ML19351F316
| ML19351F316 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 01/09/1981 |
| From: | GROUND TECHNOLOGY, INC. (FORMERLY STS D'APPOLONIA |
| To: | |
| Shared Package | |
| ML19351F314 | List: |
| References | |
| NUDOCS 8101120227 | |
| Download: ML19351F316 (27) | |
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4 SEISMIC SAFETY MARGIN EVALUATION BIG ROCK POINT NUCLEAR POWER PLANT
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CHARLEVOIX, MICHIGAN SOMMARY TABLE OF CONTENTS VOLUME I REPORT, SEISMIC SAFETY MARGIN EVALUATION BIG ROCK POINT NUCLEAR POWER PLANT FACILITIES VOLUME II APPENDIX A - CONTAINMENT SHELL AND REACTOR BUILDING APPENDIX B - PRIMARY COOLANT LOOP VOLUME III APPENDIX C - TURBINE BUILDING APPENDIX D - SERVICE BUILDING AND OFFICE ADDITION VOLUME IV APPENDIX E - REINFORCED CONCRETE STACK VOLUME V APPENDIX F - SPHERE VENTILATING ROOM VOLUME VI APPENDIX G - FUEL CASK LOADING DOCK / CORE SPRAY EQUIPMENT ROOM VOLUME VII APPENDIX H - SCREENHOUSE/ DIESEL GENERATOR ROOM / DISCHARGE STRUCTURE VOLUME VIII APPENDIX I - INTAKE STRUCTURE VOLUME IX APPENDIX J - BURIED FIRE MAIN PIPING AND LAKE BED PIPE APPENDIX K - UNDERGROUND ELECTRICAL CABLE APPENDIX L - BURIED FUEL TANKS VOLUME X APPENDIX M - LIQUID RADWASTE VAULT AND TANKS 8101120232 D%IPPOIIJONILA\\
TABLE OF CONTENTS PAGE LIST OF TABLES vii LIST OF FIGURES ix 1.0 EXECUTIVE
SUMMARY
1-1
1.1 INTRODUCTION
1-1 1.2 INVESTIGATIONAL OBJECTIVES AND CRITERIA 1-2 1.3 SCHEDULE OF ANALYSES 1-3 1.3.1 Structural Systems 1-3 1.3.2 Modeling Techniques 1-3 1.3.3 Analytical Techniques 1-4 1.3.4 Basis for Safety Evaluation 1-4 1.3.5 Floor Response Spectra 1-4 1.4 CONCLUDING EVALUATION AND ASSESSMEttr 1-4
2.0 INTRODUCTION
AND ORGANIZATION OF REPORT 2-1
3.0 BACKGROUND
INFORMATION 3-1 2.1 CENERAL DESCRIPTION OF PLANT 3-1 3.2 SYSTEMATIC EVALUATION PROGRAM 3-1 3.2.1 Seismic Safety Margin Evaluation 3-2 3.2.2 Conformance with NRC Design Criteria 3-3 3.3 SCHEDULE /RD BRIEF DESCRIPTION OF SAFETY-RELATED STRUCTURES AND COMPONE!rIS 3-4 3.3.1 Structures 3-4 3.3.2 Hechanical Systems and Components 3-6 3.3.3 Buried Structures 3-7 3.3.4 Buried Piping 3-7 3.4 SCOPE OF SEISMIC SAFETY MARGIN EVALUATION 3-7 3.4.1 Seismic Input 3-8 3.4.1.1 Sample Design Seismic Response Spectra 3-8 3.4.1.2 Sample Design Time History 3-8 3.4.1.3 Damping Values 3-10 3.4.2 Loads and Load Combinations 3-10 3.4.3 Applicable Codes, Standards, and Specifications 3-11 i
C A
ii TABLE OF CONTENTS (Continued)
PACE 4.0 MODELING TECHNIQP S 4-1 4.1 PROCEDURES USED FOR MODELING 4-1 4.1.1 Building Structures and Components 4-1 4.1.1.1 Concrete Structures 4-1 4.1.1.2 Steel Structures 4-2 4.1.1.3 Special Considerations 4-2 4.2 SOIL-STRUCTURE INTERACTION 4-4 4.3 MECHANICAL SYSTEMS AND COMPONENTS 4-4 4.4 BURIED PIPING 4-5 4.4.1 Buried Fire Main Piping System and Lake Bed Pipe 4-5 4.4.2 Underground Electrical Cable 4-5 4.4.3 Buried Fuel Tanks 4-6 4.5 LAKE BED STRUCTURE (INTAKE STRUCTURE) 4-6 4.6 DAMPING 4-6 5.0 PROCEDURES USED FOR ANALYSIS 5-1 5.1 BUILDING STRUCTURES AND COMPONENTS INCLUDING MECHANICAL SYSTEMS AND COMPONENTS 5-1 5.1.1 Time History Analyses 5-2 5.1.1.1 Mode-Frequency Analyses 5-3 5.1.1.2 Evaluation of Floor Response 5-3 5.1.1.3 Determination of Maximum Forces and Moments Due to Earthquake 5-3 5.1.1.3.1 Determination of Maximum Induced Forces in Beam and/or Spring Elements 5-4 5.1.1.3.2 Determination of Maximum Induced Forces in Pipe Elements 5-4 5.1.2 Response Spectrum Analyses 5-5 5.1.2.1 Combination of Modal Responses 5-7 5.1.2.2 Three Components of Earthquake Motions 5-7 5.1.3 Static (Dead Load) Analyses 5-8 NMYhNJbNN
iii TABLE OF CONTENTS (Continued)
PAGE 5.1.4 Thermal Analyses 5-9 5.1.5 Internal Pressure 5-9 5.1.6 Lateral Earth Pressures 5-10 5.1.7 Stress Analysis 5-12 5.1.7.1 General Guidelines for Stress Analysis of Building Structures 5-12 5.1.7.1.1 Stress Analysis of Diagonal Bracings 5-14 5.1.7.1.2 Steel Columns and Column Bases 5-14 5.1.7.1.3 Concrete Walls 5-15 5.1.7.1.4 Concrete Columns 5-16 5.1.7.1.5 Steel Beans 5-16 5.1.7.1.6 Foundation Stability Analysis 5-17 5.1.7.2 Stress Analyt,is of the Reinforced Concrete Stack 5-19 5.1.7.3 Stress Analysis of Containment Shell 5-20 5.1.7.4 Stress Analysis of PCL System 5-21 5.2 BURIED STRUCTURES AND PIPING 5-23 5.2.1 Buried Piping 5-23 5.2.2 Buried Fuel Tanks 5-24 5.3 LAKE BED STRUCTURES AND SYSTEMS 5-25 5.3.1 Intake Structure 5-25 5.3.2 Lake Bed Pipe 5-25 5.4 DEVELOPMENT OF FLOOR RESPONSE SPECTRA 5-25 5.4.1 Generation of Floor Response Spectra From Time History Integration Analyses 5-27 5.4.2 Generation of Floor Response Spectra From Modal Superposition Time History Analyses 5-28 5.5 BASIS FOR SAFETY MARGIN EVALUATION 5-31 5.5.1 Working Stress "ethod 5-32 5.5.1.1 Axially Loaded Diagonal Braces 5-33 5.5.1.2 Columns 5-34 NMYOEMNN l
iv TABLE OF CONTENTS (Continued)
PAGE 5.5.1.3 Coltaan Bases 5-34 5.5.1.4 Be ans 5-35 5.5.2 Ultimate Strength Method 5-36 5.5.2.1 Load Factor 5-36 5.5.2.2 Capacity Reduction Factors 5-36 5.5.2.3 Strength Evaluation 5-36 6.0 COMPUTER PROGRAMS USED IN ANALYSIS 6-1 6.1 FINITE ELEMEE ANALYSIS 6-1 6.2 ARTIFICIAL TIME HISTORY GENERATION 6-2 6.3 GENERATION OF FLOOR RESPONSE SPECTRA 6-2 7.0 SITE CHARACTERISTICS 7-1 7.1 SITE EXPLORATION DATA 7-1 7.2 GENERALIZED SUBSURFACE PROFILES 7-1 7.3 DETERMINATION OF SOIL-STRUCTURE INTERACTION PARAMETERS 7-2 7.3.1 Frequency and Embedment Corrections of Soil Springs 7-4 7.3.1.1 Frequency Corrections 7-4 7.3.1.2 Embedment Corrections 7-5 7. ',. 2 Corrections to Soil Damping 7-6 8.0
SUMMARY
DETAILS - STRUCTURES AND SYSTEMS ANALYZED 8-1 8.1 REACTOR BUILDING AND PRIMARY COOLANT IDOP 8-1 8.1.1 Analytical Models 8-1 8.1.1.1 Reactor Building Model 8-1 8.1.1.2 Primary Coolant Loop 8 ~z 8.1.1.3 Combined PCL/ Reactor Building Model 8-3 8.1.1.4 Containment Shell Model 8-3 8.1.1.4.1 Modeling of Support Con-dition of Shell Edge 8-4 8.1.1.4.2 Penetrations 8-4 8.1.2 Method of Analysis 8-4 8.1.3 Results of Analyses 8-5 8.1.3.1 Mode-Frequency Analysis 8 -5 EMMM NM
v TABLE OF CONTENTS (Continued)
PAGE 8.1.3.2 Stress Analysis 8-5 8.1.3.1.1 Reactor Building Internal Structure 8-5 8.1.3.2.2 Primary Coolant Loop 8-6 8.1.3.2.3 Containment Shell 8-7 8.2 TURBINE BUILDING, SERVICE BUILDING INCLUDING OFFICE ADDITION, AND RADWASTE STORAGE VAULT 8-8 8.2.1 Analytical Model 8-9 8.2.2 Method of Analysis 8-11 8.2.3 Results of Analyses 8-11 8.2.3.1 Mode-Frequency Analysis 8-11 8.2.3.2 Stress Analysis 8-11 8.2.3.2.1 Turbine Building 8-11 8.2.3.2.2 Service Building 8-12 8.2.3.2.3 Radwaste Vault 8-13 8.3 REINFORCED CONCRETE STACK 8-13 8.4 SPHERE VENTILATING ROOM 8-14 8.5 FUEL CASK LOADING DOCK / CORE SPRAY EQUIPMENT ROOM 8-15 8.6 SCREENHOUSE/ DIESEL GENERATOR ROOM / DISCHARGE STRUCTURE 8-17 8.7 INTAKE STRUCTURE 8-18 8.8 BURIED FIRE MAIN PIr*NG AND IAKE BED PIPING 8-18 8.8.1 Buried Fire Main Piping System (BFMPS) 8-18 8.8.2 Lake Bed Piping 8-19 8.9 UNDERGROUND ELECTRICAL CABLE 8-19 8.10 BURIED FUEL TANKS 8-19 9.0
SUMMARY
OF RESULTS AND CONCLUSIONS 9-1 9.1 GENERAL ASSESSMENT OF STRUCTURES AND SYSTEMS ANALYZED 9-1 9.2 CRITICAL AREAS IDENTIFIED IN THE ANALYSIS 9-2 9.2.1 Primary Coolant Loop 9-2 9.2.2 Service Building Including Of fice Addition 9-3 DD11PIPOILONIM
vi TABLE OF CONTENTS (Continued)
PAGE 9.3 GENERAL OBSERVATIONS 9-4
9.4 CONCLUSION
S 9-5 REFERENCES R-1 TABLES FIGURES a
REPORT VOLUME I SEISMIC SAFETY MARGIN EVALUATION BIG ROCK POINT NUCLEAR POWER PLANT FACILITIES CHARLEVOIX, MICHIGAN 1.0 EXECUTIVE
SUMMARY
1.1 INTRODUCTION
D'AppoloM.a Consulting Engineers, Inc. (D'Appolonia), is pleased to submit this report on the seismic safety margin evaluation of the Big Rock Point Nuclear Power Plant facilities. The investigation is part of an effort by Consumers Power Company to evaluate the capability of essential structures, systems, and components to safely shut down the reactor and maintain it in a safe shutdown condition during and after a postulated (sample problem) Safe Shutdown Earthquake (SSE).
The investigation has included 15 systems / subsystems of the plant which are:
e Reactor internal structure, Containment shell, e
Primary coolant loop, o
o Turbine building, Service building and of fice addition, e
o Re!.nforced concrete stack, e Sphere ventilating room, Fuel cask loading dock / core spray equipment room, e
Screenhouse/ diesel generator room / discharge structure, o
o Intake structure, Buried fire main piping, e
Lake bed pipe, e
Underground electrical cable, e
e Buried fuel tanks, and Liquid radwaste vault.
e The work has been performed as part of the Systematic Evaluation Progran (SEP) currently being conducted by the U.S. Nuclear Regulatory Commis-sion (USNRC) to reassess the safety of older commercial nuclear power plants located throughout the United States. This report provides part N M $ khdI( b 3hh
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1-2 of the seismic evaluation portion of the SEP investigation for the Big Rock Point Nuclear Power Plant which was commissioned in the early 1960's.
Volume I of this report (main report) describes the general guidelines with respect to the overall scope of work, and it summarizes both the methods used to analyze various systems and subsystems and the results of these analyses. Two principal assumptions for safety margin evalu-ation utilized are:
That the seismic input consists of a sample pro-e blem earthquake having a zero period horizontal ground acceleration equal to 0.12g and matches the requirements of the USNRC Regulatory Guide 1.60 (1973), and e That the essential loading condition combines dead load with seismic load.
The details of analyses of each subsystem have been included as appen-dices to this main report.
1.2 _ INVESTIGATIONAL OBJECTIVES AND CRITERIA The objectives of this investigation are:
Assessment of the behavior of the plant's struc-e tural systems during and following the occurrence of the SSE event, including quantitative and quali-tative judgment of the overall margin of safety, and Providing specific recommendations with respect e
to upgrading and/or retrofitting of structural components, as appropriate.
In accordance with the intent of the SEP, a systematic evaluation of important structural components and systems has been performed. The safety evaluation has been performed by first analyzing the structure or system subject to the stipulated seismic loading conditions. The ef fects of the earthquake have then been combined with other specific loading conditions for subsequent evaluation of safety margins.
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1-3 The safety evaluation has been performed by identifying the critical areas of the structures and systems and performing stress analyses for these critical areas. The overall adequacy and margin of safety of the individual plant structures and systems have been generalized from the results of stress analyses obtained at the critical locations and from appropriate stability analyses.
Analyses of most equipment and their support systems, e.g., mechanical and electrical equipment, were not within the scope of this investigation.
1.3 SCHEDULE OF ANALYSES Table 1-1 summarizes the schedule of analyses performed for the various systems and subsystems of the Big Rock Point Nuclear Power Plant. A sunenary of the structural systems and analytical methodology is present-ed below and defines the scope of work that has been carried out during this investigation.
1.3.1 Structural Systems In general, the structural systems analyzed may be classified as belong-ing to the following general types:
Massive reinforced concrete structures, e
e Shell structures, Steel frane structures with parts constructed of e
reinforced concrete, e Light steel-frane struc tures, Piping systems, and o
Buried structural systems, including buried piping.
e 1.3.2 Modeling Techniques The primary modeling techniques Wiich have been used in performing the analyses may be categorized as follows:
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1-4 e Finite element models for structural systems, including lumped parameter representation of massive reinforced concrete structural ele-ments, 3-D line element representations of steel frame structural elements or piping systems, and shell element representations of the containment shell; Beam on elastic foundation representation of e
buried piping systems; and Model representing the buried tanks subjected e
to pressure waves.
1.3.3 Analytical Techniques The various analytical techniques used included the following:
Linear time-history analysis, e
Response spectrum analysis, o
Analysis for the passage of seismic waves (for e
buried piping systems), and Simulation technique using studies related to the e
ef fects of blast pressure waves on underground structures (for buried tanks).
1.3.4 Basis for Safety Evaluation The basis for safety evaluation has been summarily presented in Table 1-1 and in more detail in the respective appendices for each structural
[
l system.
1.3.5 Floor Response Spectra To permit the seismic analysis of existing and additional equipment and structural subsystems, floor response spectra have been generated at critical equipment locations of the structures.
t 1.4 CONCLUDING EVALUATION AND ASSESSMENT On the basis of the analytical work performed, it is concluded that the j
structures and structural systems of the Big Rock Point Nuclear Power l
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1 i
1-5 Plant are adequate to resist the combination of dead loads and the sample problem earthquake subject to the following restrictions:
A general, visual inspection of all structural e
connections should be undertaken, with special emphasis at the base plate connections of all steel columns to verify that welding and/or bolting is in accordance with construction plans and specifications.
A stress concentration probably exists in the e
primary coolant loop (PCL) at the junctions of four-inch-diameter crossties and the 24-inch-diameter downconer.
Possible undesirable dis-placements are also present in this area. These conditions require additional study. However,
prior to such a future study, an addit'.onal investigation should be performed to establish the maximum displacement capabilities of the secondary piping systems attached to the PCL system.
Additional lateral restraints should be provided e
between the two steel columns (Column Numbers E -8 and D -7) in the passageway area.
A A
Upgrading of the bracing in the northeast corner e
and additional bracing in the southeast corner of the service building-of fice addition should be considered.
Further details regarding the basis of the above exceptions are provided in Chapter 9.0.
Because the extent of SEP criteria, as applicable to the Big Rock Point l
Plant, is not known at this time, the above qualifications are based on current understanding of regulatory practice with regard to the safety requirements of the structures and structural systems in a nuclear power plant.
It should be noted that there are degrees of redundancy in the various systems and components analyzed, full consideration of which has not been possible because of the absence of explicit guidelines for the SEP analyses.
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1-6 Finally, it is our understanding that the final safety analysis will have to be conducted with respect to a site-specific postulated earth-quake to be developed through the joint ef forts of the USNRC and Con-sumers Power Company. The preliminary indication is that a less sevare seismic design basis than that used in the present analysis will probably emerge from the study.
If such is the case, the margins of safety will be even greater than those reported here, and the exceptions noted above i
should be reevaluated.
i l
2-1
2.0 INTRODUCTION
AND ORGANIZATION OF REPORT This report has been prepared for Consamers Power Company in response to the request by the United States Nuclear Regulatory Commission (USNRC), Division of Operating Reactors, Of fice of Nuclear Reactor Regulation, for backup seismic analyses and a seismic safety margin evaluation of the Big Rock Point Nuclear Power Plant safety-related structures and components. The objectives of the analyses, first defined in a letter (Consumers Power Company,1980), dated December 1, 1977, from the Director of the USNRC Of fice of the Division of Operating Reactors to Consumers Power Company, are outlined in Section 3.2 of this report.
I The program of analyses described in this report was initiated at a meeting of representatives of the USNRC, Consumers Power Company, and D'Appolonia Consulting Engineers, Inc. (D'Appolonia), at the USNRC of fices in Bethesda, Maryland on July 26, 1979.
The report consists of two units to facilitate the description and discussion of:
Background information, e
Modeling techniques, e
Analytical methodology, including computer pro-e grams utilized, o Site characteristics, Systems and structures analyzed, and e
e Results of analyses.
l The first un it, the main body of the report designated Voluce I, fea-tures an executive summary, a general presentation of the progran and methods of analysis, a description of the site characteristics and of the structures and systems analyzed, and the results of the analyses.
The second unit, the report appendices, describes in detail the modeling, IIMAIPIINDIIADMIL4
l 2-2 I
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analyses, and results of analyses for each specific structure or system i
analyzed.
The first unit or raain body of the report includes the following chapters:
I e Chapter 1.0 - Executive Summary e Ch apter 2.0 - Introduction and Organization of Report e Chapter 3.0 - Background Information e Chapter 4.0 - Modeling Techniques e Chapter 5.0 - Procedures Used for Analysis e Chapter 6.0 - Computer Programs Used in Analysis e Chapter 7.0 - Site Characteristics e Chapter 8.0 - Summary Details - Structures and Systems Analyzed e Chapter 9.0 - Discussion of Results Additionally, the following appendices or second unit are included for detailed descriptions of the analyses of the structures and systems:
Appendix A - Containment Shell and Reactor Build-e ing Appendix B - Primary Coolant Loop e
e Appendix C - Turbine Building Appendix D - Service Building and Of fice Addition e
Appendix E - Reinforced Concrete Stack e
Appendix F - Sphere Ventilating Room e
Appendix G - Fuel Cask Loading Dock / Core Spray e
Equipment Room Appendix H - Screenhouse/ Diesel Generator Room /
e Discharge Structure Appendix I - Intake Structure e
Appendix J - Buried Fire Main Piping and Lake e
Bed Pipe Appendix K - Underground Electrical Cable e
Appendix L - Buried Fuel Tanks e
Appendix M - Liquid Radwaste Vault and Tanks e
The appendices have been prepared, to the degree possible, to stand as l
independent documents which may be reviewed individually. References to the main body of the report are made tc eliminate repetition of various analytical and design criteria which may be common to many of the struc-tures and components included in the analyses.
1 D!AllN"OIIADNIL4
2-3 The technical work has been perfomed by the following D'Appolonia personnel:
Dr. P. C. Rizzo, Principal-in-Charge; Dr. A. D. Husak and Dr. A. J. Eggenberger, Project Managers; Dr. S. Chakrabarti, Project Engineer; Dr. N. F. Allen, Mr. J. D. Bland, Mr. B. P. Bundy, Mr. W. T.
Chan, Dr. Y. Y. Cho, Mr. B. R. Doyle, Dr. J. T. Onstott, Dr. M. D.
Snyder, Mr. Mark A. Tinianow, Dr. N. R. Vaidya, and Dr. J. L. Withiam, Project Staff Engineers. Staff Project Consultants include Dr. P. P.
Christiano, Mr. D. E. Shaw and Mr. R. C. Vasko. Quality Assurance activities have been directed by Dr. D. E. Troxell, Manager of Quality Assurance, and have been performed by Mr. P. F. O'Hara and Ms. C. L.
Lanners, Project Quality Assurance Staff.
The progress of the work has been monitored by the following Constaners Power Company personnel:
Mr. W. J. Becklus, Administrator, Systematic Evaluation Program; Mr. D. E. Moeggenberg, Senior Engineer; and Mr. R.
Jenkins, Staff Engineer.
+
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9-1 9.0
SUMMARY
OF RESULTS AND CONCLUSIONS 1
9.1 GENERAL ASSESSMENT OF STRUCTURES AND SYSTEMS ANALYZED The investigation of the Big Rock Point Nuclear Power Plant is part of an effort by Consumers Power Company to evaluate the capability of es-sential structures, systems, and components to safely shut down the reactor and maintain it in a safe shutdown condition during and af ter a postulated (sample problem) Safe Shutdown Earthquake (SSE). The in-vestigation has been performed for 15 sy.ntems/ subsystems of the plant.
These are:
e Reactor internal structure, o Containment shell, Primary coolant loop, e
Turbine building, e
Service building and office addition, e
e Reinforced concrete stack, Sphere ventilating room, e
Fuel cask loading dock / core spray equipment room, e
Screenhouse/ diesel generator room / discharge structure, o
e Intake structure, o Buried fire main piping, e Lake bed pipe, e Underground electrical cable, o Buried fuel tanks, and e Liquid radwaste vault.
Two principal assumptions used for safety margin evaluation were:
e That the seismic input consists of a sample prob-lem earthquake having a zero period horizontal ground acceleration equal to 0.12g and matches the requirements of the USNRC Regulatory Guide 1.60 (1973), and e That the essential loading condition combines dead load with seismic load.
In general, the analyses indicate a high degree of overall safety margin for almost all of the structures and systems analyzed for the stipulated earthquake loadings.
The general safety margin of the plant structures is of the order of 2 to 3 on the basis of elastic analyses performed for these structures. The results of the analyses indicate that the following NS NIbbbk
9-2 13 structures and systems are considered to be adequate in their pres-ent configurations to resist the sample problem earthquake:
)
Reactor internal structure, e
e Containment shell, e Turbine building, Liquid radwaste vault, e
Reinforced concrete stack, e
Sphere ventilating room, e
Fuel cask loading dock / core spray equipment room, e
Screenhouse/ diesel generator room /diacharge structure, e
e Intake structure, Buried fire main piping, e
Lake bed pipe, e
o Underground electrical cables, and e Buried fuel tanks.
Certain critical areas have been identified in the primary coolant loop system and in the service building area which are discussed below.
9.2 CRITICAL AREAS IDENTIFIED IN THE ANALYSIS On the basis of 6ummary results presented in the report, the following sections discuss critical areas among all structural systems and compo-nents analyzed.
9.2.1 Primary Coolant Loop The allowable stresses have been exceeded under combined internal pres-sure, dead load, and earthquake loading conditions at the junctions be-tween the 24-inch-diameter recirculating pipe and the 4-inch-diameter cross-connecting loop. Possible undesirable displacements are also present in this area. Stresses up to twice the allowable stress levels under SSE conditions (Level C service limit as defined in the ASME 1977 code) have been obtained. However, these stresses and the code compar-ison are considered to be conservative because:
e A Level C service limit has been assumed for code comparison. However, in the SEP evaluation of Dresden 2 Nuclear Power Plant (NUREG/CR-0891, 1980), a Level D service condition equivalent to the faulted condition has been assumed. This permits an increase of 33 percent in the allow-able stress levels over the Level C service limit.
l 2
9-3 Thus, if a code comparison is performed on the basis of a Level D service limit, the severity of the stresses will decrease substantially; The connections of the 4-inch and 24-inch pipes e
are fitted fabricated connections (weldolets).
In an investigation by Schneider and Rodabaugh (1972), it is shown that the actual stress inten-sification factors are less than the calculated ones when specially contoured, integrally rein-forced fittings are used at branch connections.
Thus, additional conservatism is also present in the calculation of the stress intensification factors at the 4-inch and 24-inch pipe connec-tions; and e The moments used in the stress calculation for the 4-inch pipe were obtained directly from the flexibility analysis of the piping system. The evaluation of moment distribution at this junc-tion on the run pipe is dependent on the degree of detailing in the analytical model. Because the analysis of this junction has been performed through simplistic modeling, a rigorous evalua-tion of the stress distribution at this junction has not been performed. Considering the nature of a stress concentration problem, it is felt that the analysis only indicates a location of potential stress concentration and does not identify the full extent of this problem.
How-ever, pending further investigation, the junc-tions of the 4-inch and 24-inch pipes should, at best, be considered as marginal.
9.2.2 Service Building Including Office Addition Three possible critical locations have been identified in this structure.
They are:
e The northeast corner of the roof of the building where the system of bracings has been found to be inadequate, e The southeast corner of the building, where potential uplift of the column footings has l
been obtained, and e The passageway Column Number E -8.
A IDAIPILADIIJONILA
9-4 In report Appendix D, suitable modifications to the respective critical areas of the service building-of fice addition have been recommended through upgrading. However, from considerations of overall safety of plant operation during an earthquake, the first two critical areas may be defined as less critical than the passageway area.
Furthermore, from considerations of ductility levels (ductility factor equal to 2 to 3 in accordance with the recommendations of Newmark and Hall [1978] in NUREG/
CR-0098), it may be possible to show that the stress levels obtained in the northeast corner of the building by inelastic analysis are permis-sible under SSE conditions. Similarly, it has been shown that the grade beam connecting the two footings at the southeast corner of the building is adequate to carry the uplift observed at this location. Thus, recom-mendations with respect to upgrading of these two areas are based on current understanding of regulatory practice, prudent judgment, and cost effectiveness with respect to the alternative of further analysis.
With respect to the third critical area, the passageway column, the importance of safety with respect to the stability of the cable trays supported by these columns dictates that upgrading of this area through additional lateral restraints should be provided.
9.3 GENERAL OBSERVATIONS On the basis of analyses of various st? mtures, the following comments are considered to be pertinent:
e The analyses of the structures and components l
generally indicate a high safety margin (general-l ly on the order of 2 to 3) under combined dead l
load and seismic loading conditions. However, l
these safety margins have been obtained on the basis of an assumed constant level of damping.
Because damping in a structure is proportional to the energy absorption capacity of the structure, the applicability of damping levels assumed in the various analyses has been reviewed following tb performance of the stress analysis of each structural system. Based on results of stress analyses, the assumed damping levels have been justified. Where the assumed damping levels were found to be slightly high relative to the stress II]fAJP]IROIIADNILA
9-5 levels obtained in the structure or system, the effects of lower damping levels on the response of the structure have been reviewed. Such evalu-ations indicate that the level of damping that may be justified on the basis of stress levels in the structure will not affect the conclusions re-garding the structure's capability to withstand the stipulated earthquakes.
Live loads on the structures during operating e
conditions of the plant have not been considered in the analyses. However, on the basis of levels of safety margins obtained, the conservatism em-ployed in various stress analyses, and considera-tions of nominal live loads which may be realis-tically assumed to be present during an SSE event, it is reasonable to predict that adequate safety margins would be obtained in the various struc-tures if live loads are combined with the present loading conditions. This observation is for in-formation purposes only, and it generally ex-cludes critical areas where deficiencies have been identified. Final SEP evaluations may re-quire detailed analysis including live loads.
9.4 CONCLUSION
S On the basis of the analytical work performed, it is felt that the structures and structural systems of the Big Rock Point Nuclear Power Plant are adequate to resist the sample problem earthquake subject to the above-cited qualifications. Because the extent of SEP criteria, as applicable to the Big Rock Point Plant, is not known at this time, our conclusions are based upon present understanding of regulatory practice with regard to the safety requirements of the structures and structural systems in a nuclear power plant. It should be noted that there are de-grees of redundancy in the various systems and components analyzed and full consideration of these has not been possible because of the absence of explicit guidelines for the SEP analyses.
Finally, it is our understanding that the final safety analysis will have i
l to be conducted with respect to a site-specific postulated earthquake to l
be developed through the joint efforts of the USNRC and Consumers Power Company. The preliminary indication is that a less severe seismic design NMb bbbb k
s 9-6 1
basis than that used in the present analysis will probably emerge from the study.
If such is the case, the margins of safety will be even greater than those reported here, and the exceptions noted above should be reevaluated, i
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i s
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0 A
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9 TABLE l-1 SCHEDULE OF ANALYSES BIC ROCK POINT NUCLEAR POWER PLANT DET& t LE D I
I STRUCTURAL SYSTEM STRUCTURAL SYSTEM TYPE MDDELINC TECHNIQUE SEI5MIC ANALTTICAL TECHN!QUE g,
p F9I5tNTED IN seactor laternal Reinforced concrete 3-to lumped permeter Time-history analyste Individually selected Adequate Appendas A Structure st ruc t ure model coupled utth a structural component eingle essa repre-analysial foundation sentat ion of the com-etability asiement shet8 and the analytical model af abw pCL Cont ainment Shell Steel apherical shell Finite element model teaponse spect rine analysis Ef fect s of penetration Adequate At t achment Al ussag asisymmetric masses on response; Appendte A i
shoti elemente with easione strees inten-capabilit y to analyse ait y evaluation; soyenetric leading buckling consideret som including effects of major penetrations Framary Coolant imop Fiping syst em Finite element model Time-history analysis Evaluat ion of stressee Potent ial st rese ten-Appendia S IPCL) using pipe and opties in all individual piping centration probles inc luding elemente. Substruc-and support eteeente; detected at the junc-At t acheent at turing t echnique is evaluation of easimise tions of 4-inch and used for modeling of di splacement for the 24-i nc h-d t emet e r the systee for coe*
a yet ee pipes. Also, large binat ion with the re-displacement e have actor building edet been obt enned with possible effects on secondary systree Turbine Building.
Interconnected st ruct urel 1-D finit e element temponse spectrue analysis tedividually selected Minor upgraJing may Appendices C, Service Seilding systems, part steet-frame model coupled s t ruc t ur al component be necessary in the D and M a
a including Of fice framed, poet reinforced utth lumped param-analysis; foundation bracings to provide Addition, and conc r e t e eter stick modela for st abi t t t y the necessary lateral gaduaste Vault major reinforced con-stif tneee in the eer-crete component s; sub-vice building and k
st ructuring technique passageway. Ot he r used to develop the st ructures leurbine snaplete analytical building and raJmaste M
model in segment s waelt) found eJequate h
h
$g Pi?e)
Reinforced Concrete 240-foot-high reinforced 3-D lumped parameter Desponse spectrie analysis St rees analyss e of at t Adequat e Appendia E
.y Stack concrete chimney st ruc~
andel unth special element s within upper y
ture
.unaider.tt.a. f.t
- f een of t,,e st ruc.
ef fec t s of seil-t ure; f oundation st ab-et rec ture interect son ility Es b
4
TABLE l-1 (Continued)
DETAILED STRUCTURAL ST87tM STRUCTURAL ST5 TEM TTFE ML)DELING TECNN!qtfE SIl$MIC ANALTTICAL TECNNIQUE F T EV A ON E LU T OF AN L PRESENTED IN Sphere Ventitetlag Light steel st ructure 3-D f rame model Response spectrue analysie Strees analysis of Adequat e Appendam F Boom selected componente!
foundation st abilit y Fuel Cash Leedang Steel-f rame emperstruc=
Lumped parameter st ick Beeponse spectrue smalysie Strees analyste of Adequate Appeedia G Dock /Cere sprey sure with reinforced a e baseeems selected componente; e
Iquipment Room concrete besement frame model foundation etability upe rs t ruc t ure Screenhouse/ Diesel Seieforced concres oped parameter etick Response spectrtse analysis Stress analysis of Adequate Appendia N Cenerator Room /
st ruc t ur e wit h st e..
model with considere-eelected componente; Deschetse st ructure frame roof tion of torsion and foundation etability contained water Intake St ructure Light steel frene struc-Lumped parameter model Seeponse spectrue analysis
$trese analpole of Adequat e Appendia I ture submerged in take eelected componente!
qualitative assees-eest of foundation stability agaiset liquef ection, slope failure, and of joint failure between intake st ructure and intake I
P P' Buried Fire Main Cast-iroa pipios with Beme on elast ic foun-Analysis using compatibilit y Strese analysia for Adequat e Appende s J Fiping gasketed jointe running dat ion consideratione due to paa-induced forces and
(,
A>
underground around the eage of seismic waves moment e, inc lud ing
\\
perimeter of the plant consideration of dif-f(
T d
ferensist movement e
'l and flesability ef-fects of joints g
Lake Bed Pipe Reinforced concrete pipe Seem en elastic foun-Analysis usias compatibility Stress analysis for Adequate Appendia J
{
}
with gesten ed joint dataen consideratione due t o pas-induced forces enJ buried along the lake sage of seisoie waves momente, including bottom consideration of differenstel move-seets and flexibility 3
ef fects of jointe m
w b
1PV
i TABLE l-l (Continued)
DETABLED STauCTURAL STSitM STauCTURAL SYSTEM TYPE MODELING TECHNIQUE
$tISMIC ANALYTICAL TECK.tIQUE 5
V LUA ON E AL T O
L PREttNTie IN underground Elect rical embles in Beae on elset ic foun-Analysee meing compatibility Strese analysis for Adequat e Appendia R Electrical Cable Trenente Eerduct pipe detion considerat ione due to pes-induced forces and embedded in sencret e sage of splenic weees moment e, including con-sideration of delfer-ential movement s and fleeibility ef fect e of joint e Buried Fuel Tanks Ser ied s t eel s ent e Buried cylindere sub-Analysis for buckling and Mesiews strese level AJequate Appendas L ject ed to o unifere stresses using studies re-evaluat ion overpressure acted to the ef fects of bleet pressure wavee on enderground structures MV 0
9 o
%p h
)
m P-e) e o
h?'P9)
%S W
D 4
h
TABLE 3-2 SYSTEM COMPONENTS DESIGN CODES
- BIG ROCK POINT NUCLEAR POWER PLANT COMPONENT DESICN CODE Containment Shell ASME,Section III Reactor Building and Other Building Structures A.
Steel Structures and Components AISC B.
Concrete Structures and Components ACI 349-75 Primary Coolant Loop Piping ASME,Section III Reinforced Concrete Stack ACI 307 Buried Pipina ASME,Section III Buried Tanks ASME,Section III
- These are the codes used in the analyses to determine code allovable stresses.
l TABLE 5-1 LOADING CONDITIONS FOR STRUCTURAL SYSTEMS BIG ROCK POINT NUCLEAR POWER PLANT STRUCTURAL SYSTEM LOADING CONDITIONS Reactor Building D+E Containment Shell D+E+Pi Primary Coolant Loop D+E+Pi Turbine Building Complex D+E+P a
Reinforced Concrete Stack D + E + T + Pa Sphere Ventilating Room D+E Screenhouse/ Diesel Generator Room /
D + E + Pa Discharge Structure NOTES:
D = dead load of the structure, loads generated by the sample Safe Shutdown Earthquake (SSE),
E
=
thermal ef fects and loads during normal operation based on the most T
=
critical condition, but excluding loss-of-coolant accident or high-energy pipe rupture, Pa = earth pressures generated against embedded concrete structures including increases due to earthquake loading conditions, and Pi = internal pressure (piping systems).
$$YObOEN
.