ML19347F889

Seismic Review of the Oyster Creek Nuclear Power Plant as Part of the Systematic Evaluation Program

ML19347F889
Person / Time
Site:	Oyster Creek
Issue date:	04/30/1981
From:	Ma S, Robert Murray, Nelson T, Stevenson J LAWRENCE LIVERMORE NATIONAL LABORATORY
To:	Office of Nuclear Reactor Regulation
References
CON-FIN-A-0233, CON-FIN-A-0415, CON-FIN-A-233, CON-FIN-A-415 NUREG-CR-1981, UCRL-53018, NUDOCS 8105260532
Download: ML19347F889 (158)

Text

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NUREG/CR-1981 l UCRL-53018 RD,RM f

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Seismic Review of the Oyster Creek Nuclear Power Plant as Part of the Systematic Evaluation Program Manuscript Completed: November 1980 Date Published: April 1981 Prepared by R. C. Murray, T. A. Nelson ( L L N L ),

S. M. Ma (EG&G/ San Ramon Operations).

J. D. Stevenson (Structural Mechanics Associates)

Lawrence Livermore Laboratory 7000 East Avenue Livermore, CA 94550 Prepared for Office of Nuclear Peactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 NRC FIN No. A0233

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PORENORD The U.S. Nuclear Regulatory Commission (NRC) is conducting the Systematic Evaluation Program (SEP), which consistc of a plant-by-plant limited-reassessment of the safety of eleven operating nuclear reactors that rece! 'ed construction permits between 1956 and 1967. Because many safety criteria have changed since these plants were initially licensed, the purpose of the SEP is to develop a current documented basis for the safety of these older facilities.

The eleven SEP plants were categorized into two groups based upon the extent to which seismic design was originally considered and the quantity of available seismic design documentation. The Oyster Creek Nuclear Power Plant, the subject of this report, was categorized under Group 1 on the assumption that enough documentation existed to perform the SEP review.

A detailed evaluation of plant structures and the hundreds of individual components within each Group 1 plant has not been performed. Rather, the evaluations rely upon limited analysis of selected sts .acturec and ' sampling of representative components from generic groups of equit-ment. The component sample was augmented by walk-through inspections of the . facilities to select additional conponents, based upon their potential seismic fragility.

This report reflects a collective effort on the part of the following persons:

e R. C. Murray and T. A. Nelson, (Lawrence Livermore National' Laboratory (LLNL), who provided project management support and compiled the report.

( e S. M. Ma, (EGGG/ San Ramon Operations), who conducted the seismic reevaluation of structures and the soil-structure interaction (SSI)-

studies.

e N. C. Tsai, (NCT Engineering), who assisted in SSI methodology and conducted confirmatory studies.

e J. D. Stevenson, (Structural Mechanics Associates, Inc.), who conducted the seismic reevaluation of mechanical and electrical equipment and of the fluid and electrical distribution systems.

This limited assessment of the Oyster Creek facility relied in large part upon the guidance, procedures, and recommendations of recognized seismic design experts. Accordingly, a Senior Seismic Review Team (SSRT) under the direction of N. M. Newmark was established. Members of the SSRT and their affiliations are:

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Nathan M. Newmark, Chairman (Deceased, January 25, 1981)

Nathan M. Newmark Consulting Engineering Services Urbana, IL William J. Hall Nathap M. Newmark Consulting Engineering Services Urbr.a. IL Donald A. Wesley (alternate)

Structural Mechanics Associates, Inc.

Newport Beach, CA John D. Stevenson Structural Mechanics Associates, Inc.

Cleveland, OH Robert C. Murray (member since October-1, 1980)

Lawrence Livermore National Laboratory Livermore, CA The SSRT was charged with the following responsibilities:

e To develop the general philosophy of review, setting forth seismic design criteria and evaluation concepts applicable to the review of older nuclear plants, and to develop an efficient, yet comprehensive review process for NRC staff use in subsequent evaluations.

e To assess the safety of selected older nuclear power plants relative to those designed under current standards, criteria, and procedures, and to recomend generally the nature and extent of retrofitting to bring these plants to acceptable levels of capability if they are not already at such levels.

The SSRT developed its general philosophy and presented it in the first SEP report, which reviews Unit 2 of the Dresden Nuclear Power Station (Ref. 1). The limited assessment of Oyster Creek reported here is the fourth in the series of SEP seismic reviews of Group 1 plants.

This report provides partial input into the SEP seismic evaluation of the Oyster Creek Nuclear Power Plant. The results of the seismic evaluation will be documented in a Safety Evaluation Report, prepared by the NRC staff, that will address the capability of the Oyster Creek systems to respond to scismic events or to mitigate the consequences of such events.

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u A Aimited peer review of this report was conducted by the SSRT to ensure its consistency with the review philosophy established during the SSRT's review of Dresden Unit 2 and to review the results of the limited reanalyses of plant structures and the component sample.

Safety for seismic excitation implies that certain elements and components of an entire system must continue to function under normal operating and test loads. The SSRT did not review all aspects of the plant's operation and the safety margins available to assure that those elements and components needed for seismic safety would not be impaired beyond the point for which they can be counted on for seismic resistance because of unusual operating conditicas, sabotage, operator error, or other causes. These l aspects will have been studied by others. However, where unacceptable risks of essential elements not being able to function properly to resist seismic f

events were noted or inferred, greater margins of safety or provision for redundancy in the design of these elements are considered by the SSRT to be

! necessary.

The authors wish to thank T. M. Cheng, technical monitor of this work and W. T. Russell, Chief, SEP Branch, at the NRC, for their continuing support.

Thanks also go to H. F. Keedy of LLNL and R. K. Johnson of EG&G/ San Ramon Operations for publications support.

l The authors especially wish to acknowledge the accomplishments of Dr. Nathan M. Newmark (Decea*ed January 25, 1981), Chairman of the SSRT, for l

I l his guidance and direction during the course of this study. His absence will be deeply missed by all who worked with him. _

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ABSTRACT A limited seismic reassessment of the Ofster Creek Nuclear Power Plant was performed by the Lawrence Livermore National Laboratory (LLNL) for the U.S. Nuclear Regulatory Commission (NRC) as part of the Systematic Evaluation Program (SEP) . The reassessment focused generally on the reactor coolant pressure boundary and on those systems and components necessary to shut down the reactor safely and to maintain it in a safe shutdown condition following a postulated earthquake characterized by a peak horizontal ground acceleration of 0.22 g. Unlike a conprehensive design analysis, the reassessment was s

limited to structures and components deemed representative of generic classes. Conclusions and reconssendations about the ability of selected structures, equipment, and piping to withstand the postulated earthquake are presented.

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CONTENTS '

Foreword . . . . . . . . . . . . . . . . . . . . . iii Abstract . . . . . . . . . . . . . . . . . . .- . . vii List of Figures . . . . . . . . . . . . . . . . . .. . xiii List of Tables . . . . . . . . . . . . . . . . . . . xvi Chapter 1: Introduction . . . . . . . . . . . . . . . . 1 1.1 Scope and Depth of Review . . . . . . . . . . . 1 1.2 Plant Description . . . . . . . . . . . . . . . 3 1.2.1 Seismic Categorization . . . . . . . . . . . 4 1.2.2 Principal Structures . . . . . . . . . . . . 6 1.3 Organization of Report . . . . . . . . . . . . . 10-Chapter 2: Summary and Conclusions . . . . . . . . . . . . 12 2.1 Structures . . . . . . . . . . . . . . . . . 12 2.1.1 Ventilation Stack . . . . . . . . . . . . . 12 2.1.2 Condensate Storage Tank - . . . . . . . . . . 13 2.1.3 Drywell and Torus . . . . . . . . . . . . . 13 2.1.4 Turbine Building and Control Room . . . . . . . 14 2.1.5 Reactor Building . . . . . . . . . . . . . 14 2.2 Mechanical and Electrical Equipment, and Fluid and Electrical Distribution Systems . . . . . . . 14 2.3 Piping . . . . . . . . . . . . .~ . . . . . . 16 2.4 Concluding Remarks . . . . . . . . . . . . . . . 16 Chapter 3: General Basis of SEP Reevaluation of Structures and Equipment . . . . . . . . . . . . 19 l 3.1 General Approach to Reevaluation . . . . . . . . . . 19 i

3.2 Seismic Input and Site Conditions . . . . . . . . . . 20 1 3.3 Structures . . . . . . . . . . . . . . . . . 20 3.3.1 Response Spectra, Damping, and Nonlinear Behavior . . 21

! 3.3.2 Analysis Models . . . . . . . . . . . . . 22 3.3.3 Normal, Seismic, and Accident Loadings . . . . . . 22 3.3.4 Forces, Stresses, and Deformations . . . . . . . . 22 3.3.5 Relative Motions . . . . . . . . . . . . . 23 3.4 Equf pment and Distribution Systems . . . . . . . . . 23 3.4.1 Seismic Qualification Procedures . . . . . . . . 23 3.4.2 Seismic Criteria . . . . . . . . . . . . . 23 ix

3.4.3 Forces, Stresses, and Deformations . . . . . . . 24 3.4.4 Functionality . . . . . . . . . . . . . . 24 3.4.5 Nonlinear Behavior . . . . . . . . . . . . 24 3.5 Miscellaneous Items . . . . . . . . . . . . . . 24 3.6 Evaluation of Adequacy . . . .. . . . . . .. . . 25

• Chapter 4: Previous Seismic Analyses . . . . . . . . . .. . 26 4.1 Introduction . . . . . . . . . . . . . . . . . 26 4.2 Design Criteria . . . . . . . . . .- . . . . . . 26 4.2.1 _ Earthquake Motion . . . . . . . . . . . . . 26 4.2.2 Stress Criteria _ . . . . . . . . . . . . . 27 4.3 Seismic Analyses . . . . . . . . . . . . . . . . 27 4.3.1 Methods of Analysis . . . . . . . . . . . . 27 4.3.2 Damping . . . . . . . . . . . . . . ~. . 30 4.3.3 Soil-Structure Interaction . . . . . . . . . . 30 4.3.4 Combination of Response for Earthquake Directional Components . . . . . . . . . . . 31 4.4 Seismic Response of Structures . . . . . . . . . . . 31 4.4.1 Reactor Building . . . . . . . . . . . .. . 31 4.4.2 Turbine Building . . . . . . . . . . . . . 33 4.4.3 Rosctor-Turbine Buildings--

Relative Displacements . . . . . . . . . . . 36 4.4.4 Ventilation Stack . . . . . . . . . . . . . 37 4.4.5 Containment, Pressure Suppression System, and Structures Inside Containment . . . . . . . 39

, 4.5 Seismic Design of Piping Systems . . . . . . . . . . 46 4.5.1 Analyses by John A. Blume and Associates . . . . . 46 4.5.2 Analyses by Brooklyn Polytechnic Institute . . . . 49 4.5.3 Analyses by MPR Associates . . . . . . . . . . _49 4.6 Seismic Design of Equipment and Components . . .- . . . . 50 Chapter 5: Reevaluation of Selected Structures . . . . . . . . 54 l

i 5.1 Introduction . . . . . 54 j . . . . . . . . . . . .

5.2 Design Earthquake Motion . . . . . . . . . . . . . 54 l

l 5.3 Combination Of Earthquake Directionci Components . . . . . 55 5.4 Structural Damping . . . . . . . . . . . . . . . 55

! 5.5 Methodology Used in the Seismic Reanalysis . . . . . . . 59 5.6 Soil-Structure Interaction (SSI) Parameters . . . . . . 60 X

f 5.6.1 Soil Material Properties Used in SSI Analysis . . . - . 60 5.6.2 Stiffness Calculation Including Effects of Embedment . . . . . . . . . . . . . 64

. 5.6.3 Soil Damping . . . . . .- . '. . . . . .. . .65 5.7 Dynamic Modeling of Selected Structures . . . . . . :. . 65 5.7.1 Turbine Building . .. . . . . . . . . . . . 68 5.7.2 Reactor Building . . . . . .- . . . . . . . 72 5.7.3 Stack .. . . . . . . . . . . . . . . . . 76 5.7.4 ChndensateStorageTank . . . . . . . . . . . 76 5.8 Summary of SSI Analysis .

. . . .- . .. . . .. . . . 76 5.8.1 Effects of Soil Shear Modulus Variations .

.,. . . 77 5.8.2 Effects of Damping Variations . . . . . . . . .- 78' 5.8.3 Effects of Embedsent Variations . . . . .- . . 79 5.9 Analysis of Results for Selected Structures . . . . . .. 81 I 5.9.1 Turbine Building and Control Room Housing . . . . . 82 5.9.2 Reactor Building . . . . . . . . . . . . . .82 5.9.3 ventilation Stack . . . . . . . . . . . . . 85 j 5.9.4 Primary Containment (Drywell) . . . . . . . . . 88 5.9.5 Suppression Chamber (Torus) . . . . . . . . . 90 5.9.6 Condensate Storage Tank . . . . . . . . . . . 91-5.10 Seismic Input Motion for Equipment and Piping . . . . . 92 Nspter 6: Seismic Evaluation of Mechanical and Electrical Equipment and Fluid and Electrical

Distribution Systems . . . . . . . . . .- . . . 95 6.1 Introduction . . . . . . . . . . . . . . . . . 95 6.

1.1 Purpose and Scope

. . . . . . . . . . . . . 95 6.1.2 Description of Components Selected for Review . . . 98 6.2 Seismic Input and Analytical Procedures . . . . . . . . 98 6.2.1 Original Seismic Criteria . . . . . . . . . . 98 6.2.2 Seismic Criteria for Reevaluation . . . . . . . 104 6.2.3 SEP Behavior Criteria . . . . . .- . . . . . 104 6.3 Evaluation of Selected Somponents for Seismic

{ Design Adequacy . . . . . . . . . . . . . . . . 105-6.3.1 Mechanical Equipment . . . . . . . . . . . . 105 6.3.2 Electrical Equipment . . . . . . . . . . . . 111 6.4 Summary and Conclusions . . . . . . . . . . . . . 114 4

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References . . . . - * * * * * * * * * * ~ * * * *

• Appendix As Study of Turbine and Reactor Building Responses Using Different Soll Damping Representations . . . . . 124 Apperdix B: Floor Response Spectra Envelopes . . . :. . . . . . 144 Appendix C: SSRT Guidelines for SEP Soil-Structure Interaction Review . .- . . . . . . . . . . . 148 s

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LIST OF FIGURES

1. aerial photograph of the Oyster Creek Nuclear Power Plant . . . 7

2. Isometric of Oyster Creek Plant showing major structures . . . 8

3. Cross section of the reactor building . . . . . . . . . 9

4. Housner response spectra for various values of critical damping . . . . . . . . . . . . . . . . 27

5. Reactor building model, mode shapes, and acceleration response . . . . . . . . . . . . . . . 32

6. Calculated peak OBE shear and moment diagrams for the reactor building . . . . . . . . . . . . . . . 34

7. Turbine building model and acceleration response . . . . . . 35

8. Calculated peak OBE shear and moment diagrams for the turbine building . . . . . . . . . . . . . . . 36

9. Lumped mass model and seismic responses of the ventilation stack . . . . . . . . . . . . . . . . 38

10. Drywell schematic, model, and shear and moment diagrams . . . 41

11. Schematic of the drywell truss system . . . . . . . . . 43

12. Lumped mass model of the reactor pressure vessel . . . . . . 45

13. Comparison of R.G. 1.60 and Housner's seismic response spectra for 5% and 10% of critical damping . . . . . . . . 56

14. Response spectrum (24 damping) corresponding to the time-history used in the analysis superposed on the corresponding smoothed spectrum from R.G. 1.60 . . . . . 57

15. Soil composition and soil properties used in the reanalysis of the Oyster Creek Plant . . . . . . . . . . 61

16. Sizes and embedments of bases of the stack, reactor building, and turbine building . . . . . . . . . 63

17. Mathematical model used in the reanalysis of the turbine building . . . . . . . . . . . . . . . . 69

18. Mathematical model used in the reanalysis of the reactor building . . . . . . . . . . . . . . . . 75

19. SSE shear and moment responses for the turbine building . . . 83

20. SSE torsion response for the turbine building . . . . . . . 84

21. SSE shear and moment responses for the reactor building . . . 86

22. Mathematical model and SSE responses of the stack in the reanalysis . . . . . . . . . . . . . . . . 87

23. Mathematical model and SSE responses for the drywell . . . . 89

24. Turbine building floor response spectrum curves at operating floor, El. 46 ' 6" . . . . . . . . . . . . . 94 xiii

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25. Reactor building floor respon=c spectrum curves, with 3% damping, at the top floor (El. 119' 3") . . . . . . 94 A-1. Simplified two dimensional model of the turbine building . . . 126 A-2. Simplified two dimensional model of the reactor building . . . 126 A-3. Comparison of detailed and simplified model spectral curves

1 turbine building top floor, El. 46 ', modal analysis results using 34 damping . . . . . . . . . . . 127 A-4. Comparison of detailed and simplified model spectral curves: reactor building top floor, El.119',

modal analysis results using 3% damping . . . . . . . . . 127 A-5. Comparison of spectral curves (31 damping) for the turbine building under different representations of soil damping.

The modal analysis model allows no base rotation . . . . . . 129 A-6. Comparison of spectral curves (34 damping) for the

, turbine building at the operating floor, El. 46 ' . . . . . . 130 A-7. Comparison of Goil damping using SSRT guidelines; turbine building, El. 46' . . . . . . . . . . . . . . 133 A-8. Comparison of spectral curves (3% damping) using dashpot and MODSAP representations of soil damping; reactor building, El 119'3" . . . . . . . . . . . . . 134 A-9. Comparison of spectral curves (31 damping) using dashpot and MODSAP representations of soil damping; reactor building, El. 95'3" . . . . . . . . . . . . . 135 A-10. Comparison of spectral curves (34 damping) using dashpot and MODSAP representations of soil damping; reactor building, El. 75 '3" . . . . . . . . . . . . . 136 A-ll. Comparison of spectral curves (34 damping) using dashpot and MODSAP representations of soil damping; reactor building, El. 51' . . . . . . . . . . . . .. . 137 A-12. Comparison of spectral curves (3% dampingi Jsing dashpot and MODSAP representations of soil damping; reactor building, El. 23 '6" . . . . . . . . . . . . . 138 4

A-13. Comparison of spectral curves (31 damping) using dashpot and MODSAP representations of soil damping; reactor building, El. O' . . . . . . . . . . . . . . 139 A-14. Comparison of spectral curves (31 damping) using dashpot and MODSAP representations of soil damping; reactor building, El. -10 ' . . . . . . . . . . . . . 140 A-15. Ratios of spectral values obtained using different soil damping representations . . . . . . . . . . . . . 141 A-16. Comparison of soil damping using SSRT guidelines; reactor building, El.119' . . . . . . . . . . . . . 143 xiv

B-1. Spectral curves (horizontal and vertical components) with selected percentages of damping used in reanalysis of equipment in the turbine building . . . . . . . . . . 145 B-2. Spectral curves (horizontal component) with selected percentages of damping used in reanalysis of equipment in the reactor building at selected elevations from 23'6" to -19'6" . . 146 B-3. Spectral curves (horizontal component) with selected percentages of damping used in reanalysis of equipment in the reactor building at selected elevations from 119'3" to 38'5" . . 147 r

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LIST OF MBLES

1. Damping values from R.G. 1.61 compared to those recomunended for the SEP evaluation . . . . . . . . . . . 21

2. Load combinations and allowable stresses for the Oyster Creek reactor building structures . . . . . . . .. . . . 28

3. Damping values used in the seismic design analyses of Oyster Creek . . . . . . . . . . . . . . . . . 30

4. Calculated relative displacements for the reactor and turbine buildings in response to an E-W OBE level earthquake . . . . . . . . . . . . . . . . . . . 37

5. Susunary of seismic analyses of Oyster Creek safety related piping, . . . . . . . . . . . . . . . . . . . 47

6. Seismic requirements for Class I equipment and Components . . . . . . . . . . . . . . . . . . . 51

7. Original, current recossmended, and reanalysis damping values as percentages of critical damp {ng . . . . . . . . 58

8. Soil parameters used in the analysis of the turbine building . . 66 67

9. Soil parameters used in the analysis of the reactor building . .

10. Analyses made using the turbine building model . . . .. . . . 70

11. Turbine building modal frequency and damping . . . . . . . 71

12. Analyses made using the reactor building model . . . . . . . 72

13. Reactor building modal frequency and damping . . .' . . . . 73

14. Ventilation stack natural frequencies and modal damping values . . . . . . . . . . . . . 77 Variation of soil spring constant due to embedment effects 80

15. . . .

16. Mechanical and electrical components selected by the SSRT for seismic evaluation, and the basis for selection . . . . . 99 Design and construction codes for Class I piping and equipment 101

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18. SEP structural behavior criteria for determining seismic design adequacy of passive mechanical and electrical equipment and of distribution systems . . . . . . . . . . 106

19. Conclusions regarding equipment review for seismic design 115 adequacy of Oyster Creek . . . . . . . . . . . . . .

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CHAPTER 1: INTRODUCTION i

This report describes work at the Lawrence Livermore National Laboratory (LLNL) to reassess the seismic design of the Oyster Creek Nuclear P0wer Plant. This limited reassessment includes a review of the original seismic design of selected structures, equipment, and components and includes seismic analyses of selected items using current modeling and analysis methods.

The LLNL work is being performed for the U.S. Nuclear Regulatory Commission (NRC) as part of .the Systematic Evaluation Program (SEP). The purpose of the SEP is to develop a current documented basis for the safety of 11 older operating nuclear reactors, including Oyster Creek. The primary objective of the SEP seismic review program is to make a seismic safety assessment of the plants based on a limited sarple of structures, systems, and components and, where necessary, to recommend backfitting in accordance with the Code of Federal Regulations (10 CFR 50.109, Ref. 2) . The important SEP review concept is to determine whether or not a given plant meets the " intent" of current licensing criteria as defined by the Standard Review Plan (Ref. 3)--not to the letter, but rather to the general level of safety that these criteria dictate. Additional background information about the SEP can be found in Refs. 4 and 5.

1.1 SCOPE AND DEP7H OF REVIEW This review of Oyster Creek is considerably different in scope and depth f rom current reviews for construction permits and operating licenses. Its focus is limited to identifying safety issues and to providing an integrated, balanced approach to backfit considerations in accordance with 10 CFR 50.109, which specifies that backfitting will be required only if it can be demonstrated that such backfitting will provide substantial, additional protection for public health and safety. Adequate demonstration requires an assessment of broad safety issues by considering the interactions of various systems in the context of overall plant safety.

Because individual criteria do not generally control broad safety issues, this review is not based on demonstrating compliance with specific criteria in the Standard Review Plan or Regulatory Guides. However, current licensing criteria do establish baselines against which to measure relative safety

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factors to support the broad integrated assessment. Therefore, we campare the seismic resistance of the Oyster Creek facility in a qualitative fashion to that dictated by the intent of today's licensing criteria in order to determine acceptable levels of safety and reliability.

References in this report to load ratios and safety factors do not refer in an absolute sense to acceptable minimums, but to design-based levels thought to be realistic in light of current kno'wledge. In general, original levels do not represent maximum levels because such unclaimed factors as low stress and a structure's ability to respond inelastically contribute to seismic resistance. In particular, resistance to seismic motions does not mean the complete absence of permanent deformation. Structures and equipment may deform into the inelastic range, and some elements and components may even be permitted to suffer damage, provided that the entire system can continue to perform its safety function and to maintain a safe shutdown e sndition.

This seismic reevaluation of Oyster Creek centers on:

e An assessment of the integrity of the reactor coolant pressure boundary; that is, major components that contain coolant for the core and piping or any component not isolatable (usually by a double valve) f rom the core.

e A general evaluation of the capability of essential structures, systems, and components to shut down the reacter safely and to maintain it in a safe shutdown condition, including removal of residual heat, during and af ter a postulated Safe Shutdown Earthquake (SSE) . The assessment of this subgroup of equipment can be used to infer the capability of such other safety related systems as F e Emergency Core Cooling System.

Not all equipment was examined as part of this reassessment. The intent was to examine mechanical and electrical equipment representative of items installed in the reactor coolant system and safe shutdown systems at the Oyster Creek f acility for structural integrity and for electrical and mechanical functional operability. Components that potentially have a high degree of seismic f ragility were selected for review in order to estimate the lower-bound seismic capacity in generic classes of equipment. The selection was made during a site visit by representatives of the NRC, the SSRT, LLNu, and its subcontractors. The methods of selecting representative equipment for this limited assessment are described in detail in Chapter 6.

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Structures housing the selected systems were analysed to demonstrate structural adequacy and to generate seismic input to equipment. The structures reviewed for the Oyster Creek plant include the reactor building, with its related internal structures, and portions of the turbine building.

For the structural evaluation, a peak horizontal ground acceleration of 0.22 g (the original SSE value) was used along with a Regulatory Guide (R.C.) 1.60 response spectrum (Ref. 6) .

The SSE is the only earthquake level considered in the review because it represents the limiting seismic loading to which the plant must respond safely. Present licensing criteria sometimes result.in-the Operating Basia Earthquake (OBE), which is usually one-half of the SSE, controlling the design of structures, systems, and components for which operation, rather than safety, is at issue. Because a plant designed-to shut down safely following an SSE will be safe for a lesser earthquake, investigation of the effects of the OBE war, deemed unnecessary.

To ensure safety in a seismic evaluation, certain elements and components l of an t,ntire system must continue to function under normal operating and test loads both during and following an earthquake. The seismic review team did not review all aspects of the plant's operation nor did they review the safety margins available to assure that vital elements and components would withstand unusual operating conditions, sabotage, operator error, or other nonseismic events.

This report addresses structures, systems, and components in the as-built condition, including modifications made to all seismic Category'I components since the issuance of the operating license. Information about structures, systems, and conponents was primarily obtained from the Oyster Creek docket (Docket 50219) maintained by the NRC in Bethesda, MD. Additional information was supplied by the utility and the architect-engineer either through correspondence or during site visits.

1.2 PLAlff DESCRIPTION Owned and operated by Jersey Central Power and Light Company, the Oyster Creek plant (Fig. 1) is located on the Atlantic Coast, about 35 miles north of Atlantic City, NJ end about 45 riles east of Philadelphia, PA. The plant's 3

Unit No.1 is a Mtrk I boiling water reactor, commonly designated as a BWR.

The plant was designed to produce 1600 m of heat and 515 W of net electrical power.

General Electric Company, the prime contractor for the plant, engaged Burns and Roe, Incorporated, for engineering assistance and construction management. Most seismic analyses were conducted by John A. Blume and Associates, Engineers.

The Atomic Energy Connission issued Provis tonal Operating License No.

DPR-16 on August 1,1969 for a power level not to exceed 1600 MN thermal. The level was raised to 1930 W on November 5, 1971. Application for a full-term operating license is under consideration.

1.2.1 Seismic Categorization Us'ng Appendix A of the Facility Description and Safety Analysis Report (Ref. 7) as a guide, the plant equipment and structures were categorized into one of two seismic classes as follows.

Class 1 structures and equipment are those whose failure could cause -

significant release of radioactivity or which are vital to a proper shutdown of the plant and the removal of decay heat. These includes

Structures Drywell, Vents, Torus, and penetrations Reactor Building Control Room (and supporting part of turbine building)

Spent Fuel Pool Ventilation Stack Radwaste Building Equipment f Nuclear Steam Supply Systems (NSSS)

Reactor Vessel Reactor Vessel supports Control Rods and drive system- (including equipment necessary for scram operation)

Control Rod Drive Thimble supports Fuel Elements t

Core Shroud Core Supports 4

Steam Separator Steam Dryer Recirculating Piping System including valves and pumps All piping connections from the Reactor Vessel up to and including the first isolation valve e.s > nal to the drywell.

Isolation Valves Reactor Emergency Systems Isolation Condenser System Liquid Poison System' Core Spray System Reactor Building Closed Loop Cooling Containment Spray System Service Water System Standby Gas Treatment System Fuel Storage Facilities including spent fuel and new fuel storage equipment i Standby Electrical Power Systems Station Battery Diesel Generator Emergency Buses and other electrical gear that supply power to critical equipment, including startup transfortaer Instrumentation and Controls Reactor Pressure and Level Instrumentation l Standby Liquid Control System Instrumentation Manual Reactor Control System Control Rod Position Indicating System Reactor Protection System Neutron Monitor System In-Core Neutron Monitors Area Monitors Piping was here considered as one type <f equipment.

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1 Class 2 structures and equipment are those that are not essential to a proper shutdown but are related to the operation of the station. These include:

Structures Turbine Building Service Building Office Building Screenhouse Superstructures Intake and Discharge Structures Offgas Building Equipment Turbine Generator Condenser Cranes Feedwater Heaters and Pumps t

Shutdown Cooling System Condensate Storage Tanks and Pumps Station Auxiliary Power Buses Reactor Cleanup System Liquid Waste Disposal System Moisture Separators and Reheaters Condensate Demineralizer System ,

Air Compressors and Receivers All other Piping and Equipment not listed under Class 1 Note that these classifications differ from those in Regulatory Guide 1.29 (Ref. 3), which was issued af ter the design of Oyster Creek was completed.

I 1.2.2 Principal Structures l

l l The primary structures of the Oyster Creek plant are shown in Fig.1 and identified in Fig. 2. The reactor building (Fig. 3) is a reinforced concrete structure that houses the reactor and its auxiliary systems. Its base is l

approximately 140 f t by 140 f t. The reactor vessel and the recirculation system are contained inside the drywell of a pressure suppression containment f

sy stem. The primary containment system consists of the drywell, vent pipes,

! Have been reclassified as Class 1 equipment (Ref. 8).

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9

and a pool of water contained in the suppression chamber. The reactor building encloses the primary containment system, thereby providing a second containment. In addition, all refueling equipment is inside the toilding, including the spent fuel storage pool and the new fuel storage vault (Ref. 7).

The reactor service and refueling area is served by an overhead bridge crane that is designed to handle the reactor vessel head, the steam separators and dryers, the drywell head, the drywell shielding blocks, and the spent fuel shipping casks during refueling and maintenance. A refueling service platform with necessary handling and grappling fixtures serves the refueling area and spent fuel storage pool. A passenger-freight elevator provides access to the various floor levels of the building.

The turbine building is a reinforced concrete structure founded on sand.

The base of the building is approximately 272 f t by 174 f t. The control room is located at the operating floor level in the northeast corner of the building. A mechanical equipnent room is located on top of the operating floor. All portions above the operating floor are enclosed by a steel frame str ucture. The area between the reactor and turbine buildings is occupied by the office building extension.

1.3 ORGANIZATICE OF REPORT This report contains six chapters. Chapter 2 summarizes our assessment of the ability of Oyster Creek to resist the stipulated SSE event. The chapter also identifies potential deficiencies and areas that my require further study.

Chapter 3 describes the general basis for reevaluation of structures and equipment.

Chapter 4 summarizes the original facility's seismic design criteria for structures, equipment, and piping. The chapter also includes a summary of the original calculated seismic response and accep:ance criteria.

Chapter 5 compares the seismic loadings and responses for which the facility structures were originally designed with corresponding seismic loadings and responses derived using techniques thought to be more realistic in light of current knowledge.

Chapter 6 contains an evaluation of t% tapability of mechanical and electrical equipnent, and fluid- and electrical-distribution systesm to resist 10

seismic loads and to perform their necessary fanctions. Evaluations are based on the floor input generated in Chapter 5, along with other available information.

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CHAPTER 2: SIMIARY AND CONCLUSIONS I

Within' the limited scope of this reevaluation, we examined typical structures, equipment, components, and systems e To assess the adequacy of the existing plant to perform necessary safety functions during and following an SSE.

4 ,

o To qualitatively judge the overall factor of safety with regard to seismic resistance.

e To make specific recomumendations on upgrading oc retrofitting, as ,

appropriate.

i For. the SSE structural evaluation, a peak horizontal ground acceleration of 0.22 g was used along with Regulatory Guide (R.G.) 1.60 response spectra.

2.1 STRUCTURES

. Structural reassessment results are reported for the venLilation stack, condensate storage tank, drywell and torus, control room / turbine building, and 1

the reactor building.

A structure was generally judged to be adequate without the need for additional evaluation if it met one of the following three criteria:

A. Reassessment loads are less thaa original design loads. Here we assumed that the structures were designed and constructed adequately to resist the design seismic loads.

B. Stress evaluations indicate that combined static and dynamic stresses

! (including seismic) do not exceed SEP acceptance limits.

C. For cases in which structural stresses exceed yield, the estimated reserve capacity of the structures would permit inelastic 1

deformations without failure.

2.1.1 Ventilation Stack The stack was seismically reassessed since collapse of the stack may damage the surrounding reactor building and the essential equipment within.

The stack was originally designed for both wind and earthquake loads according 4

to the ACI 505 working stress method for reinforced concrete chimneys.

12

The reassessment indicated a higher seismic (SSE) moment in the lower portion of the stack compared to the original (2 x OBE) design values. The moment capacity of the critical section was then calculated using a cracked section analysis that considered axial load. These calculations indicated that the stack can, without yielding, develop sufficient mo; ....t to resist the current seismic moment.

2.1.2 Condensate Storage Tank The reanalysis indicated that some of the twelve 1-1/4 in, anchor bolts could yield under tension for the 0.22 g SSE. We assumed that the overturning moment due to hydrodynamic pressure and tank inertia force was resisted by the anchor bolts. It was reported in 1968 that the tank anchorage could not resist the calculated seismic overturning moment (Ref. 8) . Apparently no action has been taken to strengthen the anchorage. Further evaluation is needed to examine the effect of bolt yielding on the integrity of other parts of the tank structure.

Details on outlet piping were not available. Piping details should be examined to ensure integrity during a seismic event.

2.1.3 Drywell and Torus Drywell: The drywell was originally designed for both LOCA (Loss of Coc:lant Accident) and seismic loads. The reassessment seismic moments and shaars were larger than the original values. However, the total of the static stresses plus seismic stresses produced under a 0.22 g SSE was found to be substantially below the current code allowables.

Torus: Only the lateral bracing supporting the torus shell was examined. The reassessment indicated a lateral force corresponding to 0.44 g (SSE), compared to 0.35 g (2 x OBE) from the original design. The critical members are the cross bracings (8-in. schedule 40 pipes) . Results indicated that these members have enough capacity to resist 0.22 g SSE loads. To complete the reassessment of the torus system, connection details and columns should be evaluated.

An evaluation of the torus for pool dynamic loads is being conducted as part of the NRC Generic Technical Activity Program, and the seismic loads will be included in those studies.

13

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2.1.4 Turbine Building and Control Roga The control room is structurally connected to the turbine building, being located on the northeast cc rner of the operating floor. Both structures are enclosed by reinforced concrete shear walls. The reassessment analysis indicated that the walls have sufficient strength to resist the 0.22 g SSE loads. This conclusior. is based on a comparison with original design values derived f rom an acceleration diagram presented in Ref. 9. The original design values should be confirmed.

2.1.5 Reactor Building The west portion of the reactor building that is above grade (of ten referred to as the office building extension) houses equipment and cable trays leading to the adjacent control room. This part of the structure was included in the model used in the reassessment. Results indicated that the highest stressed elements in .the reactor building are the first story shear walls in the E-W direction. The walls have sufficient strength to resist the 0.22 g SSE loads.

It is recommended that adequate cable slack be verified to ensure that cables naintain their integrity during differential displacements between the turbire building and the office building extension.

2.2 MECHANICAL AND ELECTRICAL EQUIPMENT, AND FLUID AND ELECTRICAL DISTRIBUTION SYSTEMS As discussed in Chapter 6, typical mechanical and electrical equipment, components, and distribution systems were selected for review by an SEP review team conposed of the authors plus SSRT and NRC staff members. The review was largely based on the judgment and experience of team members. There is wide variation in the documentation of original specifications applied to procurement of equipment, as well as current qualification standards for equipment. While some qualifications for an item of equipment are quite specific, others are generic and apply to a class of equipment rather than specific items at Oyster Creek.

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14

Because we lacked essential seismic design and qualification data on the Oyster Creek plant, our review of the seismic design adequacy of mechanical and electrical equipment is incomplete. Additional tests and analyses that demonstrate functionality of active components must be develcped before definite conclusions can be drawn. In many cases the minimum design details needed to review the design and make appropriate calculations were unavailable. Therefore, we were unable to confirm the capability of a number of mechanical and electrical components to withstand the 0.22 g SSE without loss of structural integrity and required safety function. A summary of the qualification status of various electrical and mechanical components follows.

Chapter 6 contains detailed discussion of the adequacy of these items.

e Emergency service water pump: OK for structural integrity; no design details available to determine functional adequacy; if cast iron

material was used, part of the the pump may be overstressed.

e Emergency isolation condenser: anchor bolts appear overstressed; more data needed.

e Containment spray heat exchangers anchor bolts appear overstressed; more data needed.

e Recirculation pump support: design details unavailable; no evaluation made.

o Emergency diesel oil storage tank OK.

e Motor operated valves: evaluation of design adequacy questionable; Sarther analysis needed.

e CRD hydraulic control system: support system appears overstressed; further analysis required to ensure active function.

e Reactor vessel supports and internals: seismic input loads appear to exceed original design; design details unavailable.

l e Battery racks: OK, contingent upon completion of current modification.

e Instrument racks: OK for structural integrity; no information on func tion.

e Motor control centers: no functional qualification; design details unavailable.

e Transformers: no evaluations; design details unavailable.

e Switchgear panels: no functional qualification; design details unavailable. Panels should be positively anchored.

e Emergency generator: OK for structural integrity; functionality not demonstrated, anchorage integrity should be verified.

15

e Control room electrical panels no demonstration of functionality, design details unavailable.

e Battery room distribution panels no functional qualification, design details unavailable.

e Isolation phase ductwork supports: an analysis of the support system is needed to demonstrate seismic design adequac1' without additional lateral support.

e Electrical cable raceways: analysis or test qualification of support systems is needed to demonstrate seismic design adequacy without additional lateral support.

2.3 PIPING Piping calculations were performed for the SEP evaluation by EG&G/ Idaho and are suimaarized in Ref. 10. Portions of the main stream, feedwater, isolation condenser, and CRD return piping were analyzed by a finite element method, NUPIPE-II. Major conclusions from the summary of that study include:

e On the main stream piping, stresses are within allowable limits during an SSE event. However, possible overloading of several snubbers was indicated.

e For the subject portion of the feedwater piping, maximum stresses were all well within allowable limits, and such piping was therefore considered adequate for SSE loading.

• In the subject portions of the isolation condenser piping, calculated stresses were well within allowable limits during an SSE event.

j e In the CRD return line, maximum stresses exceed ASME Code allowable limits at several points for the " current criteria" loading. Excess stresses were primarily at fittings with high stress intensification factors. Other high stresses could probably be reduced by altering the system support configuration.

2.4 CONCLUDING REMARKS Based on the combined experience and judgment of the authors and the SSRT, reviews of the original design analyses, and comparisons with similar i 16 L___

items of equipment and components in more recently designed nuclear power plant facilities, we conclude thats e Structures and structural elements of the Oyster Creek facility are adequate to resist an earthquake with a peck horisontal ground acceleration of 0.22 g, provided additional analysis, design modification, or verification as stated above are conducted.

o No definitive statement can be made about the overall seismic design adequacy of mechanical and electrical equipment. More data is-needed before equipment seismic design adequacy can be determined in accordance with evaluation criteria in this report.

We therefore recommend e As discussed in Section 2.1, the following actions are necessary to ,

demonstrate seismic adequacy of structures:

o Ensure that the anchorage for the condensate storage tank be -

capable of carrying the SSE loading. Outhlet piping details should be examined to ensure integrity.

o Evaluate the connection details and columns of the torus for the SSE loading.

o Confirm the original design shear and moment values for' the turbine building.

l o Verify adequate cable slack between the turbine building and the office building extension.

e That modificat 6 s and/or additional analyses be made, as necessary, to the me nanical and electrical equipment items listed in Sec. 2.2, in order to demonstrate seismic design adequacy.

r e That modifications to piping be made as listed in Soc. 2.3.

e That all safety related electrical equipment in the plant be checked for adequate engineered anchorage; that is, the anchorage should be found to be adequate on the basis of tests or analysis employing

• design procedures (load, stress and deformation limits, materials, fabrication procedures, and quality acceptance) in accordance with a recognized structural code.

e That a general reconnaissance of the plant be made to identify items that are (1) overhead or suspended, (2) on rollers, or (3) capable of sliding or overturning. All items, whether permanently inatalled or ,

not, that could dislodge, fall, or displace during an earthquake and 17

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thereby impair the capability of the plant to shut down safely should either be modified or moved so that they no longer jeopardize the l plant.

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.k, - a--,L- _ - - - - - - - - - + - A-- 4"+re As- - - -AbAL 9 -+ a = A-CHAPTER 3: GENERAL BASIS OF SEP REEVALUATIG6 OF STRUCTURES AND BQUIPMENT 3.1 GENERAL APPRDACH TO REEVALUATION The seismic reevaluation part of the SEP centers ont e Assessment of the general integrity of the reactor coolant pressure boundary.

l e Evaluation of the capability of essential structures, systems, and components needed to shut down the reactor safely and maintain it in a

! safe shutdown condition during and af ter a postulated SSE. '~nis includes the capability to remove residual heat af ter the earthquake.

To accomplie this level of reevaluation, it is necessary to assess the factors of safety af essential structures, components, and systems of the older plant relative to those designed under current standards, criteria, and procedures. Sud evaluation should help define the nature and extent of l r9trofitting, if any, required to make these plants acceptable if they are not

already at acceptable levels.

As used in the previous paragraph, the term "relativc' is not to be construed as. evaluation based on the norm of current criteria, standards, and procedures, but, instead, in the light of knowledge that led to such a level of design. It would be irrational to assume that an older plant would consist of structures, equipment, components, and systems that would meet current criteria in every instance; even so, those items that do not meet current criteria may be entirely adequate in the sense of meeting the spirit of current criteria.

Within the scope of the investigation, it was impossible to reexamine every item in detail. On the other hand, by examining selected structures, equipment, components, and systems individually, it was felt possible to assess their adequacy and general margin of safety for meeting the selected SSE hazard. Thereafter, on the basis of evaluation of appropriate structures, 4.tems of equipment, or systems it should be possible to provides e Judgmental assessment cf the adequacy of the existing plant to function properly during and following the SSE hazard, including judgmental assessment of the overall margin of safety with regard to seismic resistance.

o Specific comments pertaining to upgrading or retrofitting as may be appropriate.

19

The detailed basis of the reevaluation approach to be generally followed is presented in Refs. 4 and 5. The specific bases of reevaluation are described next.

i 3.2 SEISMIC INPUT AND SITE CONDITIONS Seismicity information forms the basis for arriving at the effective peak transient ground motions (acceleration, velocity, and displacement) for use in arriving at response spectra, time histories, etc. in the reevaluation. The original regional and site geologic and seismic information used to establish 1 seismic input parameters has been reviewed in light of current knowledge. No significant changes in site conditions have been determined. Thus, the l

original 0.22 g level SSE was used for reevaluation along with R.G.1.60 spectra. Preliminary indications suggest that this criterion may be more severe than that suggested by the site specific spectra study currently being finalized. A final comparison can be made between the site specific seismic input and the seismic input assumed in this evaluation when the data become ava ilable.

Another important area for reevaluation is the treatment of SSI. More accurate methods for computing SSI are available today than were in use when Oyster Creek was designed. Existing soils information from both Oyster Creek and the nearly Fork River plant was used in the reevaluation.

3.3 STRUCTURES

! In examining a structure, the first task is to suimmarize the nature and makeup of the structure in the light of both the original design criteria and-information on the as-constructed plant. Also required is a summary of the i design analysis approaches that were used, including loading combinations,.

[

stress and dex'ormation criteria, and controlling response calculations. For evaluating the seismic design criteria, it is generally necessary to know the l

seismic input employed originally, the applicable levels of damping, and the modeling approach used in the analyses.

With the seismic criteria applicable to the reevaluation known, and'with a knowledge of other normal loading criteria - osidered necessary, the response to the seismic excitation can be est.imated. In some cases it may be e

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necessary, as deemed appropriate, to carry out new seismic analyses with the original model or new models.

4 Overall reevaluation of a structure will consider many factors, including those discussed in the following subsections.

3.3.1 Response Spectra, Damping, and Nonlinear Behavior Reevaluation of a structure will include comparison of original response sInstra to the reanalysis response spectra, along with appropriate dasping values and ductility factors. Table 1 compares the damping values specified in R.G.1.61 (Ref.11) with those reconenended in NUREG/CR-0098 (Ref. 4) for reevaluation purposes.

TABLE 1. Damping values from R.G.1.61 compared to those reconumended for the SEP evaluation.

Damping (t of critical damping)

NUREG/CR-0098 (recomunended when stresses R.G. 1. 61 (SSE) are close to yield)

Reinforced c..ncrete 7 7 to 10 Prestressed concrete 5 5 to 7 Welded assemblies 4 5 to 7 Bolted and riveted assemblies 7 10 to 15 Piping 2 or 3 2 to 3 The reason for permitting higher damping values for the reassessment is discussed in Ref. 4. Although there are limited data on which to base damping values, it is known that the R.G. 1.61 values are conservative to ensure that adequate dynamic response values are obtained for design purposes. The lower values in the NUREG/CR-0098 column of values in Table 1 are in most cases close to the R.G. 1.61 values. The upper values in the NUREG/CR-0098 column are best-estimate values believed to be average or slightly above average 21 ,

values. We recomumend that these upper values be used in evaluations of existing facilities for stresses at or near yield, and when moderately conservative estimates are made of the other parameters entering into the evaluation.

Reference 4 reconsnends that low values of ductility factors (1.3 to 2) be used for conservatian and to help ensure that no gross deformation occurs in any critical safety elements. In evaluating safety, assessing the local element deformation and its role in system performance requires careful evaluation and is largely judgmental.

3.3.2 Analysis Models The reevaluation includes a consideration of the adequacy of the models used in the original analysis. This consideration includes an assessment of possible effects of such factors ac soil-structure interaction, overturning, and torsion. In the reevaluation of Oyster Creek, state-of-the-art analysis procedures were used wherever feasible.

3.3.3 Normal, Seismic, and Accident Loadings In the reevaluation we considered the usual combinations of normal

! loadings (dead load, live load, pressure, temperature, etc., as appropriate) with seismic loadings. Design basis accident load effects were not

! considered. However, to preclude an earthquake-initiated loss-of-coolant-accident, the reactor coolant pressure boundary was examined to make certain that under an SSE event the elements comprising the boundary would remain within code prescribed behavior limits.

3.3.4 F_orces, Stresses, and Deformations The reevaluation assessed the reasonableness of the forces (axial and shear forces, and moments) and associated stresses and deformations used in the original design along with their adequacy in the light of the seismic criteria applicable to the reevaluation. Such studies involve consideration of effects arising from horizontal and vertical excitation and take into account the proportion of total effects attributed to seismic factors. Also, the amount of limited nonlinear behavior that is to be acconunodated is evaluated.

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1 3.3.5 Relative Motions The effect of any gross relative motions that might influence interaction effects between buildings is taken into account as a part of the reevaluation.

3.4 EQUIPMENT AND DISTRIBUTION SYSTEMS of particular importance in the reevaluation process is the assessment of the adequacy of critical mechanical and electrical equipment, and fluid- and electrical-distribution systems. The reevaluation centers on those items or systems essential to meeting the general criteria described earlier.

A major task of the reevaluation process is to identify the critical safety related systems and the criteria originally used for procurement and seismic qualification of equipment. For such systems selected, representative items or systems were identified on the basis of t e Physical inspection of the facility (where specific items were identified as possibly having nearly lower bound seismic resistance) .

e Representative sampling.

After system or item identification, and after ascertaining the nature of 1

the seismic criteria used during procurement or qualification, the reevaluation effort involves a detailed assessment of the original design in the -light of current knowledge about equipment vulnerability to seismic excitation. Specifically, the evaluation involves consideration of the following items.

3.4.1 Seismic Qualification Procedures The initial reevaluation assessment is concerned with the original seismic qualification of the equipment item or system, in terms of the seismic test perf ormance (level and extent of testing), or analyses that may have been made, or both.

3.4.2 Seismic criteria The second major aspect of reassessment involves comparison of the original seismic design criteria with those currently applicable. This area of assessment considers such items as the in-structure response spectra, modeling, dynamic coupling, and damping.

23

3.4.3 Forces, Stresses, and Deformations For those items of equipment for which loads, stresses, or deformations may be a major factor in design and performance, the reevaluation involves:

e Examination of the original loading combinations and analyses.

e Calculation or estimation of the situation that exists under the reevaluation criteria. Particular attention is directed to the effect of any increase in seismic component of load, stress, or deformation.

3.4.4 Functionality For those items of equipment that are defined as active in R.G.1.48, qualification testing or analysis, when performed, has been used to demonstrate the structural integrity of such equipment. Operability of such equipment has become a generic concern for all power reactors. In the interim period until the completion of this generic activity, maintenance of the structural integrity and judgments reached concerning operability of such equipment pr Jvide reason.. Die assurance that they will function following the occurrence of an earthquake up to and including the specified SSE.

3.4.5 Nonlinear Behavior Ductility factors in excess of one are not permitted in active equipment unless it can be clearly demonstrated that functional ability is not impaired and a significant margin of performance still remains. In components of passive mechanical and electrical equipment and of distribution systems made of ductile material, component ductility limits should range between three and five.

I i

i 3.5 MISCELLANEOUS ITEMS In a subsequent step of the reevaluation, it may be appropriate to evaluate such items as sources of water for emergency core cooling and to assess whether or not any potential problems

• av' 6 cccur with dams, intake structures, cooling water piping, or othS L' xnr ?. hat form part of the 5

ultimate heat sink.

24

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3.6 EVALUATION OF ADEQUACY Based on the reevaluation assessments as described above, an overall evaluation of the adequacy of the critical structures and representative equipment items and systems was made. This evaluation took into account analytical assessment of f actors of safety, as well as judgment. We also considered the adequacy of individual itent as they pertain to overall system performance.

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CHAP'IER 4: PREVIOUS SEISMIC ANALYSES

4.1 INTRODUCTION

This chapter presents the original seismic design criteria fot. Oyster Creek plant facilities. The seismic loadings and allowable stress criteria for Class 1 structures, equipment, and piping are sununarized, and the calculated seismic responses of critical structures are described. The data presented in this chapter are used to define the design basis and to form the basis for comparison with SEP acceptance criteria in Chapters 5 and 6. Most of the information has been drawn from the FDSAR (Ref. 7); detailed references are given in the sections describing the individual analyses.

The primary structures for which the original seismic design is documented are the reactor building, the turbine building, and the ventilr4 tion stack. In addition, seismic design information is available for the primary containment system (drywell and torus), the reactor pressure vessel, and various other equipment items and piping systems. -

4.2 DESIGN CRITERIA 4.2.1 Earthquake Mot;fn, The Oyster Creek design level earthquake (equivalent to the OBE) was

assigned a peak horizontal ground acceleration of 0.11 g (Ref.12) . The peak horizontal ground acceleration for the SSE was 0.22 g, twice the OBE value. A l

simultaneous vertical component equal to two-thirds of the horizontal earthquake motion throughout the frequency range was also considered in the plant design.

The response spectra at the foundation level corresponding to the original design OBE were those developed by Housner (Fig. 4). The choice of these spectra was based on a limited number of strong-motion earthquake time 1 histories (four earthquakes and eight time histories) recorded between 1934 and 1952 (Ref. 13).

26

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4.2.2 Stress criteria Stresses resulting from OBE excitation in combination with operating load stresses were held to code allowable levels. In addition, the design criteria required that the plant be able to safely shut down at double the OBE level earthquake, which corresponds to the SSE level. When Oyster Creek was designed, emphasis was placed on the lower OBE-level earthquake and as a result, the SSE acceptance criterion was not as specific. In general, stresses for the SSE were limited to yield level. The load combinations and allowable stresses used in the design as presented in the FDSAR are listed in Table 2. Additional detail for certain structures can be found in Ref.14.

4.3 SEISMIC ANALYSES 4.3.1 Methods of Analysf s In general, the seismic design of Oyster Creek was based on calculating the response of the various structures, equipment, and piping to the OBE 27

'mB2 2. Load combinations and allowable stresses for the Oyster Creek reactor building structures (from Ref. 15).

Reinforcinq steel concrete Structural steel Maximum Maximum Maximum allowable allowable Maximum Tension Shear on Compression allowable compressive shear allowable on net gross on gross Load combinations stress stress stress bearing section section section Bending

1. Dead loads, 0.5 F 0.45ff 1.1 y 0.25ff 0.60 F 0.40 Fy varies with 0.66 to plus live loads, slenderness 0.60 Fy pluz operating loads, ratio plus seismic loads (0.11 g)

2. Dead loads, 0.667 F 0.80 Fy 0.53 Fy 0.60ff 1.467/f[ 0.333ff varies with 0.88 to plus live loads, slenderness 0.80 F u plus operating loads, ratio co j plus wind loads

3. Dead loads, Safe shutdren of the plant can be achievede plus live loads, plus operating loads, plus seismic loads (0.22 g)

• F = Minimum yield point of the material.

b Y f'

c = Compressive strength of concrete.

• Where calculations indicated stress levels beyond the yield point, an analysis of energy absorption capacity was to have been made to assure that the capacity exceeds the 0.22 g earthquake energy content.

earthquake excitation described in Sec. 4.2. Most primary structures were modeled as lumped-mass vertical beams and were analyzed for elastic response by the dynamic response spectrum technique. The only exception was the turbine building, which was analyzed by a time-history method.

Original design of the drywell and torus was based on an equivalent static analysis that used a horizontal load coefficient corresponding to 0.22 g and a 0.1 g vertical coefficient. The suction header was modeled as a lumped-mass system for an OBE level equivalent static analysis that used a lateral coefficient of 0.18 g and a vertical coefficient of 0.07 g (Ref.16) .

Af ter it was built, the drywell was reanalyzed as an 18-node, lumped-mass, vertical-beam model fixed at the base; a first mode response spectrum approach both with and without water was used. A reanalysis of the torus by the response spectrum techr.ique was also made. The torus and its supports were considered to be a single-degree-of-freedom mass spring system fixed at the foundation level.

Seismic analysis of the reactor pressure vessel (RPV) system included the vessel, the surrounding sacrificial shield wall (which has a concrete core sandwiched between steel faces), the vessel support, and the concrete pedestal. The system was represented as a two-branch, vertical-beam model, for which the first mode shape and praiod were calculated by Rayleigh's method. Static methods were then used to calculate loadings.

Static methods were also used to qualify most other Class 1 equipment items for which documentation is available. Floor or in-structure response spectra were not generated because the response spectrum method (rather than the time-history method) was used in the analysis of the primary structures.

Seismic design and analysis of Class 1 piping systems was performed by several different methods, depending on the system involved and the pipe size. In general, the specifications called for the following:

e Dynamic analysis was required only for piping 10 in. in diameter and larger.

e Equivalent static analyses were permitted for piping less than 10 in.

in diameter. Static coefficients were 0.43 g horizontal and 0.29 g vertical.

e Pipe supports were to be located so as to limit support loads to less than 10 000 lb per support.

29

Design curves developed by John A. Blume and Associates were used in the generic dynamic analysis of certain Class 1 piping. Pipe spans were chosen as specified in Power Piping USAS B31.1.0 so that stresses were less than 1500 psi at equivalent static loads of 0.5 g. The period of each piping span, which depends on the length of the span and the size of the pipe, was then established from the Blume design curves. Each span was required to respond rigidly, and its period was required to be less than one-half that-of the supporting structure.

4.3.2 Damping Damping values specified for the design of Oyster Creek (Ref. 7) are given in Table 3. Linear (elastic) response of the structures and equipment was assumed.

TABIE 3. Damping values used in the seismic design analyses of Oyster Creek.

Item 4 of critical damping Reinforced concrete reactor building 10 Steel frame structures 2 Welded assemblies 1 Bolted and riveted assemblies 2 Vital piping systems 0.5 Turbine building 5 Reinforced concrete ventilation stack 5 4.3.3 Soil-Structure Interaction Dynamic models of the turbine building and ventilation stack had fixed bases, and the design earthquake motion was specified at the structure foundation. Rocking of the reactor building was analyzed by an equivalent rotational spring model for which the modulus of elasticity for the soil was 800 tons /ft 2, 30

4.3.4 Combination of Response for Earthquake Directional Components Two earthquake directions were considered in the analyses, one horizontal and one vertical. The corresponding response components were then combined on an absolute sum basis. We presumed that the worst-case horizontal load was the one chosen for the combination.

4.4 SEISMIC RESPONSE OF STRUCTURJ.S This section presents the results of analyses used in the original design. In Chapter 5 these results are compared with results from SEP criteria. Note that the original design analyses were not verified as part of f

this program.

4.4.1 Reactor Building The response of the reactor building was determined from a dynamic elastic response spectrum analysis using the lumped-m' ass vertical beam model shown in Fig. 5. (Note: This model reflects modifications to the struct,re f af ter its initial design and analysis in 1964. The modifications necessitated a new seismic analysis, which is reported in Ref.17 and summarized here.)

The top two lumps represent the steel portion of the building; the other lumps represent concrete. Each node har a horizontal (translational) and a rotational degree of freedom.

The analysis was made only for the short (i.e., the weak) E-W direction l

of the building. Natural frequencies and mode shapes were calculated for the eight-mass system, and flexural responses were obtained using the Housner spectrum for IC;. of critical damping--a value selected to account for the approximately 50 ft of embedment. Modal maximum accelerations were reportedly

combined "by taking the root mean square" of the maxima for the first three modes. However, inspection of the calculations reveals that, in fact, the square root of the sum of the squares (SRSS) method was used. For clarity, the term SRSS is used in this report even though the phrase " root mean square" may appear in the referenced original document.

Rocking of the building was analyzed separately. An equivalent rotational spring model was analyzed by the method of generalized coordinates using a linear shape function for calculating the generalized mass.

31

i Ic Ac Weight Elevation 2 (kip)

(ft )

4 (ft ) (ft)

Top 3' 700 156'9" N y f 160 i i 59.3 82.4 /s 1st mode s / -

140 538 138'0" .[ 7s 59.3 82.4 ( /\

o 7870 119'3" \ i i 120 - -

I 2.46X106 2640 '\ - -

"16 000 95'3" '\

support <

skirty '

-Shield Concrete iT r; .

.;

....5...

664. . eg?

.. .... pedestal %  :-

a .=. . . D(Ywell

R eactcr ~ .'.';.' ?.' *.

*'v. : .... s:\i .B:

.[-} building 5,j'[:6 ,

K  :. '

• .. .. . ; .

  :. *.

t.'!d'.- - ,

..**O'.,..

. .* :.. . ;

, ,;

, ,y, ,. . Drywell l ....

. :3 - -Shield wall

..:28 :.'%:

'4* '*

. . ,. .., ; . . ;r. ;;.:;: . . .:; .. r. ,

. .; :.e:. * .-: 1.

,4,: .l. ' ' ' ./. .;.j.'.

. Lateral

*. .

i.: : truss Stabilizer . .;i.s. .

L !}- support c.;;h -  ;, Reactor t' .'k' _

7

' ..;, t . Stabilizerc.i. ' -

bracketg..

pressure vessel e . ;.< ;. 8 ..

.., . '. f, _ ...,

... ;*..  ; r..*; (RPV)  :: *

. ,. ;-

.,.. . ..  ; .

. .i

• C,::. . :..; -

.t. l}y t.n.'i

^ L;. s. V. . . .

.' *p:-

,.,n.... . .. '.9'

..s-

' .~e..

.?. .;. . ... . . . . .

. ...; :.' .......:..-

. ..._;.......-

• . . .;

.,_ . . . .. . _~ _ .

..:... . t-

. . . . ;' * ?6' .

~ .

.1.f.'. c. , , :. c :;. . Drywell W.;'f .

.*.e.

, .. j;/.id.il

 ? lugs

, . .j,.1, . .'* '

M . .. . ; :..-:.i. 4..,e: *'

.* :. '.Q._ j:e.' .; .'

FIG. 11. Schematic of the drywell truss system. The lateral truss system is attached to the drywell at El. 82'2". Reactor pressure vessel supports and stabilizere are also shown.

43

1 4.4.5.4 Suction Header The suction header was modeled as a lumped Oass system connected by beam

~

elements that can account for the effects of flexural and torsional shear, as well as axial deformations. The header is assumed to be fixed at the tee connections and simply supported at snubber connections. The model was used in an equivalent-static analysis with a lateral load of 0.175 g taken from the torus analysis above. No dynamic amplification of torus response was assumed because the fundamental period of the header was found to be 0.064 s, less j than half that of the torus. Seismic loads were considered for two orthogonal horizontal directions in two separate cases and were applied concurrently with

! a vertical load of 0.07 g (2/3 of 0.11 g) . The horizontal and vertical responses were presumably added absolutely. Computer analysis results for j forces, moments, and resultant stresses under the two earthquake load cases are presented in Ref. 27, 4.4.5.5 Reactor Pressure Vessel and Supports Seismic analysis of the RPV system is reported in Ref. 25. The RPV system considered here includes the vessel, the surrounding sacrificial shield wall (the well has a concrete core sandwiched between steel faces), the vessel support, and the concrete pedestal (Fig.11) . The analysis model for the RPV system is shown in Fig. 12. Spring k is calculated from two springs in 4

l j series and represents connecting structural elesients between the vessel and the i

1 reactor building. Spring k represents 2

elements between the shield wall and j the reactor building. A schematic of these connections is shown in Fig. 11. ,

f Rayleigh's method was used to determine the fundamental period and the il first mode shape. The spectral acceleration for this frequency was then taken

from the Housner spectrum for 24 of critical damping and used to compute maximum modal accelerations relative to the building. These modal accelerations were combined on an SRSS basis with peak building accelerations.

Seismic loads were then calculated by multiplying the resulting peak l acceleration by the mass of each mode. These loads were applied as static

(

loads on the RPV system to obtain the design shear and moment for the vessel, l shield, and support pedestal. Seismic analysis results are given in Ref. 25.

Relative displacements among supports were of concern, particularly movement between the base (connected to the reactor building foundation) and f

44

ki = 43 900 kip /ft Elevation . k2= 510 000 kip /ft 93'5" <>

ki Elevation k2 82' 9"-ls

^^^

^ ^ ^ k-82' 2" 8 I

<> 76' 71' 5" d' 7 e 70' Reactor pressure f, Shield

,;;

vessel g , ,6 ,

49'5" 3<, , ,5 47'6" 38'5" 24'4" <>

7 oncrete C pedestal 10'3" imsmr FIG. 12. Lumped mass model of the reactor pressure vessel (from Ref. 25).

Connection and supports are shown in Fig.11.

the lateral support point at El. 82'2". The assumed displacement is 0.03 in.,

and th'e solution was obtained by a strain-energy method (Ref. 25) . The forces and moments produced by relative movement and by inertial effects were added by absolute sum to establish the design forces and moments.

The RPV base-support skirt and stabilizer brackets (Fig.11) were also analyzed for seismic design loads (Ref. 25). The resulting seismic stresses in these two support structures were found to be below yield stress (at 575 F) .

4.4.5.6 Reactor Vessel Internals A short discussion of the earthquake analysis of critical RPV internal structural components is given in Sec. VI of Ref. 28. A vertical seismic loading component was considered insignificant, and a 1.0 g lateral force was applied to the core shroud structure to check stress and deflection. The shroud was considered to be a vertical cantilever with the lateral force of the fuel componr . transferred to the shroud by the core plate and top gtnide.

45

1 i

Under this load, the maximum bending stress la the core shroud support was calculated to be 4400 psi, and the shroud deflection calculated to be less

! than 1/8 in. in the region of the core-spray-pipe to shroud connection. The deflection was. conservatively obtained by assuming that all the loads were applied in the region of the core-spray-pipe connection. Both bending and shear deflections were considered.

4.5 SEISMIC DESIGN OF PIPING SYSTEMS

, Seismic design and analysis of Class I piping systems were performed by several different methods, depending on the system involved and the pipe.

size. The criteria imposed by the plant architect-engineer, Burns and Roe, Inc., are contained in Ref. 49, pertinent requiremer.ts of which include: I e Dynamic analysis was required only for piping 10-in. in diameter and larger. The seismic conditions specified were given in Figures 1-9 of  ;

the specification, which provided deflections, accelerations, and  !

natural periods for the reactor vessel (on its pedestal), the containment vessel, and the reactor building.

e Equivalent static analyses were permitted for piping less than 10-in.

in diameter. Static coefficients were 0.43 g horizontal and 0.29 g vertical.

e Pipe supports were to be located to limit support loads to less than i 10 000 lh per support.

j Implementation of these requirements for various systems is described in Table 5, which shows that Oyster Creek piping seismic analyses were performed by three organizations using several different methods. A discussion of these methods follows, based on Refs. 22 and 30.

l 4.5.1 Analyses by John A. Blume and Associates i

4.5.1.1 PIFI2X/SMIS Analyses l

Except for piping with diameters of 10 in or less, the seismic analyses

in Table 5 done by John A. Blume and Associates involved the response spectrum method. The modal inertial forces were calculated using standard response

! spectrum analysis procedures, and the codirectional modal inertial forces were 46

l Table 5. Summary of seismic analyses of Oyster Creek safety related piping, from Ref. 30.

System Analyst and date Method of analysis Ref.

Recirculation Blume", 1965 (forces, PIFLEX/SMIS 31 moments, deflections only)

MPR Associates , 1979 PIPESD Recirculation MPR Associates, 1979 PIPESD bypass Suppression chamber, Blume, 1967 PIFLEX/SMIS suction header Isolation Brooklyn Poly , 1967 Model test 32 condenser Blume, 1967 PIFLEX/SMIS 33 Shutdown Brooklyn Poly, 1967 Model test 32 cooling Containment Brooklyn Poly, 1967 Model test 32 spray Core spray Brooklyn Poly, 1967 Model test 32 Core spray MPR Associates, 1979 PIPESD 34 (inside containment)

Main steam Brooklyn Poly, 1967 Hand 35 Feedwater Brooklyn Poly, 1967 Hand 35 Class 1 piping Blume, 1967 Conservative F 10 in. diameter generic pipe span criteria Emergency service Blume, 1967 PIFLEX/SMIS 36 water (including 37 buried portion)

• John A. Blume and Associates, Engineers, San Francisco, CA.

Brooklyn Polytechnical Institute, New York, NY.

MPR Associates, Inc. , Washington, DC.

47

then combined on a SRSS basis. The worst-case pipe responses to the three components of ground motion were determined by analyzing five sets of inertial forces--three sets of inertial forces corresponding to each component of ground motion acting separately and two additional sets of inertial forces corresponding to the absolute sum of the vertical and one horizontal component of ground motion.-

The two computer codes used for the analysis were PIFLEX and SMIS.

PIFLEX, a static-analysis code, was used to develop the flexibility matrix for the pipes; SMIS was then used for the dynamic analysis and for calculating the inertial forces, which were then used to calculate displacements, stresses, and other responses. PIFLEX was developed by John A. Blume and Associates in the mid 1960s, and program verification was performed at that time. SMIS is a public domain program and has been widely used. Some aspects of SMIS are self-verifying, such as orthogonality checks. A description and listing of ,

SMIS can be found in Ref. 38. The theory used in PIFLEX is completely described in Ref. 39; however, a listing of this program is no longer available because it has been superseded by newer codes.

4.5.1.2 Class I Piping Less than 10 in. in Diameter The method used to design piping with diameters of 10 in, or less for seismic loadings consisted of two parts (Ref. 9) . First, a qeneric dynamic analysis was performed by John A. Blume and Associates to determine the piping's natural periods as a function of span (i.e., distance between j supports) for various pipe diameters. Based on these analyses and the results of seismic analyses of the buildings in which the pipes are located, ranges of natural periods were determined in which the piping response would be rigid, resonant, or flexible. The general criterion was that the piping was to be supported so that its response would be essentially rigid.

Analyses were then performed to establish, for all pipe diameters used,

! the maximum allowable spans for which piping stresses would not exceed 1500 l psi for a 0.5 g static earthquake load. Based on these data, the piping support designer (Bergen-Patterson Pipe Support Corporation for most of the small piping) selected spans that kept the natural periods in the rigid range and limited the piping stress to 1500 psi. In L few cases the piping span caused the natural period to be slightly into the resonant range, in which i

48

cases the piping stresses would be greater as a result of expected amplification. No cases of piping in the flexible range were reported.

-Generic curves of the piping's natural period versus span, criteria I

establishing the rigid, resonant, and flexible period ranges, and typical i .

results. of piping span, stress, and support load calculations are given in Appendix C of Ref. 9.

4 4.5.2 Analyses by Brooklyn Polytechnical Institute t

Dynamic seismic analyses were performed by J. Curreri at the Brooklyn Polytechnical Institute using two methods:

(1) Scale-model shaker-table tests.

l (2) Dynamic analyses using classical analytical techniques.

No computer programs were used. In both methods, maximum seismic stresses and snubber reaction loads were determined and reported individually for caen of the three earthquake components (N-S, E-W, and vertical) . The method of

} combining the responses calculated by the piping and support system designers is unknown. Accordingly, the licensee reported in Ref. 30 that:

Maximum reported seismic stresses and snubber reaction loads for the worst-case horizontal and vertical earthquake motions have been combined on an absolute basis. The resulting seismic stresses were then added to

piping stresses for normal operating loads (thermal, gravity and l pressure) as specified in the applicable piping Code, B31.1. The

{ resulting combined piping stresses are within code allowables and the

, combined snubber reaction loads are less than the rated capacity of the installed snubbers for all of the piping systems evaluated by Brooklyn Polytechnical Institute.

l i

4.5.3 Analyses by MPR Associates l- Though not part of the original design, seismic analyses were performed

[

l by MPR Associates in a separate part of the Systematic Evaluation Program (Ref. 40). In each case, the analyses were performed by the response spectrum method using the computer program PIPESD, a proprietary program developed by URS/ John A. Blume and Associates, Engineers and G. Powell at the University of California. PIPESD has been verified by several benchmark analyses (Ref. 41) .

Stress and load results were summed on an absolute basis for the worst-case horizontal and vertical earthquakes. Modal responses were combined by the l JRSS method. The results were all within allowable limits at applicable codes.-

)

49

~

- - -., - - ' , - - - + ., -, -

4.6 SEISMIC DESIGN OF EQUIPMENT AND COMPONENTS In general, the design of purchased components such as pumps, valves, and motors for seismic loadings was made a vendor responsibility by the equipment purchase specifications. The general criteria established by General Electric Company for specifying seismic design loads is given in " Criteria for Earthquake Resistant Design of Structures and Equipment for Jersey Central Project" dated March 19, 1965, a copy of which appears at the end of Sec. I of Ref. 9.

Af ter reviewing selected equipment specifications, the utility reported (Ref. 9) that:

e Specified seismic static loads were somewhat inconsistent from component to component.

e Specifications generally did not require specific analysis or test methods.

e Specifications generally did not require that the seismic design basis be provided to the purchaser in the form of documented analyses or tests.

Because of the above and the length of time since the equipment was purchased, vendor documentation showing how specifications related to seismic requirements were met is limited to that shown in Table 6, plus the discussion below on equipment in the Feedwater Coolant Injection (FWCI) system.

The seismic analysis of FWCI piping was discussed in Sec. 4.5. The FWCI equipment-most of which is located on the basement floor of the turbine building-was seismically analyzed using a procedure suggested by John A.

Blume Associates (Ref. 8). The peak (OBE) value of the Hourner design response spectrum for 1% of critical damping (0.33 g) was used to check the adequacy of anchor bolts and support feet or legs, if such supports were u sed. No further analysis was made if stresses were less than those allowed (AISC elastic design formulas with 0.6 F as the allowable stress in tension and 0.66 F as the allowable stress in bending) . If stresses exceeded y

allowable levels, conservative estimates of the period of the equipment were made, and the corresponding (lower) spectral acceleration from Housner's response spectrum used to recheck the anchor bolts and support feet. If stresses still exceeded allowable levels, modifications were recommended so that the equipment would meet the requirements of Class 1. It is unclear when and how these nodifications were made.

50

TABLE 6. Seismic requirements fcr Class I equipment and components (from Ref. 9) .

Purchase Spet.lfied seismic Documentation available Equipment / component 'ipec no. requirements to demonstrate adequacy Remarks Emergency liesel Burns & Roe Normal functioning required Supplier: Electromotive (B&R) S-2299-49 for 0.12 g horizontal (H) Models MP45A 0.07 g vertical (V)

Recirculation GE 21A 1203, 0.2 g H pumps and drives Rev. 3 0.1 g V Recirc. isolation GE 21A XXXX 0.3 g H valves and operators 0.1 g V Recirc. pump MG sets GE 21 A5226 0.05 g H Core spray pumps GE 21A 5404 0.15 g H and motora 0.07 g V Main steam isolation GE 21A 5467, 0.2 g H

$ valves and operators Rev. 0 0.1 g V Emergency condensers CE 21 A1607, 0.15 g R Calculations available Supplier: Foster-Wheeler .

Rev. 1 Vertical heat exchanger B&R S-2294-41 0.24 g H B&R calculations of 6/22/79 Suppliers: 1rUBA and PERFEX (HX) and containment 0.146 g V for original HX (1rUBA) and spray HX (original equipment) new HX (PERFEX)

Transformer Standard seismic requ'ts, of ASA, ASTM, ASME, IEEE, NEMA, NBC and local codes control panels GB 21 A5294 Standard seismic reqat's. Dresden 2 seismic test data of ASA, IEEE, ASME, ASTM, using 1.5 g input AwS, NBFU and local, state, federal codes Switchgear Standard seismic reqat's.

of ASA, IEEE, AS1M & NEMA CRD control units GE 21A 5411 0.15 g H, 0.08 g V

'mBIE 6 (Continued.)

Purchase Specified seismic Documentation available Equipment / component Spec. no. requirements to demonstrate adequacy Remarks

' Station batteries B&R S-2299-50 0.12 g H, 0.07 g V Seismic test data on similar Suppliers Could A & B (old) batteries available Models FTA-21 C (new) JCP&L 3.0 g H, 2.0 g V Calculations available Gould Model NCX 1200 Reactor protectinn None local instr. Racks analyzed instrumencation racks by Blume 12/20/68. Dresden 2 seismic tests e 1.5 g input Tray and conduit B6R 0.26 g H Supplier supports S-2299-51 0.073 g V Hatzel & Baehler Duct work B&R 0.42 g H su Torts S-2299-60-C Inverter and RPS BER 0.12 g H Suppliers GE w equipment S-2299-67 0.07 g V N

Horizontal pump B&R S-2299-40 0.24 g H BER calculations of 6/28/79 Supplier: Worthington (containment spray) 0.146 g V Models 81JI-18 Concentrated vsste tank GE 21 A5264 0.05 g H (full) plus Calculations available (column supported) wind loads Liquid poison tank GE 21 A5452 0.3 g H, 0.1 g V One page calculation available Refueling platform OE 21 A5364 1.0 g H Instrumentation GE None Qualification tests after OC.

See GE Qual. Memo. 91 of 1/26/71 and 92 dated 2/70 4

4

The following FHCI equipment supports, but not the equipment itself, were checked by this procedure. Estimated periods and reduced loads are shown for

-items not able to resist the peak 0.33 g OBE load. Items 13 to 16 did not meet Class I requirements according to the procedure outlined above.

1. Condensate pumps

2. Steam jet air ejector inter-condenser

3. Steam jet air ejector af ter-condenser

-4. Steam packina exhauster

5. Condensate demineralizer cation tank (0.08 o, 0.2 g)

6. Condensate demineralizer anion tank (0.12 s, 0.215 g)

7. Condensate demineralizer resin storage tank (0.11 s, 0.225 g)

8. Dilution hot water tank (0.09 s, 0.2 g)

9. Low-pressure feedwater heater

10. Feedwater heater drain cooler

11. Feedwater pumps

12. Main condenser (period unknown, 0.11 g)

13. Condensate demineralizer mixed bed tank (7 tanks; 0.2 s, 0.3 g)

14. Intermediate p;ttpure feedwater heaters (3 heaters)

15. High-pressure feedwater heaters (3 heaters)

16. Condensate storage tank (3.88 s with sloshing, 0.11 g) .

A separate evaluation of the seismic resistance of equipment was reported to NRC in 1968 by consultant N. M. Newmark (Ref. 42) .

53

CHAPTER 5: REPNALUATION OF SELECTED STRUCTURES

5.1 INTRODUCTION

Since the completion of Oyster Creek in the early 1960s, considerable advances have been made in seismic analysis and design, particularly in:

I e Soil-structure interaction (SSI) analysis.

e competer applications for structural analysis.

e Load combination methods and acceptance criteria.

e Consideration of the inelastic actions of structures.

. As a result, seismic design and analysis are now better defined, and curr.?nt seisnic criteria require less interpretation (see, for example, Refs. 43 and 44).

This chapter reviews seismic resistance of Oyster Creek structures in the l light of current knowledge and criteria. A seismic reanalysis is made to provide a qualitatin assessment of the following safety related structures:

e The reactor building.

e The turbine building.

e The drywe11.

e The torus.

e The ventilation stack.

e The condensate storage tank.

' Seismic loads and responses from the reanalysis are then compared to original design values. These comparisons form the basis for evaluating the

structures. Either:

1. The structure meets current SEP seismic criteria.

or

2. It exhibits low margins compared to current criteria and therefore needs to be ir.vestigated further.

In-structure respos:se spectra were generated for use in reevaluating selected equipment and piping in the reactor and turbine buildings.

5.2 DESIGN FAR'HiQUAKE MOTION As dis.. ..,ed in Chapter 4, Oyster Creek structures were designed for an l OBE characterized by a hcrizontal peak ground acceleration of 0.11 g and l

54 i

l l

l checked for a SSE of J.22 g. A simultaneous vertical component of earthquake motion equal to two-thirds of the horizontal co.aponent was considered in the plant design. The original seismic input was defined by the Housner spectra (Fig. 4).

Curlent practice is to specify either site-dependent spectra, or, as ir more of ten the case, the ground response spectra specified in R.G.1.60.

Housner spectra and R.G.1.60 spectra are compared in Fig.13 for 5% and 10%

damping ratioa. At the same damping. level, the R.G. 1.60 spectral amplification is consistently higher for the indicated range of periods (0.01 to 4 s) or frequencies (0.25 to 100 Hz) . This range covers most of the frequency variations normally encountered during earthquake response.

The reanalysis of Oyster Creek's safety related structures uses a synthetic SSE ground acceleration time-history based.on R.G. 1.60 spectra (see Fig. 14). This time-history is scaled to 0.22 g maximum for the horizontal seismic input, and for the vertical input to two-thirds of horizontal values throughout the frequency range. The spectrum from this time-history closely matches, but does not envelop, the R.G.1.60 spectrum at low damping values.

For SEP reevaluation purposes, the correlation between the two spectra is considered to be acceptable (Ref. 4).

5.3 COMBINATION OF FARTHQUAKE DIRECTIONAL COMPONENTS The original design of Oyster Creek was based on the absolute addition of responses to one horizontal and one vertical load component. In the reanalysis we used the current practice of combining responses for three principal simultaneous earthquake directions by the SRSS method as described in R.G. 1.92 (Ref. 45).

5.4 STRUCTURAL DAMPING Table 7 summarizes damping values used in the original design, those from R.G. 1.61 (Ref. 11), those f rom NUREG/CR-0098 (Ref. 4), and those aCopted for

^

the reanalysis. Values shown in Table 7 reflect values shown in Tables 1 and 2 as applied to the structures and components at Oyster Creek. Reanalysis damping values are the upper NOREG/CR-0098 values .reconsnended for the SEP.

For all structures, reanalysis values for damping equal or exceed values used in t.he original design.

55

1.0 e i i i j i i e i i i i i l

. SSE -

5%

' R.G.1.60 0.22 gi ~

7n 8 5%

'g 0.1 -

Housner - -

.t, 10% -

8 d

0.01 O.01 0.1 1.0 10.0 Period (s)

FIG. 13. Comparison of R.G. 1.60 and Housner's seismic response spectra for St and 10% of critical damping.

t 56

i i i

ieiij i j i i

l i

l iisig l i

ile iei 4 - -

3 - -

en

.h _ _

e j _

i _

8

<2 - -

1 0 ' I ' ' I'I ' I ' ' I'I ' I ' '  !'

O.1 0.2 0.5 1 2 5 10 20 50 100 Frequency (Hz)

FIG. 14. Response spectrum (2% damping) corresponding to the time-history l used in the analysis superposed on the corresponding smoothed spectrum from R.G. 1.60.

l i

57

I ThBLE 7. Original, current recommended, and reanalysis damping values as percentages of critical damping.

Oyster Creek original design R.G. 1.61, NUREG/tR-0098,b Structure or OBE or SSE SSE recommended Reanalysfa component (t) (t) (t) (?)

Reactor building 10a 7 7-10 10 (reinforced concrete structure)

Turbine building 5 7 7-10 10 (reinforced concrete structure)

Stack 5 7 7-10 10 (reinforced concrete structure)

Torus 3 4 5-7 7 (welded assemblies)

Drywell 0.5 4 5-7 7 (welded (without assemblies) water)

Condensate storage -- -- --

• /

tank (aluminum) a Includes soil damping.

b The lower values for each item in the table are considered to be nearly lower bounds for structures with stress levels at or 'just below yield, and are, therefore, highly conservative and suitable for design. The upper levels are considered to be average or slightly above average values, and are acceptable for evaluation of existing structures.

Damping values were selected under the assumption that structural stresses would be close to yield during an SSE. To generate realistic in-structure response spectra for equipment and piping, calculations should be made using structural damping values that are consistent with the las vel of stress in the structure. If calculated stresses using assumed values are much below yield, a smaller damping should be assumed and another calculation made. However, when evaluating for structural integrity under SEP guidelines, only damping corresponding to stresses at or above the yield condition under an SSE need be used. .

58

l 5.5 METHODOLOGY USED IN THE SEISMIC REANALYSIS ,

l The original Oyster Creek seismic design emphasized the OBE. However, the design was evaluated at twice the OBE level of response to assure that safe shutdown of the plant could be accomplished. Emphasis in the reanalysis is on the SSE, with reanalysis of structures primarily made by the time-history method.

The original analyses, which used separate 2-D models to calculate E-W and N-S responses, did not consider 3-D effects such as torsion and coupled responses from the E-W and N-S directions, which may be significant in non-symetrical structures such as the turbine and the reactor buildings.

Mcdels constructed for the reanalysis were three dimensional.

In sumary, the reanalysis incorporat.es the following basic features e Use of 3-D vertical beam models for safety related structures.

e Use of frequency-independent soil springs and damping.

e Use of R.G.1.60 spectra for generating SSE seismic ir.put.

e Use of SEP recomended guidelines (Ref. 4) for selecting structural damping ratios.

e Varying of important SSI parameters to establish bounds on seismic responses (i.e., building shears and moments) and in-structure response spectra.

e Solution by dynamic analysis techniques.

Section 5.6 discusses the determination of SSI parameters used in the reanalysis. Next, in Sec. 5.7 we discuss the dynamic modeling of selected structures. Section 5.8 sumarizes the SSI analyses. Pinally, Sec. 5.9 discusses dynamic analyses of the structural models with soil springs attached. Analyses were made by applying a modified version of the SAP 4 program (Ref. 46) within the LLNL computing system to models of the critical structures, under various combinations of parameters, for the seismic input. ,

Analyses were made in the time domain by using normal mode superposition. Seismic input war. the synthetic time-history derived from R.G.

1.60 spectra. Seismic responses (shears, moments and displacements) in these analyses were computed for each mode and combined by the SSRS method. All modes below a 33 Hz cutoff frequency were included; all significant modes were analyzed.

59

Direct time integration analyses were also conducted for the reactor and turbine buildings, using dashpots to simulate half-space soil damping.

Results, shown in Appendix A, demonstrate the validity of using composite modal damping ratios computed from the stiffness proportional method in

] conjunction with the normal mode method of analysis. -

l The assumed foundation motion was chosen to be standard R.G.1.60 spectra. The R.G. 1.60 seismic input was applied directly at the building foundation level in the reanalysis, with no reduction for embedment effects.

Effects of embedment for SdI considerations are treated in Sec. 5.6.2.

5.6. SOIL-STRUCTURE INTERACTION (SSI) PARAMETERS 5.6.1 Soil Material Properties Used in SSI Analyses

The need to consider SSI effects depends on site soil conditions.

Suba-ade soil information of the Oyster Creek site given in Ref. 47 is depicted in Fig.15. The soil consists primarily of dense sand with the water table 23.5 ft below grade. Bedrock is about 1700 ft below grade. An investigation of the possibility of soil liquefaction during an earthquake, 1

made as part of a seismic analysis of the proposed radwaste building (Ref.

Docket 50219-784), showed an ample margin of safety.

Since the Oyster Creek plant is founded on a deep soil layer, SSI could significantly affect structural responses. The original analyses considered SSI only in the reactor building- where the contribution from the soil rocking mode was combined with fixed-base model respcases.

Soil properties were derived from recent field measurements taken at the nearby Fork River Plant site (Ref. 47). Fork River J3 situ shear wave velocity measurement data indicate that soil layers beneath the turbine and the reactor building foundations can be represented by a uniform medium.

Figure 15 shows soil properties taken from Fork River data and applied to the

reactor and turbine buildings using calculations by the procedure suggested in

' Re f. 48 .

. Determination of realistic soil shear modulus values, G, under earthquake I conditions is controversial. NUREG/CR-0098 suggests that a range of values be i

considered in making an analysis. In keeping with this suggestion, three

( combinations of soil constante were selected.

I 60

El 119'3"--

Data sources:

- Underlying soil composition: OC FDSAR Sec.11.5.2 y% V 5= shear wave velocity: Fork River (Ref.47)

- El 46'6" Turbine Ground level Reactor building ' ~' El 23' wue+g neuf //u/m building Vs ['6N///ft/s

/hf .j f :;NiediurEd'e'nsity'." i.{j7'

/

s,

/

/

/

/

x y ( . . :w: ..

an

m. .. .. ..s. . d . . . . ,r*

. ., g.., gg g.

l

////uae/af//u/,

Base 270' X 170' ElO'

/

ff Vs= 1000 ft/s 25' <

l

+ t ' '

. . . . . .. . .. . . . . . . : .;.' * .1.'

El U El - 19'6d

~* * 'E k El-11' sur//////He < .w Base 140' X 140' :q:":; fMgg .

.jli m 5(Cohansey sand)^g ll65'

" V, = 1200 ft/s 65' <

c;ij

@y _ St

................;,...

, .~. i...S. Sa. .nd. .+ c.. lay

+ silt.

. '. :." . 8.< '

Calculated average values: . A For turbine building: For reactor building: Vs = 1400 ft/s 35' <L .r . . . ....x r!* . . ..

V = 1230 ft/s V = 1266 ft/s 1 Dense sand . f ~.17,00 S S G = 5634 ksf G = 6000 ksf m ~~

3 y = 120 lb/ft 7 = 120 lb/ft3

//$/4 Bedrock FIG. 15. Soil composition and soil properties used in the reanalysis of the Oyster Creek Plant.

First typer G = 1500 ksi, 20% cutoff on composite modal da qing ratios.

The NRC recommendations from Ref. 49 of using a shear modulus of G = 1500 kai for the SSE condition was adopted for one analysis, which alw used a 20%

upper limit cutoff on all composite modal dampirq ratios. Daluping ratios suggested in Ref. 49 were believed to be projections from test data rather than exact values. The 20% cutoff was not suggested by NRC in Ref. 49, but composite modal damping results using the cutoff showed that damping ratios in predominant soil modes are in the neighborhood of the suggested values. 'Ihis type analysis forms a lower bound on the value of G.

Second typer G = Fork River G , no cutoff on modal damping ratios.

In, n situ information was not available for Oyster Creek but was available for i the nearby Fork River site. According to Ref. 50, the shear modulus computed from the i_n, situ shear wave velocity study made at Fork River should be realistic for the Oyster Creek seismic review. This type of analysis uses the low-strain soil shear modulus computed f rom field measurements of Fork River as an upper bound value for G. The actual low strain G may be increased due to the increased overburden caused by the weight of the structures. Thus, the field measured G may approach a best-estimate for that actually occurring during an SSE. As-computed soil damping and soil spring values based on elastic half-space theory are also used.

Third typer G = Fork River G g 2, no cutoff on modal damping ratios.

Results of analyses based on this type are generally bounded by those from the other two types, and therefore represent an intermediate case.

While material damping in soil is generally strain dependent, radiation (or geometrical) damping is not. In the reanalysis, material damping effects of the soil were not considered, except in the stack analysis where radiation damping in the rocking mode was negligible.

For the reanalyses, the equivalent composite modal damping was determined by Eq. (4) of SRP 3.7.2.

Because the water table is close to the ground level (Fig.16), a Poisson's ratio of 0.40 was used for the soil, a value within the range suggested by other investigators for saturated sand (Ref. 51). The soil 3

density was taken to be 120 lb/ft ,

62

Ground level 23'6"

< ! / /ty, v'///o

///ff, //U/o // /////H///o

/ / b-El O'

! L,,,,,,,,,$:II: ,iO iLEl - 19' nauimowouuuu, -e' - s'

~ ~

'//HH /H/ //HH///H/H/

Stack hexagonal base: Reactor building square base: Turbine building rectangular base:

i 45' face to face 140' X 140' 270* (N-S) X 170' (E-W)

E/b = 0.74 E/b = 0.375 E/b = 0.173 E = embedment b = least base dimension Deep embedment case: E/b > 0.15 FIG. 16. Sizes and embedments of bases of the stack, reactor building, and turbine building.

J 63

5.6.2 Stiffness Calculation Including Effects of Embedment Some current methods used to analyze SSI effects are quite sophisticated. The state-of-art is summarized in Ref. 44. According to the NRC Standard Review Plan (SRP) Section 3.7.2 (Ref. 43), the lumped parameter method (using soil springs and damping based on the elastic half-space theory) is acceptable for structures on a deep and approximately uniform soil, which describes the Oyster Creek site.

The foundation embedment and the base dimensions of the stack, reactor building, and turbine building are shown in Figure 16. The present NRC SRP 3.7.2 considers a foundation to be " deeply embedded" when E/b > 0.15, where E is the embedment depth and b the least base dimension. Under this criterion, foundationsipf these three structures are all deeply embedded.

Recent papers by Novak (Refs. 44, 52, 53) and Kausel et. al. (Ref. 54) show that embedment effects can be reasonably approximated by modifying the elastic half-space solution. The seismic reanalysis uses the lumped parameter method as modified by Novak's and Kausel's proposed formulas for considering embedment eff ects.

In the reanalysis, values of soil stiffness and damping were considered to be frequency-independent (Refs. 55 and 56) and were based on the asrumption that the underlying soil is a homogeneous half-space. Approaches proposed in Refs. 44 and 53 gave values of the spring constants, increased to account for embedment, used with lateral, vertical, and rocking modes. For the torsional stiffness, the half-space constant without embedment effects was multiplied by the ratio of the foundation / soil-contact-area to the base area. Another approach proposed in Refs. 51 and 53 gave total spring constants including the embedment effects. Both approaches were used in the reanalysis.

In applying the above approaches, it has been assumed that during earthquake response the vertical surfaces of the foundation. are perfectly bonded to the surrounding soil medium. During an actual earthquake, the effectiveness of contact along the embedded vertical foundation surfaces may be reduced as static soil pressures are overcome by the dynamic soil pressure and tension starts to develop along these surfaces. At SSE earthquake levels it is conceivable that the surrounding soil (backfill) may be pushed back plastically, leaving gaps around the foundation. Throughout the reanalysis the full embedment effect was used. Studies conducted to verify the adequacy of using full embedment effects ate reported in Sec. 5.8.3.

64

l 5.6.3 _ Soil Damping Half-space soil damping constants are colunonly converted to an equivalent composite modal damping ratio, as a percentage of critical damping. Soil damping ratios can also be determined directly from the mass ratio B (which is a function of structure mass, base radius, and soil density), by using formulas shown in Table A-2 of.Ref. 55. For the turbine building, compute 3 mass ratios lead to damping ratios of up to 133% (see Table 8). In contrast, computed mass ratios for the reactor building are much qraster and lead to damping ratios to a maximum of only 66.34 (see Table 9).

For ac61ve degrees of freedom, soil radiation damping values as computed from elastic half-space theory were used for the reanalysis models of the turbine and reactor buildings. Not all of the foundations' degrees of freedom were considered active during the reanalysis. A suunnary of damping investigations is included in Sec. 5.8.2. Soil material damping was not included because material damping values were much smaller than radiation damping values.

In the stack reanalysis, however, the half-sp- soil radiation damping ratio in the rocking mode was found to be less tha .e . For the stack, a 74 soil material damping was used for the rocking mode.

5.7 DYNAMIC MODELING OF SELECTED STRUCTURES The reactor building, the turbine building, and the stack are physically separated, reinforced concrete structures modeled independently by vertical beam models. Models used for the original analysis were 2-D. In the reanalysis, the 2-D model was used for the stack, which is axisysunetric.

Howeves, reanalysis models of the reactor and turbine buildings are 3-D to account for eccentricities of the center of mass with respect to the center of rigidity at each floor elevation.

i I

65

4 TABLE 8. Soil parameters used in the analysis of the turbine building.

Damping B, Mass 8, Soil Modal I Ratio Damping Ratio Lateral E-W 0.126 0.81 N-S 0.147 0.75 Vertical 0.102 1.33 Rocking E-W 0.0111 0.98 N-S 0.0042 1.24 Torsion 0.312 0.31 Stiffness (k, ft, rad)

Case Tla case T2b Case T3b Case T4b G = 1500 ksf Fork River, 50% of Gmax Fork River, Gmax 9 max Lateral E-W Kx 1.11 x 106 3.70 x 106 1.85 x 106 3.70 x 106 N-S Ky 1.04 x 106 3.53 x 106 1.76 x 106 3.53 x 106 Vertical Kz 1.38x106 4.66 x 106 2.33 x 106 4.66 x 106 Rocking E-W Kxp 14.7 x 109 38.52 x 109 19.26 x 109 38.52 x 109 N-S Kp y 25.4 x 109 67.47 x 109 33.98 x 109 67.47 x 109 Torsion Kt 26.1 x 109 77.2 x 109 38.6 x 109 77.2 x 109 a Assumes a uniform half-space soil medium.

b Assumes a uniform half-space below foundation and soil layering within embedment length.

i t

66

,TABIJE 9. Soil parameters used in the analysis of the reactor building.

Damping B, Mass B, Soil Modal Ratio Damping Ratio Lateral J.495 0.409 Vertical 0.411 0.663 Rocking 0.348 0.189 Torsion 1.07 0.160 Stiffness (k, ft, rad)

Case Rla Case R2b Case R3b G = 1500 ksf Fork River, 50% of 9 max Gmax Lateral Kx 0.896 x 106 3.09 x 106 1.55 x 106 Vertical Kg 0.904 x 106 3.30 x 106 1.65 x 106 Rocking K4 7.85 x 109 17.15 x 109 8.58 x 109 Torsion Kt 10.2 x 109 30.76 x 109 15.38 x 109 a Assumes a uniform half-space soil medium, b Assumes a uniform half-space below foundation and soil layering within embedment length.

67

5.7.1 Turbine Buildinq Y

The vertical beam model for the turbine building is shown in Fig.17.

Vertical walls between floors of the turbine building are represented by vertical beam elements. Shear deflection was the predominant contributor to the flexibility of the turbine building. The shear area was taken as the total area of the walls that are oriented in the direction of loading (E-W or N-S). The vertical axis of the equivalent beam representing a wall system passes through the ce cr of rigidity of the wall system, ps computed by the method of Ref. 54. The mass lumped at each floor represents the dead loads due to the floor, tributary masses of upper and lower walls related to that floor, and equipment on that floor. As-built structural drawings furnished by the licensee were used to establish section stiffnesses and mass properties.

Equipment weights were obtained f rom Ref. 57. Under operating conditions, live load effects were considered to be negligible. The control room is located at the NE corner of the operating floor (El. 46'6") in the turbine building. The cable room is located directly below the control room. The turbine pedestal shares a common foundation with the turbine building but is otherwise isolated from it. The control room is represented by a separate branch of the model. The overhead frame and the turbine pedestal are similarly modeled.

Since soil damping ratios were very high (75% or more) in the lateral, vertical, and rocking directions, soil stiffnesses for these three directions were increased to represent fixity conditions as suggested by Ref. 45. The damping ratio associated with soil motion in the torsional direction was calculated to be 31%. The torsional spring constant and damping values as calculated were used in the model.

Because of uncertainties in the soil properties, the turbine building was analyzei for four cases, all of which included full embedment effects. These cases, identified in Table 10 as Cases Tl through T4, were analyzed using indicated values of soil shear modulus G and the composite modal damping ratio.

Computed values of natural periods and associated composite modal damping i

ratios for the bounding Cases T1, T2, and T4 are shown in Table 11.

68

Top steel frame 29 28

..  ; El 109' 0"

-27 26 031  ;; El 89' 0" 25 24 30

/

Frame base is hqj2y El 74' 0" slaved [19,h El 63' 6" Control room housing to node 16 18

'g o17 .

Operating 7oy El46'6" floor oIS /

Mezzann El 23' 6" (Ground level)

, uj4 12 Turbine pedestal K 4 g Soil springs

./ /' E

/ l'U E 23 2

/

{6/"8 h/ /'8' / K*EE3 - El O' 0" (Foundation base)

Y North x  : 0*

O K

t K,

mm + = Linear spring I - = Torsional spring W = Center of mass e = Center of rigidity FIG. 17. Mathematical model used in the reanalysis of the turbine building.

69

1RBLE 10. Analyses made using the turbine building model.

Active soil Soil modulus Composite damping Embedment Case modes G ratio consideration Tl Torsional 1500 ksf 20% cutoff max. Kausel T2 Torsional Gmax As computed; Novak

= 5634 ksf no cutoff T3 Torsional, 4ax/2 As computed; Novak lateral no cutoff T4 Torsional, Qax As computed; Novak lateral no cutoff The value G is that value associated with the low strain soil condition found at Fork River.

l 7

70

1RBLE 11. Turbine building P Mal frequency and damping.

Case Tl Case T2 Case T3 Composite Composite Composite-modal modal modal Mode Period damping Period damping ~ Period damping number (s) ratio (a) ratio (s) ratio 1 0.668 0.070 0.668 0.070 0.670 0.070 2 .510 .070 .510 .070 .510 .072 3 .357 .070 .356 .070 .360 .076 4

4 .244 .200 .219 .100 .251 .418 5 .217 .100 .166 .143 .220 .752 6 .154 .110 .132 .247 .182 .395 7 .112 .100 .112 .101 .170 .146 8 .075 .070 .075 .070' .131 .252 9 .063 .079 .062 .070 .104 .126 10 .062 .094 .059 .106 .075 .070 11 .056 .071 .056 .070 .062 .070 12 .055 .105 .051 .108 .056 .070 13 .048 .102 .047 .106 .053 .093 14 .039 .107 .038 .100 .044 .107 15 .035 .101 .035 .101 .042 .107 16 .034 .104 .C33 .113 .038 .100 17 0.030 0.126 .028 .140 .034 .100 18 .028 .100 .033 .113 19 .027 .136 .028 .010 i

20 .025 .100 .025 .100 21 .022 .103 .022. .100 22 .019 .100 .020 .100 23 .019 .101 .020 .100 24 .018 .102 .017 .100 25 . .012 .100 .013 .100 26 .012 .070 .013 .070 27 .013 .070 .013 .070 28 .012 .099 .012 .099 29 .010 .099 .010 .099 30 0.010 0.100 0.010 0.100 71

5.7.2 Reactor Building The vertical beam model of the reactor building is shown in Fig.18.

This model incorporates the office building extension, the drywell, and the reactor pressure vessel (RPV), including the shield wall and the RPV support pedestal. The submodels of the drywell and RPV are similar to those used in the original analysis (Ref. 58) . The two spring constants gK and K shown 2

in Fig.19 at about El. 82' represent, respectively, the connecting element stiffness from the RPV to the drywell (stabilizers) and from the shield wall to the drywell (truss) . K values were taken from the original analysis. The spring constant K3 represents the shear lug connection between the drywell and the reactor building at El. 82' (see Fig. 9 for a sketch of the connection).

Soil properties used in reanalysis of the reactor building are given Table 9. Values of the soil shear' modulus G ?nd of the composite damping ratios used in the cases analyzed arc summarized in Table 12.

TABLE 12. Analyses made using the reactor building model.

Active soil Soil modulus Composite damping Embedment Case modes G ratio consideration R1 Lateral, rocking, 1500 ksf 20% cutoff max. Kausel vertical, torsional R2 Lateral, rocking, %ax As computed; Novak vertical, torsional = 6000 ksf no cutoff R3 Lateral, rocking, Sax /2 As computed; Novak vertical, torsional no cutoff Computed values of natural periods and associated modal composite modal

, damping ratios for the bounding Cases R1 and R2 are shown in Table 13.

i 72

TABIE 13. Reactor building modal frequency and damping.

Case R1 Case R2 Mode Period Couposite modal Period Composite modal number (s) damping ratio (s) damping ratio 1 0.522 0.330 0.331 0.257 2 .522 .330 .331 .257 3 .448 .652 .238 .634 4 .250 .180 .209 .082 5 .240 .240 .209 .083 6 .230 .207 .171 .163 7 .210 .080 .165 .213 8 .210 .083 .162 .178 9 .140 .079 .127 .151 10 .140 .079 .127 .152 11 .098 .110 .090 .083 12 .090 .082 .090 .083 13 .090 .082 .082 .127 14 .070 .081 .065 .081 15 .067 .081 .066 .081 16 .058 .095 .057 .093 17 .055 .103 .053 .110 18 .052 .110 .050 .108 19 .050 .098 .050 .108 20 .048 .087 .048 .086 21 .048 .086 .048 .089 22 .046 .100 .044 .107 23 .044 .100 .044 .089 24 .044 .088 .043 .105 25 .043 .100 .042 .103 26 .041 .084 .041 .096 27 .035 .084 .035 .085 28 .035 .084 .035 .085 29 .032 .075 .032 .074 30 .032 .074 .032 .074 31 0.031 0.100 0.031 0.104 73

TABLE 13. (Continued. )

Case R1 Case R2 Mode Period Composite modal Period Composite modal number (s) damping ratio (s) damping ratio 32 .030 .100 .030 .101 33 .029 .100 .029 .104 34 0.028 0.094 0.028 0.094 1

l i

l 74 .

l Elevation 156'9" 1129 '

> Top frames 119'3" l / Top floor Drywell Shield Reactor Elevation 95, 3,, 24'7 El wall vessel 3 022 94.81146 1135 93'5" 23..J g 3 45 2 40 K 1 ;34 82'6" 21tWV r v/ .v/

Il 20 9 1B33 71' 5" 18,,319 ll 44 1138 o 37 j j d 416, 1143 1I36 1132 49'5" 31 - -

• 31 38'5" 1H2 23'6" 1405 RPV 24'4" 120 13 pedestal II30 (Ground level) 11o=- - - - - - - - - - --

11 10'3" 19'6" O'0" y(North

/ 811

/ ,y,Y

[ 10 A El El -10' 3" 7

W~o bx <

f K

3 V1 = 48 E ksf (Foundation base) 4'Q S "" d'2 "2- = 510 000 ksf N

K 3= varies (3 cases)

= Rotational spring 3 Kt Vv' = Linear spring s E = Center of mass 1 * = Center of rigidity FIG. 18. Mathematical model used in the reanalysis of the reactor building.

75

5.7.3 Stack As-computed half-space damping ratios were used for the two stack SSI analyses. Case S1 used a shear modulus of 1500 ksf and a stiffness comprising the half-space stiffness plus the stif fness due to embedment, as calculated by Novak's method. Case S2 used the Fork River shest modulus G and as-computed modal damping ratios.

Computed values of natural periods and associated composite modal damping ratios for Cases S1 and S2 are shown in Table 14.

5.7.4 condensate Storage Tank The condensate storage tank is 45.0 f t tall and 45.0 f t in diameter.

It is made of aluminum plates that vary in thickness from 0.591 in. at the base to 0.250 in. at the top. The liquid depth in the tank was aseused to be 43.5 f t, which places the surface at the invert of a 12 in. overflow.

At ground level, the tank is secured to a concrete ring foundation by 12 equally spaced 1-1/4 in, diameter anchor bolts. The bottcun of the tank, which is not reinforced, bears upon backfilled soil within the ring foundation. The tank including the fluid weighs about 4367 kips and produces an average soil pressure of about 2.75 ksf.

The reanalysis followed methods reconunended in Ref. 59 and includes both convective and impulsive effects. The original analysis of the tank is given in Ref. 60.

5.8 SUP9tARY OF SSI ANALYSES After the major analytic work was completed, the current SSRT produced soil-structure interaction guidelines for analyzing the remaining SEP plants. These guidelines are included in Appendix C. In preparing this report we found that the effects of differences in the methods used herein and those recommended in the guidelines were minimal, and would not affect our conclusions. The results of the comparative studies are reported in Appendix A.

The analyses indicated that the soil modes predominated and that the contribution of the structure response was minimal. Because of soil 76

MBLE 14. Ventilation stack estural frequencies and modal damping values.

Case S1 Case S2 Mode Period Composite modal Period Composite modal number (s) damping ratio (s) damping ratio 1 1.791 0.094 1.791 0.094 ,

2 0.586 .097 0.586 .097 3 .266 .110 .166 .110 4 .180 .472 .179 .472 5 .155 .150 .155 .150 a

6 .112 .204 .112 .204 7 .085 .133 .085 .134 8 .080 .157 .080 .157 9 .063 .106 .062 .106

! 10 .048 .101 .048. .101 11 .043 .112 .043 .112 12 .038 .100 .038 .100 13 .031 .100 .031 .100 14 0.029 0.105 0.029 0.105 domination, limited soils data, and the uncertainties in SSI parameters, many cases were studied to determine bounding results. The effects of parameter variations are discussed in the following sections.

5.8.1 Effects of Soil Shear Modulus Variations The range in soil modulus values used included the range suggested in the SSRT guidelines. However, since the calculated radiation damping was very large for the turbine building, three of its soil spring stiffnesses (lateral, rocking, and vertical) were increased to represent fixity for the low soil shear modulus case. The rocking aM vertical degrees of freedom were always fixe 1 ,or the turbine building. Increasing the stiffnesses had the effect of negating the low soil shear modulus (1500 ksf). A simplified analysis of the turbine building, reported in Appendix A, showed that the low shear modulus 77

case without the fixity representation (increases in stiffness) followed the general trend exhibited by the reactor building results. h t is, for the in-structure spectra, decreasing the soil shear modulus shif ted the curves in the longer period directions. This effect was more pronounced in the sho:ter period region of the spectra. h magnitude of the spectral peak was essentially unchanged. h case employing G,,,/2 was bounded by the other cases.

Variations in soil shear modulus did not have a marked effect on the moments and shears for the two buildings. Moments and shears were close to the original design values for the reactor building and below the original design values for the turbine building.

5.8.2 Effects of Damping Variations Numerous studies were made to determine effects of varying the desping parameters. These varf ations were studied because high damping ratios were calculated by the half-space approach, and their use in modal analysis was controversial. Also studied were effects of limiting damping values to arbitrarily chosen values, of treating degrees of freedom as fixed when the damping ratio was high, and of more accurately calculating the response with simplified models that included dashpotG to simulate soil damping.

General trends associated with variations in composite modal damping ratios are summarized in a background report by Tsai (Ref. 61). Increasing the radiation damping ratio shif ts the fundamental mode of vibration from a predominant soil mode to a structural mode. As damping increases, a fixed base condition is approached. A study was conducted to see if a fixed base condition could adequately represent the turbine building, whose radiation

damping ratios are 804 for lateral and 100% for rocking. The fixed base

! representation greatly modified in-structure spectra in the shorter period i region and increased the zero-period acceleration. It was felt that the results from the fixed base model were not reasonable. h refore dashpot

analyses were conducted to check the fixed base results. These anaPoes also allowed a check on the use of modal analysis with high desping values.

The dashpot analyses confirmed that the fixed base model gave overly conservative results. As a result of the anomalous results from the fixity case for the turbine building, the final in-structure spectra envelopes did not include the fixed base case.

78

In general, at the higher structure elevations dashpot analysis with high radiation damping produced larger in-structure response spectral accelerations than did the normal mode analysis, using stiffness-proportional, composite modal damping. Envelope curves shown in Appendix B were modified to account for these increased accelerations.

The SSRT-SSI guidelines recommended using 75% of the theoretical radiation damping in the horizontal directions with 54 material damping added to both horizontal and rotational directions. A final study was conducted to investigate the use of these guidelines. The results of this study were f compared to results obtained using 100% of theoretical values without material I

damping. The in-structure response spectra at the top of each building were essentially the same for both cases. Therefore, no further adjustments were made to the envelope spectra in Appendix B.

5.8.3 Effects of Embedment Variations As stated in section 5.6.2, full embedment was used throughout the reanalysis. The validity of this assumption is supported by studies described in this section.

The SSI guidelines from the SSRT indicated that full embedment may not be effective as a result of possible separation of soil around the foundation.

Therefore, to account for the possibility of soil separation, the guidelines suggested using.only 50% of the calculeted effect of embedment. The embedment effect can be Jaeasured by the percent increase in soli spring constants over those f rom the elastic half-space solution. The ratios of the full embedment soil spring constants (used in the previous turbine and reactor building analyses) and those of the elastic half-space solutions are given in

( Table 15. The ratios corresponding to the 50% effect of the embedment case are also indicated. These results are for tha models using G .

The results in Table 15 for the turbine building indicate that using 50%

of the embedment effect instead of 100% will only slightly reduce the overall soil spring stiffness (less than St for lateral, rocking, and vertical modes and about lit for the torsional mode) . Therefore, the effect on overall seismic response results will not be significant. Moreover, the case of 50%

reduction on soil spring constants has been considered in the reanalysis (Case T4). Solutions corresponding to the 50% embedment should be bounded by the solution using G and the one using 50% reduction in soil stiffness.

79

TABLE 15. Variation of soil spring constant due to embedrent effects.

Turbine building Full Partial Reduction embedment embedment factor (t)

Soil mode (1)a (2)b (3)c Lateral E-W 1.08 1.04 3.7 N-S 1.09 1.045 4.1 Vertical 1.04 'l.02 1.9 Rocking E"W l.08 1.04 3.7 N-S 1.04 1.02 1.9 Torsion 1.28 1.14 10.9 Reactor building Lateral 1.31 1.155 11.8 Vertical 1.18 1.09 7.6 Rocking 1.25 1.125 10.0 Torsion 1.88 1.44 23.4

• (l) = K g/KH.S. "

Kg = Soil spring considering embedment effects; KH.S. = elastic half-space soi_ spring constant (no-embedment case) .

b

= 1. 0 + ( (l) - 1.0)/2.

c ((2)

3) = (1.0 - (2)/(1)) x 1004.

80

t .

l l

t A similar trend is observed in the reactor building results (Table 15),

~

except that the reduction factor is larger. .Again, the largest reduction occurs in the torsional mode (a factor of about 244) .- The case of 50%

reduction in all soil stiffnesses (Case R3) was 'also considered for the '

. reactor building reanalysis. Solutions using 50% of the embedment effects L

should be bounded by the solution using 100% of embedment and the one using a 504 reduction in soil stiffness.

5.9- ANALYSIS OF RESULTS FOR SELECTED STRUCTURES Seismic loads computed in .the reanalysis were used to evaluate the seismic capability of the selected safety related structures. Three measures of adequacy were considered:

e If seismic loads of the reanalysis'were less than those used in the original design, the structure in' general was judged as adequate without additional evaluation.

I e If seismic loads from the reanalysis exceeded the original loads but j resultant stresses were low compared to the limiting stresses for SSE, structures were also judged to be adequate. ,

e If seismic stresses were not low and factors of safety were j significantly less than one, the structure may still be adequate because of its ductility. In accordance with NUREG/CR-0098, stresses

{

above yield can be acceptable provided the structure can perform its required shut-down function. References 62 and 63 discuss rational ductility requirements for inslastic response of safety related structures.

Analysis of computer results indicates that frequencies of predominant soil modes (those with relatively large composite modal damping ratios) are lower than frequencies of modes associated with the concrete portions of the buildings. However, the first three modes for the stack are primarily structural modes with little soil participation, as is common for a tall, flexible - structure.

l l

i

'N 81 l l

l

5.9.1 Turbine Building and C'fatrol Roor Housing In Fig.19, reanalysis shear and moment distributions are compared to original values for E-W (transverde direction) loading. Reanal'fsis values are generally much less than the assumed original ones. That : s, the load m

ratio, defined as the ratio of load in the original deatgn to load of the reanalysis, is much greater than one. However, Fig. 19 shows that the moment, i

as reanalyzed, does not reduce to zero at El. 46'6" as in the original analysis. The non-zero moment at El. 46'6" of the reanalysis comes from seismic effects of the top frame and control room housing, structures not included in the original analysis.

Torsional loads from the reanalysis are shown in Fig. 20. No comparisons with torsions of the original analysis are possible because the original analysis used a 2-D model. The added shear load due to torsion in the E-W cross walls can be conservatively estimated by assuming that the calculated torsion is resisted by a couple in the E-W cross walls. The load in the couple is about 1100 k per wall for the first floor (El. O ' to El. 23 '6") and

-1 1260 k for the second floor (El. 23 '6" to El. 46 '6") . l Estimated shear loads due to torsional loads are about 10.4 to 16.6% of the direct shear loads. Since total shear (the sum of direct shear loads and those due to torsion) is far less than the assumed original design values, no further shear analysis was performed.

The seismic loads on the two-story control room housing were not obtained in the original design, since the original model did not include this structure. The reanalysis showed that the first-story walls of the control room housing are more highly stressed than those on the second story. The f maximum shear and in-plane stresses developed in control room walls were found ,

to be below 70 psi. Therefore, the control room housing would have an ample safety factor against the 0.22 g SSE.

l 5.9.2 Reactor Building The major resistance to lateral force is provided by the shear walls of the building. No significant torsional response was found in any of the cases studied. Further examination of the SSI mechanism (see Appendix A) revealed l

that the building seismic response is mainly controlled by soil response f

82

:- 109'0" Top frame o and control Moment from top room housing frame and control room 50 i 1 I i i i o- 46'6"

! Operating floor Operating fioor I

I 40 -

l -

,e Case T1 Original design

-(it Case T1 -

j ' '

(2 X OBE) i

! Case T2 Case ig

, T2 j 3 30 -

l

-j Qw T4 -

i '

e Case ig

,o / , T4 1 8 23'6" f  ! ' *

1. 1. 8" "* -

, 20 -

l - -

's, ' -

W l , \ Original I

is design g

l i '1 \'

w s 10 - 1 - - 't -

t I

\

E 0 '0" Base slab I

I Base slab I

i O I I I 8 0 10 20 30 0 2 4 6 8 10

- SSE Shear (10 3 kip) 6 g ', Model SSE Moment (10 Kip-ft)

- (a) (b) (c)

/

mm i FIG. 19. SSE shear and moment-responses for the turbine building. (a) Simplified schematic of the model used in the reanalysis. (b) SSE shear for three cases (of four considered) of the reanalysis is compared to values of the original design. (c) SSE moment for the same three cases of the reanalysis is compared to values of the original design.

E

109'0"

, Top frame g and control =

l i i i i room housing . N-S 50 -

ll 46'6,,

Operating floor (El 46'6") 270' E-W d l ++e i

l T 40 -

Case T4 l 8 170'

' 8 Case T2 l Bldg sect. ion

= s O I l

Case T1

.! 30 -

% I E e a I

23'6" Mezzanine '----- i, '

floor  !

20 - (El 23'6")  !

l I I 8 I I

i l 8

10 -

! 11 l

l-I i

• i Base slab

' I ' I ' ' ' I

< O'0" 0 O 1 2 3 4 ooo

[ Torsion (105 kip-ft)

(SSRS value of responses from

" E-W and N-S loadings) mm FIG. 20. SSE torsion response for the turbine building.

84

modes, with not many fixed-base modes being excited. Since the reactor building has a square foundation embedded in soil, the soil modal responses in the E-W and N-S direction are the same, and the building responds in the same manner in these two directions.

Shear and moment results from Case di and Case R2 are compared with those of the original design values in Fig. 21. The reanalysis shears stop at elevaton zero because of the soil spring influence below this point. All reanalysis results are close to original design values; when original values are exceeded, it is mostly in the lower half of the building. Although Case R2 values tend to exceed original values more than Case R1, the lowest load ratio (of values in the original design to that in the reanalysis) for shear is about 0.85 (at Ei. O', Case R2) and about 0.87 (at El. O', Case R2) for moment.

Maximum stresses for reanalysis loads occurred, however, at the first story close to El. 23'6" (ground level) . Maximum shear stress (conservatively calculated based on the areas of the cross walls in the direction of loading) was 132 psi, a value slightly above the diagonal-tension cracking limit of the concrete but considerably below the combined steel and concrete sherr capacity of the wall. The maximum tensile stress due to overturning moment is only about one-third of the concrete's flexural cracking strength.

5.9.3 Ventilation Stack The stick model for the stack, a 2-D model, is shown in Fig. 22. The stack itself was designed by using the ACI 505 working stress method for chimneys (Ref. 21), considering loadings for both wind and an OBE earthquake.

Original OBE seismic loads (shears and moments) were provided by John A. Blume and Associates. Figure 22 compares shear and moment distributions of the original analysis with those of reanalysis models using R.G. 1.60 time-history inpu ts . Differences in distribution curves are attributed to the inclusion of a soil rocking mode in the reanalysis of the stack response. Periods of the stack are listed in Table 14.

Seismic shear forces, as reanalyzed, are generally less than original design values (Fig. 22) and. correspond to shear stresses much less than the shear capacity of the stack.

85

Elevation 29 . - 156*9" ,, , ,,,,,, , ,

o - 138'0" - - -

o - 119'3" q - - -

1 I

I 45 35 -l o o o - 95'3"

_ Original design wwe E (2 X OBE) e o -

g 75'3"

- jq B l Ei l I

o - 51'3" -

i7- - -

Original design co Case R1 C f ase R2 (2 X OBE)

\

o - 23,6,, - [! .F- - -

Case R1 s

\.

s Case R2 10'3" -

I o -

0'0" -

.l. - -

N E \

' ..'" \

E _ _ 3 9.g. , , , , , , i, i i i i

[ 0 4 8 12 16 20 24 28320 SSE Shear (103 kip) 0.5 1.0 1.5 2.0 6 kip-ft) 2.5 SSE Moment (10

{z (b) (c)

/

wm

-Model (a)

FIG. 21. SSE shear and moment responses for the reactor building. (a) Simplified schematic of the model used in the reanalysis. (b) SSE shear for three cases of the reanalysis _are compared to values of the original design. (c) SSE moment for three cases of the reanalysis are compared to values of the original design.

3 g i i i i i i i 1 k Original design

\

300 -

- \ ~

\

\

\

- 4, . .

2 i b g 200 -

\ -

g 8

- { Case S1 Case S2 Original design g

=

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g -

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Case S1

~

37 o s -

\

38 o 39 o K xE

\N K E

~3 -

O 200 1 I 400 I - - t I f h -

4 600 800 1000 0 200 400 600 800 1000 g< Model SSE Shear (kip) SSE Moment (10 2 kip-ft)

(a) (b) (c)

K, mm FIG. 22. Mathematical model and SSE responses of the stack in the reanalysis. (a) Model of the stack.

(b) SSE shear responses for two cases compared with the response calculated in the original design. (c)

SSE moment responses for two cases compared with the response calculated in the original design.

1 Moment values found in the reanalysis are also shown in Fig. 22. They-exceed original values slightly at the base, and by larger amounts in the mid portion of the stack. The reanalysis shows a peak overturning moment at the base of about 78 000 k-f t. The original design showed the stack foundation capable of resisting an earthquake overturning moment of 94 000 k-f t with a f actor of safety of 1.5 (see Section 4.4.4).

Moment capacities of the stack were determined by a cracked section analycis assuming that concrete does not resist tension. . The reanalysis assumed 5000 psi concrete and 60 grade reinforcing steel, based on information given in the FDSAR amendment (see Section 4.4.4). At all sections of the stack, moment capacities at yield are higher than moments induced by the

, 0.22 g seismic loading.

4 5.9.4 Primary Containment (Drywell)

The drywell is included in the reanalysis of the reactor building model (Fig. 18). Originally, the drywell was analyzed independently as a cantilever beam and was studied both with and without lateral scpport at El. 82'2". The case of no lateral support is the more critical condition. Neither the original analysis aor the reanalysis included any lateral support from conpressive styrofoam material that fills the gap separating the drywell and the reactor building's reinforced concrete shield wall.

A sketch depicting the drywell support, which censists of eight evenly spaced shear lugs at El 82'2", are shown in Fig. 12. Corresponding female brackets are mounted on the inner f ace of the reactor building shield wall.

The reanalysis showed that Case R3 gives the most critical moments and shears for the drywell. Three lateral support cases were examined representing different lug stiffnesses. Associated lateral support stiffness values for the lugs were Case 1 K e , i.e., n lateral support 3

Case 2 K 3 = 50 000 k/ft

< Case 3 K 3 = 000 000 k/ft.

Figure 23 shows the original and the reanalyzed shear and moment diagrams for the enoty drywell. Tlw flooded condition was not evaluated because of the low probability of the max: mum earthquake occurring for this condition. Shear and moment values from the rea,alysis are significantly higher than original 88

Elevation 4

107'9" - o 1 o2 o 3 94*9" o4 ' ' ' '

i i I I I )

I l i l l' i i I o5 90 - -

90 -

l l 1 5'"'"

o6 l taterai f'**"*' '"PP " P 80 -

I l 1 N 'p"o7t?"n - 80- ' -

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l ,

71'6" - o8 l l -

70

'h Reanalysis:

5 ]

70 o9 Case R2: K3 = 0 k/f t 65'6" -' '10 l l ,.

h. - Case R2: K3 = 50 000 k/ft

- Reanalysis: I _ lt Case R2: K3 = 5 000 000 k/f t "ll 5 60 -

Cas. R2: l I

1 --

E 60 -

ils' *g C K c d'12 3Case

= 0R2:k/ft"! l

.o l' 9 nel anaIYsis:

\.kg K 3-50 N / ~~l 50 - *!'[*,*,'a',','r"J"p"pon w, -

50 g 49'3" -o 13 , l -

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l o14 40 -

l 40 -

,; i

\\

}

37'3" -

oB ,' \

o15 30 '

30 l i - -

s'

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i i i l

g

\

22'6" - o 16 20 -

originai anaiysis:

l 20 -

i, \. g -

4,l7 Without lateral support V j g k,

, ,gg With lateral support J g [g I ' l ' I I I ' I ' I ' I  ! II I ' I I 10'0" 10 10 mm ~400 0 400 800 1200 0 20 40 60 80 SSE moment (10 3 kip-ft)

Model SSE shear (kip)

(a) (b) (c)

FIG. 23. Mathematical model and SSE responses for the drymell. (a) Model used in the reanalysis. (b)

SSE shear responses for three cases are compared to the values calculated in the original design. (c) SSE moment responses for three cases are compared to the values calculated in the original design, o

design values at points below El. 82'2", primarily because part of the inertial force from the reactor pressure vessel and sacrificial shield wall is transferred directly to the drywell at El. 82'2". The portion of the seismic load transferred depends on the assumed stiffness of the lateral support: the smaller the stiffness, the larger the load on the drywell. Moments and shears corresponding to a stiffness of 50 000 k/ft, a value considered realistic for predicting drywell seismic loads, were used to compute drywell stresses.

The maximum stresses computed for SSE loads are at the base of the drywell shell where the maximum stress is 15.45 ksi in the meridian direction. The maximum dead weight stress and the pressure stress (corresponding to a 3 psig pressure differential across the shell) at the same location are respectively 0.945 ksi and 0.54 ksi. The maximum combined tensile stress due to SSE, dead weight, and pressure loadings is low compared to the yield strength of the drywell shell, with a factor of safety larger than two. However, the factor of safety in buckling is more critical, being about 1.44. This safety factor is based upon allowables from ASME Section III Subsect. NE 3133 applied to a cylinder under pure axial loading and the faulted condition.

5.9.5 Suppression Chamber (Torus)

Since sufficient details were not available, only the lateral bracing supporting the torus shell was examined. The torus is represented by a single mass stick in both the original analysis and the reanalysis. The reanalycis predicts a SSE lateral loading of 0.44 g compared to the original analysis loading of 0.175 g OBE and 0.35 g SSE (normal water level condition) . The main lateral supports for the torus are 20 outer columns with cress bracing.

Assuming that bracing will not carry compression, the maximum stress developed in the bracing (which is 8 in schedule 40 steel pipe) is about 40 ksi. The bracing material is believed to be A36 steel with a yield strength of 36 ksi.

Therefore, maximum stress exceeds yield stress by about 10%. However, in view of the rather conservative assumption that bracing members can take no

, compression, we conclude that lateral support for the torus can withstand lateral loading from a 0.22 g SSE.

l l

90

L

-5.9.6 Condensate Storage Tank The integrity of the condensate storage tank was evaluated for:

e Uplift and overturning.

e Anchor bolt forces.

, e Shell buckling at the base.

3 e Hoop stresses due to static and dynamic pressures.

e Roof integrity against sloshing.

The overturning moment (M B

) produced by lateral impulsive and convective water pressures and by lateral inertia loads of the tank itself was ,

assumed to be resisted by the 12 anchor bolts. Hydrodynamic pressures on' the tank bottom were assumed to pass directly to the soil below.

R.G.1.60 design spectral curves were used with a peak ground acceleration of 0.22 g when estimating earthquake accelerations. Calculated frequencies, damping,'and spectral acceleration for impulsive and convective s

water modes are:

Frequency . Damping Spectral Acceleretion (Hz)' (4 Critical) (g).

Impulsive mode 15.9 7 0.35 Convective mode 0.258 0.5 0.165

'3 Since the natural vertical frequency of the tank is high (much greater than 33 Hz), no amplification of ground vertical acceleration was assumed.

The reanalysis showed that tank shell stresses under a 0.22 g SSE are all well below .the yield strength (34 ksi) of the tank material. The smallest margin is in buckling. The tank shell buckling at the base was checked by using ASME Sect. III NE 3133 allowables for Class MC containment for the case of a cylinder under pure axial compression. An additional factor of 1.5 was applied to the allowable since SSE is considered to be a faulted condition..

The allowable stress was 2.1 kai versus the maximum combined compressive stress of 1.96 ksi.

91

The maximum water slosh height was estimated at 3.12 ft. With 1.5 ft of free board, the roof would be subjected to a n.*ximum noward pressure of 3.12 - 1.50 = 1.62 ft of water, or about 0.7 psi, not considering momentum.

No available drawings show structural details of the roof. However, if the roof is analyzed as a simply-supported,1/4-in.-thick circular plate, the induced bending stress would be of the order of 20 ksi. We feel an adequate margin exists even if momentum effects were included.

Analysis shows that the tank will not overturn under a 0.7.4 g SSE.

However, to withstand the maximum imposed earthquake overturning moment of 21 558 k-f t, the anchor bolts would need to take a maximum load of 157 kips, well beyond their capacity. As a result, some local uplif ting of the flexible tank base could occur as anchor bolts yield, and a portion of the tank bottom could lose soil support. As a result, plate bending stresses would increase, particularly at welds joining the sidewalls to the bottom plate. This potential source of local failure could lead to leakage of stored fluid. We recommend that attached piping be examined for potential overstressing at points of attachment.

The reanalysis found that the SSE induced a shear force of 1180 kips versus 968 kips calculated in the original analysis (Ref. 60) . As the design is such that the anchor bolts resist little of this type load, shear must be resisted by the friction force between the tank bottom and the supporting soil. A minimum average coefficient of friction of 0.27 (1180k/4367k) is needed between the bottom and the soil. Since the coefficient of friction between aluminum and soil should be at least this large, sliding is not expected to occur.

l l

5.10 SEISMIC INPUT MCTION FOR EQUIPMENT AND PIPING Seismic input motion for piping and equipment is typically defined by means of in-structure (or floor) response spectra for items vith relatively small mass. Before being used for design, floor response spectra are normally smoothed and peaks broadened to account for modeling and material uncertainties.

No floor response spectra were used in the original design. For reanalysis of the equipment and piping, horizontal and vertical floor response t

spectra were generated f.or both the reactor and the turbine buildings.

92 l

l

Spectra were generated for 34, 54, and 74 equipment damping (E.D.) for the two buildings, and an additional 24 E.D. curve was generated ~ for small size, Class I piping in the reactor building. For each building, the final spectra selected to analyze piping and equipment enveloped those resulting from analyses using three soil shear modulus values as shown below:

Value of soil modulus G Turbine building Reactor buildit.g 1500 ksf. Dashpot study Case R1 Fork River G Case T3 Case R2 ForkRiverG,g2 Case T4 Case R3 Spectra for a representative elevation in the turbine building are shown in Fig. 24, along with a smoothed, broadened envelope used for subsequent equipment and piping analyses. Spectra for an elevation in the reactor building, and the associated envelope, are shown in Fig. 25.

The final broadened spectra incorporate effects of 3-D input, torsional response, and the many SSI parameter studies reported in Appendix A. Since the soil response dominated, the variations in SSI parameters affected the floor response spectra. The following effects on floor response spectra were observed:

e The spectra were essentially the same for E-W and N-S directions.

Where a difference was found, the higher spectrum was enveloped.

e The adjustments made as a result of the dashpot analyses resulted in broadened and ira some cases higher spectra.

e Torsional response was not large; however, the final spectra were increased slightly to account for the increased response due to torsion of equipment located away from the center of rigidity of the floor.

e The predominant damping effect was not from the structure but from the soil. Relatively minor variations in structural damping values had no significant effect on the resultant spectra.

Since the resultant spectra represent smooth, broad envelopes from widely varying soil cases, peaks were not further broadened to account for structural uncertainties. All in-structure spectra for subsequent equipment and piping analyses are shown in Appendix B.

m 93

i i i i i ii i i .

i; i il, Case T4 _

3

/

Envelope

_f o1 -

Broadened h _

/f

'S envelope ~

D -

8 _ / -

f" ~

_----- - ~[ Case T3 0.1 ' ' ' ' ' 'I e i i i i e i il 0.01 01 _

1.0 Period (s)

FIG. 24. Turbine building floor response spectrum curves at operating floor, El. 46'6".

i 6 i i i i i i i i ig e i ii il

_r Envelope Case R2 [

Case R1 g Case R3 ej Broadened /.!.

f

\h

,d/

9  : -

g enveIopeg \*

t

\ /I N -

\

s

< _--------=----.p-

\g% :_

1_

l 0.1

' ' ' l i ' ' i i i ii1 0.01 01 1.0 Period (s)

FIG. 25. Reactor building floor response spectrum curves, with 3% damping, at the top floor (El. 119'3") .

94

CHAP'IER 6: SEISMIC EVALUATION OF MECHANICAL AND ELECTRICAL EQUIPMENT AND FLUID AND ELECTRICAL DISTRIBUTION SYSTEMS

6.1 INTRODUCTION

6.

1.1 Purpose and Scope

In this chapter, selected seismic-evaluation data developed to qualify certain mechanical and electrical equipment will be reviewed along with fluid and electrical distribution systems of the Oyster Creek Nuclear Power Plant.

Based on that review, the ability of the reactor to safely shut down and remain in a safe shutdown condition in the event of an SSE will be evaluated.

The SEP seismic review team purposely identified components expected to have high seismic fragility; moreover, the review team believes that the identified conponents represent not only those installed in the safe shutdown systems but those in other seismic Category I systems, such as engineered safeguards.

Thus, evaluation of those components establishes an estimated lower-bound seismic capability for the mechanical and electrical components and the distribution systems of the Oyster Creek Plant.

Considered in ternu of seismic design adequacy, nuclear power plant equipment and distribution systems fall into two main categories and two subcategories. Main categories are active and passive, and the two subcategories, which appear under both the active and passive designations, tre " rigid" and " flexible."

As discussed in R.G. 1.48 (Ref. 64) and " Standard Review Plan,"

Sa 3.9.3 (Ref. 65), active components are those that must perform a rechanical motion to accomplish a system safety function. For the purpose of this report, this definition is expanded to include electrical or mechanical components that are recuired for cafe shutdown and which must move during or af ter a seismic event in order to perform their design safety function.

Typically found in the active category are:

o Pumps.

e Valves.

e Motors and ass riated raotor-control centers.

e Switchgear.

95

Seismic design adequacy of active components should be shown by demonstration of safety function as well as by structural integrity. Adequacy may be determined by either analysis or by physical testing, but testing is generally preferred. However, because of size or weight restrictions or difficulty in monitoring function, many active components are seismically evaluated by analysis. In ensuring active component function through analysis, deformations must be limited and predictable. Therefore, total stresses in such components are normally limited to the linear elastic range of 0.5 to 0.9 times the yield stress of the material. Typically, the higher allowable stress limits are used with components constructed to meet what are generally considered to be the more rigorous requirements of the ASME Boiler and Pressure Vessel Code (BPVC), Sec. III (Ref. 66). The higher stress limits j

also tend to be used with austenitic steel materials. Other manufacturing or construction codes and standards usually have less rigorous fabrication, inspection, and test requirements than ASME, BPVC-III. Hence, components constructed or manufactured to other codes and standards tend to be assessed by using lower allowable stresses.

Components determined to be passive in this report are those components required for safe shutdown for which the only safety functions are maintenance of leak-tight pressure boundaries or structural integrity during and following the SSE. Typically found in the passive category are:

o Pressure vessels.

o Heat exchangers.

e Tanks.

l e Piping and other fluid distribution systems.

e Transformers.

l e Electrical-distribution systems.

In determining seismic design adequacy by analysis, the most important distinction between active and passive components is the stress level that the component is allowed to reach in response to the SSE excitation. For passive components, higher total stress limits, which range from 1.0 times yield to 0.7 times ultimate strength of the material, are permitted (Ref. 66) . As in the case of active components, higher stress limits are used foe components constructed or manufactured in accordance with ASME, BPVC-III. uower stress limits are used for components constructed or manufactured to other codes or I standards.

96

In selecting the magnitude of seismic input for component evaluation, it is important to determine whether a component or distribution system is rigid or flexible. Rigid systems experience the zero period floor acceleration while flexible systems will experience larger acceleration than the floor experiences. Seismic acceleration of equipment depends upon:

o Potential resonance with the supporting building structure.

e Structure and equipment damping levels, o Equipment support locations.

The desianation of " rigid" or " flexible" may also depend on how a particular component is supported. Many rigid components must be evaluated as though flexible because of the flexibility of their support.

A review of the Oyster Creek turbine and reactor building floor response spectra, shown in Figs. B-1 through B-3, shows that equipment contained in the turbine and reactor buildings may be considered rigid for frequencies greater than 20 Hz. For flexible components with fundamental frequencies less than 20 Hz, the maximum acceleration for 3% damping is approximately 15 times the SSE value of 0.22 g.

Components were first grouped as active or passive ~, and rigid or flexible; then a representative sample of each group was evaluated to establish the factor of safety or degree of adequacy of that group's seismic design. In this way, factors of safety within groups of similar components were established without detailed reevaluation of hundreds of individual conponents within each group.

A representative sample of components was selected for review by one of two methods:

e Selection was made from a walk-through inspection of the Oyster Creek facility by the review team. Based on their experience, team members selected components that appeared to have hioa seismic fragility for each component's category. Particular attention was paid to each component's support.

e Safe-shutdown components were categorized into three generic groups:

horizontally oriented tanks, heat exchangers, and pumps, vertically oriented tanks, heat exchangers, and pumps; motors and motor control centers.

The licensee was asked to provide seismic qualification data on selected components f rom each generic group.

97

The rest of this chapter reviews the seismic capacity of the selected components and recomunends, if necessary, additional analysis or hardware changes needed to qualify them for the SSE defined in this report. Based or. a -

detailed review of the seismic design adequacy of representative components, conclusions are presented as to the overall seismic design adequacy of seismic Category I equipment installed in Oyster Creek. Section 6.4 and Table 18 sununarize these conclusions.

6.1.2 Description of Components Selected for Review Table 16 describes components that the review team selected following its plant walk-through, as well as components representative of the generic groups of safety related components. Table 16 also gives the basis for each selection.

The review in this chapter emphasizes what are normally listed as auxiliary conponents. Deficiencies in seismic design tend to be found in auxiliary equipment rather than in major nuclear components. Auxiliary conponents are typically supplied by manufacturets who--unlike the nuclear steam-supply system vendors-may not have routinely designed and fabricated conponents for the nuclear power industry, particularly during the time this plant was under construction. Because of its importance to safety, the seismic design adequacy of the reactor coolant system components and support structures are also evaluated, to the extent that information was provided.

6.2 SEISMIC INPUT AND ANALYTICAL PROCEDURES 6.2.1 Original Seismic Criteria The governing design and construction codes used for Class I piping and equipment are summarized in Table 17.

f f

98 wn._ ,-7mp g -.w w - - - -

l MBLE 16. Mechanical and electrical components selected by the review team for seismic evaluation, and the basis for selection.

Item No. Description Reason for selection

1. Emergency service water pump This item has a long, vertical unsup-ported intake section that was originally statically analyzed for seismic effects.

2. Emergency isolation condenser This item is a horizontally mounted component supported by three saddles that do not appear to be seismically restrained. Concern was expressed about the saddles' ability to carry required seismic loads, particularly in the longitudinal direction.

3. Containment spray heat This item is unique in that the heat exchanger exchanger is vertically oriented and supported by four brackets. Concern was expressed about the exchanger's ability to withstand overturn'.ng eff ects.

4. Recirculation pump support This item is a vertical component sup-ported by hangers and critical to en-suring reactor coolant system integrity.

5. Emergency diesel oil Anchor bolt system for in-structure storage tank flat-bottom tanks that are flexible may be overstressed if tank and fluid contents were assumed rigid in the original analysis.

6. Motor operated valves A general concern with respect to motor operated valves, particularly for lines 4 in. or less in diameter, is that the relatively large eccentric mass of the motor will cause excessive stresses in piping attached to valves not externally supported.

7. CRD hydraulic control system Item is particularly critical to including tubing and support insuring reactor coolant system system integrity.

8. Reactor vessel supports and Same as Item 7.

internals 99

TABLE 16. (Continued. )

Item No. Description Reason for selection

9. Battery racks The bracing required to develop j lateral load capacity may not be sufficient to carry the seismic load.

10. Instrument racks The racks consist of channel and angle members that may be overstressed due to seismic loads. Anchorage to floor ,

may not be adequate.

. 11. Motor control centers Typical seismic qualified electrical l equipment. Functional. design ade-f quacy may not have been demonstrated.

In addition, anchorage to floor structure may not be adequate.

12. Transformers Same as Item 12.

13. Switchgear panels Same as Item 12.

14. Emergency generator Adequacy of anchorage was questionable. Functionality is important for safe shutdown.

i 15. Control room electrical The control panels appear adequately panels anchored at the base. However, there appear to be many components canti-levered off the front panel, and the lack of front panel st,1ffness may permit significant seismic response of the panel, resulting in high accel-eration of the attached components.

16. Battery room distribution Same as Item 15.

panels

{ 17. Isolation phase ductwork The ductwork support system does not supports appear to have positive lateral j restraint and load carrying capacity.

18. Electrical cable raceways The cable tray support system does 1

not appear to have positive lateral l restraint and load carrying capacity.

100

+

'IRBLE 17. Design and construction codes for Class I piping and equipment.

Component Design code Drywell, vents, and suppression pool ASME Sec. VIII Code Case 1272N-5. See Code Case 1276N-1 for expansion joint Reactor pressure vessel ASME Sec. I, plus Nuclear Code Case 1270 Recirculation loop, piping, ASME Secs. I and VIII, plus G.E.

and valves Specification Recirculation pump cases ASME Sec. VIII and Code Case 1274 Primary steam piping ASME Sec. I, through the first valve outside the reactor vessel. Balance:

ASA B31.1 Primary steam isolation valves ASA B 31.1, plus G.E. Specification Primary steam safety valves ASME Sec. I and Code Case 1271 N Nuclear steam supply auxiliary ASME Sec. I, through the first system piping and valves valve outside the reactor vessel.

Balance ASA B 31.1 Regenerative heat exchanger ASME Sec. III, Class C

'IEMA Standard Class R Nonregenerative heat exchanger Primary side ASME Sec. III, Class C Cooling water side ASME Sec. VIII

'IEMA Standards Class R Cleanup system vessels and ASME Sec. III, Class'2 demineralizers Isolation condenser Primary side ASME Sec. III, Class A Cooling water side ASME Sec. VIII Liquid poison tank API Standards Liquid poison pump ASME Sec. III, Class C Shutdown heat exchanger Primary side (tube) ASME Sec. III, Class C Cooling water side (shell) ASME Sec. VIII Shutdown pump ASME Sec. III, Class C 101

'mBLE 17. (Continued.)

Component Design code Containment spray cooling system ASME Sec. VIII equipment Filters (except those in the cleanup ASME Sec. VIII system) ,

Feedwater heaters (including ASME Sec. VIII, plus TEMA Standards drain coolers)

Main condenser Heat Exchanger Institute Turbine moisture separator ASME Sec. VIII Turbine steam reheaters ASME Sec. VIII Condensate demineralizers ASME Sec. VIII Control rod drive Pressure parts ASME Sec. VIII with deviations for weld joints design covered in Code Case 1361 (Sec. III)

Control rod drive housings ASME Sec. I In-core ion chamber pressure parts ASME Sec. III, Class A Gland seal exhauster condenser Heat Exchanger Institute Emergency core cooling system ASME Sec. I, through the first piping and valves valve outside the reactor vessel.

Balance ASA B 31.1 Steam jet air ejector and inter- Heat Exchanger Institute and after-condensers Scram dump piping and valves ASA B 31.1 plus APED Specifications through the first valve outside the reactor vessel. Balance ASA B 31.1 Control rod drive system Pump casing and accumulators ASMP Sec. VIII Piping and valves ASME Sec. from control rod drive to first valve. Balance ASA B 31.1 102

l The allowable stresses for Class I piping are given below:

Loading condition Allowable stress

1. Thermal Expansion S A-

2. M.O.L. + S.L. S

3. M.O.L. + 2 x S.L. (Stress such that safe shutdown can be achieved) where M.O.L. = Maximum Operating Loads including design pressure and temperature, weight of piping and contents including insulation, and the effect of supporto and other sustained external loadings, S . L. =

Seismic Loads due to the design earthquake (OBE),

2 x S.L. = Seismic loads due to twice the design earthquake (SSE),

S =

3 f(1.25 SC + 0.25 Sh )'

l and:

f = stress range reduction f actor for cyclic conditions,

[

S C

= all wable stress in cold condition per ASA B31.1,

S =

allowable stress in the hot condition (design temperature) per ASA B31.1.

For the reactor vessel supports, the allowable stresses are given as:

1. OBE: normal AISC allowable stresses.

2. OBE + Jet: 150% of normal AISC al2 owable stresses.

3. 2 x OEE: 150% of normal AISC allowable stresses.

The criteria used for instrumentation is quoted from the answer to Question N.1, Amendment 11, of the FDSAR and is given as: "The control room panels and auxiliary racks are usually shipped assembled and therefore these units must be designed for normal shipping shock which is in the order of several g's acceleration. Certain components are removed and padded to reduce vibration eff ect and excessive acceleration. In all cases, however, the design analysis is made of the panels and instruments. All relays in safety circuits are energized; and since they are capable of closing against 1.0 g, they can certainly maintain contact during an acceleration of 0.22 g."

103

6.2.2 Seismic Criteria for Reevaluation Current seismic input requirements for determining the seismic design adequacy of mechanical and electrical equipment and of distribution systems are normally based on in-structure or equipment response spectra for the various elevations at which the equipment is supported. The in-structure spectra, which are based on R.G.1.60 spectra modified by the dynamic characteristics of the buildings, are shown in Figs. B-1 through B-3. The in-structure spectra are based on the building model shown in Figs.17 and 18.

For evaluating mechanical and electrical equipment, 7% damping is used for the 0.22 g SSE. For piping evaluation, the damping associated with the SSE is lbnited to 34. These values are consistent with a recent summary of data directed toward defining damping as a function of stress level (Ref.

67). For cable trays, recent tests seem to indicate that damping levels depend greatly on the tray and support construction and on the manner in which cables are placed in the trays. Damping may be as high as 20% of critical damping (Ref. 68). A dampirg factor of 0.5% was used for analyzing the sloshing mode of fluids contained in tanks. Horizontal seismic loads are assumed to act simultaneously. Depending on the geometry of the equipment being evaluated, the resultant horizontal load will be from 1.0 to 1.4 times the individual directional component. We have applied the conservative 1.4 factor in this evalaation except where design adequacy is in question.

6.2.3 SEP Behavior Critt.ria Seismic Category I components designed to remain leak-tight or retain structural integrity in the event of an SSE are now typically designed to ASME, Sec. III Code, Class 1, 2 or 3 stress limits for Service Condition D. Stresses in supports for ASME leak-tight components have limits shown in Appendix F or Appendix XVII of the ASME, Sec. III Code (Ref. 66) .

If qualification is to be by analysis, active ASME, Sec. III components that must perform a mechanical motion to accomplish safety functions must typically meet ASME Sec. III Code, Class 1, 2 or 3 stress limits for Service Condition B. Supports for these components are also typically restricted to Service Condition B If.mits.

104

For other passive and active equipment not designed to ASME Sec. III Code requirements, and for which the design, material, fabrication, and examination requirements are typically less rigorous than ASME Sec. III Code requirements, allowable stresses for passive components are limited to yiekd values. For active components, allowable stresses are limited to normal working stresses (typically 0.5 to 0.67 of yield). SEP behavior criteria used in reevaluation of various equipment and distribution systems for Oyster Creek passive cor:ponents are given in Table 18. For electrical components such as switches, relays, etc., functional adequacy should be demonstrateo by test. '

Experience in designing such pressure retaining components as vessels, pumps, and valves to ASME Sec. III Code requirements for 0.22 g indicates that stresses induced by earthquakes seldom exceed 10% of the dead weight and pressure-induced stresses in a component (Ref. 69) . Therefore, design adequacy of such equipment is seldom dictated by seismic design.

Seismically induced stresses in nonpressurized mechanical and electrical equipment, in fluid and electrical distribution systems, and in component supports may be significant in determining design adequacy. Note that SSE loadings seldom control design of piping systems. Because of more restrictive stress and damping limits, the OBE normally controls design of piping.

6.3 EVALUATION OF SELECTED COMPONENTS FOR dBISMIC DESIGN ADEQUACY 6.3.1 Mechanical Eauipment 6.3.1.1 Emergency Service Water Pump The emergency service water pump and motor unit is oriented vertically in the Intake Structure. As shown on Byron Jackson, Inc., Drawing SK-651-N-0746, the intake portion of the pump extends downward from the discharge head and pump base for a distance of 13 ft 5 in. The seismic analysis, as given by Burns and Roe, Inc., in Ref. 70, was performed for an equivalent static load of 0.22 g acting in the horizontal direction.

The pump-motor unit is located at grade; therefore, the seismic input is essentially the R.G. 1.60 ground response spectrum normalized to 0.22 g.

Tensile and shear stresses in the pump base anchor bolts due to overturning forces were determined, as well as stresses at the attachment of the intake column pipe to the discharge head.

105

TABLE 18. SEP structural behavior criteria for determining seismic design adequacy of passive mechanical and electrical equipment and of distribution systems.

SEP criteria Components (SSE)

Vessels, S, j 0.7 Su and 1.6 S y pumps, and S,, f 0.67 S uand 1.33 S y ASME III Class 1 (Table F 1322.2.1) valves ASME III C? ass 2 (NC 3217) om,77 5 0.5 Suand 1.25 Sy AS E III Class 2 W 3321) 0, 1 0.5 Suand 1.25 S y ASME III Class 3 (ND 3321)

Piping S and 2.0 Sy a m i 1.0 S u ASME III Class 1 (Table F 1322.2.1)

S h 5 0.6 Suand 1.5 S y ASME III Class 2 and Class 3 (NC 3611.2)

Tenks No ASME III Class 1 0,g 5 S u and 1.25 S y ASME III Class 2 and Class 3 (NC 3821) g Electric Sall- $ 1.0 Sy e equipment Cable trays Sall i 1.0 S y ASME supports S

all 1 1.2 S yand 0.7 S u ASNA III Appendices XVII, F for Class 1, 2 and 3 Other S ali i 1.6 S supports Normal AISC S allowable increased by 1.6 consistent with NRC Standard Review Plan, Sec. 3.8 Bolting Sall i 1.4 S ASME Sec. III Appendix XVII for bolting where S is the allowable stress for design loads

l

)

Because the intake portion of the pump is oriented vertically as a l cantilever beam, the dynamic characteristic of the intake suction pipe was determined. The intake suction pipe and shaft were found to have a  !

fundamental frequency of 7.57 Hz. For this natural frequency, the spectral l

ccceleration f rom the R.G.1.60 response spectrum, normalized to 0.22 g for 7%

damping, is 0.44 g. Seismic accelerations were applied to the pump, considering simultaneous N-S and E-W loading to determine anchor bolt stresses. Loads of the attached pf ping nozzle under normal operation were not available. However, the emergency service water line is a cold line and therefore would tend to transfer rmall pressure and thermal loads onto the pump.

The anchorage analysis established a safety factor based on ASME Condition D stress limits of 143 for A307 anchor bolts. The stress calculated at the attachment of the discharge head to the intake column pipe is 1.475 ksi, which is within acceptable limits. It is not clear what material was used for the pump head. We recommend that all cast iron components be changed to an acceptable material at the licensee's earliest opportunity.

There is insuff ?.cient detail to evaluate the functional adequacy of the pump for motor impeller shaf t deformities and for bearing or coupling failure.

6.3.1.2 Emergency (IsolaL'on) Condenser The emergency condenser, supplied by Foster Wheeler Corporation, is located in the Reactor Building at El. 95'3". It is 44 ft long, mounted horizontally, and supported by three saddles. The original seismic design, based on 0.15 g herizontal acceleration and 0.10 g vertical acceleration, was performed by Burns and Roe, Inc., and is given in Ref. 71.

The response spectra for 7% damping tshown on Figs. B-le and B-3a) at El. 95'3" of the reactor building were used. When the component and its support system were assumed to De rigid, the resultant input horizontal and vertical seismic accelerations are 0.57 and 0.25 g, respectively.

Our evaluation assumed that only the center support would carry longitudinal shearing stress, because the bolt holes in the outer two support saddles are slotted to provide for thermal expansion. Since the center support is assumed to take the total longitudinal shear load plus one-third the transverse shear load, the shear stress in each of the four 1-in. A-307 107

support bolts indicated in Ref. 72 is 47.7 ksi for combined N-S and E-W earthquake loading. Since this stress exceeds the allowable ASME Service Conditions D shear stress of 17.4 kai, we believe that the anchorage system for the emergency condenser is inadequate to withstand the 0.22 g SSE seismic loading. The middle saddle needs to be modified so that the total shear area av.ilable is increased by a factor of 47.7/17.4 = 2.74. Therefore a minumum area of 8.61 in.2 is required instead of the 3.14 in. provided.

6.3.1.3 Containment Spray Heat Exchanger The containment spray heat exchanger, located in the Reactor Building at El. 23 '6", was supplied by McQuay-Perfex, Inc. It is a vertical component 23 ft 2 in. long and is supported by four lugs 100 in. from its top. The original seismic design (see Ref. 73) used a 0.24 g horizontal acceleration and a 0.146 g vertical acceleration.

The responsa spectra for 7% damping (Figs. B-le and B-2b) generated at EL. 23'6" of the react 7r building were used. If the heat exchanger is assumed to be rigid, the input horizontal and vertical seismic accelerations are 0.38 and 0.25 g, respectively. For load combinations that include seismic loading, the resultant anchor bolt stresses for the 1-in.-diameter A-325 bolts exceed

the ASME Condition D stress limits. For the most critical case, the combined shear stress for N-S and E-W loading is 32.6 ksi compared with an allowable shear stress of 30.4 kai. Therefore, we believe that the anchorage system for the containment spray heat exchanger should be reanalyzed in detail to determine its ability to withstand the 0.22 g SSE seismic loading.

6.3.1.4 Recirculation Pump Support No evaluation has been made since no design calculations or specifications are currently available.

j t

i 6.3.1.5 Emergency Diesel Oil Storage Tank l

The emergency diesel oil storage tank is a cylindrical vessel 14 f t 6 in.

tall and 13 f t 2 in. in diameter. The tank, which has a wall thickness of 1/4 in., is restrained by a ring anchored to the concrete floor by eight 3/4-in.

diameter bolts. The tank was originally designed by Burns and Roe, Inc.,

according to Ref. 74, for an assumed ground acceleration of 0.22 g (Ref. 75) .

108

The tank, which is supported at ground elevation of the Emergency Diesel Generator Building, was reevaluated as shown in Ref. 76 for R.G. 1.60 response spectra normalized to 0.22 g. The dynamic rialysis considered the effective ccnvection and impulsive response of contained fluid. It also determined fun &snental response frequencies of 0.48 Rs for the tank under convective loading (0.5% damping) and 48.7 Rs for the tank bending and shear deformation under impulsive loading (tank considered full) . Therefore, the tank can be considered rigid for the impulsive moment effect. .

The analysis determined gross dynamic characteristics of the tank. The evaluation showed that the oil storage tank will not slide or overturn even.

without anchor bolts. If friction were to be overcome, the resul' ting anchor bolt safety factor in shear would be 2.31, using ASME Condition D stress limits. The safety factor is 38.6 for compressive stress in the tank wall due to combined seismic overturning and deadweight stresses.

Therefore, we believe that the emergency diesel oil storage tank will withstand the 0.22 g SSE loading without loss of structural integrity.

6.3.1.6 Motor Operated valves We have reviewed the method used and conclusions reached by both the Burns and Roe calculation for valves 6 in. and larger given in Ref. 77 cnd by the MPR Associates calculation in Ref. 70. The conclusions reached in the references are based on the original seismic acceleration levels of 0.43 g.

We considered a 0.43 g level of seismic excitation for a 0.22 g SSE to be several times analler than typically would be determined if the piping systems were evaluated using in-structure response spectra. We question whether all the valve operators are installed vertically; if they are not, additional stresses would be induced in the pipe by the eccentric dead weight. We also note that there were some conservatisms introduced by assuming a simply supported span and that all the eccentric moment is carried by the pipe on one side of the valve.

Therefore, it is. recommended that the licensee, in reevaluating sample pipe runs as part of the SEP program, include at least two motor-operated valves; one larger than 4 in. and one smaller than 4 in. Resultant stresses should then be compared to those determined by the methodology shown in Refs.

77 and 78. In this way, the conclusions reached in Refs. 77 and 78 could be evaluated quantitatively.

109

e In lieu of a generic certification by analysis, we recommenf that a requirement to provide external support of valve operators be developed and implemented.

6.3.1.7 CRD Hydraulic Control System Including Tubing and Support System We reviewed the generic seismic analysis of the hydraulic control unit document No. DAR 149, dated November 1972, prepared by General Electric Co.

(Ref. 79). To verify that this generic document is applicable to Oyster Creek, the framing, mass, stiffness, and anchorage characteristics assumed in the document must be shown to be similar to those of the system actually installed at Oyster Creek.

In Table IV-I of Ref. 79, the limiting seismic capacity of the f reestanding structure is 1.27 g at 2.27 Hz, based on an allowable stress of 1.5 c . .The SEP structural behavior criterion shown in Table 18 allows only-y 1.6 S (where S is the normal AISC allowable stress, equal to 0.66 0 ), or an ,

allowable stress of 1.6 x 0.660 = 1.056 c . This reduces the capacity in y y the ratio of 1.056/1.5, giving a capacity of 0.89 g at 2.27 Hz.

The in-structure response acceleration from Fig. B-2b at 7% damping is 1.3 g at 2.27 Hz, greater then the 0.89 capacity. Therefore, modification of a freestanding type support would be required.

In addition, it is not clear whether one or two horizontal earthquake components were originally considered. If only one component was considered, the.1 resultant stresses should be increased by a factor of 1.4, giving a reduced capacity of 0.89/1.4 or 0.64 g.

We feel that if a bcan support arrangement similar to that used in Fukushima I were to be installed at Oyster Creek, the CRD system supports would be adequate.

6.3.1.8 Reactor Vessel, Supports, and Internals Results of the original analysis shown in Ref. 80 indicate that the fundamental frequency of the reactor vessel is 7.75 Hz. We assumed the corresponding design acceleration of 0.225 g to be an OBE level acceleration.

Twice this value, or 0.45 g, would have been applicable to an SSE evaluation.

110

The reanalysis value is 0.63 g for 7% damping, as shown in Fig. B-2c. Thus, seismic loads have increased by a factor of 1.4 without considering the effect of two horizontal and one vertical component.

We have insufficient information to further determine des!gn adequacy of the reactor vessel, supports, and internals.

6.3.2 Electrical Equipment 6.3.2.1 Battery Racks The battery racks used for the Oyster Creek Plant were manufactured by Gould-National Batteries, Inc., and the design calculations are given in Ref.

81. The stationary battery cells have been tested by Gould, Inc., Industrial Battery Division, according to the requirements of IEEE 323-1974, IEEE 344-1975, and IEEE Standard 535, Praft 13, and the test procedure given in Ref. 82. The response spectra for the battery racks, which correspond to the mezzanine floor of the Turbine Building, are given in Figs. B-lb and B-ld.

The original seismic design performed by Gould, Inc., is based upon peak accelerations of 3.0 g horizontally and 2.0 g vertically. All component parts, including anchor bolts, were analyzed to ensure that the stresses generated are less than allowable stresses. The analysis indicates that the battery rack and anchor bolts are adequate to withstand the specified seismic loading. For the SEP response spectra, the peak accelerations are less than the original design values. We believe that, contingent upon completion of current modifications, the battery racks will withstand the 0.22 g SSE seismic loading without loss of structural integrity.

6.3.2.2 Instrument Racks The instrument racks for the Oyster Creek Plant are frameworks 6'6" high constructed of channel and angle members. The racks were originally evaluated for seismic loads and are shown in Ref. 83. For rack RK01, which is located in the Reactor Bu".lding at El. 51'3", the fundamental frequency has been calculated as 25.0 Hz in the direction perpendicular to the rack frames and 100 Hz in the direction parallel to the frames. The original spectral accelerations were 0.135 g in the horizontal direction and 0.07 g in the 111

i 1

vertical direction. For rack RE05, which is located in the Reactor Building at El. 75'3", the fundamental frequency has been calculated as 16.5 Hz in the direction perpendicular to the rack frames and greater than 100 Hz in the direction parallel to the frames. The corresponding spectral accelerations were 0.15 g in the horizontal direction and 0.07 g in the vertical direction.

From the SEP response spectra (Figs. B-le and B-2d), the spectral accelerations corresponding to the fundamental period of rack RK01 are 0.43 g in the horizontal direction and 0.25 g in the vertical direction. Results obtained in the original analysis indicate that th'e seismically induced stresses in the members and anchor bolts are far less than the allowable stresses. If these seismically induced stresses are multiplied by the ratio of SEP seismic accelerations to the original seismic accelerations, the corresponding member stresses are still less than the allowable stresses.

Therefore, we believe that the instrument racks will withstand the 0.22 g SSE seismic loading without loss of structural integrity. However, we have no information on which to base an evaluation of the functional behavior of the instrumentation supplied and installed in the rack.

6.3.2.3 Motor Control Centers and Switchgear No evaluation has been performed since no drawings or design calculations are currently available.

6.3.2.4 Transformers No evaluation has been performed since no drawings or design calculations are currently available.

6.3.2.5 Emergency Generator and Switchgear Panels The emergency diesel generator and switchgear, which are located in the Diesel Generator Building, have been analyzed for a 0.22 g SSE seismic load,

! as given in Ref. 84. Since the equipment is located at grade, the R.G.1.60 response spectrum was used. For the diesel generator, which is considered rigid, the acceleration levels ne' assary to cause overturning and sliding cre 0.44 and 0.25 g, respectively (Rat. 84). Since these acceleration values are 112

greater than 0.22 g, we believe that the emergency diesel generator will remain stable under SSE seismie loeding.

It is our opinion that, whenever feasible, equipment should be positively anchored to resist seismic effects. We would expect the diesel generator to be positively anchored to the floor to resist starting torque and vibration eff ects. If such restraint exists, it should be considered in the analysis.

If not, we recommend that the licensee provide positive anchorage.

For the switchgear, the acceleration levels necessary to cause overturning and sliding are 0.37 and 0.25 g, respectively. However, since the switchgear cannot be considered rigid, the corresponding spectral acceleration is 0.57 g, which is greater than 0.37 or 0.25 g. We believe that the switchgear should be positively anchored to resist overturning and sliding e ff ects.

6.3.2.6 Control Room Electrical Panels No evaluation has been performed since no drawings or design calculations are currently available.

6.3.2.7 Battery Room Distribution Panels No evaluation has been performed since no drawings or design calculations are currently available.

6.3.2.8 Isolated Phase Bus Duct Supports The isolated phase bus duct supports, which are located in the turbine building at El. 23'6", were analyzed as shown in Ref. 85. Evaluation of the duct support system was made for an SSE acceleration of 0.5 g. The value of 0.5 g includes an amplification factor of at least two to account for support flexibility and possible adverse effects from higher modes of excitation. The seismic evaluatior. Indicates that additional bracing should be added to the duct supports. For 'he SEP response spectra, the spectral acceleration corresponding to the fundamental f requency of the duct support system is 0.17 g (7% damping) at a fundamental frequency of 0.6 Hz. The bending stress in the support member due to a seismic load of 2.0 x 0.17 g is 92.5 ksi, which 113

is higher than the 26.5 kai allowable value. Therefore, the original conclusion that additional bracing should be added to the duct supports is I still valid. We also note such bracing will tend to increase the fundamental frequency of the system, hence increasing the response acceleration. 'Ihis change in dynamic characteristics should be included in any support redesign.

6.3.2.9 Electrical Cable Raceways No evaluation has been performed since no drawings or design calculations are currently available.

6.4 StD9tARY AND CDNCUCIONS Tablo 19 sunniariz es our findings on the sample of mechanical and electrical cogonents and of distribution systems that were evaluated to determine the seismic design adequacy of such items required for the safe shutdown of the Oyster Creek nuclear steam supply system. As discussed in Sec. 6.1, the sample includes components the review team selected, based on judgment and experience, as representative of lower-bound seismic design capacity of Oyster Creek, as well as the grouping of components into representative categories.

Based upon the design review and independent calculations for the SEP seismic load condition, m recomumend that design modifications or reanalysis may be required for particular mechanical and electrical components to withstand the 0.22 g SSE without loss of structural integrity, as required for maintaining safety functions. In general, no information that has been ,

provided demonstrates the functional adequacy of mechanical and electrical equipment evaluated on the Oyster Creek Plant. Based on design data we have evaluated, the particular mechanical and electrical cogonents that require additional evaluation and possible design modification are as follows:

1. Emergency isolation condenser. 8. Motor control centers.

2. Containment spray heat exchanger. 9. Transformers.

3. Recirculation pu g. 10. Switchgear panels.

4. Motor operated valves. 11. Control room electrical panels.

5. CRD hydraulic control units. 12. Battery room distribution panels.

6. Reactor vessel, internals, 13. Isolated phase bus duct supports.

and supports. 14. Electrical cable raceways.

7. Emergency generator.

114

TABLE 19. Conclusions regarding equipment review for seismic design adequacy of Oyster Creek.

Item Description Conclusion and recommendation

1. Emergency service O.K. for structural integrity. Functional water pump integrity has not been evaluated due to lack of design detail. We recommend the replacement of any cast iron components used.

f

2. Emergency isolation The anchor bolts of the center saddle appear condenser overstressed in shear. Lacking a more detailed analysis that demonstrates design adequacy, additional longitudinal and lateral restraint should be provided.

3. Containment spray The anchor bolts appear overst.vssed. Lacking heat exchanger a more detailed analysis which demonstrates design adequacy, additional lateral restraints should be provided.

4. Recirculation pump No evaluation has been performed since no support design calculations or specifications are currently available.

5. Emergency diesel oil O.K.

storage tank

6. Motor operated valves Evaluation of design adequacy given in Refs.

77 and 78 assumes unrealistically low seismic accelerations. We suggest that at least two motor operated valves, one greater than 4 in, in diameter and one less, be included in the detailed reevaluation of piping for Oyster Creek and that results be compared to those esacained in Refs. 77 and 78. No information has been supplied concerning functional adequacy of motor control valves.

7. CRD hydraulic control If the Support system is of the freestanding units type, it would appear to be overstressed.

System actually installed in Oyster Creek should be compared to the systems analyzed in Ref. 79 to determine if the Ref. 79 conclusions are valid. If not, reanalysis should be performed.

8. Reactor vessel, Seismic input loads appear to be at least supports, and 40% larger than those considered in the internals original design. No detailed design calcula-tions are available to evaluate design adequacy.

115

E TRBLE 19. (Continued. )

Item Description- Conclusion and recommendaclon

9. Battery racks 0.K. (Contingent upon completion of current modifications.)

10. Instrument racks O.K. for structural integrity. No information on function.

11. Motor control No structural integrity evaluation has been centers performed since no drawings or design calcu-i lations are currently available, nor has functionality been demonstrated.

12. Transformers No evaluation has been performed since no drawings or structural integrity design calculations are currently available.

13. Switchgear panels Switchgear panels should be positively anchored to resist seismic-induced overturning and sliding effects.

14. Emergency generator O.K. for structural integrity. Functionality has not been demonstrated. Evaluate anchorage.

15. Control room No structural integrity evaluation has been electrical panels performed since no drawings or design calcu-lations are currently available, nor has functionality been demonstrated.

16. Battery room No evaluation has been performed since no distribution panels drawings or design calculations are currently available.

17. Isolation phase Lateral bracing should be added to the duct duetwork supporta supports.

18. Electrical cable No evaluation has been made since no drawing raceways or design calculations are currently avail-able. However, it is recommended that lateral restraint be provided unless design adequacy is demonstrated, i

116

}

= . . - - -- - - . -

REFERENCES i

1. .N. M. Newmark, W. J. Hall, R. P. Kennedy,'J. E. Stevenson,'and F. J.

Tokarz, Seismic Review of Dresden Nuclear Power Station--Unit 2 for the Systematic Evaluation Program , U.S. Nuclear Regulatory Constission NUREG/tR-0891 (July 1979) .*

2. "Backfitting," U.S. Code of Federal Pequlations, Title 10, Part 50.109.

3. U.S. Nuclear Regulatory Commission, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Office of Nuclear Reactor Regulations, NUREG-75-087, IMR edition (1975) .**

4. N. M. Newmark and W. J. Hall, Development of Criteria for Seit Jic Review of Selected Nuclear Power Plants, NUREG/tR-0098 (1978) . **

5. T. A. Nelson, Seismic Analysis Methods for the Systematic Evaluation Program, Lawrence Livermore National Laboratory, Livermore, Calif.,

UCRL-52528 - (1978) .

6. U.S. Nuclear Regulatory Comatission, Design Response Spectra for Seismic Design of Nuclear Power Plants, Regulatory Guide 1.60.

7. Jersey Central Power and' Light Co., Facility Description and Safety Analysis Report, vols. I, II, and NRC Docket items 50219-1 and 50219-2 (January 25, 1967).

8. Jersey Central Power and Light Co., Amendment 38 Additional Information in Response to Oral ABC Inquiry, also NRC Docket item 50219-46, prepared for Nuclear Regulatory Commission (July 8, 196G).

9. I. R. Finfrock, Jersey Central Power and Light Co., Summary of Seismic Design Information Oyster Creek Nuclear Generating Station, letter report prepared for D. L. Ziemann, Nuclear Regulatory Comatission (July 9,1979) .

10. M.'E. Nitsel, Summary of Oyster Creek Unit 1 Piping Calculations Performed for the Systematic Evaluation Program, BG&G, Idaho, Interin Report BGG-EA-5211 (July 1980).

11. t!. 4. Nuclear Regulatory Commission, Damping Values for Seismic Design of Nuclear Power Plants, Regulatory Guide 1.61-(1973).

12. G. W. Housner, California Institute of Technology, Recommended Earthquake Design Criteria-Jersey Central Nuclear Power Plant, R. B. Gile, General Electric Company (March 4,1964) .

13. G. W. Housner, and R. A. Williamson, Nuclear Reactors and Earthquakes, U.S. Atomic Energy Comatission, TID-7024 (1963).

117

14. J. E. Logan, Jersey Central Power and Light Co., Amendment 22 Additional Information Response to Category 1 Questions of October 16, 1967 Request, NRC Docket item 50219-11, prepared for P. A. Morris, Nuclear Regulatory Commission (October 16, 1967).

15. Jersey Central Power and Light Co., Facility Description and Safety Analysis Report, Vols. I and II, also NRC Docket items 50219-1 and 50219-2 (January 25, 1967).

16. John A. Blume & Associates, Jers*y Central Reactor Project Earthquake Analysis: Suppression Chamber Suction Header (November 22, 1967).

17. E. J. Keith, John A. Blume & Associates, Jersey Central Nuclear Power Plant Seismic Analysis of the Reactor Building, R. B. Gile, General Electric Company (June 18, 1965).

18. J. E. Logan, Jersey Central Power and Light, Co., Amendment 11 Answers to 109 ABC Questions Regarding Additional Plant Information, also NRC Docket item 50219-17, prepared for P. A. Morris, Nuclear Regulatory Commission (June 21,1967) .

19. R. L. Sharpe, John A. Blume & Associates, Jersey Central Reactor Proitet Earthquake Analysis of the Turbine Building, letter report prepared for R. B. Gile, General Electric Company (June 18, 1965).

20. E. J. Keith, John A. Blume & Associates, Report en the Earthquake Analysis of the Ventilation Stack for the Jersey Central Nuclear Reactor Project, letter report prepared for R. B. Gile, General Electric Company (May 6,1966) .

21. American Concrete Institute, Specification for the Design of Reinforced Concrete Chimneys, ACI-505 (circa 1966) .

22. Ralph M. Parsons Co., Primary Containment Design Report, also amendment 15 to application, NRC Docket item 50219-35 (September 1967) . -

23. D. R. Rees, Jersey Central Power and Light Co.. Amendment 32 Response to AEC Letter of January 9,1968,, also NRC Docket item 50219-43, prepared for A. Morris, Nuclear Regulatory Commission (February 20, 1968).

24. John A. Blume & Associates, Report on the Earthquake Analysis of the

[

l Drywell for the Jersey Central Nuclear Reactor Project, letter report i

prepared for General Electric Company (January - March 1965) .

[

i l

f l

i 118

i l'

25.- J. E. Logan, Jersey Central Power and Light Co., " Report on the Seismic

- Analysis of the Reactor Pressure Vessel for the Jersey Central Nuclear Power Plant", Exhibit E of Amendment 16 Reactor Pressure Vessel Design a

Report, also NRC Docket item 50219-36, prepared for the Nuclear Regulatory Commission (Septe:aber 15, 1967).

26. H. J. Sexton, John A. Blume & Associates, Jersey Central Suppression

. Chamber, letter report prepared for R. B. Gile, General Electric Company (April 15,1965) .

27. J. E. Logan, Jersey Central Power and Light Co., Amendment 28 Response to ABC Letters of October 16, 1967 and November 20, 1967, also NRC Docket item 50219-40, prepared for P. A. Morris, Nuclear Regulatory Commission.

28. Jersey Central Power and Light Co., Amendment 12 Reactor Vessel Internals Integrity Analysis, also NRC Docket item 50219-33, prepared for P. A.

Morris, Nuclear Regulatory Commission (August 30, 1967).

29. Burns and Roe, Inc., Piping Specification S-2299-60-C.

l. 30. I. R. Finfrock, Jr., Jersey Central Power and Light Co., Oyster Creek

( Nuclear Generating Station Docket No. 50-291 I. E. Bulletin No. 79-07,

~

{ letter reporc prepared for B. H. Grier, Nuclear Regulatory Commission (April 24,1977) .

~

31. John A. Blume & Associates, Report on the Seismic Analysis of the Recirculation Loops fer the Jersey Central Nuclear Power Plant (July 12, 1965).

32. J. R. Curreri, Dynamic Model Test Report of Piping Systems of The Jersey

- Central Nuclear Power Plant Oyster Creek Unit No.1, Brooklyn Polytechnical Institute (October 31, 1967).

I 33. John A. Blume & Associates, Earthquake Analysis Emergency Condenser Pipes (November 20, 1967).

34. D. G. Strawson, Oyster Creek Core Spray (Nor th) SeisIPic Analysis, MPR l Associates, Inc. (January 16, 1979).

35. J. R. Curreri, Report of the Dynamic Analysis of the Main Steam and Feedwater Piping of the Jersey Central Nuclear Power Plant, Brooklyn Polytechnical Institute (October 14, 1967).

36. John A. Blume & Associates, Earthquake Analysis: Emergency Service Water Lines (December 11, 1967).

37. John A. Blume & Associates, Zarthquake Analysis Buried Emergency Service Water Lines (November 20, 1967).

119

38. E. L. Wilson, SMIS Symbolic Matrix Interpretive System, Department of Civil Engineering, University of California, Berkeley (1973) .

39. The M. W. Kellogg Ccapany, Design of Piping Systems (John Wiley & Sons, Inc., 1955).

40. MPR Associates, Inc., High Energy Piping Systems Inside Containment -

Stress Summary (February 6, 1979).

41. PIPESD and PIPESD/ NEAT PIPE Static, Dynamic and Thermal Transient Analysis System, User Information Manual, Control Data Corporation.

42. N. M. Newmark, Letter report to the HRC (August 7,1968) .

43. NUREG-75/087, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition 1975, Office of Nuclear Reactor Regulation, NRC.**

44. Ad Hoc Group on SSI, Analysis for Soil-Structure Interaction Effects for Nuclear Power Plants, Nuclear Structures and Materials Committee of the Structural Division of ASCE (1974).

45. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.92.

46. J. J. Johnson, "A Modified version of the Structural Analysis Program SAP IV for the Static and Dynamic Response of Linear and Localized Nonlinear Structures," General Atomic Company (June 1978) .

47. Letter f rom A. S. Dam, Burns & Roe to Schmidt, MPR Assoc., dated September 11, 1979.

48. Bechtel Topical Report BC-TOP-4, Rev. 2. , Seismic Analysis of Structures and Equipment for Nuclear Power Plants.

49. Letter from C. H. Hofmayer NRC/ DOR /SEP to R. Murray/LLL, November 8, 1979. "Sunmary of Methodology and Parameters to be used in SEP Seismic Reviews for Palisades and Oyster Creek Nuclear Power Plants."

50. Letter from N. C. (Tom) Tsai to Bob Murray/LLL dated January ll 1980.

51. E. Kausel and R. Ushijhima, vertical and Torsional Stiffness of Cyli..drical Footings, MIT Department of Civil Engineering Research Report R 79-6 (February 1979).

52. M. Novak, " Vibrations of Embedded Footings and Structures," presented at ASCE National Structural Engineering Meeting, San Francisco, Calif.,

Preprint 2029 (April 1973).

53. E. Kausel, R. V. Whitman, J. P. Muray, and F. Elsabee, "The Spring Method for Embedded Foundation," Nuclear Engineering and Design 48, 337-392 (1978).

120

- . . . - . . . _ . - . .. - . ~ _ - _ .

i i

54. Blume, Newmark, and Corning, " Design of Multisto'ry Reinforced Concrete Buildings for Earthquake Motions," PCA Publications (1961).

l 55. R. E. Richart, Jr., J. R. Hall, and R. D. Woods, Vibrations of Soil and i

Foundations (Prentice Hall, Inc., 1970).

i

! 56. N. C. Tsai, Nichott, Swatta, and Hadjian r "The Use of Frequency {

Independent Soil-Structure Interaction Parameters," Nuclear Energy and Design H (2), 168-183 (1974). ,

1,.

57. Letter from Y. Nagai, JCP & L to T. Wambach/NRC, dated November 27, 1979. ,

58. John A. Blume & Associates, Seismic Analysis of Reactor Pressure Vessel for the Jersey Central Peactor Project- (February 18, 1965).

59. D. W. Coats, Recommended Revisions to Nuclear Regulatory Commission I

Seismic Design Criteria, NUREG/tR-ll61, Lawrence Livermore National Laboratory, Livermore, Calif. (December 1979) .*

60. FDSAR, Oyster Creek Unit 1, Amendment 38, Sect. V, Seismic Analysis Results of Feedwater Coolant Iniection System.

i i

61. N. C. Tsai, The Role of Radiation Dawing in the Impedance Function

' Approach to Soil-Structure Interaction Analysis, Lawrence Livermore National Laboratory Report UCRL-15233 (May,1980).

62. N. M. Neismark, "A Response Spectrum Approach for Inelastic Seismic Design j of Nuclear Reactor Facilities," in Transactions, Third International Conference on Structural Mechanics in Reactor Technology, London, Paper l K5/1 (1975).

63. N. M. Newmark, " Seismic Design Criteria for Structures and Facilities, Trans-Alaska Pipeline System," in Proceedings, U. S. National Conference

on Earthquake Engineering, Ann Arbor, Mich., Earthquake Engiceering Research Institute,94-103 (June 1975) .

64. U. S. Nuclear Regulatory Comunission, Design Limits and Loading i

Combinations for Seismic Category I Fluid System Components, Regulatory

, Guide 1.48 (1973) .

l 65. U. S. Nuclear Regulatory Commission, Standard Review Plan Sec. 3.9.3, ASME Code Class 1, 2, and 3 Components, Component Supports and Core i

Support Structures, Office of Nuclear Reactor Regulation.

66. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Sec. III (1977).

67. J. D. Stevenson, " Structural Damping Values as a Function of Dynamic Response Stress and Deformation Levels," paper Kil/l presented at the 5th SMiRT Conf., Berlin (August 14-20, 1979).

121

- w-,- .+--,,-.--_m . _ _ _ _ . - . _ _ , . , . . ..,,,-..-,___..m,..,_ - . . , , , _ _ . - _ - , , , , . - , - - , . . , . - . , - - - - _ - ,

68. P. Y. Hatago and G. S. Reimer, " Dynamic Testing of Electrical Raceway Support Systems for Economical Nuclear Power Plant Installation," paper F 79 166-0 presented at the IEEE PES Winter Meeting, New York (Februa,ry 1979'. .

69. J. J. Stevenson, Evaluation of the Cost Effects on Nuclear Power Plant Construction Resulting from the Increase in Seismic Design Level, prepared for Office of Nuclear Regulatory Research, U. S. Nuclear Regulatory Commission, Draft (May 1977).

70. J. Clapp, Emergency Service Water Pump - Seismic Review, Burns and Roe, Inc. (July 1979) .

71. Burns and Roe, Inc., Emergency Condenser - Seismic Review ( August 1979) .

72. P. J. Gallagher, Emergency Condenser - Seis:aic Evaluation, Woodward-Clyde Consultants (June 1980) .

73. PERFEX Group, Heat Exchanger Seismic Calculations (October 1978).

74. Nuclear Reactors and Earthquakes, United States Atomic Energy Commission, Division of Technical Information, TID-7024.

75. Burns and Roe, Inc., Emergency Diesel Oil Storage Tank - Seismic Review (August 1979).

l

76. L. Bergman, Diesel Oil Storage Tank - Seismic Review, Woodward-Clyde Consultants (February 1980) .

77. J. Clapp, Seismic Evaluation of Valve Eccentricity - Seismic Qualification Review, Burns and Roe, Inc. (July 1979) .

78. MPR Assoc. Inc., Seismic Evaluation of Small Pipe Valve Eccentricity -

Seismic Qualification (June 1980) .

79. General Electric Corp., Seismic Analysis of the Hydraulic Control Units Document No. 383HA853 (February 1979) .

80. John A. Blume & Associates, Earthquake Analysis: Reactor Pressure Vessel (March 1966) .

81. Gould-National Batteries, Inc., Battery Racks - Seismic Evaluation, Gould k Documvent No. RHD-064298D.

82. Gould-National Batteries, Inc., Test Procedures for the Generic Qualification of Class IE Lead-Acid Storage Batteries for Nuclear Power Generating Stations, Document No. GB-3454 (August 1978) .

83. John A. Blume & Associates, Earthquake Analysis - Instrument Racks (December 1968) .

122

84. Burns and Roe, Inc., Diesel Generator and Switchaear in Diesel Building (July 1979) .

85. Burns and Roe, Inc., Isolated Phase Bus Duct Supports - Seismic Check (July 1979) .

86. N. C. Tsai, " Modal Damping for Soil-Structure Interaction," J. Engr.

Mech. Div. , ASCE 100 (No. EM2), 323 (1974) . ,

i

• Available for purchase from the NRC/GPO Sales Program, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and/or the National Technical Information Service, Springfield, VA 22161.

• • Available for purchase from the National Technical Information Service.

123

APPENDIX A STtDY OF TURBINE AND REACTOR BUILDING RESPONSES USING DIFFERENT SOIL DAMPING REPRESENTATIONS l

l 124

A.l. INTRODUCTION AND SUlelARY The objective of the study presented in this appendix is to assess the

, accuracy of the normal mode solution method employed in the Oyster Creek soil-structure interaction analysis of the turbine and reactor buildings, as reported in Chapter 5. An additional concern gards the treatment of soil damping by using the stiffness proportional method (Section 5.8.2).

The study compared modal analysis results with those fra direct time integration analysis using dashpots to simulate soil damping. All solutions were sought in the time dmain us'ng frequency-independent soil springs and damping coefficients. The accuracy of *his approach compared to the " exact",

frequency-dependent solutions was demonstrated elsewhere; for example, in Ref.

56 for the case of a rigid footing on elastic half-space and in Ref. 57 for the coupled soil and structure systems.

Simplified 2-D turbine and reactor building models (Figs. A-1 and A-2) were used in the study. The simplified models were subjected to the same earthquake time-history and their responses were computed by the same normal mode solution procedure as for the detailed models. The comparison of the top floor response spectra curves of the simplified and the detailed models for the reactor and the turbine buildings are shown in Fig. A-3 and A-4. The good agreement for both buildings showed that the simplified models were sufficient for this purposes of this study. The simplified models for both buildings were also studied for structural damping ratios of 7% and 10%, and no significant difference in the response was observed.

The simplified models were then analyzed with dashpots to simulate radiation damping. The dashpot analysis method was that proposed by Tsai in Ref. 86. Floor spectra resulting from modal analyses employing stiffness-proportional, composite-modal damping were compared to those from the dashpot analyses. The comparison showed that in general the spectra are in good agreement. However, the magnitudes of the normal mode responses in many cases are lower, especially for the reactor building. To account for the 125

Mass (k-s2 /ft)

.Z Ashear I (ft) 2 (ft4) 8 1000 El 46.0' Kx = 3.375 X 10 k/ft o4 K,= 36.0 X 109k-ft/ rad 2500 1 X 1010 Cx = 18.59 X 104 k-s/ft C, = 9.56 X 108 k-s-ft 1000 o3 El 23.0' 2500 1 X 1010 K+ 2 K Foundation 2000 S - 1 El 0.0' X mass moment of C, inertia = 6.35 X 106 k-s2 -ft Cx f//////////////

FIG. A-1. Simplified two dimensional model of the turbine building.

I Mass Ashear I (k-s2 /ft) iI Z (ft2) (ft4) 300 o2 El 119' 1250 4.3 X 106 500 o3 El 95' 1350 4.7 X 106 6

Kx = 3.09 X 10 k/ft 500 o4 El 75' K, = 17 "i X 109 k-ft/ rad 1250 4.4 X 106 3

Cx = 94.9 X 10 k-s/ft C, = 213.5 X 106 k-s-ft 450 05 El 51' 1150 3.6 X 106 600 o6 El 23.5' 1650 5.8 X 106 700 o K, El0.0' K 3100 7.6 X 106 Foundation C+ a ,X- El - 10' 'X mass moment of '1500 & ~.

inertia = 1.87 X 10 6 k-s2 -ft Cx

///////////////'

FIG A-2. Simplified two dimensional model of the reactor building.

126

i i i i i i i i i i i i

i i ii i 2.0 -

Tn w

8

• c 2

_* 1.0

=

y .

,e -

5 e

Detailed model (case T4)

[,/' ~

S /

_ [ Simplified model -

,/ (No embedment)

/

' ' ' ' ' ' I ' ' ' ' ' ' ' '

0.2 O.01 0.1 1.0 Period (s)

FIG. A-1. Comparison of detailed and simplified model spectral curves:

turbine building top floor, El. 46', modal analysis results using 3% damping.

E-W direction.

i i i i i i ii  ; i i i i i i . .

- t .

^

cn 5

j 1.0 - -

W .

Simplified model g

g __________ __

u) - -

Detailed model (Case R2)

' ' ' ' i l i i i i i i 0.2 = = =

0.01 0.1 1.0 Period (s)

FIG. A-4. Comparison of detailed and simplified model spectral curves:

reactor building top floor, El.119', modal analysis results using 3% damping.

127

increased response observed in the dashpot analysis, the floor response spectra envelope curves for the two buildings are accordingly modified. The results are given in Appendix B.

The moments and shears oosputed by the dashpot analysis in the above-grade portion of the reactor building are notably higher than those computed by modal analysis. For the turbine building they are within the bounds of modal analysis results. Since a large factor of safety against yielding exists for the reactor building, . the increased moments and shears are still within acceptable limits. A detailed description of the dashpot studies is given in the following sections.

A.2 TURBINE BUILDING A.2.1 Analysis Method and Results The dashpot analysis model (Fig. A-1) uses half-space soil radiation damping coefficients and soil springs based on unreduced soil shear modulus, G,,,. The damping ratios corresponding to the dashpct coefficients were 0.80 for lateral and 1.0 for rocking.

In view of the high radiation damping, a modified SSI model in which the base rotation was restrained was used in the modal analysis. For this modal analysis model, composite modal damping was determined by the stiffness proportional damping method. Another modified SSI model was constructed and l analyzed. In this model, lateral translational and rotational stiffnesses of the foundation were increased to simulate a fixed base condition.

The results showed that the modified SSI model allowing only lateral motion of the base provided the best correlation with the dashpot analysis.

The comparison at different floor levels is seen in Fig. A 'i. The modal analysis of this SSI model predicted a fundamental period of 0.22 s. The building response was essentially a rigid body translation. The dashpot analysis showed somewhat higher dynamic amplification at higher frequencies due primarily to rocking.

Figure A-6 shows a comparison of the dashpot analysis results at the turbine building operating floor (El. 46') with results from normal mode analysis of the two modified SSI models. The fixed base model predicts considerably higher spectral accelerations near the period of the building's 128

2.0 i i i . . i . i . . . ..

iil i Dashpot l El 46' 1.0 -

El 23' ' -

analysis l El O' -

b ~

g -

sg -

c -

o g

,e E -

/ Modal l ElEl46' 3

y M

'y[ / analysis i i El O' 23' 0.1 i . . . . . iI i i i i i i ii 0.01 0.1 1.0 Period (s)

FIG. A-5. Comparisen of spectral curves (3t damping) for the turbine building under different representations of soil dan. ping. The Inodal analysis model allows no base rotation.

129

f-2.0 , , ,

,,,,ig i i i , , i ii A High stiffness

./ - model 1.0 -

/ N.s ,

p q.-4, m

- l -

g -

2 / v

-8

/ t

< '/y/  %\_

\

fI '

\N

g . . . - . - . - - .

, /.f - .

l

~

h. _ . ._. _m .__m e (.'. Dashpot analysis (G = 6000 ksi) ~

S Dashpot analysis (G = 1500 ksf)

~

Case T4 O.1

' ' ' ' ' il i i e i i i ii 0.01 0.1 1.0 Period (s)

FIG. A-6. Comparison of spectral curves (31 damping) for the turbine building at the operating floor, El. 46 '.

i 130

first mode (.08 s) . The building response should approach that of a fixed base structure if the soil damping becomes infinitely large. Apparently the pres int soil damping values are not large enough to simulate fixity.

Therefore, the fixed base model was not considered for floor spectra. Since the low shear modulus was negated by fixing the base, another dashpot analysis was conducted to study the effect of G = 1500 ksf. This case allowed both lateral translation and rotation. Results of this study are shown in Fig.

A-6. The spectra generated by the dashpot analysis at G,,, bounded the results of this study. Thus the G = 1500 kaf case was not considered further.

A.2.2 Implications for Reanalysis A.2.2.1 Building Moments and Shears The comparison of the dashpot and the modal analysis zero period acceleration (ZPA) results in Figs. A-5 and A-6 shows that the dashpot results are only slightly higher than the first modified SSI mod 61 results at El 46',

but are below the fixed-base results. Since the reanalysis in Chapter 5 used the larger values of the moment and shear results from both modified SSI models, no further modifications of moments and shears were necessary.

A.2.2.2 Floor Response Spectra Curves Figure A-5 shows that the results from the modal analysis at elevations 23' and 46' and for periods less than 0.4 s of the first modified SSI model are lower than those from the dashpot solution. To correct the spectra generated by modal analysis, the following modification was made (see Fig. 22 for an example of this modification):

1. Toward the shorter period side the peaks were broadened by 35% at El.

46' and by 17% at El. 23'. The peak period was taken to be 0.22 seconds (the period of the estimated soil lateral mode) .

2. From the new peak a straight line of unit slope was drawn to the point where it intercepted the flat portion of the curve.

131

A.2.3 Radiation and Material Damping Combinations The SSRT SSI Guidelines reconumended the following:

e Add soil material and radiation damping.

e Assume vertical and horizontal radiation damping to be 75% of the theoretical value.

e Assume rotational radiation damping to be 100% of the theoretical value.

For the reanalysis,100% of theoretical damphp3 was used for all directions, bat no soil material damping was included. A comparison of floor spectra for these two cases for the operating floor of the turbine building is shown in Fig. A-7. Both cases gave virtually the same results; therefore, no further consideration was given to damping combinations.

A.3. REACTOR BUILDING A.3.1 Analysis Method and Results The dashpot analysis model (Fig. A-2) uses the half-space soil daging coefficients and those soil springs derived from unreduced shear modulus, G . An identical model was used in the modal analysis except that the soil damping was expressed in terms of damping ratios and incorporated into the composite modal damping value calculated by the stiffness proportional damping method. The lateral and the rocking soil damping ratios were respectively 0.40 and 0.19.

Comparisons of floor response spectra generated by the two analysis techniques are shown in Figs. A-8 to A-14. Generally, the two response curves have the same shape. Peaks of both curves occur between 0.3 and 0.4 s, the period of the first combined soil mode. The results showed that modal analysis solutions are consistently lower than dashpot solutions at higher elevations. The ratios of ZPA and peak aesponse values are plotted in Fig.

A-15. The dashpot analysis peak values are 774 to 133% of the modal analysis values; whereas, the values of ZPA are from 834 to 130% of the modal analysis values.

132

Turbine Building El 46' 75%

l 1.0 - l',-~ N /- s -

V

- - 100%

2 -

e 'v -

,o -

i' -

• /

0 t'/

~

2e \

s

~ i O

m e/ \~

' 3 2w - , 2 i

5 g -----------------,

0.1 i i i i i i e i l i . . . i i i .

0.01 0.1 1.0 Period (s)

FIG. A-7. Comparison of soil damping using SSRT guidelines; turbine building, E l. 4ti'.

133

i i i i i , . . ..

i i i i ig i Reactor building, El: 119'3" -

/\

/ ,/ \' -

\ \

- MODSAP \

fo 1.0 -

Envelope from appendix B

/

[ Dashpot

'D - / .

e -

/

3 -

' \ )

hC  % /* \_

w- -

_ _ _ _ - _ ~

- f' 8

g - -

' ' ' ' ' ' ' 'l ' ' ' ' ' '

0.1 O.01 0.1 ' ' 1.0 Period (s)

FIG. A-8. Comparison of spectral curves (34 damping) using dashpot and MODSAP representations of soil damping; reactor bu ". ding, El.119'3".

134

,g . . . .. .

Reactor building, El: 95'3"

- A .

/ s%

/

/ \,

s k T To Envelope from appendix B MODSAP s

g 1.0 --

/ Dashpot g

~

a T -

/

y -

/ )

/

-2 (* __ /

~

0.1 O.01 0.1 1.0 Period (s)

FIG. A-9. Comparison of spectral curves (34 damping) using dashpot and MODSAP representations of scil damping; reactor building, El. 95'3".

135

1 l

1 i > > i .

.; ,

Reactor building, El: 75'3"

/

/\ \

\

/

s \

Envelope from appendix B 1.0 --

h MODSAP Dashpot 7

$ - / -

E _- / ~

)

e ,~/ -

w _ _ . _ _ _ _ _ _ _ _ _ _ _ ,

' ' ' ' ' ' ' 'I ' ' ' '..>>

0.1 0.01 0.1 1.0

Period (s)

FIG. A-10. Comparison of spectral curves (34 damping) using dashpot and MODSAP representations of soil damping; reactor building, El. 75 '3".

i s

136

I

' > i i , , ,

,I ' ' ' ' ' ' '

• Reactor building, El: 51'0" 1

f'\

a c1 30

/- \

s

~

2 -

Envelope from appendix B -

e

= MODSAP -

8 m

Dashpot

] - )

$C

a. .,~-~---

u) - - __ _

l

' ' ' ' ' ' ' 'I ' ' ' ' ' ' ' '

0.1 l 0.01 0.1 1.0 l

Period (s)

FIG. A-ll. Comparison of spectral curves (34 damping) using dashpot and MODSAP representations of coil damping; reactor building, 21. 51'.

137 l

1 i e i i i i i i i i i iiiii

Reactor building, El: 23'6" g #

/s g c 1.0 - Envelope from appendix B / \ _

3m

.5 - -

8 /v MODSAP g -

^ VV )

Dashpot

_a - / -

5 /

8 / ~

O. ( '" /

m ,/

._ _ _ _ - " Y 0.1 O.01 0.1 1.0

Period (2)

FIG. A-12. Comparison of spectral curves (34 damping) using dashpot and MODSAP represantations of soil damping; reactor building, El. 23'6".

i i

138

i i

i . i i i i i . . . . . . .

ig

- Reactor building, El: 0'0" _

/'s s i

\\

~ 1.0 - T -

5

/ '^, -

e

. ~s- w -

' ,./

.2 MODSAP

\ \ _

f Dashpot

/

s a

8 a

s I)

,/ s' i

_/

' ' ' ' ' ' ' 'I ' ' ' ' ' ' ' '

0.1 O.01 0.1 10 Period (s)

FIG. A-13. Comparison of spectral curves (31 damping) using dashpot and MODSAP representations of soil damping; reactor building, El. O'.

139

l i i i i , i ,i; i i i i i i i i Reactor building, El: - 10'0" s 's e 1.0 -

/- N -

o _ _

,,/- ' y

~ ~

0 _ e MODSAP -

8 / Dashpot 3 _

s b N_

8 - _

& \l

/

' ' ' ' ' ' ' 'I ' ' ' ' ' ' ' '

0.1 O.01 0.1 1.0 Period (s)

FIG. A-14. Comparison of spectral curves (3% damping) using dashpot and M00 SAP representations 'of soil damping; reactor building, El. -10'.

140

l I

~

120 -

@4 s

100 -

I @ ZPA _

fe $ u Peak of in-structure i response spectra 80 - ' -

g

= I

= I

,8 60 - '_ -

g _

.. 9 5 40 -

l

'I

~

20 -

['Al -

0 -

ad -

. u I I

-20 0.5 1.0 1.5 Ratio S',/S, S', = Dashpot analysis spectral acceleration S, = Modal analysis FIG. A-15. Ratios of spectral values obtained using different soil damping representations.

I 141

A.3.2 Implications for Reanalysis A.3.2.1 Building Moments and Shears The building moments and shears are directly proportional to the ZPA values. Since the factors of safety against yield computed from modal analysis results are significantly larger than 1304, (see Section A.3.1),

conclusions regarding the structural adequacy of .the reactor building remain -

unchanged.

A.3.2.2 Floor Response Spectra Curves The floor response envelope curves of the reactor building Wre modified to account for the increased response as shown by the dashpot analysis.

The envelope curves at each floor above grade were scaled up by a factor equal to the increase in peak spectral acceleration as shown in Fig. A-15.

The control point of the descending branch at 1.0 a was not adjusted since the results showed good agreement at this point. (See Fig. 23 as an example.)

The resulting envelopes are plotted on Figs. A-8 through A-12 for comparison.

A.3.3 Radiation and Material Damping Combinations i

As discussed in Section A.2.3, a comparison of radiation and material damping variations was conducted. The resulting floor spectra at the top of the reactor building are shown in Fig. A-16. Both cases gave nearly identical

! results; therefore, no further consideration was given to damping combinations.

4 142

i . . , , . . i i . , , , . .

.i Reactor Building El 119' 100%

! 31.0 - -

e _ _

.9 _ _

lii g _ _

8m _

i~ 75%

g _ _

a

(/J l

0.1 , i i i , , . il i i e i i i i 0.01 0.1 1.0 1 Period (s)

FIG. A-16. Comparison of soil damping using SSRT guidelines; reactor building, El. 119'.

l 143

0 APPENDIX B FLOOR RESPONSE SPECTRA ENVELOPES 144

{lI p~ p-

._:; . . - __ - _ _ _ _

_2 . _

i i i s t . _ - s

_ i m= ._ m ed - - .

e i m=tuoid: i

)

g i me8~R o g. ro,:. ..

i

)

g s ma5~R crZo= Io, g i

( . e ( i e n i u# c.=BE.3e n o o i

a i i .

9 it a

i u# c.=B' .3m/

s it a c# c.mBE.3o i

t r r i u# c.=BE.3 e

i c# c.eBE.3e i

e e e

le w# c.=B93e ' le s.# c.eBE3e c c c c a "o g i

l t

a "6 j i

ii t

l l a

r i

i ' t i

a r

i i

t a

t t c . t c

e s e

_ e p . i p

i e S S

__:- i

. e i L

oca o6 ~

P. "o Ps oD2oc

-(

E UD28 p.

"C

_ g- E-

. _ _ _. _ 2 t

i__:- . _

1 i i

.Z ~

._:- i o

i

)

g i

m=cwd: e

)

g i

m= Jid: u e

ce ~Io .- mgo~R c~"go r i

t. i i Io,

( . - o8o~S t ( g .. e n . t n i o , a .

i i

o . i it a

r c# c.mBE.3m t a

r i w# a a.eBE.3o t

e i m# c.mBE.3o e e >

c# o.oBE.3o t

e le o# amB93o l

a# c.mBE3c e c

c c

c '. t a a w# c.oB93otJ '

la ".o g i

l e

la "6 j i / l i

r s i r - t t v t c i c - i ep i i e

p i i i t i i S i i S - t Pw

. _ - _ _ _. 1 o6 - L ~ - _ - _:-

"o P ._ P*=g .

UD28 7.

UD2oc E

- e .. ' e -.

g- --

__zt i i i

_ ~- ~

- i i i m= c d: m=ac 1

)

g t )

g i

co&:

i mgo~Rc rga Io .-

i

(

i

( i

- n i

• i n a m g o~ Sc~r~go= I o , g i

o .

o it i it i e a , w# c.mBE.3e a r

r e o# c. e a .

le t

l e

i c# c.eBE.3o t i

c c# c.=B93m i

=BE.3e c c

i c# c.aBE.3o t

_ c a a a. # c.=B93m

' 4.#

- c.eBE3o t

"'o l

a Po l.

l t lar l

i r . t t i t t

c i i c

e i i e

p i t p i e S

i t S i i i i i i

_ _ _ ___L - L . _ _~-

Pw ob .o P. .

. Po P *= .

oD2oc e .. E UD2oc 7_-

• C 3a

• to 8,. -

n tmganae nen <@ .o0n $3gw ag engo*" oo&o343et# t0 r mo ,Ogc En e3 ge $ om @' amr 3c c$a _3 n$3%< arm g 3 ,rge3v I t  ! t7D Wc"grg g_Ha 3Q f C i _

.i wD.v lllIif- 1l

(a) (b)

, fi. 2$ 63 I '_ _ ' El: 1 'Y'I '_

3 Reactor bldg. (Hor.)

~

3 Reactor bidg. (Hor.)

~

8 2% damping 8 - 3% damping'

'5 3% damping 'S 5% damping

- 7% damping h1.0 5% dampingT 7% damping o

1.0 __

_ g _ _

i- ~_ E~

g - _ g _

S & -

0.2 ' 'l ' ' ' O.2 ''l ' ' ' ' ' ' ' '

O.1 ' ' ' ' ' 1.0 0.1 1.0 Period (s) Period (s)

El: - 19'6"' l ~

s ~ Reactor bldg. (Hor.)

g _2% damping _

.o 3% damping 5 5% damping

$ 1.0 :7% damping __

g _ _

' 'I ' ' ' ' ' ' ' '

0.2 O.1 1.0 Period (s)

(d)

( .

l

_i i i i i ii i i i ii..i i i_

s _ El: - 19'6" to 119'3" _

E Reactor bldg. (Vert.) -

8 E,1.0 - 2% damping __

~

T3  : 3% dampin g  : 5% dampin _

y _ 7% damping _

g _ _

g - -

''.I ' ' ' ' ' ' '

0.2 O1 1.0 j Period (s)

FIG. B-2. Spectral curves (horizontal component) with selected percentages of damping used in reanalysis of equipment in the reactor building at selected elevr_tions from 23'6" to -19'6"; vertical spectral curves used at all elevations analyzed.

146

(a) (b)

_ i i ii ...; i i i iiiit _ i i i iig i i i. . . i i t

, - El: 46'6" .- - El: 23'6" -

To - Turbine bldg. (Hor.) -

To - Turbine bldg. (Hor.) -

's - 3% damping -

'S ~ 3% damping 3 5% damping 5 5% damping

$ 1.0 - 7% damping - --

1.0 --7% damping --

3 3

\I  :

_ i.

3 m

8 E-W m

{  : E-W

~

N-S N-S ' ' ;

0.2 ' ' ' ' ' ' ' ' ' ' ' ' ' ' '

O.2 O.1 1.0 0.1 1.0 Period (s) Period (s)

_ ii . . . . . . .e

_ - El: 0'0" -

- Turbine bldg. (Hor.) -

e o -

'g - 3% damping t; 5% damping i 1 .0 --7% damping --

g _

s i / 2 m

1 2*./ ~

- N-S t l t I f Itl t t I f I fIf 0.2 0.1 1.0 ,

Period (s)

(d) i i i i i s i ig i

. i i i . .; i iiisi

^

m - El: 0'0" to 46'6" -

T Turbine bidg. (Vert.; 3% damping 3 1.0 -

5% damping --

j 7% damping -

8 _

8 _ -

$ -E-W -

m k - N-S -

0.2 1 ' ' ' ' ' I ' ' ' ' ' ' l ' ' ' ' '

O.1 1.0 Period (s)

Spectral curves (horizontal component) with selected percentages of -

FIG. B-3.

damping used in reanalysis of equipment in the reactor building at selected elevations f rom 119'3" to 38 '5".

147 c --

-s-w. .m., r--+r ----c e- - m- --

I APPPENDIX C SSRT Guidelines for SEP Soil-Structure Interaction Review 148

NATHAN M. NEWMARK CONSULTING ENGINEERING SERVICES 1211 CIVIL ENGINEERING BUILDING URBANA. ILLINOIS S1801 8 December 1980 Mr. William T. Russell, Chief Systematic Evaluation Program Brar.ch Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission Washington, D. C. 20555 (Mail Stop 516)

Re: SSRT Guidelines for SEP Soil-Structure Interaction Review Contract NRC-03-78-150

Dear Mr. Russell:

The Guidelines for SEP Soil-Structure Interaction Review, as prepared by the Senior Seismic Review Team, are trans-mitted herewith with signature approval.

We are appreciative of the help of the many individuals who contributed to thr: preparation of these guidelines.

Sincerely yours, 1

,k, b N. M. Newmark Chairman, SSRT dp Enclosure Distribution:

W. T. Russell - 2 T. Cheng - 1 N. M. Newmark - 2 W. J. Hall - 1 R. P. Kennedy - 1 R. Murray - 1 J. D. Stevenson - 1 149

December 8, 1980 SSRT GUIDELINES FOR SEP S0ll-STRUCTURE INTERACTION REVIEW

Background

When a structure is foundeo within or on a base of soII, it interacts with its foundation. The forces and displacements transmitted to the structure and the feedback to the foundation regions are complex in nature; the interactions that take place modify the free-field motions. Many methods for dealing with soil-structure Interaction have been proposed by a number of writers. These methods can be classified in various ways and involve generally: (1) procedures similar to those applicable to a rigid block on an elastic half-space; (2) finite element or finite difference procedures corresponding to various forcing functions acting on the combined structure-soil complex; and (3) substructure modeling techniques that may or may not include use of the direct finite element method. Another, and perhaps more convenient, classification of soil-structure Interaction analysis procedures is that of (a) direct solution techniques and (b) substructure solution techniques as described in the report entitled " Recommended Revisions to Nuclear Regulatory Commission Seismic Design Criteria", Report NUREG/CR-Il61, May 1980.

The elastic half-space theory considers a foundation plate resting on an elastic medium with harmonic oscillation applied to the plate; the few test results available to date in general have been obtained for this type of model in this excitation condition. This concept is the basis for the firs't of the three procedures described above, although for seismic excitation the problem is the inverse of the original problem formulation 150

2 in that the excitation originates in the earth. The other two methods noted also involve modeling of the structure-soil system; as such the system has intrinsic properties reflecting the make-up of the modeled system, physical properties, and especially the boundaries (for example, as they affect motion input, and reflection).

These analysis methods represent major advances in computational ability, but unfortunately all the techniques have limitations, and in many cases are not well understood. At present their use involves a great deal of Interpretive Judgment.

One principal difficulty with all of the techniques is associated with the handling of the ground input. Except for special long period waves, in most cases the ground motion is noncoherent and nonuniform.

Thus far it appears that the analysis models may not be able to handle a broad spectrum of complex wave motions. None of the techniques adequately handle nonlinear effects, which are known to be of importance. As yet no good confirmatory comparison basis exists between field observations and computations made prior to an earthquake.

This entire topic is one that requires the most careful consideration.

Exercise of Judgment as to the meaning of the results, in the light of the comments given above, is required. Reliance on any sols approach is to be avoided.

SEP Review Guideline Recommendations in keeping with the SEP approach to review existing facilities, and as reflected in the philosophy and cri teria developed to date, it appears 151

3 desirable to outline briefly one technical procedure for estima. ting soil-structure interaction effects. As a result of extensive discussions between members of the SSRT and the NRC/LLL staff, and with recognition of the many uncertainties and complexities of the topic under consideration, the general approach presented below is recommended at this time as a guideline. It will be appreciated that many decisions will have to be nede as a part of the calculational procedures described below and the exercise of Judgment obviously will be required. Justification and documentation are necessary parts of the final analysis product.

At the outset it should be noted that the simplified approach described below is not intended to preclude the use of any other procedures. The structural input motions (at the foundation level),

however developed and Justified, under no conditions shall correspond to less than 75 percent of the defined control motions (normally taken as the free-field surface motions); if a reduction in translational input motion is employed, then the rotational components of motion also should be included. If other procedures are employed they should be reviewed on a case-by-case basis.

For purposes of SEP review, one simplified approach for evaluating the effects of soil-structure interaction, involving a lumped parameter model, is deemed to be acceptable when employed under the following conditions.

l. The control motions are defined as the free-field surface motions and are input at the structure foundation level.

2. The soil stif fness, as represented by springs anchored at the 152

~4 foundation level, shall be modeled as follows.

i) To account for uncertainty in soll properties, the soll stiffnesses (horizontal..ver'tical,' rocking and torsional) employed in analysis shall include a range of soil shear moduli bounded uy (a) 50 percent of the modulus corresponding to the best estimate of the large stral.n condition, and (b) 90 percent of the modulus corresponding to the-best estimate of the low strain condition. For purposes of structural analysis three soil modulus conditions generally will suffice correspond-I ing to (a) and (b) above, and (c), a best estimated shear modulus.

For structural capacity review the analyst generally should employ the worst case condition. For equipment review the in-structure

' response spectra shall be taken as a smoothed envelope of the resulting spectra from-these three analyses.

i li) When embedment is to be considered it is recommended that the soll resistances (stiffnesses as noted above) shall correspond to 50 percent of the theoretical embedment effects. This reduction is intended to account for changes in soll properties arising from backfilling, and any

' gap effects Ili) Where it is judged necessary to model the supporting soll media as layered media, the stiffnesses are to be estimated through use of acceptable procedures.

3 The radiation and material energy dissipation (i .e. , the damping values) are considered to be additive for computation convenience.

Normally the material damping can be expected to be about 5 to 8 percent.

The geoc.etric damping (radiation energy dissipation) is recognized to be frequency-dependent. However, in order to reduce the calculational 153

5 effort (at least initially), and to be sure that excessive damping is not employed, it is recommended that values of damping be estimated theoreti-cally (on a frequency-independent basis) as follows.

I) Horizontal to be taken as 75 percen. of the theoretical value.*

11) Vertical to be taken as 75 percent of the theoretical value.*

iii) Rotation (rocking and torsional) to be taken at 100 per-cent of the theoretical value.*

In the case of layered systens the approach snployed in establishing these values needs to be Justified.

4. The following analysis approaches are caosidered to be acceptable, i) When all composite modal damping ratios ** are less than 20 percent, modal superposition approaches can be used without any valida-tion check.

ii) If in investigating the use of modal superposition approaches it is ascertained that a composite modal damping ratio ** exceeds 20 percent, one must perform a validation analysis. To perform this validation, it is generally acceptable to use a time-history analysis _in which *he energy dissipation associated with the structure is included with the structural elements, and that associated with the soil is included with the soil elements.

• As calculated by generally accepted methods, as for example given in the book Vibrations of Soils and Foundations, by F. E. Richart, Jr. , J. R.

Hall, Jr., and R. D. Woods, Prentice-Hall Inc., 1970.

• • As defined by generally accepted methods.

154

(

l

4 in-structure response spectra obtained from a modal superposition analysis employing composite modal damping throughout the frequency range of interest must be similar to or more conservative than those obtained from the validation analyses.

It is emphasized that the aforementioned procedures are Intended to be guidelines and may be subject to revision as experience is gained under the SEP Program in attempting to arrive at relatively economical and simplified techniques for estimating the possible effects of soil-structure interaction.

Respectfully submitted by the Senior Seismic Review Team:

4 . 91. bwt ad N. M. Newmark, Chairman

- M W. J. ham R. P. Kennedy F i

.[

R. C. Murray U

J. D. Stevenson 155 i . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

(7 778 U.S. NUCLEAR REGULA70filY COMMISSION NUREG/CR-1981 BIBLIOGRAPHIC DATA SHEET UCRL-53018

4. TSTLE AND SU8TsTLE (Add Volume No., of appropnaar) 2. (Leave bimkl Ssismic Review of the Oyster Creek Nuclear Power Plant as Part of the Systematic. Evaluation Program 3. RECIPIENT'S ACCESSION NO.

N/A

7. AUTHOR (Si S. DATE REPORT COMPLETED R. C. Murray, T. A. Nelson, S. M. Ma, J. D. Stevenson "[o'y" ember 1980

9. PERFORk ;NG ORGAdl2ATION NAME AND MAILING ADDRESS (include Zip Codel DATE REPORT ISSUED Lawrence Livermore. National Laboratory "}"'"pril I ^f981 P. O. Box 808 ,,,,,,,,,,,,

Livennore, California 94550

8. (Leave blankJ

12. SPONSORING ORGANIZATION N AME AND MAILING ADDRESS (include Zip Codel U. S. Nuclear Regulatory Comission Office of Nuclear Reactor Regulation 11. CONTR ACT NO.

Washington, DC 20555 Nos. A-0233 and A-0415

13. TYPE OF REPORT PE RIOD COVE RE D (in:/usdre daleff Technical N/A

15. SUPPLEMENTARY NOTES 14. (Leave otar kt

_H/A

16. ABSTRACT 000 words or lessi A limited seismic reassessment of the Oyster Creek Nuclear Power Plant was p:rformed by the Lawrence Livermore National Laboratory (LLNL) for the U. S. Nuclear Regulatory Comission (NRC) as part of the Systematic Evaluation Program (SEP). The reassessment focused generally on the reactor coolant pressure boundary and on those systems and components necessary to shut down the reactor safely and to maintain it in a safe shutdown condition following a postulated earthquake characterized by a peak horizcntal ground acceleration of 0.22 9 Unlike a comprehensive design analysis, the reassessment was limited to structures and components deemed representative of generic classes. Conclusions and recomendations about the ability of selected structures, equipment, and piping to withstand the postulated earthquake are presented.

17. KE Y WORDS AND DOCUMENT AN ALYSIS 17a DESCRIPTORS

(

17b. IDENTIFIERS /OPEN ENDED TERMS

18. AV AILABILITY ST ATEMENT 19 SE CURITY CLASS (Th,s reoort) 21. NO OF PAGES UNCLASSIFIED UNLIMITED 2gggg (This papel 22 PRICE NEC FORM 335 47 77)

- _