n ---r'- ---------------t~~
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
~
F/
\
(}' [1 + wa) ( 4 3 5 9
)
[K] = , + 9,a l 15 4 ( () Symmetric 1 1 + 25r2
)
1
\ 4 /_
The analytical solution and the solution from DYSEA are as follows: 4
- a. eigenvalues ui:
DYSEA SOLUTION ANALYTIC SOLUTION 1 5.7835 5.7837 2 30.4889 30.4878 3 75.0493 75.0751
- b. eigenvectors + : -
DYSEA SOLUTION ANALYTIC SOLUTION 1.000 1.000 1.000 1.000 1.000 1.000
-0.0319 -1.5536 -1.2105 -0.0319 -1.554 -1.211
() ,
-0.0072 -0.0666 2.0271, _-0.0072 -0.0666 2.027, j Problem 2: The second test problem compares the dynamic ; responses of the reactor pressure vessel and internals / reactor building subjected to earthquake ground motion.
The mathematical model of the reactor pressure vessel and internal / reactor building is given in Figure A-40. The inputs in the form of ground spectra are applied at the basemat level. Response spectrum analysis was used in the analysis. Natural frequencies of the system and the maximum rengnunt at key locations have been calculated by both DYSEA and SL.Nis. Result comparisons are given in Tables A-27 and A-28. It can be seen that the results calculated by DYSEA agree closely with those obtained by SAMIS. O A.8-3 .
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (]) A.9 SAP-IV General Electric an11yzes the LUMPED-MASS: models of the primary piping systems (mail: steam and recirculation) using a proprietary version of the SAP-IV computer program. Dynamic analyses of the piping system are performed using acceleration response spectra at each attachment point, i.e., distinct excitations are specified at each piping support and anchor point. The eqnstions of motion Lsed in the formulation of the multiple excitation methods conform to the fundamental laws of mechanics. The theory is based on the same assumptions used for the uniform excitation methods, that is: there is small system damping, the damping matrix is orthogonal, and it is applicable for linear systems. The technical community has generally accepted the multiple excitation method. The procedure has been described briefly in textbooks, and procedural details were presente3 in a GE publication at the Fourth SMIRT Conference, August 1977. The use of the multiple excitation method was accepted by the NRC staff on the GESSAR docket. The PISYS program, which is the SAP-IV program with simplified input format, has been verified by comparing predictions with NRC benchmark problems. Reference 22 documenting the agreement between the NRC benchmark and GE calculations has been transmitted to and approved by the NRC. () C) A.9-1
i 2PS-1-MARK II DAR AMENDMENT 13 i OCTOBER'1980 A.10 PENAN' 1
~Q PENAN (Penetration Assembly Stress Analysis Finite Element 4
Program) is an axisymmetric finite element program which performs , , . all necessary structural, thermal, and fatigue analyses required i for mechanical penetration assemblies. PENAN handles structural, thermal and fatigue analyses of axially ] ^ symmetric solids of revolution, composed of orthotropic materials with temperature-dependent properties, and subjected to i asymmetric and time-dependent heating and loading, The program has automatic finite element mesh generation for various penetration assembly configurations. It has a built-in Section l
- III material property bank. It forms optinii load and load-range.
{ combinations for the various Section Ill specified loading categories. It calculates the allowable stresses for all stress
- categories, makes the necessary stress comparisons, and generates q the entire penetration assembly stress analysis report.
1 i The structure of PENAN is a combination of two suitably modified I- at:isymmetric finite element programs. These programs are: I ! a. NOHEAT (Nonlinear Heat Transfer Analysis) by I. Farhoomand and E. L. Wilson, Structural Engineering Laboratory,_ University of California, Berkeley,
- California. Report Number UC1SESEM 71-6, April 1971.
- b. ASAL (Finite Element Analysis-Axisymmetric Solids !
with Arbitrary Loads) by R. S. Dunham and R. E.
; Nickell, Structural Engineering Laboratory, l University of California, Berkeley, California, Report Number 67-6, June 1967. ,
1 i i s l A.10-1
i-2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () A.11 PESSAL PESSAL (Penetration Design Program) is designed to assist the engineer in the design of mechanical penetration assemblies. It calculates the faulted condition loads, the required minimum sleeve thickness, determines the cooling coil requirements, and generates three different reports, the Penetration Sleeve Design . Report, the Penetration Assembly Design Report and the Penetraticr Cooling Coil Ersluation Report. In addition, the program has, at the option of the user, the capability to select the sleeve outer diameter and nominal wall thickness. O A.11-1
. . - - . - , , .- _ _ - . -y ..-_y , , - . ,, , , , . . , , , , , . , , , , , .. .,_,.,.,.__...~.9 ._
y._ , .
l 2PS-1-MARK II DAR AMENDMENT 13 ! OCTOBER 1980 i () A.12 References
- 1. S. Ghosh and E. Wilson, " Dynamic Stress Analysis of Asisymmetric Structures Under Arbitrary Loading," Report No.
EERC 69-10, University of California, Berkeley, September 1969.
- 2. S. Timosh?nko, " Theory of Plates and Shells," McGraw-Hill, New York, 1940,
- 3. S. I'. Timoshenko and J. N. Goodier, " Theory of Elasticity,"
McGraw-Hill, New York, 1951.
- 4. B. Budiansky and P. P. Radkowski, " Numerical Analysis of Unsymmetric Bending of Shell of Revolution," Journal of the American Institute of Aeronautics and Astronautics, August 1963.
- 5. H. Retsinonn and J. Padlog, " Forced Axisymmetric Motions of Cylindrical Shells," Journal of the Franklin Institute, Vol. 284, No. 5, pp. 308-319, November 1967.
- 6. S. Klein, "A Study of the Matrix Displacement Method Ac Applied to Shells of Revolution," Proceedinas, Conferenr2 on Matrix Methods in Structural Mechanics, Wright-Patterscil Air Force Base, Ohio, 1965.
- 7. J. F Abel, P. P. Cole, and D. P. Billington, " Maximum h,3/ Seismic Response of Cooling Towers," Report No. 73-SM-1, Department of Civil and Geological Engineering Research, Princeton University, March 1, 1973.
- 8. "A Program to Perform Stress Analysis of Shells of Revolution," Knolls Atomic Power Laboratory, Schnectady, New York, September 1963.
- 9. J. Lysmer, T. Udaka, H. B. Seed, and R. Hwang, " LUSH - A Computer Program for Complex Response of Soil-Structure Systems" EERC Report No. 74-4, College of Engineering, University of California, Berkeley, April 1974.
- 10. ASME Boiler and Pressure Vessel Code, Section III, 1974.
- 11. J. S. Kinney, " Indeterminate Structural Analysis," Addison-Wesley Publishing Company, Reading, Massachusetts, p. 377, 1975.
- 12. " Sample Analysis of a Piping System Class 1 Nuclear,"
prepared by Working Group on Piping of the Design Subgroup of the Nuclear Power Committee of the ASME Boiler and Pressure Vessel Committee, the American Society of Mechanical Engineers, New York, 1972.
- 13. ICES DYNAL User's Manual, McDonnell Douglas Automation Co.,
September 1971. (]} A.12-1
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
- 14. NASTRAN Theoretical Manual, NASA SP-221, September 1970.
(]} i
- 15. NASTRAN User's Manual, NASA SP-222, September 1970.
- 16. N. M. Newmark and E. Rosenblueth, Fundamentals of Earthquake Encineerino, Prentice-Hall, Inc., Englewood Cliffs, N. J., p. 15, 1971.
- 17. Strong Motion Earthquake Accelerograms, Digitized and Plotted Data, Vol. II - Corrected Accelerograms and Integrated Ground Velocity and Displacement Curves, Part A - Accelerograms IIAOUI through IIA 20, California Institute of Technology Earthquake Engineering Research LaDGratory, EERL 71-51, Pasadena, California, September 1971. j
- 18. A. G. Brady, et al., " Analysis of Strong Motion Earthquake Accelerograms, Volume III, Response Spectra, Part A, Accelerograms IIA 001 through IIA 020," Prepared for the National Science Foundation, August 1972.
- 19. P. B. Schnabel and J. Lysmer, " SHAKE: A Computer Program for Earthquake Response Analysis of Horizontally Layered Sites,"
Report No. EERC 72-12, Earthquake Engineering Research Center, Univecsity of California, Berkeley, December 1972.
- 20. I. H. Idriss, et al., "00AD-4, A Computer Program for
() Evaluating the Seismic Response of Soil Structures by Variable Damping Finite Element," Report No. EERC 73-16, rarthquake Engineering Resaarch Center, University of California, Berkeley, July 1973.
- 21. H. B. Seed and I. M. Idriss, " Soil Moduli and Damping Factors for Dynamic Response Analyses," Report No. EERC 70-10 Earthquake Engineering Retearch Center, Univers'ity of California, Berkeley, December 1970.
- 22. PIPSYS Analysis of NRC Benchwork Problems, NEDO 24210, August lo79.
O A.12-2
l g - TABLE A-1 CO!1PARISON OF DISPLACEMENTS AND FORCES FROM DYNAX AND SOR III (DYNAX) SOR-III DYNAX R-Dis- Hoop Meridional R-Dis- Hoop Meridional Z placement Rotation Force Moment placement Rotation Force Moment O. .3653-1 .6409-1 -913.4 4.43 .3723-1 .6511-1 -930.8 6.56
- 1. .1982-1 .4196-1 495.6 -42.95 .1952-1 .4256-1 488.0 -48.63
- 2. .46049-1 .1207-1 1151.2 -33.19 .462-1 .1222-1 1155.0 -32.84 y
? ?
H' w
- 3. .46049-1 .1207-1 1151.2 -33.19 .462-1 .1222-1 1155.0 -32.84 H i
- 4. .1982-1 - . 4196-1 495.6 -42.95 .1952-1 .4256-1 488.0 -48.63 h R
- 5. .3653-1 .6409-1 -913.4 4.43 .3723-1 .6511-1 -930.8 6.56 g H
O lc 8E 4 t 8s NE 1 lc M
.Hk OW I
O O O r TABLE A-2 VELOCITY IN THE Z-DIRECTION AS COMPUTED BY DYNAX AT TIME 2.2 SECONDS AT TIME 4.0 SECONDS Z-VELOCITY Z-VELOCITY NODE Z-ORDINATE NONREFLECTING NODE Z-ORDINATE NONREFLECTING 1 0. -10.1 1 0. -10.5 3 1. -12.8 3 1. -10.7 5 2. -12.2 5 2. - 8.92 s
- E e 7 3. - 9.22 7 3. -10.5 i Y T N 11 5. -10.4 11 5. - 9.44 g 15 7. - 9.2 15 7. - 9.20 H
- 8.92 17 8. -10.4 -"
17 8. M 19 9. -11.4 19 9. -10.5 23 11. - 9.12 23 11. -10.7 27 13. - 9.14 27 13. -10.7 8g am 31 14. -10,4 31 14. - 9.75 @$ mx 33 15. -10.2 33 15. - 9.92 *E 5" 37 17. -11.0 37 17. - 9.62 g[ 41 20. -11.2 41 20. - 9.58
- _ . ..~;,-.. _ . - . - ~ . - . - - . - . . . _ - -.. -.- . . . - . ~ . .
ZPS-1-MARK II DAR AMENDMENT 13 i -OCTOBER 1980 i 4 4
- TABLE A-3
) MODEL DAMPING COMPARISON l
i t i j HAND ! MODE DYNAX CALCULATIONS t i ! l 0.0352 0.0348 i i 4 1 2 0.0368 0.0367 ! l I
- 3 0.0430 0.0430 r
i I i h I t 1 P O . l i l 4 I e l 1 i o A.13-3 l 1 I I i l
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 0 TABLE A-4 PROPERTIES OF STRUCTURAL MODEL (a) MATERIAL PROPERTIES DENSITY YOUNG'S kip sec 2 MODULUS POISSON'S MATERIAL ft* (kip /ft2) RATIO Concrete 0.00466 584000 0.17 Soil 0.00420 2351.5 0.42 i * (b) CIRCUMFERENTIAL DISTRIBUTION OF LOAD FOURIER HARMONIC NUMBER 0 1 2 3 4 Coefficient .2644 .3927 .1836 .0499 .0386 O A.13-4
_. . _ _ . - - . . . ~ . _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ . . _ . _ _ . _ ZPS-1-MARK II DAR AMENDMENT 13 CCTOBER 1990 TABLE A-5 j COMPARISON OF NCDAL ACCELERATIONS IN G UNITS (a) MAXIMUM RADIAL ACCELERATION I NODE DYNAX FAST I TIME TIME l (sec) ACCLN. (sec) ACCLN. ! A 0.012 8.60 0.0.1.2 8.60 B 0.132 -21.09 0.132 -21.10 J 6 ! (b) MAXIMUM VERTICAL ACCELERATION
. NODE DYNAX FAST it
- TIME TIME I
(sec) ACCLN. (sec) ACCLN. l 4 A 0.171 -17.20 0.171 -17.21 !C:) B 0.135 12.58 0.135 12.58 ' ) 1 I i. I i i I 1 i T A.13-5
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
-( ) TABLE A-6 COMPARISON OF MAXIMUM STRESS RESULTANTS IN K, FT UNITS AT ELEMENT C, 0 = 0' DYNAX FAST TIME FORCE TIME FORCE
- COMPONENT (sec) K, ft (sec) K, ft 1
Meridional membrane force 0.10 - 9.57 0.10 - 9.57 Circumferential membrane force 0.10 - 9.79 0.10 - 9.79 Meridional moment 0.23 15.67 0.23 15.67 Circumferential moment 0.10 - 5.97 0.10 - 5.97 4 Meridional transverse
- l shear 0.10 -15.12 0.10 -15.12
() a i O A.13-6
, . - . - r, , . . - . . , . , . _ _ , . -.. -w, , , . , , _ , , _ , p.w,.-, ,,--,g. %,y_ _ , . . . , _ , , . , , . .y..- . - , - _ _ _ , .-r, -we
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 0 TABLE A-7 COMPARISON OF NODAL ACCELERATIONS IN G UNITS (a) MAXIMUM RADIAL ACCELERATION NODE DYNAX FAST TIME TIME (sec) ACCLN. (sec) ACCLN. A 0.012 8.73 0.012 8.74 B 0.132 -21.40 0.lr -21.41 (b) MAXIMUM VERTICAL ACCELERATION NODE DYNAX FAST TIME TIME (sec) ACCLN. (sec) ACCLN. A 0.171 -1745. 0.171 -1747 B 0.135 1276. 0.135 1277. O A.13-7
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
/ TABLE A-8 COMPARISON OF MAXIMUM STRESS RESULTANTS i
. IN K, FT UNITS AT ELEMENT C, 0 = 0* DYNAX FAST TIME FORCE TIME FORCE COMPONENT (sec) K, ft (sec) K, ft Meridional membrane force 0.10 - 9.71 0.10 - 9.71 circumferential - i membrane force 0.10 - 9.94 0.10 - 9.94 , Meridional moment 0.23 15.44 0.23 15.44 Circumferential moment 0.10 -36.81 0.10 -36.81 Meridional transverse shear 0.10 - 9.06 0.10 - 9.06
)
2 v 4 h h e O i ! A.13-8
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 ( TABLE A-9 STRAIN-COMPATIBLE SOIL PROPERTIES EFFECTIVE SHEAR MODULUS FRACTION OF CRITICAL SHEAR STRAIN REDUCTION FACTOR
- DAMPING (%)
Yeff (%) CLAY SAND CLAY SAND
-4'
< l. x 10 l.000 1.000 2.50 0.50
-4 3.16 x 10 0.913 0.984 2.50 0.80
-3 1.00 x 10 0.761 0.934 2.50 1.70
-3 3.50 3.20 3.16 x 10 0.565 0.826
-2 1.00 x 10 0.400 0.656 4.75 5.60
-2 3.16 x 10 0.261 0.443 6.50 10.0
-1 1.00 x 10 0.152 0.246 9.25 15.5 0.316 0.076 0.115 13.8 21.0 1.00 0.037 0.049 20.0 24.6 3.16 0.013 0.049 26.0 24.6
> 10.00 0.004 0.049 29.0 24.6 J
*This is the factor which has to be applied to the shear modulus at low shear strain amplitudes (here defined as 10 _4 percent) to obtain the modulus at higher strain levels.
/
($) b l' A.13-9
. ~ . ~ ,. _ - . . _ - . . , _ . _ . _ _ . . . . _ _ . - . - _ - .. _ _. ... .~ . - . -
l ZPS-1-MARK II DAR , AMENDMENT 13 OCTOBER 1980 - TABLE A-10 COMPARISON OF COMPUTED SOIL PROPERTIES DUE TO HORIZONTAL EXCITATION SHEAR MODULUS DAMPING RATIO ELEMENT G ksf A% NUMBER REF. 1 LUSH REF. 1 LUSH k 2 1537. 1512. 8.6 8.7 3 1409. 1388. 8.4 8.5 4 840. 828. 7.8 7.9 5 774. 763. 7.8 7.9 i O I r T i O A.13-10 4
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () TABLE A-ll COMPARISON OF STRESSES DUE TO HORIZONTAL EXCITATION psf o y psf T xy psf U ELEMENT x NUMBER REF. 1 LUSH REF. 1 LUSH REF. 1 LUSH 1 110.8 111.4 158.9 157.1 377.1 373.4 2 120.5 118.3 79.8 78.3 509.2 505.8 3 28.5 28.4 29.9 29.8 443.0 440.7 l
,4 15.8 15.3 23.3 22.8 696.8 692.2 5 39.7 38.9 42.1 41.3 648.8 644.6 i
i O r .i 1 O
\
A.13-11
O O O J TABLE A-12 COMPARISON OF NODAL POINT ACCELERATIONS DUE TO HORIZONTAL AND VERTICAL EXCITATIONS HORIZONTAL EXCITATION VERTICAL EXCITATION Acc. g Y Acc. g X Acc. g Y Acc. g NODAL POINT NUMBER REF. 1 LUSH REF. 1 LUSH REF. 1 LUSH REF. 1 LUSH 1 .1849 .1835 Fixed Fixed .1642 .1634 l 2 .2142 .2119 .0121 .0116 .1392 .1370 .2084 .2046 - 3 .1723 .1715 Fixed Fixed .1669 .1659 1
' H 4 .1444 .1443 .0000 .0000 .0000 .0000 .1322 .1299 i
> h
.g 5 .1444 .1443 Fixed Fixed .1322 .1299 @
e C .1646 .1630 Fixed Fixed .1170 .1165 U w O 7 .1708 .1694 .0050 .0049 .0547 .0572 .1101 .1085 > w 8 .1855 .1842 Fixed Fixed .1068 .1051 f 5 O 2 em i OZ tD O M 3: I' . ME Hn.
; w
@H OW il
'N l
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 , O TABLE A-13 COMPARISON OF MOMENTS FOR SELECTED MEMBERS MOMENTS FROM MOMENTS FROM i REFERENCE 2 PIPSYS
! (kip-ft) (kip-ft)
MAB 06.0 102.8 i M 33 72.0 72.5 M 133.0 131.8 BC M CB 133.0 131.8 M CD -133.0 -131.8 M -133.0 -131.8 DC MDE 133.0 131.8 M 86.0 84.2 ED i M BE -158.0 -156.6 M EB
-158.0 -156.6 bE 106.0 102.8 M yy 72.0 72.5
'l e O ; j l A.13-13. - . . . . _ - _ . - - . . . . - - - - - _ . . . . - ..--
_ m . . . _ _ _ _. . _ _ . . _ _ _ _ _ . _ . . . _
=
i y v TABLE A-14
SUMMARY
OF LOAD SETS AT GIRT!! BUTT WELD WITil CHANGE IN MATERIAL AND WALL THICKNESS (PIPSYS) LOAD NO. T P My My AT a b SET NO. LOAD SET DESCRIPTION T RA% , "S M, l (VALVE) (PIPE) 1 Zero 0 0 0
. 0 0 70 70 0 2 Cold liydro Test 3590 0 0 0 0 70 70 0 3 liot Hydro Test, Up 40 2200 251.7 141.6 -7.1 2.4 400 400 0.3 4 Hot !!ydro Test, Down 0 0 0 94 0 -2.4 70 -0.3 5 Plant Startup 100 2200 337.2 184.9 -936.0 0 70 70 0 6 Plant Shutdown 0 0 0 0 0 70 70 0 I'
t, 7 Plant Loading 18300 2200 381.6 204.4 -1169.6 0 70 70 0 N (f 8 Plant Unloading 2200 337.2 184.9 -936.0 0 70 70 0 E
% 9 10 Loss of Load, 4.1 Loss of Load, 4.2 80 2515 1500 384.2 345.7 204.4 186.4
-1183.4
-1011.4 0
0 70 70 70 70 0 0 f ' i 11 N.O. + Earthquake 50 2200 408.6 463.3 -1134.1 0 70 70 0 N *
. 12 N.O. - Earthquake 2200 265.8 -93.5 -737.9 0 70 70 0 %
ts 8E 3e EE
' "E W
=
J
O O O TABLE A-15 SIX HIGHEST VALUES OF STRESS INTENSITY, GIRTH BUTT WELD I WITH CHANGE IN MATERIAL AND WALL THICKNESS i I l VALUES FROM REFERENCE PIPSYS PROGRAM n Eq. (12) K
-LOAD SET PAIR Eq. -(13) e n Eq. (12) Eq. (12) e 3 4 52549 *
- 1.000 52600 *
- 1.000 *N t 3 9 .49883 *
- 1.000 49900 *
- 1.000 l
l 3 10 49620 * -* 1.000 49600 *
- 1.000 l 3 6 48013 *
- 1.000 48000 *
- 1.000 y l >
m i r 1 3 48013 *
- 1.000 48000 *
- 1.000' Oi w
E 3 11 47728 *
- 1.000 47700 *
- 1.000 h l N H
l ! O l
=
1 05 a re OZ
$E "E
- G d-l co w
- Because S n, alculated by Equation (10), is less than W 3S, Equations (12) and (13) are satisfied.
1
O O O i TABLE A-16
SUMMARY
OF CALCULATIONS OF CUMULATIVE USAGE FACTOR, GIRTH BUTT WELD , WITH CHANGE IN MATERIAL AND WALL THICKNESS l j VALUES BASED ON VALUES FROM REFERENCE PIPSYS PROGRAM LOAD SET PAIR SK pe USAGE pe USAGE i j 2 FACTOR 2 FACTOR 3 9 40338 0.0050 40300 0.005 4 9 34400 0.0029 34400 0.003 N 1 11 29806 0.0002 29800 0.000 g I 6 11 29806 0.0020 29800 0.002 H 5 6 7 29163 0.0023 29200 0.002 . ili g 2 10 26254 0.0002 26300 0.000 [ 10 12 93170 0.0000 93200 0.000 U N Cumulative Usage Factor 0.0126 0.0124
- 85 4
! 8!2 tp c z
j ow ? { j
^
I i i I -
l ZPS-1-MARK II EAR AMENDMENT 13 OCTOBER 1980'
/~l U
TABLE A-17 MODAL FREQUENCIES (CYCLES /SEC) MODE NUMBER PIPSYS NATRAN DYNAL 1 6.07 6.085764 6.0821088 2 10.69 10.94144 10.936468 3 11.48 11.66862 11.666215 4 14.76 15.20947 15.204282 5 20.12 22.25613 22.135260 6 23.87 28.53255 28.505264 7 25.32 30.58105 30.530972 8 28.80 31.22073 31.190062 9 30.00 32.27319 32.199679 () 10 42.39 43.14653 43.135100 11 42.95 43.50436 43.497053 12 58.02 58.19336 57.991710 13 77.78 76.62025 71.996751 2 14 90.74 93.69710 92.12974 15 91.8 96.04482 95.167976 16 93.39 97.81956 97.410131 . 17 96.96 99.40727 98.209594
- 18 101.42 104.6169 101.64513 I
19 102.14 105.4910 103.80206 I I 20 103.03 107.7136 107.52304 1 A.13-17 1
EPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 [} L s' TABLE A-18 COMPARISON OF RESPONSE SPECTRA VALUES FROM RSG AND REFERENCE A-8-18 RESPONSE SPECTRUM VALUES IN (G) UNITS 0% DAMPING 2% DAMPING 5% DAMPING 10% DAMPING 20% DAMPING PERIODS NO. (sec) R$c REF(A18) RSG REP ( A18) RSG REF(A18) RSG REF(A18) RSG REF(A18) 1 0.042 0.783 0.765 0.409 0.421 0.367 0.374 0.349 0.360 0.348 0.350 2 0.046 0.990 1.05 0.430 0.427 0.388 0.384 0.371 0.371 0.348 0.358 3 0.050 1.015 0.951 0.540 0.571 0.461 0.467 0.404 0.407 0.359 .0.369 4* 0.055 1.044 1.14 0.585 0.557 0.428 0.416 0.403 0.406 0.365 0.382 5~ 0.070 0.963 0.803 0.471 0.493 0.439 0.446 0.424 0.424 0.383 0.398 6 0.080 1.333 1.32 0.708 0.712 0.582 0.579 0.482 0.488 0.390 0.407 7 0.085 1.070 1.07 0.682 0.668 0.591 0.591 0.486 c.453 0.395 0.408 8 0.100 1.755 2.07 0.815 0.805 0.567 0.567 0.480 0.480 0.409 0.419 9 0.130 1.990 1.99 1.040 1.03 0.773 0.772 0.535 0.541 0.429 J.445 10 0.150 1.977 1.92 0.847 0.837 0.578 0.579 0.497 0.504 0.435 0.454 11 0.180 1.309 1.49 0.887 0.890 0.726 0.727 0.587 0.597 0.452 0.474 12 0.200 1.609 1.58 0.916 0.914 0.650 0.644 0.531 0.542 0.444 0.463 13 0.220 2.415 2.50 0.731 0.728 0.683 0.667 0.580 0.576 0.442 0.471
) 14 0.260 1.538 1.60 1.177 1.15 0.903 0.902 0.639 0.653 0.443 0.469 v
15 0.280 1.309 1.31 0.878 0.882 0.755 0.746 0.567 0.578 0.430 0.463 16 0.320 1.820 1.82 1.067 1.07 0.699 0.519 0.402 0.703 0.527 0.432 17 0.360 1.212 1.21 0.877 0.877 0.657 0.655 0.504 0.511 0.379 0.408 18 0.380 1.717 1.72 0.978 0.972 0.673 0.678 0.493 0.503 0.390 0.403 19 0.400 1.964 1.99 0.824 0.827 0.614 0.615 0.473 0.481 0.409 0.418 20 0.440 1.591 1.58 0.969 n.966 0.728 0.731 0.553 0.558 0.463 0.478 21 0.480 1.405 1.41 0.996 0.996 0.794 0.797 0.644 0.651 0.514 0.537 22 0.500 1.179 1.17 1.018 1.02 0.830 0.836 0.691 0.699 0.533 0.559 23 0.550 1.988 1.99 1.266 1.26 0.910 0.917 0.745 0.759 0.545 0.S88 24 0.600 1.252 1.25 0.970 0.971 0.854 0.859 0.706 0.722 0.516 0.570 25 0.700 1.846 1.84 0.898 0.900 0.619 0.622 0.534 0.546 0.408 0.459 26 0.800 1.089 1.08 0.671 0.670 0.547 0.549 0.436 0.444 0.307 0.347 27 0.900 1.176 1.17 0.754 0.755 0.536 0.539 0.385 0.393 0.267 0.289 28 1.000 0.829 0.83 0.676 0.677 0.515 0.518 0.350 0.359 0.231 0.249 29 1.200 0.818 0.818 0.440 0.441 0.330 0.331 0.236 0.241 0.173 0.186 30 1.400 0.420 0.420 0.237 0.237 0.181 0.181 0.170 0.173 0.130 0.138 31 1.600 0.327 0.327 0.242 0.243 0.194 0.195 0.159 0.162 0.124 0.136 32 1.800 0.490 0.503 0.230 0.230 0.178 0.179 0.146 0.149 0.122 0.136 33 2.000 0.353 0.353 0.226 0.226 0.178 0.178 0.148 0.152 0.120 0.135 O (/ I A.13-18
-ZPS-1-MARK II DAR AMENDMENT 13
. OCTOBER 1980 TABLE A-19
< \
I INPUT FOR FIRST THREE CONCRETE SECTION ANALYSIS PROBLEMC i i PROBLEM SECTION AND 'l 2 3 MATERIAL PROPERTIES Thickness (in.) 42.0 .30.0 42.0 Width (in.) 12.0 12.0 12.0 Area of 1st steel i layer (in2) 6.25 2.25 3.12 i i Distance of 1st steel layer (in. ) 3.0 3.0 3.0 .l Area of 2nd steel layer (in2) 6.25 4.0 3.12 Distance of 2nd steel
- layer (in.) 37.0 25.0 37.0
() Concrete unit weight (1b/ft 3) 150.0 150.0 150.0 i I Concreta compressive j strength (lb/in2) 4000,0 4000,0 4000,0 Concrete coefficient of thermal expansion 5.56 x 10 -6 5.56 x 10 -6 5.56 x 10 6 (in/in/ F) j Steel (kips /in ) yigld strength 45.0 45.0 45.0 1 Steel modulus of elasticity (kips /in 2) 29000.0 29000.0 29000.0 Material properties Nonlinear Nonlinear Linear Applied axial force (kips) .-38.25 76.53 34.65 l Applied bending moment (ft-kips) 129.75 -9.49 206.25 ) Inside temperature (oF) 82.50 67.50 247.50 I () Outside te.iperature (oF) 52.50 0.0 115.50 l l l , A.13-19 i
. . . , . , , ,_-...__...._.,.,_,,m_., . _ , , _ __..__,.x,,.-_.._,,.,_..-.,__.,,,,,,,_-.m.,_..,..,.,.7-.. .,~
ZPS-1-MARK II DAR h$hhhggNJgg} 4 () TABLE A-20 RESULTS OF FIRST THREE CONCRETE SECTION ANALYSIS PROBLEMS PROBLEM RESULTS 1 2 3 Equilibrating axial force given by TEMCO (kips) -38.25 76.53 34.65 Equilibrating axial force computed by hand (kips) -38.253 76.53 34.65 Equilibrating bending moment give by TEMCO (ft-kips) 129.75 -9.49 206.25 Equilibrating bending moment computed by hand (ft-kips) 129.752 -9,493 206.25 () Thermal moment given by TEMCO (ft-kips) -54.58 -21.07 -137.75 Thermal moment computed by hand (ft-kips) -54.585 -21.071 -137.757 i 1 l 1 l l . O - A.13-20
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 TABLE A-21 {} INPUT FOR TENSILE FORCE AND BIAXIAL BENDING PROBLEM SECTION AND MATERIAL PROPERTIES PROBLEM 4 Thickness (in.) 42.0 Width (in.) 12.0 Area of each steel bar (in ) 1.25 Number of steel bars 10.0 3 Concrete unit weight (lb/f t ) 150.0 Concrete compressive strength (lb/in2) 4000,0 Steel yield strength (kips /in 2) 45.0 Steel modulus of elasticity (kips /in2) 29000.0 Material properties Nonlinear Applied axial force (kips) 21.0 Applied x bending moment (f t-kips) 125.0 Applied y bending moment (f t-kips) 125.0 1 4 A.13-21
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 ' i O TABLE A-22 RESULTS FROM TENSILE FORCE AND BIAXIAL BENDING PROBLEM FOR TEMCO VALIDATION 4 RESULTS PROBLEM 4 l Equilibrating axial force given by TEMCO (kips) 20.99) i Equilibrating axial force computed by ! hand (kips) 22.731 Equilibrating x bending moment given by TEMCO (ft-kips) 125.000 i Equilibrating x bending moment computed by hand (ft-kips) 124.630 Equilibrating y bending moment given by ; TEMCO (ft-kips) '25.000 Equilibrating y bending moment computed by hand (ft-kips) 123.753 i i t i . O 4 A.13-22 l l
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 . TABLE A-23 INPUT FOR NONTHERMAL AND THERMAL LOADS PROBLEM-SECTION AND MATERIAL PROPERTIES PROBLEM 5 Thickness (in.) 70.92 Width (in.) 12.00 Nwnber of reinforcement layers 6 Area of each reinforcement layer (in 4) 3.96 Concrete unit weight (lb/ft 3) 150.a0 Concrete compressive strength (lb/in 2) 3 4000.00 Concrete coefficient of thermal expansion (in/in/*F) -5 0.556 x 10 O Reinforcing steel yield strength (kips /in2) 30000.00 i Material properties Nonlinear Number of liners 2 Thickness of each liner (in.) 0.375 Temperature in the first liner ( F) 200.00 Temperature in the second liner ( F) 100.00 Effective eccentricity of the first liner (in.) 20.00 Effective eccentricity of the second liner (in.) 60.00 Liner yield strength (kips /in 2) 30.00 Liner modulus of elasticity i (kips /in2) 30000.00 + Liner coefficient of thermal expansion (-)s
\- (in/in/ F) 0.65 x 10
-5 A.13-23
. _ _ _ _ _ _ - . ._ _ _. -- . . - . - _ - . . , ~ - - ,
ZPS-1-MARK II DAR AMENDMENT 13
. OCTOBER 1980 ,
I( F TABLE'A-23 (Cont'd) SECTION AND MATERIAL PROPERTIES - PROBLEM 5 Applied axial force (kips) 165.40 j- Applied bending moment-(ft-kips) -35.23 i
- Applied' thermal axial force (kips) 90.00 Applied thermal bending moment (ft-kips) 900.00 Applied shear force (kips) 160.71 Applied tangential shear force (kips) 50.00 J
l Applied. transverse thermal shear 28.00 i force (kips) t 4 ] Applied tangential thermal shear 50.00 j force (kips) i t j i 4 0 i i j ' 4 l 1 4 1 2 1 O l i l i i A.13-24
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980- t g TABLE A-24 {J
- RESULTS FROM NONTHERMAL AND THERMAL LOADS PROBLEM FOR TEMCO VALIDATION i
RESULTS PROBLEM 5 Equilibrating axial force given by program (kips) 297.7 Equilibrating axial force computed by hand (kips) 299.759 Equilibrating bending moment given by program (ft-kips) 135.26 i Equilibrating bending moment computed by hand (ft-kips) 136.44 Service Factored Required transverse shear reinforcement 0.954 0.657 area by program (in 2) () Required transverse reinforcement area computed by hand (in 2) 0.954 0.657 Required tangential shear reinforce- 0.679 0.220 ment area given f program (in 2) Required tangential shear reinforce- 0.679 0.221 ment computed by hand (in 2) f A.13-25
ZPS-1-MARK II DAR AMENDMENT'13 1 4 OCTOBER 1980
'e3 . TABLE A-25 Q
INPUT FOR NONLINEARLTEMPERATURE PROBLEM FOR TEMCO VALIDATION SECTION AND MATERIAL PROPERTIES PROBLEM 6 Thickness (in.) 70.92 Width (in.) 12.00 Number of reinforcing layers 4 Area of each reinforcing layer (in 2) 3.96 Concrete compressive strength (lb/in 2) 4000.00 Reinforcing steel yield strength (Kips /in 2) 60.00 Material properties Nonlinear Concrete and steel coefficient of thermal - i expansion -# 0.556x10 i (} Steel modulus of elasticity (Kips /in 2) 29000.00 Temperature at the reference axis T __ (po) 200.00 Temperature at the opposite axis T h (F ) 80.00 Y=5* 100-1.692y+0.0318y 2 I Temperature variation, T(y), ( F) i O A.13-26
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 gs TABLE A-26 V RESULTS FROM NONLINEAR TEMPERATURE PROBLEM FOR TEMCO VALIDATION RESULTS PROBLEM 6 Strain at y = -35.46 in
- given by program 0.000230
- computed by hand 0.000233 Strain at y = 35.46 in
- given by program -0.00430
- computed by hand -0.00434 Concrete stress at y = -35.46 in 1
2
- given by program (kips /in ) .746 b
'/ 2
- computed by hand (kips /in ) .747 Steel stress at y = -27.46 in given by program-(kips /in ) 2.036 2
- computed by hand (kips /in ) 2.002 Steel stress at y = -22.48 in given by program (kips /in ) -12.945
- computed by hand (kips /in ) -12.899 1
Thermal moment 1
- given by program (ft-kips) -235.14 i - computed by hand (ft-kips) -235.05 Thermal force given by program 0.00
- computed by hand 0.41
(~)g i A.13-27
Msd NT 13 ZPS-1-MARK II DAR OCTOBER 1980 O rastE A-27 COMPARISON OF NATURAL FREQUENCIES X-DIRECTION Y-DIRECTION FREQUENCY FREQUENCY MODE OLD ANALYSIS DYSEA OLD ANALYSIS DYSEA 1 2.810 2.727 2.678 2.649 2 3.0 2.999 2.810 2.728 3 3.764 3.76? 3.762 3.758 4 3.791 3.781 3.771 3.769 5 4.588 4.576 4.578 4.531 6 5.041 5.044 5.040 5.039 7 5.776 5.791 5.486 5.431 8 6.071 6.047 6.069 6.025 O 9 8.731 8.625 8.598 8.524 10 10.950 11.270 9.614 9.824 11 12.796 12.800 12.563 12.760 1 O A.13-28
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER.1980 TABLE A-28 COMPARISON OF MAXIMUM LOADS STRUCTURAL COMPONENT DYSEA SOLUTION SAMIS SOLUTION I. RPV and internals fuel moment 17.11 (in-K) 18.64 (in-K) Top guide shear 188 (K) 204 (K) Shroud head shear 198 (K) 213 (K) Shroud head moment 16,783 (in-K) 18,150 (in-K) Shroud support shear 479.3 (K) 503.3 (K) Shroud support moment 119,020 (in-K) 126,600 (in-K) II. Building O RPv vedeste1-Sheer 802 (x) 575.e (x3
-Moment 94,200 (in-K) 91,500 (in-K)
Containment-Shear 2,902 (K) 2,906 (K)
-Moment 1,413,000 (in-K) 1,434,000 (in-K)
Shield. building-Shear 34,037 (K) 38,060 (K)
-Moment 38,494,000 (in-K) 37,270,000 (In-K)
O A.13-29
. _ . . . . _ . _ , _ - _ _ _ . . ~ . . . . _ . . _ . . . . - . _ _ _ .
AMENDMENT 13 OCTOBER 1980 O i 2 d P =I.0 P S I I k
\ -
4
/
\ /
( = 3 IN . 7- s_\s g
\
6
- E =3XIO PSI y =0.167
~3 o 3 O
I RADIUS =90 IN.
=R WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK ll DESIGN ASSESSM ENT REPORT O FIGURE A-1 !
SHALLOW SPHERICAL SHELL ANALYZED BY DYNAX - VALIDATION PRORLEM 1 l
AMENDMENT 13 OCTOBER 1980 O Q-8. 0 - Z O F-6.0 Z
$ G CCCC^
W 0 d-4.0 - O E e DYN AX g TIMOSHENKO AND WONOWSKY - KRIEGER 2.0 - 4 i i - T~ O 10 20 30 40 0 (DEG) i i WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK li DESIGN ASSESSMENT REPORT FIGURE A-2 COMPARIS3N OF AXIAL DISPLACEMENTS OF SHALLOW SPHERICAL SHELL FROM DYNAX AND REFERENCE 2 l y -t---,e r y ymv v rwrwwww vw y --9 iy --,ww~w--,,.--wy, , y ye--m---ga-----mmg &,-,,y-m--w-g---e _am,g.-wwe9-e -- r
l AMENDMENT 13 OCTOBER 1980 I O 20-> 10-
^
O f hou* 1 d 0 DYNAX l- - TIMOSHENXO AND 6 WONOWSKY-KRlEGER 1 o 2 O a 4 ~20-Z o 9 9 e W 2 ~3o_ O
-4b- 1 I I 0 to 20 so 4o 9 - D EG.
l WM. H. ZIMMER NUCLEAR POWER STATION UNIT 1 MARK ll DESIGN ASSESSMENT REPORT O FIcuae A-3 COMPARIS0N OF MERIDIONAL M0MENTS OF SHALLOW SPHERICAL SHELL FROM DYNAX AND REFERENCE 2 r
.-..w - , ,-,,., -- .,.-----..,-.e,---,-- - - - - - . . . _._., ,-. ,....,,, ,.-,-.- ,--.--- - n-. -r,-,.-.,,._,---,,-,.n.,.- , . , , . , , - , - - .
O O O i l 4 1 ONLY VERTICAL DISPLACEMENTS [ AXIS OF C.YLINDER 24 p f'ARE RESTRAINED n 62
*--0.5 lbjiN.
- l e
, , , 0.5 iuN.
, 0.1 IN. g
,g ,
i r 7 = 551N. : : 75 (N. -
- d2
-=
> t!
3 g l
- I 4
Ea ~ i <E E 4 ppp Sz d -rm e C dom 2 6 h da5 8 *g
- i @s A $"
,GR > *8
- i. BRM co m -
a *E
= ni j r- N E E 1
95M z
- 0 RR
- ~ "s --
, a= , 4 20 o i R' o -*
3 C WMz z gH 2 8C \, - i l _
i l AMENDMENT 13 0CTOBER 1980 O l
-2.5 - - --25 E=9 00 ~ PSI.
V= 0. 49 o DYNAX
- 2 '- - TlMOSHENR0
-- 20 7 (0 AND GOODIER n
Z M v - 1. 5 - '. n H -15 l Z g m z y h 0 0 0 0 0 0 w ' 0 $ w O Y- _J "c ~~' E M Q. U O i
.5- --s :
i 55 8o id5 13 0 RADIUS-INCHES 1 i WM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASSESSM ENT REPORT 4 Q FIGURE A-5 C0fiPARIS0N OF STRESSES AND DISPLACEMENTS OF THICX-WALLED CYLINDERS FROM DYNAX AND REFERENCE 3
, . _ - . - -_ - . . . - , = . . . . - . - . . - . . , _ . - . . - . , . - . . . . . . . - . . . . - - - . . - - - , . , - _ ,
AMENDMENT 13 OCTOBER 1980 O Tm SHELL THICKNESS = 1 IN. M= 1 LB.-IN./IN.
- E= 91. LB /IN 2
'V= .3 N= FOURIER HARMONIC NUMBER E
as U.g A UR V R Z:V:O
; M so >
O , 50 1P p (M U U g =y= o R l WM. H.ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT FIGURE A-6 CYLINDER UNDER HARMONIC LOADS ANALYZED BY DYNAX - VALIDATION PROBLEM 3
l N=O o gggnogen7 13
-1. - OCTOBER 1980
,s - MERIDIONhL MOMENT (Ib.IN/IN) n v X-i 0 DYNAX
- BUDIANSKY AND
,+ - UR(IN). RADK0WSKI
-2 -
0 -
~
l I i i I do 5 so ao 30 +o 50 AXtAl. DISTANC E (,1N) O N=2
-LO '
MERIDIONAL MOMENT (Ib.IN./IND
,8 - X-l
~0~ e DYNAX
- BUDI ANSKY AND
_4_
, UR(IN.) RADKOWSKI
.2 -
- c C 0- g e
+2 i i i i i o.o 5 Io to so 40 so AXfAL DISTANCE (im)
WM. H.ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT O FIGURE A-7 COMPARIS0N OF RESULTS FROM DYNAX AND REFERENCE 4 0F MERIDIONAL M0MENTS AND DEFLECTIONS OF CYLINDER (N = 0, N = 2) l
AMENDMENT 13
, g OCTOBER 1980
- 0. 8 -
Q /
- 0. 6 -
- BUDIANSKY AND e DYNAX
- 0.4 - RADK0WSKI
~ ~
MERIDIONAL MOMENT (Ib-IN./iN.) X-1 0- 0 0 C c c.
- 0. 2 . , , , i O.0 5 10 20 30 40 50 AXIAL DISTANCE (IN.)
O -i.0 o n=20 9
- 0. 8 -
- 0.6 -
ERIDIONA L MOMENT (Ib-IN.[lN.) X-1 O DYNAX
- 0.4 - - BUDIANSKY AND
- 0. 2 - UR (IN.)
0 0 . O O O O O . O. 2 . , , , , , 5 10 20 30 40 50 AXIAL DISTANCE (IN.) WM. H.ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK ll DESIGN ASS ESSM ENT REPORT
] FIGURE A-8 COMPARIS0N OF RESULTS FROM DYNAX AND REFERENCE 4 0F MERIDIONAL MOMENTS AND DEFLECTIONS OF CYLINDER (N = 5, N = 20) ;
1
AMENDMENT 13 0CTOBER 1980 L O e v v 6 a 1 p d p L L p P t ' t P g o v I i . T 4 .0005 SEC.
- TIME HISTORY LOADING L =l8 IN.
MASS DENSITY @= 0.018 P = 500 lb. 9 = 0.3 R: 3 IN. 4 = 0.3 (N. TIME ST E P= .0000 0 5 SEC. 6 E=30X10 lb. / INf WM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 M ARK 11 DESIGN ASSESSM ENT REPORT O v FIGURE A-9 SUDDENLY APPLIED RING LINE LOAD ANALYZED BY DYNAX - VALIDATION PROBLEM 4
~ O O O t 1 I 4 l 4-3- ! W H STATIC DEFLECTION l p (i-v2) ______,_ _______________ ______,_____ l 2- - REISMANN AND PADLOG E g -E I e DYNAX
% r ~
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. MARK 11 DESIGN ASSESSM ENT REPORT O rIsuRE A-i2 SPHERICAL CAP ANALYZED BY DYNAX -
VALIDATION PROBLEM 5
i AMENDMENT 13 OCTOBER 1980
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WM. H.ZIMMER NUCLEAR POWER STATION. UNIT 1 i MARK 11 DESIGN ASSESSMENT REPORT FIGURE A-13 l COMPARIS0N OF RESULTS FROM DYNAX AND REFERENCE 6 0F AXIAL DISPLACEMENT OF l SPHERICAL CAP UNDER DYNAMIC LOAD I l _ _ _ .. . _ . . _ . . _ . ~ . ,._ __. _ ____-.._ ._..-_ _____ ._.-.. __.._ _ .,_. ,_ _...-_-._.
t-AMENDMENT 13 OCTOBER 1980 Nz I . O gg Ox 0 1 I o ' I t-l z w - l 3 - 10.*8 l 2 l 0 l F X
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l l AMENDMENT 13 i OCTOBER 1980 60ft. 0 V Softy l. -}-THICK N ESS VARIES FRO M Giu. TO 2 + in. A n (%% F E i l 1, 82..L 204.I e l= 1 l E = 3.+65x to' psi 9 = .17 0 t = 22.SI x 10 lb. saya.* THICKHESS = G in.
-270ft.Y l
--THic K NESS
-2 95 f t. y '
<<<< VARIES FR0M G in. To 95 ..s i.u $
WM. H. ZlMMER NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT O FIGURE A-15 HYPERBOLIC COOLING TOWER MlALYZED BY DYNAX - VALIDATION PROBLEM 6 l l
AMENDMENT 13 0CTOBER 1980 m 5 3 p.g:.}G,A~-R
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/] 'v FIGURE A-16 SPECTRUM OF DESIGN EARTHQUAKE USED FOR DYNAX - VALIDATION PROBLEM 6
AMENDMENT 13 50 - OCTOBER 1980 O O-i 0
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, , h , a j 2 4 6 8 10 WM. H.ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK ll DESIGN ASSESSMENT REPORT FIGURE A-17 COMPARIS0N OF COOLING TOWER MERIDIONAL FORCES 0 STAINED BY DYNAX AND REFERENCE 7
1 AMENDMENT 13 OC10BER 1980 O 100
\
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a f Finite Element Model WM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASS ESSM ENT REPORT FIGURE A-18 TYING OF SOLID AND SHELL ELEMENTS ANALYZED BY DYNAX - VALIDATION PROBLEM 7
AMENDMENT 13 Moment - Ft-k/Ft. OCTOBER 1980 10 20 fo go 80 100 O s. i i i i i i i i i 48 - O _ G 4G _
. O % 44 _ ' o e
B - o W 42 OB - t <
@ DYNAX cn 1 Analytical Solution 38 -
c' 34 _ i G , WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 M ARK 11 DESIGN ASSESSMENT REPORT O FIGURE A-19 M0 MENT DIAGRAM 0F RESULTS FROM DYNAX AND ANALYTICAL SOLUTION-
AMENDENT 13 7" OCTOBER 1980 Steel Layers o- /
~
/ E A
= 1000 =
Circular Plate 80Est- - 4.584st - 4 I l E m 3xio xs1 , l ' l E = 4.0% xl5 kus I i ~ O 2.!,e_3 132t e_3 Concrete Steel
-24 @
T _- @
@ d 10 Layers of
, Concrete 6 2' fo b Each g -
WM. H.ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK Il DESIGN ASSESSMENT REPORT Q 9 g FIGURE A-20 t24' @ CIRCULAR PLATE ANALYZED BY DYNAX - !'
. VALIDATION PROBLEM 8 Layer Idealization l
AMENDMENT 13 OCTOBER 1980 O Z A R = 10 ft. L= 5 ft. t = 0.25 ft. 9R O s i 1 l WM. H.ZIMMER NUCLEAR POWER STATION, UNIT 1 ' MARK 11 DESIGN ASSESSMENT REPORT Q FIGURE A-21 CYLINDER UNDER CONSTANT PRESSURE ANALYZED BY DYNAX AND SOR III (DYNAX)
AMENDME'NT 13 OCTOBER 1980 O l 1
-50. -50.
v v E = 100 1 p =1 v =0 20' l i a O >a 100' gg \
< N l ' l(-- ,
1 l WM. H.ZIMMER NUCLEAR POWER STATION, UNIT I MARK 11 DESIGN ASSESSMENT REPORT FIGURE A-22 CYLINDER UNDER DYNAMIC AXIAL PRESSURE FOR NON-REFLECTING BOUDARIES ANALYZED BY DYNAX - VALIDATION PROBLEM 10 l
AMENDMENT 13 OCTOBER 1980 0 44 4 Material Damping g Element Coefficiel. 8 1 0.04
" 3 2 0.05
, h 3 0.03 o 2 O
1 Trd& Q g 2' t = 0 . l' 1 WM. H.ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT Q FIGURE A-23 FINITE ELEMENT MODEL AND MATERIAL DAMPING COEFFICIENTS FOR CYLINDER l ANALYZED BY DYNAX - VALIDATION PROBLEM 11 l l
AMENDMENT 13 DCTOBER 1980 O , a1 1 s' twicx coacr e-Vertical wall o n - el m
@ c? = 2.0' thick concrete w base slab y ? ' Soil 5 (O r 32.s ' I (a) Model O
e 15
.P 1 F x/fr.-{
ir T,,,e r x/ft.' (b) Meridional Load Distribution ! WM. H.ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK ll DESIGN ASSESSMENT REPORT FIGURE A-24 MODELING AND LOAD DISTRIBUTION
AMENDMENT 13 OCTOBER 1980 4.j 4 g J ( } ;y' , 6 s.
., - .. 6 4 .- , ..-.
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WM. H. ZIMMEFI NUCLEAR POWER STATION, UNIT 1 M ARK ll DESIGN ASSESSM ENT REPORT l O Flouae A-2s TIME HISTORY OF LOAD (FAST)
AMENDMENT 13 { OCTOBER 1980 g 40' 40' Q r
= -
a d y lumped mass 20' OI = 15 kips /ft o 3 ,, 4 ,,
.5 i
20' h h o 6 7 8 o : : zo' @ @ v ,,9 ,,IO , 11 7/r/ ////////// ////i/ /// / /////////// /' Nd1 Points 4 and 5 linked by rigid element O Element Unit Wt. Poisson s She odulus Moten.o l pcf . No. Ratio ksf I Steel 15 0 0.30 6000 2 Sond iI 5 O.45 3000 3 Sand 11 5 O.45 2700 4 Clay 12 0 0.45 4OOO 5 Cloy 120 0.45 3700 WM. H.ZIMMER NUCLEAR POWER STATION. UNIT I MARK la OESIGN ASSESSMENT REPORT FIGURE A-26 LUSil VALIDATION PROBLEM .
AMENDMENT 13 OCTOBER 1980 , 1 O o' ' i i ' i >>> > ' ' ' 'i ' ' ' ' ' ' NocwL. PotNT G o,5 _ VERTICAL 6PECTRUM - vafRTicAL EXITATION
-*- EERC R8liPCRT NO.74 4 0,4 _
os.4L.LU6H - O.b - - 0 i 1 Z O . Q. - - o 1 i 2 \ _0 01 - O.0 til j NOOAL POINT Q. D O*D - HORIZONTAL 6PECTROM__ HORIZONTAL EXITATION
-+-EERC REPoftT MO,74-4 , o., _ o e.a. uum _
O s l O.S - 4 0.2 - Y - O .1 - o e 2 i i i i i ei 3 4 56761 i 1 i e i eiii 5 4 5&781 i 2 i i i i i ii 5 4 56781 0.01 0.1 1.0 10.0 PERICO IN SECONO6 WM. H. ZIMMER NUCLEAR POWER STATION. UNIT 1 MARF tl DESIGN ASSESSM ENT REPORT Q FIGURE A-27 COMPARISON OF RESPONSE SPECTRUM (EERC REPORT N0. 74-4 AND S&L LUSH)
AMENDMENT 13 0CTOBER 1980 O 1, 12'-0" d 7F 35K I = 9000 in 4 ., m c 3
~
o = , R e R 6 8n 9 N e n u h g H B E j - I = 4000 in4 h A 8 8o 6 m o J r m n n O p 8 n a v m A F gg 20'-0" . I > E = 30000 kai l l WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK 18 DESIGN ASSESSM ENT REPORT Q FIG'JRE A-28 EXAMPLE FRAME FOR PIPSYS STATIC ANALYSIS
1
*3 : AMENDMENT 13 OCTOBER 1980 7_. . l}'}
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AMENDMENT 13 OCTOBER 1980 1 25 ll 15 s L"?23,
/ 1 z/N
'Nf8J \ RESTRAINT g
/l\
/ '
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/N LUMPED MASS WM. H.ZIMMER NUCLEAR POWER STATION. UNIT 1 MARK 11 DESIGN ASSESSMENT REPORT j
O eIGuaE 4 30 STRUCTURAL MODEL 0F PIPING SYSTEM (PIPSYS)
I AMENDMENT 13 OCTOBER 1980 24 - O !;0iurs19,21,32 22 - 20 -
.H 18 -
1 *2
- C e
e 16 - JOINTS 15,17 12 10 JOINT 25 8 6 4 -
.2 -
l i I I I I I I I I
.04 .08 .12 .16 .20 .24 .28 .32 .36 i TIME (sec)
WM. H. ZIMMER NUCLEAR POWER STATION, UNIT 1 MARK 18 DESIGN ASSESSMENT REPORT FIGURE A-31 LOADTIMEHISTORY(PIPSYS) l l
AMENDMENT 13
--- OCTOBER 1980 NASTRAN
. 28 -
) -- DYNAL i O a grests
.26 -
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. 04 .08 .12 .16 .20 .24 .28 .32 .36 TIME (Sec)
WM. H. ZBMMER NUCLEAR POWER STATION, UNIT 1 MARK 18 DESIGN ASSESSMENT REPORT FIGURE A-32 DISPLACEMENT VS. TIME JOINT 8 Z DIRECTION l l \ . _ - _ - . - . . . . . . - , - - - - . - . - - - - . - - - . - - . . - - . - - _ . - - - -
AMENDMENT 13 m OCTOBER 1980 1 m.
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? m,uuc S 0 NI NO I188313338 l N01169313339 WM. H. ZlMMER NUCLEAR POWER STATION. UNIT I MARK 11 DESIGN ASSESSMENT REPORT FIGURE A-33 O. COMPARIS0N OF ACCELERATION TIME HISTORY PLOT OF THE EL CENTR 0 NORTH-SOUTH EARTHQUAKE RECORD FROM RSG AND AS PUBLISHED IN REFERENCE 7 1
AMENDMENT 13 OCTOBER 1980
@ A G.O. X DATA FROM REFERENCE 8 FOR VARIOUS DAMPINGS c')
v RSG FOR VARI 0liS DAMPINGS. FREQUENCY IN CPS 50.0 20.0 10.0 5.0 t.0 t.0 0.5 20 0 ,l ', / is,, e 'l/// ', l , l,'i/ iiii ls',/ isi i 'l/// 'i l 6 li,/is 6. 20.0 15.0~ [15.0 10 0,, _10.0
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ZPS-1-MARK I'I DAR AMENDMENT 13 OCTOBER 1980 APPENDIX B - QUESTIONS AND RESPONSES This appendix consists of NRC questions on the DAR and the applicant's responses thereto. The questions and responses have been reproduced here exactly as they appeared in the DAR through Amendment 12 and in the Closure Report, Appendix I to-the FSAR. The responses have not been altered because the areas of concern have been addressed in the body of the
; DAR in Amendment 13 and any chapter /section references in the old DAR may not be applicable to the new format intro-duced in Amendment 13.
In the future, any revisions to existing responses or the inclusion of new questions and responses will be distinguished , by having an amendment number greater than 13 in the upper right hand corner of the page. O i O B-0 I
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 APPENDIX B - NRC OUESTIONS AND RESPONSES (} INDEX TO NRC OUESTIONS SERIES 020 OUESTIONS AMENDMENT OUESTIONS KEYWORD INDEX TO QUESTIONS NUMBER PAGE 020.1 Condensation Oscillations 13 B-1 020.2 Single Downcomer Horizontal Condensation Loads 13 B-1 020.3 Multiple Downcomer Horizontal Condensation Loads 13 B-1 020.4- Downcomer Horizontal Condensation Load Time History 13 B-1 020.5 Pool Swell Surface Velocity 13 B-2 020.6 4T Test Parameter Matrix 13 B-2 020.7 4T Test 13 B-3
- 1. Scaling 13 B-3
- 2. Data Application 13 B-3
- 3. Error Analysis 13 B-3
- 4. Geometry 13 B-3
- 5. Multiple Vent Data 13 B-3 020.8 Pool Swell Waves 13 B-3
([) 020.9 020.10 Load Mitigation in Pool Plant Specific Application of 13 B-4 Pool Swell Model 13 B-4 020.11 Discrepancy in Figure 4.4-28 Identification 13 B-5 020.12 Fluid Velocity for Drag Loads 13 B-5 020.13 Bubble Pressure 13 B-5
- 1. Differential Ar.ross Equipment 13 B-5
- 2. Calculated Versus Test 13 B-5 020.14 Fallback Loads 13 B-5 020.15 Impact Load Design Margins 13 B-6 020.16 Estimated Pool Swell Parameter Correlations 13 B-6 020.17 Table of Loads 13 B-6
. 020.18 Load Combination Time-Histories 13 B-7
- 1. Vent Clearing 13 B-7
- 2. Pool Swell 13 B-7
- 3. Condensation Loads 13 B-7 -
- 4. Drag and Fallback 13 B-7 020.19 Quencher Data Multiple Regression Analysis 13 B-7 i 020.20 Quencher Data Base 13 B-8 020.21 Quencher Design Loads 13 B-8 020.22 Large Break with SRV Actuation 13 B-8 i
020.23 Asymme. ; Loads 13 B-8
/~' 020.24 4T Tasc Data 13 B-9 \ - 020.25 Air Tests 13 B-9 t B-i l
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 INDEX TO NRC OUESTIONS (Cont'd) AMENDMENT OUESTIONS KEYWORK INDEX TO QUESTIONS NUMBER PAGE 020.26 Primary and Secondary Loads 13 B-10 020.28 Pool Dynamic Effects Due to Vent Submergence and Wetwell Backpressure 13 B-10 020.29 Thrust Loads on Vent System 13 B-10 020.30 Lateral Drag Load on Vents 13 B-12 020.31 3D Test Program 13 B-12 020.32 Jet Impingement Load on Base Mat 13 B-13 020.33 Diaphragm Pool Swell Upward Load 13 B-13 020.34 High Steam Flow Downcomer and Pool Boundary Loads 13 B-13 020.35 Pool Swell Dynamic Analytic Model 13 B 14 020.36 Pool Swell Velocity Breakthrough Model 13 B-15 020.37 Downcomer Lateral Load 13 B-15 020.38 Multiple Vent Chugging 13 B-16 020.39 Pool Boundary Loads 13 B-16 020.40 Chugging Wall Load Distribution 13 B-17 020.41 PSTF Data Applicability to Mark II , Geometry 13 B-17 s 020.42 Water Impact Loading of Structures 13 B-18 020.43 Use of PSTF Impact Data 13 B-18 020.44 Pool Swell Loads 13 B-18 020.45 4T Test Report 13 B-19 020.46 4T Test Data 13 B-19 020.47 Rams Head Orientation 13 B-19 020.48 SRV Load Calculation Code 13 B-20 020.49 Multiple SRV Actuation Transient Analysis 13 B-20 020.50 SRV Bubble Dynamic Model 13 B-21 020.51 Applicability of Bubble Position As Empirical Correlation 13 B-21 020.52 Rams Head Load Computation 13 B-21 020.53 Method of Computing Bubble Frequencies Due to Multiple Valve Actuation 13 B-22 020.54 Quencher Loads 13 B-22 020.55 Calculating SRV Loads on , Submerged Structures 13 B-23 020.56 Primary vs. Secondary Loads 13 B-23 020.57 Pressure Suppression Tests 13 B-24 020.58 Pool Swell Calculations 13 B-24 020.59 Downcomer Bracing Effects 13 B-34 020.60 Underpressure Prior to Chug 13 B-35 020.61 Pool Swell Phenomena and Loads Inside Reactor Pedestal 13 B-36 ({} B-ii ) l l
)
l 1 2PS-1-MARK II DAR AMENDMENT 13' OCTOBER 1980 INDEX TO NRC QUESTIONS (Cont'd) 3 (v AMENDMENT QUESTIONS KEYWORK INDEX TO QUESTIONS NUMBER _ PAGE 020.62 Suppression Pool Temperature Monitoring 13 B-36 020.64 Lateral Loads on Down;omers During Vent Clearing 13 B-39 02r 65 Vent Exit Flanges 13 B-40 013.66 Static Equivalent Loads for Down-comers with Diameter < 24 Inches 13 B-40 020.67 Hultiple'Downcomer Loading 13 B-41 020.68 Maximum Pool Swell Elevation 13 S-42 020.69 Upward AP 13 B-42 020.70 Drag Loads on Submerged Structures 13 B-44 020.71 Calculated Drywell Pressure 13 B-44 020.72 Impact Pressures 13 B-52 020.73 Pool Swell Velocity 13 B-53 020.74 Chugging Loads 13 B-54 020.75 Main Vent Condensation Loads on Submerged Structures 13 B-54 SERIES 040 00ESTIONS - 041.54 Suppression Pool Hydrodynamic () 041.55 Loads Loads Following Pool Swell or 13 B-59 Seismic Slosh 13 B-59 041.56 Multiple Downcomer Lateral Loading Combinations 13 B-60 041.57 Support Column Loads and Downcomer Vsat Loads 13 B-61 041.58 SRV Load Calculation 13 B-61 041.59 Loads Due to Subsequent SRV Actuation 13 B-64 fdRIES 130 QUESTIONS 130.1 ERV Loads 13 B-65 130.2 Load Combination Time History 13 B-65 130.3 Load Combination Probabilities 13 B-66 130.4 Soil Modeling 13 B-66 130.5 Liner and Anchoring 13 B-66 130.6 Asymmetric SRV Loads 13 B-67 130.7 Combining SRV and Pool Loads 13 B-67 130.14 Breakdown of Contrib.: ting Member Forces 13 B-67 130.15 LOCA Cyclic Condensation Load 13 B-68 130.16 Time Phasing of Loads 13 B-69 130.17 Force and Moment Diagrams 13 B-69 {) 130.18 Maximum Force Value Plots 13 B-70 B-iii
2PS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 INDEX TO NRC QUESTIONS (Cont'd) AMENDMENT OUESTIONS KEYWORK INDEX TO QUESTIONS NUMBER PAGE 130.19 Load Assessment, Containment Liner and Interior Wetwell Structure 13 B-70 130.20 Chugging Loads 13 B-75 130.21 Thermal Effects 13 B-75 130.22 LOCA Input Into Load Combination 13 B-94 130.23 Chugging Lords 13 B-94 130.24 Thermal Effects on Load Combination 13 B-94 130.25 ABS for Structural Steel 13 B-96 130.26 FSI Effects on Floor Response Spectra 13 B-98 ; 130.27 FSI Analysis 13 B-107 O I 1 l f i O B-iv
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 ('T NRC QUESTIONS WITH REPONSES Q ,/ J QUESTION 020.1
" Clarify the statement that no load should be applied to the containment walls due to condensation osciallations. Figure 5-7 indicates that condensation osciallations should be applied to the submerged wetwell and Section 6.1.9 of NEDO-11314-08 identifies the condensation oscillation loading that should be applied to the pool boundary as determined from the PSTF tests."
RESPONSE
The response to this question is included in Appencix A to the DFRR. QUES'!ON 020.2
" Discuss the manner by which the mean and maximum horizontal condensation leads should be applied to a single downcomer. "
RESPONSE
Downcomer loads are discussed in DEFR Section 4.3 and Table 5-1. gm The static equivalent load applied to a single downcomer is 8800 (_) pounds, based on the maximum observed test load. No mean value is considered. QUESTION 020. 3
" Discuss the criteria that are used in the multiple loading of the downcomers due to horizontal condensation loads.
Specifically, identify what fraction of the downcomers experience a load acting in the same direction and identify and justify the load to be used."
RESPONSE
Multiple loading of the downcomers is discussed in subsection 4.3.2.4 of the report. In this section, a methodology is described. based on the application of a probabilistic analysis, for determining what fraction of the downcomers experience the application of in phase load. QUESTION 0 20.4 "It is not obvious how the downcomer horizontal condensation loads, loading time interval, and load period were obtained s, from test data presented in NEDE-21078P. Provide specific references and a discussion of how the foregoing parameters {'} were obtained, including any statistical analysis techniques that were used." B-1
s l ZPS-1-MARK II DAR , AMhNDMENT 13 OCTOBER 1980
RESPONSE
[) s_ The response to this question is included in Appendix A to the DFFR. QUESTION 020.5 "I1e pool swell model discusced in Section 4.4.1 of the DFFR ha3 been used to calculate the water surface velocity
- associated with the impact pressures presented in Figure 4.4-24 through 4.4-26. Discuss the adequacy of the model to conservatively predict the velocity of the pool surface considering the assumptions that the entire mass of water associated with the vent submergence must be accelerated by the bubble pressure."
RESPONSE
The pool swell model described in Subsection 4.4.1 of this report will give water surface velocities that are conservative, provided no credit is taken for any loss of energy from the drywell air. Several assumptions used in the model. lend to the credibility of this assertion. These are discussed below.
- 1. Following vent clearing, only air flows into the suppression pool rather than a mixture of air and steam.
p)x s- This maximizes the mass flowrato of the noncondensibles and, hence, the resultant pool swell will be maximum.
- 2. The mass flowrate of air through the vent is calculated based on adiabatic flow with friction. TPis will tend to maximize the air bubble pressure and, henc., the pool swell velocity.
- 3. The air in the drywell is isentropically compressed.
This maximizes the peak drywell pressure.
- 4. Frictional losses between the water and the confining walls are negligible. This will also lead to higher pool surface velocity.
The net ef fect of these assumptions is to maximize the water surface velocity calculated by the pool swell model. QUESTION 020.6
" Provide the matrix of the 4T tests that will provide data relative to the Mark II design. Identify the key pool swell parameters that were obtained from the test data. Identify tne range of independent variables that were covered by the 1
_ test program." l l B-2 l
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980
RESPONSE
(~)) This information is provided in NEDO-21297, " Mark II Supporting Program Report," Nay 1976, submitted to the NRC on June 9, 1976, and NEDO/NEDE-13442P-01, " Mark II Pressure Suppression Test Program," May 1976, submitted to the NRC on May 28, 1976. QUESTION 020.7
" Provide the following additional information related to the
'4T' tests:
- 1. Provide a detailed scaling analysis for those parameters that will not be full scale in the tests.
Specify the portions of the pool dynamics transient in which the scaling analysis is applicable.
- 2. Discuss the manner by which test data will be applied to specific plant design. Include in this discussion the influence of vent flow rate (or transient drywell pressure) , vent submergence, and wetwell airspace volume.
- 3. Provide a comprehensive error analysis for the key independent variables measured in the test program.
Discuss how these errors were factored into a (J s
).
determination of conservative dependent variables.
- 4. Discuss the potential influence of the '4T' geometry and configuration on the test results. Specifically, consider the effects of the tank walls on the measurement of the lateral loads and pool surface velocity, and the effect on the vent exit (i . e. ,
without bolt flange) on the lateral loads and bubble information.
- 5. Identify any multiple vent tests data that can or will be used to substantiate the unit cell approach used in the '4T' test facility."
RESPONSE
The response to this question is included in Appendix A to the DFFR. QUESTION 020.8
" Video tapes of tests performed on a vent system simil.ar to the Marx II design exhibited a significant amount of wave formation in the pool following the initial pool swell transient. Discuss the relevance of this phenomenon to the
( ') Mark II design, including the origin and anticipated magnitude of loads due to waves." B-3
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER.1980
RESPONSE
Additional information on this question is being developed based on further evaluation of tne 4T data and a survey of other available information. Because the pressure variations observed are less than 1.0 psi (equivalent to waves less than 2 feet through-to-crest) , the loading resulting from such a potential effect is well below the value used in containment design. QUESTION 020.9
" Discuss the design features of the Mark II containment, or potential design modifications, which would be used to mitigate pool dynamic loads."
RESPONSE
Based on the pool swell velocities obtained by applying the DFFR model to specific Nark II plants, the loads caused by pool swell on the Mark TT weLuell structure are within acceptable limits. For this reason, no need exists for design modifications of the wetwell to mitigate pool swell loads. However, tests are in progress or planned as part of the BWR Pressure Suppression containment Program that may yield beneficial design modifications that mitigate the loads acsociated with pool dynamics. While no specific load-mitigating design modifications () are currently planned for the Mark II containments, evaluation of the tests for possible application will continue. The consideration of such applications will be reported as appropriate. QUESTION 020.10 i
"In Subsection 4.4.4 of this report, all of the Mark II plants have been grouped according to key geometric
, similarities. Discuss the manner by which the solutions of the pool swell model for each of the test cases are to be applied to the other plants in each class. If the solutions for a test case are to be applied equally to all other plants in a particular class, justify the approach with respect to dif ferences in drywell response and geometry between the test case and other plants in the same class."
RESPONSE
The purpose of the grouping of all Mark II plants was to select one typical plant from each of the three groups and then analyze these plants for their pool swell response. The solution 1 obtained for each typical plant was not intended to be applied to other plants in the same class. Any specific plant whose drywell pressure response and geometrical parameters are different from
]A- that of the typical plant shall te analyzed for pool swell by using the analytical model given in Subsection 4.4.1.
B-4 l
l ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 l (} QUESTION 020.11 j
" Subsection 4.4.4 of the report identifies Figure 4-28 as being the transient suppression pool air space pressure, whereas this figure is apparently the transient bubble pressure. Clarify this discrepancy."
RESPONSE
subsection 4.4.4 of the report has been modified such that refer +nce to Figure 4-28 is no longer appropriate; however, Figure 4-28 has been changed to reflect the information shown. QUESTION 020.12
" Discuss the manner by which fluid velocity is determined for the compution of drag loads on submerged structures and piping."
RESPONSE
The response to this question is included in Appendix A to the DFFR. QUESTION 020.13 () " Subsection 4.4.5.3 of the report indicates that the bubble pressure should be applied as a uniform increase in hydrostatic pressure.
- 1. Justify this approach with respect to potential differential pressures that could be generated across equipment or structures due to bubble propagation
. through the pool, specifically consider the reactor pedestal and the drywell deck column supports. i.
- 2. Justify the use of the calculated transient bubble pressure in terms of any relevant test data available from the 4T tests."
RESPONSE
The responses to Questions 020.12 and 020.13 require the application of information currently being evaluated: additional methods for analysis are being studied and will be v.'.ded in the last quarter of 1976. QUESTION 020.14 "Section 4.4.5.4 of the report indicates the fallback loads are determined assuming the acceleration under gravity of a (} two phase fluid. Discuss the' manner by which the density of the two-phase mixture is determined. In addition, since the B-5
-- -- ., - - , , , ,. . - - . - - - . . . .l
1 ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 ' () majority of Mark II plants have an initial wetwell air space height below three times the vent submergence, justify the assumptions of acceleration under gravity with respect to a momentum exchanges due to froth impingement on the diaphram (i. e. , rebound velocity) ."
RESPONSE
Fallback loads are discussed in the revised portions of Section 4.4. A density of 1.0 (liquid phase) is used for fallback loads (subsection 4.4.5.4) . Pool swell yields no froth (Subsection 4.4.6.3) and impact on the diaphragm floor does not occur (Subsecti 1 4. 4. 6. 5) . QUESTION 020.15 "The report indicates that a 50% design margin may be applied to the impact loads determined for a structure. Discuss the criteria to be used in determining wnether a design margin shonld be applied to a particular load."
RESPONSE
Subsection 4.4.6.1 has been revised to require the application of the 50% design margin to all impact loads determined for e j ()
%J structure.
4 QUESTION 020.16
" Discuss the manner by which the material in Appendix 4.4 of the report is to be used. In addition, describe how the data points used to generate Figures A4-1 through A4-3 were obtained."
RESPONSE
The following statement has been inserted into Appendix 4.4 of the DFFR. "The foregoing correlations are not intended tc be used in design computations. Rather, the intent of this material is to allow one to be able to make an estimate of the effects of V and H n the pool swell phenomenon. These estimates are h5Apful forB qualitative assessments of the pool swell phenomenon. The data points used to generate Figures A4-1 and A4-2 were obtained by using the pool swell model at four different submergence depths with all other plant parameters held constant." QUESTION 020.17
" Provide a table which summarizes each of the loads depicted in Figures 5-1 through 5-16. For each load, specify the O experimental data and/or analysis which form the basis for B-6
~
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 ( the load. References to the test data should indicate the specific test runs."
RESPONSE
New Tables 2-1 and 5-1 of the report provide the requestel information. QUESTION 020.18
" Provide the following clarifications regarding the temporal relationships depicted on the load combination histories:
- 1. How was the 0.7 sec vent-clearing time determined?
- 2. The pool swell event is depicted to occur between 0.7 sec and 0.9 sec. The calculations in Subsection 4.4 indicate that the pool swell event takes approximately 0.6 sec. Clarify this inconsistency.
- 3. How waa it determined that condensation loads would begin at 4 seconds following a postulated LOCA?
- 4. Discuss the manner by which the loading time is determined for drag and fallback following impact or froth impingement."
(])
RESPONSE
- 1. The 0.7 second vent-clearing time was determined with tne model described in NEDM-10320, " General Electric Pressure Suppression Containment Analytical Model," March 1971, using the typical plant parameters contained in Table 4-1 of the report.
- 2. Figures 5-1, 5-2, 5-6, 5-7, 5-11, and 5-15 have been revised to be consistent with the breakthrough time of 0.6 second, which is a typical value (Figures 4-20, 4-21, and 4-22) .
Pool swell begins at 0.7 second, the time of vent clearing, and ends 0.6 second later at 1.3 seconds, the time of breakthrough.
- 3. Condensation oscillations are not of significance to Mark II containments as discussed in Section 4.2.
- 4. The loading time for drag and fallback loads is based on free fall height and velocity, as described in Subsection 4.4.5.4.
QUESTION 020.19
" Provide a multiple regression analysis for the quencher
(~)'
- x. relief valve design using the entire data base available."
B-7
i ZPS-1-MARE II DAR AMENDMENT 13 OCTOBER 1980
)
j
RESPONSE
See Reponse to Question 020.21. QUESTION 020.20
" Provide the data base being used for the quencher design evaluation. The data should be in tabular form, listing all sensitive test parameters."
RESPONSE
See Response to Question 020.21. QUESTION 020.21
" Provide the design quencher loads to be used and their bases."
RESPONSE
~
Section 3.3 of the report has been revised entire 1y and includes the responses of Questions 020.19, 020.20, and 020.21. This information is based on GESSAR Amendment 43, which was accepted by the NRC on July 19, 1976. () QUESTION 020.22 "The load combinations to be considered for the design assessment of the Mark II containment are presented in Subsection 5.2 of the report. The load ccmbinations for the large line break do not consider actuation of a single SRV concurrent with a large break. Consideration ot a single active failure will result in this load combination. Accordingly, we will require that the load combination be considered for the Mark II containment design assessment."
RESPONSE
As noted in Subsection 5.2.4 of the revised DERR, this load combination will be used in the assessment of structures. Margin tables presented in Section 4.1 of the DAR incorporate this load combination. QUESTION 020.23 "In April 1975, generic questions related to pool swell and SRV loads for Nark II typo containments were sent to utilities with Mark II containment. In this letter, we requested that information te supplied to ' describe the manner by which potential asymmetric loads were considered in (]) the containment design. Characterize the type and magnitude B-8
ZPS-1-MARK II DAR AMENDMENT lb OCTOBER 1980 (]) of possible asymmetric loads and the capabilities of the affected structures to withstand such a loading profile.. 8."
"This information was not supplied in the DFFR. Accordingly, we require that an evaluation be presented of asymmetric load
- in the Mark II containment. Potential asymmetric loads resulting from SRV actuation and from asymmetries in vent flow should be considered. In addition, provide an d
evaluation of the capability of the Nark II containment for asymmetric pool dynamic loads."
RESPONSE
This information will be provided in the last quarter of 1976. QUESTION 020.24 "The report provides an analytical evaluation of the pool dynamic loads for Mark II containment. At the April 28, 1976, Mark II meeting dealing with Mark II pool dynamic loads, the Mark II owners group stated that the 4T tests would provide experimental confirmation of the analytical methods described in the report. It is the position of the Staff that acceptance of the pool dynamic loads by the NRC Staff is contingent on the NRC review and acceptance of the rN results of the 4T test program and a comparison of the test (/ data which the analytical methods described in the report."
RESPONSE
Reports NEDE-13442P-01 (proprietary) and NEDO-13442 (nonproprietary) , " Mark II Pressure Suppression Test Program," were transmitted to the NRC on May 28 and June 7, 1976, respectively. These reports provide the 4T Phase I test data. On June 14, 1976, the " Mark II Phase I-4T Tests Applications Memorandum" was forwarded to the NRC. This memorandum briefly compares the resul+.s of the 4T tests with the assumptions and analyses presented in the DFFR. The final report on the Phase I-4T test application will be provided in November 1976, as described in NEDO-21297, " Mark II Containment Supporting Program
, Report."
QUESTION 020.25 "We have not received a detailed description of the test matrix to be conducted for evaluation of the Mark II pool dynamic loads. The description of the 4T test program we have received indicates that 4T air tests have not been covered. In the evaluation of pool dynamic loads for the Mark I and Mark III containment design, air tests were () conducted to provide data for some of the pool dynamic loads. Because of the potential for a high air fraction in the vent B-9
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 () flow during the early portion of a LOCA, we currently believe that air tests should be conducted as a part of the Mark II pool dynamic load test program."
RESPONSE
The basic Mark II Containment Test Program was described in depth in NEDO 21297, " Mark II Containment Support Program Report," sent to the NRC on June 9, 1976. More information can be found in the DFFR Appendix A response to this question. QUESTION 020.26 "The report presents a description of a number of LOCA-related hydrodynamic loads without differentiating between primary and secondary loads. Provide this differentiation between the primary and secondary LOCA-related hydrodynamic loads. We recognize that this differentiation may vary from plant to plant. We would designate as a primary load any load that has or will result-in a design modification in any Mark II containment since the pool dynamic concerns were identified in our April 1975 generic letters."
RESPONSE
(; '~ By the definition of the primary and secondary loads, all of the LOCA-related hydrodynamic loads fall under the category of the secondary loads. QUESTION 020.28 "The importance of the effect of wetwell backpressure on Mark II pool dynamic loads (i.e. , pool swell and steam loads) was discussed in the 4T test report NEDE-13442P-01 and in the June 14 1976 4T test application memorandum. The 4T test matrix including Phases I through III does not include tests that allow separation of pool dynamic effects attributable to vent submergence and wetwell backpressure. We require that additional 4T tests, with these parameters uncoupled, he performed for the purpose of developing plant specific pool swell and steam loads."
RESPONSE
This question is answered in Appendix A to the DFFR as the response to Question M020.28. QUESTION 020.29
" Thrust loads on the vent system of a Mark II containment are rs reaction forces due to vent flow caused by the LOCA pressure
(,) transient. These loads would be transmitted to the diagnragm separating the drywell and wetwell volumes through the vent E-10
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 m l, ) deflector supports. Analyses of these thrust loads have not been provided in the DFFR. We require that these thrust loads be investigated. Provide a description of the method of analyses, the magnitude, and duration of this load for each Mark II plant."
RESPONSE
The description of the method and the equation for determining the thrust load on the downcomer vent system during a LOCA is given in Appendix A, page A-39, of the DFFR. The equation for the thrust load is: AG2 F
- e -A (P D ~
T e I - "9 PB Where FT is the upward thrust load on the downcomer vent; A is the flow cross-sectional area of the vent; Ge is the mass flux at the density of the fluid at the vent exit; Pe is the pressure in the drywell; P D is the pressure at the vent exit; m is the mass of fluid contained in the downcomer vent; and g is the local acce_eration due to gravity. The numerical values of the parameters in the thrust equation as they apply to ZPS-1 for a recirculation line break are as {} follows: A = 452.39 in2 Ge = 75.24 lbm/sec-ft2 (peak mass flux) PD = 53 psia (corresponding to peak mass flux) Pe = 52.3 psia (corresponding to peak mass flux) pe = 0.88 lbm/ft3 (based on average internal pressure of 53 psia and 0.135 quality) m = 103.92 (based on 37.59-ft long downcomer and density of 0. 88 lbm/ft3) ,
- Substituting these values into the foregoing equation, the upward thrust, F, T on a single downcomer vent is calculated as FT= (452.39) (75.2412 - (45%.39) (53 - 52.3) - 103.92) (32.2)
(144) (0.88) (32.2) (32.2)
= 628 - 317 - 104 l
207 lbg.
=
O) (, At ZPS-1 there are 88 downcomer vents with a drywell floor area of 4443 ft2 Since this thrust load is eventually transmitted to B-11
ZPS-1-MARK II DAR AMENDMENT 13 l OCTOBER 1980
) the drywell floor, the total upward pressure on the floor corresponding to the thrust load of (88 x 207) lb is 0.03 psi.
f The drywell floor in the Zimmer Power Station is designed for an upward pressure differential of 9.0 psid, while the calculated maximum upward load during LOCA pool swell is about 6.3 psid. Thus the additional upward thrust load from the downcomer vents, which is 0.03 psid, is fully accommodated in the design of the drywell floor. QUESTION 020.30 "Significant differences in the pool area / vent area ratio exist from locatic 1 to location within a given Mark II plant. These differences .nay lead to cross flow and lateral drag forces on the vents during pool swell. Based on the DFFR Section 4.4.7 it would appear that this lateral drag load on the vents would be computed based on the maximum pool surface velocity and the density of water. Confirm this interpretation of the DFFR. In addition, provide the magnitude and duration of this load for each Mark II plant. Alternatively provide justification for not including this load."
RESPONSE
O This question is answered in Appendix A to the DFFR as the response to Question M020.30. QUESTION 020.31 "We require that 30 tests te performed to substantiate the pool swell loads. These loads are currently based on a one dimensional pool swell model and single vent 4T tests. The following items should be considered as a part of the 3D test program. (1) A comprehensive scaling analysis of the test facility and error analysis of the test data. (2) A determination of the sensitivity of pool swell loads to assymetries in vent flow loads and the drywell/wetwell pressure transient. (3) A determination of the effect of spatial variations of the pool area to vent area ratio within a given plant on the pool swell phenomena."
RESPONSE
This question is answered in Appendix A to the DFFR as the (~)) \_ response to Question M020.31. B-12
ZPS-1-MARK II DAR A'MENDMENT 13 OCTOBER 1980
) QUESTION 020.32 "The DFFR includes the statement on Page 4-43 that a typical jet impingement load on the basemat can be computed utilizing the velocity attenuation given in Figure 12.3 of Reference 13. Clarify this reference since Reference 13 does not contain a Figure 12.3."
RESPONSE
This question is answered in Appendix A to the DFFR as the response to Question M020.32. QUESTION 020.33 "The diaphragm pool swell upward load was based on the unheated drywell test Run 33. This test was conducted with a vent submergence of 11 feet. Figure 5-28 in Reference NEDE-13442P-01 shows that the diaphragm upward load increases with increasing vent submergence. The current peak upward design load for the diaphragm does not appear to include sufficient margin for both this effect and uncertainty in the measured load. Address this concern and provide an error analyses to substantiate the peak upward design load for the diaphragm." O RESPONSE This quer cion is answered in Appendix A to the DFFR as the responst to Question M020.33. QUESTION 020.34 "The DFFR in Section 4.2.2 states that downcomer and pool boundary loads will not be considered during periods of high steam flow since the load derived from the 4T tests are lower than corresponding low steam vent flow lateral loads. It is our position that high steam flow loads should be considered since these loads, in combination with other loads, may ba significant. It was stated in the 4T applications memorandum that no significant downcomer lateral loads were observed at high steam vent flow. However, in NEDO-21078 Figure 3-19 foreign licensee data indicate significant lateral loads at a vent flow of 20.7 lb/ft2 in tests conducted with an air mixture of 1%. Specification of a high vent flow dowacomer load should reflect this data as well as the 4T data. For structures in the pool it is our position that the i 4 psi, 4HZ load derived from PSTF tests should be used. This load should be confirmed by data from the 4T tests." O B-13
ZPS-1-MARK II DAR AMENDMENT 13 OCTOBER 1980 (~)N (_ RESPONSE This question is answered in Appendix A to the DFFR as the response to Question M0202.34. QUESTION 020.35 "With regard to the pool swell dynamic analytic model described in Section 4.4 of the DFFR, we have a number of concerns. We request modifications and/or clarification of the methodology in response to the concerns listed below:
" (1) Assumption 5 on page 4-16 of the DFFR sets the bubble air temperature equal to the (isentropic) drywell air temperature. This assumption is unrealistic from a physical standpoint, and whethe}}