ML20206B293
| ML20206B293 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 03/31/1987 |
| From: | Jeffery Griffin, Hardy G, Hashimoto P, Ravindra M EQE, INC., LAWRENCE LIVERMORE NATIONAL LABORATORY |
| To: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| References | |
| CON-FIN-A-0461, CON-FIN-A-461 NUREG-CR-4826, NUREG-CR-4826-V03, NUREG-CR-4826-V3, UCID-20948, NUDOCS 8704090063 | |
| Download: ML20206B293 (236) | |
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.NUREG/CR-4826 ;
UCID-20948 Vol. 3 l 1 Seismic Margin Review of the Maine Yankee Atomic Power Station l Fragility Analysis Prepared by M. K. Ravindra, G. S. Hardy, P. S. Hashimoto, M. J. Griffin j I EQE, Inc. I Lawrence Livermore National Laboratory Prepared for U.S. Nuclear Regulatory Commission [84000IkbbOO $9
#' s NOTICE This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability of re-sponsibility for any third party's use, or the results of such use, of any information, apparatus, product or process disclosed in tnis report, or represents that its use by such third party would not infringe privately owned rights.
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. . s.
NUREG/CR-4826 UCID-20948
- Vol. 3 s !
l 1 Seismic Margin Review of the Maine Yankee Atomic Power Station I Fragility Analysis Manuscript Completed: February 1987 Dat] Published: March 1967 Prepared by i M. K. Ravindra, G. S. Hardy, P. S. Hashimoto, M. J. Griffin EOt!, Inc. Newport Beach, CA 92600 , Under Contract to: Lawrence Uvermore Naticnal Laboratory 7000 East Avenue Uvermore, CA 94650 Prepared for e Division of Engineering Safety
., Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission W. shington, DC 20555 NRC FIN A0461
N s 1 i
'1
; ABSTRACT 1
This Fragility Analysis is the third of three volumes for the Seismic Marain Review of the Maine Yankee Atomic Power Statiga. Volume 1 is the Summary Report of the first trial seismic margin review. Volume 2 Systems Analysis, documents the results of the systems screening for the review. The three volumes
+ are part of the Seismic Margins Program initiated in 1984 by the Nuclear i Regulatory Commission (NRC) to quantify seismic margins at nuclear power plants.
The overall objectives of the trial review are to assess the seismic margins of a particular pressurized water reactor, and to test the adequacy of this review # approach,' quantification techniques, and guidelines for performing the review. ' l Results from the trial review will be used to revise the seismic margin methodology and guidelines so that the NRC and industry can readily apply them to assess the 4 inherent quantitative seismic capacity of nuclear power plants. i s
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CONTENTS , ABSTRACT _ . _.. iii LIST OF TABLES . .. v1i LIST OF FIGURES _. . .
. . viii
- 1. INTI(ODUCTION . . . _ l-1 1.1 Background ..: .. 1-1 1.2 Objectiye of the Study - . . 1-2 1.3 Scope of the Study . _ ._ 1-2 1.4 Organization of the Report . 1-3
- 2. REVIEW EARTHQUAKE LEVEL ._. _ . 2-1 2.I Screening of Components . .. 2-1 2.2 Estimation of Seismic Capaeitles . . _ . . 2-1
- 3. DESCRIPTION OF PLANT STRUCTURES, SYSTEMS, AND t
COMPONENTS . . .. 3I 3.1 Maine Yankee Plant / Structures and Systems . . _. 3-1 3.1.1 Structures . . . . 3-1 3.1.2 Systems . . .. 3-1 3.2 Seismic Desisn Criteria . . 3-2 3.3 Ayailability of P1 ant Desisn Data . .. ... _ 3-2 J
- 4. REVIEW OF PLANT INFORMATION AND WALKDOWN ... . 4-1 4.1 Initlai Screening of Components . .. . . . . 4-1 4.2 Review of Design-Analys*s and Seismic Reevaluation Reports . . . . . . .. 4-4 4.3 P1a n t Waikdown . . .. .. . . .. . .. . .. 4-6 4.3.1 Identification of Target Areas for First Wa 1 k d o w n . . . ... . .. . .. . . . 4-6 4.3.2 Wa1kdown Procedures _ .. . . . . . . . 4-22 4.3.2.I Wa1kdown Team . ... . ... 4-22 IV
M CONTENTS (Continued) 4.3.2.2 Procedures for Structures and Equipment _ 4-23 4.3.3 Walkdown Documentation .. 4 29 4.3.4 Wa1kdown Resu1ts _ __ 4-30 4.3.4.1 Walkdown Findinas 4 30 4.3.4.2 Plant Unique Features _ 4 56 4.4 Maine Yankee Component Modifications _ 4-59
- 5. EVALUATION OF SEISMIC CAPACITIES OF COMPONENTS AND PLANT . 5-1 5.1 Review of Structural Modeis _ _ 5-1 5.2 Simplified Analysis and Use of Screening Tools 5-2 5.3 Second Wa1kdown . 5-4 5.4 HCLPF Capacity of Components ..
5-4 5.4.I Frasility Analysis Method 5-4 5.4.1.1 Methodology .. .. . 5-4 5.4.1.2 Steel Structures ._ 5-23 5.4.1.3 Block Walls _ _.. . . . . 5-30 5.4,1.4 Flat Bottom Storage Tanks _ 5-32 5.4.1.5 Inyerter . _ 5-40 5.4.1.6 Diese1 Fue1 Oil Day Tank _ 5-46 5.4.1.7 Containment Spray Fans _ 5-52 5.4.2 CDFM Method 5-53 5.4.2.1 Circulating Water Pumphouse . 5-54 5.4.2.2 Block Wa11 SB 35-3 ._ _ 5-54 5.4.2.3 Refueling Water Storage Tank _ 5 55 5.5 HCLPF Capaeity of P1 ant . . _. 5 56 5.5.1 Accident Sequences . 5 56 l 5.5.2 Probabilistic Method . 5-59 5.5.3 Deterministic Method . . .. . 5-67 5.5.4 Sensitivity Studles . . _ . . . . . . . . 5 68 v
l 6 CONTENTS (Continued)
- 6. COMMENTS AND RECOMMENDATIONS ON SEISMIC MARGIN REVIEW METHODOLOGY _ 6-1 6.1 SeIection of Review Eartheuake Levei ... _ 6-1 6.2 Use of Screening Guidelines . 6-2 6.2.1 Extent of Review Needed _ . 6-2 6.2.2 Additional Screenina GuideIines _ .~ _ .. 6-2 6.2.3 Difficulties of Reviewing Certain.
Components . . _ _. 6-3 6.2.3.1 Reactor Internais and CEDM ... 6-3 6.2.3.2 Equipment Within Containment 6-3 6.3 Ayailability of Qualification Data . ... . 6-4' l 6.4 Waikdown Procedures . . 6-5 l 6.5 Guidanee on CDFM Capacity Calculation Procedures . 6-6 6.6 Staffing Requirements and Schedute .. _ 6-7 6.6.1 Staffing Requirements . 7 , 6.6.2 ScheduIe . _ 6-8 6.7 Applicability of Methodology for Other Plants . 6-8
- 7. RESULTS AND CONCLUSIONS . 7-1 7.I Screened Out Systems and Components .. . .
7-1 7.2 HCLPF Capacities of Screened in Components _ 7-2 7.3 HCLPF Capaeity of Plant 7-32 7.4 Identification of Low Capacity Components 7-32 7.5 Conc 1usions . . 7 32 , I REFERENCES . .. .. .. . R-1 l APPENDIX l A. Maine Yankee Atomic Power Station Arrangement Drawings . . . A-1 vi
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LIST OF TABLES Original Desisn Damping Vaiues _ _ _. . 3-5 3.2-1 4.1-1 Initial Screening of Maine Yankee Components by Categories Based on Seismic Capacities _ 4-2 Strueture Separations 4-8 4.3-1 List of Block Walls Targeted for Walkdown . 4-9 4.3 2 4.3-3 Maine Yankee Atomic Power Plant First Walkdown List of Equipment Components - _ 4-13 4.3-4 Example Pump Walkdown Data Sheet . _ 4-31 4.3-5 Example HVAC Components Walkdown Data Sheet . 4-33 4.3-6 Example Block Walt Data Sheet . ._ _ 4-35 5.3-1 Second Walkdown Component Review List 5-5 5.4-1 Diesel Fuel Oil Tank Critically Stressed Areas _ _ 5-51 5.5-1 Component Seismic Fragility Parameters _ _ 5-57 5.5-2 Probabilities for Nonseismic Failures . 5-58 5.5-3 Summary of Plant Level HCLPF Capacities - . 5-64 7.1-1 Summary of Group A Structures . . 7-2 7.1-2 Summary of Group A Block Walls . . .. . 7-3 7.1-3 Summary of Maine Yankee Equipment Screening and HCLPF Capaeitles _._ 7-6 vil
,- 1 ;
I LIST OF FIGURES 2.1 0.3g pga Review Earthquake Ground Response Spcctra,5% and 10% Damping . _- _ 2-2 3.2-1 0.05g Design Earthquake Groured Response Spectra 3-3 3.2-2 0.10g Hypothetical Earthquake Ground Response Spe e t ra . .. . -- -..-. .-- - 3-4 4.3-1 Block Wa11 Evaluation Procedure _ 4-38 4.3-2 Example of a PORV at the El Centro Steam Plant 4-44 4.3-3 Maine Yankee PORV PR-S-14 and PR-S-15 4-45 4.3-4 Maine Yankee CS Pump Suction MOV's CS-M-91 and CS-M-92. Valves with Long Operator Arms 4-47 4.3-5 Example of Long Operator Valves Documented in the Data Base from the Coalinga Area 4 48 4.3-6 Steam Lines Spanning Between the Steam and Feed Water Valve Area and the Turbine Building . 4-50 4.3-7 Photographs of Typical Maine Yankee Cable Trays 4-52 4.3-8 Photographs of Typical Maine Yankee Instrument Racks and Component Attaehments . 4-54 4.3-9 Photographs of the Maine Yankee Control Room Ce11ing . .- .. .. . . . 4-55 4.4-1 Block Wa11 VE 21-1 .. 4-57 4.4-2 Conceptual Seismic Retrofit to Block Wall VE 21-1 ... . - . .. 4-58 viii
LIST OF FIGURES (CONTINUED) 4.4-3 Photograph of the Alarm Message Display Mounted on the Main Contr01 Board ... 4-60 , 4.4-4 Photographs of the Maine Yankee Station Service Transformer Internal Core / Coil Assembly Before the Anchorage Modification . . 4-62 4.4-5 Proposed Anchorage Modification for the Maine Yankee Station Service Transformers Interna 1 Core / Coil Assemblles . _ _ 4-63 4.4-6 Maine Yankee Air Conditioners in the Unmodified Condition . _ _ . . 64 4.4-7 Proposed Anchorage Modifications for the Maine Yankee Air Conditioners _ 4-65 4.4-8 Maine Yankee Containment Spray Fans Before the Installation of the Anchorage Modifications ... 4-67 4.4-9 Proposed Anchorage Modifications for the Maine Yankee Containment Spray Fans . _ 4-68 4.4-10 Photographs of the Unstrapped Emergency Lighting Units in the Control Room and Welding Cart Near the Containment Spray Fanr . . 4-69 5.4-1 Fragility Curyes . 5-9 5.4-2 Median,5% Nonexceedance, and 95% Nonexceedance Fraaility Curves for a Component .. .. . .-. 5-11 5.4-3 Family of Fragility Curves for a Component . .. . 5-13 ix
E LIST OF FIGURES (CONTINUED) _ 5.4-4 CireuIating Water Pumphouse .. .. 5-24 5.4-5 Circulating Water Pumphouse, Typical Braced Frame 5-25 i 5.4-6 Turbine / Service Building, El. 35'-0"/El. 39'-0" . 5-28 5.4-7 Turbine / Service Building, El. 6l'-0" . . . . 5-29 5.4-8 Block Wa11 SB35 3 - . . - 5-31 5.4-9 Refueling Water Storage Tank 5-34 5.4-10 Model for Determination of Tank Resistance Against Seismic Base Moment . 5-36 5.41I Beam Mode 1 of Tank Bottom Plate . - . 5-37 5.4-12 Yield Line Model of Anchor Bolt Chair Top Plate _ 5-39 5.4-13 Maine Yankee 12 kVA Inverter and Battery Charger, and a Close-up of the Anchorage Addition . 5-41 5.4-14 Diesel Fue! Oil Day Tank (TK-62B) 5-47 5.4-15 Diesel Day Tank Critica1 Areas . 5-48 5.4-16 Diesei Day Tank Critica1 Areas 5-49 5.4-17 Retrofit Anchorage Design for the Day Tank .. 5-50 5.5-1 Seismic Fragility Curves of a Component .... .. 5 60 5.5-2 Fragility Curves for Small LOCA Core Damage . 5-62 5.5-3 Seismic Fragility Curves for No LOCA Core Damage .. 5-63 X
l
- CHAPTER 1
' INTRODUCTION -l 1.1 Backaround A seismic margin review methodology has becin ' developed by the NRC Expert Panel as documented i1 [Budnitz et al.,1985] and (Prassinos, et al.,1985]. The objective of the methodology is to estimate a.high confidence value of the seismic capacity 'of a nuclear power plant. The methodology has been derived using the results and insights gained in , conducting over 20 . published and unpublished seismic probabilistic risk assessments, the data on actual performance of structures and equipment of industrial facilities and power plants in major earthquakes, and the data on qualification of equipment in nuclear power plants. The methodology. _
consists of selecting a review earthquake level, screening out categories of plant structures and equipment which have seismic capacities generically higher than the review carthquake level, screening out components that are not essential for certais plant safety functions, performing plant walkdown to confirm that~ the screening of components (i.e., structures and equipment) is acceptable, and estimating the scismic capacities of the components and of the overall plant. .The seismic capacity of a component or plant in this context is the so-called High Confidence Low ' Probability of Failure (HCLPF) capacity; it is a conservative representation of capacity and in simple terms corresponds to the earthquake level at which it is extremely unlikely that failure will occur. It can be -viewed as approximately equal to the earthquake level for which :we have 95% ' confidence that the probability of failure'is less than 5%. 'At this time, seismic margin methodology has been developed for pressurized water reactors based on the review of system analysis results and insights of a number of seismic PRAs on these reactors; for < boiling water reactors additional reviews of seismic PRAs are required to develop seismic margin review procedures. The seismic margin review methodology has been developed and some preliminary - .i I guidelines for performing the margin review have been indicated in the foregoing-references. It was felt necessary to apply the methodology to a selected plant on a trial basis in order to check whether the methodology is implementable and whether the guidelines need further revision and amplification. .For this purpose, the Maine Yankee Atomic Power Plant was selected. It is expected that the methodology and margin review guidelines will be revised subsequent to this trial plant review. The NRC selected the review level earthquake to be 0.3g peak ground acceleration with a NUREG/CR-0098 50th percentile spectral shape. Maine Yankee concured with this selection. l The trial plant review is conducted' by the Lawerence Livermore National Laboratory (LLNL) and is being funded by the - U.S. Nuclear Regulatory l Commission. It consists of two major aspects: system analysis and fragility evaluation. The contracts to perform these two aspects of the trial plant review have been awarded to Energy International Incorporated and EQE Incorporated, respectively. Volume 3 of the report describes the fragility evaluation aspect of i 1-1
- - - - - - - - - - - ,-,-e,. g ,.
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.l this seismic margin review. Volume I summarizes the entire seismic margin review study on Maine Yankee; Volume II discusses the system analysis aspects of the study.
1.2 Obiective of the Study This fragility aspect of the trial plant seismic margin review is aimed at achieving the following objectives: o Apply the seismic margin review metho'dology developed by the NRC Expert Panel and identify areas where the margin review guidelines need modification and clarificatio , o Estimate the HCLPF seismic capacity of the Maine Yankee Atomic Power Plant and identify any seismic vulnerabilities in the plant. 13 Scooe of the Study The primary purpose of this study has been to evaluate the NRC Expert Panel seismic margin review methodology using Maine Yankee as the trial plant; the secondary purpose is to assess the seismic margin of Maine Yankee. The scope of the study is defined with these two purposes in view. o The methodology outlines two approaches for estimating the i component and plant-level seismic capacities: probabilistic and deterministic. The guidelines for the deterministic evaluation of the capacities are not sufficiently developed. Therefore, only the probabilistic approach is used in estimating both the component and plant-level seismic capacities in this study. Although this does not check out all the features of the Expert Panel methodology, it does provide an estimate of the seismic margin of Maine Yankee. Much further work is needed to develop more definitive guidelines for deterministic evaluation of seismic margins in the Expert Panel methodology. o The seismic capacities of structures and equipment are estimated using the structural models and qualification analysis results provided by the Maine Yankee utility. The adequacy of the l structural models and the reasonableness of the seismic responses ' of structures and equipment were confirmed by cursory review and based on judgment. The floor response spectra generated by Maine Yankee are judged to be adequate for this seismic margin review. However, this cannot be construed to be a detailed review duplicating the review done as part of the plant QA/QC 1 program in its licensing. o The limitations of.the Seismic Margin Review Methodology as outlined in the cited references by the Expert Panel also apply to 1-2 l , _ _ _ _ _ _ . . _ - - -- - - - - ~ -- - - - - - ----
r . the study described in this report (i.e., relay issues, design and construction errors, operator errors under seismic stress, etc). o The screening criteria for components developed by the Expert , Panel based on their generic high seismic capacities (i.e., components denoted by letter "C" in Table 5-1 of NUREG/CR-4334) are assumed to be applicable using minimal review and - engineering judgment. In general EQE is in agreement with the Panel's recommendations. 1.4 Ornanization of the Report Chapter 2 discusses the review level earthquake specified for this trial plant seismic margin review. A general description of the Maine Yankee plant structures, systems, and components is given in Chapter 3. The processes of plant design review, initial screening of components, and the plant nikdown are detailed in Chapter 4. Review of structural models, simplified analysis and second walkdown for additional data, and HCLPF capacity calculation for components and plant are described in Chapter 5. Feedback on the methodology in the areas of selection of review earthquake level, screenirig guidelines, walkdown procedures, HCLPF capacity calculation, staffing requirements, and applicability to other plants are discussed in Chapter 6. The results of the study in terms of HCLPF capacities of components and plant, and any seismic vulnerabilities in the plant, are highlighted in Chapter 7. Appendix A consists of a set of general arrangement drawings showing the structures and equipment in the Maine Yankee plant. l t n 1-3
o l . I CHAPTER 2 REVIEW EARTHQUAKE LEVEL For this trial application of the NRC Expert Panel seismic margin review methodology, the NRC staff has specified a review earthquake level of 0.30g pga anchored to the median NUREG/CR-0098 ground response spectrum for rock sites (Figure 2.1). In this chapter, the implications of this review earthquake level will be discussed from the viewpoint of screening of components and seismic capacity calculations. 2.1 Screenine of Comoonents
- The guidelines developed by the Expert Panel for screening.of components based j
on their generic seismic capacities are considered applicabic for review carthquakes of magnitudes less than 6.5, with 3 to 5 strong motion cycles and a total duration of 10 to 15 seconds. The spectral content of this earthquake is characterized by a broad-band spectra in the structural frequency range of I to 7 Hz. The selected review earthquake for Maine Yankee meets all of the above requirements. It is, therefore, concluded that the screening guidelines given in the Expert Panel reports NUREG/CR-4334 and 4482 are applicable to the present seismic margin review. 2.2 Estimation of Seismic Canacities The concept of HCLPF capacity requires that it is associated with a defined response spectrum and a specified nonexceedance probability. The HCLPF capacity used for screening as well as the calculated HCLPF capacities
.for particular components not initially screened out and the final plant level HCLPF capacity are considered to be valid provided ground motion from any carthquake does not exceed the review earthquake level spectrum for more than 16% of the spectral frequencies within the range of interest. The review earthquake level spectrum is a spectral shape defined by the 50% exceedance 4
spectrum specified in NUREG/CR-0098 and anchored at 0.3g pga for the initial screening. The seismic margin for the components and plant is referenced to this spectrum but anchored to the pga corresponding to the HCLPF capacity. This definition of spectra used to determine a HCLPF capacity does not in any way refer to the probability of occurrence of an earthquake. It is no more than an arbitrary spectrum used to define the HCLPF capacity that recognizes the dependency of a component capacity on the frequency content of the spectrum and not just the pga. l 2-1 f
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I t' 1 The results of the seismic margin study are interpreted as follows. The HCLPF capacity of the structures, equipment and plant are conditional on the actual site-specific spectrum not exceeding the target spectrum; exceedance is defined as the event when 16 percent of the spectral ordinates exceed the target spectrum over the frequency range of interest. It is assumed that the spectrum peak-to-peak and carthquake direction variabilities (Sec. 5.4.1.1) are removed from the hazard j analysis leading to the selection of the review earthquake. The review carthquake ! is specified by the same spectrum (Figure 2.1) in two horizontal directions and 2/3 of the horizontal spectrum in the vertical direction. It is also assumed that the review earthquake level is specified as the higher of the response spectra from the two orthogonal horizontal directions. I i l 1 l 2-3 l
t CHAPTER 3 DESCRIPTION OF PLANT STRUCTURES, SYSTEMS, AND COMPONENTS 11 Maine Yankee Plant / Structures and Systems Maine Yankee plant is located on the west shore of the Back River approximately 3.9 miles south of the center of Wiscasset, Maine. The plot plan is shown in Appendix A. Maine Yankee is a one-unit,3-loop PWR supplied by Combustion Engineering with a rated capacity of 825 MWe It began commercial operation in December 1972. 3.1.1 Structures The major structures on the site are the reactor containment, primary auxiliary
; building, fuel building, turbine building, service building, and circulating water pumphouse. They are founded on hard rock.
The reactor containment is a steel-lined reinforced concrete cylinder with a hemispherical dome and an essentially flat reinforced concrete foundation mat.
.The turbine building houses the turbine generator and the two diesel generators.
The service building consists of the main control room, switchgear room, shops, and I employee facilities. Portions of the turbine / service building, control /switchgear building, and diesel generator enclosure have reinforced concrete walls and slabs. The service building above El. 39 ft 0 in. and the remaining portions of the turbine / service building are of structural steel framing with diagonal bracing and , reinforced concrete slab with metal roof deck. The primary auxiliary building houses pumps, and tanks used for purification and processing of water from the reactor coolant system. It is made of reinforced concrete walls and slabs. The circulating water pumphouse contains the circulating water pumps and service water pumps. Below El. 21 ft 0 in., the structure has reinforced concrete walls and slabs. Above this elevation, it is fabricated structural steel framing with diagonal bracing and reinforced concrete slab. Other structures of interest to this study are the ventilation equipment room, containment spray pumphouse, and the main steam valve house. All of these structures are constructed of reinforced concrete walls and slabs except that the interior structure of the main steam valve house is of structural steel framing with diagonal bracing and metal grating. 3.1.2 Systems In this study, the main focus is on the systems (i.e., front line and supporting) that support the Group A System functions. These systems are described in Volume 2 of this report. i 3-1 l l
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3.2 Seismic Desinn Criteria The original design of the Class I structures and components was. based on a "Housner Spectrum". anchored to 0.05g for the design ~catthquake and 0.10g for'the - hypothetical earthquake (SSE). (See Figures 3.2-1 and 3.2-2.) The damping values used for the design of different structures," piping and equipment are summarized in Table 3.2-1. t
- The following structures and components are characterized as Class I
o- Reactor er w.tinment, including penetrations o Reactor vessel and its internals o Reactor coolant system o Reactor containment crane ' o o Chemical and volume control system o Residual heat removal system , i o Safety injection system , o All components affecting the ability of the control rods to scram o Containment spray system i o Spent fuel pool and racks 1 o Component cooling system o Circulating water system intake structure o Service water system
.o Emergency generators i
o Refueling water storage tank , o Control room 3 o Emergency steam generator feed pumps and piping The structures and components (including piping, cable trays, and HVAC systems) were designed and qualified to the requirements of the applicable AISC, ACI, and ASME codes of 1960 to 1970 versions.
'The plant structures and equip' ment have been reevaluated at various times for carthquakes titrger than the original SSE and for more recent regulatory standards
[Stevenson,1983]; [Hashimoto et al.,1984]; [Whittier,1986]. Certain upgrades to improve the seismic capacity of the. plant have also been made, e.g., additional anchoring of electrical equipment, and strengthenin's of block walls. The present t study has focused on evaluating the seismic capacity of.the plant in its current j state, including certain modifications installed during the course of the margin review, or to be installed during the March 1987 outage. ! i 3.3 Availability of Plant Desien Data Maine Yankee is an older plant designed and constructed before the development of quality assurance programs and seismic qualification methods presently existing l in the nuclear industry. Hence, the qualification reports on certain equipment items were not available. Because of the reevaluation efforts, additional information and new structural models have become available. However, the extent of information available is not comparable to that normally available for a modern nuclear power plant (e.g., near term operating license plant). l 1 I l 3-2 1 l 4
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I l l l l Table 3.2-1 Original Design Damping Values ! Percent of Critical Damping Design Hypothetical Earthquake Earthquake
- 1. Reactor containment structure 2.0 5.0
- 2. Reinforced concrete structure, other than containment structure, founded on rock or soil 2.0 5.0
- 3. Reinforced concrete supporting structure (not founded on soil or rock) 2.0 5.0
- 4. Steel-framed structures, including
; supporting structures and foundaticas ; Bolted or riveted 3.0 5.0 Welded 1.0 2.0
- 5. Reactor vessel, internals and control rod drives Welded assemblies 1.0 1.0 Bolted assemblies 3.0 3.0
- 6. Mechanical equipment, including pumps, fans and similar items 2.0 2.0
- 7. Piping systems 1.0 2.0 3-5
The structural models were generally reviewed by EQE; the floor response spectra ' generated for the review earthquake spectrum were provided by Maine Yankee. The spectra were judged to be realistic and representative of the seismic response ; for the review earthquake. Seismic capacities of structures and equipment were estimated by comparing the original design-analysis (or reevaluation) spectral values with the new spectral values and taking int) account the margins and
' variabilities due to differences in damping, methods of analysis and testing, and procedures for mode combination and directional components. Where the original design analysis or reevaluation information was not available or applicable, the structure or equipment was analyzed by EQE to establish its seismic capacity.
l 3-6 i
l I CHAPTER 4 REVIEW OF PLANT INFORMATION AND WALKDOWN I In this chapter, a discussion of the plant information gathered during a review' of drawings, and FSAR and seismic analysis / qualification reports will be given. The plant design information for systems and components supporting Group A functions is used to perform the initial screening of components based on the Expert Panel recommendations. The review of plant information is also aimed at identifying target areas and developing a strategy for the first plant walkdown. This chapter describes the procedures used in the plant walkdown, documentation, and salient findings. .i l 4.1 Initial Screenina of Comnonents - Energy Incorporated provided a list of systems and equipment at Maine Yankee that support the Group A functions. From this list the structures containing the equipment were identified. NUREG/CR-4334 gives guidance on screening out certain structures and equipment from further consideration based on their generically high seismic capacities. For the chosen review earthquake level of 0.33 pga, the Expert Panel report recommends that the margin review may generically screen out the component categories after satisfying some caveats as shown in Table 4.1-1. Since the scope of this study includes both an evaluation of the seismic margin review methodology and a trial application to the Maine Yankee Atomic Power Station, the initial screening could not eliminate a large number of generic categories of components from further consideration in the review, walkdown, and analysis. This is so because the Expert Panel's recommendations on screening had to be generally confirmed as appropriate to Maine Yankee. i ' The following items were screened out at this stage: i o Containment structure o NSSS supports o Soil liquefaction potential For the screened-in components, further evaluation consisted of two levels: 1
- 1. For those components identified by the Expert Panel as having 4
HCLPF capacities larger than 0.33 pga, a minimal evaluation was donc during plant review and walkdown to confirm the applicability of the Panel's recommendations - for the Maine Yankee component categories. This was done in the following manner: 1 I o Control rod drive mechanism: Review of CE design 2 o Structures: All concrete structures housing Group A systems and equipment were reviewed o Valves: Sampling review
- 4-1
... _ . - - . _ ~ , _ . - , , .~ _ - _ _ _ . _ . - _ -
Table 4.1-1 Initial Screening of Maine Yankee - Components by Categories Based on Seismic Capacities ,
)
J Expert Component Recommendation Remarks
- 1. Containment C Maine Yankee has a reinforced concrete containment structure; previous study by Hashimoto, et al (1984) confirms that the HCLPF capacity is in excess of 1g.
- 2. NSSS Supports C Studies done by LLNL as part of DEGB cvaluation included the Combustion Engineering (CE)
NSSS Supports. The NSSS Sup-ports have HCLPF capacities in excess of 0.33
- 3. Reactor Internals X To be evaluated.
- 4. Control Rod Drive C To be confirmed by a review of Mechanism the CE design.
- 5. Concrete Structure C To be confirmed by a review and Failures (Shearwalls, walkdown.
diaphragms, impact)
- 6. Steel Structures X To be evaluated.
- 7. Block Walls X To be evaluated.
i
- 8. Piping X Panel's caveats to be addressed.
- 9. Valves C To be confirmed during walkdown.
- 10. Heat Exchangers X Support and anchorage to be evaluated. i 1
- 11. Tanks X To be evaluated.
l t 4-2 .l . ' l
Table 4.1-1 Initial Screening of Maine Yankee Components by Categories Based on Seismic Capacities (Continued) - Expert Component Recommendation Remarks
- 12. Batteries and Racks X To be evaluated during review and walkdown.
- 13. Active Electrical X Anchorage to be reviewed.
Equipment
- 14. Diesel Generators C To be confirmed during walkdown.
- 15. Pumps C To be confirmed during walkdown.
- 16. Soil Liquefaction C Rock site.
- 17. HVAC Systems Fans and Cooler X ' Units on vibration isolators to Units be reviewed Ducting C To be confirmed during walkdown.
- 18. Cable Trays and C To be confirmed during walkdown.
Cabling
- 19. Control Room X To be evaluated.
Ceilings
- 20. Dams, Levees X To be evaluated.
and Dikes C = HCLPF capacity is larger than 0.3g pga. X = HCLPF capacity needs to be established through review, walkdown, and/or calculations. 4-3
o Diesel generators and peripherals: Complete review o Pumps: All pumps were evaluated o HVAC ducting: Sampling review o Cable trays and cabling: Sampling review
- 2. For those components identified by the Expert Panel as requiring a review and walkdown, a detailed evaluation.was performed as described in this report; it includes steel structures housing Group A systems. .
4.2 Review of Desian-Analysis and Seismic Reevaluation Reoorts Initial and subsennent data collection efforts concentrated on those structures and - components identified by Energy Incorporated required for the reactor suberiticality and early emergency core cooling. . The plant specific seismic qualification information was primarily made available to EQE by Maine Yankee. In a few instances, outside vendors were contacted for additional information that was either lacking or proprietary for certain . components (e.g., reactor vessel internals and the control element drive mechanism). The following provides a list of the types of information collected for the Maine Yankee Atomic Power Station as part of the margin evaluation. The data collected for use in the margin evaluation can be organized into four main categories: o Drawings 1 o Maine Yankee reports and calculations o Independent review and reports o External information Drawines Types of drawings collected from Maine Yankee include the following: o Maine Yankee structural, architectural, and excavation design drawings o Design sketches of block wall seismic retrofits o Maine Yankee plant general arrangement drawings including floor elevations showing equipment locations o Support and or anchorage drawing details for equipment o Equipment vendor drawings indicating component construction i (e.g., configuration, size, and materials used) i i Maine Yankee Reoorts and Calculations Typical types of reports and calculations collected include the following: 4-4
[ . l l b o Sections from the Final Safety Analysis Report (FSAR) o Structural steel, roof deck, and block wall construction specifications , o Tables summarizing Maine Yankee block wall information o Maine Yankee generated calculations for seismic evaluation of block wall and retrofit design o Maine Yankee generated calculations for equipment scismic qualification. o A review of the Maine Yankee FSAR Amendment No. 35, Volume II, which documents the seismic qualification of vital instrument and electrical equipment. The amendment consists primarily of certified letters from vendors regarding conformance of their components to the seismic requirements of the Maine Yankee component procurement specification; however, a few calculations , were provided and used in the component evaluations.
- Indeoendent Reviews and Reoorts Several independent reviews and reports conducted for selected structures and equipment components at Maine Yankee collected include:
o Cygna report describing structure dynamic analysis models (Cygna Report BM-Y-MY-80006-5, April 1982) ! o Cygna reports describing dynamic analyses performed for several Maine Yankee critical equipment components ) o Cygna computations for building floor spectra generation using i the 0.18g pga NUREG/CR-0098 50th Percentile Ground Response Spectra o Report by J. D. Stevenson, " Seismic Review of the Maine Yankee Nuclear Power Plant," which analyzed several critical components - 2 at Maine Yankee o Report by Structural Mechanics Associates, Inc., " Conservative 2 Seismic Capacities of the Maine Yankee Reactor Containment Including and Excluding Design Incident Pressure," [Hashimoto et al.,1984]. External information not available throuah Main *e Yankee Several examples of information collected include: 4-5
1
'~
.l o Outline drawings and seismic qualification data for the Maine Yankee reactor vessel internals and the control element drive mechanism obtained from Combustion Engineering.
4 o Information regarding dimensional data and support configuration for the Maine Yankee station service transformer , laternal core / coil assembly collected from contacts with the manufacturer, General Electric Medium -Voltage Transformer Division. o Lateral load capacity of vibration isolators supporting the Maine
~
Yankee computer room air conditioners .and the laboratory air conditioner, collected from contacts with the manufacturer, Vibration Mountings and Controls Inc During the course of the margin evaluation,. several requests were made for additional pisnt or component qualification data. The additional requests were required as a result of the following: o The equipment component list ' was being refined, adding and deleting systems and components
- o Low capacity components from the first plant walkdown were identified requiring additional component specific data to
- complete the evaluation.
i 4.3 Plant Walkdown 4.3.1 Identification of Target Areas for First Walkdown Target areas for the first walkdown were developed from the initial equipment list provided by the system analysts. This list identified preliminary equipment components as critical to reactor subcriticality and early emergency core cooling. From the preliminary equipment component list the critical structures housing this equipment were identified. The following provides a list 'and brief discussion of the structures and equipment identified as target areas for the first walkdown. Structures Structures identified for the first walkdown were those identified as housing Group A components, o Containment structure o Primary auxiliary building o Circulating water pumphouse ; o Turbine / service building l o Containment spray pumphouse j o Main steam valve house 1 o M.C.C. room o Aux feed pumphouse and purge air exhaust area i
)
i 4-6 l 1 l._-_.-
o Fire water pumphouse o Fuel oil pumphouse o Appendix R diesel room The containment internal structure was not targeted - for walkdown due to inaccessibility. The location of these structures are identified on the plant layout ' drawings in Appendix A. Structure Senarations. Based upon a review of the structural drawiny, a number of separations involving Group A structures were identified. These separations are listed in Table 4.3-1 along with their gap widths as indicated on the design drawings. Block Walls Summary tables providing information on the Maine Yankee block walls were available prior to the first walkdown. These tables describe the locations of all block walls in the plant, identify any safety-related equipment on or near the block walls,' and categorize the block walls in terms of their seismic safety status. These summary tables were used to develop a list of block walls to be inspected during the first walkdown. This list is shown in Table 4.3-2. Nearly all block walls in the plant are included in this list. Certain block walls were excluded ~ because they are located in areas that obviously did not house Group A , components, based upon a comparison of block wall and equipment locations. These walls are located in the administration building, front office, fuel building, gas house, LSA storage building, RCA building, office areas of the service ] building, and parts of the yard. Walls inside containment were not targeted since
! they were known to be inaccessible for the walkdown. Block walls affecting safety ; related equipment that were not walked down are assumed to have seismic l capacities comparable to Group A walls.
Dams. Levees. and Dikes. Review of the drawings indicated that water for the fire pond is enclosed by a dike adjacent to the fire water pumphouse. Because of the proximity of the fire water pond to the plant, this dike was targeted for walkdown to determine if it could fait during a seismic event and cause flooding. Eauinment Comnonents The Maine Yankee structural drawings and general plant layout drawings were reviewed to locate the preliminary equipment items identified by the system analyst. A general walkdown sequence organized by structure was developed for
- maximum use of time during the first walkdown. Walkdown data sheets werc
{ developed for specific classes of equipment components to be used in recording manufacturer and dimensional information necessary for a fragility evaluation 2 (reference Section 4.3.3 for a discussion and example of a typical walkdown sheet). In most cases the equipment walkdown data sheets were lengthy as sufficient vendor information had not been received from Maine Yankee prior to the first l walkdown. The first walkdown list of equipment components reviewed at the Maine Yankee Atomic Power Station are listed in Table 4.3-3. 1 4-7
l Table 4.3-1 Structure Separations Floor Separation Structures Elevations Gap 1 l Containment, containment spray El.14'-6, El. 30'-0, 3" pumphouse El. 40'-0 Containment, ventilation equipment El. 21'-0, El. 40'-0 3" room j Containment, main steam valve El. 21'-0 El. 68'-0 3" house " Containment, M.C.C. room El. 21'-0, El. 46'-0, 3" El. 68'-0 Containment, aux. feed pumphouse/ El. 9'-0, El. 22'-0, 3" purge air exhaust room El. 33'-0 Containment, fuel building El. 44'-6. , 12" Containment, equipment hatch El. 20'-0, El. 54'-0 3" shield Containment spray pumphouse, El. 21'-0, El. 40'-0 3" ventilation equipment room Ventilation equipment room, main El. 21'-0, El. 40'-0 3" steam valve house , Primary auxiliary building, El. M* 0 3" service building s a
-m 4-8
Table 4.3-2 List of Block Walls Targeted for Walkdown 4
- Maine Yankee Room Wall
- Wall ID No. Building Location Location ARD 20 Appendix R El. 21' Dividing Wall lo2 Diesel -
CS Containment Sump El. 4'-0" RHR HX 3A Sump 1 Spray FPH 20 Fire Pdmp El. 20'-0" Two walls near P-4 1-2 House - j' PAB 11 Prima'ry Door), Leading to East Door Jam 1 Auxiliary Filter Cubicles El. 11'-0" PAB 11 Primary El. 11'-0" Shielding Blocks Around 2 Auxiliary Letdown Lines PAB11 Primary El. I l'-0" Loose Blocks Near East 3 Auxiliary c Wall of the Primary Drain ' i [ Tk Cubicles PAB11 Primary El.\ I l'-0" Loose Blocks on East 4 Auxiliary Side of Seal Water Cooler . Cubicle PAB 21 Primary Boric Acid Between Lines 7 and 9 l-6 Auxiliary Storage Area and Columns F and H El. 21'-0" PAB 21 Primary Deacrator Vent East Wall 7 Auxiliary Condenser Cubicle El. 21'-0" PAB 21 Primary Waste Evaporator East Wall
.8 Auxiliary Cubicle El. 21'-0" PAB 21 Primary Sampling Cubicle North Wall 9 Auxiliary Behind Charging
. Pumps, El. 21'-0" 4
4-9
I i l Table 4.3-2 List of Block Walls Targeted for Walkdown (Continued) l l i Maine Yankee Room Wall Wall ID No. Building Location Location PAB 21 Primary Curbs Charging South End 10 Auxiliary Pump Cubicles PAB 36 Primary Degasifier Vent South and East Wall 102 Auxiliary Condenser Area Around PAB Non-Nuclear (Evaporator Safety Class Charcoal Cubicle) Filter El. 36'-0" PAB 36 Primary Waste Gas Surge Removable Shield Wall 3 Auxiliary Drum Area East Wall El. 36'-0" PAB 36 Primary El. 36'-0" Primary Vent Hi-Range 4 Auxiliary Monitor Shield Blocks SB 21 Service Control Room Control Room Entrance 1-3 El. 21'-0" SB 21 Service Corridor Along Between Column 4 and 4-7 C-Line Elevator El. 21'-0" SB 21 Service Corridor Along Between Columns 1 & 4 8-10 C-Line El. 21'-0" SB 21 Service Aux. Boiler East, West, North 11-13 Room, El 21'-0" Walls SB 21 Service Elevator Elevator Enclosure 14-15 Enclosure El. 21'-0" SB 21 Service Control Room Toilet, South, West 17-19 El. 21'-0" and East Wall i SB 35 Service Cable Tray Battery Room 3 and 4 , 1-4 Room, El. 35'-0" Area; South, North ' and West Wall 4-10 1
Table 4.3-2 List of Block Walls Targeted for Walkdown (Continued) Maine Yankee Room Wall Wall ID No. Building Location Location SB 35 Service . Elevator Enclosure Except 5-6 Enclosure North Wall El. 35'-0" SB 35 Service Elevator Elevator Enclosure East 7 Enclosure Wall - North Side l El. 35'-0" SR 35 Service Cable Tray Area East Wall 8 El. 35'-0" SB 39 Service Vent and Air Wall Along 7-Line 1 Condition Between Columns E and F Equipment El. 39'-0" SB 39 Service Cable Tray Area South Wall Along Line 7 2 El. 35'-0* Between Columns D and E - SB 39 Service Cable Tray Area South Wall Along Line 7 3 El. 35'-0" Between Columns C and D SB 45 Service Switchgear Battery No. 2 and I l-3 Room, El. 45'-6" Ares; South and West i Wall and Safety-Related SWGR Room SB 45 Service Elevator Enclosure Except East 4-5 Enclosure Wall-El. 45'-6" SB 45 Service Switchgear East Wall 6 Area, El. 45'-6"- SB 45 Service Switchgear North Wall 7 Area, El. 45'-6" 2 r 4-11
Table 4.3-2 List'of Block Walls Targeted for Walkdown (Continued) Maine Yankee Room Wall Wall ID No. Building Location - Location SB 61 Service Elevator Enclosure Except i Enclosure South Wall SB 77 El. 61'-0" and 1 El. 77'-4" SB 61 Service ' Elevator Elevator Shaft and 2 Enclosure Equipment Room - SB 77 El. 61'-0" and South Wall 2 El. 77'-4" TB 21 Turbine Lube Oil Room North . Wall 1 El. 21'-0" TB 21 Turbine Corridor Along Adjacent to PCC Hx 2 C-Line Columns 7 E-4A, E-4B -
;- to 9, El. 21'-0" TB 21 Turbine Corridor Along Corridor at Service /
38 C-Line Columns 1 Turbine Building to 7 El. 21'-0" Line TB 21 Turbine El. 21'-0" Walls Separating Feed-9-11 water Pumps P-2A and
' P-2B i
TB 21 Turbine El. 21'-0" Wall Between Turbine 12 Pedestals ?4 orth Side l TB 21 Turbine El. 21'-0" . East Wall (Inside) 13 Column 9 Doorway TB 21 Turbine El. 21'-0" : East Wall (Inside) 14 Column 10 Above Doorway VE 21 Vent El. 21'-0" Walls Near Entrance 1-4 Equipment -to Containment Area Spra*/ Building Y 28 Yard RWST RWST Shielding Blocks I 4-12
I 1 Table 4.3-3 Maine Yankee Atomic Power Plant l First Walkdown List of Equipment Components j BUILDING AND EQUIPMENT ITEM SYSTEM ELEVATION TANKS
- 1. Boric Acid Storage Tank TK-2 BAT- PAB + 36'
- 2. Boric Acid Mix Tank TK-3 BAT PAB + 11'
- 3. Refueling Cavity Water Storage Tank TK-4 HPSI Yd. + 20'
- 4. Primary Component Cooling Surge Tank TK-5 PCC SB + 61'
- 5. Volume Control Tank TK-6 CH PAB + 11'
- 6. Demineralized Water Storage Tank TK-21 AFW Yd. + 20'
- 7. Auxiliary Fuel Oil Supply Tank (buried) TK-28A FO APR + 21'
- 8. Auxiliary Fuel Oil Supply Tank (buried) TK-28B FO APR + 21'
- 9. Chemical Spray Addition Tank TK-54 CS Yd. + 21'
- 10. Secondary Component Cooling Surge Tank TK-59 SCC SB + 70' ,
- 11. Emergency Diesel Day Tank TK-62A FO AB + 21' l
- 12.
- Emergency Diesel Day Tank TK-62B FO AB + 21' i
- 13. DG-I A Compressed Air Tank TK-76Al DG AB + 21'
- 14. DG-1 A Compressed Air Tank TK-76A2 DG AB + 21'
- 15. DG-1 A Compressed Air Tank TK-76A3 DG AB + 21' )
- 16. Diesel Starting Air Receiver I A TK-76A-4 DG AB + 21'
- 17. Diesel Starting Air Receiver I A TK-76A-5 DG AB + 21' ,
- 18. Diesel Starting Air Receiver I A TK-76A-6 DG AB + 21' '
- 19. DG-1B Compressed Air Tank TK-76BI DG AB + 21'
- 20. DG-1B Compressed Air Tank TK-76B2 DG AB + 21'
- 21. DG-1B Compressed Air Tank TK-76B3 DG AB + 21'
- 22. Diesel Starting Air Receiver IB TK-76B-4 DG AB + 21'
- 23. Diesel Starting Air Receiver IB TK-76B-5 DG AB + 21'
- 24. Diesel Starting Air Receiver IB TK-76B-6 DG AB + 21'
- 25. Chemical Feed Tank TK-89 AFW AF + 20'
- 26. Sample Tank Tk-94 PCC TB + 21' i
- 27. Chemical Additive Tank (Pipe capped off) SCC PAB + 21' l
- 28. Fuel Tank DG-2 DG Yd + 21' PUMPS
- 1. Fire Pump (Diesel) P-5 ASDA FP + 20'
- 2. Boric Acid Transfer Pumps P-6A BAT PAB + 21'
- 3. Boric Acid Transfer Pumps P-6B BAT PAB + 21'
- 4. Boric Acid Transfer Pumps P-6C BAT PAB + 21'
- 5. Auxiliary Charging Pump P-7 CH PAB + 11' 4-13
l Table 4.3-3 Maine . Yankee Atomic Power Plant-First Walkdown List of Equipment Components (Continued) 1 BUILDING AND~ EQUIPMENT ITEM SYSTEM ELEVATION. ! i PUMPS (Continued)
- 6. Primary Component Cooling Pump . P-9A .PCC. _TD + 21'.
- 7. Primary Component Cooling Pump P-9B PCC TB + 21'
- 8. Secondary Component Cooling Pump P-10A -- SCC TB + 21'-
- 9. Secondary Component Cooling Pump P-10B SCC TB + 21' -
- 10. Charging Pump . P-14A HPSI _ PAB + 21'
- 11. Charging Pump P-14B _ , . HPSI PAB + 21'
! 12. Charging Pump P-14S ' HPSI PAB + 21'
- 13. Emergency Feed Pump P-25A
- _AFW; AF + 21'
- 14. Auxiliary Feed Pump P-25B .AFW VA ' + 21'
- 15. Emergency Feed Pump P-25C AFW AF + 21'
- 16. Service Water Pump P-29A' SW-_
CW + 7'
- 17. Service Water Pump P-29B SW CW + 7'
- 18. Service Water Pump P-29C SW CW + 7'
- 19. Service Water Pump P-29D SW CW + 7'
- 20. Auxiliary Fuel Oil Transfer Pump P-33A FO Yd + 21' 5
- 21. Auxiliary Fuel Oil Transfer Pump P-33B . .FO Yd + 21'
- 22. Service Water Sampling Pump P-38 . SW TB + 21' 4
- 23. Containment Spray Pump P-61 A CS CS +: 14'
- 24. Containment Spray Pump P-61 B CS CS + 14'
- 25. Containment Spray Pump P-61 S CS . CS + 14' :
- 26. Boric Acid Mix Tank Pump P-81 BAT PAB + 11' a
- 27. SWS Mussel Control Pump P-86 .SW- CW + 7'
- 28. Chemical Feed Pump P-IIS AFW AF + 20'
- 29. Service Water Seal Pit Sample Pump P-Il6 SW j 30. Fuel Pumps (4 pumps total on DG skid) DG -on Diescl
- 31. Lubrication Oil Pumps (8 pumps on DG skid) DG- on Diesel
- 32. Water Pumps (4 pumps total on DG skid) DG . on Diesel
- 33. DG-2 Fuel Pump ASDA Yd + 21'-
HEAT EXCHANGERS l 1. Pressurizer E-2 PCC RC + 20'
- 2. Residual Heat Removal Heat Exchanger E-3A PCC CS + 14'
- 3. Residual Heat Removal Heat Exchanger E-3B SCC CS + 14'-
- 4. Primary Component Cooling Heat Exchanger E-4A PCC. TB + 21' 4-14
Table 4.3-3 Maine Yankee Atomic Power Plant First Walkdown List of Equipment Components (Continued) BUILDING AND i SYSTEM ELEVATION EQUIPMENT ITEM HEAT EXCHANGERS (Continued)
- 5. Primary Component Cooling Heat Exchanger E-4B PCC TB + 21'
- 6. Secondary Component Cooling Heat Exchanger E-5A SCC TB + 21'
- 7. Secondary Component Cooling Heat Exchanger E-5B SCC TB + 21'
- 8. AC Electric Compressor Fan Cooler E-20 Elec TB + 39'
- 9. Seal Water Heat Exchanger E-34 PCC PAB + 11'
- 10. Diesel Generator Heat Exchanger E-82A PCC AB + 21'
- 11. Diesel Generator Heat Exchanger E-82B SCC AB + 21'
- 12. Oil Coolers E-86A AFW AF ' + 20'
, 13. Oil Coolers E-86B AFW VA + 20'
- 14. Oil Coolers E-86C AFW AF ' + 20'
- 15. Charging Pump Seal Leakage Cooler E-92A SCC PAB + 11'
- 16. Charging Pump Seal Leakage Cooler E-92B PCC PAB + 11' ,
MISCELLANEOUS COMPONENTS
- 1. Control Air Compressor C-10A ASDA VA + 21'
- 2. Control Air Compressor C-10B ASDA VA + 21'
- 3. DG-1 A Starting Air Compressor C-51 A DG AB + 21'
- 4. DG-1B Starting Air Compressor C-51B DG AB + 21'
- 5. Diesel Generator DG-1 A DG AB + 22'
- 6. Diesel Generator DG-1B DG AB + 22'
- 7. Diesel Generator DG-2 ASDA AB + 21' l
- 8. Primary Ejector EJ-2A SPC 'TB + 39'
- 9. Primary Ejector EJ-2B SPC TB + 39' i
- 10. Atmospheric Steam Dump Valve Silencer S-1 SPC VA + 40' l
- 11. Traveling Screen SR-1 A SW CW + 21'
- 12. Traveling Screen SR-1B SW CW ' + 21' l
- 13. Traveling Screen SR-IC SW CW + 21'
- 14. Traveling Screen SR-ID SW CW + 21'
- 15. Turbine for P-25B T-1 AFW VA + 21'
- 16. Charging Pump Seal Leakage Cooler E-92B PCC PAB + 11' l
f 4-15
, Table 4.3-3 Maine Yankee Atomic Power Plant First Walkdown List of Equipment Components (Continued) BUILDING AND EQUIPMENT ITEM SYSTEM ELEVATION ELECTRICAL DISTRIBUTION SYSTEMS
- 1. 4160V Emergency Bus 5 Elec SB + 46'
- 2. 4160V Emergency Bus 6 - Elec SB + 46'
- 3. 480V Emergency Bus 7 Elec SB + 46'
- 4. 480V Emergency Bus g - Elec SB + 46'
- 5. 480V Emergency Motor Control Center MCC-7A Elec SB + 46'
- 6. 480V Emergency Motor Control Center MCC-7B Elec RMC + 21'
- 7. 480V Emergency Motor Control Center MCC-7BI Elec CS + 20'
- 8. 480V Emergency Motor Control Center MCC-8A Elec SB + 46'
- 9. 480V Emergency Motor Control Center MCC-8B Elec RMC + 21'
- 10. 480V Emergency Motor Control Center MCC-8BI Elec CS + 20'
- 11. 120V AC Vital Bus 1 Elec SB + 21'
- 12. 120V AC Vital Bus I A Elec SB + 21'
- 13. 120V AC Vital Bus 2 Elec SB + 21'
- 14. 120V AC Vital Bus 2A Elec SB + 21'
- 15. 120V AC Vital Bus 3 Elec SB + 21'
- 16. 120V AC Vital Bus 3A. Elec SB + 21'
- 17. 120V AC Vital Bus 4 Elec SB + 21'
- 18. 120V AC Vital Bus 4A Elec SB + 21'
- 19. 125V DC Bus 1 Elec SB + 46'
- 20. 125V DC Bus 2 Elec SB + 46' ,
21'. 125V DC Bus 3 Elec SB + 46'
- 22. 125V DC Bus 4 Elec SB + 46'
- 23. Station Battery No.1 (Lead Antimony) Elec SB + 46'
- 24. Station Battery No. 2 (Lead Antimony) Elec SB + 46'
- 25. Station Battery No. 3 (Lead Antimony) Elec SB + 35'
- 26. Station Battery No. 4 (Lead Antimony) Elec SB + 35'
- 27. Battery Charger No. BC-1 Elec SB + 46'
- 28. Battery Charger No. BC-2 Elec SB + 46'
- 29. Battery Charger No. BC-3 Elec SB + 46' l 30. Battery Charger No. BC-4 Elec SB + 46'
- 31. Inverter No. INVR-1 Elec SB + 46'
- 32. Inverter No. INVR-2 Elec SB + 46'
- 33. Inverter No. INVR-3 Elec SB + 46'
- 34. Inverter No. INVR-4 Elec SB + 46'
- 35. Station Service Transformer X-507 Elec SB + 46'
- 36. Station Service Transformer X-608 Elec SB + 46' i
4-16 t _ _ _ - - . - - _ -. ,-, .__ -
l Table 4.3-3 Maine Yankee Atomic Power Plant First Walkdown List of Equipment Components (Continued) BUILDING AND SYSTEM ELEVATION i EQUIPMENT ITEM , . ELECTRICAL DISTRIBUTION SYSTEMS (Continued)
- 37. 480V MCC-9B1 (normally off MCC-9B) Elec APR + 20'
- 38. 120V Vital Bus 7 Elec AF + 20'
- 39. 125V Bus 6 Elec APR + 20'
- 40. Station Battery 6 Elec - APR '+ 20'
- 41. Battery Charger No. 6 Elec APR + 20'
- 42. Inverter No. 7 Elec APR + 20'
- 43. Alternate Shutdown Panet Elec AF + 21'
- 44. 480V MCC 1IB (mix-tank agitator) Elec FB + 21'
- 45. 480V MCC-IID (for pump P-86) Elec CW + 21'
- 46. 480V MCC 9B (mix tank pump) Elec PAB + 21'
- 47. Diesel Generator Control Board DG-1A Elec AB + 27'
- 48. Diesel Generator 480V Distribution Cab 1 A Elec AB + 22'
- 49. Diesel Generator Control Board DG-1B Elec AB + 22'
- 50. Diesel Generator 480V Distribution Cab IB Elec AB + 22'
- 52. 120V Distribution Cabinets (Diesel Backed) Elec SB + 46'
- 53. Main Control Board Elec SB + 21' HVAC
- 1. DG-1 A Room Exhaust Fan FN-20A HV AB + 31
- 2. DG-1B Room Exhaust Fan FN-20B HV AB + 31' i 3. Protected SWGR Room Supply Fan FN-31 HV SB + 39' 4
- 4. Protected SWGR Room Exhaust Fan FN-32 HV SB + 55'
- 5. AC Electric Compressor Fan Fan FN-33 ? TB + 39'
- 6. Fan FN-61 CH PAB + 21' VALVES l
i
- 1. Aux Feedwater Regulating Valve AFW-A-101 AOV AFW AF + 23'
- 2. Aux Feedwater Regulating Valve AFW-A-201 AOV AFW AF + 23'
- 3. Aux Feedwater Regulating Valve AFW A-301 AOV AFW AF + 23'
- 4. Block Valve for AFW-A-101 AFW A-338 AOV AFW AF + 23'
- 5. Block Valve for AFW-A-201 AFW-A-339 AOV AFW AF + 23'
- 6. Block Valve for AFW-A-301 AFW A-340 AOV AFW AF + 23'
- 7. Boric Acid VCT Isolation valve BA-A-32 BAT
- 8. Boric Acid VCT Isolation valve BA-A-80 BAT l 9. Boric Acid Flow Control valve BA-F-30 BAT i
- 10. Emergency Boration Isolation valve BA-M-36 BAT i
4 17
Table 4.3-3 Maine Yankee Atomic Power Plant First Walkdown List of Equipment Components (Continued) BUILDING AND EQUIPMENT ITEM SYSTEM ELEVATION ! l VALVES (Continued)
- 11. Emergency Boration Isolation valve BA M-37 BAT.
- 12. HPSI Pump B Discharge to charging header HPSI PT + 13' CH-A-32 i
- 13. HPSI Pump A Discharge to charging header HPSI PT + 13' CH-A 14. VCT Discharge to HPSI Pumps CH-M-1 MOV CH PAB + 21'
- 15. VCT Discharge to HPSI Pumps CH-M-87 MOV CH PAB + 24' 2
- 16. Containment Spray Header Isolation Valve CS-M-1 CS CS + 19'
- 17. Containment Spray Header Isolation Valve CS-M-2 CS CS + 19'
- 18. Spray Chem Tank Isolation Valve CS-M 66 MOV ,CS Yd + 29'
- 19. Spray Chem Tank Isolation Valve CS M-71 MOV CS Yd + 29' 4
- 20. CS Pump Containment Suction CS M-91 MOV CS CS - 08'
- 21. CS Pump Containment Suction CS-M-92 MOV CS CS - 08'
] 22. HPSI Discharge to Loop 1 HSI-M-Il MOV HPSI PAB + 23' , j 23. HPSI Discharge to Loop i HSI M-12 MOV HPSI PAB + 23'
- 24. HPSI Discharge to Loop 2 HSI-M-21 MOV HPSI PAB + 23'
- 25. HPSI Discharge to Loop 2 HSI-M-22 MOV HPSI PAB + 23'
- 26. HPSI Discharge to Loop 3 HSI-M-31 MOV HPSI PAB + 23' i 27. HPSI Discharge to Loop 3 HSI M-32 MOV HPSI PAB + 23 j 28. HPSI Discharge to SI Header HSI-M-40 MOV
' HPSI PAB + 23'
- 29. HPSI Pump Discharge HSI M-41 MOV HPSI PAB + 23' j 30. HPSI Pump Discharge HSI M 42 MOV - HPSI PAB + 23'
- 31. HPSI Discharge to SI Header HSI-M-43 MOV HPSI PAB + 23'
- 32. HPSI Suction from RWST HSI-M-50 MOV HPSI Yd + 21'
! 33. HPSI Suction from RWST HSI M-51 MOV HPSI Yd + 21'
- 34. CS Discharge to HPSI Pump HSI M-54 MOV CS CS + 19'
- 35. CS Discharge to HPSI Pump HSI-M-55 MOV CS CS + 19'
- 36. RWST Discharge to LPSI LSI M-40 MOV CS Yd + 28'
- 37. RWST Discharge to LPSI LSI M-41 MOV CS Yd + 28'
- 38. Decay Heat Release Valve MS-A-162 AOV ASDA VA + 43'
, 39. AFW Pump B turbine throttle valve MS-A-173 AFW VA + 21' ! 40. Main Steam Stop Check Valve MS-M-10 MOV MS VA + 49'
- 41. Main Steam Stop Check Valve MS-M-20 MOV MS VA + 49'
- 42. Main Steam Stop Check Valve MS-M-30 MOV MS VA + 49'
- 43. Decay Heat Release MS-M-161 MOV SPC VA + 43'
- 44. Auxiliary Steam Supply Valve MS-M-255 MOV MS VA + 43'.
- 45. Turbine Steam supply pressure control MS-P-168 AFW VA + 21'
- 46. Steam Generator Safety Valve MS-S-12 SPC VA + 39',
- 47. Steam Generator Safety Valve MS-S-13 SPC VA + 39' 4
4 18
i I Table 453-3 ' Maine Yankee Atomic Power Plant-First Walkdown List of Equipment Components (Continued) i BUILDING AND-SYSTEM ELEVATION EQUIPMENT ITEM
= VALVES - (Continued)
Steam Generator Safety Valve MS-S-14 SPC VA + 39' 48. Steam Generator Safety Valve MS-S-15 SPC ' V A + 39'. - 49. Steam Generator Safety Valve MS-S-16 SPC: VA + 39'
;' 50. VA + 39' Steam Generator Safety Valve MS-S-17 SPC 51.
Steam Generator Safety Valve MS-S-22 SPC VA + 39': 52. Steam Generator Safety < Valve MS-S-23 SPC VA + 39' 53. f i 54. Steam Generator Safety Valve MS-S-24 SPC VA + 39' Steam Generator Safety Valve MS-S-25 SPC . VA ^ + 39' .
- 55.- !
Steam Generator Safety Valve MS-S-26 SPC : V A - + 39' 56.
- 57. Steam Generator Safety Valve MS-S-27 SPC' VA + 39' Steam Generator Safety Valve MS-S-32 .SPC. VA -+ 39'-
58.
- 59. Steam Generator Safety Valve MS-S-33 SPC VA + 39' Steam Generator Safety Valve MS-S-34 SPC VA ' + 39' ~
60.
- 61. Steam Generator Safety Valve MS-S-35 .SPC VA + 39'
- 62. Steam Generator Safety Valve MS-S-36 SPC VA + 39' : 4
- 63. Steam Generator Safety Valve MS-S-37 SPC- VA + 39' '
- 64. Return from Penetration Coolers PCC-A-216 PCC PT + 12' 2
- 65. Return from Penetration Coolers PCC-A-238 PCC- PT + 12' i 66. Return from RCP Coolers PCC-A-252 PCC - RC + O l' ' -
- 67. Return from RCP Coolers PCC-A 254 'PCC - PT + 12' ' *
- 68. Return from CEA Air Coolers PCC-A-268 PCC RC +' 01'.
. RC + Ol'
- 69. Return from CEA Air Coolers PCC-A-270 PCC
] Return from Drain Cooler PCC-A-299 7 70.
- 71. Return from Drain Cooler PCC-A-300 PCC- RC + Ol'
- 72. Return from Drain Cooler PCC-A-302 PCC' RC + Ol'
- 73. Diesel 1 A Cooling Water Outlet PCC-A-493 -PCC AB + 24'
- 74. PCCW Outlet from RHR Heat Exchanger PCC-M-43 PCC CS + 02
, 75. PCCW Isotation to BR & LW Coolers PCC-M-90 PCC PAB + 11' ~
- 76. PCCW Isolation to Letdown Heat Exchangers ~ ~ PCC; PAB + 21' l
, PCC-M-150
- 77. PCCW Isolation to Containment PCC-M-219 PCC PAB + .11' !
j l
- 78. Pressurizer Safety Valve PR S-Il SRV RC ..+ 65' - -
- 79. Pressurizer Safety Valve PR-S-12 SRV- RC + 65'
- 80. Pressurizer Safety Valve PR S-13 'SRV RC + 65'
- 81. Power-Operated Relief Valve PR-S-14 PORV. RC + 66'
- 82. Power-Operated Relief Valve PR-S-15 PORY RC + 66' 4 83. Power Operated Block Valve MOV PR M 16 PORV RC + 64'
- 84. Power-Operated Block Valve MOV PR-M-17 PORY RC + 64'
- 85. Nonscismic Return Header Stop SCC-A-460 SCC TB + 43' i
! 4-19 i
l 1 l Table 4.3-3 Maine Yankee Atomic Power Plant First Walkdown List of Equipment Components (Continued) BUILDING AND EQUIPMENT ITEM SYSTEM ELEVATION VALVES (Continued) i 86. Nonseismic Return Header Stop SCC-A-461 SCC TB + 42'
- 87. RCP 1 Seal Water Return MOV SL-M-29 SL RC + 2'
- 88. RCP 2 Seal Water Return MOV SL-M-40 SL RC + 2'
- 89. RCP 3 Seal Water Return MOV SL-M-51 SL RC + 2' PIPING CABLE TRAY AND CONDUIT INSTRUMENT RACKS CONTROL ROOM CEILING Legend SYSTEM AFW Auxiliary Feedwater ASDA Alternate Shutdown Decay Heat Removal BAT Boric Acid Transfer CH Charging CS Containment Spray DG Diesel Generator Starting l
FO Fuel Oil HPSI High Pressure Safety Injection HV Area Heating and Ventilation PCC Primary Component Cooling PORY- Power Operated Relief Valve
, PPC Primary Pressure Control SCC Secondary Component Cooling SL Seal Water SPC Secondary Pressure Control SRV Safety Relief Valves SW Service Water 4
i 4-20 l - _ . . . - , _ - - - - - - - - - - - - - . - . - - - - - - - - - - - - - - - ---- - ---- - - - - - - - - -
Table 4.3-3 Maine Yankee Atomic Power Plant First Walkdown List of Equipment Components (Continued) l BUILDING AND
' SYSTEM ELEVATION EQUIPMENT ITEM 1
Legend (Continued ; BUILDING ' AB Turbine Building Auxiliary Bay AF Auxiliary Feed Pumphouse APR Appendix R Diesel CS Containment Spray Pumphouse CW Circulation Water Pumphouse
; FP Fire Pumphouse l PAB Primary Auxiliary Building -
- PT Pipe Tunnel PV Purge Air Valve Room RC Reactor Coolant RMC Reactor Motor Control Center Room SB Service Building TB Turbine Building VA Steam and Feed Water Valve Area YD Yard 4
k a
- I 4-21 ,
v Pinine and Valvine. For margin review levels up to 0.3g pga the Panel's guidelines recommend a sample review of accessible piping systems to verify that no problems exist, sah as inflexibility of piping runs between adjacent buildings. A sample piping system to be reviewed in detail was selected by mutual agreement between ; the system analysis and the fragility analysis teams prior to the first walkdown. l l . Additionally, piping as encountered during the course of the equipment component walkdown was reviewed, verifying no anomalies exist. A sample review of valves from the preliminary valve list was determined to be i the most effective way to confirm the Panel's guidelines regarding valve capacities. The selection of valves to be reviewed during the walkdown was based on a study j performed by Maine Yankee [Henrics et al., March 1986]. The study qualified the MYC critical valves using EQE's earthquake experience data for valves. Valves not enveloped by experience data were targeted for walkdown review. A ~ more descriptive discussion on the targeted valves occurs under Section 4.3.2.2, Walkdown Procedures for Valves. Cable Travs. Cable tray systems throughout the plant were targeted for general survey to determine if they could be screened out generically in conformance with i the review guidelines. This general survey was planned to be supplemented by a
- j. detailed inspection of a representative cable tray run, with the selection of this run j to be made during the walkdown.
Instrument Racks. A sample of instrument racks, although not specifically required for review by the Expert Panel guidelines, were targeted for walkdown j review. The instrument racks support critical system actuation instruments and components. Although many of the critical instrument racks are located inside i containment, a sample review was conducted on similar instrument racks encountered during the course of the equipment walkdown. l Control Room Ceilina. The control room suspended ceiling system was targeted for 1 walkdown since inspection is required by the panel's guidelines. { 4.3.2 Walkdown Procedures
! 4.3.2.1 Walkdown Team The fragility analysis team that conducted the plant walkdown consisted of engineers experienced iri structural and equipment fragility analysis, scismic analysis and design of nuclear power plants, assessment of actual earthquake experience, and seismic margin studies, and probabilistic risk assessments. The analysis team was divided into two groups as follows:
Ravindra (EQE) Hardy (EQE) Swan (EQE) 'i Hashimoto (EQE) Quilici (EI) Moore (EI) Prassinos (LLNL) Murray (LLNL) Griffin (EQE) As can be observed, each group consisted of fragility analysts and system analysts. The close interaction between the two aspects of the review was considered important. During the course of the walkdown, some of the members of the two groups were switched to confirm the findings of the other group and to ensure that l 2 l ] 4-22
.. _m _.. -..-_---.m. ~ , . , , . . , . _ , _ . . . _ . . . _ . - _ . , _ . . . , . . ..~,.----.-,_m., _ ,__,_mm_my.,_ ,_e-.. ..,,-_,.,.
certain items are not missed by either of the groups. At the end of each day, the groups met to comparc notes and to identify the areas and items to cover in the next day's walkdown. 4.3.2.2 Procedures for Structures and Equipment Structures Information necessary for seismic evaluation of civil structures is normally obtained from the design drawings rather than a walkdown. A complete set of drawings for the Maine Yankee structures was available prior to the~ first walkdown. These drawings were reviewed to obtain a general understanding of construction and configuration of the structures and to identify any specific data to be obtained during the walkdown. Walkdown of the targeted Group A civil structures was performed to determine the following: o Verify that the structures are in general conformance with the design drawings. o Identify any gross deficiencies that wo'uld imply a reduction in seismic capacity. o Confirm that structure separations indicated on the drawings were provided. o Obtain structural details not available from the drawings. The first two items above were obtained in a general manner, rather than performing a rigorous walkdown of all seismic load resisting structural members. The latter two items were specifically identified for inspection. Targeted structure separation gaps were determined before the walkdown and tabulated. Specific structural data not contained in the design drawings were obtained in the walkdown. For example, as-built sketches of weld and bolt details for certain structural steel connections were developed. Block Walls In preparation for the first walkdown, the following tasks were performed: o A target list of block walls in the plant was compiled using Maine Yankee summary tables, o Detailed walkdown data sheets were prepared to facilitate the compilation of block wall information. o Locations of block walls were highlighted on mechanical. layout or architectural drawings to permit wall identification during the walkdown. 4 23
I L F 6 L o Information, if already available, was entered into the walkdown i data sheets in advance of the wsikdown. For example, wall i t identification numbers and locations were recorded based upon-the Main.e Yankee summary tables. Also included were any available sketches of seismic retrofits. i Block walls identified on the target list were inspected to the extent possible. The following information was typically obtained and recorded on the walkdown data sheets: 1 o Location o Dimensions - o Boundary conditions o Seismic retrofits,if any o Other physical conditions (any cracking, gaps at boundaries, openings or penetrations) , o Identification of Group A components or lifelines directly attached or nearby
- These data were supplemented by photographs and sketches.
j During the walkdown, it was possible to identify several walls which obviously do ) not pose hazard to Group A components. Since the purpose of the walkdown was only to verify that their failure will not damage Group A components, the detailed { data listed above were not recorded for them. i Eauioment In general, preparation for the first walkdown began with a review of the available equipment data. This prereview was used to accomplish two goals: l 1. I To obtain as much familiarity with the equipment component as possible prior to the first walkdown.
- 2. Identify areas where additional details were required to assess component capacity and any possible low capacity items requiring a detailed review during the walkdown.
, Typical data reviewed prior to the first walkdown included a review of: I o Plant general layout drawings determining equipment location. j o Structural drawings determining equipment support and anchorage details. o If available, equipment vendor drawings or data to determine configuration, size, and material properties. 1 Equipment walkdown data sheets were developed for each class of equipment 4 l (reference Section 4.3.3 for examples of walkdown data sheets used). These data i sheets were used as checklists to verify details identified during the data review , 4-24 4
--a, ,--.- ,, . - ~ me-----,--~e+~ ~m-w a - -- - - < - .r ~ ,,--,--www--- ,- - - - - , - - - - - - - - - , - -,-w .-.c-----e------ -,.,r, + -.+~---w- ,- - . - - -- -
' - and to record walkdown inspection notes and details. To expedite and make efficient use of the time available during the walkdown, the data sheets were filled out as completely as possible prior to the first walkdown. l The majority of the equipment components identified by the system analysts were ! reviewed during the first walkdown. Components located in highly radioactive 4 areas (containment was not accessible at Maine Yankee) were not reviewed. Generically reviewed components included piping, valving, ducting, cable trays, , and instrument racks. The following provides specific procedures used for the
' review of different classes of equipment inspected during the walkdown. ! T.anka. Design drawings for the tanks and their foundations and or supports were available prior to the first walkdown. These drawings were reviewed to obtain a general understanding of the tank configurations and anchorage details.
Walkdown procedures for the tanks included the following: o Verification that the overall tank configuration and anchorage details conform with the design drawings. I o Review of piping and other attachments to identify any potential sources of damage due to seismic anchor point motion. o Inspect any unique features, which are not common to tanks, but were identified during review of the drawings. l An example of the latter item is the concrete enclosure surrounding the demineralized water storage tank (DWST). The enclosure was inspected to confirm the separation gap from the tank itself and to determine if the enclosure access room was scaled to prevent loss of tank contents to the environment in case of I tank failure. I
! Pumns. Historical performance during past earthquakes of horizontal and vertical pumps have shown HCLPF capacities greater than 0.3g acceleration levels ; (reference NUREG/CR-4334). The Panel's recommendations for horizontal and vertical pumps are that for a margin review level of 0.3g, a high HCLPF capacity exists. The walkdown procedures concentrated on verifying pump and motor anchorage, type of anchorage, foundation configuration and integrity, as well as any interaction potential from attached or adjacent components. This review ' aimed at a confirmation of the Panel's recommendations based on our judgment as well as documentation of the pump configuration and anchorage via the walkdown data sheets and photographs.
Heat Exchanners. Walkdown procedures for heat -exchangers concentrated on reviewing the supports, support saddles, anchorage details and interaction potential from attached or adjacent components. This is consistent with the Panel's recommendations for establishing heat exchanger capacity. Heat exchanger Internals were not reviewed as past PRAs have established their capacity to be ! greater than that of their supports or anchorage. Data sheets were used to record l' configuration and dimensional data from the walkdown for heat exchanger support, anchorage, and attached or adjacent component interaction potential l ) 4 l 4 25 l
. - - . - . - , , ---,..-,..--.-w.-m ...--.--.e-,,-, ---c -_-,..,-c-- +..m-, --.-:,--.-,-------ey-,~~-e.r. - - - ,-v--,-. , , , - - -r.v.,.---1 ,m
l-details that were not available from the plant _ data - reviewed prior to the walkdown. Diesel Generators. -The Panel's guidelines for diesel generators recommends a review similar to . that for pumps. Past. performance of diesel generators demonstrates HCLPF capacity levels of at least 0.5g pga. The Maine Yankee diesel generators were reviewed for anchorage and support integrity, noting if any vibration isolators were present, and the review of the peripherals for positive anchorage. The two major' peripherals reviewed during the walkdown were the engine control panel and the heat exchanger, both mounted on the diesel generator skid. Walkdown data sheets were developed and used during the walkdown to record any problem areas encountered. Photographs were also used to document items reviewed during the walkdown. Electrical Distribution Eauipment. The Panel's recommended walkdown procedures for a 0.33 pga review earthquake for electrical distribution equipment include reviewing that the cabinet or enclosures and internal instruments and components are positively anchored. Past performance of electrical distribution equipment during earthquakes suggests HCLPF capacities near 0.5g, providing the equipment . and internals, instruments, breakers, contactors, etc., are positively anchored. This supports the Panel's recommendation regarding the walkdown evaluation of electrical distribution equipment. It should be noted that relays have been specifically excluded from the Maine Yankee seismic margin study. The walkdown procedures at Maine Yankee concentrated on reviewing and collecting cabinet or enclosure anchorage details on the larger equipment items for a subsequent analytical review. The smaller (wall-mounted distribution cabinets) equipment items were reviewed for positive anchorage, but typically very few details were recorded. These smaller items were judged to have a HCLPF capacity of greater than 0.3g pga during the walkdown review. Internal component anchorage was inspected for positive attachment to the cabinet framing or cabinet walls for all electrical equipment components reviewed during the walkdown. Additionally, any interaction problems or seismic' deficiencies observed were recorded. Walkdown data sheets and photographs were used to record and document the walkdown findings. HVAC. The Panel's recommended walkdown procedures for a margin evaluation of HVAC equipment include reviewing the component for positive anchorage for a 0.3g pga review earthquake level acceleration _ margin level. Additionally, if a component is supported from vibration isolators, then an evaluation is required to establish lateral stability. The historical performance of HVAC equipment supports the Panel's recommendations. HVAC equipment positively anchored, as well as vibration isolator supported equipment having positive lateral restraints installed, have performed well during past earthquakes. The procedures for the walkdown review of the Maine Yankee HVAC equipment followed the Panel's recommended guidelines. Two procedures for reviewing the HVAC equipment were used. 1. For HVAC equipment found mounted on vibration isolators, a detailed walkdown review was performed. 4 26
L
- 2. For HV'AC equipment found positively anchored to a supporting structure, an engineering judgmental evaluation was performed during the walkdown review.-
The review for HVAC ' equipment mounted on vibration isolators. included recording the dimensional data and support configuration sufficient to perform an analytical evaluation after - the walkdown. Comprehensive data sheets were developed to record and document details and sketches. Photographs were also used to record the walkdown findings. The review of components in the second case included several-air intake and exhaust dsmpers, and exhaust fans. The.walkdown review assessed anchorage and any seismic deficiencies present in order to judge that the component had a HCLPF capacity greater than 0.3g pga. The predominant form of documentation for these components was the use of photographs to record the walkdown findings. Little details or notes were recorded on the walkdown date sheets. HVAC Ductina. The Panel's guidelines state that for HVAC ducting, a HCLPF 4 capacity exists for acceleration levels up to 0.3g pga. The walkdown procedures
- used to confirm the panel's guidelines consisted of two approaches:
- 1. Inspect the ducting in close proximity to the HVAC equipment j components to be reviewed.
; 2. Inspect a sample ducting system selected during the walkdown.
Inspection items included reviewing the vertical- and lateral-load-resisting members j of the ducting system and any possible anchor point displacements that could impart significant loads to connecting equipment. Documentation consisted of noting any anomalies and taking several photographs. Valves For review earthquake levels up to 0.33 pga the Panel's recommended j walkdown procedures require a review of valves encountered during the sample 1 piping system walkdown. Areas of concern to be reviewed during the walkdown
- include observing for interaction potential between the valve operator and adjacent
- structure or component and reviewing possible anchor point displacements between
- piping and valve. - Historical performance of valves during earthquakes at this 1 acceleration level supports the Panel's walkdown review recommendatious.
In addition to reviewing valves during the piping walkdown, a sample list of
- valves targeted for review was developed. The method used to select a valve for review was determined after a prior review of a Maine Yankee (MYC) document
[ Henries et al., March 1986], which evaluated the seismically critical valves using seismic experience data. The document recorded items such as valve description i and function, height above grade, cast iron body or yoke, pipe diameter, type of j operator, actuator weight, distance between the pipe center line and the top of the i operator, and the evaluation conclusions. The margin study valve list was i reviewed against the MYC document. The selected valves included those valves not ' previously shown to be within the bounds of the experience data or were not ! included in the document list. i 4 27
I + l The review of- the targeted valves was conducted during the course of two !
. walkdov 's. _ The review procedures con 3isted of inspecting each valve-on the list-i' _
accessibit to the walkdown teams. Seven valves targeted for review were located -
.in containment and could not be reviewed during the walkdown. For these valves, vendor literature and any previous photographs taken by Maine Yankee were requested for review- to confirm : the Panel's recommendations. Walkdown :
procedures consisted predominantly of inspecting the valve and area around it, , noting any potential interaction or anchor point displacement problems. l } Documentation consisted of taking several' photographs and in some instances l
- recording specific valve data on walkdown data sheets. This was limited to valves not included in the Maine Yankee document, because those valves' listed had this
{- information tabulated. Pioina. Past seismic PRA studies and earthquake experience data have shown that welded ' steel piping systems have a very high resistance to seismic loads. NUREG/CR-4334 contains the' Expert Pariel recommendation that " piping systems I l in nuclear power plants have HCLPF capacities greater than 0.5g pga." The panel recommends that for a 0.3g pga review level earthquake, a walkdown of a sample piping system should be conducted and that piping between buildings should be inspected. i Specific areas associated with the piping which were reviewed during the plant walkdown for Maine Yankee were: 4
; o Walkdown of the auxiliary feedwater system (AFW)
{ o Assessment of piping which spans between two buildings Underground piping systems could obviously not be addressed on the plant l walkdown, and were evaluated based on their design drawings. Two other piping failure modes that were addressed during the walkdown included the impacting. failures of valve operators and the damage of ' piping caused by the failure of anchorage of attached equipment. The valve clearance issue and the equipment-anchorage issue are addressed as a part of the margin evaluation for the specific equipment component and not as a part of the piping margin review. 1 The sample system to be walked down for the scismic margins review was the AFW system. The AFW system was chosen based on its importance to'the safety of the plant and because of the variety of piping sizes, supports, and components (branches, cibows, reducers, tee connections, etc.) in the system. The procedure for walking down the piping system included following the piping layout drawings to 3 verify support locations, assessing system interaction potential to the piping, and 4 taking detailed configuration information for piping details that are judged to bc
- of potential concern.
Cable Travs. In the first walkdown, inspection of the cable trays was performed at two levels: o General survey of cable tray systems in the plant l o Detailed inspection of a representative cable tray run 4 4 28 _ _ _ __ - _ _ _ _ _ ~ - . - _ _ _ _ . _ _ _ . . _ _ _ . _ - _ _ _ _ _ . _ _ .
l l-The general survey was performed to obtain an overview of cable tray construction throughout the plant. ' This included a review of the variety of cable tray system layouts, support configurations, and construction details. The inspection also considered items identified in'the review guidelines as being of potential concern, including failure.of taut . cables due to large relative displacement, severing of cables caused by sharp edges at the ends of cable trays, and weld failure. As a part of the walkdown of - cable tray systems, a representative run was inspected in detail to' obtain specific information on the construction. This run, located in the cable spreading room, was selected since it exhibited many of the~ features common to cable trays throughout the plant. Information was documented on walkdown data sheets. Information collected included run location and layout , supplemented with sketches, support configuration details and spacings, cable tray j configuration and loading, tray and support connection details, interfaces with the i structure and other components, and any potential problems. Instrument Racks. Instrument racks were not specifically addressed by the Panel > in establishing its margin review guidelines. Historical performance during i carthquakes of a variety of instrument rack configurations documented in the 1 experience data base suggests that a HCLPF capacity of greater than 0.33 pga j exists. Walkdown procedures of the Maine Yankee instrument racks consisted of i reviewing a sample of racks encountered during the course of the equipment walkdown. The reviewed racks were inspected for positive anchorage and ! similarity to documented data base racks. Attached instruments and components I were also inspected for positive anchorage to the rack. Walkdown documentation I consisted of taking several photographs. 1 i Several of the most critical instrument racks at Maine Yankee are located inside the containment building and could not be reviewed during either the first or the i second walkdown. Yankee Atomic engineers have photographic records of these instrument rack installations which were taken on previous plant outages. These j photographs were used as a basis for evaluating the seismic margin inherent in the
- instrument racks located in containment and the components (transmitters, i transducers, sensors, etc.) supported on these racks. The impulse lines and -
electrical leads which enter and exit these instrument racks were assessed by the systems analysts on the basis of multiple trains being separated from one another. Control Room Ceilina. Sketches provided prior to the first walkdown illustrated j design modifications incorporated in the suspended ceiling system over the control j room. These modifications consisted of safety wiring the T-bar ceiling and light
; fixtures to the concrete slab overhead. The ceiling system was-inspected during
' the walkdown to verify that the safety wiring was installed consistent with the I sketches and appeared adequate. In accordance with the Panel's guidelines, other fixtures above the control room were inspected to identify any other potential hazards to personnel and equipment below, i l 4.3.3 Walkdown Documentation Walkdown documentation for equipment and structures consisted of recording the findings using walkdown data sheets and photographs. The walkdown data sheets were developed for each particular class of component indicating specific t i } 4 29
_- - . - . -. - .- -~- . . . . - - . - - -. . - + 4 i < information required to confirm and verify the Panel's recommendations as well as y to record details sufficient to perform a seismic fragility evaluation if necessary. Typical examples of equipment data sheets used during the Maine Yankee
~
walkdowns for equipment are presented in Tables 4.3-4, 4.3-5, and 4.3-6 for pumps, HVAC components _and. block walls, respectively. The data sheets reflect the , j varying levels of information required between different classes of equipment { depending on their seismic ruggedness (e.g. pumps require little review other than i to verify anchorage and interaction potential whereas HVAC components supported j by vibration isolators require a detailed review recording greater- degrees -of
; information necessary for a fragility evaluation).
Photographs were also used to record details of the walkdown. Photographs i provide a permanent record of what was reviewed and support any notes or details
- taken during the walkdown. System interaction concerns are typically documented t with photographs. Additionally, photographs are used in the fragility evaluation to confirm details taken or to provide additional clarification. Photographs are a valuable part of the complete walkdown documentation.
i 4.3.4 Walkdown Results 4.3.4.1 Walkdown Findings j Structures ' i
! The original seismic design criteria for the Maine Yankee civil structures are l described in Section 3.2. With the exception of the 10 CFR 50, Appendix R diesel room, and the turbine / service building, all of the Group A structures listed in , Section 4.3.1 were categorized as Class I structures by the original design basis.
However, the following areas within the turbine / service building were considered
- Class 1: control room, cable room, switchgear room, service building area housing the control room air conditioning, breathing air, and switchgear room ventilating j equipment, diesel generator enclosure, and turbine building portion housing the t
component cooling heat exchangers, pumps and air compressor receivers. The Appendix R diesel system was subsequently deleted from the Group A systems by j the systems analyst. 1 Seismic load resisting systems for the Group A civil structures are composed of i reinforced concrete and/or structural steel. HCLPF. capacities were not established-I by the Expert Panel for steel frame structures. The following steel frame i structures are therefore screened in and require a seismic capacity evaluation:
- o Circulating water pumphouse, portion above El. 21'-0" l o Turbine / service buildinE, steel framed portions
- o Main steam valve house, interior steel structure
! During the first walkdown, one of the diagonal braces for the mainsteam valve j house steel structure was found to be missing. This was considered in the cvaluation of the structure HCLPF capacity. l The Expert Panel concluded that concrete containments, concrete shear walls, j diaphragms, and footings, special nonductile details, and impact between buildings l 4-30
TABLE 4.3-4 EXAMPLE PUMP WA'LKDOWN DATA SHEET i':: % ]/ ENGINEERING, PLANNFIG AND MANAGEMENT CONSULTANTS 8227-09 LLNL Seismic Margin Review BY DATE JOBNO JOB LLNL Plant Walkdown Data Sheet CHKI) DATE UEM SUBJECT FACILITY: Maine Yankee Atomic Power Station COMPONENT: LOCATION: Building - Elevation - 1.0 COMP 0tlENT DATA: Plant ID Number - Manufacturer - Model - Function - Photograph (overall) Roll No. Frame No.s 2.0 AREAS REQUIRING DETAILED REVIEW:
2.1 Anchorage
Number and size of anchor bolts - Type of anchor bolts - Description of foundation = Photograph Roll No. Frame No.s Note: Provide a sketch of anchorage plan with dimensions and indicate any foundation deficiencies observed in space provided below. 4-31 1 1
TA8LE4.3-4(CONTINUED) EXAMPLE PUNP #ALKDOWN DATA SHEET
+M "M ENGINEERING, PLANNING AND MANAGEMENT CONSULTANTS 2/
SHEETNO 8227-09 LLNL Seismic Margin Review SSNO joe BY. DATE CUENT L "L SUBJCCI Plant Walkdown Data Sheet CHKD DATE 3.0 ADDITIONAL COMMENTS OR OBSERVATIONS: Note any system interactions. i 1 4-32
TABLE 4.3-5 EXAMPLE HVAC COMPONENT WALK 00WN DATA SHEET
- Q -
ENGINEERING. PLANNING AND MANAGEMENT CONSULTANTS }/ g g ,8227-09 g LLNL Seismic Margin' Review By DATE CUENT LLNL SUBJECT Plant Walkrinwn nata theat CHKD DATE FACILITY: Maine Yankee Atomic Power Station COMPONENT: LOCATION: Building = Elevation = 1.0 COMPONENT DATA: Plant ID Number - Manufacturer - Model - Function - Photograph (overall) Roll No. Frame No.s 2.0 AREAS REQUIRING DETAILED REVIEW:
2.1 Housing
Overall dimensions = Anchorage: Type of anchorage (vibration isolators?) = Lateral restraints on isolators - Number and size of anchor bolts - Type of anchor bolts - Descriptionoffoundation/ supt.- Photograph Roll No. Frame No.s Note: Provide a sketch of anchorage plan with dimensions and indicate any foundation / support deficiencies observed in space provided below. If vibration isolated sketch lateral restraint. l 4-33
TABLE 4.3-5 (CONTINUED) EXAMPLE HVAC COMPONENT WALKDOWN DATA SHEET si E ENGINEERING, PLANNING AND MANAGEMENT CONSULTANTS g/
) a no 8227-09 a LLNL Seismic Maroin Review BY DATE CUENT l l NI SUBJECT plant usl yeun nsy ekeet CHKD DATE 3.0 ADDITIONAL COMMENTS OR OBSERVATIONS:
Note any system interactions. i 4-34
TABLE 4.'3-6 EXAMPLEBLOCKWALLDdTASHEET 'ciQ ENGINElmNG.RANNING AND MANAGEMENT CONSATANTs SHEET NO DATE JOB NO ?d - Z y2__17. i u m:e. Lb * . - s OY CHKD DATE CUEN! k ML SUS.1Cf LtMEdeon h b BLOCK WALL DATA SHEET Wall ID Number: Building: Floor: Location: Ref. Drawing Number: Film Roll Number: Frame Numbers: Wall Dimensions: Wx Hx T Lateral Supports: Any Visible Cracking? Any Gaps At Boundaries? Any Openings and/or Penetrations? Group A Components Directly Attached: Group A Components Nearby: Supported or Adjacent Lifelines: Group A Components or Lifelines Likely to Bo Damaged By Wall Failure: 4-35
TABLE 4.3-6 (CONTINUED) EXAMPLE BLOCK WALL DATA SHEET G' & [NGINEERING. PLANNING AND MANAGEMENI CONSULTANTS SHEEI NO. g No .9117 -Oin ilmi c. Vo- ;- x BY DATE CLIENT U UL ESJECT MAkdevh % % CHKD DATE BLOCK WALL DATA SHEET Wall ID Number: PLAN 4 f ELEVATION 4-36
4 should have HCLPF capacities greater than 0.3g. The Group A concrete structures were reviewed to confirm that this conclusion is appropriate for Maine Yankee. The containment structure consists of.a base mat, a cylindrical shell, and a hemispherical dome. The wall . is reinforced in a two-way pattern. The containment structure is comparable to other reinforced concrete containments
- evaluated in past fragility evaluations. In addition, HCLPF capacities have 4
already been developed in [Hashimoto et al.,1984] using approaches essentially the same as the fragility analysis and CDFM methods. Even with the concurrent ' effects of the design incident internal pressure, the HCLPF capacity was found to
-be equal to or greater than 1.0g.
The other reinforced ' concrete structures are typically composed of integral walls and slabs. A review of the design drawings was performed to verify that they are adequate to resist the review level earthquake. This review considered the ability 1 of the shear walls to transmit the overall seismic loads into the foundation, the ~ availability of local load paths .to deliver inertial loads to the. walls, and the
- presence of any nonductile detailing. The concrete structures were found to be comparable to other structures analyzed in past fragility evaluations. No inherent weaknesses in the seismic load resisting capabilities were noted. The Group A concrete structures were concluded to have HCLPF capacities greater than 0.3g and were therefore screened out.
A review of the structural drawings indicates that the buildings are typically separated by gaps of three inches or more. Readily accessible building separations were inspected during the first walkdown. The walkdown confirmed that these separations are present. Three-inch separation is greater than gaps provided for most nuclear plant structures. Since the Maine Yankee structures are typically of shear wall construction and founded on rock, lateral seismic displacements will be small. It is concluded that impact between buildings is very unlikely at the review carthquake level of 0.3g. Block Walls. The Maine Yankee block walls are typically unreinforced. A seismic evaluation of the block walls was conducted in response to I & E Bulletin 80-11. As a result of this evaluation, seismic retrofits were installed to increase the resistance of certain block walls supporting or adjacent to safety-related equipment. - 4 The review guidelines specify that a margin evaluation is required for these types j- of block walls. However, an evaluation should not be necessary for all walls in the plant. Walls that can be screened out include those that are not supporting or adjacent to Group A components, and those that can collapse but not cause damage to adjacent Group 'A components. To identify the. subset of block walls requiring a HCLPF calculation, the block wall evaluation procedure in Figure 4.3-1 was developed. Prior to the walkdown, block walls in the plant were identified and located. Based upon the walkdown, the walls were screened in or out depending on their potential hazard to Group A components. All walls supporting attached Group A components were screened in. For a wall adjacent ta(i.e., within about one wall height) Group A components, an assessment was made of the likelihood of damage to the components should the 4 -
l REVIEW AVAILABLE INFORMATION] IDENTIFY ALL BLOCK WALLS USING l MAINE YANKEE
SUMMARY
TABLE I LOCATE BLOCKWALLS ON EQUIPMENT LAYOUT DRAWINGS l WALKDOWN l l I I I I I WALLS SUPPORTING WALLS WHOSE WALLS SUPPORTING WALLS WHOSE OTHER GR0!'P A COMPONENTS COLLAPSE LIFELINES (I.E., COLLAPSE WALLS COULD RESULT CABLE TRAYS, COULD RESULT IN IMPACT PIPING,ETC.) IN IMPACT ONTO GROUP DNTO LIFELINES SCREENED A COMPONENTS (I.E., CABLE OUT TRAYS, PIPING, ETC.) l SCREENED IN l WILL IMPACT ARE THESE ARE THESE CAUSE DAMAGE LIFELINES LIFELINES TO GROUP A PART OF GROUP PART OF GROUP COMPONENTS? A SYSTEMS? A SYSTEMS? BY SYSTEMS BY SYSTEMS i ANALYST ANALYST I 4 Y N Y i i N l SCREENED INI l SCREENED OUTl WILL IMPACT [SCREENEDOUTl CAUSE DAMAGE TO GROUP A LIFELINES? , Y I I N ISCREENED INl l SCREENED OUTl Y I I M l SCREENED INl l SCREENED OUT PERFORM FRAGILITY ANALYSIS TO CALCULATE HCLPF Figure 4.3-1 Block Wall Evaluation Procedure 4-38
wall collapse. Examples of walls screened out at this point include walls shielded from adjacent components by built up steel framing and walls adjacent to components judged to be capable of withstanding the impact without damage. 1 Block walls affecting lifelines such as piping and cable trays were screened in the same manner. If necessary, confirmation of the Group A status of these lifelines was provided by the systems analyst. All other block walls not supporting or adjacent to Group A components were screened out. Block walls affecting safety . related equipment that were not walked down are assumed to have seismic ! capacities comparable to the Group A walls. Review of the block walls inside of containment was conducted 'on the basis of architectural and design modification drawings. Dams. Dikes. and Levees. Walkdown of the fire water pond indicated that grading around the pond would cause the water to flow away from the plant even if the surrounding dike should fait during an earthquake. The dike was therefore screened out since its failure would not have any impact on plant safety systems. Eauioment Tanks Except for the buried diesel fuel oil storage tanks, TK-28A and B, all the tanks identified on the Maine Yankee margin review equipment list required a walkdown review and capacity evaluation. The tanks reviewed during the walkdown were all found to be anchored (i.e., no unanchored tanks were observed). The majority of tanks observed were consistent with the design drawings and details reviewed prior to the walkdown. Few anomalies affecting tank capacity
! were found during the course of the walkdown. The results of specific tank l findings are discussed in greater detail below. l The auxiliary fuel oil supply tanks, TK-28A and B, were reviewed during the walkdown. The fuel oil supply tanks are buried with only the upper portion of the tank visible; consequently, only the visible connecting piping was reviewed for (
possible failure caused by large relative displacements of the ground surrounding ' the buried tanks. The tank fill lines were the only observable piping. The fill l lines are attached to the top of the tank; consequently, if a piping failure occurred , a loss of diesel oil would not result. The emergency diesel day tanks, TK-62A and B, were reviewed in detail during the walkdown. The review found the tank elevated and welded to a braced steel frame. The support frame was bolted to the concrete floor with two expansion ) anchor bolts per column. Under each column was observed to be a shim plate and 3 ' approximately one inch of grout. This installation detail varied between the four columns due to the sloping floor with the distance between the top of concrete and bottom of column base plate varying from 1.75 in to 2.5 in. This observation brought up the concern of how much of the anchor bolt was embedded in the concrete floor. To verify bolt embedment, Maine Yankee plant personnel ultrasonic tested (UT) each of the anchor- bolts. The UT inspection confirmed that the bolt embedment did not extend into the concrete floor. Maine Yankee subsequently modified the day tank anchorage by installing new anchor bolts which are fully embedded into the concrete floor slab (Section 4.4 describes the modification). 4-39
I I PumDS. The critical pumps identified on the margin equipment list were reviewed during the walkdown. For a 0.33 pga review level earthquake the Panel requires a review of the pump anchorage and any potential interaction problems. Horizontal pumps at Maine Yankee were determinui to be well anchored on the basis of the first - walkdown. Piping anchor point displacement and interaction potential concerns were reviewed for each pump with no areas of concern observed. Maine 4 Yankee horizontal pumps were subsequently screened out with a HCLPF capacity judged to be greater than 0.3g pga. The Expert Panel guidelines recommend a capacity evaluation for vertical pumps if the shaft length is cantilevered greater than 20 ft. There were three sets of vertical pumps identified on the Maine Yankee margin review equipment list: o Service Water Pumps o Containment Spray Pumps i o Auxiliary Fuel Oil Transfer Pumps The Maine Yankee. service water pumps were identified from the initial drawing and vendor data review as vertical pumps with long cantilevered shafts greater than 20 ft in length. The service water pumps shaft base support could not be reviewed during the plant walkdown as required by the Panel's recommendations; however, the vendor drawings identified the pumps'to have 26 ft shaft lengths with pin supports at the base. Consequently, with respect to this issue, the service water pumps were screened out per the Panel's guidelines. The two other sets of vertical pumps identified on the equipment list for review were observed to have shaft lengths less than 20 ft in length. These pumps were subsequently screened out per the Panel's guidelines. Two vertical pumps (service water and containment spray pumps) were observed during the walkdown to have a possible weak connection between the motor and pump interface due to a small number of bolts and smali bolt circle dimension. Dimensional data and motor name plate data was recorded in order perform a capacity calculation for these pumps. Heat Exchanners. All heat exchangers identified on the margin review equipment list were reviewed during the walkdowns except for the reactor containment air recirculation coolers, the reactor coolant regenerative heat exchanger, the seal water heat exchanger and seal water heater due to inaccessibility (high radiation areas). Per the Panel's guidelines the exchangers were reviewed for support and anchorage integrity as well as any potential interaction or seismic anchor point displacement problems. The supports and anchorage require an evaluation to show ' capacity greater then 0.3g pga. The following briefly describes the types of heat exchangers reviewed at Maine Yankee and the walkdown findings. The heat exchangers reviewed at Maine Yankee can be classified into four categories.
- 1. Vertically oriented heat exchangers.
l 4-40 l . _ _ _ _ _
- .-. ._ -~ . -_ .
l
- 2. Horizontally oriented heat exchangers supported from a concrete-pier or steel support frame.
- 3. Small heat exchangers mounted directly to a larger supporting component.
- 4. Heat exchangers not accessible during the walkdown due to high local radiation levels.
The only vertical heat exchangers reviewed were the residual heat removal (RHR) heat exchangers which were observed to be supported from anchor lugs at - approximately the midpoint of the exchanger. shell. . Additionally, bottom lateral supports in both horizontal directions were observed. An analytical evaluation of the supports and anchorage was performed. f The standard horizontal heat exchangers were all observed to have an adequate support system (concrete pier or braced steel frame). All were observed to be well anchored with no anomalies noticed. An analytical evaluation of the supports and anchorage was performed. 1 The small skid-mounted heat exchangers included as a peripheral component of the component to which it was attached. These heat exchangers were observed to typically provide a bearing cooling function, and varied from small diameter
~
(approximately 5 to 6 inches) cylindrical units to just helical coils attached to the side of the pump casing. Capacity for these exchangers was judged to be greater than 0.3g pga during the walkdown. The heat exchangers not accessible for the walkdown were evaluated via a drawing and vendor data review with the support from photographs previously taken by Maine Yankee engineers. The reactor containment air- recirculation coolers were the only component that fell into this category requiring an evaluation. The - reactor coolant regenerative heat exchanger, the seal water heat exchanger, and seal water heater did not require a review, because valves PCC-M-90 and PCC-M-219 have a HCLPF capacity greater than 0.3g pga, which will isolate the PCC system if these components fail. , Diesel Generators. The walkdown findings verified the Panel's recommendations l that a high capacity exists for diesel generators at the 0.3g pga review level earthquake . The diesel generator skid assembly rests on a large grout pa'd which is used for leveling purposes. The anchors for the skid consist of large "J" bolts ' embedded well into the concrete foundation. These i 1/4-inch-diameter "J" bolts
- are judged to have a very high seismic capacity and the effcct of these anchors passing through the grout pad is felt to be minimal. The diesel generator peripheral components which were mounted on the skid were reviewed for structural integrity and observed to be well anchored. No potential interaction j problems were observed from surrounding components. The diesel generators and peripheral components were subsequently screened out.
Electrical Distribution Eauioment. In general the electrical equipment reviewed for the margins -program were observed consistent with the Panel's guidelines stating that active electrical equipment can survive ground accelerat. ions up to 0.5g l 4 41
l
. pga, provided that the cabinets and instruments are anchored. The equipment
- reviewed at Maine Yankee was well anchored except for one anomaly that .
- occurred with the station service transformers which is described below. Most of the critical electrical equipment at Maine Yankee had seismic anchorage upgrades installed during one of several previous ' seismic evaluations conducted. by the i Utility. Upgraded anchorage and supports were evident on the switchgear, motor control centers, inverters, distribution cabinets, main control board, and battery chargers. The instruments and internal components of most electrical cabinetry were all observed to be well anchored. However, one noncritical component in the main control board was observed unanchored. This component has been subsequently anchored by Maine Yankee (component feviewed during the second walkdown was observed positively anchored). The fellowing discusses several walkdown findings with regard to particular types of electrical equipment !
j reviewed at Maine Yankee. The station battery cells at Maine Yankee were observed to be of lead-antimony flat-plate construction. Data on the performance of lead-antimony batteries during earthquakes is lacking at the present time. However, extreme corrosion degraded-electrical performance has been recorded for aged cells such that a HCLPF
- capacity cannot be confidently established. The Expert Panel (NUREG/CR-4334) 4 suggests that battery cells are not vulnerable at a. 0.3g earthquake and can be screened out. EQE feels that this recommendation has not been proven to be reasonable for all emergency batteries in nuclear power plants. Maine Yankee batteries proved to be a case where further examination / review was necessary.
Section 4.4, Maine Yankee Component Modifications, discusses this finding further. The inverters were one of the components which had an upgraded anchorage from. its originally installed condition. The anchorage modifications were somewhat unorthodox due to adjacent equipment constraining the placement of the anchorage addition. Details were recorded and an anchorage capacity evaluation was performed. Additionally, the back shear panel of the cabinet had ventilated cutouts which raised some concern about the lateral load resistance of the cabinet frame. After the front doors were opened it was observed that the transformer was bolted to the bottom base framing with few component attachments occurring above the base framing level. The station service transformers enclosure framing was observed to be well anchored and also had an anchorage upgrade installed. However, when the - internal core / coil assembly was reviewed it was found to be unanchored. The core / coil assembly was observed to be supported by rubber isolators in four places . with no vertical uplift or lateral restraints installed. The transformer internal
- core / coil assembly has been subsequently identified as requiring - additional anchorage to be installed during the next refueling outage (March 1987). Section 4.4, Maine Yankee Component Modifications, describes the modifications which are scheduled to be installed.
Numerous wall-mounted panels were reviewed during the course of the walkdown. All of these panels were observed to be light in weight and well anchored with a minimum of four bolts per panel, one at each corner. The HCLPF capacities for the wall mounted panels were determined to be greater than 0.3g pga review level 4-42
~_____ _ -- _. .
l carthquake using the methods described under Section 5.2, Simplified Analysis and Use of Screening Tools. HVAC Components. The walkdown review consisted of inspecting several fan units, air conditioner units and dampers. Except for the diesel generator exhaust fans, all the fan units and air conditioner units were observed to be supported from vibration isolators; consequently, per the Panel's guidelines the fans and air conditioners required an evaluation to establish lateral stability. An initial evaluation determined the lateral load capacity to be deficient for both items and Maine Yankee subsequently scheduled both items to be upgraded during the March 1987 refueling outage. Section 4.4 Maine Yankee Component Modifications describes these modifications. l The diesel generator exhaust fans and the diesel generator air intake and exhaust dampers were observed to be well anchored into the building walls. A capacity evaluation based on the walkdown inspection judged the fans and dampers to have a HCLPF capacity greater than 0.3g pga, thus were subsequently screened out. Overhead HVAC ducting in the Maine Yankee plant was observed to be well supported with lateral bracing in both horizontal directions. The vertical and lateral supports were found to be threaded rods anchored into threaded inserts in the concrete ceiling and adjacent walls. The ducting reviewe,d at each component was observed to have flexible joints at the interface connections. Consequently, lateral movement of the ducting will not impart significant loads to the connected
! component. Maine Yankee ducting was judged to have a HCLPF capacity greater than 0.33 pga, thus was subsequently screened out.
Valves. The sample list of valves reviewed during the walkdown were all observed , to be seismically adequate such that a high confidence exits that the valves i addressed in the Maine Yankee margin study will survive a 0.3g pga margin earthquake. These findings agree with the Panel's guidelines on valve performance in a 0.3g pga margin earthquake. The walkdown findings of the sample list of valves is discussed below for specific types of valves reviewed. The power operated relief valves (PORV's) could not be reviewed during the walkdown due to inaccessibility (high local radiation). This included several other valves as well (Reference Table 7.1-3 in Chapter 7 for the capacity evaluation of the valves). However, vendor data and photographs were provided by Maine Yankee to aid in EQE's review of the valves. The vendor data illustrated the construction and operation of the PORV's. The function is essentially a solenoid operator cantilevered from the valve body, not unlike a standard MOV, actuating the valve upon the presence of an operation signal. The Maine Yankee PORV is i quite similar to the PORV's located at the El Centro Steam Plant which was subjected to the 1979 Imperial Valley earthquake. Figures 4.3-2 and 4.3-3 show a comparison of the Maine Yankee PORV's and those from the El Centro Steam Plant. There were several valves reviewed which had long " reach rods" between the operator and the valve body. The configuration of these valves is such that i operators can operate the valves from a safe distance and not be exposed to dangerous levels of radiation. The valve operator was well anchored to a pedestal i 4-43
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1 1 while the oper'ator rod was connected to the operator with a universal joint. - The operator rod was then routed through the _ floor' supported .at regular spacings before connecting to the valve body. At _each support point and at the valve body ~ l
- -- was a universal joint. From the _walkdown review it was judged that these valves 4
have a HCLPF -: capa' city greater .than the 0.3g pga review level earthquake. Earthqu'ake experience data supports our judgment' on these -" reach rod" valves. Valves similar to the long " reach rod" valves were found at several Coalinga area-sites subjected to the 1983 earthquake. - Figures 4.3-4 and 4.3-5 show a comparison 1 of the Maine Yankee valves and a sample of those from the Coalinga sites.'
.l j
Several Maine Yankee valves fell out of the'. bounds of the: earthquake _ experience data based on conservative estimates = of ' operator weight and . operator height.
- These' valves were._ reviewed.and found to be-' reasonably close to the experience data bounds. . Seismic anchor point displacements and. interaction potential.were.
, assessed for all . valves addressed on the walkdown. No problems were observed and - the valves subsequently judged to have a HCLPF capacity of greater than 0.3g pga.
' Table 7-3 in Chapter 7 identifies the Maine Yankee valves and also indicates which valves were judged to have a HCLPF capacity' greater than 0.3g pga compared to those where experience data supported the evaluation.
~
Pioina. The auxiliary feed water -system (AFW) was 'the sample piping system selected for a detailed walkdown review. The AFW extends from the Containment penetration to the demineralized water storage tank. I Three Maine Yankee drawings were used (verification of support' types, support' spacing, unsupported spans, etc.) during the walkdown to aid in the walkdown:
.o Drawing 12365.10-MKS-103RI-4 shows the piping . from the l containment penetration to anchor H-102
- o l>rawing 12365.10-MKS-103N1-4 shows the piping from anchpr H- -
102 to the discharge side of the emergency feedwater pump (P-25A)
- o Drawing 12365.10-MKS-103N1-4 shows the piping from the suction side of ' Pump 25A to the point 'where the piping goes underground to the DWST.
The'walkdown of the AFW piping -produced nothing which altered the Panel's assessment of a high confidence exits that the piping will survive a 0.3g pga review
- level carthquake. The piping was observed to be of all welded steel construction with standard fittings (tees, cibows, and branch connections). The supports were i
typically welded steel frames constructed of angles or boxbeams and attached to , the walls /cciling/ floors by welding to embedded or structural . steel, and on occasion with expansion anchors. Interaction with surrounding components was assessed throughout the AFW piping system, but no possible damage scenarios we're found. Piping systems which span between two structures were also found to have high capacities. Maine Yankee is a rock site, thus very little relative motion between f 4-46
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Valves with Long Operator Arms. 4-47
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the foundations of different. structures would be expected to occur in a 0.3g pga earthquake. The areas which are of a possible concern for a plant on a rock site are places where short spans of piping traverse two buildings at a location which will experience large relative displacements. This could occur high up in a relatively flexible structure. Maine Yankee has very few piping systems which
-traverse two structures at elevations above grade. The most critical location at Maine Yankee exists for the steam lines running between the steam and feedwater valve area and tiic turbine building (Figure 4.3-6). These interbuilding pipes were
, judged to have a high capacity for two reasons: o The length of piping is long relative to the diameter of the piping. The relative building displacements required to fail the piping would have to be very large given the distance between buildings and the enormous ductility which has been demonstrated for welded steel piping, o The steam and feed valve area structure is a very stiff concrete structure and will have a small displacement at the.mid-height location of these steam pipes. Displacements of the turbine building on the other end of the pipe will be limited by the stiffer control building. The cast iron service water header from the service water pumps to the PCC and SCC Heat Exchangers was found to be predominately buried. The only exposed J portion of the piping was from Pump Discharge to ground penetration in the Pumphouse. The exposed portion of the cast iron service water piping was retrofitted with large scismic supports which were anchored back to the pumphouse structure. Short span lengths and adequate bracing were observed during the walkdown, such that capacity was judged greater than the margin earthquake level of 0.3g. The buried portion of the service water header could not be inspected; however, Maine Yankee calculations were reviewed with the subsequent capacity of the buried service water header determined to be greater than 0.3g. Cable Trays and Cablina. Under the Panel's guidelines,~ cable trays and cabling are , screened out for review carthquake levels less than 0.3g pga. The Expert Panel i recommended that example cable trays be inspected to verify ' that they are
- adequately anchored and braced and to confirm that taut cables will not be affected by anticipated relative displacement or any sharp edges at the ends of the trays.
Cable tray walkdown effort was focused on the following areas which -contain much of the cable trays in the plant: o Turbine / service building Cable vault, El. 21'-0" Cable tray room, El. 35' 0' Cable tray area, El. 35'-0* Switchgear room, El. 45'-6" Turbine hall, El. 21'-0" j o Circulating water pumphouse o Primary auxiliary building, El. 21'-0"
- 4-49
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o M.C.C. room M.C.C. rooms,~ El. 21'-0" and El. 33'-4" Containment penetration area, El. 46'-0" i Ventilation equipment area l o. 7 . i Cable trays in other areas are generally similar to cable trays in the areas listed above. J Typical Maine Yankee cable trays are shown in Figure 4.3-7. The following general description of the Maine Yankee cable trays. is obtained from the ~ walkdowns:
~
Cables are typically contained in 24-inch-wide aluminum, ladder
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o type cable trays. I o Maximum cable tray loading was found in the service building cable vault where inserts were installed within the trays to permit i cable fill somewhat in excess of 100% relative to the tray itself. J
; o Cable tray supports are typically unbraced rod hung trapezes,-
screwed into Phillips Redhead concrete inserts. Braced cantilever bracket floor to ceiling column supports are used in the cable
^
4 vault. Wall-mounted cantilever brackets were found at a few locations in the plant. j 1 o Support spacings are six feet or less. j o Supports carry a maximum of six tiers of cabic trays, ! Cable tray: systems similar to those found at Maine Yankee have been subjected to ; shake table testing and actual carthquakes. The test programs have shown that i cable tray systems are capable of withstanding significant seismic input levels - l l withcut gross damage that would compromise- cable integrity.' ' [EQE 'Inc.,1986] ! I
' presents a summary of. the perfonnance of cable tray and conduit systems in past -
earthquakes. Experience data shows that ~ cable tray and . conduit systems ~ constructed according to normal industrial standards have a large capacity for the - l absorption of seismic inertial loads. There have only been a few instances of local i damage to cable tray systems due to actual earthquakes. These occurrences did not compromise the structural integrity of the cable tray systems or the integrity = or
- function of the supported cables. The single instance of - cable tray collapse involved an anomalous support configuration not observed at Maine Yankee.
To confirm that the Maine Yankee cable trays can be screened out at the 0.3g pga review carthquake level, they were compared to cable trays in the earthquake experience data base [EQE Inc.,1986). This comparison was performed for a number of parameters important to cable tray seismic adequacy, including the - following associated with seismic input, seismic response, and system integrity: ! o Input Parameters l
- Peak ground acceleration
- Duration of strong ground motion ,
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- Frequency content of ground motion
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-- Elevation in building l 0 Response Parameters
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- . Cable tray loading
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The Maine Yankee cable trays were found to be enveloped by the experience data base for these parameters,' thereby demonstrating that they can be screened out for
.the 0.3g pga review earthquake.
Potential concerns identified by the Expert Panel include failure of taut cables due to large relative displacements, seveting of-cables caused by sharp edges atLthe ends of cable trays, and failure of welds. Review of the Maine Yankee cable trays did not identify any of these conditions. Based upon a comparison of the Maine Yankee cable trays with cable trays in the earthquake experience data base [EQE Inc.,1986], and walkdown of the systems for particular problem areas, it is concluded that the cable trays and cabling can be
- screened out for the review earthquake level.
1 Instrument Racks. The sample of instrument racks reviewed during the course of : the walkdown were all judged to have a HCLPF capacity of greater than the 0.3g pga review level earthquake. The racks reviewed were observed to be similar to rack configurations documented in the experience data base. The racks were constructed of structural angle with bracing near the base well anchored to the 4 concrete floor or walls with expansion anchors. The rack member connections were of all welded construction. The instruments and attached components were i all observed to be well anchored to the rack face plate. Typical Maine Yankee instrument racks are shown in Figure 4.3-8. i i Control Room Ceilina. Control room ceilings are screened out for review earthquake levels less than 0.33 pga. This is subject to verification that the ceilings are adequately braced and that other overhead fixtures are properly anchored. i i l l 4-53
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/ Component Attachments. 4-54
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The suspended ceiling system over the control room is the conventional T-bar type with fibrous acoustic panels typically 2'-0" square. The T-bars are safety wice,d to the concrete slab above. The safety wiring is judged to be sufficient to prevent collapse of the T-bar ceiling to the floor below. The acoustic panels are not positively attached to the T-bars. Suspended ceiling panels have been dislodged in past carthquakes. However, this has not been a source of damage to control panels or injury to operators. Furthermore, attempts to push the Maine Yankee ceiling panels out met with significant resistance due to the presence'of the safety wire loops at the T-bar intersections. Damage resulting from falling ceiling panels is
. considered unlikely.
The control room and battery room light fixtures are made of sheet metai ahd m suspended from the slab above. They are safety wired to the slab similar to the T-bar ceiling (Figure 4.3-9). This should be sufficient to prevent the light fixtures from dropping due to the 0.3g review level carthquake. It may be possible. for translucent panels and light tubes to become dislodged during an earthquake. However, as with the ceiling panels, the likelihood of damage to equipment and 3 personnel below is small. Other ceiling fixtures above the control room include HVAC ducting and con'luit. As with the rest of the plant, the ducting is typically rod hung and braced by tods attached to the ceiling by concrete inserts. . Conduit above the control roon: is typically about two inches in diameter or less. It is supported by threaded rod hangers anchored to the ceiling by concrete inserts, similar to the cable trays. Ducting .and conduit are judged to be adequate for the review level earthquake of 0.3g. In conclusion, review of the control room ceiling and other overhead fixtures confirmed that they could be screened out. 1 4.3.4.2 Plant Unique Features During the review and walkdown of the plant, particular attention was paid to identify any unique features. The unique features are defined as the~ ones that either have proved to be important contributors in the past seismic PRAs or have not been considered in the previous PRAs. The example of the first kind is the Jocasse dam at Oconee. At Maine Yankee, the earth dike enclosing the fire water pond was examined for potential safety significance to the plant. It was found that terrain around the plant is such that the water from the pond would flow f away from the plant in case of a dike failure. Other features that were not addressed in the previous seismic PRAs and that should be examined in future seismic margin studies as a result of this study are: o Mussel pump o Steel structures o Lead-antimony batteries o Anchorage of the transformer core / coil assembly 1 4-56 1 l w- , m, - - - , , , < r- , ._.-, , - ,-,--.w -
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4 57
Q 18" SOLID CONC. BLOCK WALL 12'-11 1/2" (REF.) X-
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The report by Energy Incorporated (Vol.2) discusses the significance of mussel pump and where similar situations could occur in other plants. The Panel report should include some discussions and guidance on reviewing steel structures and about their seismic capacities as discussed in Chapter 6. The topics of lead-antimony batteries and transformers are covered in later sections. 4.4 Maine Yankee Comoonent Modifications Several Maine Yankee components were identified during the margin evaluation as potentially having a relatively low capacity. These components included critical equipment items identified as part of the margin review and noncritical items that posed a possible interaction hazard to critical equipment. During the course of the margin review Maine Yankee took upon itself - to modify those components identified -as having a potentially low capacity. The installation of the modifications occurred in two stages. Modifications which could be installed without affecting plant operation were undertaken. Modifications which would be I disruptive and pose a risk to plant operation and maintenance personnel were scheduled for the March 1987 refueling outage. The modifications completed before the refueling outage were verified during the second walkdown; however, a third walkdown during or after the March 1987 outage is necessary to verify the modifications scheduled to be completed during that time period (reference Section 6.4). The specific component modifications described below are reflected in the structures, equipment and plant capacities presented in Section 5. Structures Block Wall VE 21-1. Block Wall VE 21-1 is located in the ventilation equipment room. This wall is adjacent to the containment spray fans, Fans 44A and 44B, and , their ducting and filter (Figure 4.4-1). It is freestanding and built integral with an intersecting wall at one side. Structural drawings show this wall to be lg in. thick by 10 ft 0 in. high, and constructed of solid concrete block. Failure of Wall VE 21-1 due to out-of-plane seismic response could result in
; damage to the adjacent ducting and filter, thus causing the fans to draw air from the ventilation equipment room itself rather than the containment spray pumphouse. This wall may be particularly vulnerable to seismic effects since it is freestanding. Conceptual sketches for seismic retrofits call for structural steel restraints to be added at the top and bottom. Top supports will consist of steel angles spanning back to the concrete wall of the containment spray pumphouse j (Figure 4.4 2). Calculations based upon the preliminary retrofit sketches provided indicate that the modifications should be sufficient to provide the wall with a HCLPF capacity greater than 0.3g.
Eauioment j The Emernency Diesel Day Tanks TK-62 A & B. The day tanks were observed
! during the walkdown to be elevated tanks supported from a braced steel frame.
1 Each column of the frame was observed anchored to the concrete floor with two expansion anchor studs with provisions for the installation of two additional studs. Under each column was a leveling plate and grout pad which raised concern about a j 4-59
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; Fi8ure 4.4-4 Photographs of the Maine Yankee Station Service Transformer l l
Internal Core / Coil Assembly Before the Anchorage Modification. I 4-62 l i l
5/8"4 JACK STUD TRANSFORMER CORE & OIL ASSEMBLY NEW NUT AND WASHER _, EXISTING 3/4"f SHIPPING STUD NEW 3/8" THICK BY 4"4 OR SQ ' PLATE WASHER \ EXISTING 4'x1' CROSS BAR S b. E NEW 1/2" THICK x 4" sq SHIM PLATE N ' 6,9 . ' g ; I r/////s/I ' A _ _ _ _ _ ' ~~~F/_ _ i _ _g ,7 7
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- 1/8" (TYP. )
y (/ ,, h F 3/16 (TYP.) TOP & BOTTm f FLOOR l MAY BE SIMILAR 1 13/16" g -O TO SECTIOu C-C-NEW L 5" ' SEW ION 0-D SECTION B-B Figure 4.4-7 Proposed Anchorage Modification for the Maine Yankee Air - Conditioners. 4-65:
scheduled the anchorage modification to be installed during the March 1987 refueling outage. The transformer and plant HCLPF capacities represent the modified condition. The station service transformers should also be inspected during the third walkdown to verify the modification has been installed. HVAC Comouter Room and Lab Air Conditioners. The air conditioners reviewed during the walkdown were found to be supported from vibration isolators. Conversations with the isolator manufacturer determined the isolators to have no vertical uplift capacity and marginal lateral load resisting capacity. Maine Yankee developed a modification that would provide vertical and lateral load capacity to be installed during the March 1987 refueling outage. Sketches of the modification were provided to EQE for the capacity evaluation of these components. Figures 4.4-6 and 4.4-7 show photographs of one air conditioner and a sketch of the proposed modification provided to. EQE, respectively. The HCLPF capacities of air conditioners represent this modified condition. The air conditioners should also be inspected during the third walkdown to verify the modification has been installed in accordance with the proposed design provided to EQE. HVAC Containment Sorav Fans. The fans were observed during the walkdown to be supported from vibration isolators. The configuration of the isolators was such that vertical uplift was provided. Lateral stability of the isolator configuration was identified as marginal in resisting the seismic loadings. The existing containment spray fan isolator assembly was scrutinized by both the fragility analysts and by the peer review group members. The earthquake experience data base was not useful as a resource for this particular isolator configuration due to a lack of similarity with the data base isolator systems. A rigorous fragility analysis of the cold formed steel attachment strip and the attachment bolt was deemed to be nearly impossible to complete due to low cycle fatigue, lack of ductility, and stress concentration issues associated with the attachment strip. Maine Yankee engineers agreed to design an anchorage upgrade based on the good engineering judgment that a modification was necessary in order to ensure that the fan could withstand (with high confidence) the 0.3g review level carthquake. Maine Yankee provided EQE with a sketch of the proposed modification for the margin evaluation. Figures 4.4-8 and 4.4-9 show photographs of the fans and a sketch of the proposed modification provided to EQE, respectively. The fans and plant HCLPF capacities represent this modified condition. The fans should also be inspected during the third walkdown to verify that the modification has been installed. Miscellaneous Comoonents. Miscellaneous components include noncritical components which were identified during the walkdown to posses a potential risk to plant safety by failing and impacting a critical equipment component. All of these components fall into the general category of system interaction which the Expert Panel identifies in its review guidelines. These components included the alarm message display discussed above, emergency lighting units throughout the ,
- plant, compressed gas boilers, and a welding cart on wheels near the containment spray fans. Figure 4.4-10 shows photographs of the emergency lighting units and the welding cart. All components that were identified during the first or second I walkdown have been subsequently modified, secured, or removed by Maine Yankee. These modified components were evaluated during the second walkdown (or based on verbal communication by Maine Yankee engineers in the case of the welding cart) and judged to have a HCLPF capacity greater than the 0.3g pga.
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Figure 4.4-8 Maine Yankee Containment Spray Fans Before the Installation of the Anchorage Modification. 4-67
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SECTION 6 B Figure 4.4-9 Proposed Anchorage Modification for the hon Yankee Containment Spray Fans. 4-68
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Figure 4.4-10 Photographs of the Unstrapped Emergency Lighting Units in the Control Room and Welding Cart Near the Containment Spray Fans. 4-69
4 J0 dJA ,# r #3 - I 4' J r C h l m 4 v b pl d b t i i I I l J 4-
i-CHAPTER 5 EVALUATION OF SEISMIC CAPACITIES OF COMPONENTS AND PLANT 5.1 Review of Structural Mods]1 In-structure response spectra for Maine Yankee were generated by Cygna specifically for use in the trial plant application. Input consisted of acceleration time-histories matching .c median NUREG/CR-0098 ground response spectrum. The dynamic analyses were performed with structural models recently developed for use in other Maine Yankee analyses. These models were reviewed to verify that they are adequate to predict responses for the review level earthquake. The floor response spectra generated by Cygna were judged to be adequate for this margin review. Since seismic analyses of structures and components developed for the original design were generally not used to calculate HCLPF capacities, the original design structure dynamic models were not obtained or reviewed. [Cygna,1982] describes the dynamic models generated for the following structures: o Containment structure o Containment internal structure o Containment spray pumphouse o Main steam valve house o Primary auxiliary building o Turbine / service building This report contains the following information: o Computer programs used o General modeling assumptions ' o Description of individual structure modeling considerations l 0 Sketches of seismic load resisting elements l o Overall structure mass properties o Calculated frequencies, modal participation factors, and mode shapes This information was reviewed to verify the adequacy of the dynamic models in the following manner:
- 1. Review overall approach and assumptions.
- 2. Review individual structure models.
2.1 Review general model layout. 2.2 Verify that the major load paths are included. 2.3 Verify accuracy of the overall masses. 2.4 Review reasonableness of the eigen solutions. The overall approach and assumptions were judged to be consistent with practices within the nuclear industry and generally adequate to predict overall seismic response. All structures were modeled as being linear clastic with fixed base boundary conditions. 5-1
The containment was analyzed by a three-dimensional finite-element model using shell elements with uniform mass densities. The effects of concrete cracking for the containment analysis were conservatively modeled- by using upper and. lower - bounds on .the modulus of elasticity. ' Accuracy of the containment model was confirmed by an independent analysis described in [Hashimoto,1984]. The other structures were analyzed using dynamic models with masses lumped at i major floor elevations.- Stiffness . matrices for the _ lumped mass models were i generated from more _ refined finite-element representations of the load resisting elements. -Uncracked stiffness properties were used for concrete shear walls. Floor
^ slabs were modeled as being infinitely stiff in their own planes.
With the possible exceptions noted below, the dynamic models of the structures were found to be consistent with practices used in the nuclear industry and = l 4 generally adequate to predict overall seismic response. Model descriptions and sketches in [Cygna, 1982) .were reviewed to verify that major load-resisting elements were included. Independent, approximate calculations were performed when possible to confirm accuracy of the lumped masses. The eigen solutions were reviewed using engineering' judgment to assess whether they were reasonable given ' the structure configurations. Review of the turbine / service building model indicates that the diesel generator enclosure walls contribute significantly to the overall stiffness of the lowest story. However, the load path -from the rest of the structure to the diesel generator enclosure is relatively flexible. Although decoupling_the diesel generator enclosure from the remainder of the structure reduces the overall structure stiffness, the net effect on seismic response is lessened since exclusion of the diesel generator enclosure also reduces the overall structure - mass. Rather than _' develop a new structure model, correction of the dynamic response was estimated in assessing the structural response factor for the fragility evaluations. Stiffnesses of the shear wall structures other than containment _wcre based upon uncracked properties. On-going scale model testing being conducted for the NRC
- has indicated that this may result in an overestimation of the actual stiffness. The effect of this issue on the plant HCLPF capacity was assessed by sensitivity studies described in Section 5.5.4.
i Group A structures that were not dynamically analyzed by Cygna are the circulating water pumphouse and the fuel oil pumphouse. . Review. of the circulating water pumphouse indicates that the concrete portions at and below the - operating floor can be considered essentially rigid since the structure is supported by several heavy concrete walls and the exterior north, south, and west walls were poured against excavated rock. Dynamic analysis of- the steel superstructure supporting the roof slab was independently generated for the fragility evaluation. Dynamic response of the fuel oil pumphouse was n'ot required since no Group A components are housed in or directly attached to it. 5.2 Simolified Analysis and Use of Screeninn Tools i NUREG/CR-4334 developed a set of initial screening criteria for seismic margin 4 studies as outlined in Section 4.1 of this report. These initial screening criteria are l l 5-2 i t - _ . - - - - - _ - , - . - _ _ - --- .-
~ - - - __ -_ . _ _ _
L l based on the results of past probabilistic risk assessments and on actual earthquake experience data. These ' initial screens identified classes of equipment and structures which have consistently demonstrated high seismic capacities. Within ' the remaining classes of equipment and structures which have not been screened i 'out there typically exists a range of seismic capacities. Simplified analysis
! techniques can be utilized to separate out components within a category which have very high seismic capacities and do not warrant a detailed fragility analysis.
Simplified analyses for the Maine Yankee margin study were conducted using either of two methods: o Fragility derivation using conservative response and capacity parameters and an estimate for the variability ($R +kU = 0.7) { o Deterministic evaluation using conservative values similar to the : CDFM approach outlined in NUREG/CR-4482-a The 120-V ac vital bus I A through 4A is a good example of a Maine -Yankee component whose seismic margin was evaluated on the basis of a simplified analysis. These 120-V buses are wall-mounted panels containing circuit breakers I and are located in the control room (service building at +21 ft). NUREG/CR-4334 does not explicitly screen out active electrical equipment and states that anchorage
; should be verified for the cabinet and for individual components in the cabinet.
The walkdown of these 120-V buses showed them to be relatively lightweight I components with overdesigned wall anchorage. A simplified conservative fragility analysis was conducted utilizing the following parameters: o Weight = 245 lb (specified on drawing) 1 o Spectral acceleration = peak spectral value for 7% damping for the appropriate floor spectra (cabinet is judged to be rigid) o Center of gravity = 8 in. away from the wall (conservative since the cabinet is only 10 in. deep) o Anchor bolt capacity = industry specified allowable of one-quarter of the ultimate strength
;=
0 $R + MU = 0.7 The simplified fragility analysis on these panels resulted in a HCLPF capacity greater than 0.5g. Components whose HCLPF capacities were calculated via simplified analysis methods to be greater than 0.3g were automatically screened out, and detailed fragility derivations were judged to be unnecessary. The conservatisms most commonly utilized in the simplified analyses are: o Natural frequency - conservative estimate on the frequency or use of the frequency corresponding to the peak spectral acceleration i 5-3 ; i I
r o _ Capacity - conservative estimate such as the Code' al'Iowable, 70% of the ultimate,'or failure at the yield strength o Combined randomness - and ' uncertainty variability - conservatively - estimated based on calculations for similar
- components at Maine Yankee 5.3 Second Walkdown The second walkdown was performed to accomplish the following tasks:
- o Obtain additional detailed information on equipment and
- structures inspected during the first walkdown.
i o Survey components added to the Group A equipment list after the
- first walkdown was performed.
HCLPF capacities were calculated for certain components immediateiy following the first walkdown. In determining these capacities, additional information needs were identified. The first task listed above was performed to obtain this - information for finalization of component HCLPF capacities. A minor amount of effort was also devoted towards additional documentation' of items reviewed in the first walkdown. A limited number of components were added. to the Group A equipment list after the first walkdown was completed based upon more detailed
- system analyses. The second task was performed to obtain information on these added components for screening and/or HCLPF capacity determination.' Structures
- and equipment reviewed in the second walkdown are listed in Table 5.3-1.
5.4 HCLPF Canacity of Comnonents i In the following, the fragility analysis methodology is described with some , illustrative examples. 4 5.4.1 Fragility Analysis Method . 5.4.1.1 Methodology In this method, the component HCLPF capacity is calculated using the fragility of
, the component. The seismic fragility of a structure or equipment is defined as the
- conditional probability of its failure at a given value of peak ground acceleration.
4 The methodology for evaluating seismic fragilities of structures and equipment is documented in [Ravindra and Kennedy,1983], [PRA Procedures Guide,1983],' and [ Kennedy and Ravindra,1984] and has been developed and applied in over 20 seismic PRAs. 1 i 4 r 54 i _ - - - - - - _ - _ ._ _ _ -_,_ ~ ~ _- _ _ _. . . . - _. . _ . _ . _ _ _ . _ . . _ . _ . _ _ . . _ _ . - . . _ _ _ _ - . . _ . . . _ _
Table 5.3-1 Second Walkdown Component Review List TURBINE BUILDING COMPONENT FLOOR ELEVATION
- 1. Structural details
- 2. Block wall TB 21-2 21' - 0"
- 3. Primary Component Cooling Heat Exchanger E-4A 21' - 0"
- 4. Primary Component Cooling Heat Exchanger E-4B 21' - 0"
- 5. Secondary Component Cooling Heat Exchanger E-5A 21' - 0"
- 6. Secondary Component Cooling Heat Exchanger E-5B 21' - 0"
- 7. Cooler Supply Temperature Control Valve PCC-T-19 37' - 0"
- 8. Cooler Bypass Temperature Control Valve PCC-T-20 37' - 0" SERVICE BUILDING COMPONENT FLOOR ELEVATION
- 1. Structural details
- 2. Block Wall SB 21-7 21' - 0"
- 3. Block Wall SB 21-17 21' - 0"
- 4. Block Walls SB 35-1 to 4 35' - 0" i 5. Block Walls SB 39-1 and 2 39' - 0" 1
- 6. Block Wall SB 45-6 45' - 0"
- 7. Block TVall SB 61-2 61' - 0"
- 8. Block TVall SB 77-2 77' - 0"
- 9. General inspection of the Switchgear Room 4 5' - 6"
- 10. Protected SWGR. Room Supply Fan FN-31 39' - 0" ll. Computer Room Air Conditioner AC-I A 39' - 0"
- 12. Computer Room Air Conditioner AC-1B 39' - 0"
- 13. Lab Air Conditioner AC-2 39' - 0"
- 14. General inspection of the Control Room 21' - 0"
- 15. Cable Trays
- 16. Control Room Ceiling 21' - 0" ,
TURBINE BUILDING AUXILIARY BAY COMPONENT FLOOR ELEVATION I
- 1. Auxiliary Fuel Oil Supply Tanks Tk-28 A & B 21' - 0"
- 2. Diesel Fuel Oil Day Tank Tk-62 A & B 21' - 0"
- 3. DG Compressed Air Tanks Tk-76A & B1-6 21' - 0"
- 4. DG Room Exhaust Fans FN 20A & B 31' - 0"
- 5. Diesel Generator Air Intake & Exhaust Dampers 31' - 0" l 6. DG 1B Cooler Inlet Temperature Control Valve 23' - 0" i
SCC-T-305 5-5
~
Table 5.3-1 Second Walkdown Component Review List (Continued) PRIMARY AUXILIARY BUILDING COMPONENT FLOOR ELEVATION
- 1. Charging Pump Seal Leakage Coolers E-92A & B 11' - 0"
- 2. Seal Water Heater E-96
- 3. Auxiliary Charging Pump P-7 Lube Oil Cooler 11' - 0"
- 4. Return from Penetration Coolers PCC-A-238 12' - 0"
- 5. Seal Water Supply Filter FL-34B
- 6. PCCW Isolation from the RHR Heat Exchanger 11' - 0" PCC-M-90
- 7. PCCW Isolation to Containment PCC-M-219 11' - 0"
- 8. Cable Trays STEAM AND FEED WATER VALVE AREA COMPONENT FLOOR ELEVATION 1
- 1. Structural details
- 2. AFW Pump B Turbine Throttle Valve MS A-173 21' - 0"
- 3. Turbine Steam Supply Pressure Control MS-P-168 21' - 0" -
CONTAINMENT SPRAY PUMP HOUSE COMPONENT FLOOR ELEVATION
- 1. Residual Heat Removal Heat Exchangers E-3A & B 14' - 0"
- 2. Reactor Coolant Regenerative Heat Exchanger E-67
- 3. Safeguards Pumps Seal Leakage Cooler E-91 A & B 00 - 0"
- 4. LPSI Pump Coolers 21' - 0" l S. Containment Spray Pump P-61 A, B & S Coolers 14' - 0"
- 6. Containment Penetration Cooling Lines
- 7. Block Wall VE 21-1 21' - 0" FUEL BUILDING COMPONENT FLOOR ELEVATION
- 1. Block Walls
- 2. Fuel Pool IIcat Exchanger E 25 21' - 0" 56
.i i l i Table 5.3-1 Second Walkdown Component Review List (Continued) l l MCC ROOM COMPONENT FLOOR ELEVATION
- 1. Cable Trays MISCELLANEOUS COMPONENTS i
COMPONENT FLOOR ELEVATION
- 1. Instrument Racks various E
T } I 5-7
- (
n The_ objective of fragility evaluation. is to estimate the . ground acceleration '
~
capacity- of a given component.. This capacity _ is defined as the' peak . ground l acceleration value at which the seismic response of a given component located at a specified point in the structure exceeds the component's resistance, resulting in its. failure. The ground acceleration capacity of the component is estimated. using
; information on plant design bases, responses calculated at the design-analysis stage, l as-built dimensions, and material properties. The ground acceleration capacity is a
- random variable which can be described completely by its probability distribution.
However, there is uncertainty in the estimation. of the parameters of this distribution, the exact shape of this distribution, and in'the appropriate failure model for the component. For any postulated failure model and set of parameter values and shape of the probability distribution, a fragility curve depicting the ~ , conditional probability of failure' as a function of ground acceleration can be obtained. Hence, for different models and parameter assumptions,- one could
, obtain different fragility curves. A satisfactory way to consider ' these uncertainties is to represent the component fragility by means of. a family of ; fragility curves obtained as above; a subjective probability value is assigned to each curve to reflect the analyst's degree of belief in the model that yicided the 4 particular fragility curve. When represented in this fashion, the fragility curves 1 need not appear to be smooth S-shaped curves, approximately parallel to each other; they could intersect 'each other and they may not even be nondecreasing ; functions of peak ground acceleration. The only requirement is that fragility being a probability should be between 0 and I (see Figure 5.4-1).
l j At any acceleration value, the component fragility (i.e., conditional probability of failure) varies from 0 to 1; this variation is represented by a subjective probability distribution.' On this distribution we can find a' fragility value (say,~0.01) that corresponds to the cumulative subjective probability of SE We have 5% cumulative subjective probability (confidence) that the fragility is less than 0.01. Similarly, we can find a fragility value for which we have a confidence of 95%
; Note that these statements can be made without reference to any probability model.
Using this procedure, the median and high (95%) and low (5%) confidence fragility curves can be drawn. On the high confidence curve, we can locate the fragility i value of 5%; the acceleration corresponding to this fragility on the high confidence
- curve is the so called HCLPF capacity of the component. By characterizing the i component fragility through a family of fragility curves, the analyst has expressed
; all his knowledge about the seismic capacity of the component along with the i uncertainties. Given the same information, two analysts 'with similar experience l and expertise would produce approximately the same fragility curves. '
Development of the family of fragility curves using different failure models and j parameters for a large number of components in a seismic margin review or seismic
; PRA is impractical if it is done as described above. Hence, a simple model for the j fragility was proposed as described in the above cited references. In the following this fragility model is described.
i Fragility Model !- The entire fragility family for an element corresponding to a particular failure mode can be expressed in terms of the best estimate of the median ground acceleration capacity, A m, and two random variables. Thus, the ground acceleration capacity, A, is given by 58 i l
0 1.0 m x
=>
d 6 A o D E 4 2 8 a. a
' o H
C
=a o
l 1 0 PEAK GROUND ACCELERATION 1 Figure 5.41 Fragility Curves. 1 I I 59 i
A=A mR 6 E U, (5.4-1) in which h respectively,R and hU are random variables with the inherent randomness about the median andunit medians, inrepresenting, the uncertainty the median value. In this model, we assume that both 6R and 6U are lognormally distributed with logarithmic standard deviations,$R and $g, respectively. The formulation for fragility given by Eq. (5.4-1) and the assumption of lognormal distribution allow easy development of the family of fragility curves which appropriately represent fragility uncertainty. For the quantification of fault trees in the plant system and accident sequence analyses, the uncertainty in fragility needs to be expressed in a range of conditional failure probabilities for a given ground acceleration. This is achieved as explained below: With perfect knowledge (i.e., only accounting for the random variability,6R), the conditional probability of failure, f ,ofor a given peak ground acceleration level, a, is given by
,. f n ga I -
fo"4 (5.4-2) where @(') is the standard Gaussian cumulative distribution function. The relationship between of and a is the median fragility curve plotted in Figure 5.4-2 for a component with a median ground acceleration capacity A m = 0.90g and $ 0.30. For the median conditional probability of failure range of 5% to 95%,Rthu " ground acceleration capacity would range from 0.55g to 1.48g. When the modeling uncertainty 6 is included, the fragility becomes a random variable (uncertain). At each acceferation value, the fragility f can be represented by a subjective probability density function. The subjective probability, Q (also known as " confidence") not exceeding a fragility f' is related to f' by (a i 1
*In f, =4
- O Y (Q)
U (5.4-3)
- #R where Q = P[f < f' / a] i.e., the subjective probability (confidence) that the conditional probability of failure, f, is less than f for a peak ground acceleration a
@ ~I(') =
the inverse of the standard Gaussian cumulative distribution function. For example, the conditional probability of failure f' at acceleration 0.4g that has a 95% nonexceedance subjective probability (confidence) is obtained from Eq. (5.4-
- 3) as 0.22. The 5% to 95% probability (confidence) interval on the failure at 0.4g is 0 to 0.22. Subsequent computations are made easier by discretizing the random 5 10
1.0 e 95% B Nonexceed. liw0 .8 Probability ~ TConfidence Median o Curve Fragility x Curve 50.6 - S
.o O
d[0.4 - i Tu - 5% i E Nonexceed, 3 0 .2 - Probability (Confidence) 1
; E g Curve
{ 1 I I 0.4 0.8 1.2 1.6 2.0
, SSE Peak Ground Acceleration, g i
1 i l l
- Figure 5.4 2 Median,5% Nonexceedence, and 95% Nonexceedence Fragility )
Curves for a Component. l l l l 5-11 _ . . _ . - ~ . . . _ . . _ - . _ _ _ . _ _ , _, .f
. .=
variable probability of failure f into different intervals and deriving probability q; for each interval (Figure 5.4-3). Note that the sum of q; associated with all the intervals is u'nity. The process' develops a family of fragility curves, each with an associated probability q;. 4 The median ground acceleration capacity Am, and its variability estimates hR and fg are evaluated by taking into account tiie safety margins inherent in capacity
, predictions, response analysis, and equipment qualification, as explained below.
Failure Modes The first step in generating fragility curves such as those in Figure 5.4-3 is to develop a clear definition of what constitutes failure for each of the critical elements in the plant. This definition of failure must be agreeable to both the structural analyst generating the fragility curves and the systems analyst who must judge the consequences of component failure. Several modes of failure (each with a different consequence) may have to be considered and fragility curves may have to be generated for each of these modes. For example, a motor-actuated valve may fail in any of the following ways: 4 l 1. Failure of power or controls to the valve (generally related to the seismic capacity of the cable trays, control room, and emergency
! power). Since they are not related to the specific item of ! equipment (i.e., motor actuated valve) and are common to all I
active equipment, such failure modes are most easily handled as failures of separate systems linked in a series to the equipment.
- 2. Failure of the motor.
I
- 3. Binding of the valve due to distortion and, thus, failure to operate.
- 4. Rupture of the pressure boundary.
It may be possible to identify the failure mode most likely to be caused by the scismic event by reviewing the equipment design and considering only that mode. Otherwise, fragility curves are developed based on the premise that the component could fail in any one of all potential failure modes. Identification of the credible modes of failure is largely based on the analyst's ! experience and judgment. Review of plant design criteria, calculated stress levels in relation to the allowabic limits, qualification test results, seismic fragility I evaluation studies done on other plants, and reported failures (in past earthquakes, in licensee event reports and fragility tests) are useful in this task. I Structures are considered to have failed functionally when they cannot perform their designated functions. In general, structures have failed functionally when inciastic deformations under seismic load are estimated to be sufficient to potentially interfere with the operability of safety related equipment attached to
- the structure, or fractured sufficiently so that equipment attachments fall. These j failure modes represent a conservative lower bound of seismic capacity since a 1
5 12 i
l l l l i l e 1.0 u , 2 To
' O.8 -
o 50.6 D q1 q2 j q3 Q4 [ 0.4 - QS To C
.$ 0.2 -
1 sC O I f O
- 1 0.4 0.8 1.2 1.6 2.0
SSE Peak Ground Acceleration, g i 4 Figure 5.4 3 Family of Fragility Curves for a Component. 4 1 5-13 i e- -- -, . - ,-.w - - -- " - - - - - -e- - --:-
?
4 larger margin of safety' against collap.e exists for nuclear structures. Also, a
- structural failure has been generally a.sumed to result in a common cause failure of multiple safety systems, if these ar,'h~oused in the same structure. For exampic, the service water pumps in Zion we e assumed to fail when the crib house pump enclosure roof collapses.
For piping, failure of the supp;rt system and plastic collapse of the ' pressure boundary are considered dominant failure modes. Failure modes of equipment examined may include structurri failure modes (e.g., bending, buckling of supports, anchor bolt pullout, etc.), .unctional failures (binding of valve, excessive 4 deflection), and relay trip or , hatter. Consideration should also 8.e given to the potential for soil failure modes (e.g., liquefaction, toe bearing p essure failure, base slab uplift, and slope failures). For i
' buried equipment (i.e., pl.;ing and tanks), failure due to lateral soil pressures may be an important mode. Seismically induced failures of structures or equipment i
under impact of another structure or equipment (e.g., a crane) may also .be a consideration. Seismically induced failures of dams, if present, resulting in either i flooding or loss-of-cooling-source, should also be investigated. I Estimation of Fragility Parameters I
! In estimating fragility parameters, it is convenient to work in terms of an
- intermediate random variable called the factor of safety. The factor of safety, F, j on ground acceleration capacity above the safe shutdown carthquake level 1 specified for design, ASSE, is defined as follows
- ;
A=FA SSE j p , Actual seismic canacity of element Actual response due to SSE 1
, Actual canacity Calculated capacity Calculated canacity X
Design response due to SSE i X Desian resnonse due to SSE Actual response due to SSE 1 I i l 5 14 I
. _ . , . , - - - -..._,-..--.,--.,..--._.-,--__-m-
l F is further simplified as: Actual canacity l p, Design response due to SSE y Desian response due to SSE Actual response due to SSE (5.4-4) F=FFC RS. Note F can also be defined with reference to a different earthquake such as the i review earthquake level in this margin study.
- t The median factor of safety F m, can be directly related to the median ground j acceleration capacity, Am, as:
Am F (5.4-5) m=A SSE The logarithmic standard deviations of F, representing inherent randomness and uncertainty, are identical to those for the ground acceleration capacity A. 4 For structures, the factor of safety can be modeled as the product of three random variables: i j F=F3 Fy FRS (5.4 6) i j The strength factor, F3, represents the ratio of ultimate strength (or strength at loss-of function) to the stress calculated for ASSE. In calculating the value of FS,
; the nonseismic portion of the total load acting on the structure is subtracted from the strength as follows:
{ S-PN F , (5.4 7) 4 3,PT-Pg I where S is the strength of the structural element for the specific failure mode, P is the normal operating load (i.e.,' dead load, operating temperature load, etc.) an$ ) PT s i the total load on the structure (i.e., sum of the seismic load for ASSE and the i normal operating load). For higher earthquake levels, other transients (e.g., SRV - { discharge, and turbine trip) may have a high probability of occurring ;
- simultaneously with the carthquake; the definition of Pg in such cases should be extended to include the loads from these transients.
l The inelastic energy absorption factor (ductility), Fy, accounts for the fact that an earthquake represents a limited energy source and many structures or equipment l items are capable of absorbing substantial amounts of energy beyond yield without i loss of function. A suggested method to determine the deamplification effect 5-15 4 I
. - - . _ . . ~ . . , - _ , , . ~ _ , , . . . . _ - . . . _ . - , . _,- --,,-,,. ,._.,y _ _ , , . . . _ , - , - - . , . , , , . - _ . _ _ - - _ - - , , , . _ . ,
.- .- .- . - _ ~ . - - --
resulting from inciastic energy dissipation involves the use of ductility modific'd response spectra [Newmark,1977). The desmplification factor is primarily a I function of the ductility ratio ' defined as the ratio of maximum displacement to displacement at yield. - More recent analyses [Riddell and Newmark,19791 have shown the deamplification factor to be a function of system damping. One might estimate a median value of for low rise concrete shear walls (typical of auxiliary building walls) of 4.0. The corresponding median Fy value would be 2.4. The vari bilities in the inciastic energy absorption factor, Fy, are both . estimated [ as R = 0.21 and pg= 0.21, taking into account the uncertainty in the predicted rela onship between Fp,y, and system damping. 1 f i The structure response factor, Fgg, recognizes that in the design . analyses structural response was computed using specific (often conservative) deterministic
- response- parameters for the structure. Because many of these parameters are -
1 random (often with wide variability) the actual response may differ substantially j from the design analyses calculated response for a given peak ground acceleration. 1 The structure response factor, FRS, is modeled as a product of factors influencing
- the response variability
{ FRS = FSAF,F Fgy MC F FEC FSD F33 , '(5.48) 1 where ] ! FSA = spectral shape factor representing --variability in ground motion and associated ground response spectra j F = direction factor representing the variability in the two
- Y earthquake direction response spectral values about the mean j value
, Fg =
damping factor representing variability in response due to '
~
difference between actual damping and design damping
- FM =
modeling factor accounting for uncertainty in response due to { modeling assumptions
- FMC= mode combination factor accounting for . variability in
} response due to the method used in combining dynamic modes
- of response I FEC = earthquake component combination factor accounting for variability in response due to the method used in combining i earthquake components FSD = factor to reflect the reduction with depth of seismic input
, F33 = factor to account for effect of soil structure interaction. I { The median and logarithmic standard deviations of F are expressed as: i i ) 5 16 l
F F (5.4-9) F, = FS, Fp , FSAm F,F6m 9 Mm F MCm ECm Fgg, F33 , and
." p S SA * *** 5 (5.4-10)
. The logarithmic standard deviation ppi s further divided into random variability, i
phe, t and uncertainty, $g. m To obtain the median ground acceleration capacit; l i peak ground acceleration J l For equipment and other components, the factor of safety is composed of a j capacity factor, F C; a structure response factor, FRS; and an equipment response l (relative to the structure) factor, FRE. Thus, f (3 4*II) ' F=FC FRE FRS !' The capacity factor FC or f the equipment is the ratio of the acceleration level at which the equipment ceases to perform its intended function to the seismic design i level. This acceleration level could correspond to a breaker tripping in a i switchgear, excessive deflection of the control rod drive tubes, or failure of a steam generator support. The capacity factor for the equipment may be calculated as the product of F3 and Fp . The strength factor, F ,S is calculated using Eq. (5.4- ' j 7). The strength, 5, of equipment is a function of the failure mode. Equipment failures can be classified into three categories: I
- 1. Elastic functional failures
- 2. Brittle failures i
- 3. Ductile failures 1
Elastic functional failures involve the loss of intended function while the component is stressed below its yield point. Examples of this type of failure j include the following: o Elastic buckling in tank walls and component supports 1 o Excessive blade deflection in fans o Shaft seizure in pumps The load level at which functional failure occurs is considered the strength of the jl component. 1 1 Brittle failure modes are those which have little or no system inelastic energy absorption capability. Examples include the following: l
) ! o Anchor bolt failures i o Component support weld failures o Shear pin failures l
Each of these failure modes has the ability to absorb some inelastic energy on the l component level, but the plastic zone is very localized and the system ductility for an anchor bolt or a support weld is very small. The strength of the component l 5 17 I
i l ' failing in a brittle mode is therefore calculated using the ultimate strength of the
- material.
1 i Ductile failure modes are those in which the structural system can absorb a . significant amount of energy through inelastic deformation. Examples include the ! following: o Pressure boundary failure of piping o Structural failure of cable trays and ducting o Polar crane failure ~ The strength of the component failing in a ductile mode is calculated using the yield strength of the material for tensile loading. For flexural loading, the ' strength is defined as the limit load or load to develop a plastic hinge. The inelastic energy absorption factor, Fp, for a piece of equipment is a function of the ductility ratio,p. The median value Fy is considered close to 1.0 for brittle and functional failure modes. For ductile failure modes of equipment that respond in the amplified acceleration region of the design spectrum (i.e.,2 to 8 Hz): l F - e (2 - 1)I , (5.4-12) j where 6is a random variable reflecting the error in Eq. (4-16) and has a median value of 1.0 and a logarithmic standard deviation, !l (increasing with the ductility ratio). For rigid equip $U, ranging from 0.02 to 0.10 ment, Fy is given by F -e 0.13 (5.4-13) f Again, E is a random variable or median equal to 1.0 and logarithmic standard deviation ranging from 0.02 to 0.10. The median and logarithmic standard deviation of ductility ratios for different j equipment are calculated considering recommendations of [Newmark,1977]. This reference gives a range of ductility ratios to be used for design. The upper end of this range might be considered to represent approximately the median value, while
' the lower end of the range might be estimated at about two logarithmic standard 4
deviations below the median. The equipment response factor FRE, is the ratio of equipment response calculated in the design to the realistic equipment response; both responses are calculated for design floor spectra. FRE is the factor of safety inherent in the computation of ! l ' equipment response. It depends upon the response characteristics of the equipment ' and is influenced by some of the variables listed under Eq. (5.4 '8). These variables differ according to the seismic qualification procedure. For equipment qualified by dynamic analysis, the important variables that influence response and variability are as follows: i
; o Qualification method (QM) o Spectral shape (SA) including the effects of peak broadening
- and smoothing, and artificial time history generation
- o Modeling (affects mode shape and frequency results)(M) 5 18 6
-.._.-__....___,__v,7,,._...-,_,mmmw.,,.-, _
, , , y, . , , , . . . ._m,,- - , . _ _ ~ _ , . , ,-m.m., ..,,s_- .m...,w,._,.,,
o Damping (6) o Combination of modal responses (for response spectrum method)
- (MC) o Combination of earthquake components (EC)
For rigid equipment qualified by static analysis, all variables, except the qualification method, are not significant. The equipment response factor is th: ratio of the specified static coefficient divided by the zero period acceleration of the floor level where the equipment is mounted. If the equipment is flexible and was designed via the static coefficient method, the dynamic characteristics of the equipment must be considered. This requires estimating the fundamental frequency and damping, if the equipment responds predominantly in one mode. The equipment response factor is the ratio of the static coefficient to the spectral acceleration at the equipment fundamental frequency. Where testing is conducted for scismic qualification, the response factor must take into account the following: o Qualification method (QM) o Spectral shape (SA) o Boundary conditions in the test versus installation (BC) o Damping (6)
- o Spectral test method (sine beat, sine sweep, complex waveform, etc.) (STM) o Multi directional efrects (MDE)
The overall equipment response factor is the product of these factors of safety 1 corresponding to each of the variables identified above. The median and logarithmic standard deviations for randomness and uncertainty are estimated following Eas. (5.4 9) and (5.410). The structural response factor, FRS, is based on the response characteristics of the structure at the location of component (equipment) support. The variables pertinent to the structural response analyses used to generate floor spectra for equipment design are the only variables of interest to equipment fragility. Time- , history analyses using the same structural models used to conduct structural response analysis for structural design are typically used to generate floor spectra. The applicable variables are as follows: o Spectral shape o Damping o Modeling o Soil-structure interaction l For equipment with a seismic capacity level that has been reached while the l structure is still within the clastic range, the structural response factors should be calculated using damping values corresponding to less than yield conditions (e.g., about 5% median damping for reinforced concrete). The combination of earthquake components is not included in the structural response since the variable is to be addressed for specific equipment orientation in the treatment of equipment i response. 5-19
5 Median parametersF, a ffecting capacity and response factors of . These safetyund median and variability'$R a variability ' estimates arc then combined using the properties 'of .lognormal distribution in accordance with Eqs. (5.4-6), (5.4-8), and (5.4-11) to obtain the
- overall median factor of safety Fmand variabilityhR andhU estimates required t
to define the fragility curves for the structure or equipment. For each variable and uncertainty ) variabilities affecting the factor must be separately of safety, estimated. the random g) tion is somewhat The differentia ju gmental, but can be based on general guidelines. -Essentially,hR represents variability due to i the randornness of the carthquake characteristics for the same acceleration and to the structural response parameters which relate to these characteristics. The { dispersion represented byhUi s due to factors such as the following: o Our lack of understanding of structural material properties such' as strength, inciastic energy absorption, and damping. o Errors in calculated response due to use of approximate modeling of the structure and inaccuracies in mass and stiffness > representations. o Usage of engineering judgment in lieu of complete plant-specific data on fragility levels of equipment capacities, and responses. l Information Sources For structures such as concrete shear walls, prestressed concrete containment, steel i i frames, masonry walls, field erected tanks, and buried structures, the fragility parameters are generally estimated using plant-specific'information. For major I passive equipment (e.g., reactor pressure vessel, steam generator, reactor coolant pump, recirculation pump, major vessels, heat exchangers, and major piping), it is i preferable to develop plant-specific fragilities using original design analyses. Because of the large quantities of other types of passive equipment (e.g., piping and supports, cable trays and supports, HVAC ducting and supports, conduit, and miscellaneous vessels and heat exchangers), it is generally necessary to use generic. fragilities. For active equipment, use of a combination of. generic and plant-specific information is needed to develop fragilities.
' Several sources of information are utilized in developing plant-specific and generic fragilities for equipment. These sources include the following:
- 1. Scismic qualification design reports
, 2. Seismic qualification test reports ! 3. Plant safety analysis reports 4 Seismic qualification review team (SQRT) submittals } 5. Seismic qualification report summaries
- 6. Past earthquake experience and expert opinion
- 7. United States Corps of Engineers shock test reports j 8. Specifications for the seismic design of equipment i
Sources I to 5 are plant specific; sources 6 to 8 are generic data collected for
- similar types of equipment.
i i ! 5 20 5
l Equipment fragility development is accomplished by grouping equipment into a number of categories. In seismic margin studies, a relevant quantity is the HCLPF acceleration capacity . of the component. This quantity considers both the uncertainty and randomness variabilities and is the acceleration value for which we have 95% confidence that the conditional probability of failure is less than 5%. That is, it is an acceleration i value for the component for which we are highly confident there is only a small l chance of failure given this ground acceleration level. l HCLPF capacity = A m exp( - 1.65 (fU +hR)} * (5.4-14) Spectral Shape Factor - FSA i The review carthquake level is specified in this study as the NUREG/CR 0098 median ground response spectrum (rock) anchored to 0.3g pga. It is interpreted as 3 I the 84% nonexceedance spectrum for Maine Yankee. For calculating the median capacity and variability, we need to estimate the median spectrum for Maine Yankee and the variability in the spectral ordinates. It is interpreted that the review earthquake level (target spectrum) is a 84% nonexceedance level, in that a future earthquake will exceed the target spectrum only if 16% of the spectral ordinates exceed the target over the frequency range of interest. l In calculating the spectral shape factor FSA we need to consider the peak-to-peak response spectrum variability which reflects the observation that response spectra in each horizontal direction for real carthquakes have hills and valleys relative to a mean spectrum. In other words, the spectral ordinates at a given frequency will occur randomly either above or below the mean spectrum. The peak-to-peak variability for multidegree of freedom structure is different from a single-degree of freedom structure. In general, the logarithmic standard deviation due to peak-to peak variability is less for a multidegree of freedom structure. In fragility analysis, the response spectrum logarithmic standard deviation, assuming independent dynamic modes, is given by the following equation [ Reed et al.,1986): 1 R. 4
#x l fpp (5.4-15) xm where gx = logarithmic standard deviation for building response in the x-i direction (similar for y-direction) 5 21
_ - _ _ _ _ _ . _ _ _ _ _ . _ - . _ _ _-_ __ - , ~
R;, = median building response for.the ith mode for an earthquake-in the x-direction R,- 3 median building response combined for all modes for an earthquake in the x direction When a single mode is domi.nant, I h " @pp j The.value of ppp has been estimated to be 0.20 [ Reed, et al.,' 1986]. i i In this present study, p is conservatively assumed to be 0.20. Since the review l level spectrum is assudEd to be at a . 84% nonexceedance value, the median - spectrum is obtained by multiplying the target spectrum by exp(-ppp). Therefore, the median spectral shape factor for structural response factor l calculation is given by Fgg = exp(0.20) = 1.22 - An additional factor to be considered is the horizontal carthquake direction variability This occurs since at a given frequency the response spectrum value in one horizontal direction is different from the direction 90 degrees away. If the response of the critical structural element depends equally on the two horizontal earthquake components, then the directional variability has little. effect on the response. In contrast, if one direction dominates, then direction variability- may , significantly effect response variability. I When co linear responses for two horizontal earthquake directions are combined, the logarithmic standard deviation for total response due to peak-to peak and direction variability, becomes for the SRSS peak response rule:
~
2 2 i
/R ,f pg 2 2 [p2 ,
[
,, 3,. /
where Rm) R g2 ) y (5.4-16) h logarithmic standard deviation for direction variability from a statistical analysis of the ratios of spectral accelerations for a fixed direction ' (e.g., north / south) to the corresponding geometric mean values. Since we are assuming that the maximum of the two horizontal components is used in r specifying the r;: view earthquake, the resulting distribution of , the ratio will be approximately lognormal with a median larger than 1.0 5 22
1
~l In the absence of such a detailed study. hue to directional effect is judged to be' between 0.10 and 0.20 and is taken af 0.15 in this ' margin study. The total variability due to peak to-peak variability and directional effects is taken as hSA" hpp + h = 0.202 + 0.152 = 0.25 -
It is assumed that these variabilities (peak-to peak and direction) are removed from any hazard analysis leading to selection of margin earthquake selected for review. It should be noted that the issue of exceedance of the target response spectrum (e.g. '16% assumed in the Maine Yankee specific ~ spectrum) is not the same as the confidence levels associated with the target spectrum as determined from the hazard curves. In the case of the hazard curves, the confidence is associated with ; j ' the uncertainty of the underlying geophysical parameters, while the exceedance of the target response ' spectrum is a randomness consideration associated with the variabilities of earthquake time history " signatures" [ Reed et al.,1986]. 5.4.1.2 Steel Structures HCLPF capacities were determined for the following steel framed structures: i o Circulating water pumphouse steel superstructure o Turbine / service building, steel framed portions o -Main steam valve house steel interior structure r The circulating water pumphouse (Figure 5.4-4) houses the service water pumps and other components of the service water system. This structure is constructed of - both reinforced concrete and structural steel framing. In the vicinity of the structure, grade is located at El 20'-0" and the rock line is at approximately El. O'- 0". The portion of the structure at 'and below El. ~21'-0" consists of thick reinforced
. concrete walls and slabs. The exterior north, south, and west walls were poured against rock and the exterior east wall faces the Back River. As discussed in ~
Section 4.3.4.1, the concrete portion of the pumphouse is screened out for the 0.3g review level carthquake. The 12-inch-thick concrete roof slab at El. 37'-0" is supported by structural steel beams and ' columns. The columns are typically anchored to the concrete floor at
;. El. 21'-0". The steel superstructure is enclosed by metal siding. Resistance to lateral seismic loads is provided primarily by diagonal bracing (Figure 5.4-5). The diagonal braces are built up from two channel sections arranged to form a T shape.
A single' diagonal brace is provided at each side of the building.
'l Review guidelines are not established for steel framed structures. Consequently, -
the circulating. water pumphouse steel superstructure was analyzed to determine its fragility. There are no Group A components attached directly to the steel framing. ' Damage to the Group A service water system components can only occur if the roof slab collapses. 1 5-23
1 i l 1 p....., . . _ a-. s
. ! f. .-a ".. ., ., i n , . r .
i >
'_ 1
'3 3 i 1 3
1 l : vg<] k,. u . wa -- , _g .;. -
- x. , . t c ., .. .
- t4 a:....,.
, e4 .g.%.__
, ou,. ,, .
rk~"-
. _ _ _ j ,p __.m.......,u i
q,,4 .,
- 7. , ,
6*t to **aou . .- y g , . , , a-M ./, , ,,, _- 7 P, (/ .-r6 T I.Il,e..m . I u., u c...
/
E
. .s ,,
. .n t:
- _ u ,;, y.
!... ,,..L F,I m ,.
o
= ..,
e i.,p A __r
" -~
,_.T j!
( 5 : E ' " _.. N. /y} L m adin.u ir Y* i n _= >j
= ex.3; w, a - )
l ,. u . ,, . _. *i<-~. ; I r lg )*j g
.r, I,wru.e ' "'
- f. / , .j 4.qtswg wA
{, i. s \ w , O awa ma rw.ia'.o ,. sueno j ett.i
- 6' 9-!f ][.-e
- t. e- _r e-3 s.c Stitvitt j .--,
,ALL. 5-.wsw c ll [ ; "w ,I,"
ji e rtt .,.FtLw k f jl- w.e e l . l'l ll
..e , ,
tt-to-a,
- 31.,
O l' I' d ' I., i U
- I'"'"
6 ,t [ 06dt'4 y
...... = ,
._ e .t.,,.., .
> M'
- U
.+in.ii4"0 Pila vu -
e <- . . e- ,o . o. _ a ,._ Figure 5.4-4 Circulating Water Pumphouse. 5-24
>A I v 1~
^
11 11 11 s , C9x13.4 m SECTION A-A l j Figure 5.4-5 Circulating Water Pumphouse, Typical Braced Frame. l l 5-25
Seismic analysis of the steel superstructure above El. 21'-0* was performed using a single-degree-of-freedom dynamic model. The concrete portion of the pumphouse was considered to be essentially rigid since it is built into rock and the shear walls are very stiff. Mass lumped at the roof level included contributions from the slab, steel framing, siding, and attached equipment. The roof slab was treated as a rigid diaphragm. Horizontal stiffnesses for the dynamic model were based on the stiffnesses associated with the braced frames. Seismic responses in the two horizontal directions were analyzed independent of each other since little coupling is expected. Fundamental frequencies in the two horizontal directions were both estimated to i be about 3.7 Hz. Seismic input to the steel superstructure consisted of the specified revie* level earthquake ground response spectrum since the concrete structure is essentially rigid. Median damping was estimated to be approximately 15% based upon the recommendations of NUREG/CR-0098 for bolted steel structures at or near the yield point. Resulting overall seismic inertial loads were distributed to i ^ the braced frames in proportion to their relative rigidities. Accidental torsional moments were based upon a minimum 5% eccentricity. A number of potential failure modes were considered in the evaluation of the
- circulating water pumphouse, including the following
o Diagonal brace buckling
- o Column buckling -
~ o Diagonal brace connection failure o Column base connection failure o Roof diaphragm shear failure Buckling of the diagonal braces was found to be the initial failure mode. The results of testing on carbon steel compression members cre compiled in [ Hall,1981]. This study derives empirical equations to determine average buckling stresses and associated variabilities. These data were used to estimate the buckling capacity of the diagonal braces and its logarithmic standard deviation. Inelastic energy absorption due to ductile response was not included since the structure is essentially clastic until buckling occurs. A HCLPF capacity of 0.19g was calculated for initial diagonal brace buckling. As previously noted, damage to the service water system components can cccur only if the roof slab collapses. Known damage to steel framed structures in past carthquakes with estimated peak ground accelerations in the range of 0.5g is uncommon. There have only been a few instances of gross collapse of steel framed structures in actual carthquakes of magnitudes larger than expected for the Maine. Yankee site. The overall evidence on the performance of steel framed structures in past earthquakes indicates that the collapse capacity should be at least 0.3g. Furthermore, the results of static and dynamic testing demonstrate that additional reserve capacity past buckling is provided by the following two sources: o Even in the buckled state, a steel member can still resist axial compression by developing a plastic hinge at midlength. Although the capacity degrades with increased axial displacement, significant displacements in excess of that at buckling are still attainable. 5-26
o The reversing nature of the dynamic seismic input causes displacements to cycle rather than monotonically increase to a total loss of stability. Based upon past carthquake performance of steel structures and test data, the HCLPF capacity for collapse of the pumphouse steel structure is judged to be at least 0.30g; corresponding estimated fragility parameters are shown in Table 5.5-1. The turbine / service building (Figures 5.4-6 and 5.4-7) was built as a single integral structure. Most of the Group A components housed in this building are located in the control /switchgear building and the diesel generator enclosure. These portions of the structure are constructed of heavy reinforced concrete shear walls and slabs which have been screened out for the review level earthquake of 0.3g. Only the steel framed portions of the turbine / service building that could fail.with resulting damage to Group A components were subjected to detailed evaluation. These areas included the following: , o Service building floor at El. 39'-0", bounded by column lines 1/2, 7, C, and F. o Service building roof at El. 61'-0" o Turbine building floor at El. 35'-0" over the PCC/ SCC system Approximate comparisons of initial load distributions and load carrying capacities indicated that the turbine building diagonal braces may buckle. However, even if this occurs, overall structural integrity of the turbine building will be maintained since lateral displacements will be limited by the massive concrete turbine pedestals once the one inch separation gap with the operating floor is closed. The portion of the service building floor at El. 39'-0" bounded by column lines 1/2, 7, C, and F was evaluated since its failure may result in damage to the PCC lines running to the chillers. Resistance to seismic load in the N-S direction is provided by diagonal bracing, W36 building columns on Column Line C, diesel generator enclosure shear walls, and control /switchgear building shear walls. The capacity of the load path to the diesel generator enclosure is limited by the shear capacity i of the adjacent roof deck. Because the El. 39'-0" floor is discontinuous at Column Line 7, the capacity of the load path to the control /switchgear building is limited by the weak axis bending capacity of the columns on Line 7 which must transmit E-W seismic loads to the floors at El. 35'-0" and El. 45'-0". Accounting for the i available load paths and their ductilities, a HCLPF capacity of 0.38g was conservatively estimated. The steel building roof at El. 61'-0" over the chillers and the turbine building steel framing over the PCC/ SCC components at grade both have relatively low masses. They were evaluated using conservative methods. Steel framing in these two areas were found to be capable of maintaining their integrity at peak ground i acceleration levely, well in excess of 0.3g. The steel framed structure within the main steam valve house provides support for the main steam lives before they enter containment and associated components. The structure consists primarily of platforms with metal grating supported by steel beams and columns. It is essentially independent from the exterior concrete structure with the exception of three horizontal struts spanning between the steel framing and the east exterior concrete wall. These struts were included as pipe supports only. 5-27 _ - -- . _- _ \
NORTH I SERVICE BLDG FLOOR SERVICE BLDG ROOF EL 39' EL 35', EL 45'6" O' CONTROL BLDG.
@ SERVICE BLDG. FLOOR EL 39' EL 35' EL 45'-6" N B G. MAZZANINE M ZZ N R EL 35 DIESEL GENERATOR TURBINE PEDESTAL ENCLOSURE O @@
Figure 5.4-6 Turbine / Service Building, El. 35'-0"/El. 39'-0" 5-28
l NORTH
@g i
OL SERVICE BLDG ROOF BLDG. ROOF TURBINE BLDG. OPERATING FLOOR 4 TURBINE O @@ Figure 5.4-7 Turbine / Service Building, El. 61'-0". l 5-29 t
j Resistance to lateral seismic load is provided primarily' by diagonal bracing at each 4 side of the structure. This bracing consists of either double angles or composite channel sections welded together to form-a T shape. During the first walkdown, it i l- was noted that the diagonal brace for the north side of the structure at the lowest
! story was missing. This member was apparently removed due to interference.with the access door.
Reevaluation of the steel structure in the as-built condition was performed by. Cygna. Results consisted of overall seismic loads from the dynamic model and localized load distributions to individual structural members from more detailed static models.- The following potential failure modes were evaluated . using conservative methods:
; o Column failure under combined axial load and bending 'I o Diagonal brace buckling l
- o Brace connection failure
; o Column base connection failure o Beam failure due to biaxial bending o Pullout of pipe support anchorage Particular attention was given to members in the vicinity of the missing brace since their loads may be increased to accommodate the load path discontinuity.
The HCLPF capacity for this structure was found to be greater than 0.3g based upon conservative methods. 5.4.1.3 Block Walls A list of block walls supporting or adjacent to Group A components is presented in Chapter 7. These walls are separated into two categories: o Walls that were screened out on the basis that their failure will not damage Group A components. o Walls for which HCLPF capacities were determined.
, Evaluation of the Group A walls did not identify any with HCLPF capacities less than 0.3g.- This can be attributed to the scismic retrofits already installed or to be '
added during the next outage. Verification - that the block wall capacities are - greater than 0.3g was performed either by determining the actual fragilities or by utilizing simplified, conservative analyses. As an example, the fragility evaluation of Wall SB 35-3 is described in the following. Wall SB 35-3 (Figure 5.4-8) is located at the west side of 'the service building l battery room at El. 35'-0". Out-of-plane collapse of this wall will result in damage to Battery Group 3 mounted on the floor next to the wall. It is built of 12-inch-thick, hollow block units which are assumed to be normal weight. -The wall is 7'-0" high by 19'-8" wide. It is mortared up into the overhead concrete beam. l Original block wall construction specifications for Maine Yankee were available. l Concrete block units were specified to meet ASTM C90 requirements. The mortar was specified to meet C270, Type M or N requirements. Testing on a limited number of unit and mortar samples taken from the battery room at El. 45'-6" determined the following average compressive strengths: o Block unit, f =m3800 psi on the net area l l 5-30 l l l
'*FM '
p"- o
't
'~
__ ~
;.y ,
s 4
~
lT 4. i + < k W'- s .: gp cj : 4 ; cir}(
%.gg
,.gE-
- a. Looking from the Battery Room
*4 3
3 e
; ye 4
@$${Ih5k[7 <
l f 4#i%- ' Ls 5M&fe;
.W;5 w
fp 7 /1'] f , st3' - i E ggu r M : r.f7
- g -
nc;-6Ja .: 3, a c. .. o. a f <:A l y
)'- :jp
.c
_p
+, I
.I j[ L .. MfN g~ il lf f ;
< p:- .'., ;
, 3 ,
Tgm f 5' I
,7, _E- , , m :. at l R =; i- I$
d: . D gy N- '
' , .& jj f
- b. Looking from the Cable Spreading Room Figurc 5.4-8 Block Wall SB35-3.
5-31 l
; :l
.j . i o Mortar, m g- 2700 psi The mortar strength would imply the use of Type M mortar at the El. 45'-6" battery room. Because use of Type M mortar throughout the plant could not be confirmed,
~
Type N was assumed for other walls. An average Type N mortar strength of 825 - psi was estimated by scaling the results from the test samples. This wall was analyzed to determine its resistance against out-of-plane seismic response. Because _ of its low height-to-width ratio, it was analyzed as a one-way member spanning vertically. The top and bottom boundary conditions; were-represented as simple supports.; The fundamental out of-plane response frequency, was calculated using uncracked stiffness properties and found to be into the rigid response range of the floor spectra. . The maximum ' stress due to out-of-plane response was based upon inertial loads due to the floor ZPA and the clastic section modulus. This stress was combined with those due.to in-plane and vertical seismic I inertial loads by SRSS. Additional in-plane loads may be developed by relative structure displacements between the top and bottom of the wall. However, these loads are stiffness dependent and will tend to _ dissipate once the wall cracks. Relative structure displacements are small and are considered insufficient to cause wall collapse. Initial cracking was evaluated by comparing the calculated clastic stress against I the median mortar modulus of rupture, including additional resistance provided by 4 decd load compression. Conservative allowable stresses for _use in the design of masonry structures are presented in ACI 531. The commentary to ACI 531 notes that factors of safety in masonry are generally held at a minimum of three. Review of available test data indicated that the average factor of.. safety is greater than this value. The median cracking stress was estimated to be three times the ACI 531 allowable value to account for potential reduction in capacity due to
! differences between laboratory and field workmanship. A relatively_ large coefficient of variation of 0.3 was assigned to the cracking stress to account for 4 scatter in the test data, workmanship effects, etc. Inelastic energy absorption effects were not considered since behavior prior to cracking is essentially linear.
A HCLPF capacity of 0.60g was estimated for initial cracking of Wall SB 35-3.- Even if the wall cracks through at midheight, stability .can be maintained by arching action which has been exhibited in testing of confined walls.- The wall _ will tend to behave as two rigid bodies pivoting about the midheight crack.' This 3 rotation causes the wall to push outward against the top and bottom supports, thus developing compression stresses at the points of contact which _ oppose the transverse loading. The resistance developed by arching action may be limited by compression or shear in the wall, or by bending of the floor beams above and below. An approximate analytical model was developed to' account for arching action. While this model was not correlated against test results, it demonstrated that, even using very conservative assumptions, this wall could potentially develop a HCLPF capacity well in excess of 0.3g through arching action. ! 5.4.1.4 Flat Bottom Storage Tanks The following flat bottom storage tanks were evaluated: o Refueling water storage tank, TK-4 (RWST) o Demineralized water storage tank, TK-21 (DWST) 5-32
o Primary water storage tank, TK-16 (PWST) o Spray chemical addition tank, TK-54 (SCAT) While the SCAT is not actually part of the Group A systems, its failure could result in the failure of interconnecting piping to the RWST. Fragility parameters
. for the RWST, DWST, and PWST are shown in Section 5.5-1. Calculated HCLPF capacities for these tanks are listed in Chapter 7. These capacities were determined using essentially the same methodology for all tanks. A detailed
- description of the evaluation of the RWST follows.
I !- An overall view of the RWST and a typical anchor bolt detail are shown in Figure 5.4-9. The RWST has an inside diameter of 40'-0". The shell has a total height of 38'-0" to the springline and varies in thickness from 3/16 in. to 3/8 in. The spherical shaped head has a total rise of 6'-10" and is 5/16 in. thick. The bottom
- plate is 1/4 in, thick. The shell, head, and bottom plate are fabricated from A240 Type 304 stainless steel. The RWST is anchored to its concrete ring wall foundation by a total of 8 anchor bolts spaced at 45 degrees. These bolts are 2
- inches in diameter and are fabricated from A307 steel. They are attached to the tank shell by chairs built up from stainless steel plate (Figure 5.4-9).
Seismic response of the RWST was calculated following the recommended approach described in NUREG/CR-il61. The following response modes were included: o Horizontal impulsive o Horizontal convective (sloshing) o Vertical fluid o Vertical tank 4 Effective masses and mass centroids to determine overall tank seismic loads due to l [ the horizontal impulsive and convective modes were calculated from equations in l NUREG/CR-1161. The impulsive mode fundamental frequency of 5.6 Hz' was based upon coefficients developed by [Haroun,1981]. The frequency of the convective mode was found using equation for rigid tanks. The median impulsive mode damping of 7% was based upon the recommendations of NUREG/CR-0098 i
; for welded steel structures at or near the yield point. Damping of 0.5% was assigned to the convective mode. Overall tank seismic loads due to the impulsive i and convective modes were found by factoring the effective masses, mass centroid heights, and spectral accelerations corresponding to the modal frequencies.
Contributions from the two horizontal modes were then combined by SRSS. Tank response due to vertical excitations accounted for amplified fluid response associated with tank shell flexibility in the breathing mode and vertical response of the tank itself. Failure of the RWST is controlled by buckling of the shell due to'overall seismic-
, moment at the tank base. Conventional design practice for anchored tanks utilizes a stress distribution at the tank b:se derived from elementary bcam theory. In this stress distribution, plane sections are assumed to remain plane with the neutral axis at the tank centerline under pure bending. This distribution results in unrealistic seismic capacities and is considered to be overly conservative. When the vertical 5-33
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- a. Overall View b. Anchor Bolt Chair Figure 5.4-9 Refueling Water Storage Tank.
force provided by the tank weight is overcome by the base moment, the tank will
, uplift and mobilize anchor bolt resistance. Because the anchor bolts are not as
,*iff as the shell, the neutral axis will shift toward the compression side to l maintain equilibrium between the applied and resisting forces. Also, if the anchor bolts or their. chairs can deform in a ductile manner, additional load redistribution will occur as uplift increases and anchor bolts around the tank perimeter are progressively . stressed past yield. This behavior is acceptable since ductile 3
deformation of the anchor bolts or their chairs will not directly cause a loss of
-tank contents, unless local deformations become excessive. .
To more accurately determine a realistic tank capacity against overall seismic base l moment, the analytical model shown in Figure 5.4-10 was developed. Tank ; deformations at the base and at the elevation through the top of the anchor bolt chairs are assumed to be linearly ' distributed. This is appropriate since the foundation and the portion of tank above the top of the chairs are stiff compared to the anchor bolts. The distribution of forces at the base of the tank resulting from this assumption is shown in Figure 5.4-10. Compressive strains in the shell are found by distributing the displacements over the height of shell between the tank base and the top of the anchor bolt chairs. Shell compressive stresses are , found by factoring the strains by the clastic modulus. So long as the anchor bolts or their chairs are clastic, anchor bolt strains are determined by assuming that the total uplift displacement results in a uniform strain over the length of the bolt
- from the top of the chair to the anchor plate embedded in concrete.
! Uplift of the tank base also mobilizes some additional resistance associated with the contained fluid bearing on the tank bottom plate. The fluid hold-down force - t provided by the tank bottom plate is accounted for in the design of unanchored tanks by modeling the plate as a beam subjected to a vertical displacement at one l end and uniform pressure along its length by the fluid weight. . A plastic I mechanism is assumed to develop in this beam. While this representation is ; considered to be conservative for unanchored tanks, it may be unconservative for ' an anchored tank whose uplift displacements may not be sufficient to develop this plastic mechanism. Accordingly, the fluid holdown force-was based on the same i beam representation for unanchored tanks described in [Wozniak,1978], but limited , to elastic behavior only (Figure 5.4-11). The magnitude of the hold-down force l varies around the tank perimeter according to the distribution of uplift displacements. The overall tank base moment capacity is found by limiting the maximum shell corapressive stress to the value causing buckling and solving for force equilibrium between the applied loads and resisting forces. Design buckling stresses for cylindrical shells subjected to bending moment were developed in [ NASA,1966). The NASA design approach was adapted to predict the median RWST buckling stress. The design buckling stress coefficients in [ NASA, 1966] are values having a 90% confidence of exceedance. Based upon buckling stress data reported in [Weingarten, 1965] and [Gerard, 1957], the design coefficients for buckling due to bending are estimated to have a median factor of safety of 1.4. Increase in the tank buckling stress due to internal pressurization by the contained fluid was included. Stainless steel does not have a sharply defined yield plateau, and instead exhibits a rounded stress-strain curve. To account for potential inelastic buckling, the plasticity correction factor was derived from a Ramberg-Osgood-Hill representation of the material stress-strain curve. The tensile capacity of the anchor bolts may be limited by any of the following: 5-35
l l External Loads
" ^C 3
~
- ..? . - T.3 m- .- -
l i ! t l Neutral Axis I a. Assumed Displacement Distribution
- b. Distribution of Shell Compressive Stresses
~ l I P s Anchor Bolt / Bolt Chair Yield Force
- c. Distribution of Anchor Bolt Forces -
l
- d. Distribution of Fluid Holddown Force on Tank Bottom Plate Figure 5.4-10 Model for Determination of Tank Resistance Against -
Seismic Base Moment. 5-36
i WEIGHT OF WATER
/ l 1
iP 1F if if tr gr 1F a REACTION ON TANK SHELL TANK SHELL BOTTOM PLATE POINT OF CONTACT WITH GROUND ] UPLIFT DISPLACEMENT l, I
- c. BOTTOM R. DEFORMED SHAPE i
- b. BOTTOM R. M0 MENT DI AGRAM Figure 5.4-11 Beam Model of Tank Bottom Plate.
i 5-37
- ~ w
o Anchor bolt yield o Anchor bolt fracture at the threads o Anchor bolt pullout from the concrete foundation .
- o Yield of the tank shell locally at the anchor bolt chair-o Bending of the anchor bolt chair top plate o Failure of anchor bolt chair welds o Stability of the anchor bolt chair stiffener plates Based upon a review of these items, capacity of the anchor' bolts was found to be limited .by bending of the anchor bolt chair top plate. Under_. applied loading imposed by the anchor bolt, the top plate is subjected to bending as it spans between its transverse supports _ consisting of the vertical stiffener plates and the I tank shell. The available resistance was based upon yield line theory with yield i lines developing as shown in Figure 5.4-12. At the supports, -hinge moment capacities are governed by the bending capacities of the vertical stiffener plates and the tank shell since they are thinner than the top plate itself.
l A240 Type 304 stainless steel has a minimum specified yield strength at 0.2% offset of. 30 ksi and a minimum specified tensile strength of 75 ksi. The minimum i elongation is specified to be 40%. Based upon values reported by [ Smith, ASME] median yield and tensile strengths are estimated to be 37 ksi and 84 -ksi, respectively. Because stainless steel does not exhibit a sharply defined yield i plateau characteristic of carbon steel and because of the large strain ductility available, derivation of plastic moment capacities from the yield strength would be overly conservative. Instead, an effective yield stress was estimated. The moment-curvature relationship was derived for a flat plate using a Ramberg-Osgood-Hill fit to the median strength properties. From a-review of the resulting curve, an effective yield stress of 69 ksi was judged to provide a reasonable approximation to the moments at the relatively high curvature demands to which the top plate is i likely to be subjected. The median anchor bolt tensile capacity as limited by j bending of the chair top plate was estimated using this effective yield stress in 1 conjunction with the yield line representation. I Following this methodology, the RWST was found to have a HCLPF capacity of ' 0.21g against shell buckling. This capacity included a median system ductility of l.1 to account for some limited benefit of inelastic energy absorption due to nonlinear response of the tank as the anchor bolt chair top plates yield around the perimeter. Because the scismic response and capacity'of the tank is the same in
- any horizontal direction, the prescribed ground spectrum which is the larger of the i two horizontal components is the one that will lead to tank failure. Consequently, values for Fp and $ gp of 1.0 and 0, respectively, were used. Other potential tank failure modes considered and found to have greater capacity than shell buckling included sliding at the base and yield of the shell due to hoop stress induced by hydrostatic and hydrodynamic pressure.
The other flat bottom storage tanks were evaluated using essentially the same methodology. Although the DWST is smaller than the RWST and has the same type of anchorage, its HCLPF capacity is lower because it is fabricated from B209-5052-O aluminum. The PWST is similar to the DWST, but it is anchored by twenty , anchor bolts, one inch in diameter. t 5-38 L _ . - _ - . - _ _ _. . _ - - ~
ii i ll II /
^
s N I
!l s l
I i , I-1I sN / l Il i N/ l ll l I
/
YIELD LINES
- a. YIELD LINE PATTERN
~~~_ ~~
~~
~~~ ~
_4_ -t-
- b. DEFORMED SHAPE l
Figure 5.4-12 Yield Line Model of Anchor Bolt Chair Top Plate. 5-39
I i l l 5.4.1.5 Inverter }- i' The Maine Yankee inverters are located in the protective switchgear room of the turbine / service building at elevation 45'-6" (25 ft. above grade). The inverters are-12-kVA static inverters manufactured by Solid State Controls Inc. (SCI). They are . constructed of an all welded steel tubular frame with bolted shear panels on three~ sides and bolted doors on the front. The back panels have ventilated slits to-
~
. provide cooling for the inverter internals. SCI provided the initial qualification of the inverters by comparing the Maine Yankee 12-kVA units to a larger ,15-kVA
! unit shake table tested as part of the qualification program for the Three Mile Island Nuclear Station (reference Maine Yankee FSAR Amendment No. 35, Volume II). The tested 15-kVA inverter was similar in configuration and size to the Maine Yankee inverter, but approximately 1000 pounds heavier. Specific test data was . c not provided such that a fragility evaluation could be extracted from the tests. I 4 The results however, did confirm that the critical failure mode of the inverter is its anchorage. The test results further confirmed the high capacity of the internals , and attached components. The inverters were also compared to'the'GERS, October ' 1986 draft which supported the capacity of the internals and attached components. The Maine Yankee inverters are one electrical component which has undergone an 4 anchorage upgrade since its original installation. The original installation (which is still present) bolted the inverter to four base channels which were in turn was 4 expansion bolted to the. supporting floor. In 1983 Maine Yankee installed an anchorage upgrade to the inverters increasing the capacity. Figure 5.4-13 shows two photographs of the inverter and anchorage addition. A conservative static j analysis was performed by Maine Yankee in evaluating the anchorage addition. To l arrive at a more realistic capacity, a fragility evaluation was performed in the present study using available data and as-built details collected - during the walkdowns. The anchorage of the inverter.to the concrete floor was identified as the critical - failure mode. The failure mechanism was defined as the existing anchorage in combination with the anchorage upgrade. Canacity Factor. The failure of the existing anchorage to the concrete floor was
; determined to be the most critical failure mode. A flexibility study was performed , to demonstrate the existing anchorage was much stiffer than the anchorage ! upgrade. Calculations determined the existing anchorage was 99% stiffer than the i
anchorage upgrade. The primary area of flexibility found in the anchorage upgrade was the long extension of the angle required to clear the base channels in order to bolt the entire assembly to the floor. The method used to arrive at the inverter capacity consisted of computing two bounding cases for anchorage failure: 1 o Lower bound: The existing 1/2 in, diameter Red Head self-drilling anchors resist the entire seismic load. I 4 5-40 E . - _ . _ _ _ _ - - _ . _ _ - _ - . ._ __ _ __ ._ _ . _ _ - _ .-
I
?
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Figure 5.4-13 Maine Yankee 12 kVA Inverter and Battery Charger, and a Close-up of the Anchorage Addition. 5-41
-. - -- -. --~ . .--. _. _ - _ _ . - - - - - - _ _ - _ _
o Upper bound: The existing Red Head anchors combined with the anchorage upgrade fail at the same time. The actual failure mechanism will lie between these two bounding cases. The theorized failure mechanism will be the existing anchors loading up first since this load path is more stiff. However, some slippage in the existing anchors will occur as ultimate load is reached. This condition will begin to transfer more load into the anchorage upgrade as the displacement increases until failure occurs. It is difficult to ascertain _ how much slippage (thus load transfer) will occur; consequently, the second bounding condition was assumed. The median failure capacity was assumed to lie more closely to the lower bound, since this is the ' stiffer load path. Consequently, median capacity was assumed to be defined by the following: o A upper represents a +3 on A m (5.4-17) o A lower represents a -1 on A m (5.4-18) , Strenath Factor. The inverter was modeled as a single-degree-of-freedom system , where the strength factor was computed using static analysis techniques. The l turbine / service building floor spectra at 7% damping generated from the 0.18g NUREG/CR-0098 50th percentile ground response spectra was used to obtain the spectral accelerations at the inverters fundamental frequency. A conservative horizontal fundamental frequency was ' calculated using only 'the: welded frame and neglecting the bolted shear panels. This yields a lower bound on
- the median horizontal fundamental frequency. A horizontal fundamental
, frequency of approximately 13 Hz was computed. This was used for both j horizontal directions as the side to side direction was not believed to bc significantly stiffer. The rear shear panels were observed to have ventilating slits and the front doors had cut-outs for instruments. The vertical fundamental frequency of the inverters was judged to be greater than 30 Hz. Vertical response spectra was not developed for points in the floor' slab. In or. der to account for the amplification of the floor spectra at points in the slab, the amplification factors developed from a study by [ Structural Mechanics Associates, 1983] were used. The amplification factors are based -on the the floor vertical j fundamental frequency. A conservative floor vertical fundamental frequency of 18 Hz was computed for the slab in the protective switchgear room which yielded
!. an amplification factor of 1.5. The factor was applied to the ground response spectra at the inverters vertical fundamental frequency to arrive at the spectral acceleration. Since the inverters vertical fundamental frequency was - judged greater than 30 Hz, the 1.5 factor was applied to the 0.18 ZPA of the ground response spectra.
j The following presents the strength factor computed for the inverter HCLPF capacity determination based on the the combined failure mode of the existing 4 anchorage and the anchorage upgrade. The strength factor discussion below - l presents the strength factor determination for the existing anchorage, the anchorage upgrade, and the combined strength factor resulting from both. l l 5-42
Existina anchorane strenath factor. The existing anchorage consists of the 1/2 in. Red Head self-drilling anchors bolted through the inverter base channels into the I concrete floor. The lower bound strength factor ~ determination assumed this I failure mode to transfer the entire seismic loading. The seismic loads were applied to the C.G. of the inverter assumed located at the geometric center of the cabinet. This is slightly conservative because the majority of the inverter weight is bolted l to the bottom base framing (the heaviest component, transformer, was observed to be positively bolted to the inverter base framing). The seismic loadings were applied in the three orthogonal directions considering 100-40-40 as median centered to result in the most severe loadings. One hundred percent of the east-west seismic accelerations yielded the highest bolt loads. The loads were resolved using standard static techniques to obtain bolt shear and tension loads. Criteria for determining the strength factor for expansion anchors was that c recommended by [URS Corporation / John A. Blume & Associates, Draft 1986]. ! Shear-tension interaction was based on the quadratic formulation as the inverter bolt loadings were in the area of high tension and _ low shear (V/Vm <0.4). In this range of high tension and low shear the quadratic and the linear *[ormulation produce negligible differences in the results. The median anchor bolt values tabulated in the Blume report were considered median. These l values were used only if the Blume criteria was satisfied regarding edge distance requirements and anchor spacing. The uncertainty was computed based on a factor of 2 on the median values as representing a 95% confidence level. Anchorane Unarade Strenath Factor. The critical failure mode identified for the anchorage apgrade was shear in the fillet weld attaching the angle to the inverter - base channels. The fillet weld is heavily stressed due to the configuration of the anchorage upgrade. The angle attaching to the inverter base channels extends out away from the inverter, thus creating a long lever arm which induces shear in the weld. l The strength factor for the weld was computed based on considering median shear strength as 70% of the weld materials ultimate tensile capacity. Median ultimate tensile capacity for material strength was judged to be 20% higher than the code value. Combined Strennth Factor. The combined strength factor considered both of the anchorage systems as resisting the seismic loadings. As discussed previously, the existing anchorage is stiffer than the added anchorage upgrade; consequently, will begin to transfer the loading initially. After slippage occurs in the existing 4_ expansion anchors, the anchorage addition will begin to share in the load transfer until idealistically both anchorage systems will fait simultaneously. This represents an upper bound on the actual failure mechanism. The strength factor for this failure mode was computed by adding the individual strength factors computed previously. The realistic failure mode lies between the lower bound, the existing anchor bolts, and the upper bound where the combined anchorage systems fail together. The median strength factor was computed from Eqs. (5.4-17) and (5.4-18). i 5-43 1
i e Ductilitvc Inelastic energy absorption was considered as part of the strength-factor. Median ultimate material strength properties were used in computing the strength factor. The inverter overall capacity factor, random variability and uncertainty computed are: F = 12.26
= 0.00
= 0.46 kU Eauinment Resoonse Factors Oualification Factor. The inverters respond in a single-degree-of-freedom manner.
The static analysis used to compute the strength factor was considered median centered with no uncertainties or random variabilities. Soeetral Shane Factor. The unsmoothened and broadened floor response spectra were not provided for the margin study. There is typically. conservatism, particularly in the amplified regions of the spectra, in the smoothing and broadening. No attempt _ was made to recover this . conservatism; however,' . conservatism in the development of the synthetic time history for input to the dynamic structural models for the computation.of floor response spectra did occur. Cygna performed the structural modeling and subsequent floor response spectra generation. A synthetic time history was developed such that the response spectra enveloped the 0.18g NUREG/CR-0098 50th percentile ground response spectra. Cygna computed the statistical results of the conservatism in the enveloped spectra over the NUREG spectra. This conservatism was assumed to be included in the ~ floor response spectra. The spectral shape factor used and associated random variabilities and uncertainties derived from the statistical results are: FSS = 1.15 hR
= 0.066 hU
= 0.00 Damoinn Factor. Seven percent damping was considered median centered for 3 welded and bolted structures. The inverter analysis used 7% damping. Five j
percent damping was considered as representing a minus one beta uncertainty on the median. The random variability was considered to be two tenths of the ! uncertainty. The resulting values used are: i FD = 1.0
= 0.01
= 0.05 Modelina Factor. A median horizontal fundamental frequency was computed in determining the spectral accelerations for the inverter analysis.~ To establish the
- uncertainty on the computed median frequency a range of frequencies was considered. A range of 10 to 20 Hz was considered in order to compute the i uncertainty. The highest spectral acceleration in the frequency range was considered to represent a 95% confidence on the median spectral acceleration. The l
I ) l 5-44 l - . . _ _ - . . - _ . . - - - . - - - - - - - - - - - . --- - - - - = - - - - - - - - -
.Q s random variability was considered to be equal to zero. The modeling factor and the associated random variability, and uncertainty are J FM = 1.0
= 0.00
= 0.17
; Mode Combination. The inverter was modeled as a single-degree-of-freedom system. There is some uncertainty as to the contribution of higher modes to the overall response of the inverter not considered in the simplified static analysis. An uncertainty of 0.10 was assumed to reasonably represent the contribution of higher -
modes. .The mode combination factor, random variability, and uncertainty are: F MC = 1.0
= 0.00
= 0.10 Earthauake Comoonent Combination Factor. 100-40-40 rule was used to compute the strength factor. This was considered median centered. Random variability -
based on past .PRA studies, considering median coupling and both horizontal directions contributing to failure equal to 0.10 was used. The factor, random-
; variability, and uncertainty are:
1.0 FECC =
= 0.10 hR
~ 0 00
$U The combined equipment response factors for the inverter are:
1 F ER = 1.15 j R
= 0.12
= 0.20 l
Structural Resoonse Facton The structural response factors for the turbine / service building included: o Soil-structure interaction o Modeling l o Damping o Spectral shape o Directional effects affecting equipment response Refer to Section 5.4.1.1 for a discussion of the structural response factors. The combined structural response factors used for the inverters are: 5-45
F RS = 1.35
$R
= 0.25
$U = 0.24 Inverter Ground Acceleration Canacity. Combining the capacity factor, equipment response factor, and the structural response factor the resulting median ground acceleration capacity and HCLPF capacity for the inverter was computed as:
Am = 3.43g
$R = 0.28
$U = 0.59
- HCLPF Capacity = 0.82g l 5.4.1.6 Diesel Fuel Oil Day Tank The Maine Yankee diesel fuel oil day tank is a horizontal cylindrical tank j supported about 5. feet off the floor by a braced frame support structure. The day
! tank is located at grade in the diesel generator building. Figure 5.4-14 shows photographs of the tank taken. during the first walkdown. The critical failure il mode evaluation for the tank was based on a previous finite element analysis
[Impell,1985]. The results of that analysis identified the seven most critical areas of the tank / support system and compared the resulting seismic stresses to ASME Code allowables. Table 5.4-1 specifies the factor of safety (allowable stress divided by the seismic stress) for the seven critical areas assessed. Figures 5.415 and 5.4-16 are outline drawings of the day tank and show locations for these seven critical areas which were analyzed. The most critical day tank failure mode is shown from the factors of safety to be the anchor bolts, while the second most critical failure mode is the one-half-inch bolts on the 2" x 2" x 1/4" bracing. The demonstrated
; factor of safety on the bracing bolts is over three times the factor of safety on the anchor bolts. Thus, the anchor bolts are judged to be the primary failure mode 1
and the contribution to risk of all other failure modes is judged to be negligible
-due to their much higher capacity. It should be noted that in addition to the areas j analyzed in the Impell analysis, the fuel piping (threaded) exiting the day tank bottom was evaluated for displacement induced loads and 'found to have a relatively high seismic capacity.
Canacity Factor 1 l The capacity factor for the day tank anchor bolts is determined by the ratio of the bolt ultimate failure threshold to the response at the bolt due to the review level earthquake. The capacity factor and its associated variability are described in detail in Section 5.4.1.1. The ultimate failure threshold is based on the anchor bolt test data presented in [URS,1986]. Close inspection of the Maine Yankee day tank ' anchor bolts revealed that the expansion anchors utilized for the original plant design were not fully embedded into the floor slab (see Section 4.4 for complete description). Maine Yankee engineers devised a retrofit to ensure that the full strength of these anchor bolts would be developed by installing eight new Hilti Kwik bolts (two per leg) as shown in Figure 5.4-17 (On one tank only six bolts could be installed; The calculated capacity is still greater than 0.3g). The day
- tank fragility is based on this retrofitted condition.
i 5-46
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Day Tank Support Legs Day Tank with Attached Piping and Original Anchor Bolts and Splash Shield l t Figure 5.4-14 Diesel Fuel Oil Day Tank (TK 62B). 5-47
I i
~
-1/4"FILLEh
,e, lI ,8 et . .l ,
a n m , , , d u " 2'x2'x 1/4" ANGLE
*:
- 5 FOUR 3/8"
' BOLTS PER LEG 1/2" BOLTS /
(FOUR PLACES) ' : 2 1/2" NOMINAL PIPE TWO 1/2" j f ANCHORS PER LEG '. h" . r D,
/
h 1/4 FILLET WELD
- FAILURE MODES CIRCLED CORRESPOND TO THOSE ON TABLE 5.4-1
' Figure 5,415 Diesel Day Tank Critical Areas. 5-48
43" > FULL PENETRATION WELD Z - dir f FOUR 3/8"f BOLTS-1/2" NOMINAL PER LEG PIPE 21/2" NOMINAL PIPE TWO 1/2" ANCHORS PER LEG n; , '"i 1/4" FILLET WELD
- FAILURE MODES CIRCLED CORRESPOND TO THOSE ON TABLE 5.4-1 Figure 5.4-16 Diesel Day Tank Critical Areas.
l 5-49
,7 ) b! ,f i
'5.
+ 4 ti FUEL O!L DAY TANK (TK - 62A.8). ,y s
,; F 3: '
NEW 1/2"d HILT! IN' EXISTING HOLE (TYP.) - NOTE 3 +A-.
'~ 'DI 8' EXISTING 1/2"d STUD AND NUT 1 EXISTING 1/2"d STUD & NUT
/
- NEW 1/2" HILTI-( SUPPORT LEG GROUT (DEPTH VARIES)
(L SUPPORT LEG 1/2"A +A cm 3/4" FLANGE (BOLT HOLES'AT 90*, AZMITHAL ORIENTATION ABOUT' SUPPORT LEG VARIES) U rh rh I gfl . u r/l ///// s- s s s s s
! /1
's u n' d ob d 44
'[ * *
,' 4" MIN EMBE0 If I
Figure 5.4-17 Retrofit Anchorage Design for the Day Tank. z I
, 5-50
Table 5.4-1 Diesel Fuel Oil Day Tank Critically Stressed Area Critical Area Factor of Safety
- 1. Welding on 1 1/2" Pipe Braces 5.95 3.28 2.' 1/2" Diameter Bolts at 2"x2"xl/4" l Longitudinal Bracing
- 3. Fillet Weld Attaching 2"x2"xt/4" Bracing High to the Center Stiffener Plate
- 4. ~ Weld of Tank to Saddle Plates High
- 5. Bolting of Vertical 21/2" Pipe Posts 2.44 to the Saddle Plates
- 6. Welds at top of Tank Saddle 7.14*
- 7. Anchor Bolts 1.03
- Conservatively based on a 1/4" fillet weld instead of the actual full penetration weld.
i 5-51
r The individual median capacities of the Hilti Kwik bolts are reported in [URS L 1986] to be 5.97 kips in tension and 8.80 kips in shear. These values are reportedly
-based on the 2.25-inch minimum embedment for these bolts. Although the Maine
- Yank'ee day tank anchors have a specified minimum embedment of 4 inches, the 2.25-inch embedment strengths are conservatively judged to be applicable for this situation. The reason for the use of the 2.25 inch embedment data is that the relatively close proximity between anchors (4.5 inches between centers) on each day -
tank leg will lead to some degradation of the rated strength as embedment goes l past this 2.25-inch embedment depth. Although the derated strength of a four inch embedded bolt in this situation is greater than the 2.25-inch embedment strength, the 2.25 inch embedment strength has conservatively been assumed for this fragility analysis, h a The uncertainty on the median tensile and shear strength values stipulated above is quantified by estimating the 95% probability of exceedance level to be at 50% of the median values. This factor of one-half was judged appropriate as a method of-quantifying the various uncertainties associated with expansion anchor capacities . (i.e., concrete pours, installation procedures, drill tolerances, dynamic loads and material properties). The shear and tensile loads have been-combined using the method recommended in [URS,1986]. Section 5.4.1.5 describes this interaction equation for the Inverter fragility description. ~ Resoonse Factors The day tank natural frequencies were rederived from those presented in the ' [Impell,1985] analysis because the weight of the tank contents was neglected in the frequency calculation. This error is significant in calculating the tank frequency and subsequently in deriving the seismic accelerations since the weight of the
~
contents is much greater than the weight of the tank itself. The actual tank I frequencies shifted out of the ZPA region and into the amplified section of the spectra (14 Hz and 17 Hz). Spectral accelerations for all three earthquake components were based on the most recent floor spectra (0.18g NUREG/CR-0098) scaled to the 0.3g review earthquake level. Seven percent median damping was used with 5% as the 95% probability of exceedance value. Earthquake components were combined using the median centered 100%, 40%, 40% technique. The remaining equipment and structural response factors were calculated using the methodology presented in Section 5.4.1.1. Ground Acceleration Canacity The median ground acceleration capacity of the day tank was calculated to be 1.41g based on the failure of the anchor bolts. The HCLPF capacity is calculated to be 0.43g. Since this HCLPF capacity is greater than the 0.3g review level earthquake for Maine Yankee, the conservative assumption of 2.25-inch embedded ' anchors is considered acceptable. 5.4.1.7 Containment Spray Fans The Maine Yankee containment spray fans are located in the ventilation equipment room at elevation 21'-0"(grade). The fans are Westinghouse Silent Vane, size 8030, 4 5-52
1 13500 cfm units. The fan units are supported from vibration isolators with seismic i restraints installed in all directions. The seismic restraints are to be installed ! during the next Maine Yankee refueling outage (see Section 4.4). Figure 4.4-8 and l Figure 4.4-9 show photographs of the fans in the unmodified condition and a sketch of the proposed seismic restraints, respectively. The fragility evaluation ; was based on this modified condition. The entire fan assembly is anchored to a reinforced raised concrete pad. The anchorage of the fans was identified as the critical failure mode to investigate and will be presented below. However, another possible failure mode was identified during the walkdown. The fan units have cantilevered shafts. It was postulated that the cantilevered fan blade on the end of the shaft could fail or-cause premature bearing failure as a result of seismic excitations. The concern for the cantilever shaft on the containment spray fans was alleviated based on numerous similar configurations which had survived large magnitude earthquakes. Several vendors also verified the fact that this cantilever variety fan is very - common and represents approximately 50% of. the fans used in power plant applications. These cantilever fans are designed to withstand the high operating loads, and, the vendor felt the safety factor on the operating loads was more than - adequate to offset the seismic load. The containment spray fans anchorage HCLPF capacity was computed using a conservative simplified analysis as outlined in Section 5.2. Several conservative assumptions were used in the analysis to show a high HCLPF capacity. These included: o The existing anchorage was conservative ~y assumed to carry no loads. o Peak spectral acceleration values were used since the vibration isolation system frequency was not known. o Conservative dimensional data for the anchorage modification was used. Only a conceptual sketch was provided to the fragility team by Maine Yankee for the anchorage modification. Figure l 4.4-9 shows the sketch received from Maine -Yankee for the i anchorage modification. This modification will be verified during the third walkdown. l o Overturning was assumed to be resisted by the outer anchor bolts alone. Even with these conservative assumptions the fragility evaluation for the containment spray fans showed a HCLPF capacity greater then 0.5g pga which is greater than the margin carthquake for Maine Yankee. 5.4.2 CDFM Method For the purpose of comparison, HCLPF capacities were determined using the '
-CDFM method for the following selected components:
5-53
o Circulating water pumphouse o Block wall SB 35-3 o Refueling water storage tank Application of the CDFM method was based on the recommendations in NUREG/CR-4482. HCLPF capacities found by. the CDFM and fragility analysis methods are compared in the following discussion. One major difference between the two methods common to all of the examples is the use of the median NUREG/CR-0098 ground response spectrum to determine structure loads and in-a structure response spectra for the CDFM evaluation. For the fragility analysis, conservatism in this spectrum was accounted for in the structure' response factor < described in Section 5.4.1.1. The structure damping value of 7% that was included in the generation of in-structure response spectra is judged to be conservative for the structures considered. Other major differences between the two approaches in their application to the selected examples are discussed below. 5.4.2.1 Circulating Water Pumphouse Differences between the CDFM and fragility analysis methods for the circulating water pumphouse included the following: o Damping o Diagonal brace buckling capacity o Effective inelastic energy absorption factor Seismic response of the steel superstructure was determined using 10% damping. This value is recommended by NUREG/CR-0098 as being a conservative value for bolted steel structures at or near the yield point. The diagonal brace buckling capacity was determined using two different formulations: ^ o Factored AISC allowable o Conservatively biased value based on test data The AISC allowable buckling capacity for working stress design was factored by 1.7 as permitted by 'Section 2 of the AISC specifications for plastic design. , Alternatively, the test data in [ Hall,1981] used to establish the median capacity I was reviewed to establish the 84% exceedance value. These two formulations were found to provide essentially the same conservative capacity. a The HCLPF capacity against initial brace buckling was calculated- to be 0.22g.- This value is higher than the corresponding capacity of 0.19g determined by the fragility analysis method. As noted in Section 5.4.1.2, the HCLPF capacity for collapse of the pumphouse steel structure is judged to be greater than 0.3g. This corresponds to an increase factor accounting for the reserve capacity against collapse of 1.35. 5.4.2.2 Block Wall SB 35-3 Conservative seismic response of Wall SB 35-3 for the CDFM method was based upon in-structure response spectra generated for the median NUREG/CR-0098 5-54 l
l ground spectrum. A conservative estimate of block wall damping prior.to initial cracking was not required since clastic response of this wall occurs in the high frequency range of the floor spectra. The primary source of conservatism in the HCLPF capacity is the selection 'of a conservative modulus of rupture. The Standard Review Plan permits the use of ACI 531 allowable- block wall stresses with certain increase factors for extreme load combinations. However, no increase is permitted for the modulus of rupture perpendicular to the bed joint. This is considered to be- overly conservative since ACI 531.itself permits a one-third. increase in working stress allowables when stresses due to earthquake loading are included. Review of available test data indicates that the increase factor should be at least 1.33. The modulus of rupture perpendicular to the bed joint was conservatively based upon the ACI 531 allowable-value increased by a factor of 1.33. The HCLPF capacity using the CDFM method was determined to be 0.678 This value is greater than the capacity of 0.573 found by the fragility analysis - l . method. l 5.4.2.3 Refueling Water Storage Tank Major differences between the CDFM' and fragility - analysis methods' in the determination of HCLPF capacities for the refueling water storage tank were associated with the following quantities: i
! o Damping o Shell buckling stress o Material and component strengths o Effective ductility A conservative damping value of 5% was estimated -for the horizontal' impulsive -
i mode response. This is the value recommended by NUREG/CR-0098 for welded steel structures at or near the yic!d point. The same analytical model developed in the fragility analysis to determine the resistance against' overall seismic base moment was used. As in the fragility analysis,' input to this analytical model
. consisted of the shell buckling stress, bending capacity of the anchor bolt chair top plate, and fluid holdown force on the tank bottom plate. The shell buckling stress was determined using the NASA design coefficients which have a 90% confidence of exceedance. Plastic moment capacities for the anchor bolt chair top plate were based on an effective yield stress taken as the average of the minimum specified yield and tensile strengths. This value corresponds to an equivalent clastic- ,
perfectly plastic representation of the material stress-strain relationship. ' Additional conservative bias was introduced into the top plate capacity to account for uncertainty in the location of the loading imposed by the anchor bolt nut .tnd washer. Additional seismic capacity implied by nonlinear behavior of the tank as it uplifts and yields the bolt chairs prior to shell buckling was conservatively neglected. A HCLPF capacity of 0.21g was determined for the RWST. using the CDFM method. This value is the same as the capacity determined by the fragility analysis method. [ Manos, 1986] has developed an empirical method for the seismic design of unanchored tanks. Manos' comparison of his approach with data compiled on damage and undamaged tanks from actual earthquakes indicates that this method is CDFM equivalent. HCLPF capacities for the RWST, DWST, and PWST based l 5-55
1 l L upon Manos' approach are listed in Chapter 7. :For example, a cap.acity of 0.24g is determined for the RWST. This capacity is greater than the 0.21g HCLPF capacity - found by tiie fragility analysis and CDFM methods. Intuitively, the seismic n capacity of an anchored tank would be expected to be greater than that of an unanchored tank. However, the mechanics of seismic resistance of unanchored tanks is not well understood at this time. Because the nature of seismic resistance of anchored tanks may be much different than the source of resistance for unanchored tanks, ,it is not possible to derive firm conclusions on the HCLPF - 4
' capacities of the Maine Yankee tanks based upon Manos' work.
! 5.5 HCLPF Canacity of Plant j 5.5.1 Accident Sequences Following the review of plant information, a list of the components that make up the front line and support systems required to perform the plant safety functions (Group A functions) was developed. This list is given in Chapter 7 and provided i the basis for the remainder of the margin review. After the plant walkdowns and - subsequent analyses by the systems analysts and fragility analysts, the list of components was reduced by screening out those components that were found to have HCLPF capacities greater than 0.30g . pga. The components for which fragilitics a:id HCLPF capacities were determined are discussed in Section 5.4.
] These remaining components were used in the development of the event trees and fault trees for the scismic induced core damage accident sequences as described in "
Volume 2. The event trees and fault trees were analyzed to determine the cut sets for each accident sequence that could lead to core damage. From these cut sets,'- the Boolean expressions for the accident sequences were developed. The component failures that are significant to these accident sequences are given in Tables 5.5-1 and 5.5-2. Table 5.5-1 gives the seismic induced failures along with the fragility parameters used to represent their capacity. Table 5.5-2 gives the non seismic failures and their failure probability. Note that the component items and the non seismic failure events have been numbered consecutively; the missing numbers represent the items that were screened out in the final pruning of the event and fault trees. , The Boolean expressions for the two dominant accident sequences are: Small LOCA , . = 4 + 7 + 20 (5.5-1) No LOCA
= ( 4 + 20 ) * ( 8 + 15 + 17 + 22 )
+ 8 * ( 14 + 16 ) + 7
- 15 , (5.5-2) where each number in these expressions corresponds to the failure of a component given in Table 5.5-1 or Table 5.5-2. In the above expressions, the notation "+"
5-56
Table 5.5-1 Component Seismic Fragility Parameters l l Item HCLPF Am(8) Capacity (g) No. Item hR kU 4 Transformers 0.84 0.30 0.32 0.30 7 RWST 0.45 0.20 0.25 0.21 8 DWST 0.36 0.20 0.26 0.17 20 Circulating Water 0.69 0.24 0.27 0.30 Pumphouse 21 PWST* 0.57 0.20 0.26 0.27
- See Page 5-67 5-57
Table 5.5-2 Probabilities for Nonseismic Failures Item Median Error No. Description Unavailability Factor * (per demand) 10 Operator Failure to Close 8.0E-02 2-PCC Isol. Valves 11 Random Failure of DG-1B 4.2E-02 5 12 Random Failure of DG-1 A' 4.2E-02 5 13 Operator Failure to Place AFW l.5E-01 2 Pump Train B in Service Locally l l 14 Nonseismic Common Cause 1.6E J3 5 Failure of DGS 15 Nonseismic Common Cause 1.2E-04 5 Failure of AFW 16 Operator Failure to Refill DG 8.0E-03 3 Fuel Tanks by Opening Valve or Running P-33A,B 17 Operator Failure to Place AFW 4.0E-02 3 Pump Train B in Service from MCR I 22 Random Failure of the Turbine 3.0E-02 5 Driven Aux. Feedwater Pump
- Error factor equals (95% Confidence Value/ Median Value).
5 's I i . . - . . . -
t i L denotes probabilistic addition (union) and "*" indicates probabilistic multiplication
-(intersection).
4 Because the impulse lines inside the containment could not be inspected to confirm that small LOCA cannot occor, the Boolean expression for small LOCA and No - LOCA were combined in two different wsys.- 1 The first method uses split fractions to express the conditional probability of a seismic induced small LOCA given the seismic event. The two Boolean Eqs. (5.5-1) and (5.5-2) are combined using the split fraction p, and sensitivety studies are performed on this split fraction. Core Damage *
= p. [ Small LOCA] +
( 1 - p ) . [No LOCA] (5.5-3) The second method uses an additional term in the small LOCA Boolean expression 1 that represents the initiating event of seismic induced small LOCA and the other terms in Eq. (5.5-1) represent the failure of mitigating systems. The core damage Boolean expression is then the logical combination of the Boolean expressions for two accident sequences i.e., [Small LOCA] * [Small LOCA Boolean Expression] + [No LOCA] * [No LOCA Boolean Expression]. The HCLPF capacity of the plant
, against core damage is a function ofthe HCLPF capacity against small LOCA Unitiating event) if this capacity is less than 0.21g. The plant level core damage HCLPF capccity is governed by the HCLPF capacity of RWST. If the small LOCA HCLPF capacity is larger than 0.21g, the plant level HCLPF capacity is controlled i by tne small LOCA HCLPF capacity. The HCLPF capacity against small LOCA was not determined since we could not access containment. However, the plant level HCLPF capacity is expected to be greater than 0.21g.
In the following, the quantification of these Boolean expressions for the purposes of deriving the plant HCLPF capacities is discussed. 5.5.2 Probabilistic Method i Following the rules of Boolean algebra, the component fragilities are combined i using the numerical procedure proposed by Kaplan (1981). For this purpose, the l component fragilities are discretized into a family of fragility curves with a subjective probability estimated for each curve as discussed in Section 5.4.1. The probability assigned to each curve is developed based on the uncertainty distribution on the median capacity. Each fragility curve is assumed to be completely described by the median capacity and the value of hR. The fragility I curve is not truncated in either the lower or upper tail. The discrete probability distribution (DPD) for a component, say 4 , is expressed as: 4 4 ( <q; ,f; , (a) > ) (5.5-4) 5-59
and is shown graphically in Figure 5.5-1. Here, a is the peak ground acceleration, f; is the conditional probability of failure of the component under acceleration a and and fjq;+ is the For Af;/2. subjective example,probability the aggregation that the value of f;ies for two components 4is in of fragilit
+ 7 as given by a convenient numerical integration scheme [Kaplan,1981):
4+7 4+7 {<qik *Iik , (a) > ), (5.5-5) where 4+7 4 7 Gik " Ai 4k 4+7 4 7 4 7 fik " I -(I-f )(I*Ik i ) or max (f; , fk }' l l
- .w ,, ,
s' ,'
/ / <
l l /
/
s' 0 / / 2 / ^/ /
- 0.75 -
W/ / u
$ i 2'
/
9 01 /
/
I / q
$ i /
'B /
i y 0.50 -
, i o j l t , area _
g / / 94 f j+afg /2,,' g I ,' /
.5 / f j-Afj /j '
2 y o 0.25 -
,' / , Subjective
/ - Probability U ' ' '
- Distribution of
,/,/,! ,, / Failure Probability, f;
. ,W Peak Ground Acceleration, a Figure 5.5-1: Seismic Fragility Curves of a Component t
l 5-60
In the expression for fik, two bounds are given corresponding to perfect
' independence or perfect dependence between component failure events. These bounds'are used to assess the significance of correlation between failure events.
Expressions similar to Eq. (5.5-5) are derived for the case of joint events 4
- 8 By taking two components at a time,-the plant level fragility is developed'
{ <q;P, p P, (a)>). (5.5-6) Figure 5.5-2 -hows a plot of the plant level (small LOCA core damage) fragility curves in which the family of fragility curves is reduced to the 5%,50%, and 95% confidence fragility curves. From this plot, the HCLPF capacity (defined as peak. ground acceleration corresponding to 5% probability of failure at 95% confidence) for small LOCA core damage is obtained as 0.21g. ! Inspection of the small LOCA core damage Boolean expression [Eq.5.5-1] given ! above indicates that the dominant components are the singletons. The singleton component with the lowest HCLPF capacity is the refueling water storage tank (RWST) with a HCLPF capacity of 0.21g. Failure of this tank results in no coolant being available for reactor vessel injection following a LOCA. The other singleton components have HCLPF capacity equal to 0.30s. The capacity of the transformer is estimated based on the proposed upgraded condition. The seismic fragility of the circulating water pumphouse is conservatively j estimated. However, its failure has a less significant effect on the small LOCA core damage HCLPF capacity because of the smaller hR and g._ The HCLPF capacity for No LOCA core damage accident sequence was estimated i using the Boolean expression given in Equation 5.5-2 as 0.32g (Figure 5.5-3). The ! higher capacity against this sequence compared to the small LOCA sequence is , l because of the absence of low capacity singletons in the expression. Although DWST, i.e., component 8, with its HCLPF capacity of 0.173 appears in - this sequence, its failure has to occur simultaneously with one of the higher capacity I components i.e., transformer, and circulating water pumphouse. i
- Core Damage HCLPF Capacity As explained above, the core damage HCLPF capacity calculation requires a knowledge of the split fraction between the two accident sequences (i.e., small LOCA and Transient). For different assumed split fraction values, the core damage HCLPF capacities were obtained as shown in Table 5.5-3.
1 The conclusion regarding the dominance of RWST failure in the HCLPF capacity estimation (displayed in the small LOCA accident sequence) is a function of the split fraction assumed. If the plant HCLPF. capacity needs to be increased, it is not necessary to concentrate only on RWST. A walkdown and review of small . impulse lines within the containment may be performed to estimate .* eir fragilities
! in order to assign a realistic split fraction. By this procedure, the plant HCLPF capacity may be concluded to be higher without the necessity of any upgrading of
{ the components. 4 5 61 i _ _ . . . _ . _ _ _,. . _ . _ _ , . - _ _ . . - . . _ _ _ . _ . , , . _ _ _ . _ _ _ . _ . ~ .,._--_,_ .--m ._ _ _
to E o s.>- CM. .
.95 tt. .o .*
P-M J H . e-- m m o C
- c. .+.
J z o M .>- H H O o u , J -
".o :s :< :s :e
,sy a PEAK GROUND ACCELERATION (g) i l
Figure 5.5-2 Fragility Curves for Small LOCA Core Damage. t 5 62
.a Z
P n.>- .s. :
~
s 3 C / [
- .+.
m m o C
- a. .-
- E z a o
>=
- I a l u .
' ~ ;a .a .< :s :e s.o r io '
PEAK GROUND ACCELERATION (g) l l Figure 5.5-3 Seismic Fragility Curves for No LOCA Core Damage, 5-63
Table 5.5 Summary of Plant Level HCLPF Capacities A Case Description HCLPF Capacity (g) 1 Small LOCA 0.21
- Independent Seismic Failures 2 Small LOCA 0.21 i
- Dependent Seismic Failures 3 No LOCA 0.30
- Independent Seismic Failures with Nonseismic Failures 4 No LOCA 0.30
- Independent Seismic Failures without Nonseismic Failures 5 Core Damage I - split fraction p = 0.01 0.30 p = 0.10 0.28
- p = 0.50 0.23 6 Small LOCA
- RWST 0.26 Reduced Fluid Level 3
+ 5-64
, , ,, , - - , , - - . , , - . . , - , - - - . , , - , , . - - . , , , , - - , , - - - , , , - - - - - , - - - . - - , - , , , . , , , .m. . ~ . , , - , + .-.-r - - - - - -,.
Effect of Nonseismic Failures The overall plant HCLPF capacity was calculated using the Boolean r expressions f core failures.damage accident sequences which contained both seismic and nonseismic failures in this calculation.In the following we describe the reasons for including the nonsei There are two kinds of unavailability. The run. other is that the component is unavailableto error.start on deman unavailability per unit time (e.g., year).The first is expressed as failur For the seismic event, the component fragilities may be considered as failure rate per demand (varies with carthquake level). nonseismic unvailabilities, thenonscismic failure onFor demand the rates may be technique is to convert them into conditional probabilities component by multiplying is needed to perform. the failure rate by a time interval during which the Example is that the diesels are expected to start on seismicseinduced loss of off it power and continue to operate for some hours. Probability of failure of diesel system = P p3 + A.t + Pg (a) where Pg3 = probability of failure to start A = rate of failure to run t = required time for the diesels to be running Pg(a) = seismic fragility of dicscis The probabilities for nonseismic failures given in Table 5.5-2 include the The uncertainties in the unavailability (per run.demand) is ex of the error factor unavailability. which is equal to the ratio of the 95% confidence value to the media distribution. The unavailability is modeled by a lognormal probability Need for Consideration of Nonseismic Failures in Seismic Margin Studies in the seismic margin studies, we are interested in estimating the largest earthquake that the plant can withstand with a high degree ofA confidence particular component may be able to withstand this carthquake . the Assume that plant will not suffer core damage if any one of the two components survives thi carthquake. s dependent. Hence, the plant level HCLPF capacity may be con to the higher of the two component capacities. However, the plant HCLPF capacity failure rate. will be different if one of the components is affected by high nonscismi One component may fail in an earthquake, yet the other component 5 65
Therefore, it is may not be available for the plant to survive the carthquake. important to consider nonseismic failures in estimating seismic component is not of particular significance in a seismic margin study. The effect of including nonseismic failures is observed in the No LOCA core damage sequence in that the HCLPF capacity for the sequence changes from 0.33g (when the nonseismic failures are not included) to 0.32g (when the nonscismic Since the components with HCLPF capacities >0.3g have failures are included). been screened out, these plant HCLPF capacities have been represented in Table 5.5-3 as >0.3g. In the small LOCA sequence, there are no nonseismic failures. Correlation between Seismic Failures The above calculations were performed assuming perfect independence l'etween seismic failures of different components; i.e., the seismic capacities are assumed to be statistically independent both in randomness and uncertainty. This is a realistic assumption because the components involved in the core damage Boolean expression are yard tanks (RWST, and DWST), transformer, and circulating water pumphouse. These are dissimilar items of structures and equipment, their locations in the plant and within the structures are different, and their dynamic Hence correlation in the seismic responses and characteristics are also different. seismic capacities of these components is judged to be minimal. It is realistic to expect some correlation in the component failures. Assumption of perfect dependence in both uncertainty and randomness is an extreme case. Assumption of perfect dependence in the uncertainties of different component fragilities means that the median ground acceleration capacities of all components are known if the median ground acceleration capacity of one component is known. Since the uncertainty arises from insufficient understanding of structural material the properties, approximate modeling of the structure, inaccuracies in representation of mass and stiffness, and the use of engineering judgment in lieu of plant specific data, it is expected that all components will be affected to some Therefore, some probabilistic dependence between degree by these uncertainties. component median capacities may be expected. Perfect dependence in uncertainty is however an extreme assumption. generating the Dependence responses in different in the randomness arises from a common earthquakecompo Assumption of dependence in the randomness means that ifpeak for a given the fragility ground (conditional probability of failure) of a componentacceleration is known, the somewhat modified by that knowledge if it were known. The plant level fragility, and therefore, the HCLPF capacity depends on the degree of dependence in randomness and uncertainty between the component One approach failures. is to However, the degree of dependence is difficult to estimate. bound the core damage HCLPF capacities by assuming perfect dependence as This calculation was performed for opposed to the case of perfect independence. i The small LOCA the small LOCA core damage Boolean expression S ven above. core damage HCLPF capacity is estimated to be 0.21g i.e., governed by the capacity i 5 66 l
3 Effect of Nonseismic Failures The overall plant HCLPF capacity was calculated using the Boolean expressions for core damage accident sequences which contained both seismic and nonscismic failures. In the following we describe the reasons for including the nonseismic failures in this calculation. There are two kinds of unavailability. The first kind is that the component fails to start on demand due to random failure, common cause failure or operator error. The other is that the component is unavailable due to maintenance or failure to run. The first is expressed as failure rate per demand; the second is in terms of-i unavailability per unit time (e.g., year). For the seismic event, the component fragilities may be considered as failure rate per demand (varies with earthquake level). Hence component fragilities and nonseismic failure on demand rates may be added probabilistically. For the nonseismic unvailabilities, the technique is to convert them -into conditional probabilities by multiplying the failure rate by a time interval during which the component is needed to perform. Example is that the diesels are expected to start on seismic induced loss of offsite power and continue to operate for some hours. Probability of failure of diesel system = Pfs + A.t + Pp(a) where P fs = probability of failure to start 4 A = rate of failure to run t = required time for the diesels to be running Pg(a) = seismic fragility of diesels i The probabilities for nonseismic failures given in Table 5.5-2 include the probability of failure to start, operator error and the probability of failure to run. The uncertainties in the unavailability (per demand) is expressed in terms of the
- error factor which is equal to the ratio of the 95% confidence value to the median unavailability. The unavailability is modeled by a lognormal probability distribution.
Need for Consideration of Nonseismic Failures in Seismic Margin Studies In the seismic margin studies, we are interested in estimating the largest carthquake that the plant can withstand with a high degree of confidence. A l particular component may be able to withstand this earthquake. Assume that the l plant will not suffer core damage if any one of the two components survives this l earthquake. Conservatively, we may treat the component failures as statistically dependent. Hence, the plant level HCLPF capacity may be considered to be equal to the higher of the two component capacities. However, the plant HCLPF capacity will be different if one of the components is affected by high nonseismic failure rate. One component may fail in an earthquake, yet the other component 1 5-65
-~-- _- ... - - _ . . -. - ..
i 4 I l may not be available for the plant to survive the earthquake. Therefore, it is important to consider nonseismic failures in estimating seismic margins. . Yet, any i I nonseismic failure that does not occur si'multaneously with a seismic failure of a l component is not of particular significance in a seismic margin study. The effect of including nonseismic failures is observed in the No LOCA core - damage sequence in that the HCLPF capacity for the sequence changes from 0.33g (when the nonscismic failures are not included) to 0.323 (when the nonseismic . failures are included). Since the components with HCLPF capacities >0.3g have - been screened out, these plant HCLPF capacities have been represented in Table 5.5-3 as >0.33. In the small LOCA sequence, there are no nonscismic failures. ,' Correlation between Seismic Failures )' The above calculations were performed assuming perfect independence between seismic failures of different components; i.e., the seismic capacities are assumed to
; be statistically independent both in randomness and uncertainty. This is a realistic assumption because the components involved in the core damage Boolean l
expression are yard tanks (RWST, and DWST), transformer, and circulating water pumphouse. These are dissimilar items of structures and equipment, their locations in the plant and within the structures are different, and their dynamic characteristics are also different. Hence correlation in the seismic responses and seismic capacities of these components is judged to be minimal. i It is realistic to expect some correlation in the component failures. Assumption of
- perfect dependence in both uncertainty and randomness is an extreme case.
- Assumption of perfect dependence in the uncertainties of different component fragilities means that the median ground acceleration capacities of all components I
are known if the median ground acceleration capacity of one component is known. l Since the uncertainty arises from insufficient understanding of structural material 4 properties, approximate modeling of the structure, inaccuracies in the representation of mass and stiffness, and the use of engineering judgment in lieu
- of plant specific data, it is expected that all components will be affected to some i degree by these uncertainties. Therefore, some probabilistic dependence between j component median capacities may be expected. Perfect dependence in uncertainty l is however an extreme assumption.
I Dependence in the randomness arises from a common earthquake generating the i responses in different components and common structural / material properties. i Assumption of dependence in the randomness .means that if the fragility. (conditional probability of failure) of a component for a given peak ground
- acceleration is known, the probability of failure of the other components would be somewhat modified by that knowledge if it were known.
The plant level fragility, and therefore, the HCLPF capacity depends on the degree ! of dependence in randomness and uncertainty between the component failures. i However, the degree of dependence is difficult to estimate. One approach is to j bound the core damage HCLPF capacities by assuming perfect dependence as opposed to the case of perfect independence. This calculation was performed for
- the small LOCA core damage Boolean expression given above. The small LOCA core damage HCLPF capacity is estimated to be 0.2131.e., governed by the capacity i
5 66 1 i
l 1 of RWST. The plant level HCLPF is determined to be approximately 0.213 When ! the Boolean expression is dominated by singletons, the assumption of perfect independence is more severe than the assumption of perfect dependence between failures if the fragilities are approximately equal; if there is a single component with a very low capacity compared to the rest of the components in the Boolean expression consisting of singletons, both the assumptions give about the same plant level HCLPF capacity. The question of dependence between failures is important when there are similar . components experiencing common seismic excitation. The case in point is the yard tanks (i.e., RWST, DWST, and PWST). By reviewing the cut sets, it was found that a tripleton cut set could lead to core damage. It is DWST
- PWST
- RWST
- The HCLPF capacity for this cutset would be higher than the highest of the three tank HCLPF capacities (i.e.,0.273 for PWST).
5.5.3 Deterministic Method This approach is based on the assumption that the HCLPF capacities of components estimated in Section 5.4 are true lower bound values. The HCLPF capacity of the plant is obtained directly by studying the Boolean expressions for small LOCA and no LOCA: Small LOCA
= 4 + 7 + 20 No LOCA
= ( 4 + 20 ) * ( 8 + 15 + 17 + 22 )
+ 8 ' ( 14 + 16 ) + 7
- 15
- For calculating the plant level HCLPF capacity in this method, the nonseismic i failures are ignored. Also, the cutset that includes a low probability nonseismic j failure is also omitted from this estimation. Therefore, the simplified Boolean
- expression is
No LOCA
= ( 4 + 20 )
- 8 In this method the HCLPF capacity of a "doubleton" cut set is calculated as the higher of the two component HCLPF capacities. The HCLPF capacity of a
'tripleton" cut set is calculated as the highest of the three component HCLPF capacities. The HCLPF capacity of a union of singleton cut sets is estimated to be l the lowest of all the component HCLPF capacities. Using this procedure, the HCLPF capacities of the core damage accident sequences are:
5-67
HCLPF capacity against small LOCA core damage
= ' min [ 0.30, 0.21, 0.30 ]-
= 0.21g HCLPF capacity against No LOCA core damage
= max [ min ( 0.30, 0.30 ), 0.17 ]
- = max [ 0.30, 0.17]
= 0.30g Note that the failure of DWST with its HCLPF capacity of 0.173 is not governing the plant level HCLPF capacity because DWST has to fait simultaneously with one j
of two nonseismic failures to Igad to core depage. Since these failures have per demand respectively, it is median probabilities of 6 x 10~ and 8 x 10~ .
- appropriate to ignore the DWST failure in the plant level HCLPF capacity
. calculation.
5.5.4 Sensitivity Studies
- Two sensitivity studies were performed to assess the effect of certain assumptions j on the plant HCLPF capacity. These sensitivity studies addressed the following:
- 1. Effect of shearwall stiffness reduction on structure response and equipment seismic input.
2 Reduction of RWST fluid level.
; On-going scale model testing being conducted by the NRC has indicated potential reductions in the stiffness of concrete shear walls of up to a factor of four, from elastically calculated values due to cracking. This would imply a reduction in the ; elastic structure frequencies of up to 50%.
i ! For the case with an assumed small LOCA, the RWST is the dominant contributor to the plant HCLPF capacity. RWST constructed of stainless steel and is located in the yard on its own foundation. Its HCLPF capacity is not affected by the a potential structural frequency shift. For the case No LOCA, the 4160/480 V transformer is the dominant contributor to the plant HCLPF capacity. The
; transformer has a 12-13 Hz fundamental frequency which corresponds to a spectral 4
acceleration on the downward slope of the floor response spectra. A reduction of
; the building frequencies would result in a reduction of the seismic input to the l
transformer with a correspodig increase in its HCLPF capacity.- As a further check on the effect of the building frequency shift, a similar evaluation was made on { each of the components within the final Boolean expression (Table 5.5-1).' The
- potential building frequency reduction did not lower the HCLPF capacity _ for any of these components. Thus, for the purpose of the Maine Yankee seismic margin study, the shcar wall stiffness reduction and resulting building frequency shift does not affect the plant seismic margin.
i { 5 68 1
~ . ,- - .-~ _ ,-.-- - - - .. ..r,... - . - - _ . . - . _ . _ , - _ . . . . . , - . - . - . - - - - . , . - - . . . . - . . . _ - , . . . _ . , - . - - . ~ __ - - - - . - .
Because the seismic capacity of the RWST controls the plant HCLPF capacity for the small LOCA case, the effect of reducing the fluid level within the tank was investigated. A reduction of the fluid level to a total height of 33 ft, which is approximately equal to 10% reduction from the current level of 37 ft, was assumed. This change leads to a reduction in the effective fluid weight, mass centroid height, and overall tank seismic loads. The RWST HCLPF capacity is increased to 0.283 with a corresponding increase of the small LOCA HCLPF capacity to 0.263 i l + 5 69
~-- -- -rv -
y e p--- ,, ,. . , _ . ,__ _ _ , , _ _ , _ _
CHAP,TER 6 COMMENTS AND RECOMMENDATIONS ON SEISMIC MARGIN REVIEW METHODOLOGY 6.1 Selection of Review Earthauake Level In the course of this trial application of the seismic margin review methodology, it was found that sufficient guidance was not included in NUREG/CR-4334 and NUREG/CR- 4482 for the selection of the review earthquake level. NUREG/CR-4334 states that:
"The Panel has focused its efforts on earthquakes that could occur in the eastern part of the U.S., specifically east of the Rocky Mountains. Because of limited data on large magnitude events, the assessment of component capacities is limited to carthquakes of less than a Richter magnitude of about 6.5, which are characterized by three to five strong motion cycles with a total duration of 10 to 15 seconds. As the Richter magnitude increases above 6.5, the Panel recommendations may be slightly non-conservative. The frequency content ut the earthquake round motion is assume to be represented by median broadband response spectra, and the structures are assumed to have fundamental frequencies above i Hz. Note that for high frequency, high acceleration, low magnitude carthquakes, the Panel's recommendations are overly conservative."
The report goes on to state that:
"The margin review must begin with a target carthquake, in order to provide focus. It is assumed for the purposes of the review activity that some external source (perhaps the NRC staff, or the utility) has designated the carthquake level (in terms of peak ground acceleration, pga) which is the level at which the review is aimed".
The Panel in the NUREG/CR-4482 has stated that:
"The choice of the review level is a critical one since it is ,
used as a basis for screening out components. The review level should be specified in terms of pga and enough spectral information to assure the applicability of the material of Chapter 5 in NUREG/CR 4334 which is ! partially given in Table 2-1 of this report. If the review carthquake spectral content is not consistent with the assumptions made in NUREG/CR 4434, then this difference j needs to be taken into account. Table 2-1 was 4 6-1
3 i constructed 'to cover most spectra generated by magnitude I- 6.5 earthquakes or less. . In addition, spectral information will be needed to calculate HCLPF values." For the seismic margin review of Maine- Yankee, the NRC- specified . that the
- review earthquake level . would be .0.3g with a
- median Newmark-Hall ground '
response spectrum as defined in NUREG/CR-0098 for rock sites. : Since this 1 spectrum is applicable generically for all rock sites, it was concluded by the NRC l that the selected spectrum is a 84% confidence spectrum for Maine. Yankee (See -
- Chapter 2). The requirements on the review earthquake-level are also somewhat i implicit in the Panel's recommendations.' Yet, for. future applications, it is
; ' suggested based on this trial plant review that more explicit guidance be provided.
4 ) 6.2 Use of Screeninn Guidelines i l 6.2.1 Extent of Review Needed i As discussed before, this study has two primary objectives: to confirm the j appropriateness and adequacy of guidelines given in the Expert Panel reports and 1 to estimate the scismic margin for Maine Yankee. Because of this dual objective, i the study team had to expend many engineering hours in reviewing the components l that were supposedly to have been screened out in the initial screening, i.e., those components identified as having seismic capacities generically higher than the- !~ review carthquake level and denoted by letter "C". For example, the review by the i study team included components such as valves, diesel generators,' pumps, HVAC
! ducting, cable trays and cabling although the Panel screening guidelines state that j these components have seismic capacities larger than 0.30g pga. It may be pointed I out that the Panel screening guidelines are invariably conservative in that the j specific components in a plant may actually have capacities larger than indicated by the Panel. Also, the collective experience and judgment of the Expert Panel in ., developing the guidelines cannot be expected to be matched by any single team ; performing the seismic margin reviews in future. Hence, in keeping with' the primary objective of minimizing the review effort needed in a seismic margin ; review, it is recommended that the components identified as having capacities
! larger than the review carthquake level be screened out with no further review in-future seismic margin studies. ) j 6.2.2 Additional Screening Guidelines i The Expert Panel reports have not explicitly discussed the seismic capacities of steel structures. In the trial plant review on Maine Yankee, three steel structures ! were analyzed to assess their HCLPF capacities. These structures do have i capacities in excess of 0.30g. At the same time, they could not be screened out j based on a walkdown review because of unusual connection details and structural { arrangement. Therefore, some guidance is necessary in the margin review methodology for steel i structures. i f l } 4 6-2 ) %
. _ , , _ _ _ , . , , . ~ _ . . _. _ ,,, - _ .._ . _ _ _ _ . . _ _ _ . _ _ _ . . _ . _ . . . - - _ . . . _ . - . _ _ _ _ . _ .
- . _. .~ . -. . . -_ - .
6.2.3 Difficulties of Reviewing Certain Components The majority of the components identified for review in the Maine Yankee seismic margins study were reviewed during the plant walkdowns. Component design e reports and seismic qualification analyses were also revicwed on a majority of the ' Maine Yankee equipment as a part of. the fragility assessment. Difficulties in i reviewing certain Maine Yankee components fell into one of two categories: o Components whose qualification reports were available only from the NSSS supplier o Components located inside the containment structure 6.2.3.1 Reactor Internals and CEDM The reactor internals and the control element drive mechanism (CEDM) are two components requiring fragility assessments for the Maine Yankee margin study. The utility (MYAC) did not have access to the seismic qualification documentation i for two reasons: , +
- 1. This information is considered proprietary by the NSSS vendor f
- 2. Due to the vintage of the Maine Yankee Plant, seismic qualification was not required for the CEDM and for portions of
< the reactor internals 1 5 The NSSS supplier, Combustion Engineering (CE), is the only credible source of capacity / response information on the internals and the CEDM. Discussions l between CE, MYAC and EQE resulted in a contract being set up through LLNL for CE to collect the qualification data from their files and to meet.with EQE and Maine Yankee representatives. CE was very cooperative during this meeting and provided the following: + o Summaries of the qualification report results } o Similarity assessments for components like the CEDM which had no previous seismic qualification This interfacing with the NSSS supplier will generally be required on future : seismic margin studies, just as it has been required on almost all of the previous seismic PRA's. In addition, for plants where the utility does not take such an active role in the design / margin review, meetings with the architect-engineer to obtain plant / component information may also be required. 6.2.3.2 Equipment Within Containment Due to radioactivity concerns, critical components inside the Maine Yankee containment are accessible for review only at the time of a plant outage. The margin study schedule precluded the possibility of one of the plant walkdowns coinciding with the March 1987 plant refueling outage. This fact necessitated the use of previous photographs and design drawings for fragility derivation purposes. 63
d l Fortunately, Maine Yankee personnel have maintained a relatively complete file of component photographs based on previous seismic integrity studies. The experiences at Maine Yankee point out the need in future margin studies to plan for the review of components within containment. Difficulties in reviewing plant components will be more severe for plants which do not maintain organized documentation on the components within containment. i Impulse line failures were assumed to be the source of_ a small LOCA at Maine i Yankee. This conservative assumption was required 'due to the tremendous number
- - of hours which would be required to walk down each of these impulse lines and -
; assess potential system interaction problems. These lines originate from the primary pressure boundary (i.e., RPV, steam generator, pressurizer, primary coolant loop piping, etc.) and are field routed to instrument racks inside containment. The amount of work required to demonstrate the seismic margin in each of these lines plus the fact that a walkdown of these lines would have to take place during a plant outage necessitated the assumption of a small LOCA as an initiating event. ) 6.3 Availability of Oualification Data In structure response spectra for the median NUREG/CR-0098 ground response spectrum were generated for use in the trial plant review by Maine Yankee during the course of this study. The structure building models were developed recently for other Maine Yankee programs. The availability of in-structure response-i spectra generated by current techniques can reduce the amount of effort necessary to quantify component fragilities since correction'or regeneration of responses was not necessary. Adequate floor spectra may not always be available for older , plants. In such cases, it may be necessary to develop new dynamic models and j
perform dynamic analysis to define the seismic input to equipment. This will increase the amount of time and funding necessary to perform seismic margin i studies. Most of the Maine Yankee Group A structures were screened out, so detailed 1
- seismic load distributions to the structural members were not necessary for the '
- trial plant review. For the screened in structures, seismic load distributions were either available from recent dynamic analyses, or could be estimated with j relatively little effort due to simplicity in structure load paths. Consequently, j
determination of seismic load distributions was a minor consideration. However, this would not have been the case had it been necessary to rely upon load
' distributions from the original design calculations. The original structure design calculations for Maine Yankee comprised a large volume of 'information.
l Determination of the seismic loads and verifying their accuracy within this body of material would have been time consuming and tedious. Typically, original j structure design calculations have not been used in past PRAs for this reason. When necessary, approximate load distributions have been independently } developed. Generation of structure load distributions with the level of accuracy i appropriate for a seismic margin review could require additional effort. Detailed information on the Maine Yankee block walls was available. These data were very useful in conducting the trial plant review since the data provided locations of all block walls in the plant and identified any safety related { components that could be affected by their failure. In particular, this latter set of i s i j 6-4 l 1 -. - - . . - , . -_- - - - -_- - - . - .-. - - - -. - -
- - . - ~ . . . . . - . , - . . - . - - . .. - - - . - . --.
information was valuable since it permitted quick identification of several Group A block walls. It is not known if comparable information exists for most older ; plants. Seismic qualification information for Maine Yankee equipment was available for selected critical components. Maine Yankee engineers and their consultants have conducted specific qualification tests and analysis as the -need to do 'such an - analysis has been identified. There is very little seismic qualification analysis available from the original plant design documentation. This fact stems from the l fairly relaxed requirements in the seismic qualification area at the time of. the Maine Yankee plant construction. Seismic qualification in many cases consisted of simply a letter from the vendor that his equipment ' component . had adequate seismic capacity without any supporting documentation. Examples of equipment components for which seismic qualification information was available included the diesel generator day tank, hattery racks, RHR heat ' exchanger, emergency service water pump,- component cooling heat exchanger, j demineralized water storage tank and the diesel generator remote control panel. Seismic anchorage calculations were also available for all' of . the electrical- ! equipment which had their anchorage / supports upgraded from their original design - i (e.g., inverters, motor control centers, control room cabinets, transformers, etc.). j Combustion Engineering provided seismic analysis data for some of the reactor internal components, but the remaining reactor internals structures and the control 1 element drive mechanism did not have qualification data because there was no i detailed seismic analysis requirement for Maine Yankee vintage plants. Equipment l for which seismic qualification data was not available included the majority of-the : valves, fans, dampers, heat exchangers, small tanks, pumps, and air conditioners at Maine Yankee. l I j 6.4 Walkdown Procedures i j Based upon the trial plant review, suggestions on improvements to the review i methodology in the area of walkdown procedures would include the following: o Third walkdown , j o Access to personnel with specialized expertisc j o Data sheets i o Level of walkdown for bulk items i During the course of the trial plant review, certain components with relatively low HCLPF capacities were identified. The utility elected -to develop conceptual 3 modifications to be installed in the next outage. Plant HCLPF capacities reported i for this study reflect the increased component capacities based upon the conceptual. i modifications. A third walkdown will be necessary to confirm - that these l modifications are installed, and that they provide seismic capacity at least l comparable to the conceptual retrofits upon which the' calculated component - capacities were based. 1 J' 65 i
i
\
During the walkdowns, plant and utility personnel having expertise or specialized
)
knowledge in particular fields were interviewed. ' Specific areas of expertise ) included. i ! o - Fire water systems b F o Electrical o HVAC o Instrumentation and control
; o Control room personnel o Block walls The insight gained from discussions with these individuals . proved to be very useful in assessing the overall state of the plant and resolving any particular seismic issues. Scheduled meetings with plant experts in each of these areas allow the fragility and systems analysts to gain in depth understanding of plant specific 4 parameters such as design criteria, construction practices, equipment locations and -
functions, and operator actions in emergency events. i ! Walkdown data sheets (Section 4.3.3), while not noted in the review suidelines, were developed prior to the first walkdown. These data sheets served as a checklist of. items to review during the walkdown and facilitated the gathering of
- information necessary ' for HCLPF evaluation. While no set format need be -
) established for future margin reviews, comment on the use of data sheets in the review guidelines should be considered. 1 I Bulk items such as cable trays, piping, and ducting extend throughout the plant, and local configurations and support details can vary. Detailed walkdown of these l bulk commodities would be time-consuming. The extent of walkdown required for l review level carthquakes less than 0.33 is typically specified in general terms. As an example, the Expert Panel recommends that example cable trays be inspected to j verify that they are adequately anchored and braced and that specific ^ vulnerabilities do not exist. In the trial plant review, a sineral survey of the cable tray systems was performed and supplemented by detailed inspection of a typical system. More guidance in the seismic margins methodology on the level of walkdown for bulk items would be useful.
- 6.5 Guidance on CDFM Canacity Calculation Procedures The second report by the Expert Panel [Prassinos et al.,1985] rece nmends two
. candidate methods for calculating the HCLPF capacities for components: the conservative deterministic failure margin (CDFM) method and the fragility analysis (FA) method. The fragility analysis method was used in this study as was
- done in over 20 seismic PRAs. .This method, although very familiar to the present i analysis team, calls for subjective judgments to estimate parameters such as Am' j k and % for each variable needed in the seismic capacity estimation. The
'cDFM method, on the other hand, prescribes the parameter values and procedures to be used in calculating the HCLPF capacities. It is conceptually very attractive
, because it aims to avoid subjective judgments on the part of the analysts: first, the i analysts may not be equipped to make such judgments and second, there may be j inconsistencies between the subjective fragility assignments by different analysts with varied background and expertise. For the sake of consistency and + 66
l repeatability, we need a prescriptive method to estimate the seismic capacitics of structures and equipment. Ideally, the CDFM method would fit the bill when it is fully developed. The guidan:e on the use of CDFM method in NUREG/CR-4482 is not sufficient for this purpose. In the course of this study, five example components (i.e., refueling water storage tank, steel structure, diesel day tank, inverter, and block wall) were analyzed using the two candidate methods (i.e., CDFM and FA). Our experience was that several judgmental decisions had to be made in arriving at the parameters of the CDFM method. In each case we were not sure whether we met the intent of the method, i.e., conservative estimation of the . capacity, yet more liberal than the SRP requirements: In some cases we may have been overly conservative as was pointed out by the Peer Review Group. The difficulties arise because of two factors: o The CDFM method has not been fully developed for all structures and equipment items, o The parameters of-the CDFM method such as damping, material strength, static capacity equations, system ductility, and methods for floor spectra generation are not explicitly specified;- even where they are specified they may be overly. conservative. Also, the appropriate conservatism in the selection of the CDFM parameters needs to be determined using calibration methods. Development of deterministic procedures such as the CDFM method may be f.:complished using probabilistic models in such a way that over a large number of components both the CDFM and FA metinods yield identical HCLPF capacities. Once such a calibration is achieved, the CDFM method may be confidently used in, future seismic margin reviews. The calibrition should start with a fragility evaluation of a representative set of components using the models and parameters agreeable to the group (i.e., Seismic Margins Expert Panel). The parameters of the CDFM method are to be selected such that over this representative set of components the HCLPF capacities derived using the CDFM method and the fragility analysis method are approximately equal. We recommend that such a n % ration study be performed. 6.6 Staffina Reauirements and Schedule 6.6.1 Staffing Requirements NUREG/CR-4482 gives estimated staffing requirements for a seismic margin review; for a plant founded on rock with a review carthquake level of 0.303 pga, the estimate to perform the fragility analysis is 19 engineering-months. The actual engineering months expended in the present study is about 25% more than the Panel's estimate. Several factors need to be considered in this regard: o The dual objective of the study, i.e., seismic margin review of Maine Yankee and a trial application of seismic margin review methodology, required additional review efforts and interface m'.etings not to be expected in future studies. 6-7
o Maine Yankee is an older plant designed and constructed before the development of quality assurance- programs and seismic qualification methods presently existing in the nuclear industry. Hence, the qualification reports on certain equipment items were not available. Because of the reevaluation efforts, additional
= information and new structural models have become available.
However, the extent of information available is not comparable to that aonaally evailable for a modern nuclear power plant (e.g.,NTOL plant). The staffing requirements in the NUREG/CR-4482 should therefore be revised to reflect the availability of design and qualification data in the particular plant. o The utility performed the seismic response analysis for structures in the plant and developed the floor response spectra for the selected review earthquake. The engineering hours spent on this effort should studies, be included in the estimates for future margin o The plant design information was provided by the utility with some assistance from the NSSS vendor. Also, the utility personnel assigned for the interface were extremely familiar with the plant arrangement, systems and design. Typically, the interface will extend to the plant architect-engineer and constructor. The level of effort to collect plant design information and perform the plant walkdown may - be. much larger for other plants as more organizations become involved. Our seismic PRA experience indicates that the Maine Yankee interface experience is very-fortunate and not typical. 6.6.2 Schedule This study has been conducted over an eight month period; the plant walkdowns have also been conducted as per the schedule outlined in the Panel's report (Prassinos et al.,1985].' The schedule appears to oc reasonable. 6.7 Aonlicability of Methodolony for Other Plants The methodology has been developed using the insights obtained from seismic PRAsII. Mark on six PWRs supplied by Westinghouse and Babcock & Wilcox and a BWR This study on Maine Yankee has added to this collective experience in that a PWR supplied by Combustion Engineering has been analyzed. From a fragility standpoint, the CE reactor internals and the control rod drive mechanism have augmented the PRA fragility information base. The screening guidelines may be revised to reflect this additional information. The trial plant review has been for a plant on a rock site and for a review carthquake level of 0.30s. The methodology, the review guidelines, and the staffing requirements have not been verified for other conditions, i.e., soil . conditions and higher review earthquake level. Also, the methodology has not beendeveloped for screening systems in a BWR. In a BWR, the effect of 6-8
high frequency hydrodynamic loads on the equipment needs to be considered. The systems and components to be reviewed in a BWR would be different and the level . of effort and staffing requirements would vary, The applicability of the Panel's
- recommendations to these reactor / site / review levels needs to be demonstrated.
I l i l 6-9
. .. - - - =- - . . - - .. .-
a CHAPTER 7 RESULTS AND CONCLUSIONS ! The primary objective of this study has been to apply the seismic margin review methodology developed by the NRC Expert Panel on a trial basis in order to confirm the applicability of the methodolor,y for future seismic reviews and to provide input to the Panel for modifications in the methodology and review procedures. The other objective of the study has been to determine the seismic margin of the Maine Yankee Atomic Power Station. This report focuses on the fragility aspects of the seismic margin review. The report has described the plant review, screening of the components based on their generic high seismic capacities, plant walkdowns and calculations of the HCLPF capacities of the components, systems, accident sequences, and the plant. During this review, potential seismic vulnerabilities have been identified and for some of them the utility has proposed / incorporated certain modifications. 7.1 Screened Out Systems and Components Based on the screening guidelines of the Expert Panel, systems (both front-line and supporting sytems) not supporting Group A functions were screened out by Energy International Inc. These systems are discussed in Volume 2 of this Report. Among those components supporting the Group A function systems, some of them were
! screened out based on their seismic capacities as generically larger than 0.3 g pga, the selected review level earthquake. Table 7.1-1 lists the Group A structures, the screening procedures used, and the HCLPF capacities. Table 7.1-2 provides a summary of Group A block walls; included are the wall ID number, the Group A 1 i
components that may be affected by the wall failure and the bases for screening and HCLPF capacity. Table 7.1-3 lists the equipment in Maine Yankee by generic categories, the systems they support, their location, the screening procedures used (i.e., screening tables in NUREG/CR-4334, walkdown review, calculation review) and the HCLPF capacity. 7.2 HCLPF Caoacities of Screened-in Comoonents The definition of screened-in components depends on the stage of the reiew. Screening was done at four stages: o Preliminary screening based on the plant design informttion and I using the Panel guidelines o Preliminary screening confirmed by first walkdown and additional items screened out based on walkdown observation o Screening based on conservation capacity calculations using simplified methods o Screening based on final capacity calculations that show the HCLPF capacities to exceed 0.30g 7-1
l l Table 7.1-1 Summary cf Groip A Structures Screenino NUREG Watkdows Fragility NCLPF Structure 4334 Review Evatustion Capacity Comments Contalrument structure C >0.3g Reinforced concrete strur.ture. Contairunent Internal structure c >0.3e Reinforced concrete structure. Primary auxiliary building / fuel building C ,0.3g 2 Reinforced concrete structure. Circulating water pumphouse X X >0.30s Cottepse capacity judged to be >0.3e beoed upon post earttupake perfoneance, test data, and freellity evaluation. Turbine / service building, concrete portions c Reinforced concrete structure. Turbine / service building, steel framed X 0.38g NCLPF capacity bened span detailed w floor at EL 39'-0=
$a freettity evolustion.
Turbine / service building, other steel X >0.3g framed portions NCLPF capacity >0.3g bened span conservative fragility evetuation. Containment spray playhouse C >0.3g Reinforced concrete structure. Ventitation equipment room C >0.3g Reinforced concrete structure. Main steam valve house, exterior concrete C >0.3g Reinforced concreter structure, structure Main steam valve house, interior steet X >0.3g MCLPF capacity >0.3g based toon structure conservative fragility evaluation. M.C.C. room C >0.3g Reinforced concrete structure. Aux feed ptmphouse, purge air exhaust crea C >0.3g Reinforced concrete structure. Fuel oil porphouse C >0.3g Reinforced concrete structure.
Tabte 7.1-2 Sumary of Gro@ A Stock Watts NCLPF Walt ID No. Gro @ A Components Screening Capacity Comuments C 1, Assorted safety class piping and out These wells are loose blocks shields. C 20-2, 3 cw ipment Adjacent components are protected ty shletding consisting of steel framing. C 2, 3 Containment tiner out Breech of the containsect Liner dae to tapact by the block ustt is considered unlikely C 0.5 1, Pressuriser instrssesntation and In >0.3s IICL/F capacity beoed won simplified C 20-1 t eing freettity evetustion. steem generator enubbers and t eing In >0.3s seismic retrofits are installed. IICLPF capacity C 46-1, 2 C 61 1 to 3 based gent sleptified freallity evolustion. SB 21 7 PCC and SCC piping, PCC surge Line out The PCC and SCC Lines are about 24" In diameter. The PCC surge Line is about 6" in diameter. These piping systems are of welded construction. Failure due to ispect by the block watt is considered highly smtlkely. Main control board (aux feed penets) In >0.3g Seismic retrofits are instatted. NCLPF capacity $8 21-17 based won simplified fragility evaluation. Hein control boord (aux feed panels), In >0.3g Seismic retrofits are instatted. NCLPF capacity se 21 18 aux Logic penets based upon aisplified fraglLity evaluation. Battery gro g e 3 and 4, cable trays In >0.3g HCLPF capacity judged greater than value for ustt S8 35-1
$8 35-4.
Tabte 7.1-2 sumary of Group A stock Watts (Continued) I 2 1 HCLPF
; ustl ID No. Group A Components Screening Capacity Comuments .I l
1 ss 35-2 sattery gro w s 3 and 4, cable trays In >0.3g NCLPF capacity based on detailed fragility evolustion. SS 35-3 settery gro w s 3 and 4, cable trays In >0.3g NCLPF capacity beoed on detailed fraglLity evolustion. r Se 35-4 sattery gro w s 3 and 4 In >0.3g NCLPF capacity based on detailed frag lLity evaluation. ss 35 7 PCC surge line, PCC temperature In >0.3g NCLPF capacity judged greater than value for well controtter sa 35-3. SS 45-1 125v DC distribution cabinets 1 to 4, In >0.3e selenic retrofits are instetted. NCLPF capacity bened sattery gro g #2, Inverters et and #2, on sleptified fragility evetuation. Battery chargers #1 and #2, sus 8 ss 45-2 settery gro w #1, MCC 84, 480v In >0.3s setemic retrofits are instetted. NCLPF capacity beoed [ emergency switchgear on sleptified freellity evaluation. 4 se 45-3 Battery grog #1 In >0.39 seismic retrofits are instatted. NCLPF capacity based on sleptified freettity evolustion. S8 45 6 PCC surge line Out The PCC sarge line is shielded from the well by a seismic retrofit consisting of two steet angles risining paratlet with the pipe. Falture of both the anstes and the surge Line dse to lupact by the block well is considered highly tsilikely, ss 61-2 PCC surge tuA TK-5 out The PCC surge tank is shielded from the block wett by a solemic retrofit that consists of welded wire fabric spenning between steel framing. The shielding is judged to be sufficient to restrain the blocks from lapacting the tank.
- l Table 7.1 2 Sumery of Group A Block Walls (Continued)
HCLPF , Group A Components Screenire repecity Comments Walt ID No. Out The surge tanks are shletded from the S8 77-2 PCC surge tank TK-5, SCC surge tank TK 59 block well by a seismic retrofit that consists nf welded wire fabric spenning between steel framing. The shielding is judged to be sufficient to restrain the blocks from lapacting the tanks. PCC heat exchangers E-4A and 48 Out The heavy construction of the heat T8 21 2 exchangers is judged to be capable of withstanding impact by the block wett. No other vulnerable co p ts are exposed. Fans 44A and 44g, ducting and filter In >0.3g Seismic retrofits are to be instatted VE 21-1 [ in the next outage. Detailed fragility evetuntion of conceptusL modification sketches indicates that the HCLPF capacity is >0.3g. In >0.3g MCLPF capacity based upon simplified VE 21-4 SCC lines fragility evaluation.
4 Table 7.1 3 Suunary of Maine Yankee Equipment Screening and NCLPF Capacities. Equipamt Qualification Screening Building and MUREG Watkdown Calculation NCLPF EgJipment Item System Elevation 4334 Review Review Capacity Comuments TANKS
- 1. Refueling Cavity Water Storage Tank NPSI Yd. + 20' X FA 0.21g (1)
TK 4 i CDFM 0.21g (2)
.(0.24g) (3)
- 2. Primary Component Cooling Surge Tank PCC 58 + 61' x >0.3g Coretive Fragitity Cateviation TK 5 Determined NCLPF Capacity is >0.3g.
- 3. Primary Water Storage Tank AFW Yd. + 20' Tk-16 X 0.77s (1)
(0.253 ) (3) y 4. Deelneralized Water Storage Tank Arv Yd. + 20' X 0.17s (1)
& Tk 21 (0.25g) (3)-
- 5. Air Start Receiver Tank AFW VA + 21' X >0.3s Tk 25 conservative Fragilitf Calculatics:
Determined PCLPF Ca;s:ity is >0.3g.
- 6. Auxiliary Fuel Oil St@ ply Tank (buried) F0 APR + 21' C
>0.3s Tk 28A NCLPF Capacity J.dged >0.3g for guried Tanks.
- 7. Auxitir.ry Fuet oft St@pty Tank (buried). F0 APR + 21' C >0.3g Tk 28g NCLPF Capacity .;udged >0.33 for Buried Tanks.
(1) FA
- NCLFF capacity calculated per the Fragility Analysis approach.
(2) CDFM - NCLPF capacity calculated per the Conservativw Deterministle railure h rgin apy occh. (3) NCLPF capacity for unenchored vertical fluid stc, rage tanks per Manos' approach, o
Table 7.1-3 Sumnery of Maine Yankee Equ!pment Screening and HCLPF Capacities (Continued) Equipment Qualification Screening Buitding and NUREG Walkdown Calculation NCLPF Equipment Item System Elevation 4334 Review Review Capacity Comments TANKS (Continued) Spray Chemical Addition Tank Yd. + 20' X >0.3g conservative Fragility Calculation
- 8. HPSI Determined NCLPF Capacity is >0.3g.
Tk 54 Secondary Component Cooling Surge Tank 58 + 7G' X >0.3g Conservative Freeltity Calculation
- 9. SCC Determined NCLPF Capacity is >0.3g.
ik-59
- 10. Emergency Dieset Day Tarh FO A8 + 21' X FA 0.43g Reflects the feodified Condition.
CDFM 0.54g Tk-62A
- 11. Emergency Diesel Day Tank Fo A8 + 218 X FA 0.43g Reflects the Modified Condition.
CDFM 0.54g y Ti. 628 U DG-1A Compressed Air Tank DG A8 + 21' X >0.3g Conservative Fragility Calculation 12. Determined NCLPF Capacity is >0.3g. Tk-76At DG-1A Compressed Air Tank DG A8 + 21' X >0.3g Cunservative Fragility calculation 13. Determined NCLPF Capacity is >0.3g. Tk-76A2 (same s et. as A1) DG-1A Compressed Air fa.1 DG A8 + 218 X >0.3g Conservative Fragility calculation 14. Determined NCLPF Capacity is >0.3g. Tk-76A3 (Same s @ t. as A1) Dieset Starting Air Receiver 1A DG A8 + 21' X >0.3g Conservative Fragility Calculation 15. Determined NCLPF Capacity is >0.3g.' Tk-76A4 Dieset Starting Air Receiver 1A DG A8 + 21' X >0.3g . Conservative Fragility calculation 16. Determined NCLPF Capacity is >0.3g. Tk-76A5 (same supt, as A4) l
Table 7.1 3 Suenary of Maine Yankee Equipment Screening and MCLvF Capacities (Continued) Equipment Qualification Scresning Building and NUREG Walkdown Calculation HCLPF Equipment Item System Elevation 4334 Review Review Capacity Comunents TANKS (Continued)
- 17. Diesel Starting Air Receiver 1A DG A8 + 21' >0.3g X conservative Fragility calculation Tk-76A6 (Same s@t. as A4) Determined NCLPF Capacity is >0.3g.
- 18. Og 1B Compressed Air Tank DG AB + 21' >0.3g X Conservative Freellity calculation ik-7641 Determined NCLPF Capacity is >0.3g.
- 19. Og 1B Compressed Air Tank DG AS + 21' X >0.39 Conservative FraglLity Calculation ik-7652 (same s @t. as B1) Determined NCLPF Capacity is >0.3g.
- 20. Og te Compressed Air Tank DG A5 + 21' X >0.3g Conservative Fragility Calculation Tk-7653 (Same s@t. as 81) Determined NCLPF Capacity is >0.3g.
- 21. Dieset Starting Air Receiver 1B DG As + 21' >0.3g X conservative Fragility calculation ik 7654 Determined NCLPF Capacity is >0.3g.
- 22. Diesel Starting Air Receiver 18 DG A8 + 21' >0.3g X Conservative Fragility Calculation Tk 7685 (Same s @t. as 84) Determined NCLPF Capacity is >0.3g.
- 23. Dieset Starting Air Receiver 18 DG A8 + 21' X >0.3g Conservative Fragility Calculation Tk 7686 (same supt as 54)
Determined NCLPF Capacity is >0.3g.
I Table 7.1-3 Summary of Maine Yankee Ecpipment Screening and HCLPF Capacities (Continued) Equipment Quellfication Screening Building and NUREG Watkdown Calculation HCLPF Equipment item System Elevation 4334 Review Review Capacity Comuments PUNPS Primary Component Cooling Pump TB + 21' C >0.3g NCLPF Capecity Judged >0.3g
- 1. PCC P 9A for Horizontal Puups.
Primary Component Cooling Puup TB + 21' C >0.39 HCLPF Capacity Judged >0.3g
- 2. PCC for Horizontal Pumps.
P 98 Secondary component Cooling Ptap TB + 21' >0.3g McLPF Capacity Judced >0.3g
- 3. SCC C P-10A for Horizontal Ptaps.
Secondary Component Cooling Ptsp TB + 218 C >0.3g MCLPF Capacity Judged >0.3g
- 4. SCC P-108 for Horizontal Puups.
i y 6
- 5. Charging Pump HPSI pas + 21' C >0.3g HCLPF Capacity Judbed >0.3g for P 14A Vertical Ptap W/ Shaft Length <20'.
Charging Puup PA8 + 21' C >0.3g McLPF Capacity Judged >0.3g for
- 6. HPSI P-148 vertical Pump W/ Shaft Length <20'.
l I PAS + 21' C >0.3g MCLPF Capacity Judged >0.3g for-
- 7. Charging Pump HPSI P-14S Vertical Ptsp W/ Shaft Length <20'.
Emergency Feed Ptap P-25A AF + 21' >0.3g Horizontal Ptsp, HCLPF Cap. >0.3g.
- 8. AFW C X >0.3g 011 Cooler MCLPF Cap. Judged Oil Cooler (Skid mounted) E 86A
>0.3g Assessed on Watkdown Review.
Emergency Feed Turbine Ptsp P-25B VA + 21' >0.3g Horizontal Ptap, HCLPF Cap. >0.3g.
- 9. AFW C X >0.3g Oil cooler HCLPF Cap. Judged Oil Cooter (Skid mounted) E-868
>0.3g Assessed on Watkdown Review.
Table 7.1 3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) Equipment Qualification Screening Buttding and MUREG Watkdoun Calculation NCLPF Equipment Item System Elevation 4334 Review . Review Capacity Comuments PUMPS (Continued)
- 10. Emergency Feed Puup P 25C HPSI AF + 21' C >0.3g norizontal Pump, NCLPF Cap. >0.3g.
Olt Cooler (Skid mounted) E-86C X >0.3g oft Cooter NCLPF Cap. Ju W
>0.3g Assessed on Welkdoun Review.
- 11. Service Water Puup SW CW + 7' C X >0.3e Conservative Fragility Calculation P-294 Determined Motor to Puup Botts KLPF Capacity is >0.33
- 12. Service Water Pump SW CW + 7'- C X >0.3e Conservative Fragility calculation P 295 Determined Motor to Puup Botts y KLPF Capacity is >0.33 o
- 13. Service Water Puup SW CW + 7' C X >0.3g Conservative Fragility Calculation P 29C Determined Motor to Puup Botts ELPF Capacity is >0.3g.
14 Service Water Pump SW CW + 78 C X >0.3g Conservative Fragility Calculation P 290 Determined Motor to Pump Botts KLPF Capacity is >0.3g. l
- 15. i Auxiliary Fuel 011 Transfer Pump FO Yd. + 21' C >0.39 HCLPF Capacity Judged >0.3g for P-33A (Vertical Pump over Tk-28A)
Vertical Pump W/Shaf t Length <20'.
- 16. Auxiliary Fuel Oil Transfer Puup FO Yd. + 21' C >0.3g HCLPF Capacity Judged >0.3g for P 338 (Vertical Puup over Tk-28A)
Vertical Puup W/Shaf t Length <20'. i
- 17. Containment Spray Puup CS CS + 14' C X >0.3g Conservative Fragility Calculation P-61A Determined Motor to Pump Botts HCLPF Capacity is >0.3g.
Table 7.1 3 Suesary of Maine Yankee Equipment Screening and NCLPF Capacities (Continued) Equipment Quellfication Screening Buttding and NUREG Wetkdown Catculation NCLPF Egaipment Item System Elevation 4334 Review Review Capacity Comments PUMPS (Continued)
- 18. Containment spray Pu p Cs Cs + 14' C X >0.3g Conservative Fragility calcutetten P-618 Determined Motor to Pu p Solts NCLPF Capacity is >0.3g.
- 19. Containment spray Pump Cs Cs + 14' C X >0.3e Conservative Fragility calcutetton P-615 Determined Motor to Pig Botts NCLPF Capacity is >0.33 NEAT EXCNANGERS Y
O 1. Residant Neet Removat Neet Exchanger
- PCC Cs + 148 X >0.3s Conservative Freettity calcutetten E-3A Determined the Sipports and Anchor Bolts NCLPF Capacity is >0.3g.
- 2. Reeldsel Neat Removat Neet Exchanger
- SCC CS + 14' X >0.3s Conservative Fragility calculation E 35 Determined the Sipports and Anchor Bolts NCLPF Capacity is >0.3g.
- 3. Primary Component Cooling Neat Exchanger PCC TB + 21' X >0.3s Conservative Fragility calculation 1 E-4A Determined the sipports and Anchor Bolts NCLPF Capacity is >0.3g.
- 4. Primary Component Cooling Neet Exchangerr PCC Te + 21' X >0.3g Conservative Frasitity Calculation E-45 Determined the supports and Anchor Botts McLPF Capacity is >0.3g.
- 5. secondary Component Cooting Heat SCC ts + 21' X >0.3g . Conservative Fragility Calculation Exchanger E-5A Determined the Supports and Anchor Botts NCLPF Capacity is >0.3g.
Table 7.1-3 Sunnary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) l Equipment Qualification Screening Building and NUREG Watkdown Calculation HCLPF Equipment Item System Elevation 4334 Review Review Capacity Conssents
)
HEAT EXCHANGERS (Continued)
- 6. Secondary Component Cooling Heat SCC TB + 21' X >0.3g Conservative Fragility Calculation Exchanger E-58 Determined the Supports and Anchor Bolts HCLPF Capacity is >0.39
- 7. Fuel Pool Heat Exchanger
- PCC FB + 21' X >0.3g Conservative Fragility calculation E-25 Determined the Supports and Anchor Botts HCLPF Capacity is >0.3g.
- 8. Seal Water Heat Exchanger ** PCC PAB + 11' ----
Valves PCC M-PO & 219 Have a HCLPF E 34 Capacity >0.3g and will Isolate y this Component. i3
- 9. Reactor Containment Air Recirculation
- PCC RC + 46' X >0.3g Conservative Review Determined Cooler E 54 1 the Cooler Supports and Anchorage HCLPF Capacity is >0.3g.
- 10. Reactor Containment Air Recirculation
- PCC RC + 468 X >0.3g Conservative Review Determined Cooter E 54-2 the Cooter Supports and Anchorage HCLPF Capacity is >0.3g.
- 11. Reactor Containment Air Recirculation
- PCC RC + 46' X >0.3g Conservative Review Determined Cooler E- W 3 the Cooler Supports and Anchortge HCLPF Capacity is >0.3g.
- 12. Reactor Contairunent Air Recirculation
- PCC RC + 46' X >0.3g Conservative Review Determined Cooler E-54-4 the Cooter supports and Anchorage HCLPF Capacity is >0.3g.
- 13. Reactor Containment Air Recirculation
- PCC RC + 46' X >0.3g Conservative Review Determined Cooter E-54-5 the Cooter Supports and Anchorage HCLPF Capacity is >0.3g.
I
Table 7.1 3 Summary of Maine Yankee Equipment Screening and NCLPF Capacities (Continued) Equipment Quotification Screening ButLding and NUREG Walkdoem Calculation NCLPF Ecpalpment Item System Elevation 4334 Review Review Capacity ra===nts NEAT EXCHANGERS (Continued)
- 14. Reactor Contalrument Air Rulrculation
- PCC RC + 46' X >0.3g Conservative Review Determined Cooter E 54-6 the Cooler Styports and Anchorage NCLPF Capacity is >0.3g.
- 15. Reector Cootar.: Regenerative Heat *
- NPSI RC + 218 --
Valves PCC-N 90 & 219 Nave a NCLPF Exchanger E 67 Capacity >0.3g and will Isolate this component.
- 16. Safeguards Pumps Seat Leakage Cooter
- SCC CS - 02' X >0.3g Cooter NCLPF Capacity Judged E-91A >0.3g Assessed on Watkdoun Review.
4
-- 17. Safeguards Ptaps Seat Leakage Cooter
- PCC CS + 00* X >0.3g Cooter NCLPF Capacity Judged E 918 >0.3s Assessed on Walkdown Review.
- 18. Charging Pump Seal Leakage Cooter
- SCC PAS + 118 X >0.3g Conservative Fragility Calculation E-92A Determined the Stpports and Anchor Sotte NCLPF Capacity is >0.3g.
- 19. Charging Puup Seal Leakage Cooler
- PCC PAB + 118 X >0.3g Conservative Fragility Calculation E-928 Determined the !;tyports and Anchor Botts NCLPF Capacity is >0.3g.
- 20. Seal Water Neater ** HPSI PA8 + 118 ---
Valves PCC-N 90 & 219 Nave a NCLPF E-96 Capacity >0.3g and will Isolete this Component.
- 21. Auxiliary Chargir3 Ptap P 7 Lube Oil
- CM PAB + 11' X >0.3g Cooter NCLPF Capacity Judged Cooter (on skid) >0.3g Assessed on Walkdown Review.
~
i Tabie 7.1 3 Sumary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) Equipment Qualification Screening
' Building and NUREG Walkdown Calculation HCLPF l Equipnent item Capacity l System Elevation 4334 Review Review Comments CEA1 EXCHANGERS (Continuco) l
- 22. Containment Penetration Cooling Lines
- PCC various X >0.3g Cooter HCLPF Capacity Judged (Line integrity only) >0.3g Assessed on Walkdown Review.
- 23. LPSI Purrp Coolers LPSI
- CS + 21' x >0.39 Cooler HCLPF Capacity Judged (coolers are coilt *munted on the pumps) >0.3g Assessed on Watkdown Review.
- 24. Containment Spray Pm p Coolers
- CS CS + 14' x >0.3g Cooter HCLPF Capacity Judged (cooters are colts mounted en the P-61 Pw ps) >0.3g Assessed on Watkdown Review.
Jt!SCELLANEOUS COMPONENTS Diesel Generator DG-1A DG Ac + 22' C >0.3g HCLPF Capacity is Judged >0.3g Diesel Generator Heat Exchanger E-82A
- for the Dieset Gen. and DG 1A Engine Control Panet Peripherals.
- 2. Diesel Centrator DG-1B DG A9 + 22' C >0.39 HCLPF Capacity is Judged >0.3g Diesel Generator Heat Exchanger E 8?S
- for the Diesel Gen. and DG 18 Engine Control Panet Peripherats.
- 3. Seal Water Supply Filter ** HPSI unknown ----
Valves PCC-M-90 & 219 Have a HCLPF FL-348 Capacity >0.39 and will Isolate this Corponent.
, rj . s '~' . l . /; -
Q , s.
-r '
c.- .. , , . l 1 Table 7.% 3 Summary of Maine *ankee Equipment Screening and NCLPF Capacities (Continued) Equipment Quellfication Screening Building and NUREG Walkdown Calculation NCLPF Feuipment Item Systeun Elevation 4334 Review Review Capacity Cosmaants ELECTRICAL DISTE! Buff O SYSTEMS
- 1. 4160V Emers w.y sus Elec SB + 46' X >0.3g Conservative Fragility Calculation Bus 5 Determined the Anchorage NCLPF Capecity is >0.3g.
- 2. 4160V Emergency Bus Elec SB + 46' X >0.3g Conservative Fragility calculation isus 6 Determined the Anchorage NCLPF Capacity is >0.3g.
- 3. 480V Emergency Bus Elec $8 + 46' X >0.3g Conservative Fragility calculation Bus 7 Determined the Anchorage y NCLPF Capecity is >0.3g.
U
- 4. 480V Emergency sus Elec S8 + 46e X >0.3g Conservative Fragility Calculation Bus 8 Determined the Anchorage NCLPF Capacity is >0.3g.
- 5. 480V Emergency Motor Control Center Elec SB + 46' X >0.3g Conservative Fragility Calculation MCC 7A Determined the Anchorage HCLPF Capacity is >0.3g.
- 6. 480V Emergency Motor Control Center Elec RMC + 218 X >0.3g Conse vetive Fragility Calculation MCC-78 Determined the Anchorage MCLPF Capacity is >0.3g.
- 7. 480V Emergency Motor Control Center Elec CS + 20' X >0.3g Conservative Fragility Calculation MCC 781 Determined the Anchorage HCLPF Capacity is >0.33
i Tabte 7.1-3 swunary of Maine Yankee Equipment Screening ard HCLPF Capacities (Continued) s Equipnent Qualification Screening Building a.d NUREG Valkdown Calculation HCLPF Equi;wnent Item System Elevation 4334 Review Review Capacity Conments ELECTRICAL DISTRIBUTIcel SYSTEMS (Continued)
- 8. 480V [mergency Motor control Center Elec S8 + 46' X >0.39 Conservative Fragility Calculation MCC 8A Determined the Anchorage HCLPF CaFacity is >0.39
- 9. 480V Emergency Motor Control Center Elec RMC + 36' X >0.3g Conservative Fragility calculation MCC-8G Determined the Anchorage HCLPF Capacity is >0.3g.
- 10. 480V Emergency Motor Control Center Elec CS + 20' X >0.3g Conservative Fragility calculation MCC-88! Determined the Anchorage y HCLPF Capacity is >0.3g.
E
- 11. 120V AC Vital rus (WALL mounted) Elec $8 + 21' X >0.3g Bus HCLPF Capacity Judged >0.3g Bus 1A Assessed During Watkdown Review.
- 12. 120V AC Vital Bsrs (Wall momted) Elec 58 + 21' X >0.39 Bus NCLPF Capacity Judged >0.3g Bus 2A Assessed During Watkdown Review.
- 13. 120V AC Vital Bus (Wall mounted) Elec $8 + 21' X >0.39 Bus HCLPF Capacity Judged >0.3g Sus 3A Assessed During Watkdown Review.
14 120V AC Vit41 Bus (WmLL mounted) Elec FB + 218 X >0.3g Bus HCLPF Capacity Judged >0.3g E,us 4A Assessed During Walkdoen Review.
- 15. 12SV DC Bus Distribution Cacinet Elec $8 + 46' X >0.39 Conservative FraglLity Calculation sus 1 Determined the Anchorage NCLPF C @acity is >0.3g.
l
Table T.1-3 Summary of Maine Yankee Equirment Screening and NCloF Capacities (Continued) l l Equipment oualification Screening Suilding and NUREG Watkdo r. Calculation NCLPF l Equipment Itea System Elevation 4334 Review Review Capacity Comments j l ELECTRICAL DISTRIBUTION SYSTEMS (Continued)
- 16. 125V OC Bus Distribution Cabiret Elec S8 + 4' X >0.3g Conservative Fragility Calculation Bus 2 Detemined the Anchorege NCLPF Capacity is >0.3g.
- 17. 125V DC Bus Distribution Cabinet Elec S8 +M' X >0.33 conservative Fragility calculation E.as 3 Determined the Anchorage NCLPF Capacity is >0.3g.
- 18. 125V DC Sa Distribution Cabinet Elec $8 +M' h *0.3g Conservative Fragility Calculation Bus 4 Determined the Anchorage y NCLPF Capacity is >0.3g.
C
- 17. Station Battery No. 1 III Elec $8 + 4' X >0.3g Conservative Fregility calculation New Lead Calciua Batteries Determined the Anchorage NCLPF Capacity is >0.3g.
- 20. Station Battery No. 2 (Lead Antimony) Elec S8 + 4' X (2)
- 21. Station sattery No. 3 IlI Elec $8 + 35' X >0.33 C e.servative Fragility calculation New Lead Calcius Batteries Determined the Anchorage NCLPF Capacity is >0.3g.
- 22. Station Battery No. 4 (Lead Antimony) Elec S8 + 35' X (2)
(1) New Lead Catclun Batteries being installed during the next MYC outage (Spring 1987), see Section 4.4. (2) Existing MYC Lead Antimony batteries not schechsted to be changed out during the next outage. Lacking data on the seismic performance of lead antimony batteries preclufes the determination of a HCLPF Capacity, see Section 4.4. MM
Table 7.1-3 Surinary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) Equipment Qualification Screening Building , and NUREG Watkdown Cal utation MCLPF l Equipment Item System Elevation 4334 Review Review Cepecity Comments ELECTRICAL DISTRIBUTION SYSTEMS (Continued)
- 23. Bettery Charger No. 1 Elec $8 +4' X >?.3s Conservative FreglLity calcutetten BC-1 Determined the Anchorage NCLPF Capacity is >0.33
- 24. Bettery Charger No. 2 Elec sg + 4* X >0.3g Careervative Fragility Calculation SC 2 Determined the Anchorage NCLPF Capacity is >0.33
- 25. Bettery Charger No. 3 Elec SS + 46' X >0.?g Conservative Fragility Calculation eC 3 Determined the Anchorage w NCLPF Capacity is >0.3g.
on
- 26. tettery Charger No. 4 Elec SS + 46' X >0.3g conservative Fragility Calcutetten DC-4 Determined the Anchorage NCLPF Capacity is >0.33
- 27. Inverter No.1 Elec SS + 46' X FA 0.82s INVR-1 CDFM 1.13g
- 28. Inverter No.2 Elec $8 + 46' X FA 0.82g INVP. 2 CDFM 1.1?g
- 29. Inverter No.3 Elec $8 + 46' X FA 0.82g INVR-3 CDFM 1.13g
- 30. Inverter No.4 Elec SS + 46' X FA 0.82g INVR-4 CDFM ,1.e3g
Table 7.1 3 Sunnary of Maine Yankee EgJipment Screening and HCLPF Capacities (Continued) Equipment Qualification Screening Building and NUREG Watkdown Calculation NCLPF Equipment Item System Elevation 4334 Reviaw Review Capacity Comunants ELECTRICAL DISTRIOUTION SYSTEMS (Continued)
- 31. Station Service Transformer Elec $5 + 46' X 0.30s KLPF Calculated for Modified X-507 (located adjacent to sus 7 & 8) Archora2e configuration.
- 32. Station Service Transformer Elec Ss + 46' X 0.30g NCLPF Calculated for Modified X 608 (Located adjacent to sus 7 & 8)
Anchorage Configuration. Elec A8 + 22' X >0.3g Panet NCLPF Capacity Judged
- 33. Dieset Generator 1A 480V Distribution
>0.3g Assessed During Walkdown Penet (welL sourited)
Review. Diesel Generator 10 450V Distribution Elec A8 + 225 X >0.3g Panel ELPF Capacity Judged a 34.
>0.3g Assessed During Watkdown Panet (welt mounted) h Review.
- 35. Dieset Generator Control Panet 1A Elec A8 + 22' X- >0.3g Conservative Fragility Catculation Determined the Anchorage NCLPF Capacity is >0.3g.
Elec As + 22' X >0.39 Conservative FragfLity Calculation
- 36. Diesel Generator Control Panet 1B Determined the Anchorage NCLPF Capacity is >0.3g.
Elec >0.3g Cabinet KLPF Capacity Judged l 37. Distribution Cabinet 1 (well mounted) 58'+ 21' X I >0.3g Assessed During Watkdown DC-1 Review. Elec $8 + 21' X >0.3g Cabinet KLPF Capacity Judged
- 38. Distribution cabinet 2 (wati mounted)
>0.3g Assessed During Watkdown DC-2 Review.
l
Table 7.1 3 Sumary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) Equipment Qualification Screening Building and NUREr, Watkdown Calculation HCLPF Equipment Item System Elevation 4334 Review Review Capacity Comments ELECTRICAL DISTRIBUTION SYSTEMS (Continued)
- 39. Distribution Cabinet 3 (wall mounted) Elec S8 + 21' X >0.3g Cabinet HCLPF Capacity Judged DC-3
>0.3g Assessed During Walkdown Review.
- 40. Distribution Cabinet 4 (watt mounted) Elec S8 + 21' X >0.3g Cabinet MCLPF Capacity Judged DC 4
>0.3g Assessed During Walkdown Review.
- 41. Main Control Board Elec S8 + 21' X >0.3g Conservative Fragility Calculation 120V AC Vital Bus 1 4 Determined the Anchorage y MCLPF Capacity is >C.3g.
E$
- 42. Electrical Control Board Elec S8 + 21' X >0.3g Conservative Fragility Calculation DC-1A & 18 Start 1 & 2 Circuits and Control Power Determined the Anchorage HCLPF Capacity is >0.3g.
- 43. Aux 1Llary Logic Cabinets Elec $8 + 21' X >0.3g Conservative FragfLity Calculation Determined the Anchorage HCLPF Capacity is >0.3g.
- 44. ESF Auxiliary Panets A & 8 Elec S8 + 21' X >0.3g Conservative Fragility calculation Determined the Anchorage HCLPF Capacity is >0.3g.
- 45. Air Condition Control Panet ACCP Elec S8 + 21' X >0.3g Conservative Fragility Calculation Determined the Anchorage HCLPF Capacity is >0.3g.
- 46. Safety Parameter Display System Cabinets Elec $8 + 21' X >0.3g Conservative Fragility Calculation Determined the Anchorage HCLPF Capacity is >0.3g.
Tabte 7.1-3 Sumary of Maine Yankee Equipment Screening and MCLPF Capacities (Continued) Equipment Qualification Screening Building and NUREG Walkdown Calculation HCLPF Equipment item System Elevation 4334 Review Review Capacity ccaunents HVAC
- 1. DG-1A Room Exhaust Fan MV A8 + 31' x >0.3g Fan HCLPF Capacity Judged FN 20A >0.3g Assessed During Walkdown Review.
- 2. DG-18 Room Exhaust Fan MV A8 + 31' x >0.39 Fan HCLPF Capacity Judged FN 208 >0.3g Assessed During Walkdown Review.
- 3. Conputer Room Air Conditioner
- SCC $8 + 39' x 0.38g HCLPF Calculated for Modified AC-1A St@ port Configuration.
4
* $8 + 39' x 0.38g HCLPF Calculated for Modified h 4. Conputer Room Air Conditioner PCC St@por'; Cor. figuration.
AC-18 Lab Air Conditioner
- 58 + 39' x 0.38g HCLPF (st'utated fc,.- Modified
- 5. SCC .
AC 2 Support f mfiguration. l
- 6. Diesel Generator Air Intake & Exhaust DG A8 + 31' x >0.39 Danper Hi1PF Capacity Judged Danpers >0.3g A64essed During Walkdown Review.
- 7. Containment Spray Fan SCC CS + 20' x >0.33 Conservaths Fragility Calculation Determined the idiorege FN 44A (Modified Anchorage) HCLP' f*,acity is >0.3g.
- 8. Contairunent Spray Fan SCC CS + 20' x >0.3g conservative Fragility calculation FN-448 Determined the Anchon ge (Modified Anchorage) HCLPF Capecity is >0.3g.
Table 7.1-3 Sumnary of Haine Yankee Equipment Screening and MCLPF Capacities (Continued) Equipment Qualification Screening Building and NUREG Walkdown Calculation MCLPF Equipment Item System Elevation 4334 Review Review Capacity Comments UALVES
- 1. Aux. Feedwater Regulating Valve A0V AFW AF + 23' C >0.39 Valve within bomds AFW A-101 of experience data.
- 2. Aux. Feedwater Regulating Valve AOV AFW AF + 23' C >0.3g Valve within bounds AFW A-201 of experience data.
- 3. Aux. Feedwater Regulating Valve A0V AFW AF + 23' C >0.3g valve within bomds AFW A-301 of experience data.
- 4. Stock Valve for AFW-A-101 A0V AFW AF + 23' C >0.3g Valve within bounds y AFW-A-338 of experience data.
U
- 5. Block Valve for AFW-A-201 A0V AFW AF + 23' C >0.3g Valve within bounds AFW-A-339 of experience data.
- 6. Block Valve for AFW-A-301 A0V AFW AF + 23' C >0.3g Valve within bomds AFW-A 340 of experience data.
- 7. HPSI Pump 8 Discharge to charging header HPSI PT + 13' C >0.3g l
valve within bounds CH-A-32 A0V of experience data.
- 8. HPSI Puup A Discharge to charging header HPSI PT + 13' C >0.3g Valve within tnmds CH-A-33 A0V of experience data.
- 9. Inlet to Charging Header ADV ** HPSI PA8 + 18' C >0.3g valve within bounds CH F 38 of experience data.
- 10. VCT Discharge to HPSI Pumps MOV CM PA8 + 24' C >0.3g valve within bomds CH-M-1 of experience data.
1 Table 7.1 3 Suomry of Maine Yankee Equipacnt Screening and NCLPF Capacities (Continued) Equipment Qualification Screening guitding and NUREG Walkdown Calculation NCLPF Equipment Item System Elevation 4334 Review Review Capacity comments VALVES (Continued) 1). VCT Discharge to NDSI Pumps MOV CM PAS + 24' C >0.3g Valve within bounds CH M 87 of experience data.
- 12. Containment Spray Header Isolation Velve CS CS + 19' C >0.3g Valve within bounds CS-N 1 MOV of experience data.
- 13. Containment Spray Header Isolation Valve CS CS + 19' C >0.3g Valve within bo mds CS N-2 MOV of experience data.
- 14. CS Puup Containment Suction MOV CS CS - 06' C >0.3g Vstve within bounds of experience data.
y CS N 91 U CS - 06' C >0.3g valve within bom ds
- 15. CS Ptsp Containment Suction MOV CS CS N 92 of experience data.
- 16. NPSI Discharge to Loop 1 MOV NPSI PAS + 23' C X >0.3g Valve encaeds experience NSI-M-11 data bounds by 171.
1 NPSI Discharge to Loop 1 MOV NPSI PAS + 23' C X >0.3g Valve exceeds experience i 17. HSI M-12 data bounds by 17X.
- 18. NPSI Discharge to Loop 2 M0i NPSI pas + 23' C X >0.3g Velve exceeds experience NSI-M-21 data bounds by 171.
Table 7.1-3 Summary of Maine Yankee Equipment Ecreening and HCLPF Capacities (Continued) Equipment Qualification Screening Building and NUREG Walkdown Cciculation HCLPF Equipment Item System Elevation 4334 Review Review capacity conenents VALVES (Continued)
- 19. HPSI Discharge to Loop 2 MOV HPSI PA8 + 23' C X >0.3g Valve exceeds experience HSI M 22 data bounds by 17%.
- 20. HPSI Discharge to Loop 3 MOV HPSI PA8 + 23' C X >0.3g valve exceeds experience HSI-M 31 data bomds by 17%.
- 21. HPSI Discharge to Loop 3 MOV HPSI PA8 + 23' C X >0.3g Valve exceeds experience HSI M-32 data bomds by 17%.
- 22. HPSI Disenarge to SI Header MOV HPSI PA8 + 23' C X >0.3g Valve exceeds experience y NSI-M-40 data bounds by 17%.
o a
- 23. HPSI Pusp Discharge MOV HPSI PA8 + 23' C X >0.3g valve exceeds experience HSI M-41 data bomds by 17%.
- 24. HPSI Ptap Discharge MOV HPSI PA8 + 23' C X >0.3g valve exceeds experience HSI-M 42 data bounds by 17%.
- 25. HPSI Discharge to SI Header MOV HPSI PA8 + 23' C X >0.3g Valve exceeds experience HSI-M 43 data bomds by 17%.
- 26. HPSI Suction from RWST MOV HPSI Yd + 21' C
>0.39 Valve within bounds MSI-M-50 of experience data.
27 HPSI Suction from RWST MOV HPSI Yd + 21' C >0.3g valve within bounds HSI M-51 of experience data. l l
Table 7.1 3 Summary of Maine Yankee Equipment Screening and MCLPF Capacities (Continued) Equipment Quotification Screening Building and NUREG Walkdown Calcu'.ation HCLPF Equipment item System Elevation 4334 Review Review Capacity Comments VALVES (Continued)
- 28. CS Discharge to NPSI Ptsp MOV CS CS + 19' C >0.3g Valve within bounds HSI M-54 of experience dets.
- 29. CS Discharge to NPSI Puup MOV CS CS + 19' C >0.3g Valve within bounds MSI-M-55 of experience date.
- 30. RWST Discharge to LPSI MOV CS Yd. + 28' C >0.3g vatve within bouids LSt-M-40 of experience data.
- 31. RWST Discharge to LPSI MOV CS Yd. + 28' C >0.3g Valve within bounds LSI-M 41 of experience data.
y U VA + 438 >0.3g vetve within bouids
- 32. Decay Neat Release Velve A0V ASDA C MS A 162 of experience date.
- 33. AFW Puup B Turbine Throttle Valve AFW VA + 21' C >0.3g Valve within botads MS A-173 of experience deta.
- 34. Decay Heat Release MOV SPC VA + 43' C >0.3g Valve within bouids MS M 161 of experience dets.
- 35. Auxiliary Steam St@ ply Valve MOV MS VA + 43' C >0.3g Valve within boteds MS N-255 of experience data.
n
Table 7.1 3 Sunmary of Maine Yankee Equipment Screening snt W** CWities (Continued) l
-A %e esW. AC Equipment Quell? w ilon Screening 8dilding ~~
and MUREG Walkdown Calculation NCLPF Etpaipment item System Elevation 4334 Review Review Capacity Comuments VALVES (Continued)
- 36. Turbine Steam Surply Pressure Control AFW VA + 21' C >0.3e Valve within bom ds MS P-168 of experience data.
- 37. I Turbine Steam Supply Temperature Control ASDA VA + 418 C >0.3s valve within bounds l I MS-T 163 ADV of experience data.
- 38. PCCW S q ly to P-12A Seal Cooler A0V PCC PA8 + 04' C >0.3g valve within bom ds PCC A-53 of experience data.
- 39. Return from Penetration Coolers A0V PCC PT + 12' C >0.3g va ve within bom ds y PCC-A 216 of experience data.
- 40. Return from Penetration Coolers ADV PCC PT + 12' C >0.3s valve within boe ds PCC A 238 of experience data.
- 41. Return from CEA Air Coolers PCC RC + 018 C >0.3g Valve within boeds PCC-A 270 of experience data.
- 42. DG-1A Cooling Water Outlet A0V PCC A8 + 21' C >0.3s valve within bo eris PCC A 493 of experience data.
- 43. PCCW DJtlet from RNR Neat Exchanger PCC CS + O2' C >0.3s valve within bounds PCC N 43 of experience data.
k Table 7.1-3 Summary of Maine Yankee Equipment Screening and NCLPF Capacities (Continued) l Equipment Quotification Screening l Building and MUREG Watkdown Calculation NCLPF E pipment Itse System Elevation 4334 Review Review Capacity Comments VALVES (Continued) t 44 PCCW ! solation to BR & LW Coolers MOV PCC PAS + 11' C >0.3s vetve within beimds PCC-N 90 of experience data.
- 45. PCCW Isolation to Letdown Neat Exchangers PCC PAS + 21' C >0.3g vetve within beimds PCC-M 150 MOW of experience data.
- 46. PCCW Isolation to Contairment PCC PAS + 11' C >0.3g vetve within tursu6 PCC-N-219 Mov of experience stata.
- 47. Cooler Sigply Temprature Control A0W PCC TB + 37+ C >0.3g vetve within boimds y PCC-T-19 of experience data.
U >0.3g Valve within beimde
- 48. Cooter typess Temperature Control ADV PCC TS + 37' C PCC-T 20 of experience date.
- 49. Pressurizer Safety Velve SRV RC + 65' C >0.3g Verdor data review foamd PR S-11 vetve within boimda of emperience date.
- 50. Pressurizer Safety Valve SRV RC + 658 C >0.3g vendor data review foamd PR-S-12 valve within bounds of experience date.
- 51. Pressurizer safety valve SRV RC + 65' C >0.3g vendor data review foimd PR S-13 velve within bounds of experience data.
.m.. - -
Table 7.1 3 Suunary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) l l Equipment Qualification Screening Building and NUREG Watkdown Calculation HCLPF Equipment item System Elevation 4334 Review Review Capacity Comments VALVES (Continued)
- 52. Power-operated Relief Valve POR4 RC + 66' C >0.3g vendor data review found PR-S-14 valve within bom ds of experience data.
- 53. Power Operated Relief Valve PORV RC + 668 C >0.3g Vendor data review fo md PR S-15 valve within bounds of experience data.
- 54. Power-Operated Block Valve MOV PORY RC + 648 C >0.3g Vendor data review fo md valve
? PR M-16 within bomds of experience data.
N
- 55. Power-operated Block Valve MOV PORY RC + 64' C >0.3g Vendor data review fo o d valve PR M 17 within bo m ds of experience data.
- 56. Non-Seismic Return Headar Stop Valve A0V SCC TB + 21' C >0.3g Valve within bounds SCC-A-460 of e p lence data.
- 57. Non-Seismic Return Needer Stop Valve A0V SCC TB + 43' C >0.39 volvo within bom ds SCC-A-461 of experience data.
- 58. Cooter typass Temperature control A0V SCC TB + 37' ; >0.3g Valve within bounds SCC T-23 of experience data.
- 59. Cooler sypess Temperature control A0V SCC TB + 37' C >0.3s Valve within bounds SCC 7-24 of experience data.
Table 7.1 3 susmery of Maine Yankee Egsipment Screening and NCLPF C pecities (Continued) Equipment Ouellfication Screening guitd' ,J and NUREG Watkdown Calculation NCLPF system Elevation 4334 Revleu Revleu Capacity Ceements Equipment Itas VALVES (Continued) DG 13 Cooler Intet Temperature Control Ag + 23' C >0.3e Velve within bo n ds
- 60. SCC SCC-T-305 Aoy of emperience date.
RCP seet Water Intet Aoy ** PCC sg + 23' C >0.3s Vetve within bounds 61. of emperience date. st.P.3 X >0.3g Piping NCLPF Capacity Judged PIPING
>0.3g a=====M During Watkdom Revleu.
Y
$TUATIONS
*a Racks NCLPF Capacity Judged
- 1. Instrument Rocks various various X
>0.3g Assessed During Walkdoun Revleu.
various various X >0.3s Cable Tray & Conduit NCLPF Judged
- 2. Cable Trays & Conduit
>0.3g Assessed During Welkdown Revleu.
- 3. Ispatse Lines (EI)
- Component h failure may breach critical system pressure boundary.
*
- Valves PCC-N 90 and PCC-N-219 have been determined to have a capacity of 0.3 e's or higher, thus, these components do not reg 4 ire a seismic revleu as the valves will isolate att components downstream from these valves.
table 7.1 3 Sumnery of Maine Yankee Equipment Screening and NCLPF Capacities (Continued)
~
Equipment Quellfication Screening Building and NUREG Watkdown Calculation NCLPF EgJipment Item System Elevation 4334 Review Revleu capacity Comuments Legend:I SYSTEM SYSTEM AFW Auxiliary Feeduster SL Sett Water ASDA Alternate Shutdown Decay Heat Removat SPC Secondary Pressure Control CN Charging SRV Safety Relief Valves CS Containment Spray SW Service Water p OG Diesel Generator Eterting o ~ F0 Fuet Olt NPSI Nigh Pressure Safety Injection NV Aree Meeting and Ventitation i PCC Primary Component Cooling PORV Power Operated Relief Valve SCC Secondary Component Cooling
Table 7.1 3 Summary of Maine Yankee Egalpment screening and NCLPF Capacities (Continued) Equipment euellfication Screening Sullding and NUREG Welkdoun Catculation NCLPF E @lpment Itse System Elevation 4334 Rev'ow Review Capacity Commente Legend: (Continued) Building AS Turbine sullding Auxiliary Bay AF AuglLiery Food Ptap Nouse CS Containment spray Pig House CW Circulation Water Pump Nouse
? PAS Primary Atatillary Building M
PT Pipe itsinet
- PV Purge Air Valve Room RC Reacter Coolant RMC Reactor Motor Control Center Room
$8 Service tullding TB Turbine Building VA Steam and Feed Water Valve Area YD Yard
.. . . ~.
. The screened-in components are the ones that are left in in the final event and -l fault trees using which the accident sequence Boolean expressions are derived. The j seismic fragility parameters and the HCLPF capacities of these ccmponents are : given in Table 5.5-1 in Chapter 5. The nonseismic unavailabilites are listed in l Table 5.5-2 in Chapter 5. ' 7.3 HCLPF Canacity of Plant Two dominant accident sequences that could lead to core damage were studied in this review: small LOCA and transient. The Boolean expressions were derived for thesetsequences consisting of seismic failures and nonscismic unavailabilities. The seismic failures were quantified by the seismic fragilites reported in Table 5.5-1. For the accident sequence of small LOCA, the HCLPF capacity was determined to be 0.21 g pga. This is almost entirely governed by the failure of the refueling water storage tank. For the accident sequence of no LOCA, the HCLPF capacity 2 was determined to be in the range of 0.32 3 to 0.33 3 depending on whether the nonseismic unavailabilities are considered or not. The component failures contributing to this sequence are the transformers, and circulating water pumphouse. 1 These accident sequences could not be combined together to estimate the core damage HCLPF capacity since the split fractions for the two dominant accident sequences are not known. Parametric studies were done wherein the different split fractions were assumed. The resulting HCLPF capacities were seen to vary from , 0.23 g to 0.32 g for split fraction ranging from 50% to 1% for the small LOCA. , 7.4 Identification of Low Canacity Comnonents j J During the course of this review, certain items were identifec tobe potential seismic vulnerabilites at Maine Yankee. Anons these are the aged lead antimony batterics, the internal seismic supports for transformers, vibration-isolation supports for containment spray fans, anchorage of diesel day tank, and block wall near the containment spray fans. The utility has proposed that certain modifications would be made for the components in the next outage in March 1987. The seismic capacities of these components have been estimated in this study in their upgraded condition. 4 7.5 Conclusions This seismic margin review has been performed with the following assumptions and limitations: o The review earthquake level was specified by the NRC as the NUREG/CR-0098 median spectrum. We have interpreted this spectrum to be a 84 percent confidence site specific spectrum for i Maine Yankee. The spectral values are treated as corresponding to the higher ones from the two orthogonal horizontal directions. 7-32
o The structural models and the floor response spectra generated by Maine Yankee have been cursorily reviewed by EQE and judged
=-to be adequate for the purposes of this margin review.
o In general EQE is in agreement wi:h the Panel's recommendations; where the Panel's recommendations were not specific or did not cover a particular item, they have been identified and the. review has been accomplished based on our own . experience -from past . seismic PRAs, earthquake damage ,
. investigations, and judgment.:
o ,[Since the analysis team cou!d not perform'the walkdown inside the containment, the seismic capacity of components inside the
, containment could not be determined. , In some instances, Maine
- Yankee provided. photographs for certain equipment items.inside
> ;the containment. - However, we could not confirm the absence of. potential system interaction effects that may make the impulse
. . lines inside the containment vulnerable in earthquakes and lead
'to a small LOCA.
o Our review and walkdowns. concentrated on components
- supporting Group A system functions; however, a review of non- .
, Group A system components (block walls, for example) led us to believe that there are no anomolies at Maine Yankee which would violate the Expert Panel syttems screening guidelines. o In keeping with the Expert Panel philosophy, the screening of components was performed using conservative procedures. For the screened-in components, the seismic capacities have been j calculated using conservative methods since the scope did not permit detailed response analysis and thorough investigation of
, failure modes and their capacities. In all cases, the factors contributing to the seismic margin 'and their varitabilities are 4
identified and quantified using procedures normally used within the state-of-the-art. o Th: plant level HCLPF capacity was calculated based "on two accident sequences: small LOCA and no LOCA. Since the small LOCA could not be dismissed at this review carthquake level of 0.30s, it was postulated to occur: the contributing component to this sequence HCLPF- capacity is RWST. Sensitivity ~ studies indicate that the plant level HCLPF capacity is higher than the HCLPF capacity of the RWST. o The HCLPF capacity of the plant has been determined based on the seismic capacities of components in ,th d r existing or proposed modified conditions. Maine Yankee has proposed that certain modifications or' replacements would be made for station batteries, transformer internal core / coil assembly anchorage, vibration-isolation supports for containment spray fans and air conditioners, anchorage of diesel day tank, and block wall near 7-33 -
m i I the containment spray fans. We recommend that a third walkdown of the plant be ' performed to assure that these i modifications have been carried out and that the HCLPF l capacities of these components calculated in the present study are l still applicable. i 1 Based on this margin review and subject to the above restrictions, EQE confirms that the high confidence low probability of failure capacity of Maine Yankee plant is at least 0.213 refered to the ground response spectrum specified above;it is also our judgment that the actual capacity of the plant is much higher since the assumption of a small LOCA occuring is conservative, s J ] i 7-34 t
w . 4:
" REFERENCES l
EAmerican , Concrete Institute 3(1981), Buildinn Code ' Reauirements for- < Concrete Masonry' Structures (ACI 531-79) (Revised 1981) and Commentarv -
~ ACI 531R-79. -
i ANCO, Engineers,'Inc., Generic Seismic Oualification Usinn- Existinn~ Test DaliL, 4 Bernreuter, D.L., -J.B. L Savy, R.W. Mensing, J.C. Chen, and" B.C. Davis (April 1985), Seismic Hazard Characterization of the Eastern United States i Volume 1- Methodolony and Results for Ten Sites. - UCID-20421 Vol.1.
' Lawrence Livermore National Laboratory, Livermore, California.
Budnitz, R.J., et - al., (July 1985), An Anoroach to the Ouantification of 4 Seismic Marnins in Nuclear Power Plants. NUREG/CR-4334. Campbell. R.D.,' R.P. Kennedy,' and R.P. Kassawara May 1986, Seismic Marnin Review Methodolony. Draft EPRI Report. I Coats, D.W., May 1980, Recommended Revisions to Nuclear Renulatory l' Commission Seismic Desian Criteria. NUREG/CR-Il61. CYGNA Energy Services -(April 1982), Dynamic Modeline of Class I
- Structures. Maine Yankee Atomic Power Station. Report BM-Y-MY-80006-5, i Rev. O.
I EQE Incorporated (October 1986), The Performance of Cable Trav ' and Conduit Systems In Actual and Simulated ~ - Seir.mic Motion, Report i 82001.05.001 Draft B, prepared for the Seismic Qualification Utility Group. 1 Gerard, G. and H. Becker (1957), Handbook of Structural Stability. NACA 1 TN 3783. j Hall, D.H. (April 1981), Pronosed Steel Column Strenath Criteria. Journal of The Structural Division, ASCE. l
- Haroun, M.A. and G.W. Housner (April 1981), Seismic - Desian of Liould i Storane Tcnks. Journal of The Structural Division, ASCE.
Hashimoto, P.S., R.P. Kennedy, R.B. Narver, and D.A. Wesley (December 1984), Conservative Seismic Canacities of the Maine Yankee Reactor Containment Includine and Excludinn Desian Incident Pressure. NTS/SMA 36001.01, prepared for-Maine Yankee Atomic Power Company by Structural Mechanics Associates. Henries, W.E., and S.R. Boeschen (March 1986), Evaluation of AOVs and MOVs to a Greater Than Desian Base 'Earthouake in the Maine vankee' ' Nuclear Power Plant, Prepared for Maine Yankee Atomic Power Company.. Kaplan, S. (1981), On the Method of Discrete Probability Distribution in Risk and Reliability Calculations. Risk Analysis, Volume 1. R-1
=,
7 -
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]
;y
~ _
6 m Kennedy, R.P.' and M.K. Ravindra (May 1984); Seimmic Franil'ities for NuclearI
- Power Plant Risk Studies. Nuclear Engineering and Design,- Vol.- 79,.No.1, pp.
- 47-68. _
Manos, G.C.E (Augusti 1986),' Earthauake Tank-Wall ~ Stability of Unanchored ~ Tanks. Journal of the Structural Engineering, ASCE. National Aeronautics and Space- Administration (June 1966), Shell Analvsis : Manual.
. Newmark, ;N.M. and W.J.? Hall (May- 1978),2 Development' of Criteria For-Seismic Review of Selected Nuclear Power Plants. NUREG/CR-0098.
~
Prassinos, P.G., M.K. Ravindra,- and J.B. Savy (October 1985), Trial Guidelines for Seimmic Marnin -Review's ' of Nuclear' Power Plants. -NUREG/CR-4482, Lawrence Livermore National Laboratory, Livermore, California. MA Procedures Guide. NUREG/CR-2300,'2 volumes, January -1983. Ravindra, M.K.,' and R.P. Kennedy (August .1983), " Lessons Learned from .
. Scismic PRA Studes," Paner M6/4, Proceedings of the-Seventh Conference on c .
Structural Mechanics in Reactor Technology Chicago, Illinois. Reed,- J.W., R.P. Kennedy, and M.K.- Ravindra (December 5,1986),' Memo to
~
Fragility - Hazard Interaction Subgroup Joint . Working Group on Seismic PRA, ASCE Dynamic Analysis Committee. Smith, - G.V., An Evaluation of the Yield. Tensile. Creen. and ' Tructure : Strenaths of Wrought 304. 316.~ 321. and 347 Stainless Steels At Elevaterl '
.Temocratures. ASTM Data Series DS.5S2.
~
Stevenson, J.D., and W.E. Henries (January 1983), Seismic - Review of the Maine Yankee Nuclear Power Plant.~ Volume I, prepared ' for the Maine" Yankee Atomic Power Company. Structural Mechanics Associates (February 1983), Seismic Marnin -Review Midland Enerny Center Pro ject. Volume I. Methodolony and - Criteria, ' prepared for Consumers Power Company. ' URS Corporation / John A. Blume & Associates, Engineers, Seismic Anchorane Guidelines for Nuclear Power Plant Eauioment. Volume I: Develooment and Use of Guidelines. prepared t'or Electric Power Research Institute, Palo Alto, California. , + j- . Weingarten, V.I., E.J. Morgan, an'd P. Sc.ide (March 1965), Elastic Stability of j Thin Walled Cylindrical and Conical Shells Under Axial Comoression. AIAA ! Journal. Wozniak, R.S. and W.W...Mitchell (May 9,1978), Basis of Seismic Desian Provisions For Welded Steel Oil Storane TanksfAmerican Petroleum Institute, ! 43rd Midyear Meeting, Toronto, Ontario, Canada. i i-j- f-R-2 4 4 '~
APPENDIX A MA NE YANKEE ATOMIC POWER STATION ARRANGEMENT DRAWINGS CONTENTS STRUCTURE Site Plan -- A-2 Plot Plan .- A-3 Reactor Containment Plan, El. 46'-0" . A-4 Containment Plan, El. 20'-0" . A-5 Primary Auxiliary Building Arrangement -.- . A-6 Primary Auxiliary Building Arrangement A-7 Turbine Building El. 21'-0" - .- A-8 Turbine Building El. 39'-0" . . A-9 Turbine Buitding P1an . A-10 l Service Buitding Addition A-I I CircuIating Water Pump House . A-12 Turbine Building El. 21'-0" - --. . A-13 i I A-1
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' '- NU G -4826 2#2 BIBLIOGHAPHIC DATA SHEET ut iairavem O., t ..' Vol. 3 2 TITLE ANO Suti,TLt 3 LEAvegLahn Seismic Margin R 'ew of the Maine Yankee Atomic ou ragilit Analysis (,'""""""',,,,
Aur O a. Fe ry 1987 M. K. Ravindra, G. S. rdy, P. S. Hashimoto, and ,,,,,,,,"""""",,,,, J. J. Griffin 4 [ arch 1987
- 7. Pt FORMamQ ORGACKAl'ON MAME AND MAsLi GOmaSSfiacwe leC m/ 8 PROJACTsT ASE/ WORE UNil hum 88R EQE, Inc.
Newport Beach, CA 92660 4 . >>~ e ca A~ r ava.a i Under Contract to: Lawrenc Livermore National Labo tory Livermor , CA 94550 / A0461 i 10. SPON.OmeNG ORGANe2 AfiON NAwa Amo uAsLt4G ADCRt rarws. te Cases _ tia TVPE 08 REPORT Division of Engineering Safety \ Office of Nuclear Regulatory Re arch Technical
' " " ' ' " " ' " ~ ' ' '
- U.S. Nuclear Regulatory Commissi ,
Washington, DC 20555 j L June 1986-March 1987 12 SUPPLtutNT ARY NOf tS A This Fragility Analysis is the thir of ae volumes for the Seismic Margin Review of the Maine Yankee Atomic Power Stat- n. me 1 is the Summary Report of the first trial seismic margin review. Vol le 2, S $ ems Analysis, document the results of the systems screening for the review The three, volumes demonstrate how the seismic nargins review guidance (NUREG/ -4482) of t' NRC Seismic Design Margins Program l can be applied. The overall objectives of t trial review are assess the seismic margins of a particular pressurized wat reactor, and to tes he adequacy of this review approach, quantification chniques, and guidelin' for performing the review. Results from the trial review wi be used to revise the sh smic margin methodology and guidelines so that the Candindustrycanreadilytapplythemtoassesstheinherent quantitative seismic pacity of nuclear power plant
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