ML20113C913
| ML20113C913 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 06/28/1996 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20113C911 | List: |
| References | |
| REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR NUDOCS 9607020187 | |
| Download: ML20113C913 (136) | |
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ATTACHEMENT 1 FOR ESK 96-128 A-46 SEISMIC EVALAUTION REPORT FOR QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 9607020187 960628 PDR ADOCK 05000254 P
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Final Report 93C2806-03.A46 USNRC USl A-46 Resolution Seismic Evaluation Report Quad Cities Nuclear Station Units 1 and 2 Prepared for Commonwealth Edison Company Quad Cities Nuclear Power Station 22710 206th Avenue, North Cordova, Illinois 61242-9740 prepared by Stevenson & Associates 10 State Street Woburn, Massachusetts 01801 June 19,1996 I
m Quad Cities A-46 Final Report June 19,1996 List of Acronyms CEA Concrete Expansion Anchor Comed Commonwealth Edison Company EPRI Electric Power Research Institute GERS Generic Equipment Ruggedness Spectra GIP Generic implementation Procedure for the Seismic Verification of Nuclear Station Equipment GL Generic Letter GRS Ground Response Spectrum lAEA international Atomic Energy Agency IPEEE Individual Plant Examination for Extemal Events ISRS In-structure Response Spectra QCNS Quad Cities Nuclear Station l
LAR Limited Analytical Review MCC Motor Control Center OSVS Outlier Seismic Verification Sheet i
PASS Plant Area Summary Sheet PSD Power Spectral Density QCNS Quad Cities Nuclear Station, Units 1 & 2
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S&A Stevenson & Associates 1
SCE Seismic Capability Engineer SEWS Screening Evaluation Work Sheet l
SQUG Seismic Qualification Utility Group l
SRT Seismic Review Team l
SSE Safe Shutdown Earthquake SSEL Safe Shutdown Equipment List SSER Supplemental Safety Evaluation Report l
SVDS Screening Verification Data Sheet i
USl Unresolved Safety issue i
NRC Nuclear Regulatory Commission l
ZPA Zero Period Acceleration i
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Quad Cities A-46 Final Report June 19,1996
- 1. INTRODUCTION AND SEISMIC VERIFICATION METHODOLOGY 1-1 1.1 Introduction 1-1 1.2 Seismic Verification Methodology 1-1 1.3 Report Organization 1-2
- 2. QUAD CITIES NUCLEAR STATION SAFE SHUTDOWN PATH 2-1
- 3. QUAD CITIES NUCLEAR STATION SEISMIC DESIGN BASIS 3-1
3.1 DESCRIPTION
OF INPUT MOTIONS 3-1
3.2 DESCRIPTION
OF DYNAMIC MODELING AND BASES FOR THE SELECTION OF KEY MODELING PARAMETERS 3-1 3.3 IN-STRUCTURE r.ESPONSE SPECTRA 3-2
- 4. RESULTS OF SCREENING VERIFICATION AND WALKDOWN -
EQUIPMENT CLASSES 0 THROUGH 20 4-1 4.1 Seismic Evaluation Guidelines 4-1 4.1.1 Seismic Capacity Vs. Demand 4-2 4.1.2 Caveat Compliance 4-2 4.1.3 Anchorage Adequacy 4-3 4.1.4 SeismicInteraction Checks 4-6 4.2 Outlier Resolution 4-7 4.3 Seismic Capability Engineers and Peer Reviewer 4-7 4.4 Other Types of Seismic Evaluations and Interfaces 4-8 4.5 Documentation 4-8 4.6 Evaluation Results - Equipment Classes 0 Through 20 4-10
- 5. GIP DEVIATIONS AND COMMENTARY ON MEETING THE INTENT OF CAVEATS 5-1
- 6. RESULTS OF THE TANKS AND HEAT EXCHANGER REVIEW 6-1 6.1 EVALUATION METIlODOLOGY 6-1 6.2 Summary of Evaluation Results 6-3 i
m Quad Cries A-46 Final Report June 19,1996
- 7. RESULTS OF THE CABLE TRAY AND CONDUlT RACEWAY REVIEW 7-1 7.1 Summary of the Review 7-1 7.2 Scope of the Review 7-1 7.3 Limited Analytical Reviews (LARs) 7-2 7.3.1 General Description ofQuad Cities Raceways 7-2
- 7. 3. 2 Limited A nalytical Review Results 1-3 l
7.4 Summary of Outliers 7-5
- 8. RESOLUTION OF OUTLIERS 8-1
- 9. REFERENCES 9-1
- 10. APPENDIX A: SAFE SHUTDOWN EQUIPMENT LIST (SSEL) i
- 11. APPENDIX B: SEISMIC DESIGN BASIS SPECTRA
- 12. APPENDIX C: WALKDOWN PERSONNEL RESUMES
- 13. APPENDIX D: SCREENING VERIFICATION DATA SHEETS (SVDS)
- 14. APPENDIX E: PEER REVIEW ASSESSMENT 4
- 15. APPENDIX F: ANCHOR BOLT TIGHTNESS AND EMBEDMENT CHECKS ii
Quad Cities A-46 Final Report June 99,1996
- 1. Introduction and Seismic Verification Methodology 1.1 Introduction This report provides the final documentation of the seismic adequacy evaluations performed at Commonweal:S Edison Company's (Comed's) Quad Cities Nuclear Station (QCNS), Units 1 and 2, for the resolution of Unresolved Safety issue (USI) A-46, " Seismic Qualification of Equipment in Operating Plants". USl A-46 was issued by the United States Nuclear Regu! story Commission (NRC)in December,1980 to address the concern with the seismic adequacy of mechanical and electrical equipment in older nuclear power plants. This report describes the methodology used for and the results of the seismic reviews of active mechanical and electrical equipment, selected tanks and heat exchangers, and cable and conduit raceways.
1.2 Seistnic Verification Methodology Utilities affected by USI A-46 formed the Seismic Qualification Utility Group (SQUG)in 1982 to develop a consistent industry approach for resolving USI A-46. SQUG utilities, including Comed, and along with the technical and financial assistance of the Electric Power Research Institute (EPRI) conducted research and studies regarding this issue in order to formulate a thorough and reasoned program to resolve the identified concem. In February,1987, the NRC issued Generic Letter 87-02, " Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety issue (USI) A-46," requesting USI A-46 licensees to commit to a detailed approach for resolving USl A-46 [1].
Subsequently, further research conducted by SQUG (and its contractors) and reviewed by the NRC staff resulted in a detailed procedure developed by SQUG called the
" Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Station Equipment"(2]. Specifically, the NRC staff reviewed Revision 2 of the GIP and accepted (with provisos) the approach in Supplement No.1 to Generic Letter (GL) 87-02 that Transmits Supplemental Evaluation Report No. 2 (SSER #2) on SQUG Generic Implementation Procedure, Revision 2 as Corrected on February 14,1992 (GIP-2) (3}.
This GIP version and the clarifications, guidance and additional requirements provided by the NRC in SSER #2 are the basis for the seismic evaluation of mechanical and electrical equipment at Quad Cities for resolution of USl A-46. The GIP Revision 2 referred to as GIP-2 by the NRC is referred to as the GIP in this report.
Separate, but related issues pertaining to methods of analysis for above-ground flexible tanks identified in USl A-40, " Seismic Design Criteria" (4), and seismic adequacy of proximity items above and around important-to-safety equipment identified in USI A-17
[5] are explicitly addressed and resolved by implementation of the GlP.
The GlP approach relies on developing a safe shutdown equipment list (SSEL) which identifies equipment needed to achieve and maintain safe hot shutdown conditions as defined by a nuclear power plant's Technical Specifications. This equipment is then seismically reviewed in accordance with the GlP methodology. By means of plant walkdowns to specifically observe and evaluate each equipment item on the SSEL, an 1-1
Quod Cities A-46 Final Report June 19,1996 assessment can be made conceming its seismic adequacy. By evaluating seismic demand criteria, selected caveats to ensure similarity to the GIP equipment classes, an anchorage evaluation, and a seismic interaction proximity assessment, the trained walkdown engineer can be satisfied that the equipment will survive the plant's design basis seismic event. The basis for this approach is rooted in detailed observations of representative, if not identical, equipment in industrial facilities that have survived earthquakes of similar or greater magnitude in Califomia and throughout seismically active regions around the world. Each equipment assessment is documented on a Screening Evaluation Work Sheet (SEWS). Any deficiencies are documented on an Outlier Seismic Verification Sheet (s) (OSVS).
1.3 Report Organization The following section of this report discusses the development of the safe shutdown path and the resulting Safe Shutdown Equipment List (SSEL) for Quad Cities. The SSEL is provided in Appendix A. The seismic design basis of Quad Cities and the assessment of it by the NRC are discussed in Section 3. The design basis spectra are contained in Appendix B. The Quad Cities equipment walkdown and results are provided in Section 4. These assessments resulted in summary level screening verification data sheets, SVDS, (Appendix D)
SSER #2 requires explicit documentation of any deviations from the caveats or their intent in the GIP. Section 5 provides a detailed listing of exceptions to the rules taken for any equipment item assessment. Section 6 discusses the results of the Tanks &
Heat Exchangers assessment. They are also summarized under Class 21 - which is the Tanks & Heat Exchangers class -in the SVDS given in Appendix D. Cable Tray &
Conduit Raceway assessments are provided in Section 7. Section 8 provides a listing of identified outliers, the reasons for which they are outliers, and their proposed resolution.
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i Quad Cities A-46 Fin:1 Rtport June 19,1996
- 2. Quad Cities Safe Shutdown Path The Seismic Qualification Utility Group (SQUG) Safe Shutdown Equipment List (SSEL) contains all mechanical and electrical equipment (excluding relays) needed to achieve and maintain safe shutdown conditions at Quad Cities station. The equipment listed contains those components that are needed to support the four SQUG and Safe Shutdown Functions.
The equipment was chosen based on the requirements defined in Section 3 of the SQUG GIP (Generic Implementation Procedure), Rev. 02-corrected 2/14/92 and the NRC's SSER No. 2.
Components have been identified that are in the primary and backup shutdown paths and have been grouped together by both function and system. Components have also been listed in the order of the flowpath.
2.1 DESCRIPTION
OF SSEL BASIS The purpose of the Seismic Qualification Utility Group (SQUG) Safe Shutdown Equipment List (SSEL) is to identify all mechanical and electrical equipment (excluding relays) that is needed to achieve and maintain safe shutdown conditions at Quad Cities station. Based upon
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a review of the Quad Cities station operating procedun:s, systems were chosen to be utilized for the primary and backup safe shutdown paths to accomplish the four safe shutdown functions identified in the SQUG. This list was generated to identify those components that are part of these systems as well as those components that are needed for support. This list contains essential information related to the safe shutdown function of these components.
The following systems have been chosen to satisfy the four safe shutdown functions for s
primary and backup shutdown paths.
Safe Shutdown Function Primary Shutdown Path Backup Shutdown Path Reactivity Control Reactor Protection System; N. A. - Note 1 Control Rod Drive i
i Pressure Control ADS Valves (B, C)
ADS Valves (E, D) 1 Decay Heat Removal Residual Heat Removal RHR Loop B; (RHR) Loop A; RHRSW IAop B RHR Service Water (RHRSW)IAop A Reactor Water Makeup RHR IAop A RHR IAop B Note 1 -
No single failure identified that would prevent bringing the reactor suberitical (i.e., a failure may keep a single rod from inserting but not multiple rods).
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Quad Cities A-46 Final Report l
June 19,1996
2.2 DESCRIPTION
OF SSEL For each piece of equipment / component identified in the list, all of the appropriate information fields were completed. The following is a description of the information that is included in the SQUG SSEL.
LINE_NO A unique number used to identify the order in which the equipment was identified.
SYSTEM Identifies the system the equipment is a part of.
EQUIP _ID An identifying number for the equipment; usually the equipment tag number.
EQUIP _DESC Provides a description of the function of the equipment.
LOC _ CODE Provides the column-row of the equipment in the plant or the room it is in.
BUILDING Identifies the building the equipment is in.
ROOM _ELEV Identifies the plant elevation of the room the equipment is found in.
EVAL _REQD Identifies whether the equipment is:
S=
Seismic R=
Relay SR = Seismic and Relay B=
A " Rule of the Box" item BR = A " Rule of the Box" item requiring a relay review.
Appendix A provides the Seismic Qualification Utility Group (SQUG) Safe Shutdown Equipment List (SSEL) for equipment required to be walked down (EVAL _REQD = S or SR).
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_.__v Quad Citirs A-46 Fin:1 Report June 19,1996 2.3 BASIS FOR SELECTION Quad Cities Station has elected to use the Seismic Qualification Utility Group (SQUG)
Generic Implementation Procedure (GIP) to select and evaluate the seismic adequacy of the mechanical and electrical equipment needed to bring the station to a safe shutdown condition following a safe shutdown eanhquake. Per section 3.3.3 of the GIP, " safe shutdown is defined as bringing the plant to a hot shutdown condition and maintaining it there for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following an eanhquake. Some plants may not have sufficient water inventory to stay in hot shutdown for three days, wnile other plants may prefer to be brought to a cold shutdown condition during this period of time instead of staying in the hot shutdown condition."
The four safe shutdown functions that must be accomplished to achieve safe shutdown are as follows.
1.
Reactor Reactivity Control 2.
Reactor Coolant Pressure Control 3.
Reactor Coolant Inventory Control 4.
Decay Heat Removal The systems / functional paths chosen are to meet the following criteria.
1.
Achieve and maintain the plant in a hot shutdown condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following the SSE.
2.
A LOOP is considered to have occurred coincident with the SSE and lasts for the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
i 3.
No other design basis event is considered to occur other than the SSE.
REACTOR REACTIVITY CONTROL To achieve reactor mactivity control under this safe shutdown path, the Control Rod Drive (CRD) System was chosen. This system was chosen for the following masons:
o The CRD System is a safety-related system.
O The control rods are designed to be driven in automatically upon a loss of power.
O The CRD System will actuate automatically.
O The CRD System is designed to fail in the safe position, that is, with the control rods in the fully inserted position.
o The CRD System is a Class I system.
O The CRD System is located in the Reactor Building on elevation 595'.
o No major support systems are required for CRD system operation.
2-3
Quad Cities A-46 Fin.1 Report June 19,1996 The CRD System is used as both the primary and backup system for the Reactor Reactivity Control function. This is because each of the 177 hydraulic control units (HCUs) per unit are completely separate and independent. Therefore, the system is designed to accommodate a single failure, and still accomplish safe shutdown.
REACTOR COOLANT PRESSURE CONTROL To achieve reactor coolant pressure control under this safe shutdown path the Automatic Depressurization System was chosen, using Electromatic Relief Valves (ERVs) for Unit 1 and Target Rock Power Operated Relief Valves (PORVs) for Unit 2. The B and C valves are used for primary shutdown paths and the D and E valves for backup shutdown paths.
These valves are sized to rapidly remove steam generated from the reactor upon closure of the turbine stop valves and coincident with failure of the turbine bypass system. The blowdown from each miief valves is routed through a separate line below the torus water line.
The valves require control power from the Direct Curmnt system.
The advantages of all of these valves are as follows:
o These valves are pan of a safety-related system.
O These valves will actuate automatically.
O These valves are pan of a Class I system.
O These valves are located on the second floor of the drywell.
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v REACTOR COOLANT INVENTORY CONTROL The system used for this function is the Iow Pressum Coolant Injection i
(LPCI) mode of the Residual Heat Removal (RHR) system in conjunction with ADS (described above). ADS will depressurize the reactor vessel so that the i
LPCI mode of RHR could be initiated for reactor coolant inventory control.
I The advantages of the LPCI and ADS systems are as follows:
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The LPCI and ADS systems are safety-related systems.
o The LPCI and ADS systems will actuate automatically..
j The LPCI and ADS systems are a Class I system.
o q
The ADS system is located on the second floor of the drywell.
o o
The LPCI system is located in the corner rooms which are in the 1
Reactor Building at elevation 554'.
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Quad Citi:s A-46 Fin *.! Report June 19,1996 DECAY HEAT REMOVAL To achieve decay heat removal under this safe shutdown path the Low Pmssure Coolant Injection (LPCI) mode of the RHR system and the RHR Service Water (RHRSW) system were chosen. Loop A was selected as the primary path and Loop B as the backup path.
l In this alignment the RHR pumps take suction from the suppression pool, routing the water through the RHR heat exchanger, and injecting the water back into reactor vessel. RHRSW provides the cooling water to the RHR heat exchanger and takes suction from the cribhouse bay.
The advantages of the RHR and RHRSW systems are as follows:
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The RHR and RHRSW systems are safety-related systems.
o The RHR and RHRSW systems will actuate automatically.
The RHR and RHRSW systems are Class I systems.
o o
The RHRSW system is located in the Turbine Building at elevation 554'.
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The RHR system is located in the corner rooms which are in the Reactor Building at elevation 554'.
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Quad Cities A-46 Fird Report June 19,1996
- 3. Quad Cities Nuclear Station Seismic Design Basis This section describes the seismic motion used for the Quad Cities USI A-46 e +
For purposes of the A-46 study, Quad Cities is using the Dresden plant ISRS fu m 4.? to achieve for the Quad Cities safe shutdown earthquake (SSE) peak grc d ameteration (PGA) of 0.24g. The ISRS and the development of the spectra were pesented to the NRC by Comed in response to Generic Letter 87-02 as " conservative dewgn" spectra (6). The NRC reviewed the design basis ISRS and declared that they could be utilized as " conservative design"ISRS [7]. The following sections describe the basis and development of the design basis spectra.
3.1 DESCRIPTION
OFINPUTMOTIONS The Quad Cities power block structures are virtually identical to those at the Dresden site. The Quad Cities SSE-PGA is 0.24g as opposed to 0.20g determined for the Dresden site. For the A-46 study, Quad Cities is using the Dresden ISRS multiplied by 1.2 to adjust the SSE-PGA to 0.24g. The following describes the development of the Dresden ISRS since they are the basis for the ISRS that are being used for Quad Cities for the A-46 evaluation study.
The input motions used to create the seismic design of DNS [6] are based on the Housner-like Ground Response Spectrum (GRS) and the north-south component earthquake record of El Centro of May 18,1940. For the seismic analysis of the Seismic Class I structures and development of ISRS, the Housner-like GRS (referred to as " smooth response spectrum") was used [8, (Figure 3.7-2)]. The El Centro earthquake time history was used to verify that when using the time-history method the maximum Operating Basis Earthquake (OBE) loadings did not occur in the valleys of the unsmoothened (El Centro) spectrum. This DNS Operating Basis Earthquake (OBE) is defined in the horizontal direction by the Housner GRS scaled to 0.10g peak ground acceleration (PGA) and the ISRS developed from the El Centro earthquake time history scaled to 0.10g. The OBE in the vertical direction is defined by 2/3 of the Housner GRS with a resulting PGA value of 0.0679. The DNS Safe Shutdown Earthquake (SSE)is defined by multiplying the OBE acceleration by a factor of 2 resulting in a horizontal direction GRS PGA value of 0.20g.
3.2 DESCRIPTION
OF DYNAMIC MODELING AND BASES FOR THE SELECTION OF KEY MODELINC, PARAMETERS The Quad Cities Nuclear Station is made up of the following Seismic Class 1 structures:
- 1. Drywell Containment Structure & Intemals including Reactor Pressure Vessel
- 2. Reactor-Turbine building
- 3. Suppression Chamber
- 4. Control Room
- 5. Chimney Safe shutdown equipment is located in structures 1 through 4 listed above. The seismic structural response was determined for structures above. The buildings are founded on bedrock so any soil-structure interaction effects were considered negligible, 3-1
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Quad Cities A-46 Final Report Juro 19,1996 and the Housner response spectra described above were used as the input at the base of the structures.
The mathematical model of the above Seismic Class 1 structures was constructed in terms of lumped masses and stiffness coefficients on a fixed base. Damping values used were based upon the evaluation of the materials and mode shapes. The damping used is as shown in Table 3-1 shown below:
Table 3-1 Quad Cities Design Basis Damping Values Structural / Component Type Damping Welded Assemblies 1%
Steel Frame Structures 2%
Bolted & Riveted Assemblies 2%
Reinforced Concrete Structures 5%
Vital Piping Systems 0.5%
3.3 IN-STRUCTURE RESPONSE SPECTRA The horizontal response spectra curves are based on a lumped mass, fixed base model for the Reactor-Turbine building. Appendix B shows the north-south and east-west floor spectra used for the A-46 evaluation. The vertical in-structure response const nt acceleration or seismic coefficient is defined by 2/3 of the Housner-like GRS PGA (t SE - 0.08g, SSE - 0.16g).
Additionalin-structure response spectra at additional damping values have been developed for other programs such as the review of masonry block walls (IE Bulletin 80-11). As described in the SQUG GIP, Revision 2, the use of 5 percent damped in-structure response acceleration curves are allowed for characterizing seismic demand for equipment.
The NRC staff reviewed the original and subsequent modeling performed by Comed and its contractors and determined that the building modeling was adequate. The NRC staff concluded that the resulting in-structure response spectra could be utilized as conservative desian ISRS spectra as defined in the GIP [2] as opposed to realistic, median centered ISRS [7].
The SSE site ground response spectrum and the generated ISRS are provided in Appendix B.-
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Quad Cities A-46 Final Report June 10,1996
- 4. Results of Screening Verification and Walkdown - Equipment Classes 0 Through 20 The purpose of this section is to describe the Screening Verification and Walkdown performed to verify the seismic adequacy of active mechanical and electrical equipment identified in the Quad Cities Safe Shutdown Equipment List (SSEL) report [24]. The guidelines contained in this section were used to screen the equipment for seismic adequacy. If the equipment did not pass this screen, it was declared an outlier and discussed in Section 8. Outlier Resolution, also described in Section 8, is accomplished by
- 1) more refined or sophisticated methods for verifying seismic adequacy, or
- 2) equipment / anchorage modification.
4.1 Seismic Evaluation Guidelines The procedure for performing the Screening Verification and Walkdown is based on the following four seismic screening guidelines, as defined in the GIP [2):
1.
Seismic Capacity Compared to Seismic Demand - The seismic capacity of the equipment, based on earthquake experience data, generic seismic testing data, or equipment-specific seismic qualification data, should be greater than the seismic demand imposed on the equipment by the safe shutdown earthquake (SSE).
2.
Caveats - In order to use the seismic capacity defined by the earthquake experience Bounding Spectrum or the generic seismic testing GERS, the equipment should be similar to the equipment in the earthquake experience equipment class or the generic seismic testing equipment class and also meet the intent of the specific caveats for that class of equipment. If equipment-specific seismic qualification data is used, then any specific restrictions or caveats for that qualification data apply instead.
3.
Anchoraae - The equipment anchorage capacity, installation, and stiffness should be adequate to withstand the seismic demand from the SSE at the equipment location.
4.
Seismic Interaction - The effect of possible seismic spatialinteractions with nearby equipment, systems, and structures should not cause the equipment to fail to perform its intended safe shutdown function.
The evaluation of equipment against each of these four screening guidelines at Quad Cities is based upon walkdown evaluations, calculations, and other supporting data.
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ouad Cities A-46 Final Report i
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4.1.1 Seismic Capacity Vs. Demand Quad Cities determined the seismic capacity of safe shutdown equipment using:
Earthquake experience data with capacity defined by the Bounding Spectrum, or Reference Spectrum depending of the demand spectrum used, Generic seismic test data which have been compiled into Generic Equipment Ruggedness Spectra (GERS), or Equipment-specific seismic qualification data.
l The seismic demand imposed on an item of equipment depends on whether or not the Housner ground response spectrum or amplified in-structure response spectra was used, and how it is compared to the capacity data.
Conservative ISRS were compared to 1.5 times the bounding spectrum (i.e., reference spectrum) or the ground spectrum was compared to the bounding spectrum. To a lesser extent, the ground spectrum was compared to the bounding spectrum for equipment within 40' of grade with an estimated fundamental frequency greater than 8 l
Hz. The GERS were used in only a few cases for the capacity vs. demand comparisons. Finally, newer, upgraded equipment that had been seismically qualified in accordance with the IEEE 344 Standard,1975 Edition or later, was accepted based on this testing documentation and anchorage design calculations, and was supplemented only by a seismic interaction review by the SRT.
For purposes of determining the 40' Above Grade elevation, effective grade for the site and/or each building must be determined. " Effective grade" at a Nuclear Station is defined as the average elevation of the ground surrounding the building along its perimeter. As Quad Cities effective grade was established at 595'.
4.1.2 Caveat Compilance The second screening guideline which must be satisfied to verify the seismic adequacy of an item of mechanical or electrical equipment is to confirm that (1) the equipment characteristics are generally similar to the earthquake experience equipment class or the generic seismic testing equipment class and (2) the equipment meets the intent of the specific caveats for the equipment class. This review is only necessary when the Bounding Spectrum or the GERS is used to represent the seismic capacity of an item of equipment. If equipment-specific seismic qualification data is used instead, then only the specific restrictions applicable to that equipment-specific qualification data need be l
applied.
Another aspect of verifying the seismic adequacy of equipment included within the scope of this procedure is explained by the " rule of the box." For the equipment 4-2
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Quod Cities A-46 Final Report 4
June 19,1996 f
included in either the earthquake or testing equipment class, all of the components mounted on or in this equipment are considered to be part of that equipment and do not have to be evaluated separately. However, the walkdown engineers did look for 4
j suspicious details or uncommon situations which could make the equipment item vulnerable, i
j An item of equipment should have the same general characteristics as the equipment in the earthquake experience equipment class or the generic seismic testing equipment class. The intent of this rule is to preclude items of equipment with unusual designs l
and characteristics which have not demonstrated seismic adequacy in earthquakes or s
tests.
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" Caveats" are defined as the set of inclusion and exclusion rules which represent specific characteristics and features particularly important for seismic adequacy of a particular class of equipment. Appendix B of the GIP contains a summary of the caveats for the earthquake experience equipment class and for the generic seismic testing equipment class.
i The " intent" of the caveats should be met when evaluating an item of equipment as they are not fixed, inflexible rules. Engineering judgment is used to determine whether the specific seismic concem addressed by the caveat is met. Each item of equipment should be evaluated to determine whether it meets the specific wording of the applicable caveats and their intent. However, if an item of equipment meets the intent i
of the caveats, but the specific wording of the caveat rule is not met, then that item is considered to have met the caveat. At Quad Cities, a small number of SSEL items were judged to meet the intent, but not the exact wording of a caveat, and these cases are reported in Section 5 of this report.
4.1.3 Anchorage Adequacy Quad Cities verified anchorage adequacy with an approach incorporating three elements:
Comparison of the anchorage capacity with the seismic demand.
Evaluation of the anchorage to verify that it is free of gross installation defects.
Evaluation of the equipment anchorage load path to verify that there is adequate stiffness and strength.
The screening approach for verifying the seismic adequacy of equipment anchorage is based upon a combination of inspections, analyses, and engineering judgment.
Inspections consist of.T.Gawaments and visual evaluations of the equipment and its anchorage, supp!amented by use of plant documentation and drawings. Analyses compare the anchorage capac4y to the seismic loadings (demand) imposed upon the anchorage. These analyses v/ere done using the guidelines in Section 4 and Appendix C of the GIP. Enaineerina f adament is also an important element in the evaluation of equipment anchorage. As L general rule, all significantly sized equipment was rigorously analyzed using the ANCHOR software package developed by Stevenson &
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i Quad Cities A-46 Final Report June 99.1996 Associates [19). Small equipment, weighing usually 50 lbs or less was accepted by
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judgment and a " tug test". The tug test simply involves pulling on the device (say a wall-mounted transmitter) with a force to exceed 3 times the expected seismic demand for the equipment location. This demonstrates, as a minimum, a factor of safety of 3 for the equipment anchorage, consistent with the anchorage evaluation criteria in the GIP.
l The four main steps used to evaluate seismic adequacy of equipment anchorages at Quad Cities followed the guidance of the GIP and are shown below:
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1.
Anchorage Installation Inspection i
2.
Anchorage Capacity Determination 3.
Seismic Demand Determination 4.
Comparison of Capacity to Demand The first main step in evaluating the seismic adequacy of anchorages is to check the anchorage installation and its connection to the base of the equipment. This inspection consists of visual checks and measurements along with a review of plant documentation and drawings where necessary, and an anchor bolt tightness and embedment check for anchorage utilizing concrete expansion anchors.
All accessible anchorages were visually inspected. A check of the following equipment anchorage attributes was made:
1.
Equipment Characteristics 2.
Type of Anchorage 3.
Size and Location of Anchorage 4.
Installation Adequacy 5.
Embedment Length 6.
Gap atThreaded Anchors 7.
Spacing Between Anchorages 8.
Edge Distance 9.
Concrete Strength and Condition 10.
Concrete Crack Locations and Sizes 11.
Essential Relays in Cabinets 12.
Equipment Base Stiffness / Prying Action 13.
Equipment Base Strength / Structural Load Path 14.
Embedment Steel and Pads For expansion anchors, a tightness check was performed to detect gross installation defects (such as oversized concrete holes, totallack of preload, loose nuts, damaged subsurface concrete, and missing plug for shell types) which would leave the anchor loose in the hole. The tightness check for expansion anchors was accomplished by applying a torque to the anchor by hand until the anchor was " wrench tight," i.e.,
tightened without excessive exertion. If the anchor bolt or nut rotates less than about 1/4 tum, then the anchor is considered tight. The tightness check was performed on all accessible expansion anchors for floor mounted equipment where the anchorage adequacy is performed by analysis rather than " tug test" as described above. Wall mounted equipment was excluded as allowed by the GIP because the anchors 4-4
Quad Cities A-46 Final Report June 19,1996 experience some tensile loading due to gravity. A random (" spot") embedment check on selected anchors was performed, inspecting them to ensure that the shell anchor and equipment base are not in contact so as to invalidate the results of the tightness check.
j In summary, one loose anchor each was found on control panels 901-50,902-49,902-32 and 901-39. The anchors were tightened on the first three panels. Panel 901-39's loose anchor could not be tightened and a repair order has been issued to immediately repair the anchor; thus, it will have been repaired by the time this report is issued. This panel was, however, declared an outlier for reason of the loose anchor problem.
Lastly, one gapped anchorage has been found on Panel 902-48. This anchor will be disassembled (as it is a shell type anchor) and will be inspected for embedment and any other problems that may become evident. The panel has also been declared an outlier. Detailed results of the tightness checks and embedment checks are found in Appendix F. The rest of CEAs for which the tightness check was performed were found to be adequate. Based on the embedment checks and plant documentation, the predominant expansion anchor type at Quad Cities is the WEJ-IT drop-in shell anchor which requires no knock-down factor for anchor type. Newer installations utilize Hilti-Kwik bolt wedge type expansion anchors.
The second main step in evaluating the seismic adequacy of anchorages is to determine the allowable capacity of anchors used to secure an item of equipment. The allowable capacity is obtained by multiplying ths nominal allowable capacities by the applicable capacity reduction factors. The nominal capacities and reduction factors are obtained from Appendix C of the GIP, based on the results of the anchorage instaliation inspection checks. The nominal allowable tensile and shear capacities are established in EPRI Report NP-5228-SL (15]. The nominal allowable capacities incorporate a design safety factor of 3 between the ultimate and allowable (working) capacities.
The pullout capacity allowable is based on the product of the nominal pullout capacity and the applicable capacity reduction factors based on identifying the appropriate anchor type and make:
Pon = Pnom RT, RLp RS, RE, RF, RC, RR, Where: P,n = Allowable Pullout capacity of installed anchor (kip)
Pnom= Nominal allowable Pullout capacity (kip)
RTp= Heduction factor for the Type of expansion anchors R Lp = Reduction factor for short embedment Lengths RSp= Heduction factor for closely Spaced anchors rep = Reduction factor for near Edge anchors RF,= Beduction factor for low strength (fe) concrete RCp= Eeduction factor for Cracked concrete RRp= Heduction factor for expansion anchors securing equipment with essential Eelays The shear capacity allowable is based on the product of the nominal shear capacity and the applicable capacity reduction factors:
4-5
i Quad Cities A-46 Final Report June 19.1996 Von = Vnom RT. RL. RS. RE. RF. RR.
Where: V n = Allowable shear capacity of installed anchor (kip)
Vnom= Nominal allowable shear capacity (kip)
RT.= Beduction factor for the Type of expansion anchors RL.= Reduction factor for short embedment Lengths RS.= Beduction factor for closely Spaced anchors RE.= Reduction factor for near Edge anchors RF.= Beduction factor for low strength (f.) concrete RR.= Reduction factor for expansion anchors securing equipment l
with essential Relays Note that the pullout and shear capacities for anchors given above are based on having adequate stiffness in the base of the equipment and on not applying significant prying action to the anchor. If Check 12, Base Stiffness and Prying Action, from Part II, i
Chapter 4 of the GIP shows that stiffness is not adequate or that significant prying action is applied to the anchors, then the Seismic Capability Engineers lowered the allowable capacity loads accordingly, normally by completely discounting the affected l
bolt.
The third step in evaluating the anchorages is to determine the seismic demand imposed on the equipment. The demand load is established based on the type of demand spectrum used, if the amplified ISRS are used, no additional factors of l
conservatism are used to establish the demand load since the ISRS are deemed i
" conservative design" by review of the NRC. The demand load is the product of the l
appropriate spectral acceleration value times the weight of the equipment item. Table C.1-1 of the GIP is used, in general, to establish the fundamental frequency and equipment damping for the given classes of equipment. If the item is deemed rigid, the zero period acceleration (ZPA) is used. If the item is deemed flexible, the peak of the response spectrum is used. If the fundamental frequency is given in the SEWS, then l
the largest spectral acceleration in the range from that estimated frequency to the ZPA is used, if the ground spectrum is used for demand, then 1.875 times the appropriate spectral acceleration is used where 1.875 is the product of 1.5, the median amplification factor, and 1.25, the additional anchorage factor of conservatism for non-conservative demand spectra.
The fourth and final step to complete the evaluation determines the seismic de ad on the equipment anchorage and compares the seismic demand to the anchorage capacity. The demand on the anchorage is calculated by applying the demand load at the equipment center of gravity. If the demand is less than the capacity, the anchorage l
is acceptable; otherwise, the equipment item is declared an outlier.
l l
4.1.4 Selsmic Interaction Checks l
The fourth and fina' screening guideline used to verify the seismic adequacy of an item of mechanical or electrical equipment was to confirm that there were no adverse seismic spatial interactions with nearby equipment, systems, and structures which could 4-6
1 i
l Quad Cities A-46 Final Report 3 --
June 19,1996 cause the equipment to fail to perform its intended safe shutdown function. The interactions of concem are (1) proximity effects, (2) structural failure and falling, and (3) flexibility of attached lines and cables. Guidelines forjudging interaction effects when verifying the seismic adequacy of equipment are presented in Appendix D of the GIP, During the plant walkdowns at Quad Cities, the SRT's identified only a few interaction concems. These particular issues and their resolution are discussed in detailin Section 8.
Overhead piping systems and ductwork were closely examined in all plant areas containing A-46 and seismic IPEEE equipment. The SRT's identified no vulnerabilities and noted that the systems were well supported.
4.2 OutilerResolution An outlier is defined as an item of equipment which does not meet the screening guidelines noted above. An outlier may be shown to be adequate for seismic loadings by performing additional evaluations such as the seismic qualification techniques currently being used in newer nuclear power plants. These additional evaluations and alternate methods were thoroughly documented on the OSVS forms.
4.3 Seismic Capability Engineers and Peer Reviewer The guidelines described in this section were applied by Seismic Capability Engineers as defined in Section 2 of the GIP. These engineers exercised engineering judgment based upon an undemtanding of the guidelines given in this document, the basis for these guidelines given in the reference documents and presented in the SQUG training course, and their own seismic engineering experience.
The station walkdowns were conducted from April 3-7, May 15-17, July 10-14, and September 18-22,1995; and February 26-29 and April 1-4,1996. The seismic capability engineers for the Quad Cities walkdown were Dr. J. D. Stevenson, Mr. G. G.
Thomas and Mr. W. Djordjevic of S&A, Dr. R. P. Kennedy of RPK Structural Mechanics, Inc. and Messrs. B. M. Lory and K. Adlon of Commonwealth Edison Company. All have been SQUG trained and cert;fied. Their resumes and/or SQUG Walkdown Course Completion Certificates are previded in Appendix C.
An independent evaluation and peer review of the walkdown process was performed by Mr. Harry Johnson of Programmatic Solutions during May 6-8,1996. As required by the GIP, the review included an assessment of the walkdown and analyses by audit and sampling to identify any gross errors. Mr. Johnson personally conducted two days of walkdowns to ascertain completeness and correctness of the A-46 walkdown. His review included comparing completed SEWS with equipment previously inspected by the SRTs. Mr. Johnson also reviewed the documentation packages the SRTs used to determine equipment design details that could not be readily determined by walkdown.
- Mr. Johnson concluded that the walkdowns were being conducted competently and the 4-7
. ~.. -. - ~ -. - -.. -.. - =. - - - -. -
1 Quad Cities A-46 Final Report June 19.1996 1
i findings made were appropriate. Appendix E provides documentation of Mr. Johnson's peer review.
)
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y 4.4 Other Types of Seismic Evaluations andInterfaces
\\
l in addition to the seismic evaluations covered in this section for active mechanical and 4
1 electrical equipment, seismic evaluations for two other types of equipment are covered in other sections as follows:
4 Section 6 - Tanks and Heat Exchangers Review j
Section 7 - Cable and Conduit Raceways Review A separate Relay Evaluation Report will be issued to document the results of the relay functionality review required in Section 6 of the GlP.
While these other seismic evaluations can generally be performed independently from those for active mechanical and electrical equipment, there are a few areas where an interface with the Relay Functionality Review is appropriate:
. Any cabinets containing essential relays, as determined by the relay review, j
should be evaluated for seismic adequacy using the guidelines contained in this section.
A capacity reduction factor should be applied to expansion anchor bolts which secure cabinets containing essential relays as well as a "no gap" provision for base anchorage. These provisions are discu: sed in Section 4.4 and Appendix C of the GlP.
Seismic interaction, including even mild bumping, is not allowed on cabinets containing essential relays. This limitation is discussed in Section 4.5 of the GIP.
In-cabinet amplification factors for cabinets containing essential relays may be estimated, using the guidelines in Section 6 of the GIP, by the Seismic Capability Engineers for use in the Relay Functionality Review.
4.5 Documentation Quad Cities documented the results of the Screening Verification and Walkdown on Screening Verification Data Sheets (SVDS)in Appendix D and Screening Evaluation Work Sheets (SEWS).
As discussed in Section 4.4, the discussion of the review of Heat Exchangers & Tanks and Cable Tray & Conduit Raceways is given in Sections 6 and 7, respectively. SEWS 4-8
. ~.. _. -...
Quad Cities A-46 Final Report June 19,1996 1
l for Class 21 equipment, Heat Exchangers and Tanks, and Plant Area Summary Sheets (PASS) for the Cable Tray & Conduit Raceway Reviews were also developed.
Outliers for all equipment are discussed in Section 8. The Relay Functionality j
Assessment is given in a separately bound report entitled, "USNRC USl A-46 Resolution, Relay Evaluation Report"[21].
3 i
i 4-9
e 1
Quad Cities A-46 Final Report June 19,1996 4.6 Evaluation Results - Equipment Classes 0 Through 20 The seismic review SSEL list contains approximately 880 equipment items excluding heat exchangers & tanks. Of this population,114 items were declared outliers. Based on the electrical raceway walkdowns and subsequent analytical reviews,9 electrical raceway outliers were also declared. For a discussion of the outliers see Section 8.
i
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Quad Cities A-46 Final Report June 19,1996
- 5. GIP Deviations and Commentary on Meeting The Intent of Caveats No significant or programmatic deviations from the GIP were made while performing the walkdowns and seismic adequacy evaluations at Quad Cities for resolution of USI A-
- 46. Very few interpretations were made with respect to the wording of the GIP caveats I
versus the caveat's intent. This section lists those interpretations or measures taken to meet the intent of the caveat in Table 5.1 below. All other equipment not listed in this i
table met the caveat rules as stated in the GlP.
Table 5-1 Commentary Regarding GIP Deviations Equipment ID and Description Commentary MOTOR CONTROL CENTERS Seismic demand does not significantly Uceed capacity.
(MCCS)
MCC 18/19-5 MCC 28/29-5 AIR HANDLERS (COOLERS)
Seismic demand does not exceed capacity at frequency of 2-5746A hung cooler.
2-5746B 1-5748A 1-5748B 2-5748A 2-5748B
- Coolers AIR OPERATED VALVES Mounted on 3/4" line which is less than 1" minimum 1 &2-0220-44-AO required by GlP. Valve has small offset and line is lightly 1&2-0220-45 loaded, so intent of caveat is met.
1&2-8801 A-AO 1&2-88018-AO (In the case of 1&2-0220-45, the line is well supported at or 1&2-8801C-AO near the valve body so that line bending and torsional 1&2-8801D-AO stress is minimized, so the intent of the caveat is met.)
1&2-8802A-AO 1&2-88028-AO 1&2-8802C-AO 1&2-8802D-AO TRANSFORMER AND RACKS These anchors are grouted-in-place; however, cast-in-place anchorage allowables (for tensile capacities) were used. It 2251-97 was established by Drawings M-684 and M-354 that this 2251-98 grout is a non-shrink type. The grouted anchorages were 2251-100 themselves closely inspected and found to be in excellent condition with no shrinkage or cracking is evident.
2252-97 2252-98 1
5-1
Quad Cities A-46 Final Report June 19,1996
- 6. Results of the Tanks and Heat Exchanger Review Tanks and heat exchangers were evaluated in accordance with the rules and procedures given in Section 7 of the GIP (2).
This section gives the results of the tank and heat exchanger reviews performed. In total,24 tanks and heat exchangers were evaluated. Seven (7) tanks and heat exchangers were declared outliers generally due to exceeding anchorage allowables.
1 6.1 EVALUATIONMETHODOLOGY The screening evaluations described in this section for verifying the seismic adequacy of tanks and heat exchangers cover those features of tanks and heat exchangers which experience has shown can be vulnerable to seismic loadings. These evaluations include the following features:
l Check that the shell of large, flat-bottom, vertical tanks will not buckle. Loadings
+
on these types of tanks include the effects of hydrodynamic loadings and tank wall flexibility.
Check that the anchor bolts and their embedments have adequate strength against breakage and pullout.
Check that the anchorage connection be'. ween the anchor bolts and the tank
+
shell (e.g., saddles, legs, chairs, etc.) nave adequate strength.
Check that the attached piping has adequate flexibility to accommodate the motion of large, flat-bottom, vertical tanks.
The Seismic Capability Engineers reviewed these evaluations to verify that they meet the intent of these guidelines. This review included a field inspection of the tank, the anchorage connections, and the anchor bolt installation against the guidelines described in this Section 7, Section 4.4, and Appendix C of the GlP [2].
The derivation and technical justification for the guidelines utilized were developed specifically for: (1) large, flat-bottom, cylindrical, vertical, storage tanks; and (2) horizontal cylindrical tanks and heat exchangers with support saddles made of plates.
The types of loadings and analysis methods described in this section are considered to be appropriate for these types of tanks and heat exchangers; however, a generic procedure cannot cover all the possible design variations. Other design features such a
as wall mounted heat exchangers, heat exchangers and vertical tanks supported on legs not covered by the GIP were evaluated using the same procedures and loading conditions as given in Section 7 of the GIP.
Other types of tanks and heat exchangers (e.g., vertical tanks supported on skirts and structural legs) which were not specifically covered by the guidelines in Section 7 of the 6-1
^
L Quad Cities A-46 Final Report l
June 19,1996 l
GIP were evaluated by the Seismic Capability Engineers using an approach similar to l
that described in Section 7 of the GIP.
The other types of tanks covered by the screening guidelines in Section 7 of the GIP are l
cylindrical steel tanks and heat exchangers whose ax9s of symmetry are horizontal and are
)
supported on their curved bottom by steel saddle plates. The screening guidelines are based on the assumption that the horizontal tanks are anchored to a stiff foundation, which l
has adequate strength to resist the seismic loads applied to the tank. All the base plates under the saddles are assumed to have slotted anchor bolt holes in the longitudinal direction l
to permit thermal growth of the tank, except for the saddle at one end of the tank which is fixed. The saddles are assumed to be uniformly spaced a distance S apart, with the two ends of the tank overhanging the end saddles a maximum distance of S/2.
(
A simple, equivalent static method is used to determine the seismic demand on and capacity of the anchorages and the supports for horizontal tanks. The screening guidelines contained in Section 7 of the GlP specifically addressed only the seismic loads due to the inertial response of horizontal tanks. If, during the Screening Verification and Walkdown of a tank, the Seismic Capability Engineers determined that the imposed nozzle loads due to the seismic response of attached piping may be significant, then these loads were included in the seismic demand applied to the anchorage and supports of the tank. The nozzle loads were obtained from existing Comed piping analysis.
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Quad Cities A-46 Final Report June 19,1996 l
l 6.2 Summary of Evaluation Results The results of the A-46 evaluations are summarized below:
l Tsble 6-1 Tank & Heat Exchanger Evaluation Results l
No.
ID Description Type Resulta 1
1-5201 Fuel Oil Tanks Buried OK -Meets Design Basis in Accordance 1/2-5201 Horizontal with GlP Section 7. Rules 2-5201 Storage Tank 2
1-5202 Fuel Oil Day Horizontal Outlier - Fuel piping may not be flexible Tank Storage Tank enough to withstand the resulting displacement and one bolt may subject to bending.
3 1-1003A RHR7 Elevated Outlier-Commonwealth Edison is 1-1003B Erna gets Vertical Heat currently evaluating the adequacy of the 2-1003A Exchanger support steel. The A-46 evaluation of 2-1003B this equipment is pending the completion of Comed's review and any i
l modification to be made to this support.
4 1-4605 Blowdown Tanks Horizontal OK - Meets Design Basis in Accordance 1/2-4605 Storage Tank with GIP Section 7 Rules 2-4605 5
1-0203-1 AA Air Accumulator Honzontal OK - Meets Design Basis in Accordance 1-0203-1 BA Tanks Storage Tank with GIP Section 7 Rules 1-0203-1CA 1-0203-1DA 1-0203-2AA 1-0203-2BA i
1-0203-2CA 1-0203-2DA 6
1-6662 Expansion Tanks Horizontal OK-Meets Design Basis in Accordance 1/2-6662 Storage Tank with GIP Section 7 Rules 2-6662 I
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Quad Cities A-46 Fine' Report June 19,1996
- 7. Results of the Cable Tray and Conduit Raceway Review l
l 7.1 Summary of the Review This section documents the cable and conduit raceway review performed at the Quad Cities Generating Station, using the Generic Implementation Procedure (GIP) by the Seismic Qualification Utility Group (SQUG)[2]. The review includes both Unit 1 and Unit 2.
The raceway review was performed as specified in GIP Section 8. Raceway systems were walked down, checked against the Inclusion Rules and Other Seismic Performance Concems as specified in Section 8.2 of the GIP, and examined for seismic spatial interactions with adjacent equipment and structures. Twelve (12) l representative, worst-case raceway supports were selected and as-built. These supports then received a Limited Analytical Review per GIP Section 8.3 of the GIP.
Outliers were identified and documented.
This section (Section 7.1) summarizes the raceway review. The scope of the review is described in Section 7.2. Section 7.3 contains the results of the Limited Analytical Reviews (LARs). Section 7.4 summarizes the raceway outliers. Supporting l
documentation consists of the Plant Area Summary Sheets (PASS), the calculations l
performed for the Limited Analytical Reviews (LARs), and the Outlier Seismic Verification Sheets (OSVS).
7.2 Scope of the Review i
The scope of the review was determined by studying the plant layout drawings and l
walking through the plant to determine how cabling is routed to the areas of the plant containing the Safe Shutdown Equipment List (SSEL) equipment. It was concluded that the review should include:
all elevations of the reactor buildings excluding elevations 666' and 690', and a e selected non-safety areas of elevations 639' and 647',
selected vital areas of the turbine building including the electrical equipment room, e
diesel generator rooms, safe shutdown pump room,4 KV switchgear area, RHR and HPCI rooms, and battery rooms and electrical bus areas of elevations 615' and -
623',
the control room, e
and, in the crib house, the raceways suspended from elevation 595' to 554' above e
l the circulating water pumps, and the raceways in the service water pumps area.
The bulk of the walk down was performed in August,1995. The remainder of the walk down, which comprised the Unit 2 Torus Comer Rooms, the Unit 1 and Unit 2 Turbine Building Common Area, El. 611 and the Unit 1 and Unit 2 Diesel Generator and f.
Switchgear Rooms, was performed in February,1996.
i 7-1
)
1 Quad Ctties A 46 Final Report June 19,1996 The walk down of each plant area was documented on a PASS. Table 7.1 summarizes the PASS. The first column contains the ID used to track the PASS, the second column describes the area of the plant covered by the PASS, and the third column lists the LARs (if any) associated with the PASS.
Table 7.1 Summary of the Plant Area Summary Sheets (PASS)
PASS #
Plant Area L.AR #
RACE 001 Unit 1 and Unit 2 Cable Spreading Rooms LAR 001,002 RACE 003 Auxiliary Electrical Equipment Room RACE 004 Unit 1 Cable Tunnel LAR 004,005 RACE 005 Unit 2 Cable Tunnel LAR 006 RACE 006 North End of Turbine Building, El. 615 LAR 007 RACE 007 Unit 1 and Unit 2 Turbine Building, El. 639 LAR 008,009 RACE 008 South End of Turbine Building, El. 619 LAR 010 RACE 009 Unit 1 and Unit 2 Reactor Building, El. 595 LAR 011,012 RACE 010 Unit 1 and Unit 2 Reactor Building, El. 623 RACE 011 Unit 1 and Unit 2 Reactor Building, El. 647 RACE 012 Unit 1 and Unit 2 Reactor Building, El. 666 RACE 013 Unit 1 Torus, Comer Rooms RACE 014 Unit 2 Torus, Comer Rooms RACE 015 Unit 1 and Unit 2 Turbine Euilding Common Area, El. 611
)
RACE 016 Unit 1 and Unit 2 Diesel Generator and Switchgear Rooms 7.3 Limited AnalyticalReviews (LARs) 7.3.1 General Description of Quad Cities Raceways Quad Cities' raceway systems are primarily of light steel strut frame, steel angle frame, or rod-hung construction. The hangers of frame construction vary from the very simple i
wall mounted bracket supporting a few conduit to floor, ceiling, floor-ceiling, floor-wall and ceiling-wall mounted multi-tier frames supporting mostly cable trays and some conduit. The rod-hung trapeze hangers vary from single to multi-tier depths and also support mostly cable trays and some conduit.
Cable trays are primarily of ladder or solid bottom type construction with 6" side rails and vary in width from 12" to 44" The trays are sometimes sprayed with fire retardant or have steel covers. Conduits vary in size up to 5" nominal diameter and are of rigid steel material (standard schedule pipe). The trays and conduits are secured to hangers using standard tray clamps (clips), pipe clamps, or botting. No missing or damaged hardware was noted during the walkdowns The light steel strut frame posts are constructed of double channel members with double channel cross members interconnected with single and double-bolted clip 7-2
i I
Quod Cities A-46 Final Report June 19.1996 1
i angles. The s' eel angle frame supports are constructed with single angle posts and j
cross members interconnected with single bolts. Ceiling anchorages for both types of i
frame construction vary from embedded struts to building steel weldments. Floor anchorages are welded t ase plates with concrete anchor bolts. Wall anchorages are embedded struts and welded base plates with concrete anchor bolts. Rod hanger cross members are construt.+ed with single or double channel light steel struts or steel angle sections. Ceiling anchorages for rod hangers consist of embedded struts, shell
- type anchors and building steel weldments.
r 7.3.2 Limited Analytical Review Results The Limited Analytical Reviews (LARs) evaluate the seismic capability of the raceway supports which were selected as representative, worst case examples of the major types of raceway support configurations. Twelve (12) supports were selected for evaluation. The evaluations were performed per the guidelines in GIP Section 8.3.
Table 7.2 summarizes the LARs:
The first column lists the ID used to track the LARs.
The second column contains a brief description of the subject raceway support.
l The third column contains the maximum acceptable raceway weight for which the
)
e GIP acceptance criteria can be met. For trays, the weight is expressed in pounds j
per square foot (psf). For conduit, the weight is expressed as a percentage - 100%
)
corresponds to the weights in GlP Section 8.3.9 for steel conduit.
The fourth column contains the SRT's field estimate of the actual weight. For trays, e
the estimate is the SRT's conservative estimate of the actual percent fill and tray height scaled by the GIP Section 8.3.9 value of 25 psf for a 4" high tray with 100%
fill, plus a upper bound value of 5.7 psf for the dead weight of a 6" x 24" tray (scaled from Reference 5.8,4" x 12" tray = 8.9 psf). For conduit, the " estimate" is always 100% fill. If the fill estimate (4th column) is less than the acceptable fill (3rd column), then the support meets the GlP's acceptance criteria.
The fifth column lists the goveming load case.
GIP Section 8.3.9 states: " Conservative estimates should be made for the weights of other miscellaneous items attached to the raceway support, such as HVAC ducting, piping andlighting " The SRT did not observe any instances of significant additional weight due to attachments, therefore no additional weight was included in the LARs.
The Limited Analytical Review produced eight (8) outliers (LARs 001,002,004,005, 008,010,011 and 012). They are discussed in Section 7.4 and shown as shaded entries in Table 7.2.
73
Quad Cities A-46 Final Report June 19,1996 Table 7.2 Summary of the Limited Analytical Reviews conservative Acceptable Weight Goveming LAR #
Description Weight Estimate Load Case LAR 0017.6/8";thewaded rod trapeze, three (3) tiers ;;
3..,
W_. =J m h e t w tr m C E :a = r.
'. k--
_Getenee'tenswa;e:60_jo_toeMa.__ aug___ _ _ ss_p_g _ _,,gge!g_ Lea _dA Welded anchorage ~to Building Steels 24}r
<46.0 psf
, 45.0 psf <
iModfatiguee
..to top tierl
~
V?4
-.AR 002 Floor. mounted. Unistrut frame, 9; tall. 6j a23 psf 23.3 per,,,
Laters Loads
- L ~-
1 g g,,, gg) 44 g,,,jgg,gg 4
m w
hsQW Eg%
/
/
LAR 0041 Floor-to-colling suel._ angle frameg12' tallA n<<23 pef 3 W;23.3ps#y Lateral, Load 3 f
-- '~
h(8) tiets with one'(1) 44" tray / tier " ~
~~^
~ ' ' ' ~ ~ ~
LAR 005 Wall-ceiling mounted steel angle support, 22.8 psf 23.3 psf Dead Load four (4) tiers with one (1) 44' tray / tier LAR 006 Wall-floor mounted steel angle support, 40 psf 23.3 psf Dead Load 9' tall, eight (8) tiers with one (1) 44" tray / tier t.AR.00?i ;5/8* threaded md trapezer2 tiers with d n5"40 perd s^42.6 psip" Rodfatigue W1) 32' tray /tW.t8" to top tieM " F
' ' " ~
~~
~
~ 1.[L,
LAR.006j floor,mountedynistrut frame (5: tall,with; 1*1L5 psf 3 MSS.psg' ^
- ,LateralLoad3 2 tiers with one~(1)'24" tray / tier
~~"
~
~~
LAR 009 Wall bracket, two (2) 2" conduit and two
> 500 %
100 %
Dead Load (2) 5* conduit LAR 010; ;5/81 threaded rod trapeze,four.(4) tiers;,
,c_;,,cggggg$x
.$l.DMUlE lrgp_ler,[,4,,3, ";,,tp,, t,c, glierj;_,,,7; ;,;,;,;; ; ; _
. : Gateway,1,,rtser,t,,anctio,raf i. 11.. u.. 515.6 g(r,,,,,., 3,,$,,gst,6,;
83x Dead t.osd a
,,, 42
..: 1. '... -.. _:.:.1
.11.1:1:1 :.iLin.: ;?n
. ~.. _.
Shell anchoraosn ~' ~
' ' ' ' ' ' ~
321;7 asfr
' 4 42.6 ostr ~ I~3 DEIdW
~
LAR.011 5/8" threaded rod trapeze; three (3) ?
tieraltwo boys with one (1) 24' tray / tier, 24' J 60",to,t,,op tier,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,
Getene*'tenewaje__ ___ ____ _,_ _ ju g_ __ _ _ up_g;_ _,_3tgead1.ge_d_,
~
Weided anchorace to Building Steel
<32 per 32.0 psf Rod fatious LAR. 012 Floor mounted cable duct support. 3' pipe
<30% ~
100% c LateralLoad column. 9-1/2' tall supports 8'x24' cable 4
duct containing equivalent of fifteen (15) 3" conduit Outliers are shown as shaded entries in table above.
7-4
I l
ound Cities A-46 Final Report June 19,1996 I
7.4 Summary of Outliers A total of nine (9) outliers resulted from the cable and conduit raceway reviews. A l
description of each raceway outlier is provided in Table 7.3.
l Table 7.3 Raceway Outliers ID and Location Outlier issue and Resolution PASS # RACE 001 Outlier issue: An enveloping support in the Unit 1/2 Cable Spreading Cable Spreading Room Room was chosen for limited analytical review. The support is a 3-tier, rod l
LAR 001 hung trapeze supporting 3 trays. The support has two different types of l
ebiling anchorages: embedded strut and weldment to building steel. The l
embedded strut version loads exceed allowables for the vertical capacity check.
1 Resolution: This outlier is resolved by outlier analysis. A limit analysis per Section 8.4.8 was performed and the hanger passes. The system has 'short" hangers interspersed among longer hangers. Using the Redundancy and l
Consequence approach of Section 8.4.8, the short hangers were assumed to fail.
Rod fatigue evaluations were then performed on the longer rods. Rod fatigue data obtained from actual cyclic testing of threaded rods including the weldment anchorage plate from SEP Project 8050 (Ref. 27) was used to evaluate the rods.
Based on the
- Generic Rod Acceptability Curves", it was shown that the Quad Cities rods will sustain the SSE demand loads.
PASS # RACE 001 Outlier issue: A second enveloping support in the Unit 1/2 Cable Cable Spreading Room Spreading Room was chosen for limited analytical review. The support is LAR 002 an unbraced floor mounted Unistrut frame with 6 tiers supporting B trays.
The loads in the bolted post connections to the floor baseplates exceed allowables for the lateral load check.
Resolution: HCLPFsse is 0.12g peak ground acceleration (PGA). This hanger may require lateral bracing.
PASS # RACE 004 & 005 Outlier issue: An enveloping support in the Unit 1 Cable Tunnel was.
Unit 1&2 Cable Tunnels chosen for limited analytical review. The support is an eight tier, floor LAR 004 mounted with interspersed floor-to-ceiling, unbraced frames with 3x3 steel angle posts and 2x2 steel angle cross members supporting four trays. The welded floor anchorage is not ductile. The stress in the welds exceeds j
allowables for the lateral load check.
Resolution: HCLPFsse is 0.04g PGA. This hanger may require lateral bracing.
ID and Location Outlier issue and Resolution PASS # RACE 006 Outlier issue: An enveloping support in the Turbine Building, El. 615, Turbine Building North was chosen for limited analytical review. The support is a 2-tier, rod hung End trapeze supporting 2 trays. The support does not meet the rod fatigue LAR 007 check.
j Resolution: Rod fatigue data obtained from actual cyclic testing of threaded rods I
75
Quod Cities A.46 Final Report June 19.1996 including the weldment anchorage plate from SEP Project 8050 (Ref. 27) was used to evaluate the rods. Based on the
- Generic Rod Acceptability Curves", it was shown that the Quad Cities rods will sustain the SSE demand loads. A hmit analysis per Section 8.4.8 of the GIP was performed and the hanger is shown to be able to meet the SSE loadings.
PASS # RACE 007 Outlier issue: A second enveloping support in the Turbine Building, El.
Turbine Building Units 1/2 639, was chosen for limited analytical review. The' support is an unbraced, LAR 008 floor mounted, Unistrut frame with 2 tiers supporting 2 trays and 2, 3' conduit. The left post welded floor anchorage is not ductile. The stress in the welds exceeds allowables for the lateral load check.
Resolution: HCLPFsse is 0.24g PGA which meets the SSE demand with no additional margin per the CDFM method of NP-6041 Seismic margins Report.
T PASS # RACE 008 Outlier issue: An enveloping support in the Turbine Building, El. 619, Turbine Building So. End was chosen for limited analytical review. The support is a 4-tier, rod hung LAR 010 trapeze supporting 4 trays. The support has three different types of ceiling anchorages, the loads for two of which, the embedded strut and shell anchor versions, exceed allowables for the vertical capacity check.
F Resolution: A limit analysis per Section 8.4.8 of the GIP was performed and the hanger is shown to be able to meet the SSE loadings.
PASS # RACE 009 Outlier issue: An enveloping support in the Reactor Building Units 1/2, Reactor Building Units 1/2 El. 595, was chosen for limited analytical review. The support is a 2 bay,3-LAR 011 tier, rod hung trapeze supporting 6 trays. The support has two different types of ceiling anchorages: embedded strut and weldment to building steel. The embedded strut version loads exceed allowables for the vertical i
capacity check and the rod fatigue evaluation criteria. The welded attachment to building steel version does not meet the rod fatigue evaluation criteria.
Resolution: HCLPFSSE s 0.12g peak ground acceleration (PGA). This i
hanger may require lateral bracing. Additional walkdowns to identify this hanger type may be required in the Reactor building.
PASS # RACE 009 Outlier issue: A floor mounted HCU cable duct support in the Reactor Reactor Building Units 1/2 Building Units 1/2, El. 595, was chosen as an enveloping support for limited j
LAR 012 analytical review. The support consists of a 3' pipe column,9-1/2' tall i
supporting a 8'x24' cable duct containing the equivalent of fifteen (15) 3" conduit. The anchorage loads exceed the allowables for the lateral load j-check.
I 1
l Resolution: HCLPFsse is 0.08g peak ground acceleration (PGA). This hanger may require lateral bracing.
i 8.
L,
)
I i
7-6 i
i
,m.
..-.-4
-. +..
,r_---
Quad Cities A-46 Final Report Juro 19,1996
- 8. Resolution of Outliers This section discusses the outliers identified during the USI A-46 walkdowns conducted at Quad Cities. The outliers are identified from the Twenty Classes of Equipment discussed in Section 4, the Tanks & Heat Exchangers Review discussed in Section 6, and the Cable Tray & Conduit Raceway Review given in Section 7. Relay outliers are discussed in the Quad Cities Relay Evaluation Report (21].
An outlier is an item of equipment which does not comply with all of the screening guidelines provided in the GIP. The GIP screening guidelines are intended to be used as a generic basis for evaluating the seismic adequacy of equipment. If an item of equipment fails to pass these generic screens, it may still be shown to be adequate by additional evaluations.
Finally, a discussion of the disposition or corrective action, as appropriate, for each outlier is discussed below.
TABLE 8.1 GENERIC OUTLIER ISSUES ID EQUIPMENT OUTLIER FINDING RESOLUTION G1 numerous Supply cabinets, copies, Better seismic tables, carts, tool boxes, etc.
" housekeeping" needs to be are found near/ adjacent to implemented by the plant.
SSEL equipment.
j TABLE 8.2 A46 EQUIPMENT OUTLIERS ID EQUIPMENT OUTLIER FINDING RESOLUTION A1 Hydraulic Control Unit
- 1) No seismic capacity
- 1) Unit consists of valves and U1 HCU based on earthquake accumulators which are in expenence or genene the earthquake experience seismic testing ruggedness and generic seismic testing data is available.
ruggedness data. Rack load
- 2) Red gas bottles at South path and anchorage are bank restrained by only one separately analyzed and are chain as well as an orange OK. CRD piping and scram gas bottle and an header seismically unrestrained green trash supported. Outlier is can are impact hazards.
resolved.
- 2) Gas bottles need a second chain near the bottom. Re-locate trash can.
A2 Hydraulic Controi' Unit No seismic capacity based Unit consists of valves and U2 HCU on earthquake experience accumulators which are in 8-1
=
Quad Cities A-46 Final Report June 19,1996 ID EQUIPMENT OUTLIER FINDING RESOLUTION or generic seismic testing the earthquake experience ruggedness data is and generic seismic testing i
available for class O ruggedness data. Rack load equipment.
path and anchorage are separately analyzed OK.
CRD piping and scram header seismically supported. Outlier is resolved.
A3 Nitrogen Bottle Nitrogen bottle is restrained Gas bottles need a second N/A (1), N/A (2) only by top chain.
chain nearthe bottom.
A4 Exhaust Silencer The silencer saddle not Analyze piping adequacy for 1-6667 positively supported.
support of silencer.
1/2-6667 2-6667 AS Fan and Damper Seismic demand exceeds Develop realistic, median 1-5727 capacity.
amplified floor spectra to 1/2-5727 potentially reduce seismic 2-5727 demand to an acceptable Motor Control Center level.
1 and Panel MCC 28-3 MCC 29-1 250VDC MCC 1 A 250VDC MCC 18 250VDC MCC 2A 250VDC MCC 28 Transformer MCC 281 A-1 TR MCC 29-1-1 TR SWGR 18 TR SWGR 19 TRANSFMR SWGR 28 TR SWGR 29 TR A6 Unit 2 250 VDC Battery Right rear anchor bolt install missing anchor bolt.
Charger missing.
A7 Motor Control Center
- 1) Seismic demand
- 1) Develop realistic, median and Panel exceeds capacity.
centered floor spectra.
MCC 18-1 A
- 2) Unit has missing or MCC 19-1-1 loose sheet metal screws
MCC 28-1 A to the base channel.
A8 Motor Control Center
- 1) Seismic demand
- 1) Develop realistic, median MCC 18-3 exceeds capacity.
centered floor spectra.
- 2) MCC is not bolted to adjacent distribution panel
- 2) Bolt MCC to adjacent (3/16" gap).
panel.
- 3) Missing several comer
- 3) Install missing screws and sheet metal screws which tighten loose screws.
8-2
Quad Cities A-46 Final Report June 19,1996 ID EQUIPMENT OUTLIER FINDING RESOLUTION attach the MCC units to the base channel.
A9 Motor Control Center
- 1) Seismic demand
- 1) Develop realistic, median en sp e a.
2 The e d ng in ludes both 19-1 and 19-6 and is
- embeds, should be upgraded on the
- 3) Install m,ssing mounting i
left end unit and the 2 right screws.
end units (next to 19-6).
- 3) Missing some sheet metal screws which attach the MCC unit to the base channel.
A10 Motor Control Center
- 1) Seismic demand
- 1) Develop realistic, mettian MCC 10-4 exceeds capacity.
centered floor spectra.
- 2) The weiding for 19-4 is not well distributed. Welds
- 2) Upgrade MCC to embed should be upgraded on the welding.
left end of 19-4 (next to 19-1-1).
A11 Panel
- 1) Seismic demand
- 1) Develop realistic, median 2201-32 exceeds capacity.
amplified floor spectra to p tentially reduce seismic
- 2) 1/4" QaPped anchora9e demand to an acceptable on one side of panel.
level.
- 2) Shim the gap closed since panel contains essential relays.
A12 Panel
- 1) Seismic demand
- 1) Develop realistic, median 2202-32 exceeds capacity, centered floor spectra.
A13 Damper Damper on small duct line Evaluate or modify damper.
2-9472-32 which may disconnect.
A14 Motor Control Center
- 1) Seismic demand
- 1) Develop realistic, median MCC 18-1 A-1 exceeds capacity.
amplified floor spectra to MCC 28-1 A-1 potentially reduce seismic MCC 18-1 A-1 PNL
level.
PNL #1
- 2) Bolt (or tie) units together PNL #2 since cubicles contain essential relays.
A15 Motor Control Center
- 1) Seismic demand
- 1) Develop realistic, median and Panel exceeds capacity.
centered floo. spectra.
- 2) MCC or Panel is not
(there is only a 3/8" gap).
8-3
Quad Cites A-46 Final Report June 19,1996 ID EQUIPMENT OUTLIER FINDING RESOLUTION A16 Motor Control Center
- 1) Seismic demand
- 1) Develop realistic, median e
Spe da.
MC 81B i right next to (touching) MCC 1 A-
- 2) Bolt MCC units together.
1 but not bolted together.
- 3) Unit has very deficient
added.
- 4) In; tall missing mounting
- 4) Missing two internal 1/4" screws.
machine screws that attach the unit to the base channel that is bolted to the embedded angle.
A17 Bus Panel is 1/2" away from Install neoprene or like 125 VDC BUS 1 A block wall on rear and it is material between wall and 125 VDC BUS 1 A-1 an interaction hazard.
panel.
A18 Bus
- 1) Anchorage demand
- 1) Evaluate anchorage 125 VDC Bus 1B exceeds capacity based on capacity based on more 125 VDC Bus 1B-1 current conservat,ve weight refined weight value.
i estimate.
- 2) Close S-hooks on lights
- 2) There are lights above above bus, the units with open hooks and it is an interaction issue.-
A19 Motor Control Center
- 1) We&d to em ded
Y welds.
even but unsymmetric.
Anchorage is unacceptable
inspection.
- 2) There are lights above the units with open hooks and it is an interaction issue.
A20 Panel Nearby cable tray support Attach cable tray support to 2202-70B is an impact hazard.
panel to eliminate impact.
A21 Panel Adjacent gas bottle with Add lower restraint to ensure 2212-32 single loose chain is an bottle does not slip out.
Impact hazard.
A22 Panel
- 1) Pa isn ed o
- 1) Bolt panel to adjacent E*"*
~
an impact hazard.
- 2) All anchors are
- 2) Design positive unwelded friction clips.
attachment to support steel.
- 3) Bookcase behind panel
- 3) Restrain or move is an interaction hazard.
bookcase.
A23 MCC & Battery Chargers Overhead fluorescent light Repair (close S-hooks) on MCC 18-2 above hung with an operi overhead light.
1-8300-1 A S-hook 8-4
Quad Cities A-46 Final Report June 19,1996 ID EQUIPMENT OUTLIER FINDING RESOLUTION 1-8350 2-8300-1 2-8300-1 A 2-8350 1-8300-1 A24 Cooler Cooling water line lacks Coolers are rod hung and flexibility and since cooler thus flexible. Laterally 1-5746A s rod hung so water line restrain coolers to preclude 1-57468 may be in danger of piping rupture potential or 2-5746A breaking due to perform a detailed analysis of 2-5746B displacement potential of piping flexibility / loading on 1-5747 rods.
cooler or evaluate effects of 2-5747 loss of service water 1-5748A inventory and loss of cooling 1-5748B capability.
2-5748A 2-5748B A25 Cooler Information on bolt type is Capacity of Cinch Anchoris 1-5745A not available. Bolt type based on " Lead Expansion 1-5745C and size are assumed to be Anchor Load Capacity in 2-5745A the same as the ones used Reactor Building at the 2-5745C in Dresden which are 1/2*
Savannah River Site",
1-5749 Cinch Anchor. Bolt type is Westinghouse Savannah 1/2-5749 not covered by the GIP -
River Company, RTR-2661, 2-5749 Cinch Anchor.
Aug.15,1989 (Refs 22 &
23). The resulting anchorage capacity was shown to exceed the design basis seismic demand loads and outlier is resolved.
A26 Switchgear Overhead trolley hoist is an Use a clamp or similar SWGR 28 Impact hazard.
device to prev 3nt the hoist SWGR 18 from rolling freely.
i SWGR 19 A27 Switchgear Overhead trolley hoist is an Restrain trolley to eliminate SWGR 29 impact hazard and needs impact hazard.
to be parked.
A28 Switchgear
- 1) Could not open units to 1&2) Schedule an SWGR 13 determine if the units are appropriate time for an SWGR 14 plug welded at the base.
Intemal inspection.
- 2) Not sure end two units of
- 3) Remove breaker or SWGR 14-1 are tied to positively restrain.
SWGR 23-1 SWGR 24-1 other 10 units.
i
- 3) Spare breaker stored i
8-5
Quad Cities A-46 Final Report Jurt] 19,1996 i
ID EQUIPMENT OUTLIER FINDING RESOLUTION near SWGR 24-1 is an impact hazard.
l A29 Switchgear
- 1) Could not open units to
- 1) Schedule an appropriate SWGR 23 determine if the units are time for an intemal j
SWGR 24 plug welded at the base.
inspection.
j
- 2) A light with open S-
- 2) Repair (close S-hooks) hooks above the overhead light.
switchgear is an impact hazard.
- 3) Spare breaker stored near SWGR 24 is an impact hazard.
A30 Panel Anchor at comer of panelis Tighten /repairloose anchor.
901-39 loose.
A31 Panel Panel has one gapped Tighten / repair anchor.
902-48 anchor (loose washer).
A32 Panel Panelis not bolted to Connect (bolt) panels 2202-70B adjacent panel together to preclude impact 901-32 potential.
901-46 901-47 901-48 902-39 902-46 A33 Panel Cabinet is too close to a insert neoprene or like 2252-10 conduit snugly against the material between wall and wall which thus poses an panel, or positively secure impact hazard.
panel to wall.
A34 Panel Tool Box (on wheels)
Move tool box and stress 2251-12 located next to Panels the seismic housekeeping to 2251-98 Exciter which thus poses plant personnel.
an impact hazard.
A35 Panel Adjacent ductwork Insert neoprene or like 902-33 supported on rod hangers material between duct and is 1/4" away (gap) from top panel, or positively secure of cabinet and poses an panel to duct.
interaction hazard.
A36 Panel Panelintemal PC card Modify internel rack (stiffen 901-27 racks for which cards it) and restrain PC cards with 902-27 falling out of very flexible strap or bar to preclude cards intemal rack is a concem.
from falling out.
A37 Panel
- 1) Panel is not bolted to
- 1) Connect (bolt) panels 901-33 adjacent panel 901-47, together to preclude impact thus it is an impact hazard.
potential.
- 2) Adjacent ductwork
- 2) Connect (bolt) panels supported on rod hangers together to preclude impact 8-6
-. _ ~
Quad Cities A-46 Final Rcport June 19,1996 ID EQUIPMENT OUTLIER FINDING RESOLUTION is 1/4" away (gap) from top potential. Insert neoprene or of cabinet and poses an like material between duct interaction hazard.
and panel, or positively secure panel to duct.
l A38 Battery Rack Some Styrofoam spacers insert " full height" Styrofoam 125 VDC BATT 1 on the front or ends of the spacers, or secure (glue or 250 VDC BATT 1 rack are short and can fall tie) spacers to rack (or i
125 VDC BATT 2 loose (that is, through) and battery cells) so that they a few already have even cannot slide out.
though there has been no earthquake, A39 Battery Rack
- 1) Some Styrofoam
- 1) Insert " full height" spacers on the front or Styrofoam spacers, or secure 250 VDC BATT 2 ends of the rack are short (glue or tie) spacers to rack and can fall loose (that is, (or battery cells) so that they through) and a few already cannot slide out.
have even though there has been no earthquake.
- 2) Determine qualified life of i
cells or replace.
- 2) Battery age is unknown.
?
A40 Valve Valve is located in RWCU Review of photographs l
1-0220-45 room which is inaccessible shows supports at or near 2-0220-45 therefore Caveats 7-9 and valve (body); thus, the line i
the review for seismic will not be subjected to large interaction cannot be bending forces and is resolved. Valve is located adjudged acceptable. This 1
on a 3/4"line which lies outlier is considered I
i outside of the GlP
- resolved, l
1 database thus Caveats 4 and 5 cannot be resolved.
l A41 Rack Rack mounted (enclosed in Replace it with Mercoid Snap PE-1 panel) Mercoid switch Action (Orange High) which is designated as mechanical switch or
" bad actor".
equivalent.
A42 Pressure Switch Missing one of the 4 Repair (replace) missing 1-4641-42A screws attaching it to the screw.
support.
A43 RHR Heat Exchanger Commonwealth Edison is SQUG review pending.
1-1003A currently evaluating the 1-1003B adequacy of the support 2-1003A steel, hence the A-46 2-1003B evaluation of this equipment is pending the completion of Comed's review and any modification to be made to j
this support.
8-7
Quad Cities A-46 Find R port i
l Juni 19,1996 l
l 1
i ID EQUIPMENT OUTLIER FINDING RESOLUTION A44 Cable Spreading Poom~
An enveloping support in This outlier is resolved by outlier LAR 001 the Unit 1/2 Cable analysis. A limit analysis per Spreading Room was Section 8.4.8 was performed chosen for limited and the hanger passes.
l analytical review. The support is a 3-tier, rod hung
)
trapeze supporting 3 trays.
The support has two I
(
different types of ceiling anchorages: embedded stmt and weldment to building steel. The embedded strut version loads exceed allowables for the vertical capacity check.
A45 Cable Spreading Room A second enveloping HCLPFsse is 0.12g peak LAR 002 support in the Unit 1/2 ground acceleration (PGA).
Cable Spreading Room This hanger requires more was chosen for limited detailed analysis to analytical review. The determine if lateral bracing is support is an unbraced required to achieve the SSE floor mounted Unistrut demand level.
frame with 6 tiers supporting 6 trays. The loads in the bolted post j
connections to the floor baseplates exceed allowables for the lateral load check.
Unit 1&2 Cable Tunnels An enveloping support in HCLPFsse is 0.04g PGA.
A46 LAR 004 the Unit 1 Cable Tunnel This hanger type requires was chosen for limited further investigation to analytical review. The determine the extent of the support is an eight tier, problem and whether or not floor to ceiling, unbraced lateral bracing will be frame with 3x3 steel angle required to achieve the SSE posts and 2x2 steel angle demand level.
cross members supporting four trays. The welded floor anchorage is not ductile. The stress in the welds exceeds allowables for the lateral load check.
This hanger type is prevalent in the Unit 1 Cable Tunnel. In the Unit 2 Cable Tunnel this hanger type is found only on the east wall at the Unit 1 side.
A47 Turbine Building Units An enveloping support in This outlier is resolved by outlier 1/2 the Turbine Building, El.
analysis. A limit analysis per LAR 007 639, was chosen for limited Section 8.4.8 was performed 8-8
1 Quad Cities A-46 Final Report June 19,1996 ID EQUIPMENT OUTLIER FINDING RESOLUTION analytical review. The and the hanger passes.
support is a 2-tier, rod hung trapeze supporting 2 trays.
The support loads in the welded steel anchorage exceed the allowables for the vertical capacity check.
does not pass The rod fatigue check was also not met.
A48 Turbine Building Units A second enveloping HCLPFsse is 0.24g PGA 1/2 support in the Turbine which meets the SSE LAR 008 Building, El. 639, was demand with no additional chosen for limited margin per the CDFM analytical review. The method of NP-6041 Seismic j
support is an unbraced, margins Report. A more floor mounted, Unistrut detailed analysis is needed to frame with 2 tiers determine if more design supporting 2 trays and 2,3" margin is available or conduit. The left post whether or not a design j
welded floor anchorage is modification will be needed, not ductile. The stress in the welds exceeds allowables for the lateral load check.
A49 Turbine Building So. End An enveloping support in This outlier is resolved by outlier i
LAR 010 the Turbine Building, El.
analysis. A limit analysis per 619, was chosen for limited Section 8.4.8 was performed i
analytical review. The and the hanger passes.
support is a 4-tier, rod hung trapeze supporting 4 trays.
The support has three j
different types of ceiling anchorages, the loads for j
~
two of which, the embedded strut and shell anchor versions, exceed allowables for the vertical capacity check.
A50 Reactor Building Units An enveloping support in HCLPFue is 0.12g peak 4
1/2 the Reactor Building Units ground acceleration (PGA).
..1 LAR 011 1/2, El. 595, was chosen i
This hanger may require for limited analytical lateral bracing. Additional review, The support is a 2 walkdowns to identify this e
bay,3-tier, rod hung hanger type may be required trapeze supporting 6 trays.
in the Reactor and Turbine The support has two buildings.
different types of ceiling anchorages: embedded strut and weldment to 4
8-9 4
Quad Cities A-46 Final Report June 19,1996 l
lD EQUIPMENT OUTLIER FINDING RESOLUTION l
building steel. The I
embedded strut version loads exceed allowables for the vertical capacity check and the rod fatigue evaluation criteria due to short hanger rods interspersed among longer rod hangers. This general problem of mixed long and short hangers exists in the Turbine building, as well.
A51 Reactor Building Units A floor mounted HCU cable HCLPFsse is 0.08g peak 1/2 duct support in the Reactor ground acceleration (PGA).
LAR 012 Building Units 1/2, El. 595' This hanger may require was chosen as an lateral bracing.
enveloping support for
]
limited analytical review.
The support consists of a 3" pipe column, 9-1/2' tall supporting a 8"x24" cable duct containing the equivalent of fifteen (15) 3" conduit. The anchorage loads exceed the allowables for the lateral load check.
8-10
Quad Cities A-46 Final Report June 19,1996
- 9. References i
1.
Generic Letter 87-02," Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety issue (USI) A-46",
USNRC, Washington, D.C., February 19,1987.
2.
" Generic implementation Procedure (GlP), for Seismic Verification of Nuclear Station Equipment", Revision 2, Corrected,2/14/92, Seismic Qualification Utility Group.
3.
" Supplemental Safety Evaluation Report No. 2 (SSER #2) on GIP-2", USNRC, Washington, D.C., May 22,1992.
4.
USl A-40 " Seismic Design Criteria Short-Term Program", USNRC, Washington, D.C.
5.
USl A-17 " Systems Interactions in Nuclear Power Plants" USNRC, Washington, D.C.
6.
Comed Response to GL 87-02, Supplement 1 on SQUG Resolution of USl A-46 Dresden Nuclear Power Station, Units 2 and 3, Quad Cities Nuclear Power Station, Units 1 and 2, and Zion Nuclear Power Station, Units 1 and 2, Letter to USNRC, dated September 21,1992.
7.
USNRC Letter " Evaluation of Dresden Nuclear Power Station, Units 2 and 3, Quad Cities Nuclear Power Station, Units 1 and 2, and Zion Nuclear Power Station, Units 1 and 120-day Response to Supplement No.1 to Generic Letter 87-02 (TAC Nos. M69442, M69443, M69476, M69477, M69492 and M69493)", C.
~
P. Patel (USNRC) to T. J. Kovach (Comed), dated November 201992.
4 8.
Comed, " Updated Final Safety Analysis Report for Quad Cities Nuclear Station, Units 1 and 2.
9.
SPECTRA Software Package, Stevenson & Associates, Version 2, November, 1992.
4
- 10. EPRI Report NP-7146, " Development of In-Cabinet Amplified Response Spectra for Electrical Panels and Benchboards." Revision 0, Electric Power Research Institute, Palo Alto, CA, prepared by Stevenson & Associates, December,1990.
- 11. " Cable and Conduit Raceway Limited Analytical Review (LAR) for USl A-46 at Quad Cities Nuclear Station" Calc. 93C2806.04-C-002, Rev.0,93C2806.04 by 1
Stevenson and Associates.
- 12. SSRAP Report,"Use of Seismic Experience Data to Show Ruggedness of l
Equipment in Nuclear Power Plants," Senior Seismic Review and Advisory Panel, Revision 4.0, February 28,1991.
1 91
a l
i Quad Cities A-46 Final Report June 19,1996 I
- 13. Commonwealth Edison Company, Quad Cities Nuclear Station - Drawings. (Dwg.
j numbers are specified where referenced).
i 2
- 14. EPRI Report NP-5228-SL, " Seismic Verification of Nuclear Station Equipment Anchorage (Revision 1)." Electric Power Research Institute, Palo Alto, CA, I
prepared by URS/ John A. Blume & Associates, Engineers, June,1991.
1
- 15. ACI 318-83, " Building Code Requirements for Reinforced Concrete", American Concrete Institute,1983.
l l
- 16. EPRI Report NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin (Revision 1).", Electric Power Research Institute, Palo Alto, CA, prepared by JR Benjamin Associates et. al., August,1991.
1 i
- 17. ANCHOR 3.0 Software Package (with Verification and Users Manuals), Rev.0, 8/16/90 by Stevenson and Associates.
)
- 19. Comed, " Safe Shutdown Equipment List (SSEL) Report for Quad Cities Nuclear j
Station, Units 1 and 2", May,1996.
l
- 20. Comed Quad Cities Nuclear Station Units 1 and 2, "USNRC USl A-46 Resolution, q
Relay Evaluation Report," Rev. O, June 1996.
?
- 21. West, D. A and Griffin, M. J., " Lead Expansion Anchor Load Capacity," Current issues Related to Nuclear Power Plant Structures, Equipment, and Piping, Proceedings of the Fifth Symposium, Orlando, Florida, December 1994.
- 22. Westinghouse Savannah River Company, " Lead Expansion Anchor Load Capacity in Reactor Buildings at the Savannah River Site," August 15,1989.
- 23. URS/J.A. Blume & Associates, " Shaking Table-Testing for Seismic Evaluation of Electrical Raceway Systems, April 1983 24.
F. Elsabee, S. Anagnostis, and W. Djordjevic, " Seismic Evaluation of Electrical Raceway Systems," proceedings, ASME Boiler and Pressure Vessel Conference, ASME Paper 83-PVP-18, Portland, Oregon, Juno, 1983
- 25. URS/ John A. Blume & Associates " Analytical Techniques, Models, and Seismic Evaluation of Electrical Raceway Systems", August 1983.
- 26. URS/ John A. Blume & Associates," Seismic Investigations of Electric Raceways at the SEP Plants", June 1984.
j 9-2
"'"^ iud $$
- 10. Appendix A: Safe Shutdown Equipment List (SSEL)
- 11. Appendix B: Seismic Design Basis Spectra
- 12. Appendix C: Walkdown Personnel Resumes
- 13. Appendix D: Screening Verification Data Sheets (SVDS)
)
- 14. Appendix E: Peer Review Assessment i
- 15. Appendix F: Anchor Bolt Tightness and Embedment Checks j
Quad Cities A46 FincJ Report June 19,1996 Appendix A Quad Cities Nuclear Station Safe Shutdown Equipment List
thed cm eDasr Ias, rt Jose 19,1996 UNE_NO SYSTEM lEQUtP_ID
' EQUIP _DESC l LOC _ CODE l BUILDING, ROOM _ ELE EVAL _REQD 18037 120/208VAC JPNL 19-1-1 PNL, MCC 19-1-1 DIST PNL lN-18 lRB 18128 120/208VAC 18-1 A-1 PNL,120/208VAC PNL IN MCC 18-1 A-1 lN-16
'RB
~j623 S
623 S
28025 120/208VAC
- PNL 29-1-1 PNL, MCC 29-1-1 DIST
'N-9 RB 623 S
18205 120/240VAC j901-63 PS, UNINTERRUPTIBLE 120-240VAC AUX ELEC RM SB 595 SR 28204 120/240VAC W-63 PS, UNINTERRUPTABLE 120/240VAC F25-UPS RM TB j639 S
18400 125VDC 125VDC BATT #1 BATT, #1 G-24 TB j628 S
18401 125VDC 125VDC CHGR #1 CHGR, BATT #1 G-24 TB
$15 SR 18403 125VDC 125VDC CHGR #1 A CHGR, BATT #1 A G-24 TB l624 SR 18405 125VDC 125VDC BATT BUS #1 BUS, BATT #1 G-24 TB j615 S
18407 125VDC 125VDC, TB MN BUS #1 A BUS. TB MN #1 A H-24 TB j615 S
18410 125VDC 125VDC RB DIST PNL #1 PNL 125VDC RB DIST N-15 RB j623 S
18416 125VDC 125VDC, TB MN BUS 1 A-1 BUS, TB MN 1 A-1 H-24 lTB j615 S
18419 125VDC 125VDC, TB RES BUS 1B BUS. TB RES 1B H-24 TB
[615 S
18421 125VDC 125VDC TB RES BUS 1B-1 BUS, TB RES 18-1 H-24 TB j615 S
28400 125VDC 125VDC BATT #2 BATT,#2 G-2 TB j628 S
28401 125VDC 125VDC CHGR #2 CHGR, #2 BATT G-1 TB i615 SR 28403 125VDC 125VDC CHGR #2A CHGR, #2A BATT G-1 TB 615 SR 28405 125VDC 125VDC BATT BUS #2
, BUS, #2 BATT G-1 TB 615 S
28407 125VDC 125VDC TB MN BUS 2A
'615 S
28409 125VDC 125VDC RB DIST PNL #2 PNL, RX BLG DIST PNL #2 M-9 RB 623 S
28415 125VDC 125VDC TB MN BUS 2A-1 BUS, TB MN 2A-1 H-1 TB 615 S
28418 125VDC 125VDC TB RES BUS 2B BUS. TB RES 2B H-1 TB j615 S
28420 125VDC 125VDC TB BUS 28-1 BUS, TB RESERV 28-1 H-1 TB j615 S
18034 208VAC 19-1-1 MCC,208VAC RB 19-1-1 N-18 RB l623 S
18125 208VAC MCC 18-1 A-1 MCC,208VAC RB #18-1 A-1 N-16 jRB l623 S
28023 208VAC 29-1-1 MCC,208V RB 29-1-1 N-9 IRB j623 S
28125 208VAC 28-1 A-1 MCC,28-1 A-1 N-8 RB
'623 S
18209 250VDC 250VDC BATT #1 BATT UNIT 1250VDC G-24 TB 628 S
18300 250VDC 250VDC CHGR #1 BATT CHGR, #1 G-24 TB 615 SR 18302 250VDC 250VDC CHGR #1/2 BATT CHGR,1/2 G-24 TB 615 SR 18305 250VDC 250VDC MCC 1 MCC, TB 250VDC #1 G-24 TB 615 S
18308 250VDC 250VDC MCC 1 A MCC, RX BLG 250VDC #1 A L-19 RB 623 S
28300 250VDC 250VDC BATT #2 BATT, UN!T 2 250VDC G-1 TB 628 S
28301 250VDC 250VDC CHGR #2 BATT CHGR, #2 G-1 TB 615 SR 28305 250VDC 250VDC MCC 2 MCC, TB 250VDC #2 G-2 TB 615 S
28308 250VDC 250VDC MCC 2A MCC, RX BLG 250VDC #2A L-11 RB 623 S
i 18305A 250VDC 250VDC MCC 1B lMCC, RX BLG 250VDC #1B N-18 RB l623 S
28305A 250VDC 250VDC MCC 2B iMCC, RX BLDG 250VDC #2B M-10 RB
' 23 S
6 18024 4160/480VAC TR 19 TR,4160VAC/480VAC SWGR 19 SERVICE H-15 RB 647 S
18108 4160/480VAC TR 18 TR,4160/480VAC SWGR 18 SERVICE H-13
,RB 647 S
28013 4160/480VAC TR 29 TR,4160/480VAC SWGR 29 SERVICE E-13
'RB 647 S
28110 4160/480VAC TR #28 TR,4160/480V SUPLY TO SWGR 28 H-11 RB 647 S
18003 4160VAC 14-1 iSWGR,14-1 H-16 RB 639 S
lSWGR,14
{RB 18006 4160VAC 14 G-22 B
615 S
18101 4160VAC 13-1 ISWGR,13-1 H-14 647 S
Page A-1
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m th.scame A4erealRepet June 19,1998 UNE_NO SvSTEM
! EQUIP _ID iEQUIP_DESC LOC _ CODE
~ BUILDING ROOM _ ELE EVAL _REQD 11004 CRD lAO 1-0302-21D VLV, SOUTH BANK SDV VNT L-12 RB 595 +20 S
11005 CRD
'A010302-22C VLV, SOUTH BNK SDV DRN H-14 RB 554 +35 S
i 11006 CRD
'A01-0302-22D VLV, SOUTH BANK SDV DRN J-18 lRB j554 +35 S
11007 CRD A01-0302-21 A VLV, NORTH BANK SDV VNT L-13 lRB 1595 +20 S
11008 CRD A01-0302-218 VLV, NORTH BANK SDV VNT L-13 lRB 595 +20 S
11009 CRD AD 1-0302-22B VLV, NORTH BANK SDV DRN H-14
- RB 554 +35 S
11010 CRD A01-0302-22A VLV, NORTH BANK SDV DRN H-14 RB 554 +35 S
11011 CRD S0-1 4302-198 VLV, B/U SCRAM L-17 RB 595 +04 SR 3
11012 CRD SO 1-0302-19A VLV, B/U SCRAM L-17
'RB 595 +05 SR 11015 CRD SO 1-0302-20A VLV, SDV VNT & DRN L-17 RB 595 +5 SR 11016 CRD SO 1-0302-208 VLV, SDV VNT & DRN L-17 RB d595 +3 SR 21001 CRD UNIT 2 HCU HYDRAULIC CONTROL UNIT (177 TOTAL)
RB 595 S
21004 CRD A0-2-0302-21 C VLV, SOUTH BNK SDV VNT L-12 RB 595-20 S
21006 CRD A0-2-302-21 D VLV, SOUTH BNK SOV VNT L-12 RB 595 +20 S
21007 CRD AO-2-0302-22C VLV, SOUTH BNK SDV DRN J-12 RB 554 +35 S
21009 CRD A0-24302-22D VLV, SOUTH BNK SDV DRN J-12 RB "554 +35 S
21012 CRD A0-2-0302-21 A VLV, NORTH BNK SDV VNT L-7 RB 595 +20 S
i 21014 CRD A0-2-0302-21 B VLV, NORTH BNK SDV VNT L-7 lRB 595 +20 S
21015 CRD A0-2-0302-22A VLV, NORTH BNK SDV DRN J-7
'RB 554 +35 S
21017 CRD A0-2-0302-228 VLV, NORTH BNK SDV DRN J-7 RB 554 +35 S
21019 CRD
'S O-2-0302-208 SCRAM DUMP VLV L-8 RB 595 +3 SR 21020 CRD SO-2-0302-20A SCRAM DUMP VLV L-8 RB 595 +3 SR 21023 CRD SO-2-0302-19A BACK-UP SCRAM VLV L-8 RB 595 +5 SR 21025 CRD SO-0302-198 BACK-UP SCRAM VLV L-8 RB 595 +4 SR 12029 CS A01-1402-9A CK VLV, CS LOOP A INJ J-16
,RB fi40 +03 S
12030 CS A01-1402-98 CK VLV, CS LOOP B INJ K-16 RB j640 +03 S
I 22040 CS AO-2-1402-9A TESTABLE CK VLV, LOOP A INJ AZ-345 RB
$43 S
22041 CS A0-2-1402-9B TESTABLE CK VLV, LOOP B iNJ AZ-195 RB l643 S
8000 DG 1/2-6601 DG, UNIT 1/2 N-13 1/2 DG RM '595 SR 18000 DG 1-6601 DG, UNIT 1 G-24 TB 595 SR 28000 DG 2-6601 DG, UNIT 2 G-2 TB 595 SR 5201 DG AIR 1/2-4609A DG STARTING AIR COMP 1/2A N-13 1/2 DG RM 595 SR 5203 DG AIR 1/2-4600A DG A!R RECEIVER TANK 1/2 A N-13 1/2 DG RM 595 S
3 5205 DG AIR RV-1/2-4699-306A jREUEF VLV, DG AIR RECEIVER TNK 1/2 A N-13 1/2 DG RM 595 S
5206 DG AIR 1/2-46008 lDG AIR RECEIVER TNK 1/2 B N-13 1/2 DG RM 595 S
5208 DG AIR RV-1/2-4699-3063
'REUEF VLV, DG AIR RECEIVER TNK 1/2 B N-13 1,2 DG RM 595 S
5211 DG AIR 1/2-4609B DG STARTING AIR COMP 1/2 B N-13 1/2 DG RM 595 SR 5213 DG AIR 1/2-4600C DG AIR RECEIVER TNK 1/2 C N-13 1/2 DG RM 595 S
5215 DG AIR RV-1/2-4699-306C REUEF VLV, DG AIR RECEIVER TNK 1/2 C N-13 1/2 DG RM 595 S
5216 DG AIR 1/2 e600D
,DG AIR RECEIVER TNK 1/2 D N-13 1/2 DG RM 595 S
5218 DG AIR
.RV-1/2-4699-3060 lREUEF VLV, DG AIR RECEIVER TNK 1/2 D N-13 1/2 DG RM 595 S
5221 DG AIR 1/2-4604
%IR DRYER
i 5224 lDG AIR 1/2-4605 l BLOWDOWN TNK lN-13 1/2 DG RM 595 S
5226
[DGAIR PCV-1/2-4699-224 l PRESS CONTROL VLV, DG STARTING AIR lN-13 1/2 DG RM 595 S
5230 lDG AIR S O-1/2-4699-310 l STARTING SOLENOID VLV
!N-13 1/2 DG RM 595 SR Page A-3
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thes Otus A 48 Feel Report has 19,1998 LINE,,_NO jSYSTEM
- EQUIP _ID l EQUIP _DESC l LOC _ CODE
5244 lDG CLG 1/2-6665
'ENG COOLING PIPE MANIFOLD N-13 1/2 DG RM 595 S
15130 lDG CLG 1-6661/A HT EXCH, DG CLG WTR G-23 TB 595 +05 S
15131 lDG CLG 1-6661 B HT EXCH, DG CLG WTR G-23 TB 595 +05 S
15132
'DG CLG 1-6663 VLV, TEMP REG THEMOSTATIC G-23 TB 595 +07 S
15133 DG CLG 1-6665
,MANIFLD, ENG CLG PIPE G-23 TB
[595 +06
,S iHT EXCH, LUBE OIL G-23 TB l595 lS 15134 DG CLG 1-6654 15135 lDG CLG 1-6662 TANK, EXPANSION G-23 TB 595 +07 iS 15136 lDG CLG 1-6664 HTR, IMMERSION G-23 TB 595 +03 S
15138 iDG CLG 1-666A PMP, ENG DRIVEN COOLANT G-23 TB 595 S
15139
'DG CLG 1-6668
'PMP, ENG DRIVEN COOLANT G-23 TB 595 S
25137 DG CLG 2-6663 VALVE, TEMP REGULATING l#2 DG RM TB 595 S
25138 DG CLG 2-6661 A HT EXCH, DG 2 COOLING WATER j#2 DG RM TB 595 S
25139 DG CLG 2-6661 B HT EXCH DG 2 COOLING WATER l#2 DG RM TB 595 S
25140 lDG CLG 24654 LUBE OIL COOLER
lDG CLG 2-6662 EXPANSION TANK
25141 25142 lDG CLG 2-6664 lMMERSION HEATER
i 25144 lDG CLG 24665
' ENGINE COOLING PIPE MANIFOLD
5246 lDGCW 1/2-3903 PMP, DG COOLING WTR D-21 TB 547 SR 5249 IDG CW 1/2-5749 COOLER, DGCW PMP CUBICLE D-21 TB 547 S
15141 lDG CW 1-3903 PMP, DGCW C-21 TB
- 547 +02 SR 15145
!DG CW 1-5749 COOLER, DG CLG PMP CUB C-22 TB l552 +00 S
15155 lDG CW 1-5746 AIR HANDLER, N CS EQU RM EMER CLR G-14 RB l560 S
15158 lDG CW 1-5748 l%IR HANDLER, S CS EQU RM EMER CLR G-18 TB l560 S
15164 IDG CW 1-5747
' AIR HANDLER, HPCI RM EMER CLR
[G-12 TB j564 S
15167 DG CW 1-5746A AIR HANDLER, N RHR RM EMER CLR jN-14 RB 1580 S
25146 DG CW 2-3903 PUMP, DG 2 COOLING WATER lC-5 TB 547 SR 3
25149 DG CW 2-5749 COOLER, DGCW PUMP 2 CUBICLE C-5 TB 547 S
25162 lDG CW 2-57488 COOLER, SOUTH CS ROOM G-12 RB 554 S
25165 lDG CW 2-5748A COOLER, NORTH CS ROOM G-8 RB 554 S
25168 lDG CW 2-57468 COOLER, SOUTH RHR HX ROOM M-13 RB 554 S
25171 lDG CW 2-5746A COOLER, NORTH RHR HX ROOM M-8 RB 554 S
25177 lDG CW 2-5747 COOLER, HPCI ROOM G-12 RB 554 S
15159B lDG CW 1-5748B CLR, S RHR RM M 13 RB 560 S
5289 lDG EXHAUST 1/2-6667 EXHAUST SILENCER N-13 1/2 DG RM 595 S
15201 lDG EXHAUST 1-6667 SILENCER, EXHAUST G-24 TB 639 +04 S
25216
!DG EXHAUST 2-6667 EXH SILENCER G-24 TB 639 S
5260 DG FO 1/2-5201 TNK, DG FO STORAGE OUTSIDE UNDERGRFS 5262 DGFO 1/2-5203 PMP, DG FO TRANS
,N-13 1/2 DG RM 595 SR 5264 iDG FO RV-1/2-5201 RELIEF VLV, DG FO TRANS PMP DISCH
5268 lDG FO SO-1/2-5201 SOV, DG FO DAYTNK FILL N-13 1/2 DG RM 595 SR 5270
'DG FO 1/2-5202 DAYTNK, U 1/2 DG FO i -13 1/2 DG RM j595 S
N 5275 DGFO 1/2-5208 PMP, FUEL PRIME lN-13 1/2 DG RM l595 SR 5279 DG FO 1/2-5206 DUPLEX FUEL STRAINER jN-13 1/2 DG RM I595 S
15172 DGFO 1-5201 TANK, DG FO STORAGE l
OUTSIDE IUNDERGRt S Page A-5
thed Cihme A46 Feel Report Jhes 19, He8 LINE_NO l SYSTEM
' EQUIP _10 l EQUIP _DESC LOC _ CODE BUILDING l ROOM _ ELE l EVAL _REQD 15174
'SR 15180 DGFO SO 1-5201 SOV, DG FO DAY TNK FILL G-23 TB l595 +19 SR
' 95 +11 S
15182 DG FO 1-5202 TNK, DG FO DAY G-23 TB 5
15188 DG FO 1-5208 PMP, FUEL PRIME G-23 TB 595 +04 SR 15190 DG FO 1-5209 PMP, ENG DRVN FUEL G-23 TB 595 S
15192 DGFO 1-5206 FILTER, DUPLEX FUEL G-23 TB 595 +06 S
25182 DG FO 2-5201 TANK, DG FO STORAGE OUTSIDE UNDERGRPS 25184 DGFO 2-5203 PMP, DG FO TRANS
25193 DGFO SO-2-5201 lVLV, DG FO DAY TANK FILL
252C2 DGFO 2-5208 PMP, FUEL PRIME
5101 DG HVAC 1/2-5772-87 DAMPER, OUTSIDE AIR INTAKE N-13 1/2 DG RM 595 S
5102 DG HVAC 1/2-5772-68 DAMPER, DG VENT FAN INTAKE MODULATING
5103 DG MVAC 1/2-5727 FAN, DG RM VENTILATION IN-13 1/2 DG RM l595 SR 5104 DG HVAC 1/2-5772-86 DAMPER, DG RM EXHAUST N-13 1/2 DG RM j595 S
'N/A REGULATOR, DAMPER lA SUPPLY N-13 If2 DG RM $95 S
5108 DG HVAC EP-1 (SO 1/2-5799-549)
'N-13 1/2 DG RM 595 SR 5109 DG HVAC N/A INITROGEN BOTTLE, DAMPER PNEUM. BACKUP N-13 1/2 DG RM l595 S
5110 DG HVAC jN/A NITROGEN BOTTLE, DAMPER PNEUM BACKUP N-13 1/2 DG RM j595 S
I 5113 DG HVAC lR-1 REGULATOR, DAMPER PNEUM. SUPLY N-13 1/2 DG RM iS95 S
5114 DG HVAC lR-2 REGULATOR, DAMPER PNEUM. SUPLY N-13 1/2 DG RM 595 S
SOV, INLET & OUTLET ISOL DAMPER CNTRL N-13 1/2 DG RM 595 SR 15301 DG HVAC 1-9472-035 DAMPER, OUTSIDE AIR INTAKE MODULATION G-23 TB 595 S
15304 DG HVAC 1-5772-87 DAMPER, DG VENT FAN INTAF'i G-23 TB 595 S
15305 DG HVAC SV-1 (EP-2/SO 1-5799-553) SOV, DAMPER CONTROL G-23 TB 595 SR 15307 DG HVAC 1-5727 FAN, DG ROOM VENTILATION
15309 DG HVAC EP-3 (SO 1-5772-89)
SOV, DAMPER CONTROL G-23 TB 595 SR 15310 DG HVAC 1-9472-40 DAMPER, DG ROOM EXHAUST G-23 TB 595 S
15311 DG HVAC 1-9472-41 DAMPER, DG ROOM EXHAUST G-23 TB 595 S
25301 DG HVAC 2-9472-035 DAMPER, OUTSIDE AIR INTAKE MODULATION
25304 DG HVAC 2-5772-87 DAMPER, DG VENT FAN INTAKE
25305 DG HVAC SV-1 (EP-2/SO2-5799-553) SOV, DAMPER CONTROL
!#2 DG RM TB 595 SR 25308 DG HVAC 2-9472 432 DAMPER, NORMAL DG RM VENT.
25309 DG HVAC EP-3 (SO 2-5772-89)
25311 DG HVAC 2-9472-41 DAMPER, DG ROOM EXHAUST
5288 DG INTAKE 1/2-6668 FILTER, DG INTAKE AIR N-13 1/2 DG RM j595 S
15200 DG INTAKE 1-6668 FILTER, A INTAKE G-23 RB j615 +01 S
25215 DG INTAKE 2-6668 FILTER, DG INTAKE AIR G-24 TB
'639 S
. FILTER, LUBE OIL N-13 If2 DG RM 595 S
Page A-6
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bd ciam 646 reef heart June 19.1998 LINE_NO iSYSTEM
- EQUIP _ID iEQU!P_DESC
' LOC _ CODE iBUILDING jROOM_ ELE: EVAL _REQD 22012
- MS A0-2-0203-1 B lMSIV, LN B INBD AZ-10 lRB l593 S
AO-2-0203-2B G-9 lRB l591 +4 S
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22020 iMS lAO-2-0203-1C MSIV, LN C INBD AZ-350 RB 593 S
22021 lMS lAO-2-0203-2C MSIV, LN C OUTBD G-10 RB 591 +4 S
22022 jMS lRV-2-0203-4D SAFETY VLV, LN D AZ-310 RB t620 S
22023 MS
'RV-2-0203-4H SAFETY VLV, LN D AZ-320 RB 620 S
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12000A iMS
'RV 1-2003-4A PRV, A SAFETY VLV J-15 RB 614 +06 S
12005A
!MS SO 1-0203-1 A-1 SOV, MSIV 1-0203-1 A H-15 RB 591 SR fMS ISO 1-0203-1 A-2 SOV, MSIV 1-0203-1 A H-15 RB 591 SR 120058 12006A lMS
'SO 1-0203-2A-1 SOV, MstV 10203-2A G-15 RB 591 SR 12006B MS SO 1-0203-2A-2 SOV, MStV 1-0203-2A G-15 RB 591 SR 12014A MS SO 1-0203-1B-1 SOV, MStV 1-0203-1B H-15 RB 591 SR 120148 MS SO 1-0203-1B-2 SOV, MSIV 1-0203-1B H-15 RB 591 SR 3
12015A MS S O 1-0203-28-1 SOV, MSIV 1-0203-2B G-15 RB 591 SR 4
MS SO 1-0203-2B-2 SOV, MStV 1-0203-28 G-15 RB 591 SR 12015B i
12021A iMS SO 1-0203-1C-1 SOV, MStV 1-0203-1C H-15 RB 591 SR 12021B iMS SO 1-0203-1C-2 SOV, MStV 1-0203-1C H-15 RB 591 SR 12022A lMS SO 1-0203-2C-1 SOV, MSIV 1-0203-2C G-15 RB 591 SR SO 10203-2C-2 SOV, MSIV 1-0203-2C G-15 RB 591 SR 12022B iMS g
12027A MS SO 1-0203-1 D-1 SOV, MSIV 1-0203-1D H-15 RB 591 SR i
12027B lMS SO 1-0203-1 D-2 SOV, MSIV 10203-1D H-15 RB 591 SR 12028A lMS SO 1-0203-2D-1 SOV, MSIV 1-0203-2D G-15 RB 591 SR 12028B
'MS SO 1-0203-2D-2 SOV, MSIV 1-0203-2D G-15 RB 591 SR 22005A MS SO 2-0203-1 A-1 SOV MStV 2-0203-1 A AZ-5 RB 591 SR 22005B MS SO 2-0203-1 A-2 SOV, MSIV 2-0203-1 A AZ-5 RB 1591 SR i
22006A
[MS SO 2-0203-2A-1 SOV, MSIV 2-0203-2A G-9 RB j591 SR 22006B jMS SO 2-0203-2A-2 SOV, MSIV 2-0203-2A G-9 RB 591 SR 22012A MS SO 2-0203-1B-1 SOV, MSIV 2-0203-1B AZ-10 RB 591 SR 220128 MS SO 2-0203-1 B-2 SOV MSIV2-0203-1B AZ-10 RB 591 SR 22013A MS SO 2-0203-2D-1 SOV, MSIV 2-0203-2B G-9 RB 591 SR 22013B MS SO 2-0203-2B-2 SOV, MSIV 2-0203-2B G-9 RB 591 SR 22020A MS SO 2-0203-1C-1 SOV, MSIV 2-0203-1C AZ-350 RB 591 SR i
22020B jMS SO 2-0203-1 C-2 SOV, MSIV 2-0203-1C AZ-350 RB 591 SR 22021A jMS SO 2-0203-2C-1 SOV, MSIV 2-0203-2C G-9 RB 591 SR 22021B lMS SO 2-0203-2C-2 SOV, MSIV 2-0203-2C G-9 RB 591 SR 22027A lMS SO 2-0203-1D-1 JSOV, MStV 2-0203-1D AZ-355 RB 591 SR 22027B
- MS SO 2-0203-1D-2 lSOV, MSIV 2-0203-1 D AZ-355 RB 591 SR 22028A lMS SO 2-0203-2D-1 SOV, MSIV 2-0203-2D G-9 RB 591
,SR 220288
!MS SO 2-0203-2D-2 SOV, MSIV 2-0203-2D G-9 RB 591 iSR Page A-8
thed Ohne A 46 Feel Report June 19,1996 LINE_NO iSYSTEM jEQUIP_ID
- EQUIP _DESC l LOC _ CODE
' BUILDING jROOM_ ELE: EVAL _REQD 9000 l PROCESS INST
{912-8 lPNL,912-8
'S 15502
{ PROCESS INST jlT 1-1641-5A LT, TORUS WIDE RNGE LVL, DIV I L/M 13/14 RB l585 S
15505
- PROCESS INST
- LT 1-1641-5B LT, TORUS WIDE RNGE LVL, DiV ll UM 18/19 RB 585 S
i -16 RB 554 S
15508 l PROCESS INST TE 1-1641-200 TE, TORUS WATER DIV 1 K
15509 l PROCESS INST TE 1-1641-201 TE, TORUS WATER DIV I
'K-16 RB 554 S
i 15510
' PROCESS INST TE 1-1641-202 TE, TORUS WATER DIV I K-16 RB 554 lS 15511 PROCESS INST TE 1-1641-203 TE, TORUS WATER DIV I K-16 RB 554 lS 15512 PROCESS INST TE 1-1641-204 TE, TORUS WATER DIV I K-16 RB 554 lS 15513 PROCESS INST T E 1-1641-205 TE, TORUS WATER DiV I
- K-16
'RB 554 S
15514 PROCESS INST TE 1-1641-206 TE, TORUS WATER DIV I
!K-16 RB 554 S
15515 PROCESS INST TE 1-1641-207 TE, TORUS WATER DIV I lK-16 RB 554 S
15516 PROCESS INST TE 1-1641-208 TE, TORUS WATER DIV 11 lK-16 JRB
'554 S
554 S
15517 iPROCESS INST TE 1-1641-209 TE, TORUS WATER DIV 11 lK-16
15519 PROCESS INST TE 1-1641-211 TE, TORUS WATER DIV 11 lK-16 lRB 554 S
15520 jPROCESS INST TE 1-1641-212 TE, TORUS WATER DIV 11 lK-16
[RB 554 S
15521 l PROCESS INST TE 1-1641-213 TE TORUS WATER DIV 11
!K-16 iRB 554 S
15522 l PROCESS INST TE 1-1641-214 TE, TORUS WATER DIV li lK-16 jRB 554 S
15523 iPROCESS INST TE 1-1641-215 TE, TORUS WATER DIV ll lK-16 lRB 554 S
15542 l PROCESS INST
'PT 1-1641-6A PT, DW DIV I jM-15 lRB 623 S
15572 jPROCESS INST TE 1-1046A TE, HT EXCH 1 A INLET
[M/N 13/14 lRB 554 S
i 15573 l PROCESS INST TE 1-1046B TE, HT EXCH 1B INLET lM/N 18/19 RB 554 S
15574 l PROCESS INST TE 1-1047A TE, HT EXCH 1 A OUTLET lM/N 13/14 RB 554 S
15575 jPROCESS INST TE 1-1047B TE, HT EXCH 1B OUTLET l MIN 18/19 RB 554 S
901-27 jRPIS l AUX ELEC RM SB 595 SR 19017
[ PROCESS INST 1
19018 l PROCESS INST l901-3 lMN CONT PNL 901-3 lMN CNTRL RM SB 623 S
19019 (PROCESS INST 501-2
'MN CONT PNL 901-2
19020 l PROCESS INST IR 2201-70B INST RACK 2201-70B G/H 25/26 RB 595 S
19021 jPROCESS INST 901-36 PNL,901-36 MN CNTRL RM SB 623 S
19022 PROCESS INST tR 2201-5 INSRT RACK 2201-5 K-14 RB 623 S
19023 PROCESS INST IR 2201-6 INSRT RACK 2201-6 L-16 RB 623 S
19024 PROCESS INST 901-5 MN CONT PNL,901-5 jMN CNTRL RM SB 623 S
3 19025 jPROCESS INST IR 2201-7
'INSRT RACK 2201-7 p-14 RB 595 S
19026 l PROCESS INST IR 2201-8 INSRT RACK 2201-8 K-17 RB
-595 S
19027 l PROCESS INST IR 2201-73A INSRT RACK 2201-73A CABLE SPRD RM;SB 609 S
19028 jPROCESS INST lR 2201-73B INSRT RACK 2201-73B CABLE SPRD RM SB 609 S
i 19029 l PROCESS INST jlR 2201-59A INSRT RACK, RHR SVS M-14 RB 554 S
!IR 2201-598 (NSRT RACK, RHR SVS M-18 RB 554 S
19030 l PROCESS INST 25508 l PROCESS INST TE-2-1641-200 TEMP ELEM, TORUS WTR DIV I K-10 RB 554 S
25509 jPROCESS INST TE-2-1641-201 TEMP ELEM, TORUS WTR DIV I K-10 RB j554 S
25510
! PROCESS INST TE-2-1641-202 TEMP ELEM, TORUS WTR DIV I K-10 RB
'554 S
g 25511 l PROCESS INST TE-2-1641-203 TEMP ELEM, TORUS WTR DIV l K-10 RB 554 S
25512 l PROCESS INST TE-2-1641-204 TEMP ELEM, TORUS WTR DIV 1 K-10 iRB 554 S
25513 l PROCESS INST TE-2-1641-205 TEMP ELEM, TORUS WTR DIV I K-10
!RB 554 S
25514 l PROCESS INST T E-2-1641-206 TEMP ELEM. TORUS WTR DIV I
_K-10 lRB
- 554 S
Page A-9
Osed Ones A 48 Feel Report June 19J996
.Ll_NE_NO iSYSTEM
' EQUIP _ID
- EQUIP _DESC jLOC_ CODE
' BUILDING 1 ROOM _ ELE. EVAL _REQD 25515 l PROCESS INST TE-2-1641-207
- TEMP ELEM, TORUS WTR DIV I
'K-10 RB l554 lS 25516 l PROCESS INST TE-2-1641-208 TEMP ELEM, TORUS WTR DIV 11 K-10 RB 554 S
25517
[ PROCESS INST TE-2-1641-209 TEMP ELEM, TORUS WTR DIV ll K-10 RB 554 S
25518 jPROCESS INST TE-2-1641-210 TEMP ELEM, TORUS WTR DtV ll
,K-10 RB 554 S
25519 l PROCESS INST TE-2-1641-211 TEMP ELEM, TORUS WTR DIV 11 lK-10 iRB 554 S
25520 l PROCESS INST TE-2-1641-212 TEMP ELEM, TORUS WTR DIV 11 lK-10
'RB 554 S
25521 PROCESS INST TE-2-1641-213 TEMP ELEM, TORUS WTR DIV 11 lK-10 RB iS54 S
i 25522 PROCESS INST TE-2-1641-214 TEMP ELEM, TORUS WTR DIV 11 lK-10 RB j554 S
25523 PROCESS INST TE-2-1641-215 TEMP ELEM, TORUS WTR DtV tl lK-10 RB 554 S
25568
! PROCESS INST TE-2-1046-A TEMP ELEM, RHR HX 2A INLET jM/N 7/8 RB 554 S
26569 PROCESS INST TE-2-1046-B TEMP ELEM, RHR HX 2B INLET iM!N 12/13 RB G4 S
25570 PROCESS INST TE-2-1047-A TEMP ELEM, RHR HX 2A OUTLE1 lM-9 RB 623 S
25571 PROCESS INST TE-2-1047-B TEMP ELEM, RHR HX 28 OUTLET lM-11 RB 623 S
29018 PROCESS INST l902-3 iMN CNTRL PNL 9024 jMN CNTRL RM SB 623 S
29019 PROCESS INST l902-2 MN CNTRL PNL 902-2 lMN CNTRL RM SB 623 S
3 29020 l PROCESS INST llR2202-70A INSTR RACK 2202-70A lE/F 25/25 Is3 595 S
29021 l PROCESS INST jlR2202-708 INSTR RACK lE/F 25/25 RB 595 S
29022
! PROCESS INST j902-36 PNL 902-36
i 29023 PROCESS INST llR2202-5 llNSRT RACK J-11 RB 623 +2 S
29024 PROCESS INST llR2202-6 llNSRT RACK L-11 RB 623 +4 S
i 29025
' PROCESS INST l902-5 tMN CNTRL RM PNL 902-5
29026 PROCESS INST llR2202-7 INSRT RACK J-8 RB S95 S
i 29027 PROCESS INST lR2202-8 INSRT RACK K-11 tRB
[595 S
i i
29028
j 29029 PROCESS INST IR2202-73B INSRT RACK CABLE SPRD RM,SB 609 S
29030 PROCESS INST tR2202-59A INSRT RACK, RHR SYS
.M-7 RB 554 S
29031
15523A l PROCESS INST IR-2201-70B INST RACK 2201-70B G'H 25/26 RB 595 S
9001 l RACK / PANEL 2212-46 CAB,1/2 DG EXC N-13 1/2 DG RM 595 S
9005 l RACK / PANEL 2212-123 PNL,4160 VAC SYS N-13
,RB 595 S
f006
' RACK / PANEL 2212-125 PNL,1/2 DG HVAC N-13
'RB 595 S
9007 RACK / PANEL 2212-127 PNL,1/2 DG HVAC N-13 RB 595 S
1 9008 RACK / PANEL 2212-45 lPNL,1/2 DG RELAY N-13 1/2 DG RM 595 S
9009 jRACK/ PANEL 2212-50 lPNL,1/2 DG HVAC N-13 1/2 DG RM 595 S
19000 RACK / PANEL 2201-32 lPNL, AUTO BLOWDOWN
,K-14 RB 623 S
19001 RACK / PANEL 2251-12 l CAB, DG #1 EXC jG-23 TB 595 S
19002 RACK / PANEL 2201-75 lPNL, HVAC M-18 RB 554 S
19003
t 19004 jRACK/ PANEL 2251-112 lPNL,4160VAC SYS E-26 TB 595 S
19005
! RACK / PANEL 2251-113 jPNL, DG #1 G-23 TB 595 S
19006 RACK / PANEL 2251-86 lPNL,5160VAC SYS H-14 TB 639 S
19007 RACK / PANEL
'2251-87 lPNL,4160VAC SYS H-16 TB 623 S
19003 RACK / PANEL 2251-97 lPNL, DG #1 AUX FD TRANS G-23 TB 595 S
19009
! RACK / PANEL 42251-98 lPNL, DG CLG WTR PMP FD G-23 TB 595 S
19010 l RACK / PANEL i901-32 lPNL, CONTROL AUX ELEC RM SB 595
.S Page A-10
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Good Ciese A 46 Feed Report June 19,1996 LINE_NO l SYSTEM
' EQUIP _ID
- EQUIP _DESC l LOC _ CODE iBUILDING ROOM _ ELE' EVAL _REQD 24050 lRHR 2-1002C PMP 2C
'STH RHR RM lRB 554 SR 24051 lRHR 2-1006C CYC SEP, PMP 2C STH RHR RM lRB
,554 S
24052 lRHR 12-1002C SEAL COOLER, PMP 2C STH RHR RM lRB l554 S
24059 lRHR lMO-2-1001-18B
< MIN FLO VLV, PMP 2C L-11 lRB l554 SR 24065
'RHR
'MO-2-1001-168 BYP VLV, HT EXCH 2B M-12 lRB l554 SR 24067 RHR 2-10038 HT EXCH 2B STH RHR RM
'R B
}554 S
j554 24068 RHR RV-2-1001-166B RELIEF VLV, HT EXCH 28, RHR SIDE M-12 RB S
24070 RHR RV-2-1001-22B RELIEF VLV, LOOP B L-12 RB 554 S
1 i
I 24074 RHR
- MO-2-1001-348 VLV, TORUS SUPLY L-12 RB 554 SR 24076 RHR IMO-2-1001-36B VLV, SUP POOL COOLING L-11 RB 554 SR 25014 RHR 2-1001-1458 HT EXCH, PMP 28 MTR COOLER NTH RilR RM RB 554 S
25015 RHR 2-1002B SEAL COOLER, PMP 2B NTH RHR RM RB 554 S
^
25040
25041
[RHR 2-1002D SEAL COOLER, PMP 2D STH RHR RM RB 554 S
5000 lRHRSW 1/P 1/2-5741-333 1/P TRANS FOR FCV-1/2-5741-333
,E-24 TB 615 S
5045 lRHRSW XCV-1/2-5741-319A VLV, RHRSW SUPLY TO CR HVAC E-24 TB l615 S
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15015 RHRSW 1-1001-145B
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15020 RHRSW RV 1-1001-165A PRV,1 A HT EXCH iM-13 RB 554 +38 S
3 i
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~MO-1-1001-5A VLV, HT EXCH DISCH 15027 iRHRSW 1-5745A COOLR CUBICLE C-18 RB 547 00 S
15033 lRHRSW 1-1001-65C PMP 65C C-19 TB 547 +03 SR 15039 iRHRSW 1-1002D HT EXCH, PMP 2D SEAL COOLER M-18 RB 554 +04 S
15041 lRHRSW 1-1001-145D HT EXCH, PMP 2D SEAL C,LR M-18 RB 554 +05 S
15044 lRHRSW RV 1-1001-1658 PRV,1B HT EXCH M-18 RB 580 +09 S
15048 lRHRSW MO-1-1001-58 VLV,1 A HT EXCH DISCH M-18 RB 554 +19 SR 15051 lRHRSW 1-5745C COOLR - CUBICLE C-20 TB 547 00 S
25005 lRHRSW 2-5745A RHRSW PMP CUB COOLER 2A C-6 TB 547 S
25007 lRHRSW l2-1001-65A PMP 2A C-5 TB 547 SR 25020
'RHRSW lRV-2-1001-165A RELIEF VLV, RHR HT EXCH 2A, RHRSW SIDE M-7 RB 554 S
3 I
25022 RHRSW
'MO-2-1001-5A VLV, RHR HT EXCH 2A DISCH M-7 RB 554 SR 25030 RHRSW 2-5745C RHRSW PMP CUB COOLER 2C C-7 TB 547 S
25032 RHRSW 2-1001-65C PMP 2C C-7 TB 547 SR i
25046 RHRSW
25048 RHRSW lMO-2-1001-58 VLV, RHR HT EXCH 2B DISCH M-12 RB 554 SR 12032 RWCU iMO-1-1201-2 VLV,INBD SUCT ISOL AZ-270 RB 614 SR 12033 RWCU lMO-1-1201-5 VLV, OTBD SUCT K-18 RB 623 SR 22031 RWCU iMO-2-1201-2 VLV,INBD SUCT ISOL K-9 RB 623 SR 5
!MO-2-1201-5 VLV, OUTBD SUCT ISOL K-8 RB 623 SR Page A-14
.. ~. -. -.. -
Quad Citirs A46 Find Report June 19,1996 l
l l
Appendix B l
Quad Cities Nuclear Station SQUG Seismic Design Basis Spectra l
l l
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Fig 1re B-17 Page # B-17
s i
Quad Citi s A-46 Fin:3 R port June 19,1996 4
1 i
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i Appendix C Quad Cities Nuclear Station l
Walkdown Personnel Resumes i
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i
'l KARL L. ADLON Comed - Principal Engineer Quad Cities Nuclear Station EXPERIENCE and RESPONSIBILITIES Mr. Adlon is a degreed mechanical engineer currently responsible for the implementation of the SQUG program at Quad Cities Nuclear Station. He has 20 years of experience in the analysis of nuclear power plant equipment. This experience includes:
equipment seismic qualification, procurement engineering including part classification, component and part procurement and commercial grade item dedication, and procedure writing a
Mr. Adlon's seismic qualification experience includes performing all facets of equipment seismic qualification, from writing the seismic requirements section of equipment specifications to technical review of and commenting on the acceptability of vendor seismic analysis and test reports and to performing finite element analyses for the calculation of total stress and deflections due to seismic, operating and other l
loadings.
Where applicable, he has reviewed vendor ASME Stress Reports, reviewing and performing Code Reconciliations, as required. He has participated in the initial design team meetings for modifications and plant changes, ensuring that seismic and Code requirements would be met. He has evaluated as-found field conditions for seismic acceptability, recommending upgrades as warranted. Mr. Adlon has provided FSAR input, such as load combinations and stress limits used, the methods used to qualify equipment and the seismic levels to which they were qualified.
Mr. Adlon has prepared Comed's Technical Information Documents for seismic qualification for station engineers' use. These included " Dynamic Qualification Requirements," " Qualification Using Analytical Methods," " Qualification Using Test Methods," and " Qualification Using a Combination of Test and Analysis."
Procurement engineering experience encompasses performing and reviewing evaluations to support the purchase of parts and equipment.
Procedure Writing experience includes preparing Technical Information Documents for seismic qualification, authoring a shelflife procedure for evaluation of items being purchased and for items in stores and writing and updating engineering procedures for equipment procurement.
EMPLOYMENT 1995 to Present: Comed 1976 to 1995:
Sargent & Lundy Engineers PROFESSIONAL American Nuclear Society and MEMBERSIIIPS American Society of Mechanical Engineers EDUCATION Degree:
Bachelor of Science in Mechanical Engineering, The University of Akron,1976
l l
STEPHEN ANAGNOSTIS l
EDUCATION:
B.S. - Civil Engineering, Columbia University School of Engineering,1974 M.S. - Structural Engineering, Massachusetts Institute of Technology,1976 PROFESSIONAL HISTORY:
(
Stevenson & Associates, Inc., Project Manager,1983 - present URS / John A. Blume & Associates, Engineers, Boston, Massachusetts, Project Engineer 1982 -
1 1983; Senior Engineer, 1980-1982 Charles Stark Draper Laboratory, Cambridge, Massachusetts, Technical Staff, 1976 - 1980; Draper Fellow, 1974 -1976 PROFESSIONAL EXPERIENCE:
Mr. Anagnostis has eighttsen years of experience, principally in the areas d seismic and dynamic l
cngineenng analysis. He joined Stevenson & Associates in February 1983 as a Project Manager in l
the Boston area office. He is currently the Project Manager, lead walkdown engineer and analyst for i
the resolution of USl A-46 and the seismic IPEEE for the James A. Fitzpatrick and Cooper Nuclear Station projects. He has over a thousand hours of seismic walkdowns at nuclear power facilities throughout the US and abroad implementing the SQUG GIP and EPRI NP-6041 screening methodologies. Mr. Anagnostis is the principal author of S&A's GIPPER and IPEXPERT, computer programs specifically designed to gather and analyze data for the implementation of the USI A-46 and l
seismic IPEEE programs respectively.
i Mr. Anagnostis was extensively involved in both analysis (frequency domain and time domain structural dynamics) and testing (in-situ modal and full-scale shaking-table) at URS/Blume's Boston office. He had lead technical responsibility for a two year program to develop a seismic evaluation criteria for electrical raceway systems at eight of the oldest United States nuclear power stations. This program included the design, supervision, and data analysis of shaking-table tests of full-scale raceway systems, cyclic / fatigue tests of raceway components, and the development of analytical evaluation techniques incorporating the test results.
As a member of the technical staff of Charles Stark Draper Laboratory, Mr. Anagnostis was involved l
in the assessment of space based surveillance (infra-red and radar) and defense systems for the l
Defense Advanced Research Projects Agency. He was a major author of a software simulation l
system to assess the capabilities of spaced based optical systems including structural vibrations, l
control dynamics, and optical performance.
i PROFESSIONAL GROUPS:
Committee, Working Group for the Analysis and Design of Electrical Cable Support Systems
4 1
Member, American Society of Civil Engineers Nuclear Structures and Materials PUBLICATIONS AND REPORTS:
"Vioration Engineering in the Semiconductor Industry," with W. Djordjevic and T. M, Tseng, Test and Measurement World, May,1984 "EDASP: Structural Modification Program," with W. Djordjevic and T. M. Tseng, Proceedings, Second International Modal Analysis Conference, Orlando, Florida (February 1984).
" Implementation of Software to Account for Equipment Modifications," with W. Djordjevic, C.
Gangone, R. Jenkins and A. Marion, Transactions, ANS 1983 Win'er Meeting, San Francisco, California (October 1983)
" Theory and Implementation of Analytical Tools to Calculate Response Changes in Equipment Previously Evaluated by Testing," with W, Djordjevic, Transsctions, 7th Intemational Conference on Structural Mechanict in Reactor Technology (August 1983)
" Seismic Ecluation of Electrical Raceway Systems," with W. Djordjevic and F. Elsabee,1983 ASME Pressure Vessel and Piping Conference, Portland, Oregon (June 1983) i
" Space Radar Large Aperture Simulation / Analysis," with F. Ayer, CSDL R-1413 (October 1980)
"Large Beam Expander Technology Design, Analysis and Simulation Development Program" (U),
with K. Soosaar, et al., CSDL R-1224 (Secret), (April 1979)
J "High Altitude Large Optics Integrated Simulations" (U). with K. Soosaar. et al.,
CSDL R-1286 (Secret), (July 1979) i
" Passive and Active Suppression of Vibration Response in Precision Structures"(U), with K. Soosaar et al., CSDL R-889 (Secret) (February 1978)
" Optimal Actuator Locations for Mirror Surface Control," M.S. Thesis Massachusetts Institute of Technology (May 1976)
)
1 l
WALTER DJORDJEVIC EDUCATION:
B.S. - Civil Engineering, University of Wisconsin at Madison,1974 M.S. - Structural Engineering, Massachusetts Institute of Technology,1976 I
REGISTRATION:
i State of Califomia, State of Wisconsin, Commonwealth of Massachusetts, State of Michigan PROFESSIONAL HISTORY:
Stevenson & Associates, Inc., President 1996 - present; Vice President and General Manager of the Boston area office, 1983-1995 URS/ John A. Blume & Associates, Engineers, Boston, Massachusetts, General Manager,1980 -
l 1983; San Francisco, California, Gupervisory Engineer, 1979-1980 l
Impell Corporation, San Francisco, Califomia, Senior Engineer, 1976 -1979 Stone & Webster Engineering Corporation, Boston, Massachusetts, Engineer, 1974 - 1976 PROFESSIONAL EXPERIENCE:
Mr. Djordjevic founded the Stevenson & Associates Boston area office in 1983 and serves as President arid General Manager.
He has performed over a thousand hours of onsite seismic walkdowns for using t!'e EPRI-SQUG methodology for resolution of the USI A-46 and seismic IPEEE issues. He is one of the most experienced seismic walkdown, screening and verification engineers having personally participated in seismic walkdowns at 26 U.S. nuclear units.
l In 1994 he performed seismic walkdowns and analysis of the Tooele Chemical Demilitarization Facility in support of a seismic quantitative risk assessment. Prior to the formulation of the current seismic screening criteria, Mr. Djordjevic performed seismic analyses at the eight SEP nuclear plants, and prototype seismic screening walkdowns on the Hanford Purex facility, and the Savannah River L and l
P reactors.
l l
Under contract to the SQUG, Mr. Djordjevic authored sections of the Generic implementation Procedure, now in broad use for seismic walkdown screening methodologies. Together with other S&A engineers, Mr. Djordjevic developed GENRS, a of the software product sponsoreri by the SQUG l
which establishes in-cabinet amplification factors for GIP relay evaluations.
Mr. Djordjevic is expert in the area of seismic fragility analysis and dynamic qualification of ele.;trical and mechanical equipment. He has participated in and managed over twenty major projects 'nvolving the evaluation and qualification of vibration sensitive equipment and seismic hardening of equipment.
4 As demonstrated by his committee work and publications, Mr. Djordjevic has participated in and contributed steadily to the development of equipment qualification and vibration hardening methodology.
~
1
.. _ _ = -
l PROFESSIONAL GROUPS:
Member, Institute of Electrical and Electronics Engineers, Nuclear Power Engineering Committee Working Group SC 2.5 (IEEE-344)
Chairman, American Society of Civil Engineers Nuclear Structures and Materials Committee, Working Group for the Analysis and Design of Electrical Cable Support Systems Member, Ameiican Society of Mechanical Engineers Operation, Application, and Components Committee on Valves, Working Group SC-5 l
%ma
PROGRAMMATICSOLUTIONS,Inc.
Ilarry W. Johnson i
Education Columbia University - Mechanics Engineer,1968 University ofMiami - Master of Science in Civil Engineering,1967 University ofMiami - Bachelor of Science in Civil Engineering,1965 Bettis Atomic Power Laboratory - Reactor Engineering School,1972 SQUG Walkdown Screening and Seismic Evaluation Training Course,1992 Licenses and Registrations California: Civil Engineer Professional History l
Programmatic Solutions, Inc.1992 Robert L. Cloud & Associates, Inc, Senior Associate, 1990-1992 EQE Engineering, Senior Associate, 1986-1990 Impell Corporation, Section Manager, 1974-1986 Westinghouse Bettis Atomic Power Laboratory, Senior Engineer, 1969-1974 Comell Aeronautical Laboratory, Project Engineer, 1968-1969 Experience Mr. Johnson has over 20 years experience in engineering consulting in the power industry. In particular, Mr. Johnson has extensive experience in project management and design engineering.
Mr. Johnson is heavily involved in applying SQUG methods to nuclear procurement programs, and is active in industry initiatives in this area. To date Mr.
Johnson has managed the development of these methods for eight plants. Mr. Johnson is a contractor for EPRI/PSE in the development of Seismic Technical Evaluation of Replacement Items (STERI) and Generic STERI (GSTERI).
Mr. Johnson participated in the development of the EPRI training programs for New and Replacement Equipment (NARE ) and Seismic Technical Evaluation of I
replacement items (STERI), and is a trainer for both of these course. Mr. Johnson has l
participated in the development of SQUG training programs, and developed the original draft of training for the SQUG seismic interaction module. Mr. Johnson (along with Paul i
j Smith of TRO) developed a 2 day training course in SQUG walkdown methods which has l
i w
PROGRAMMATICSOLUTIONS been presented to GPU Nuclear, PSE&G of New Jersey, Rochester Gas & Electric, New York Power Authority, Carolina Power & Light, Martin Marietta, Stone & Webster, and Gilbert / Commonwealth.
Mr. Johnson is currently performing Peer Reviews for USI A-46 and/or IPEEE for Peach Bottom 2 & 3 and Limerick 1 & 2, Salem Units 1 & 2, Arkansas Nuclear One Units 1 & 2, Grand Gulf, Waterford, Nine Mile Point 2, Byron, Braidwood, LaSalle, Quad Cities, and Dresden. Mr. Johnson is also performing Peer Review of experience based piping re-evaluation for the Brookhaven National Laboratog HFTR, performing a QA Audit of Seismic Programs for Indian Point Unit 2, performing USI A-46 walkdown for Indian Point Unit 3, helping to develop a plant specific seismic program and performing walkdowns to address USI A-46 for Crystal River, participating in the implementation of the Seismic IPEEE for Pery, and providing EPRI/PSE sponsored training for Seismic Technical Evaluation of Replacement Items and for Use of SQUG Methods for New and Replacement Equipment.
Mr. Johnson has performed SQUG type seismic verifications for Peach Bottom, Vermont Yankee, Pilgrim, Three Mile Island, Oyster Creek, Maine Yankee, Princeton Tokamak (TFTR), Savannah River Project, Cruas (France), Sequoyah, and Indian Point Unit 2.
For TVA Bellefonte, Mr. Johnson developed an implementation strategy for the overall IPEEE program (all external events).
Mr. Johnson has performed Independent Peer Reviews of comprehensive seismic interaction programs at Watts Bar and Comanche Peak and of the equipment seismic qualification program at Watts Bar.
Project work included development of new methods for seismic evaluations of mechanical and electrical equipment using SQUG methods. For GPU Nuclear, Mr.
Johnson developed and managed all work associated with the seismic verification of TMI and OCNGS control room cabinets, including verification of many modifications.
At Impell Mr. Johnson was Manager of Engineering Mechanics. This included design, project management and engineering evaluations.
At Bettis Mr. Johnson headed a reactor engineering stress / test group for an advanced submarine project.
At Cornell Aeronautical Laboratory Mr. Johnson performed projects related to automobile crash injury research.
PROGRAMMATICSOLUTIONS i
Selected Publications
" Guideline for the Seismic Technical Evaluation of Replacement Items for Nuclear Power l
Plants (PSE-01)", EPRI NP-7484.
"Use of Seismic Experience Data for Replacement and New Equipment", Nuclear l
Engineering and Design,1990.
" Application ofIndustry Experience Data to Equipment Qualification", Third Symposium
- Current Issues Related to Nuclear Power Plant Structures, E9uipment and Piping, December 1990.
" Impact of General Procurement and Commercial-Grade Item Initiatives", Third Symposium - Current Issues Related to Nuclear Power Plant Structures, Equipment and Piping, December 1990.
" Integration of Seismic Technical Requirements fu Replacement items into Nuclear Plant i
Design Basis", to be presented at ASME PVP Conference, June 1992.
" Role of Seismic Engineers in CGI Technical Evaluations", Ea.thquake Safety &
Licensing Report, June 15,1992.
" Site Specific Procedures for Seismic Verification Walkdown for a Nuclear Plant". Fourth Symposium - Current Issues Related to Nuclear Power Plant Structures, Equipment and Piping, December 1992.
" Peer Review for USI A46 and Seismic IPE", prernted at ASME PVP Conference, July 1993.
" Integration of Seismic Qualification in the Engineering Change Process", presented at Fifth Symposium - Current Issues Related to Nuclear Power Plant Structures, Equipment and Piping, December 1994.
t f
l l
P. Kannsdy ROBERT P.
KENNEDY SEISMIC EXPERT AND WALKDOWN TEAM MEMBER EDUt'ATION:
B.S.
- Civil Engineering, Stanford University M.S.
- Structural Engineering, Stanford University Ph.D. - Structural Engineering, Stanford University REGISTRATION:
State of California, State of Tennessee, State of Texas, State of Alabama PROFESSIONAL HISTORY:
RPK Structural Mechanics Consulting, Yorba Linda, California, President, 1987 to Present National Technical Services Engineering, Los Angeles, California, Vice President, 1985 - 1987 Structural Mechanics Associates, Newport Beach, California, President, 1979-1985 Engineering Decision Analysis Corp., Newport Beach, California, Vice President, 1977
- 1979 Homes and Narvier, Inc., Los Angeles, California, Manager of Engineering Mechanics Division, 1970 - 1979 PROFESSIONAL EXPERIENCE:
Dr. Kennedy has over twenty years experience in static and dynamic analysis plus design of special purpose civil and mechanical-type structures, particularly for the nuclear, petroleum, and defense industries: design of structures to resist extreme loadings including seismic, missile impact, extreme wind, impulsive loads, and nuclear environmental effects; development of computerized structural analysis methods: administrative and program management; and teaching.
Seismic Ruocedness - Nuclear Facilitie_g Chairman, Senior Seismic Review and Advisory Panel (SSRAP), jointly advising both nuclear power utilities and the U.S. NRC on issues relating to seismic ruggedness of existing nuclear power plants.
Member of NRC Expert Panel on Seismic Margin for nuclear power plants. Co-author of Electric Power Research Institute (EPRI) Seismic Margin Research Program.
Provided technical direction on Seismic fragility portion of seismic probabilistic risk assessments for 23 nuclear power plants.
Developed the methodology most commonly used for such studies and author of many technical papers thereon.
Taught short courses on seismic PRA methodology in U.S., Spain, Taiwan, and People's Republic of China.
Consultant on seismic evaluation or design for more than 40 nuclear facilities through the world.
Directed seismic analysis of many nuclear power plant buildings and components.
Directed many nonlinear seismic response analyses investigations.
Evaluated effects of differential earth movement (faulting) on nuclear facility.
Performed a number of dynamic soil-structure interaction analyses of nuclear reactor containment buildings accounting for the nonlinear effects of base slab uplif t.
Directed nonlinear seismic evaluation of nuclear facility to demonstrate increased seismic capacity.
Evaluated concepts for
P. Ksnnsdy seismic response mitigation and increased energy absorption. Has participated in 13 nuclear power plant seismic walkdown.
Dynamic Loads - Nuclear Facilities Extensive experience in the analysis of nuclear facilities subjected to extreme dynamic loads including effects of external missile and aircraft impact, and impulsive loading resulting from loss-of-coolant accident and SRV discharge.
Prime developer of the method currently in extensive use by the nuclear industry in the U.S.
for evaluating the local effects of missile impact on concrete.
Consultant on the effects of aircraft impact for several nuclear plants.
Consultant to General Electric on effects of pool swell loads resulting from LOCA, and on the increased dynamic reserve margin available in structures subjected to pulsive loads.
Consultant to G.E. and Mark I, Mark II, and Mark III Owner's Group on combination of respcnses from multiple dynamic loadings.
Consultant on Mark I and Mark III evaluations to address the conservatism and uncertainty associated with standard structural analyses for SRV loadings. Consultant on methods of response combination and expert witness at Black Fox hearings.
Consultant to Mark I and Mark III groups on conservatism, uncertainty, structural modeling, and load definition for new dynamic loads.
Consultant on three Mark III BWR plants with free-standing steel containment, Leibstadt, Allens Creek, and River Bend, in order to evaluate realistic containment response to SR; loadings as current approaches are overconservative and i
lead to serious design problems.
Developed floor response spectra for final design of attached piping for Leibstadt plant by coupled analysis such that beneficial effects of energy feedback are included.
Developed method to account for the coupling of equipment and piping to the main structure and to account for energy feedback from the subsystem to the structure.
Developed method to account for random phasing of multiple harmonics of condensation oscillation loading in order to compute responsas more compatible with measured results.
Member ASCE committee on impact and impulse analysis of nuclear facilities, and ACI committee which developed code for the design of nuclear safety-related concrete structures subjected to impact and impulse loads.
Dr. Kennedy has personally performed seismic walkdowns of 12 nuclear power plants and serves on numerous government (NRC) and industry advisory boards (EPRI) dealing with seismic qualification of nuclear power plant facilities.
4 l
PROFESSIONAL GROUPS:
Chairman, Seismic Analysis, Nuclear Structure and Materials Committee, Structures Division, ASCE.
Chairman, Seismic Analysis of Safety Class Structures Standard Committee, Technical Council on Codes and Standards, ASCE.
Former Chairman, Gas and Liquid Fuel Lifelines Committee, Technical Council on Lifeline Earthquake Engineering, MCE.
Member, Nuclear Structures and Materials Technical and Administrative Committee, Structures Division, ASCE.
4 Member, Impact and Impulse Analysis, Nuclear Structures and Materials Committee l
Structures Division, ASCE.
Member, Editing Board, ASCE Report entitled " Structural Analysis and Design of Nuclear Plant Facilities."
Member, Ad Hoc Group on Soil-Structure Intersection, Nuclear Structures and Materials Committee, Structures Division, ASCE.
Member, ACI 349, " Subcommittee on Standard Requirements for Nuclear afety-Related Concrete Structures," Design Committee and Working Group 5 -
"Impactive and Impulsive Loads."
1 P. K&nnedy Member, AWWA DlOO Revision Task Force, charged with revising the AWWA Standard for Welded Steel Tenka for Water Storage.
Member, National Research Council Subcommittee on Probabilistic Seismic Hazard Assessment, l
Bruce M. Lory Education State University of New York at Buffalo - B.S.
Mechanical Engineering - 1982 Graduate Studies in Environmental Acoustics Memberships American Society of Mechanical Engineers Committee Memberships ASME-PVP Division - Operations, Applications, and Components Committee EPRI - Structural Reliability and Integrity Working Group SQURTS member and former Technical Chairman EPRI - GSTERI Committee (Phases I - 111)
EPRI - SEQUAL Committee STERl/NARE Software Development Utility Steering Group ChairmanASCE - Nuclear Energy Employment History Comed: 1991 to Present ABB Impell 1989 to 1991 Sargent & Lundy Engineers: 1979 to 1989 Experience Mr. Lory is currently a member of the Mechanical & Structural Engineering Group of the Nuclear Engineering Department of Comed. In this role, he assists the Chief Mechanical / Structural Engineer in providing technical oversight and technical assistance to the six Comed nuclear sites. He has serving as the corporate project engineer in assisting the sites develop their reports for the SQUG and seismic IPEEE projects. In this role he apprised site reps of ongoing SQUG and seismic IPEEE issues, served as part of the walkdown team and performed evaluations as a part of this effort..
In addition, he currently is the corporate seismic equipment qualification engineer, assisting all six nuclear sites address various seismic qualification challenges related to equipment and components, in this capacity, he witnesses seismic qualification testing on an as-needed basis in support of plant startup and mentors to the sites so that they are capable and perform seismic qualification activities within the site organization.
From 1989 through 1991 he assisted Comed in coordinating the engineering activities related to their maintenance of the EQ (Environmental Qualification) program for compliance with 10CFR50.49. He has performed many EQ binder updates and calculations, prepared / reviewed EQ test plans and reports, and has witnessed several EQ tests.
Earlier, Mr. Lory had been involved on Comed projects related to Environmental and I
Seismic Qualification of equipment and components. He has performed EQ and SQ activities on all Comed station projects. He was also part of the S&L team which per-formed check valve calculations in response to Comed's report on the INPO SOER regarding check valve fatigue. He was part of the S&L team which provided resolution to EDSFl issues in the seismic field for Dresden & LaSalle stations.
D. Stsysnson JOHN D.
STEVENSON SEISMIC EXPERT AND WALKDOWN TEAM MEMBER EDUCATION:
B.S.
- Civil Engineering - Virginia Military Institute, 1954 M.S.
- Civil Engineering - Case Institute of Technology, 1962 Ph.D. - Civil Engineering - Case Institute of Technology, 1968 REGISTRATION:
Commonwealth of Virginia, State of Ohio PROFESSIONAL HISTORY:
1 Stevenson & Associates, Cleveland, Ohio, President: 1981 - present Structural Mechanics Associates, Cleveland, Ohio, Vice President: 1980 - 1981 Weodward Clyed Consultants, Cleveland, Ohio, Vice President: 1979 - 1980 j
A. G. McKee & Co., Cleveland, Ohio, Vice President: 1976 - 1979 Case Western Reserve University, Cleveland, Ohio, Assoc. Prof.: 1974 - 1976 Westinghouse Electric Co., Pittsburgh, Pennsylvania, Consultant: 1972 - 1974 University of Pittsburgh, Pittsburgh, Pennsylvania, Adjunct Professor: 1970 - 1972 Westinghouse Nuclear Energy Systems, Manager Structural System Engineering: 1966 -
1970 Virginia Military Institute, Assistant Professor: 1957 - 1962 PROFESSIONAL EXPERIENCE:
Since November 1981, Dr. Stevenson has managed and has served as President and Senior Consultant to Stevenson & Associates.
The firm specializes in high technology consulting and engineering services associated with failure analysis of structural and mechanical systems; extreme loads; and nonlinear,
- dynamic, probabilistic and high temperature analyses.
His years of expertise include structural and mechanical design and qualification of nuclear power plant structures and components.
He serves on several committees of the ASCE, ASME, ANS, ACI and AISC charged with the development of standards devoted 4
to design of nuclear plant facilities.
Dr. Stevenson's relevant experience includes I
seismic walkdowns of 13 - nuclear power stations and 23 years as a structural-mechanical engineer with particular application to earthquake design and analysis.
A list of earthquake related projects which Dr. Stevenson performed or directly supervised is as follows:
1.
Developed seismic design criteria for 5
nuclear power stations:
Westinghouse Turnkeys 2.
Reviewed and approved seismic design adequacy from plans and specifications for 5 nuclear power plants: Westinghouse Turnkeys
D.
S tsysn::on i
3.
Evaluation of seismic design adequacy of mechanical equipment for French nuclear power station:
Fessenhein station l
4.
Evaluation of seismic design adequacy of liquified natural and petroleum gas storage facilities:
U.S. General Acccanting Office l
5.
Testing of electrical racks to demonstrate seismic design adequacy:
Federal Pioneer, Canada 6.
Analytical seismic qualification of spent fv.el racks nonlinear analysis for 10 nuclear power plants 7.
Quality Assurance audit (technical) of the Tokomak fusion test reactor tritium retention structures to resist seismic loads:
U.S.
Department of Energy 8.
Prepare preliminary assessment of requirements and installation specification for anchorage of electrical equipment in the Monticello Nuclear Plant in anticipation of meeting I&E 80-21 requirements: Northern States Power Co.
9.
Review and evaluation of the Purex facility seismic capabilities at Hanford Plant:
U.S. Department of Energy 10.
Survey and evaluation of the L reactor mechanical and electrical equipment seismic capabilities for Savannah River Plant:
E.
I. DuPont 11.
Systematic evaluation of the seismic capacity of a 600 MW Candu reactor station for safety shutdown in Argentina 12.
Review of literature and develop recommendations for piping system damping
)
values for Kraftwerk Union and Mitsubishi 13.
Detailed evaluation of seismic design adequacy for selected mechanical and electrical equipment of five nuclear power stations as part of the U.S.
NRC systematic evaluation program:
U.S. Nuclear Regulatory Commission 14.
Review of seismic design adequacy of nuclear plant facilities for the D.C.
I Cook Nuclear Power Plant: American Electric Power Corporation 15.
Systematic evaluation of the seismic capacity of mechanical and electrical equipment for the Connecticut Yankee Nuclear Power Plant:
Northeast Utilities 16.
Systematic evaluation of the seismic capacity of mechanical and electrical equipment for the Maine Yankee Nuclear Power Plant:
Yankee Atomic Co.
17.
Consultant to EPRI to develop criteria for OBE Exceedance based on measure and observed site behavior: Electric Power Research Institute 18.
Consultant to NRC to develop an experience data base and criteria for design and analysis of piping systems to resist seismic loads:
19.
Consultant to NRC to evaluate procedures for HCLPF assessment of equipment in nuclear power plants (NUREG/CR-5270) :
U.S.
Nuclear Regulatory Commission 20.
Review of earthquake resistance of plutonium fabrication facility:
Rockwell International Corp.
21.
Consultant to EPRI to develop procedures for plant start-up following a damaging earthquake at a nuclear power plant site: Electric Power Research Institute
. ~..
~... -. - - _ _. -
1 D. Stavanson i
i I
j 22.
Personally performed seismic walkdowns of 16 nuclear power plants in the U.S.,
Europe, Asia and South America.
4 i
I PROFESSIONAL GROUPS:
l I
Member, American Society of Civil Engineers, Structural Division Committee on Nuclear Safety; Structural Division Committee on Nuclear Structures and Materials; t
j Steering Committee on Development of a Manual of Profersional Practice for Quality l
in the Constructed Project 1
Former Chairman, American Society of Civil Engineers, Executive Committee Technical
. Council Codes and Standards l
Chairman, American Society of Civil Engineers, Nuclear Standards Committee
{
Member, American Concrete Institute, Joint ACI-ASME Subgroup on Design of Concrete 2
Components in Nuclear Service, ASME BPVC-Section III-Div. 2 i.
l Member, American Society of Mechanical Engineers, Subgroup on Design of ASME BPVC-l-
Section III-Div. 1 Nuclear Components Subcommittee on Qualification. of Mechanical Components in Nuclear Service Member, Nuclear Standards Management Board of ANSI representing ASCE
- Member, U.S.
Representative International Standards Committee SC 85/3/7 on Seismic Criteria for Nuclear Plants j
- Member, U.S. Representative International Atomic "nergy Agency Working Group on the l
Development of Seismic Design Standards
- Member, ANS-2, American Nuclear Society Committee on Site Evaluation; NUPPSCO, i
American Nuclear Society Committee on Nuclear Power Plant Codes and Standards 1
(
Member, AISC, American Institute of Steel Constre.ction Committee on Specifications for Structural Steel in Safety Class Nuclear Strr.ctures Member, Earthquake Engineering Research Institute
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GEORGE G. THOMAS I
EDUCATION:
B.S. - Civil Engineering - Purdue University,1976 M.S. Civil Engineering - Purdue University,1978 REGISTRATION:
Registered Engineer-in-Training: Indiana Passed Principals and Practices Exam: Texas PROFESSIONAL HISTORY:
Stevenson & Associates, Cleveland, Ohio, Project Engineer,1982 - Present Cleveland State University, Cleveland Ohio, Engineering Instructor, 1981 -1987 Davy McKee Company, Cleveland, Ohio, Lead Engineer, 1980-1982 Exxon Production Research Company, Texas, Research Engineer,1978 - 1980 McDermott-Hudson Engineering Con., Engineer / Draftsman, 1976 -1977 PROFESSIONAL EXPERIENCE:
Mr. Thomas has served as a Project Manager for Stevenson & Associates on a variety of projects involving the evaluation and qualification of nuclear safety-related structures, equipment and piping.
He has been responsible for the detailed seismic analysis, testing and qualification of a variety of mechanical and electrical equipment and piping systems, including the anchorage and support structure evaluations. He has performed analysis of safety-related piping systems subject to extreme loadings of earthquake, tomado wind, and missile impact. He has performed a failure analysis of a piping system due to water hammer loading. He has performed in-situ modal testing of nuclear components to determine their dynamic characteristics, and to determine structure amplified response spectra.
Mr. Thomas developed a large portion of the Generic implementation Procedure, GIP, for the Seismic Qualification Utilities Group, SQUG, that defined the generic walkdown requirements for USl A-46. Mr.
Thomas was a walkdown participant for both the SQUG Zion and Nine Mile Point 1 trial plant walkdowns for SQUG. Mr. Thomas prepared and presented the training modules on the GIP for the SQUG training program, and was one of the Subject Matter Experts for the development of the training program and the training program tapes.
Mr. Thomas was the Project Manager and a Lead Walkdown Engineer on a Seismic Review Team for the D.C. Cook Nuclear Station Units 1 and 2 for USl A-46, Arkansas Nuclear One Units 1 and 2 for a combine USI A-46 and Seismic Margins Assessment for IPEEE, and Waterford 3, Grand Gulf 1 and V.C. Summer 1 for the Seismic Margins Assessment for IPEEE. In addition, Mr. Thomas was the Project Engineer for the Turkey Point Unit 3 and 4 and St. Lucie Unit 1 and 2 USI A-46 and seismic IPEEE efforts using Florida Power & Light's
%tL
Utility specific program. Mr. Thomas was Project Manager and participated on the walkdown for the seismic IPEEE Probabilistic Risk Assessment for Beaver Valley Unit 2.
Mr. Thomas as Project Manager for these evaluations is the primary author of the final IPEEE and USl A-46 reports.
Mr. Thomas has developed background material and a seismic criteria document for a utility client for l
an older nuclear power plant to be used by the utility in seismic evaluations for modifications and j
additions to plant structures, equipment and piping. He has served as a Project Engineer for a pilot l
snubber reduction program for a utility and has provided expert consulting services to another utility l
for their in-house snubber reduction program.
Mr. Thomas developed program COMPARE which consisted of the assembly of a data base of l
nuclear power plant components that have been previously seismically qualified. He developed the computer software on an IBM-PC computer to store and retrieve seismic qualification data on these components. The program is used to facilitate seismic qualification of components not previously qualified by comparison to those components qualified in the data base.
Mr. Thomas has served as a part time instructor in the School of Civil Engineering and Engineering Technology at Cleveland State University.
On the undergraduate level, he has taught Static, Dynamics, Material Science, Structural Analysis and Concrete Design. In the graduate program, he has taught Advanced Steel Design. In all of the teaching assignments he was responsible for developing the course outline, lecture notes, problems and tests.
Mr. Thomas served as Lead Engineer in the Piping Engineering Group of Davy McKee Company. His work consisted of the design supervision and design of piping networks in a number of different petrochemical facilities. His responsibilities comprised the following:
Designing and analyzing piping networks subjected to thermal, weight, wind, earthquake, and pressure loadings using both manual and computerized techniques.
Design and analysis of pipe supports and pipe support structures. Preparing specifications for expansion joints, and providing an overall support and expansion joint package. Preparing hydrotest procedures and planning of hydrotest circuits.
Mr. Thomas served as a Research Engineer in the Offshore Structures Division of Exxon Production Research Company. His work consisted of development of design and analysis procedures for the Guyed Tower, a type of deepwater offshore production platforms. He performed the dynamic, structural, and fatigue analysis necessary for the Guyed Tower design of three proposed structures.
He also wrote, revised, and maintained computer programs used in the Guyed Tower analysis and design procedures.
Mr. Thomas served as an Engineer and Draftsman for Robert W. Crooks Consulting Engineer and McDermott-Hudson Engineering Company.
His work consisted of structural design, drafting, foundation design, railroad layout, and site development for chemical facilities.
PUBLICATIONS:
Mr. Thomas, G.G. and Finn, L.D., "A Guyed Tower for North Sea Production," Presented at 4th Offshore North Sea Technology Conference in Stavanger, Norway,1980.
Thomas, G.G. and Starck, R.G., " Overview of SQUG Generic Implementation Procedure (GIP),
"Second Symposium on Current Issues Related to Nuclear Power Plant Structures, Equipment, and h
l
Piping with Emphasis on Resolution of Seismic issues in Low-Seismicity Regions, EPRI NP-6437-D Proceedings, May 1989.
Thomas, G.G. and Starck, R.G., " Overview of SQUG Generic implementation Procedure (GlP),"
Nuclear Engineering and Design, Vol. 123 (1990), Nos. 2&3, October (ll) 1990, Pgs 225-231.
h
Quad Citi s A-46 Fin:1 Report June 19,1996 Appendix D l
Quad Cities Nuclear Station l
Screening Verification Data Sheets (SVDS) l l
i Quad Cities Nuclear Power Station i
G3/19/96 02.20 PM SCREENING VERIFICATION DATA SHEET OVDS)
Page 01 i
r Eq Eq. ID Rev Sys/Eq. Desc Bldg.
F1 El.
Rm or Rw/Cl Base El. <40'?
Cap.
Demd.
Cap >
Caveats Anchor Interad Equip Cl No Spec.
Spec Demd?
OK?
OK7 OK7 OK7 0
U1HCU 0
CRD/HCU RB 595.00 595.00 N/A DOC RRS No Yes Yes No No 0
U2 HCU 0
CRD/HCU RB 595.00 595 00 N/A DOC RRS No Yes Yes Yes No 1
125 VDC BUS 1 A 0
125 VDC / BUS 1 A,125 VDC BATTERY TB 615 00 H-24 615.50 N/A ABS CRS Yes Yes Yes No No 1
125 VDC BUS 1 A-1 0
125 VDC / BUS 1 A-1,125 VDC TB 615.00 H-24 615.50 N/A ABS CRS Yes Yes Yes No No BATTERY
,m p h
1 125 VDC BUS IB 0
125 VDC / BUS. TB RES 18 TB 615.00 H-24 615.50 N/A ABS CRS Yes
"'W es-
"8P(b-Yes
.Mee-1 125 VDC BUS 18-1 0
125 VDC / BUS. TB RES 1B-1 TB 615.00 H-24 615.50 N/A ABS CRS Yes Jyt8es-ifdeo-Yes
. Vee. fpA 1
125 VDC BUS 2A 0
125 VDC / BUS 2A.125 VDC BATTERY TB 615.00 H-1 615.50 N/A ABS CRS Yec Yes Yes Yes Yes 1
125 VDC BUS 2A-1 0
125 VDC / BUS 2A-1,125 VDC TB 615.00 H-1 615.50 N/A ABS CRS Yes Yes Yes Yes Yes BATTERY 1
125 VDC BUS 2B 0
125 VDC / BUS. TB RES 28 TB 615.00 H-1 615.50 N/A ABS CRS Yes Yes Yes Yes Yes 1
125 VDC BUS 28-1 0
125 VDC / BUS, TB RES 28-1 TB 615 00 H-1 615.50 N/A ABS CRS Yes Yes Yes Yes Yes 1
250VDC MCC 1 0
250 VDC / MCC TB 615 00 G-25 615.50 N/A ABS CRS Yes Yes Yes Yes Yes 1
250VDC MCC 1 A 0
250 VDC / MCC RB 623.00 L-19 623 00 N/A AEcS CRS No Yes Yes Yes No j
1 250VDC MCC 18 0
250 VDC / MCC RB 623.00 N-18 623.00 N/A ABS CRS No Yes Yes Yes No 1
250VDC MCC 2 0
250 VDC / MCC TB 615.00 G-2 615.50 N/A ABS CRS Yes Yes No No No 1
250VDC MCC 2A 0
250 VDC / MCC RB 623.00 L-11 623 00 N/A ABS CRS No Yes Yes Yes No 1
1 250VDC MCC 2B 0
250VDC / MCC RB 623.00 M-10 623.00 N/A ABS CRS No Yes Yes Yes No 1
MCC 18-1 A 0
480 VAC / MCC RB 623 00 M-17 623.00 N/A ADS CRS No No No Yes No 1
MCC 18-1 A-1 0
208 VAC / MCC RB 623.00 N-16 623.00 N/A ABS CRS No No Yes No No 1
MCC 18-1 A-1 PNL 0
120/208 / MCC DIST PANEL RB 623.00 N-16 623.00 N/A ABS CRS No No Yes No No 1
MCC 18-1B 0
480 VAC / MCC RB 623.00 N-17 623.00 N/A ABS CRS No No No No No 1
MCC 18-2 0
480 VAC / MCC TB 615.00 G-22 615.50 N/A ABS CRS Yes Yes Yes No No 1
MCC 18-3 0
480VAC / MCC RB 623.00 N-15 623.00 N/A ABS CRS No No No No No 1
MCC 10/19-5 0
480VAC / MCC RB 595.00 N-17 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 1
MCC 19-1 0
480 VAC / MCC RB 623.00 M-17 623.00 N/A ABS CRS No No No Yes No i
l1 MCC'I9 t-1 PNL 0
208 VAC / MCC DIST PANEL RB 623.00 N-18 623.00 N/A ABS CRS No No No Yes No Certification:
Certification:
All the information contained on this Screening Venfication Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capabi!ity Engineers regarding systems and operations our knowledge and belief, correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief. correct and conclusion (whether venfied to be seismically adequate or not).
accurate.
Approved: (Signatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operations Engineer is required if the Seismic Capabihty are required, there should be atleast two on e SRT. All signatones should agree with all the Engineers deem it necessary.)
entnes and conclusions. One signatory sho a licensed fe i al en 'neer.)
f l
R. P. Kennedy l
l M
[l l
l l
Pnnt or Type Name
' Signature Date Pnnt or Type Name Sqnature Date l
1 D Stevenson l
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i Date Pnnt or Type Name Sqnature Date Pnnt or Type Name
~
Signature l
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Pnnt or Type Name Sqnature Date Pnnt or Type Name Sgnature Date m
l l
Quad Cities Nuclesr Power St: tion 06/19/96 02:20 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page # 2 l
Eq.
Eq. ID Rev Sys/Eq Desc Bldg.
FI El.
Rm or Rw/Cl Base El. <407 Cap.
Demd.
Cap >
Caveats Anchor interact Equip Cl No Spec.
Spec Demd?
OK7 OK?
OK?
OK7 j
1 MCC 19-2 0
480 VAC / MCC TB 615.00 G-23 615.50 N/A ABS CRS Yes Yes Yes Yes Yes 1
MCC 19-4 0
480 VAC / MCC RB 623.00 N-18 623.00 N/A ABS CRS No No No Yes No 1
MCC 28-1 A 0
480 VAC / MCC RB 623 00 M-8 623.00 N/A ABS CRS No No No Yes No 1
MCC 28-1 A-1 0
208 VAC / MCC RB 623.00 N-8 623 00 N/A ABS CRS No No Yes No No 1
MCC 28-1B 0
480 VAC / MCC RB 623.00 N-9 623.00 N/A ABS CRS No No Yes No No 1
MCC 28-2 0
480VAC / MCC TB 615 00 G-4 615 50 N/A ABS CRS Yes Yes Yes Yes Yes 1
MCC 28-3 0
480VAC / MCC RB 623.00 N-11 623 00 N/A ABS CRS No Yes Yes Yes No 1
MCC 28/29-5 0
480VAC / MCC RB 595 00 N-8 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 1
MCC 29-1 0
480 VAC / MCC RB 623.00 M-10 623.00 N/A ABS CRS No Yes Yes Yes No 1
MCC 29-1-1 0
208 VAC / MCC RB 623.00 N-9 623 00 N/A ABS CRS No No Yes No No 1
MCC 29-1-1 PNL 0
120/208 / MCC DIST PANEL RB 623.00 N-9 623 00 N/A ABS CRS No No Yes No No 1
MCC 29-2 0
480VAC / MCC TB 615.00 G-3 615.50 N/A ABS CRS Yes Yes Yes Yes Yes l
1 MCC 29-4 0
480 VAC / MCC RB 623.00 N-10 623.00 N/A ABS CRS No No Yes No No 1
2 SWGR 18 0
480VAC / SWGR 18 RB 647.00 H-13 647.50 N/A GERS CRS Yes Yes Yes No No
(
2 SWGR 19 0
480VAC / SWGR 19 RB 647.00 H-15 647.50 N/A GERS CRS Yes Yes Yes No No 2
SWGR 28 0
480VAC / SWGR 28 RB 647.00 H-11 647.50 N/A GERS CRS Yes Yes Yes No No 2
SWGR 29 0
480VAC / SWGR 29 RB 647.00 H-13 647.50 N/A GERS CRS Yes Unk Unk No No
[
3 SWGR 13 0
4160VAC / SWGR 13 TB 615 00 G-22 615.50 N/A ABS CRS Yes {,-Vee-f#ee-Yes Wes-Me 3
SWGR 13-1 0
4160VAC / SWGR 13-1 TB 639 00 H-14 639.00 Yes BS GRS Yes Unk Unk Yes Unk 3
SWGR 14 0
4160VAC / SWGR 14 TB 615 00 H-22 615.50 N/A ABS CRS Yes)5 Veo
,V, Vee-Yes Vee-MO 3
SWGR 14-1 0
4160VAC / SWGR 14-1 TB 639.00 H-16 639 00 Yes BS GRS Yes Unk Unk Yes Unk 3
SWGR 23 0
4160VAC / SWGR 23 TB 615.00 G-4 615.50 N/A ABS CRS Yes No No No No 3
SWGR 23-1 0
4160VAC / SWGR 23-1 TB 639.00 H-10 639.00 Yes BS GRS Yes Unk Unk Yes Unk 3
SWGR 24 0
4160VAC / SWGR 24 TB 615 00 H-4 615 50 N/A ABS CRS Yes No No No No 3
SWGR 24-1 0
4160VAC / SWGR 24-1 TB 639.00 H-12 639.00 Yes BS GRS Yes Unk Unk No No 4
480 VAC / TRANSFORMER 28-1 A-1 RB 623.00 M-8 623.00 N/A ABS CRS No Yes Yes Yes No 4
480VAC / TRANSFORMER 28-2 TB 615.00 G-4 615.50 N/A ABS CRS Yes Yes Yes Yes Yes Certification:
Certification:
All the information contained on this Screening Venfication Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capability Engineers regarding systems and operations our knowledge and belief, correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and befef, correct and conclusion (whether venfied to be seismically adequate or not).
accurate.
Approved: (Signatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operations Engineer is required if the Seismic Capability are required; there should be atleast two on e SRT. All signatones should agree with all the Engineers deem it necessary.)
l entries and conclusions. One signatory sho d a licensed rofe nal engineer.)
L l
R.P. Kennedy l
/ -
l 2
l l
l l
Pnnt or Type Name Sgnature Date Pnnt or Type Name Sgnature Date l
ldht_f % l l
l l
l J. D. Stevenson l
~
6gnature I Datb Pnnt or Type Name Sqnature Date f m-u 3
Pnnt or Type Name i
I I
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Pnnt or Type Name Sgnature Date Pnnt or Type Name Sgnature Date
Quad Cities Nuclear Power Station C3/19/96 02.20 PM SCREENING VERIFICATION DATA SHEET (SVDS)
PageO 3 Eq Eq. ID Rev Sys/Eq. Desc Bldg.
Fi El.
Rm or Rw/Cl Base El. <407 Cap.
Demd.
Cap >
Caveats Anchor interact Equip Cl No Spec.
Spec Demd?
OK7 OK7 OK7 OK7 4
480/208 / TRANSFORMER 29-1-1 RB 623 00 N-9 623.00 N/A ABS CRS No Yes Yes Yes No 4
4160VAC / TRANSFORMER 18 RB 647.00 H-13 647.50 N/A ABS CRS No Yes Yes Yes No 4
SWGR 19 TRANSFMR 0
480VAC / TRANSFORMER 19 RB 647.00 H-15 647.50 N/A ABS CRS No Yes Yes Yes No 4
4160VAC / SWGR 28 RB 647.00 H-11 647.50 N/A ABS CRS No Yes Yes Yes No 4
480VAC / SWGR 29 RB 647.00 H-13 647.50 N/A ABS CRS No Yes Yes Yes No 5
1-1001-65A 0
RHRSW; PUMP 1 A TB 547.00 C-17 547.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1-1001-65C 0
RHRSW / PUMP 1C TB 547.00 C-19 547.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1-3903 0
DG CLG / PUMP, DG COOLING TB 547.00 C-21 547.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1-5203 0
DG FO / PUMP, FUEL OIL TRANSF DG 595.00 G-23 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1-5208 0
DG FO / PUMP, FUEL PRIME DG 595.00 G-23 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1-5209 0
X-DG FO / PUMP, ENG DRVN FUEL DG 595.00 G-23 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1-6650 0
X-DG FO / PUMP, ENG DRVN SCAVEN DG 595.00 G-23 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1-6651 0
X-DG FO / PUMP, ENG DRVN MN L &
DG 595.00 G-23 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1-6666A 0
X-DG CLG / PUMP, ENG DRVN CLG DG 595.00 G-23 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1-6666B 0
X-DG CLG / PUMP, ENG DRVN CLG DG 595.00 G-23 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1/2-3903 0
OG CLG / PUMP, DG COOLING TB 547.00 D-21 547.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1/2-5203 0
DG FO / PUMP, FUEL OIL TRANSF DG 595.00 N-13 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1/2-5208 0
DG FO / PUMP, FUEL PRIME DG 595 00 N-13 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1/2-5209 0
DG FO / PUMP, ENG DRVN FUEL DG 595.00 N-13 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 5-1/2-6650 0
X-DG FO / PUMP, ENG DRVN SCAVEN DG 595 00 N-13 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1/2-6651 0
X-DG FO / PUMP, ENG DRVN MN L &
DG 595.00 N-13 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1/2-6666A 0
X-DG CLG / PUMP, ENG DRVN CLG DG 595.00 N-13 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1/2-66668 0
X-DG CLG / PUMP, ENG DRVN CLG DG 595.00 N 13 595 00 N/A ABS Ge<S Yes Yes Yes Yes Yes 5
2-1001-145A 0
RHR / COOLER RB 554.00 M-8 554.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
2-1001-1458 0
RHR / COOLER RB 554 00 M-8 554.00 N/A ABS CPS Yes Yes Yes Yes Yes 5
2-1001-65A 0
RHRSW / PUMP 2A TB 547.00 C-5 547.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
2-1001-65C 0
RHRSW / PUMP 2C TB 547.00 C-7 547.00 N/A ABS CRS Yes Yes Yes Yes Yes Certification:
Certification:
All the information contained on this Screening Venfication Data Sheet (SVDS) is, to the best of The informaticn provided to the Seismic Capability Engineers regarding systems and operations our knowledge and behef, correct and accurate. "All information" includes each ertry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and conclusion (whether verified to be seismically adequate or not).
accurate.
Approved: (Signatures of all Seismic Capabihty Engineers on the Seismic Rev' w Team (SRT) Approved: (One signature of Systems or Operations Engineer is required if the Seismic Capabihty e
are required; there should be atleast two o he SRT, All signatones should agree with all the Engineers deem it necessary.)
entries and conclusions. One signatory sh a licensed pro.. ional engineer.)
l R.P. Kennedy
/W l
l l
l Pnnt or Type Name
- Sgnature '
Date Pnnt or Type Name Sgnature Date l
J. D. Stevenson l
- mnW l
t'd Nl l
l l
- Date Print or Type Name Sqnature Date Pnnt or Type Name
{
Sqnature l
I I
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Pnnt or Type Name Sqnature Date Pnnt or Type Name Sgnature Date
Quad Cities Nuclear Power Station 06/19/96 02:20 PM SCREENING VERIFICATlON DATA SHEET (SVDS)
PageC4 Eq.
Eq. ID Rev Sys/Eq. Desc Bldg.
FI El.
Rm or Rw/Cl Base El. <40'?
Cap.
Demd.
Cap >
Caveats Anchor interact Equip 1
Cl No Spec.
Spec Demd?
OK7 OK7 OK7 OK7 5
2-1006A 0
X-RHR / SEPARATOR, CYCLONE RB 554.00 M-8 554.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
2-3903 0
DG CLG / PUMP, DG COOLING TB 547.00 C-5 547.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
2-5203 0
DG FO / PUMP, FUEL OlL TRANSF DG 595.00 G-3 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
2-5208 0
DG FO / PUMP, FUEL PRIME DG 595.00 G-3 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
2-5209 0
X-DG FO / PUMP, ENG DRVN FUEL DG 595 00 G-3 595 00 N/A ABS CRS Yes Yes Yes Yes Yes j
5 2-6650 0
X-DG FO / PUMP, ENG DRVN SCAVEN DG 595.00 G-3 595.00 N/A ABS CRS Yes Yes Yes Yes Yes j
5 2-6651 0
X-DG FO / PUMP, ENG DRVN MN L &
DG 595.00 G-3 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
2-6666A 0
X-DG CLG / PUMP, ENG DRVN CLG DG 595 00 G-3 595.00 N/A ABS CRS Yes Yes Yes Yes Yes i
5 2-6666B 0
X-DG CLG / PUMP, ENG DRVN CLG DG 595 00 G-3 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
6 2-1002A 0
RHR / PUMP,2A RB 554.00 M-8 554.00 N/A ABS CRS Yes Yes Yes Yes Yes l
7 1-0302-21 A 0
CRD / VALVE, NO. BNK SDV VNT RB 595 00 L-13 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0302-210 0
CRD / VALVE, NO. BNK SDV VNT RB 595.00 L-13 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0302-21C 0
CRD / VALVE, SO BNK SDV VNT RB 595 00 L-18 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0302-21D 0
CRD / VALVE, SO. BNK SDV VNT RB 595.00 L-12 595.00 Yes BS GRS Yes Yes N/A Yes Yes I
7 1-4699-226 0
DG AIR / VALVE, AIR START RELAY DG 595.00 G-23 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-4699-306A 0
DG AIR / VALVE, RELIEF DG 595.00 G-23 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-4699-3068 0
DG AIR / VALVE, RELIEF DG 595.00 G-23 595.00 Yes BS GRS Yes Yes N/A Yes Yes
[
7 1-4699-306C 0
DG AIR / VALVE, RELIEF DG 595.00 G-23 595.00 Yes BS GRS Yes Yes N/A Yes Yes
[
7 1-4699-306D 0
DG AIR / VALVE, RELIEF DG 595 00 G-23 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-5201-RV 0
DG FO / VALVE, RELIEF DG 595.00 G-23 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-6663 0
X-DG CLG / VALVE, TEMP REG DG 595 00 G-23 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
1/2-4699-226 0
DG AIR / VALVE, AIR START RELAY DG 595 00 N-13 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
1/2-4699-306A 0
DG AIR / VALVE, RELIEF DG 595.00 N-13 595.00 Yes BS GRS Yes Yes N/A Yes Yes l
7 1/2-4699-3068 0
DG A!R / VALVE, RELIEF DG 595.00 N-13 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1/2-4699-306C 0
DG AIR / VALVE, RELIEF DG 595.00 N-13 595.00 Yes BS GRS Yes Yes N/A Yes Yes f
7 1/2-4699-306D 0
DG AIR / VALVE, RELIEF DG 595.00 N-13 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
1/2-5201-RV 0
DG FO / VALVE, RELIEF DG 595.00 N-13 595.00 Yes BS GRS Yes Yes N/A Yes Yes Certification:
Certification:
l All the information contained on this Screening Venfication Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capability Engineers regarding systems and operations our knowledge and belief, correct and accurate "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and conclusion (whether verified to be seismically adequate or not).
accurate.
Approved: (Signatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operations Engineer is required if the Seismic Capabikty
~
are required; there should be atleast two on the SRT, All signatories should agree with all the Engineers deem it necessary.)
[
entries and conclusions. One signatory sh be a licen prof ssionalengineer.)
a l
l l
l l
l R.P. Kennedy Pnnt or Type Name Signature
/
Date Pnnt or Type Name Sgnature Date l
J. D. Stevenson l
)
l 1
hl l
l l
Pnnt or Type hama l
Sgnature
~
l Date Pnnt or Type Name Sqnature Date l
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Pnnt or Type Name Sgnature Date Pnnt or Type Name Sgnature Date l
l
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l Quad Cities Nuclear Power St:. tion 1
06/19/96 02 20 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page#5 I
Eq Eq.1D Rev Sys/Eq. Desc Bidg.
FI El.
Rm or Rw/Cl Base El.
<40*?
Cap.
Demd.
Cap >
Caveats Anchor interact Equip Cl No Spec.
Spec Demd?
OK7 OK?
OK7 OK?
7 1/2-5741-319A 0
RHRSW / VALVE TB 615.00 E-23 615.50 Yes BS GRS Yes Yes N/A Yes Yes 7
1/2-5741-333 0
RHRSW / VALVE, FLOW CONTRL TB 615.00 E-24 615.50 Yes BS GRS Yes Yes N/A Yes Yes 7
1/2-5741-345 0
RHRSW / VALVE, RELIEF TB 615.00 E-24 615.50 Yes BS GRS Yes Yes N/A Yes Yes 7
1/2-6663 0
X-DG CLG / VALVE, TEMP REG DG 595 00 N-13 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0302-21 A 0
CRD / VALVE, NO. BNK SDV VNT RB 595 00 L-7 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0302-218 0
CRD / VALVE, NO. BNK SDV VNT RB 595.00 L-7 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0302-21C 0
CRD / VALVE, SO. BNK SDV VNT RB 595 00 L-12 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0302-21 D 0
CRD / VALVE, SO. BNK SDV VNT RB 595.00 L-12 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-1001-125A 0
RHR / VALVE, RELIEF RB 554 00 M-8 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-1001-166A 0
RHR / VALVE, RELIEF RB 580.00 M-7 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-4699-226 0
DG AIR / VALVE, AIR START RELAY DG 595.00 G-3 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-4699-306A 0
DG AIR / VALVE, RELIEF DG 595 00 G-3 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-4699-306B 0
DG AIR / VALVE, RELIEF DG 595 00 G-3 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-4699-306C 0
DG AIR / VALVE, RELIEF DG 595 00 G-3 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-4699-306D 0
OG AIR / VALVE, RELIEF DG 595 00 G-3 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-5201-RV O
DG FO / VALVE, RELIEF DG 595 00 G-3 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-6663 0
X-DG CLG / VALVE. TEMP REG DG 595.00 G-3 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-4699-310 0
DG AIR / SOV, START AIR DG 595.00 G-23 595.00 Yes BS GRS Yes Yes N/A fes Yes 8
1-5201-SOV 0
DG FO / SOV, DAYTANK FILL DG 595 00 G-23 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1/2-4699-310 0
DG AIR / SOV, START AIR DG 595.00 N-13
$95.00 Yes BS GRS Yes Yes N/A Yes Yes 8
1/2-5201-SOV 0
DG FO / SOV DAYTANK FILL DG 595 00 N-13 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1/2-5741-319A (S) 0 RHRSW / SOV, PILOT TB 615.00 E-23 615.50 Yes BS GRS Yes Yes N/A Yes Yes 8
2-0302-19A 0
CRD / SOV, B/U SCRAM RB 595.00 L-8 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-0302-20A 0
$95.00 L-8 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-0302-20B 0
CRD / SOV, SDV VNT & DRN RB 595.00 L-8 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-1001-16A 0
RHR / VALVE. HX 2A BYPASS RB 554.00 M-8 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-1001-19A 0
RHR / VALVE, LOOP A CROSS TIE RB 580.00 M-8 595 00 Yes BS GRS Yes Yes l N/A Yes Yes Certrfication:
Certification:
All the information contained on this Screening Venfication Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capabihty Engineers regarding systems and operations our knowledge and belief, correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and conclusion (whether venfied to be seismically adequate or not).
accurate.
Approved- (Sgnatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operations Engineer is required if the Seismic Capability are required; there should be atleast two on the SRT. All signato es should agree with all the Engineers deem it necessary.)
entries and conclusions. One signatory she a licens ro onal engineer.)
()
f l /
l l
l l
l R.P. Kennedy
~
SgnstuTe-Date Pnnt or Type Name Sqnature Date Pnnt or Type Name l
1 D. Stevenson l
) l
%bh l l
l l
Pnnt or Type Name h
Signature L Date Pnnt or Type Name Sqnature Date i
I I
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I I
Pnnt or Type Name Signature Date Pnnt or Type Name Sqnature Date
Quad Cities NucleIr Power Station 06/19/SS 02.20 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page # 6 Eq.
Eq ID Rev Sys/Eq. Desc Bldg.
F1 El.
Rm or Rw/Cl Base El. <40'?
Cap.
Demd.
Cap >
Caveats Anchor Interact Equip Cl No Spec.
Spec Demd?
OK7 OK7 OK7 OK7 8
2-1001-198 0
RHR / VALVE, LOOP CROSS TIE RB 580.00 M-12 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-4699-310 0
DG AIR / SOV, START AIR DG 595 00 G-3 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-5201-SOV 0
DG FO / SOV, DAYTANK FILL DG 595.00 G-3 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
EP-1 0
DG HVAC / SOV, N2 BACKUP DG 595.00 N-13 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
EP-2 0
DG HVAC / INLET & OUTLET ISOL DG 595.00 N-13 595.00 Yes BS GRS Yes Yes N/A Yes Yes DAMPER CONTROL 8
EP-3 0
DG HVAC / INLET & OUTLET ISOL DG 595.00 G-3
$95.00 Yes BS GRS Yes Yes N/A Yes Yes DAMPER CONTROL 8
EP-3 (DG-2) 0 DG HVAC / INLET & OUTLET ISOL DG 595.00 595.00 Yes BS GRS Yes Yes N/A Yes Yes DAMPER CONTROL 8
N!A 0
DG HVAC / REGULATOR, DAMPER 1A DG 595 00 N-13 595 00 Yes BS GRS Yes Yes N/A Yes Yes SUPPLY 8
R-1 0
DG HVAC / REGULATOR DG 595.00 N-13 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
R-2 0
DG HVAC / REGULATOR DG 595 00 N-13 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
SV-1 (DG 2) 0 DG HVAC / SOV, DAMPER CONTROL, DG 595.00 G-3 595.00 Yes BS GRS Yes Yes N/A Yes Yes DG-2 9
1-5727 0
DG HVAC / FAN, ROOM VENTILAT DG 595.00 G-23 615.50 N/A ABS CRS No Yes Yes Yes No 9
1/2-5727 0
DG HVAC / FAN. ROOM VENTILAT DG 595.00 N-13 615.50 N/A ABS CRS No Yes Yes Yes No 9
2-5727 0
DG HVAC / FAN, ROOM VENTILAT DG 595.00 G-3 615.50 N/A ABS CRS No Yes Yes Yes No 10 1-5745A 0
RHRSW / COOLER. CUBICLE TB 547.00 C-18 595.00 N/A ABS CRS Yes Yes No Yes No 10 1-5745C 0
RHRSW / COOLER, CUBICLE TB 547.00 C-20 595 00 N/A ABS CRS Yes Yes No Yes No 10 1-5749 0
HVAC / COOLER. CUBICLE TB 547.00 C-22 595 00 N/A ABS CRS Yes Yes No Yes No 10 1-5772-87 0
DG HVAC / DAMPER DG 595 00 G-23 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 10 1-9472-32 0
DG HVAC / DAMPER DG 595.00 G-23 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 10 1-9472-40 0
$95 00 G-23 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 10 1-9472-41 0
DG HVAC / DAMPER DG 595.00 G-23 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 10 1/2-5749 0
HVAC / COOLER, CUBICLE TB 547.00 D-21 595.00 N/A ABS CRS Yes Yes No Yes No 10 1/2-9400-100 0
CR HVAC / AHU TB 615.00 E-24 615.50 N/A ABS CRS Yes Yes Yes Yes Yes Certification:
Certification:
All the information contained on this Screening Verification Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capabihty Engineers regarding systems and operations our knowledge and belief, correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and conclusion (whether verified to be seismically adequate or not).
accurate.
Approved: (Signatures of all Seismic Capabihty Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operations Engineer is required if the Seismic Capabihty are required; there should be at! east two o he SRT. All signatories should agree with all the Engineers deem it necessary.)
entries and conclusions. One signatory sh be a licen pr f ssionalengineer.)
l R.P. Kennedy
/4" p
l[
fl l
l l
Pnnt or Type Name
/
Sgn6ture
/
Date Pnnt or Type Name Signature Date l6,T.I)Tbl l
l l
l 1 D. Stevenson l
Sgnature Date Pnnt or Type Narne Sgnature Date Pnnt or Type Name j
i I
I I
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l Pnnt or Type Name Sqnature Date Pnnt or Type Name Sgnature Date
Ouad Cities Nuclef r Power Station 06/19/96 02 20 PM SCREENING VERIFICATION DATA SHEET (SVDS)
PageC7 EQ Eq ID Rev Sys/Eq. Desc Bldg.
FI El.
Rm or Rw/Cl Base El. <40'?
Cap.
Demd Cap >
Caseats Anchor interact Equip Cl No Spec.
Spec Demd?
OK?
OK7 OK7 OK?
10 2-5745A 0
RHRSW / COOLER. CUBICLE TB 547.00 C-6 595.00 N/A ABS CRS Yes Yes No Yes No 10 2-5745C 0
RHRSW / COOLER, CUBICLE TB 547.00 C-7 595.00 N/A ABS CRS Yes Yes No Yes No 10 2-5746A 0
HVAC / COOLER, CUBICLE RB 554.00 M-8 595 00 N/A ABS CRS Yes No Yes No No 10 2-5748A 0
HVAC / COOLER, CUBICLE RB 554 00 G-8 595 00 N/A ABS CRS Yes No Yes No No 10 2-57488 0
HVAC / COOLER. CUBICLE RB 554.00 G-12 595.00 N/A ABS CRS Yes No Yes No No 10 2-5749 0
HVAC / COOLER. CUBICLE TB 547.00 C-5 595.00 N/A ABS CRS Yes Yes No Yes No 11 1/2-9400-102 0
CR HVAC / UNIT TB 615.00 E-24 615.50 N/A ABS CRS Yes Yes Yes Yes Yes 12 1-4600A 0
DG AIR / TANK. AIR RECEIVER DG 595 00 G-23 595 00 Yes BS GRS Yes Yes Yes Yes Yes 12 1-4600B 0
DG AIR / TANK, AIR RECElVER DG 595.00 G-23 595.00 Yes BS GRS Yes Yes Yes Yes Yes l
12 1-4600C 0
DG AIR / TANK, AIR RECEIVER DG 595 00 G-23 595 00 Yes BS GRS Yes Yes Yes Yes Yes l
12 1-4600D 0
DG AIR / TANK. AIR RECEIVER DG 595 00 rG-2E 595 00 Yes BS GRS Yes Yes Yes Yes Yes 12 1-4609A 0
DG AIR / COMPRESSOR DG 595 00 C-23 595 00 Yes BS GRS Yes Yes Yes Yes Yes 12 1-46098 0
DG AIRICOMPRESSOR DG 595.00 G-23 595.00 Yes BS GRS Yes Yes Yes Yes Yes 12 1 1/2-4600A 0
DG AIR / TANK. AIR RECEIVER DG 595.00 N-13 595.00 Yes BS GRS Yes Yes Yes Yes Yes 12 1 1/2-4600B 0
DG AIR / TANK. AIR RECEIVER DG 595.00 N-13 595 00 Yes BS GRS Yes Yes Yes Yes Yes 12 1/2-4600C 0
DG AIR / TANK, AIR RECEIVER DG 595 00 N-13 595 00 Yes BS GRS Yes Yes Yes Yes Yes 12 1/2-46000 0
DG AIR / TANK, AIR RECEIVER DG 595.00 N-13 595.00 Yes BS GRS Yes Yes Yes Yes Yes 12 1/2-4609A 0
DG AIR / COMPRESSOR DG 595.00 N-13 595.00 Yes BS GRS Yes Yes Yes Yes Yes 12 1/2-4609B 0
DG AIR / COMPRESSOR DG 595 00 N-13 595 00 Yes BS GRS Yes Yes Yes Yes Yes 12 2-4600A 0
DG AIR / TANK. AIR RECEIVER DG 595.00 G-3 595 00 Yes BS GRS Yes Yes Yes Yes Yes 12 2-4600B 0
DG AIR / TANK. AIR RECEIVER DG 595 00 G-3 595 00 Yes BS GRS Yes Yes Yes Yes Yes 12 2-4600C 0
DG AIR / TANK. AIR RECEIVER DG 595 00 G-3 595 00 Yes BS GRS Yes Yes Yes Yes Yes 12 2-4600D 0
DG AIR / TANK. AIR RECEIVER DG 595.00 G-3 595 00 Yes BS GRS Yes Yes Yes Yes Yes 12 2-4609A 0
DG AIR / COMPRESSOR DG 595.00 G-3 595 00 Yes BS GRS Yes Yes Yes Yes Yes 12 2-46098 0
DG AIR / COMPRESSOR DG 595 00 G-3 595 00 Yes BS GRS Yes Yes Yes Yes Yes 15 125 VDC BATT 1 0
125 VDC / BATTERY.125V TB 628 00 G-24 639.00 Yes BS GRS Yes No Yes Yes No 15 125 VDC BATT 2 0
125 VDC / BATTERY,125V TB 628.00 G-2 639.00 Yes BS GRS Yes No Yes Yes No Certification:
Certification:
All the information contained on this Screening Venfication Data Sheet (SVDS) is, to the best of The informstion provided to the Seismic Capability Engineers regarding systems and operations our knowledge and belief. correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, co. rect and conclusion (whether venfied to be seismically adequate or not).
accurate.
Approved: (Sqnatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operations Engineer is required if the Seismic Capabil!ty are required, there should be atleast two on the SRT. All signatories should agree with all the Engineers deem it necessary.)
entries and cor;clusions. One signatory sh be a licen e ro ionalen ineer.)
l kl l
l l
l R.P. Kennedy
,,4r y
aSqnature
/
Dite Pnnt or Type Name Sgnature Date y
v v v
Pnnt or Type Name hl l
l l
l L
l J. D. Stevenson l
~
Sqnature i
"U Date Pnnt or Type Name Sgnature Date Pnnt or Type Name 1
I I
I I
I I
Pnnt or Type Name Sqnature Date Pnnt or Type Name Sgnature Date
_m
.m m _..
_.m Quad Cities Nuclear Power Station 06/19/96 02.20 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page#8 Eq Eq ID Ra~
Sys/Eq. Desc Bldg.
FI El.
Rm or Rw/Cl Base El. <40'?
Cap.
Demt Cap >
Caveats Anchor interact Equip Cl No Spec.
Soca Demd?
OK7 OK?
OK7 OK?
15 250 VDC BATT 1 0
250 VDC / BATTERY. 250V TB 628.00 G-24 639 00 Yes BS GR5 Yes No Yes Yes No 15 250 VOC BATT 2 0
250 VDC / BATTERY,250V TB 628.00 G-1 639.00 Yes BS GRS Yes Unk Yes Yes No 16 1-8300-1 0
125VDC / CHRGR #1,125V TB 615 00 G-24 615.50 N/A ABS CRS Yes Yes Yes No No 16 1-8300-1 A 0
125VDC / CHRGR #1 A,125V TB 615.00 G-24 615.50 N/A ABS CRS Yes Yes Yes No Na 16 1-8350 0
250VDC / CHRGR #1,250V TB 615.00 G-24 615.50 N/A ABS CRS Yes Yes Yes No No 16 1/2-8350 0
250VDC / CHRGR #1/2. 250V TB 615.00 G-24 615.50 N/A ABS CRS Yes Yes Yes Yes Yes 16 2-8300-1 0
125VDC / CHRGR #2,125V TB 615 00 G-1 615.50 N/A ABS CRS Yes Yes Yes No No 16 2-8300-1 A 0
125VDC / CHRGR #2A,125V TB 615 00 G-1 615.50 N/A ABS CRS Yes Yes Yes No No 16 2-8350 0
250VDC / CHRGR #2. 250V TB 615.00 G-1 615 50 N/A ABS CRS Yes Yes Yes No No 17 1-4604 0
DG AIR / DRYER. AIR DG 595.00 G-23 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1-4614 0
X-DG AIR / LUBRICATOR, AIR LINE DG 595.00 G-23 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1-5206 0
X-DG FO / FILTER FUEL OIL DG 595.00 G-23 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1-5207 0
DG FO / STRAINER, FUEL OIL DG 595 00 G-23
$95 00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1-6601 0
DG / DG. UNsT 1 DG 595.00 G-23 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1-6654 0
X-DG CLG / HX, DG LUBE OIL DG 595 00 G-23 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1-6655 0
X-DG FO / FILTER, LUBE OIL DG 595 00 G-23 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1-6661A 0
X-DG CLG / HX. DG COOLING DG 595.00 G-23 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1-6661B 0
X-DG CLG / HX. DG COOLING DG 595.00 G-23 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1-6664 0
X-DG CLG / HEATER, IMMERSION DG 595.00 G-23 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1-6665 0
X-DG CLG / MANIFOLD, ENG CLG DG 595.00 G-23 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1-6668 0
DG INTK / FILTER, INTAKE AIR TB 615.00 G-23 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1/2-4604 0
DG AIR / DRYER. AIR DG 595 00 N-13 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1/2-4614 0
X-DG AIR / LUBRICATOR AIR LINE DG 595 00 N-13 595 00 N/A A.BS CRS Yes Yes Yes Yes Yes 17 1/2-5206 0
X-DG FO / FILTER, FUEL O!L DG 595.00 N-13 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1/2-5207 0
$95.00 N-13 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1/2-6601 0
$95.00 N-13 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1/2-6654 0
X-DG CLG / HX, DG LUBE OIL DG 595.00 N-13 595 00 N/A ABS CRS Yes Yes Yes Yes Yes Cert:ficatm Certification:
All the ir**ns.non cantained on this Screening Verification Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capability Engineers regarding systems and operations our knnWqe and belief, correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and conclusion (whether verified to be seismically adequate or not).
accurate.
Approved: (Signatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operations Engineer is required if the Seismic Capability are required; there should be atleast two on the SRT. All signatories should agree with all the Engineers deem it necessary.)
entries and conclusions. One signatory sho be a license to s ionalengineer.)
l l
R.P. Kennedy j/
l 1
l l
l l
%gnature Date Pnnt or Type Name Sgnature Date
~
V ~ ' ' -
Pnnt or Type Name l
J. D. Stevenson l
buii d )
l MN l l
l l
' Datd Pnnt or Type Name Sgnature Date Pnnt or Type Name Signature I
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Pnnt or Type Name Signature Date Pnnt or Type Name Signature Date
Quad Cities Nucleu Power Station 06/19/96 02:20 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page0 9 Eq Eq ID Rev Sys/Eq. Desc Bldg.
F1 El.
Rm or Rw/Cl Base El.
<40*?
Cap.
Demd.
Cap >
Caveats Anchor interact Equtp CI No Spec.
Spec Demd?
OK?
OK?
OK?
OK7 17 1/2-6655 0
X-DG LO / FILTER, LUBE O L DG 595.00 N-13 59500 N/A ABS CRS Yes Yes Yes Yes Yes 17 1/2-6661A 0
X-DG CLG / HX, DG COOLING DG 595.00 N-13 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1/2-6661B 0
X-DG CLG / HX, DG COOLING DG 595 00 N-13 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1/24664 0
X-DG CLG / HEATER, IMMERSION DG 595.00 N-13 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1/2-6665 0
X-DG CLG / MANIFOLD, ENG CLG DG 595.00 N-13 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1/2-6668 0
DG INTK / FILTER, INTAKE AIR DG 615.00 N-12 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 17 2-4604 0
DG AIR / DRYER, AIR DG 595.00 G-3 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 17 2-4614 0
X-DG AIR / LUBRICATOR, AIR LINE DG 595.00 G-3 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 17 2-5206 0
X-DG FO / FILTER, FUEL OIL DG 595.00 G-3 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 17 2-5207 0
DG FO / STRAINER FUEL OIL DG 595.00 G-3 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 17 2-6601 0
DG / DG. UNIT 2 DG 595.00 G-3 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 17 2-6654 0
X-DG CLG / HX, DG LUBE OIL DG 595 00 G-3 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 17 24655 0
X-DG LO / FILTER, LUBE OIL DG 595 00 G-3 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 17 2-6661A 0
X-DG CLG / HX, DG COOLING DG 595.00 G-3 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 17 2-6661B 0
X-DG CLG / HX, DG COOLING DG 595.00 G-3 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 17 2-6664 0
X-DG CLG / HEATER, IMMERSION DG 595.00 G-3 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 17 2-6665 0
X-DG CLG / MANIFOLD, ENG CLG DG 595 00 G-3 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 17 2-6668 0
DG INTK / FILTER, INTAKE AIR TB 615 00 G-2 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 18 1-1641-6A-PT 0
PR SUPP / TRANSMITTER. PRESS.
RB 623 00 M-15 623.00 Yes BS GRS Yes Yes Yes Yes Yes 18 1-4641-42A 0
DG AIR / SWITCH, PRESS.
DG 595.00 G-23 595.00 N/A ABS CRS Yes No No Yes No 18 1-4641-42B 0
DG AIR / SWITCH, PRESS.
DG 595.00 G-23 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 18 1/2-4641-42A 0
DG AIR / SWITCH, PRESS.
DG 595 00 N-13 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 18 1/2-4641-42B 0
DG AIR / SWITCH. PRESS.
DG 595 00 N-13 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 18 1/2-5741-333 (PT) 0 RHRSW/ TRANS l/P (PT)
TB 615.00 E-24 639.00 Yes BS GRS Yes Yes Yes Yes Yes 18 2-464142A 0
DG AIR / SWITCH. PRESS.
DG 595 00 G-3 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 18 2-4641-42B 0
DG AIR / SWITCH, PRESS.
DG 595.00 G-3 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 18 2201-32 0
RACK / RACK, AUTO BLOWDOWN RB 623 00 K-14 623.00 N/A ABS CRS No
--Ves--
No Yes No Certification:
Certification:
All the information contained on this Screening Venfication Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capabihty Engineers regarding systems and operations our knowledge and belief, correct and accurate. "All information' includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and behef, correct and conclusion (whether venfied to be seismically adequate or not).
accurate.
Approved: (Signatures of all Seismic Capabihty Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operations Engineer is required if the Seismic Capabihty are required, there should be atleast two on the SRT. All signatories should agree with all the Engineers deem it necessary )
entries and conclusions. One signatory sh, be licen pr f sional engineer.)
l R.P. Kennedy
/
M /.
M l
W l
l l
l Pnnt or Type Name
(/ N
'Sqnature Date Prtnt er Type Name Sgnature 1 ate
)
l 21 % l l
l l
l 1 D. Stevenson l
' Datd Pnnt or Type Name Sgnature Date Pnnt or Type Narne Sqnature i
I I
I I
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Pnnt or Type Name Sqnature Date Pnnt or Type Name Sgnature Date i
Quad Cities Nuclesr Power St; tion 06/19/93 02 20 PM SCREENING VERIFICATION NATA SHEET (SVDS)
Page # 10 Eq Eq. ID Rev Sys/Eq. Desc Bldg.
FI El.
Rm or Rw/Cl Base El.
<40'?
Cap.
Demd.
C3p >
Caveats Anchor interact Equip Cl No Spec.
Spec nen.d?
OK?
OK7 OK?
OK7 18 2201-5 0
RACK / RACK RB 623.00 K-14 623.00 Yes BS GRS hes Yes Yes Yes Yes 18 2201-6 0
X-RACK / RACK RB 623 00 L-16 623 00 Yes BS GRS Yes Yes Yes Yes Yes 18 2201-7 0
RACK / RACK RB 595.00 J-14 595.00 Yes BS GRS Yes Yes Yes Yes Yes I
18 2201-708 0
RACK / RACK ATWS INVERTER B SB 595 00 G-25 595.00 N/A ABS CRS Yes No Yes No No I
18 2201-73A 0
RACK / RACK SB 609 00 G-26 609.00 Yes BS GRS Yes Yes Yes Yes Yes l
18 2201-73B 0
RACK / RACK SB 609 00 G-26 609 00 Yes BS GRS Yes Yes Yes Yes Yes 18 2201-8 0
RACK / RACK, AUTO ULO"VDOWN RB 595 00 K-17 595 00 Yes BS GRS Yes Yes Yes Yes Yes 18 2202-32 0
RACK / RACK, AUTO BLOWDOWN RB 623 00 J-12 6^3 00 N". ABS CRS No FM h
Yes No 18 2202-5 0
RACKIRACK RB 623 00 J-11 623.00 f 9-3S GRS Yes Yes Yes Yes Yes
(
18 2202-59A 0
RACK / RAf,K RB 554.00 M-7 554 00 N,..
- JS CRS Yes Yes Yes Yes Yes 18 2202-598 0
RACK / RACK, RHR ROOM INSTR RB 554 00 M-13 554 00 N/A nBS CRS Yes Yes Yes Yes Yes t
18 2202-6 0
X-RACK / RACK RB 623.00 L-11 623 00 Yes BS GRS Yes Yes Yes Yes Yes 18 2202-7 0
RACK / RACK RB 595.00 J-8 595 00 Yes BS GRS Yes Yes Yes Yes Yes 18 2202-70A 0
RACK / RACK. ATWS DIV.1 SB 595 00 E-26 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 18 2202-70B 0
RACK / RACK. ATWS DIV. 2 SB 595 00 E-25 595 00 N/A ABS CRS Yes Yes Yes No No 18 2202-73A 0
RACK / RACK, ANALOG TRIP SB 609.00 F-25 609 00 Yes BS GRS Yes Yes Yes Yes Yes 18 2202-73B 0
RACK / RACK, ANALOG TRIP SB 609.00 F-25 609 00 Yes BS GRS Yes Yes Yes Yes Yes 18 2202-75 0
RACK / RACK RB 554.00 M-12 554 00 N/A ABS CRS Yes Yes Yes Yes Yes 18 2202-76 0
RACKIRACK RB
$54.00 M-7 554 00 N/A ABS CRS Yes Yes Yes Yes Yes 18 2202-8 0
RACK / RACK RB 595.00 K-11 595 00 Yes BS GRS Yes Yes Yes Yes Yes j
18 2212-123 0
RACK / RACK,1/2 DG SYNC RELAY DG 595 00 N-12 615.50 N/A ABS CRS Yes Yes Yes Yes Yes 18 2212-125 0
RACK
18 2212-127 0
RACK / RACK DG 595.00 N-13 615.50 N/A ABS CRS Yes Yes Yes Yes Yg 18 2212-32 0
DG HVAC / Wall mounted Panel DG DG 595 00 N-13 615.50 N/A ABS CRS Yes Yes Yes No No F
I 18 2212-45 0
RACK / RACK DG 595 00 N-13 595 00 N/A ABS CRS Yes Yes Yes Yes Yes I
18 2212-46 0
RACK / RACK,1/2 DG EXCITER DG 595.00 N-13 595 00 N/A ABS CRS Yes Yes Yes Yes Yes Certification:
Certification:
All the information contained on this Screening Venfication Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capability Engineers regarding systems and operations f
our knowledge an i belief, correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and conclusion (whe+her venfied to be seismically adequate or not) accurate.
Approved. (Signatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operations Ercineer is required if the Seismic Capability are required, there should be atleast two on e SRT, All signatories should agree wth all the Engineers deem it necessary.)
entnes and conclusions. One signatory sho a licensed fes o alengineer.)
l R.P. Kennedy l
[., b l
l l
l l
/ U V SRjnature ~
j Date Pnnt or Type Name Sgnature Date Pnnt or Type Name l
2L M l l
l l
l J D. Stevenson l
Sgnature Date Pnnt or Type Name Sgnature Date Pnnt or Type Name l
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l
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Pnnt or Type Name Sgnature Date Pnnt or Type Name Sqnature Date
. = - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
Quad Cities Nuclear Power Station 06/19/93 02.20 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page # 11 i
Eq Eq ID Rev Sys/Eq. Desc Bldg.
F1 EL Rm or Rw/Cl Base EL
<40"?
Cap.
Demd.
Cap >
Caveats Anchor interact Equip l
Cl No Spec.
Spec Demd7 OK7 OK7 OK7 OK7 i
18 2212-50 0
RACK / RACK,1/2 DG AUX CONTROL DG 595.00 N-13 615.50 N/A ABS CRS Yes Yes Yes Yes Yes
[
18 2251-10 0
RACK / RACK, U-1 DG RELAY &
DG 595 00 F-23 595.00 N/A ABS CRS Yes Yes Yes Yes Yes l
METERING 18 2251-112 0
RACK / RACK, DG SYNC RELAY DG 595 00 F-23 615 50 Yes BS GRS Yes Yes Yes Yes Yes 18 2251-113 0
RACK / RACK, DG ENGINE CONTROL DG 595 00 G-23 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 18 2251-12 0
RACK / RACK, DG EXCITER DG 595 00 G-23 595 00 N/A ABS CRS Yes Yes Yes No No l
18 2251-86 0
RACK / RACK TB 639 00 H-14 639 00 Yes BS GRS Yes Yes Yes Yes Yes 18 2251-87 0
RACK / RACK TB 639 00 H-16 639 00 Yes BS GRS Yes Yes Yes Yes Yes 18 2251-97 0
RACK / RACK. DG AUX FD TRANSFER DG 595 00 G-23 595 00 N/A ABS CRS Yes Yes Yes Yes Yes l
18 2251-98 0
RACK / RACK. DG CWP FEED DG 595 00 G-23 595.00 N/A ABS CRS Yes L Yes Yes No No TRANSFER 18 2252-10 0
RACK / RACK, DG RELAY & METERING DG 595.00 G-3 595.00 N/A ABS CRS Yes Yes Yes No No 18 2252-112 0
RACK / RACK, DG SYNC RELAY DG 595.00 F-23 615.50 Yes BS GRS Yes Yes Yes Yes Yes 18 2252-113 0
RACK / RACK, DG ENGINE CONTROL DG 595.00 G-3 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 18 2252-12 0
RACK / RACK, DG EXCITER DG 595 00 G-3 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 18 2252-86 0
RACK / RACK TB 639 00 H-10 639 00 Yes BS GRS Yes Yes Yes Yes Yes 18 2252-87 0
RACK / RACK TB 639.00 H-12 639 00 Yes BS GRS Yes Yes Yes Yes Yes 18 2252-97 0
RACK / RACK, DG AUX FEED DG 595 00 G-3 595 00 N/A ABS CRS Yes Yes Yes Yes Yes j
TRANSFER 18 2252-98 0
RACK / RACK, DG CWP FEED DG 595.00 G-3 595 00 N/A ABS CRS Yes Yes Yes Yes Yes TRANSFER 18 PE-1 0
DG HVAC / SWITCH DG 595 00 N-13 615.50 N/A ABS CRS Yes No Yes Yes No a
19 2-1641-200-TE O
PR INST / TE RB 554.00 K-10 554 00 N/A ABS CRS Yes Yes N/A Yes Yes f
19 2-1641-201-TE O
PR INST / TE RB 554.00 K-10 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 19 2-1641-202-TE O
$54.00 K-10 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 19 2-1641-203-TE O
PR INST / TE RB 554 00 K-10 554.00 N/A ABS CRS Yes Yes N/A Yes Yes 19 2-1641-204-TE O
PR INST / TE RB 554 00 K-10 554 00 N/A ABS CRS Yes Yes N/A Yes Yes Certifcation:
Certification:
i i
All the information contained on this Screening Venfication Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capability Engineers regarding systems and operations our knowledge and belief, correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and conclusion (whether venfied to be seismically adequate or not).
accurate.
Approved. (Signatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operations Engineer is required if the Seismic Capability i
are required, there should be atleast two on e SRT. All signatories should agree with all the Engineers deem it necessary.)
entnes and conclusions. One signatory sho a licensed p fes i al engineer.)
I l
R.P. Kennedy l
//
l 2 #[
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Pnnt or Type Name
/
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/S(nattfre Date Pnnt or Type Name Sqnature Date l
J. D. Stevenson l
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}
Sgnature
' Datd Pnnt or Type Name Sgnature Date i
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Pnnt or Type Name Sgnature R4te Pnnt or Type Name Sgnature Date
~ -
Quad Cities Nucleir Power Station 06/19/96 02.20 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page#12 Eq Eq ID Rev Sys/Eq Desc Bldg Fi El.
Rm or Rw/Ci Base Et <40'?
Cap.
Demd.
Cap >
Caveats Anchor Interact Equip Cl No Spec.
Spec Demd7 OK7 OK7 OK7 OK?
19 2-1641-205-TE O
PR INST / TE RB 554.00 K-10 554.00 N/A ABS CRS Yes Yes N/A Yes Yes 19 2-1641-206-TE O
PR INST / TE RB 554 00 K-10 554.00 N/A ABS CRS Yes Yes N/A Yes Yes 19 2-1641207-TE O
PRINST/TE RB 554 00 K-10 554.00 N/A ABS CRS Yes Yes N/A Yes Yes 19 2-1641-208-TE O
PRINST/TE RB 554 00 K-10 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 19 2-1641-209-TE O
PR INST / TE RB 554.00 K-10 554.00 N/A ABS CRS Yes Yes N/A Yes Yes 19 2-1641-210-TE O
PR INST / TE RB 554 00 K-10 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 19 2-1641-211-TE O
PR INST / TE RB 554.00 K-10 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 19 2-1641-212-TE O
PRINST/TE RB 554 00 K-10 554.00 N/A ABS CRS Yes Yes N/A Yes Yes 19 2-1641-213-T E O
PR INST / TE RB 554.00 K-10 554.00 N/A ABS CRS Yes Yes N/A Yes Yes 19 2-1641-214-TE O
PRINST/TE RB 554.00 K-10 554.00 N/A ABS CRS Yes Yes N/A Yes Yes 19 2-1641-215-TE O
PR INST / TE RB 554.00 K-10 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 20 901-2 0
PANEL / PANEL SB 623 00 623.00 Yes BS GRS Yes Yes Yes No No 20 901-27 0
PANEL / PANEL SB 595 00 G-26 595 00 N/A ABS CRS Yes No Yes No No 20 901-3 0
PANEL / PANEL SB 623.00 623 00 Yes BS GRS Yes Yes Yes Yes Yes 20 901-32 0
PANEL / PANEL SB 595 00 G-26 595.00 N/A ABS CRS Yes No Yes No No 20 901-33 0
PANEL / PANEL SB 595 00 H-26 595.00 N/A ABS CRS Yes No Yes No No 20 901-36 0
PANEL / PANEL SB 623.00 623.00 Yes BS GRS Yes Yes Yes Yes Yes 20 901-39 0
PANEL / PANEL SB 595 00 G-25 595 00 N/A ABS CRS Yes No No Yes No 20 901-46 0
PANEL / PANEL SB 595.00 G-26 595.00 N/A ABS CRS Yes No Yes No No 20 901-47 0
PANEL / PANEL SB 595.00 G-26 595.00 N/A ABS CRS Yes No Yes No No 20 901-48 0
PANEL / PANEL SB 595.00 H-26 595.00 N/A ABS CRS Yes No Yes No No 20 901-49 0
PANEL / PANEL SB 595 00 G-27 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 20 901-5 0
PANEL / PANEL SB 623.00 623 00 Yes BS GRS Yes Yes Yes Yes Yes 20 901-50 0
PANEL / PANEL SB 595 00 G-27 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 20 901-63 0
PANEL / PANEL SB 595.00 H-26 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 20 901-8 0
PANEL / PANEL SB 623.00 623 00 Yes BS GRS Yes Yes Yes Yes Yes 20 902-2 O
PANEL / PANEL SB 623 00 623.00 Yes BS GRS Yes l Yes Yes Yes Yes Certification:
Certification-All the information contained on this Screening Venfication Data Sheet (SVDS) is, to the best of The information provided to the Sersmic Capability Engineers regarding systems and operations our knowledge and belief, correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and conclusien (whether venfied to be seismically adequate or not).
accurate.
Approved: (Signatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operations Engineer is required if the Seismic Capability are required; there should be atleast two on t SRT. All signatories should agree with all the Engineers deem it necessary.)
entnes and conclusions. One signatory sho license rofe ional engineer.)
/f/
_j D M l[ f l
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R. P. Kennedy l ~~
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J D. Stevenson l
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Pnnt or Type Name Sgnature Da.e Pnnt or Type Name Sgnature Date i
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Quad Cities Nucleir Power Statiort 06/19/96 02.20 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page#13 Eq.
Eq ID Rev Sys,Eq%,c Bldg.
F4 El.
Rm er Rw/Cl Base EL
<40"?
Cap.
Demd.
Cap >
Caveats Anchor Interact Equip Cl No Spec.
Spec Demd?
OK7 OK7 OK7 OK7 20 902-3 0
PANELIPANEL SB 623.00 623 00 Yes BS GRS Yes Yes Yes Yes Yes 20 902-32 0
PANEL / PANEL SB 595.00 F-26 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 20 902-33 0
PANEL / PANEL SB 595 00 E-26 595.00 N/A ABS CRS Yes Yes Yes No No 20 902-36 0
PANEL / PANEL SB 623.00 623 00 Yes BS GRS Yes Yes Yes Yes Yes 20 902-39 0
PANEL / PANEL SB 595.00 E-26 595 00 N/A ABS CRS Yes No Yes No No 20 902-46 0
PANEL / PANEL SB 595 00 F-26 595 00 N/A ABS CRS Yes No Yes No No 20 902-47 0
PANEL / PANEL SB 595 00 E-26 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 20 902-48 0
PANEL / PANEL SB 595 00 E-26 595 00 N/A ABS CRS Yes Yes No Yes No 20 902-49 0
PANEL 1 PANEL SB 595 00 F-27 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 20 902-5 0
PANEL / PANEL SB 623.00 623.00 Yes BS GRS Yes Yes Yes Yes Yes
~20 902-50 0
PANEL / PANEL SB 595 00 F-27 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 20 902-63 0
PANEL / PANEL SB 595.00 F-25 595.00 -
N/A ABS CRS Yes Yes Yes Yes Yes 20 902-8 0
PANEL / PANEL SB 623.00 623.00 Yes BS GRS Yes Yes Yes Yes Yes 20 912-8 0
PANEL / PANEL SB 623.00 623 00 Yes BS GRS Yes No No No No 20 NGBR1 0
RACK / Neutral Ground Breaker, Unit 1 DG 595.00 F-23 595.00 N/A ABS CRS Yes Yes Yes Yes Yes 20 NGBR1/2 0
RACK / Neutral Ground Breaker, Unit 1/2 DG 595 00 F-23 595 00 N/A ABS CRS Yes No Yes No No 20 NGBR2 0
RACK / Neutral Ground Breaker, Unit 2 DG 595.00 F-23 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 20 PNL #1 0
PANEL / PANEL RB 623.00 N-15 623 00 N/A ABS CRS No No Yes No No 20 PNL #2 0
125VDC / PANEL RB 623.00 M-9 623 00 N/A ABS CRS No Unk Yes Unk No Certification:
Certification:
All the information contained on this Screening Verification Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capability Engineers regarding systems and operations our knowledge and belief, correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and l
conclusion (whether venfied to be seismically adequate or not).
accurate.
Approved. (Signatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) Approved. (One signature of Systems or Operations Engineer is required if the Seismic Capability are required; there should be atleast two on the SRT. All signatories should agree with all the Engineers deem it necessary.)
entnes and conclusions. One signatory sho a licensed to s 'onalengineer.)
l R. P. Kennedy l
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]
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Date Pnnt or Type Name Sgnature Date
{
Pnnt or Type Name l
J. D. Stevenson l
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]
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Quad Cities Nuclear Power Station 06/19/96 02.33 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page # 14 Eq Eq ID Rev Sys/Eq Desc Bldg.
FI El.
Rm or Rw/Cl Base El.
<40"?
Cap.
Demd.
Cap >
Caveats Anchor Interact Equip Cl No Spec.
Spec Demd?
OK7 OK7 OK?
OK7 5
1-1001-145A 0
RHR I COOLER RB 554 00 M-13 554 00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1-1001-1458 0
RHRSW / COOLER RB 554 00 M-13 554.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1-1001-145C 0
RHR / COOLER RB 554 00 M-18 554 00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1-1001-145D 0
RHRSW / COOLER RB 554.00 M-18 554.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1-1006A 0
X-RHR / SEPARATOR, CYCLONE RB 554 00 M-13 554 00 Yes BS GRS Yes Yes Yes Yes Yes 5
1-1006C 0
X-RHR / SEPARATOR CYCLONE RB 554.00 M-18 554 00 Yes BS GRS Yes Yes Yes Yes Yes 5
2-1001-145C 0
RHR / COOLER RB 554 00 M-12 554 00 N/A ABS CRS Yes Yes Yes Yes Yes 5
2-1001-1450 0
RHR / COOLER RB 554.00 M-12 554.00 N/A ABS CRS Yes Yes Yes Yes Yes l
5 2-1006C 0
X-RHR / SEPARATOR CYCLONE RB 554 00 M-12 554.00 Yes BS GRS Yes Yes Yes Yes Yes 6
1-1002A 0
RHR / PUMP,1 A RB 554 00 M-13 554 00 N/A ABS CRS Yes Yes Yes Yes Yes 6
1-1002C 0
$54 00 M-17 554 00 N/A ABS CRS Yes Yes Yes Yes Yes 6
2-1002C 0
RHR / PUMP, 2C RB 554 00 M-12 554 00 N/A ABS CRS Yes Yes Yes Yes Yes 7
1-0203-1 A 0
MS / VALVE, MSIV (NBOARD RB 592.00 H-15 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-18 0
MS / VALVE, MSIV INBOARD RB 592.00 H-15 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-1 C 0
MS / 'vaM, NislV INBOARD RB 592.00 H-16 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-1 D 0
MS / VALVE, Mb.? 8NBOARD RB 592.00 H-16 595.00 Yes BS GRS Yes Yes N/A Yes Yes s
7 1-0203-2A 0
MS / VALVE, MSIV OUTBOARD RB 592 00 G-15 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-28 0
MS / VALVE, MSIV OUTBOARD RB 592 00 G-15 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-2C 0
MS / VALVE, MSIV OUTBOARD RB 592.00 G-16 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-2D 0
MS / VALVE, MSIV OUTBOARD RB 592 00 G-16 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-3A 0
MS / VALVE, ERV RB 614 00 K-15 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-38 0
MS / VALVE, ERV RB 614 00 K-15 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-3C 0
MS / VALVE, ERV RB 614 00 J-17 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-3D 0
MS / VALVE, ERV RB 614 00 J-16 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-3E O
MS / VALVE, ERV RB 614 00 K-15 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-4A 0
MS / VALVE, SRV RB 614 00 J-15 623 00 Yes BS GRS Yes Yes N/A Yes Yes t
7 1-0203-48 0
MS / VALVE, SRV RB 614 00 J-15 623 00 Yes BS GRS Yes Yes N/A Yes Yes Certification:
Certification:
All the information contained on this Screening Venfication Data Sheet (SVDS) is, to the best of The information provided to the Sersmic Capabihty Engineers regarding systems and operations our knowledge and behef, correct and accurate. "All information" includes each entry.2nd of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and conclusion (whether venfied to be seismically adequate or not).
accurate.
Approved. (Signatures of all Seismic Capabihty Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operahons Engineer is required if the Seismic Capabihty tre required, there should be atleart two on the SRT, All signatories should agree with all the Engineers deem it necessary.)
entnes and conclusions. One signatory should be a licensed professional engineer.)
l K. Adlon l
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W. Djordjevic l
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Pnnt or Type Name
$gnature D6te Pnnt or Type Name Signature Date i
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Pnnt or Type Name Sqnaturp Date Pnnt or Type Name Sgnature Date i
_~ _. - _ _ _
Quad Cities Nuclear Power Station 06/19/CS 02:33 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page # 15 l
Eq.
Eq ID Rev Sys/Eg Desc Bldg.
FI Ei.
Rm or Rw/Cl Base El
<40'?
Cap.
Demd.
Cap >
Caveats Anchor interact Equip C1 No Spec.
Spec Demd?
OK?
OK?
OK?
OK?
7 1-0203-4C 0
MS / VALVE, SRV RB 614 00 J-17 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-4D 0
MS / VALVE, SRV RB 614 00 J-16 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-4E O
MS / VALVE, SRV RB 614 00 J-15 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-4F 0
MS / VALVE, SRV RB 614.00 J-15 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-4G 0
MS / VALVE, SRV RB 614 00 J-17 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-4H 0
MS / VALVE, SRV RB 614 00 J-16 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0302-22A 0
CRD / VALVE, NO. BNK SDV DRN RB 595.00 H-14 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0302-22B 0
CRD / VALVE, NO. BNK SDV DRN RB 595.00 H-14 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0302-22C 0
CRD / VALVE, SO. BNK SDV DRN RB
$95.00 H-14 595 00 Yes BS GRS Yes Yes N/A Yes Yes J
1-0302-220 0
CRD / VALVE, SO. BNK SDV DRN RB 595 00 J-18 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-1001-125A 0
RHR / VALVE, RELIEF RB 554 00 M-13 554 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-1001-125C 0
RHR / VALVE, RELIEF RB 554 00 M-17 554 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-1001-165A 0
RHRSW / VALVE, RELIEF RB 554 00 M-13 554 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-1001-1658 0
RHRSW / VALVE, RELIEF RB 580.00 M-18 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-1001-166A 0
RHR / VALVE, RELIEF RB 554 00 M-13 554.00 Yes BS GRS Yes Yes N/A Yes Yes 7
1 1001-1668 0
RHR / VALVE, RELIEF RB 554 00 M-18 554 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-1001-22A 0
RHR / VALVE, RELIEF RB 554.00 L-13 554.00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-1001-22B O
$54 00 L-18 554 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-1001-59 0
RHR / VALVE, RELIEF RB 591 00 G-16 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-1001-68A 0
RHR / VALVE, CHECK, TESTABLE RB 579 00 K-15 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-1001-68B 0
RHR / VALVE, CHECK, TESTABLE RB 579 00 K-17 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-2301-7 0
HPCI / VALVE, CHECK RB 591.00 G-16 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1/2-5741-341 A 0
CR HVAC / VALVE, CONTROL,1-3/8 TB 615 00 E-24 615 50 Yes BS GRS Yes Yes N/A Yes Yes 7
1/2-5741-341B 0
CR HVAC / VALVE, CONTROL,1-3/8 TB 615 00 E-24 615 50 Yes BS GRS Yes Yes N/A Yes Yes 7
1/2-5741-341C 0
CR HVAC / VALVE. CONTROL,1-3/8 TB 615.00 E-24 615 50 Yes BS GRS Yes Yes N/A Yes Yes 7
1/2-5741-341D 0
CR HVAC / VALVE, CONTROL,1-3/8 TB 615.00 E-24 615 50 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-0002AH25 0
MS / VALVE, PNUEMATIC RB 592.00 msrv 595 00 Yes BS GRS Yes Yes N/A Yes Yes Certification.
Certification:
All the information contained on this Screening Venfication Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capabihty Engineers regarding systems and operations our knowledge and behef, correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, " the best of our knowledge and behef, correct and conclusion (whether venfied to be seismcally adequate or not).
accurate.
Approved. (Signatures of a!I Seismic Capabihty Engineers on the Seismic Review Team (SRT) Approved: (One sgnature of Systems or Operations Engineer is required if the Seismic Capabihty are required, there should be atleast two on the SRT, All signatones should agree with all the Engineers deem it necessary.)
cntnes and conclusions. One signatory should be a heensed professional engineer.)
l K. Adlon l
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W. Djordjevic l
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Pnnt or Type Name Sgnatute
' Dan 6 Pnnt or Type Name Sgnature Date i
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Pnnt or Type Name Sqnature Date Pnnt or Type Name Sqnature Date I
l Quad Cities Nuclear Power Station j
06/19/96 02.33 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page#16 Eq Eq ID Rev Sys/Eq Desc Badg FI El.
Rm or Rw/Ci Base El.
<40'?
Cap.
Demd.
Cap >
Caveats Anchor Interact Equip Cl No Spec.
Spec Demd?
OK?
OK7 OK?
OK?
7 2-0203-0002BH25 0
MS / VALVE, PNUEMATIC RB 592 00 msiv 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-0002CH25 0
MS / VALVE, PNUEMATIC RB 592 00 msiv 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-0002DH25 0
MS / VALVE, PNUEMATIC RB 592.00 mssv 595 00 Yes BS GRS Yes Yes N/A Yes Yes I
7 2-0203-2A 0
MS / VALVE, MSIV OUTBOARD RB 592.00 G-9 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-2B 0
MS / VALVE, MSIV OUTBOARD RB 592.00 G-9 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-2C 0
MS / VALVE, MSIV OUTBOARD RB 592.00 G-10
$95 00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-2D 0
MS / VALVE, MSIV OUTBOARD RB 592 00 G-10 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-1001-125C 0
RHR / VALVE RELIEF RB 554.00 M-11 554 00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-1001-165A O
$54 00 M-7 554 00 Yes BS G_RS Yes Yes N/A Yes Yes 7
2 1001-1658 0
$54 00 M-12 554.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-1001-166B 0
RHR / VALVE, RELIEF RB 554.00 M-12 554.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-1001-59 0
RHR / VALVE, RELIEF RS 591.00 G-10 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-2301-7 0
$91.00 G-16 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0203-1 A-1 0
MS / SOV, FOR MSIV RB 592.00 H-15 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0203-1 A-2 0
MS / SOV, FOR MSIV RB 592.00 H-15 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0203-18-1 0
MS / SOV, FOR MSIV RB 592.00 H-15 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0203-1 B-2 0
MS / SOV. FOR MSIV RB 592.00 H-15 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0203-1C-1 0
MS / SOV, FOR MStV RB 592.00 H-16 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0203-1C-2 0
MS / SOV, FOR MSIV RB 592 00 H-16 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0203-1D-1 0
MS / SOV, FOR MSIV RB 592.00 H-16 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0203-1 D-2 0
$92.00 H-16 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0203-2A-1 0
MS / SOV, FOR MSIV RB 592.00 G-15 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0203-2A-2 0
MS / SOV, FOR MSIV RB 592.00 G-15 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0203-28-1 0
MS / SOV, FOR MSIV RB 592.00 G-15 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0203-28-2 0
MS / SOV, FOR MStV RB 592.00 G-15 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0203-2C-1 0
$92.00 G-16 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0203-2C-2 0
MS / SOV, FOR MSIV RB 592.00 G-16 595 00 Yes BS GRS Yes Yes N/A Yes Yes Certification:
Certification:
All the information contained on this Screening Venficahon Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capabihty Engineers regarding systems and operations our knowledge and belief, correct and accurate. *All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and conclusion (whether venfied to be seismically adequate or not) accurate.
Approved: (Signatures of all Seismic Capabahty Engineers on the Seismic Review Team (SRT) Approved- (One signature of Systems or Operations Engineer is required if the Seismic Capabihty tre required; there should be atleast two on the SRT. All signatories should agree with all the Engineers deem it necessary.)
ent-ies and conclusions. One signatory should be a hcen professional engineer.)
l K. Adlon l
M l [Eb l
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Date Pnnt or Type Name Sgnature Date l
W. Djordjevic l
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' Da(e Pnnt or Type Name Sqnature Date l
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}
l Quad Cities Nuclear Power Station 06/19/96 02:33 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page#17 s
Eq Eq ID Rev Sys/Eq Desc Bldg FI El.
Rm or Rw/Cl Base El.
<4 0'?
Cap.
Demd.
Cap >
Caveats Anchor interact Equip Cl No Spec.
Spec Demd?
OK7 OK7 OK?
OK7 8
1-0203-2D-1 0
MS / SOV, FOR MSIV RB 592.00 G-16 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0203-2D-2 0
MS / SOV, FOR MSIV RB 592.00 G-16 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0302-19A 0
CRD / SOV, B/U SCRAM RB 595.00 L-17 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0302-198 0
CRD / SOV, B/U SCRAM RB 595 00 L-17 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0302-20A 0
CRD / SOV, SDV VNT & DRN RB 595 00 L-17 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-0302-200 0
CRD / SOV, SDV VNT & DRN RB 595 00 L-17 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-1001-16A 0
RHR / VALVE, HX 3A BYPASS RB 554.00 M-13 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-1001-168 0
RHR / VALVE, HX 3B BYPASS RB 554 00 M-18 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-1001-18A 0
RHR / VALVE, LOOP A MIN FLOW RB 554 00 L-14 554 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-1001-188 0
RHR / VALVE, LOOP B MIN FLOW RB 554 00 L-17 554 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-1001-19A 0
RHR / VALVE, LOOP A CROSS TIE RB 580 00 M-14 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-1001-198 0
RHR / VALVE, LOOP B CROSS TIE RB 580 00 M-18 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-1001-28A 0
RHR / VALVE, LOOP A OUTBRD RB 554 00 J-13 554 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-1001-288 0
RHR / VALVE, LOOP B OUTBRD RB 554.00 K-18 554 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-1001-29A 0
RHR / VALVE, LOOP A INBRD RB 554 00 J-14 554 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-1001-298 0
RHR / VALVE, LOOP B INBRD RB 554 00 K-17 554 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-1001-36A 0
RHR / VALVE, TORUS CLG RB 554 00 L-14 554 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-1001-36B 0
RHR / VALVE, TORUS CLG RB 554 00 L-17 554 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-1001-5A O
RHRSW / VALVE, HX DISCH RB 554 00 M-13 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-1001-58 0
$54 00 M-18 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-1201-2 0
RWCU / VALVE RB 623.00 K-17 623 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-1201-5 0
RWCU / VALVE RB 623 00 K-18 623 00 Yes BS GRS Yes Unk N/A Unk Unk 8
1-2301-4 0
HPCI / VALVE, ISOLATION RB 592 00 AZO 95 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-2399-40 0
HPCI / VALVE, ISOLATION RB 554 00 G-14 554 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1-2399-41 0
HPCI / VALVE, ISOLATION RB 554 00 G-14 554.00 Yes BS GRS Yes Yes N/A Yes Yes 0
1/2-5741-326 0
CR HVAC / SOV SB 623 00 G-27 623 00 Yes BS GRS Yes Yes N/A Yes Yes 8
1/2-5741-330 0
CR HVAC / SOV TB 615.00 E-24 615 50 Yes BS GRS Yes Yes N/A Yes Yes 1
Certification:
Certification-All the information contained on this Screening Venfication Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capability Engineers regarding systems and operations our knowledge and belief, correct and accurate. "All informat!on" includes each entry and of the equipment contained in the SVDS is, to the best of our knowiedge and behef, correct and conclusion (whether venfed to be seismically adequate or not).
accurate.
j Approved: (Signatures of all Seismic Capabihty Engineers on the Seismic Review Team (SRT) Approved. (One signature of Systems or Operations Engineer is required if the Seismic Capability are required, there should be atteast two on the SRT. All sgnatories should agree with all the Engineers deem it necessary.)
cntnes and conclusions. One signatory should be a l'
. of ssionalengineer.)
l K. Adlon l
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W. Djordjevic l
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Quad Cities Nuclear Power Station 06/19/96 02 33 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page #18 Eq Eq ID Rev Sys/Eq Desc Bldg.
FI El Rm or Rw/Cl Base El.
<40'?
Cap.
Demd Cap >
Caveats Anchor interact Equip Cl No Spec.
Spec Demd?
OK?
OK?
OK7 OK?
U 2-0203-2A-1 0
MS / SOV. FOR MSIV RB 592.00 G-9 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-0203-2A-2 0
MS / SOV. FOR MSIV RB 592 00 G-9 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-0203-2B-1 0
MS / SOV, FOR MSIV RB 592 00 G-9 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-0203-28-2 0
MS / SOV, FOR MStV RB 592.00 G-9 595 00 Yes BS GRS Yes Yes N/A Yes Yes 6
2-0203-2C-1 0
MS / SOV FOR MSIV RB 592.00 G-10 595 00 Yes BS GRS Yes Yes N/A Yes Yes 0
2-0203-2C-2 0
MS / SOV, FOR MSIV RB 592 00 G-10 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-0203-2D-1 0
MS / SOV FOR MSIV RB 592 00 G-10 595 00 Yes BS GRS Yes Yes N/A Yes Yes C
2-0203-2D-2 0
MS / SOV, FOR MSlV RB 592.00 G-10 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-0302-19B O
CRD / SOV, B/U SCRAM RB 595 00 L-8 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-1001-168 0
RHR / VALVE, HX 2B BYPASS RB 554 00 M-12 595 00 Yes BS GRS Yes Yes N/A Yes Yes O
2-1001-18A 0
RHR / VALVE. LOOP A MIN FLOW RB 554 00 L-8 554 00 Yes BS GRS Yes Yes N/A Yes Yes u
2-1001-180 0
RHR / VALVE, LOOP B MIN FLOW RB 554.00 L-11 554 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-1001-28A 0
RHR / VALVE. LOOP A LPCI INJ RB 554 00 J-7 554 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-1001-288 0
RHR / VALVE, LOOP B OUTBRD RB 554 00 J-12 554 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-1001-29A 0
RHR / VALVE, LOOP A LPCI INJ RB 554 00 J-8 554 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-1001-298 0
RHR / VALVE, LOOP B INBRD RB 554.00 K-11 554 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-1001-36A 0
RHR / VALVE, TORUS CLG RB 554 00 L-8 554 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-1001-36B 0
RHR / VALVE. TORUS CLG RB 554 00 L-11 554 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-1001-5A 0
RHRSW / VALVE RHR HX 2A DISCH RB 554 00 M-7 595 00 Yes BS GRS Yes Yes N/A Yes Yes ti 2-1001-58 0
RHRSW / VALVE, HX DISCH RB 554 00 M-12 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-1201-5 0
RWCU / VALVE RB 623 00 K-8 623 00 Yes BS GRS Yes Unk N/A Unk Unk 8
SV-1 (DG 1) 0 DG HVAC / SOV, DAMPER CONTROL, DG 595 00 N-13 595 00 Yes BS GRS Yes Yes N!A Yes Yes DG 1 10 1-5746A 0
HVAC / COOLER, CUBICLE RB 580 00 N-14 595 00 N/A ABS CRS No No Yes No No 10 1-57468 0
HVAC / COOLER. CUBICLE RB 580 00 M-18 595 00 N/A ABS CRS No No Yes No No 10 1-5747 0
HVAC / COOLER, CUBICLE TB 554 00 G-14 595 00 N/A ABS CRS Yes No Yes No No 10 1-5748A 0
HVAC / COOLER, CUBICLE RB 560 00 G-18 595 00 N/A ABS CRS Yes No Yes No No Certification:
Certification:
All the informaton contained on this Screening Venficaten Data Sheet (SVDS) is, to the best of The informaton provided to the Seismic Capability Engineers regarding systems and operatens our knowledge and belef, correct and accurate. "All informaton" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belef, correct and conclusion (whether venfied to be seismically adequate or not).
accurate.
Approved (Sgnatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operatens Engineer is required if the Seismic Capability tre required, there should be atleast two on the SRT. All signatories should agree with all the Engineers deem it necessary.)
entnes and conclusions One signatory should be a licer's professional engineer.)
l K. Adlon l
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/
Sgnaturp Date Pnnt or Type Name Sgnature Date l
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' bate' Pnnt or Type Name Sgnature Date l
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-. - ~. -. ~..
Quad Cities Nuclear Power Station 06/19/96 02.33 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page # 19 Eq Eq. ID Rev Sys/Eq Desc Bkig.
FI El.
Rm or Rw/Cl Base El.
<40'?
Cap.
Demd Cap >
Caveats Anchor interact Equip Cl No Spec.
Spec Demd?
OK?
OK7 OK?
OK7 10 1-57488 0
HVAC / COOLER, CUBICLE RB 560.00 M-13 595 00 N/A ABS CRS Yes No Yes No No 10 1-9472-35 0
DG HVAC / DAMPER DG 595 00 G-23 615.50 Yes BS GRS Yes Yes Yes Yes Yes 10 1/2-5741-326E O
CR HVAC / DAMPER SB 623 00 G-26 639 00 Yes BS GRS Yes Yes Yes Yes Yes 10 1/2-5741-330A 0
CR HVAC / DAMPER TB 615 00 E-24 639 00 Yes BS GRS Yes Yes Yes Yes Yes 10 1/2-5741-330B 0
CR HVAC / DAMPER TB 613 00 E-24 639 00 Yes BS GRS Yes Yes Yes Yes Yes 10 1/2-5772-88 0
DG HVAC / DAMPER DG 595 00 N-13 615 50 N/A ABS CRS Yes Yes Yes Yes Yes 10 2-57468 0
HVAC / COOLER. CUBICLE RB 554.00 M-13 595 00 N/A ABS CRS Yes No Yes No No 10 2-5747 0
HVAC / COOLER. CUBICLE TB 554 00 G-12 595 00 N/A ABS CRS Yes No Yes No No 10 2-9472-32 0
DG HVAC / DAMPER DG 595 00 G-3 615 50 Yes BS GRS Yes No No Yes No 10 2-9472-35 0
DG HVAC / DAMPER DG 595 00 G-3 615 50 Yes BS GRS Yes Yes Yes Yes Yes 10 2-9472-40 0
DG HVAC / DAMPER DG 595.00 G-3 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 10 2-9472-41 0
DG HVAC / DAMPER DG 595 00 G-3 595 00 N/A ABS CRS Yes Yes Yes Yes Yes 17 1-6667 0
DG EXH / SILENCER. EXHAUST TB 639 00 G-24 639 00 N/A ABS CRS No No 17 1/2-6667 0
DG EXH I SILENCER, EXHAUST DG 615 00 N-12 623 00 N/A ABS CRS No No
~No Yes No N1 vn No 17 2-6667 0
DG EXH / SILENCER, EXHAUST TB 639 00 G-2 639 00 N/A ABS CRS No NJ C
Yes No 18 1-1641-5A-PT 0
PR SUPP / TRANSMITTER, PRESS.
RB 585 00 L-13 595 00 Yes BS GRS Yes Yes Ya s Yes Yes 18 1-1641-58-PT 0
PR SUPP / TRANSMITTER PRESS.
RB 585 00 L-13 595 00 Yes BS GRS Yes Yes Yos Yes Yes 18 2-1462-A-PS 0
MS / SWITCH, PRESSURE RB 554 00 G-7 595 00 Yes BS GRS Yes Yes
'. es Yes Yes 18 2-1462-8-PS O
MS / SWITCH, PRESSURE RB 554.00 G-12 554 00 Yes BS GRS Yes Yes Yes Yes Yes 18 2201-59A 0
RACK / RACK RB 554 00 M-14 554 00 N/A ABS CRS Yes Yes Yes Yes Yes 18 2201-59B 0
RACK / RACK RB 554.00 M-18 554 00 N/A ABS CRS Yes Yes I Yes Yes Yes 18 2201-75 0
X-RACK / RACK RB
$54.00 M-18 554 00 N/A ABS CRS Yes Yes Yes Yes Yes 18 2251-100 0
RACK / RACK,1/2 DG CLG AUX FD TB 547.00 C-19 547.00 N/A ABS CRS Yes Yes Yes Yes Yes XFER 19 1-1046A-TE O
RHR/TE RB 554 00 M-13 595 00 Yes BS GRS Yes Yes N/A Yes Yes 19 1-1046B-TE O
RHR/TE RB 554.00 M-18 595 00 Yes BS GRS Yes Yes N/A Yes Yes 19 1-1047A-TE O
RHR/TE RB 554 00 M-13 595 00 Yes BS GRS Yes Yes N/A Yes Yes Certification:
Certification:
All the information contained on this Screening Verification Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capabihty Engineers regarding systems and operations our knowledge and belef correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belef, correct and conclusion (whether venfied to be seismically adequate or not).
accurate.
Approved: (Signatures of att Seismic Capabihty Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operations Engineer is required if the Seismic Capabihty are requirsd; there should be atleast two on the SRT. All signatories should agree with all the Engineers deem it necessary.)
entnes and conclusions. One signatory should be a heen rofessional engineer.)
l K. Adlon l
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Pnnt or Type Name Signgtur D te' Pnnt or Type Name Signature Date l
W. Djordjevic l
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Pnnt or Type Name
$1 re
/ D(te Pnnt or Type Name Signature Date i
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.____--m Quad Cities Nuclear Power Station 0S/19/96 02.33 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page # 20 Eq Eq ID Rev Sys/Eg Desc Bldg.
FI El.
Rm or Rw/CI Base El. <40'?
Cap.
Demd.
Cap >
Caveats Anchor interact Eque Cl No Spec.
Spec Demd?
OK?
OK?
OK7 OK7 19 1-10478-TE O
RHR/TE RB 554 00 M-18 595 00 Yes BS GRS Yes Yes N/A Yes Yes 19 1-1641-200-TE O
PRINST/TE RB 554 00 K-16 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 19 1-1641-201-TE O
PRINST/TE RB 554.00 K-16 554.00 N/A ABS CRS Yes Yes N/A Yes Yes 19 1-1641-202-TE O
PR INST / TE RB 554 00 K-16 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 19 1-1641-203-TE O
PR INST / TE RB 554 00 K-16 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 19 1-1641-204-TE O
PR INST / TE RB 554.00 K-16 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 19 1-1641-205-TE O
PR INST / TE RB 554.00 K-16 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 19 1-1641-206-TE O
PR INST / TE RB 554.00 K-16 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 19 1-1641-207-TE O
PR INST / TE RB 554.00 K-16 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 19 1-1641-208-TE O
PR INST / TE RB 554 00 K-16 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 19 1-1641-209-TE O
PRINST/TE RB 554 00 K-16 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 19 1-1641-210-TE O
PR INST / TE RB 554 00 K-16 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 19 1-1641-211-TE O
PR INST / TE RB 554 00 K-16 554.00 N/A ABS CRS Yes Yes N/A Yes Yes 19 1-1641-212 TE O
PR INST / TE RB 554 00 K-16 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 19 1-1641-213-TE O
PR INST / TE RB 554 00 K-16 554 00 N/A ABS CRS Yes Yes N/A Yes Yes 19 1-1641-214-TE O
PR INST / TE RB 554.00 K-16 554.00 N/A ABS CRS Yes Yes N/A Yes Yes 19 1-1641-215-TE O
PR INST / TE RB 554 00 K-16 554.00 N/A ABS CRS Yes Yes N/A Yes Yes 19 2-1046A-TE O
RHR/TE RB 554.00 M-7 595 00 Yes BS GRS Yes Yes N/A Yes Yes 19 2-10468-TE O
RHR/TE RB 554 00 M-12 595 00 Yes BS GRS Yes Yes N/A Yes Yes 19 2-1047A-TE O
RHR / TE RB 554 00 M-7 595 00 Yes BS GRS Yes Yes N/A Yes Yes 19 2-10478-TE O
RHR/TE RB 554.00 M-12 595 00 Yes BS GRS Yes Yes N/A Yes Yes Certification:
Certification.
All the information contained on this Screening Venfication Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capabihty Engineers regarding systems and operations our knowledge and behef, correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and behef, correct and conclusion (whether verifed to be seismically adequate or not).
accurate.
Approved. (Signatures of all Seismic Capabihty Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operations Engineer is required if the Seismic Capability are required, there should be atleast two on the SRT. All signatories should agree with all the Engineers deem it necessary )
entnes and conclusions. One signatory should be a licen, to essional engineer.)
l K. Adlon l-9 l
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Pnnt or Type Name Sgna re D te Pnnt or Type Name Sgnature Date l
W. Djordjevic l
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Pnnt or Type Name "SighM e
Datd Pnnt or Type Name Sgnature Date l
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Pnnt or Type Name Sgnature Date Pnnt or Type Name Sgnature Date
Quad Cities Nuclear Power Station OEf1496 3.17 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page # 14 Eq Eq. ID Rev Sys/Eq. Desc Bldg FI EL Rm or Rw!Cl Base EL
<40'?
Cap.
Demd.
Cap >
Caveats Anchor Interact Equip Cl No Spec.
Spec Demd?
OK7 OK7 OK7 OK7 5
1-1001-145A 0
RHR / COOLER RB 554 00 M-13 554 00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1-1001-145B 0
RHRSW / COOLER RB 554.00 M-13 554 00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1-1001-145C 0
RHR / COOLER RB 554.00 M-18 554.00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1-1001-145D 0
RHRSW / COOLER RB 554.00 M-18 554 00 N/A ABS CRS Yes Yes Yes Yes Yes 5
1-1006A 0
X-RHR / SEPARATOR, CYCLONE RB 554 00 M-13 554 00 Yes BS GRS Yes Yes Yac Yes Yes 5
1-1006C 0
X-RHR / SEPARATOR, CYCLONE RB 554 00 M-18 554.00 Yes BS GRS Yes Yes Yes Yes Yes 5
2-1001-145C 0
RHR / COOLER RB 554 00 M-12 554 00 N/A ABS CRS Yes Yes Yes Yes Yes 5
2-1001-1450 0
RHR / COOLER RB 554.00 M-12 554 00 N/A ABS CRS Yes Yes Yes Yes Yes 5
2-1006C 0
X-RHR / SEPARATOR, CYCLONE RB 554 00 M-12 554 00 Yes BS GRS Yes Yes Yes Yes Yes 6
1-1002A 0
RHR / PUMP,1 A RB 554 00 M-13 554 00 N/A ABS CRS Yes Yes Yes Yes Yes 6
1-1002C 0
RHR / PUMP,1C RB 554.00 M-17 554 00 N/A ABS CRS Yes Yes Yes Yes Yes 6
2-1002C 0
RHR / PUMP,2C RB 554 00 M-12 554.00 N/A ABS CRS Yes Yes Yes Yes Yes 7
1-0203-1 A 0
MS / VALVE, MSIV INBOARD RB 592.00 H-15 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-1 B 0
MS / VALVE, MSIV INBOARD RB 592.00 H-15 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-1 C 0
MS / VALVE, MSIV INBOARD RB 592.00 H-16 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-1 D 0
MS / VALVE. MSIV INBOARD RB 592.00 H-16 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-2A 0
MS / VALVE, MSIV OUTBOARD RB 592.00 G-15 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-2B 0
MS / VALVE, MSIV OUTBOARD RB 592 00 G-15 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
10203-2C 0
MS / VALVE, MStV OUTBOARD RB 592.00 G-16 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-20 0
MS / VALVE, MSIV OUTBOARD RB 592.00 G-16 595 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-3A 0
MS / VALVE, ERV RB 614 00 K-15 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-3B 0
MS / VALVE, ERV RB 614 00 K-15 623.00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-3C 0
MS / VALVE, ERV RB 614 00 J-17 623.00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-3D 0
MS / VALVE ERV RB 614 00 J-16 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-3E O
MS / VALVE, ERV RB 614 00 K-15 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-4A 0
MS / VALVE, SRV RB 614 00 J-15 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
1-0203-4B 0
MS / VALVE, SRV RB 614.00 J-15 623 00 Yes BS GPS Yes Yes h/A Yes Yes 7
1-0203-4C 0
MS / VALVE, SRV RB 614 00 J-17 623.00 Yes BS GRS Yes Yes N'A Yes Yes Certification:
Certification:
All the information contained on this Screening Verification Data Sheet (SVDS) is, to the best of our The informahon provided to the Seismic Capability Engineers regarding systems and operations of the knowledge and belief, correct and accurate. "AH informatxm" includes each entry and conclusion equapment contained in the SVDS is, to the best of our knowledge and belief, correct and accurate.
(whether verified to be seismically adequate or not).
Approved: (Signatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) are Approved: (One signature of Systems or Operations Engineer is required if the Seismic Capability required, there should be atleast two on the SRT. AB signatories should agree with all the entries and Engineers deem it necessary.)
conclusions. One signatory should be a licensed professional engineer.)
i K. Adlon l
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w Diordievic I
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Print or Type Name Signature Date Print or Type Name Signature Date
Quad Cities Nuclear Power Station CS/19/96 02:37 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page021 Eq Eq. ID Rev Sys/Eq Desc Bldg.
FI El.
Rm or Rw/Cl Base El.
<40"?
Cap.
Demd.
Cap >
Caveats Anchor interact Equip Cl No Spec.
Spec Demd?
OK7 OK7 OK?
OK7 7
2-0203-0001AH25 0
MS / VALVE, PNUEMATIC RB 592.00 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-0001BH25 0
MS / VALVE, PNUEMATIC RB 592.00 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-0001CH25 0
MS / VALVE. PNUEMATIC RB 592.00 595.00 Yes BS GRS Yes Yes N!A Yes Yes 7
2-0203-0001DH25 0
MS / VALVE. PNUEMATIC RB 592.00 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-1 A 0
MS / VALVE, MSIV !NBOARD RB 592.00 A?005 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-1 B 0
MS / VALVE. MSIV INBOARD RB 592.00 AZO 10 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-1C 0
MS / VALVE. MStV INBOARD RB 592 00 AZ350 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-1 D 0
MS / VALVE, MSIV INBOARD RB 592.00 AZ335 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-3A 0
MS / VALVE. ERV RB 620 00 AZO 20 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-38 0
MS / VALVE, ERV RB 620 00 AZO 55 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-3C 0
MS / VALVE. ERV RB 620 00 AZ290 623.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-3D 0
MS / VALVE. ERV RB 620.00 AZ335 623.00 Yes BS GPS Yes Yes N/A Yes Yes 7
2-0203-3E O
MS / VALVE. ERV RB 620.00 AZO 90 623.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-4A 0
MS / VALVE, SRV RB 620.00 AZO 50 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-4B 0
MS / VALVE, SRV RB 620 00 AZ075 623.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-4C 0
MS / VALVE, SRV RB 620 00 AZ305 623.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-4D 0
MS / VALVE, SRV RB 620.00 AZ310 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-4E O
MS / VALVE, SRV RB 620 00 AZO 30 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-4 F 0
MS / VALVE, SRV RB 620 00 AZO 40 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-4G 0
MS / VALVE, SRV RB 620 00 AZ310 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0203-4H 0
MS / VALVE, SRV RB 620 00 AZ320 623 00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0302-22A 0
CRD / VALVE, NO. BNK SDV DRN RB 595 00 J-7 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0302-228 0
CRD / VALVE, NO. BNK SDV DRN RB 595.00 J-7 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0302-22C 0
CRD / VALVE, SO. BNK SDV DRN RB 595 00 J-12 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-0302-220 0
CRD / VALVE, SO. BNK SDV DRN RB 595.00 J-12 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-1001-22A 0
RHR / VALVE, RELIEF RB 554.00 L-7 554.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-1001-22B 0
RHR / VALVE. RELIEF RB 554 00 L-12 554.00 Yes BS GRS Yes Yes N/A Yes Yes Certification:
Certification:
All the information contained on this Screening Verification Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capability Engineers regarding systems and operations our knowledge and belief, correct and accurate. 'All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and conclusion (whether venfied to be seismically adequate or not).
accurate.
Approved. (Signatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) Approved. (One signature of Systems or Operations Engineer is required if the Seismic Capabihty are required; there should be atleast two on the SRT. All signatories should agree with all the Engineers deem it necessary.)
entnes and conclusions. One signatory should be a heensed prefepional engineer.)
l K. Adlon l
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Pnnt or Type Name Sig atu Q
Dat Pnnt or Type Name Signature Date l
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W. Djordjevic l
' $d@ ure
[ fate Pnnt or Type Name Signature Date Pnnt or Type Name
Quad Cities Nucleir Power Station 06/19/96 02.37 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page # 22 Eq Eq ID Rev Sys/Eq. Desc Bldg.
FI El.
Rm or Rw/Cl Base El
<40*?
Cap.
Demd.
Cap >
Caveats Anchor in+eract Equip Cl No Spec.
Spec Demd7 OK?
OK7 OK7 OK7
'7 2-1001-68A 0
RHR / VALVE, CHECK, TESTABLE RG
$89 00 AZO 90 595.00 Yes BS GRS Yes Yes N/A Yes Yes 7
2-1001-68B 0
RHR / VALVE, CHECK. TESTABLE r<B 589 00 AZ270 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-0203-1 A-1 0
MS / SOV. FOR MSIV RB 592.00 AZ005 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-0203-1 A-2 0
$92.00 AZ005 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-0203-18-1 0
MS / SOV. FOR MSIV RB 592.00 AZO 10 595.00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-0203-18-2 0
MS / SOV, FOR MSIV RB 592.00 AZO 10 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-0203-1 C-1 0
MS / SOV, FOR MSIV RB 592.00 AZ350
$95.00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-0203-1 C-2 0
MS / SOV. FOR MSIV RB 592.00 AZ350 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
24203-1 D-1 0
MS / SOV, FOR MSIV RB 592.00 AZ335 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-0203-1 D-2 0
MS / SOV, FOR MSIV RB 592.00 AZ335 595 00 Yes BS GRS Yes Yes N/A Yes Yes 8
2-1201-2 0
RWCU / VALVE RB 623.00 K-9 623 00 Yes BS GRS Yes Yes N/A Yes Yes Certification:
Certification:
All the information contained on this Screening Venfication Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capability Engineers regarding systems and operat:ons our knowledge and belief, correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and beiief, correct and conclusion (whether verified to be seismically adequate or not).
accurate.
Approved. (Signatures of all Seismic Capability Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operations Engineer is required if the Seismic Capabihty are required, there should be atleast two on the SRT. All signatones should agree with all the Engineers deem it necessary )
entries and conclusions. One signatov should be a licensed professional engineer )
l K. Adlon l
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G G Thomas l
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Quad Cities Nuclear Power Station 03/19/95 02:38 PM SCREENING VERIFICATION DATA SHEET (SVDS)
Page # 23 t
Eq.
Eq ID Rev Sys/Eq. Desc Bkjg FI El.
Rm or Rw/Cl Base El. <40'?
Cap.
Demd.
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Caveats Anchor Interact Eque Cl No Spec.
Spec Demd?
OK7 OK7 OK?
OK?
1 MCC 18-4 0
480VAC / MCC TB 615 00 E-24 615.50 N/A ABS CRS Yes Yes Yes Yes Yes 1
MCC 19-1-1 0
208 VAC / MCC RB 623.00 N-18 623.00 N/A ABS CRS No No No Yes No 10 1/2-5772-86 0
DG HVAC / DAMPER DG 595.00 N-13 615.50 N/A ABS CRS Yes Yes Yes Yes Yes 10 1/2-5772-87 0
DG HVAC / DAMPER DG 595.00 N-13 615.50 N/A ABS CRS Yes Yes Yes Yes Yes i
10 2-5772-87 0
$95.00 G-3 615.50 N/A ABS CRS Yes Yes Yes Yes Yes l
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Certification:
Certification:
i All the information contained on this Screening Verification Data Sheet (SVDS) is, to the best of The information provided to the Seismic Capabihty Engineers regarding systems and operations i
our knowledge and belief, correct and accurate. "All information" includes each entry and of the equipment contained in the SVDS is, to the best of our knowledge and belief, correct and conclusion (whether verified to be seismically adequate or not).
accurate.
Approved: (Signatures of all Seismic Capabdity Engineers on the Seismic Review Team (SRT) Approved: (One signature of Systems or Operations Engineer is required if the Seismic Capability are required; there should be atleast two on the SRT. All signatories should agree with all the Engineers deem it necessary.)
l entries and conclusions. One signatory thould be a lice sed professional engineer.)
l R. Janowick l
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I i
1 Programmatic Solutions 085-96-004-L 24 Woodbine Avenue, Suite 12 Northport, New York 11768 June 21,1996 j
Mr. Karl Adlon Commonwealth Edison Comnpany Quad Cities Nuclear Power Station 22710 206th Avenue North Cordova, IL 61242-9740
SUBJECT:
Third Party Audit of USI A Quad Cities Nuclear Power Station Dear Mr. Adlon This letter report summarizes the activities performed and the conclusions drawn from the Third Party Audit of the USI A-46 Implementation program for Quad Cities Units 1 & 2.
1.0 Scope of Third Party Audit The scope of the third party audit covers all US A-46 topics that relate to the capacity of the equipment reviewed and the seismic demand that applies to these items.
Specifically, the following activities were reviewed as a part of this third party audit.
1.1 The Quad Cities Units 1 & 2 Seismic Design Bases Site seismicity e
Seismic input to structures and equipment e
Seismic design of mechanical and electrical equipment e
Seismic design of tanks e
Seismic spatialinteraction issues e
1.2 The USI A-46 Evaluation The overall approach used e
The screening criteria used e
The qualifications of the Seismic Review Team e
Screening walkdown implementation Results of the screening walkdown e
e Outlier resolution Design enhancements 1.3 Control of Work
. Documentation of work
Mr. Karl Adlon June 21,1996 Page 2 Imerface of the seismic activities with the systems activities e
1.4 Conclusions and Results Reasonableness of results e
Appropriateness of recommended resolutions 1.5 Structures, Systems, and Components Included in the Third Party Audit Electrical and mechanical equipment included on the SSEL Tanks and heat exchangers included on the SSEL e
Electrical raceways e
Items which may have potential spatial interaction with SSEL items e
2.0 Summary of Third Party Audit Activities The following summarizes the activities performed for the third party audit of USI A-46 activities for Quad Cities Units 1 & 2.
2.1 Third Party Audit Walkdown of Quad Cities Units 1 & 2 On May 6 and 7,1996 a third party audit walkthrough of the plant was performed by Harry Johnson (Reference 3). The objectives of the walkthrough were the following:
Gain familiarity with the typical design and installation of equipment used at Quad Cities Units 1 & 2.
Review the seismic interaction potential at Quad Cities Units 1 & 2.
Discuss the SRT walkdown activities.
e Review at least one of each equipment classes on the SSEL 2.2 Third Party Audit of Methods Used i
l The third party audit of the methods was based on insights obtained during the third party audit walkdown of Quad Cities Units 1 & 2 and on review of the draft final reports (References 1 and 4).
Observations regarding the methods used were provided in Reference 2.
~
2.3 Third Party Audit of Sample Documentation During the site visit, a sample of the SEWS forms being filled out by the SRT were reviewed to determine if the documentation being developed was appropriate. Selected Programmatic Solutions 516 261 5536 fax: 516 2615732
Mr. Karl Adlon June 21,1996 l
Page 3 l
additional SEWS for items walked down for the peer review were reviewed following the walkdown.
l The documentation of walkdown and results provided in the project report (References 1 and 4) was reviewed. This documentation includes the following items:
Summary of Quad Cities Units 1 & 2 seismic design basis Example SEWS forms e
Seismic Verification Data Sheets (SVDS) e Walkdown Resumes e
Walkdown Comments by SRT Summary of Outlier Recommendations e
Photographs taken during the walkdown e
2.4 Third Party Audit of Results The results of the Quad Cities Units 1 & 2 USI A-46 implementation are provided in the project reports. A third party review of the seismic evaluation report was l
performed on the draft reports (References 1 and 4).
3.0 Results of Third Party Audit 3,1 Methods Used The method used for Quad Cities Units 1 & 2 conforms with the SQUG GIP method. This method has been accepted by the NRC as described in SSER Revision 2.
3.2 Implementation of Program The Quad Cities Units 1 & 2 USI A-46 implementation has been correctly implemented and documented.
3.3 Reasonableness of Results Based on the third party audit walkdown and review of the methods used for the Quad Cities Units 1 & 2 USI A-46 implementation, the results appear to be reasonable and consistent with those anticipated.
Programmatic Solutions 516 261 5536 fax: 516 2615732
Mr. Kerl Adlon i
June 21,1996 Page 4 3.4 Recommended Actions The recommended actions resulting from USI A-46 are documented in the project reports in Sections 4.6 (equipment),6.2 (tanks and heat exchangers), and 7.3 (electrical raceways) of the Seismic Evaluation Report (Reference 4). These recommended actions appear to be reasonable and consistent with those anticipated.
4.0 Conclusions The methods used for the Quad Cities Units 1 & 2 implementation are appropriate, the methods have been correctly implemented, and the results and the recommended actions appear to be reasonable and consistent with those anticipated.
l 5.0 References
)
i 1.
"USNRC USI A-46 Resolution Seismic Evaluation Report Quad Cities Nuclear Station Units 1 and 2", Report 93C2806-03.A46, Draft, April 30,1996.
j 2.
Programmatic Solutions letter 085-96-002-L, dated May 20,1996.
l 3.
Programmatic Solutions letter 085-96-003-L, dated May 23,1996.
4 "USNRC USI A-46 Resolution Seismic Evaluation Report Quad Cities Nuclear Station Units 1 and 2", Report 93C2806-03.A46, Final Report, June 19,1996.
Sincerely, f
a Johnson enior Consultant 1
Programmatic Solutions 516 261 5536 fax: 516 2615732 i
I i
{
Quad Cities A46 Final Repon June 19,1996 l
l l
i Appendix F Quad Cities Nuclear Station Anchor Bolt Tightness and Embedment Checks l
l l
Quad Cities A-46 Final Report June 19,1996 INTRODUCTION This report summarizes the results of the expansion anchor bolt tightness checks performed at the Quad Cities Nuclear Power Plant on June 10,1996. The checks were performed in order to satisfy the SOUG Generic Implementation Procedure (GlP) requirements contained in Appendix C, Section C.2.3.
The tightness check was performed using a standard open-ended wrench or a socket wrench on the bolt head or nut and applying a torque by hand until the bolt or nut is " wrench tight".
The bolt tightness checks were performed on June 10,1996. The checks were performed by Mr.
Mike Wassels (Comed - Electrical Maintenance) and witnessed by Mr. Karl Adlon (Comed - Design Engineering).
The GIP allows for a sampling program to check the tightness of expansion anchors. However, in order to be as complete as possible the tightness checks at Quad Cities were performed on all accessible expansion anchors where the base anchorage may be subjected to a tensile load during an earthquake.
RESULTS OF BOLT TIGHTNESS CHECKS Table F.1 indicates the results of the bolt tightness checks for each item of equipment. When all the bcits visually identified were not tested, the reason for this is given in the results and notes column. All accessible expansion anchors where the base anchorage may be subjected to a tensile load durinc an earthquake were tested.
There were a total of 596 anchors tested out of the 717 expansion anchors visually confirmed for the equipment.
The expansion anchor tightness check found two instances of loose anchors; one could not be tightened and had a gapped anchorage and one could not be tightened and had a loose washer.
These anchors will be disassembled, inspected and repaired as necessary.
GIP Section C.2.3 states that "a small amount of initial rotation (about 1/4 turn) is acceptable provided the nut or bolt will tighten and resist the applied torque. If a bolt turns more than about 1/4 turn, but does eventually resist the torque, it should be re-torqued to the manufacturer's recommended installation torque and then considered acceptable." in seven instances the bolt turned more than about 1/4 turn but less than 1 full turn. Since these are shell-type anchors with bolt heads, these are considered acceptable. In five instances, the bolt or nut turned more than 1 turn but tightened. Again, these were shell type, where the nuts were installed on cut threaded rod. Likewise, these are acceptable. For conservatism, however, these anchors will be disassembled, inspected and repaired as necessary.
The sample could not be described as homogeneous as defined in the GIP.
However with only two non-conforming anchors out of the 596 tested, the general objective of meeting the 95/5 criterion is satisfied.
F-1
Table F.1 Expansion Anchor Tightness Check Results TAG DESCRIPTION BLD ELEV LOCATION NUMBER NUMBER RESULTS AND NOTES BOLTS CHECKED 2202-32 RACK, AUTO BLOWDOWN RB 623 12-J 4
4 ALL TIGHT.
2201-32 RACK, AUTO BLOWDOWN RB 623 14-K 4
4 ALL TIGHT.
2201-70A RACK, ATWS INVERTER B SB 595 A.E.R.
14 7
ALL TIGHT. PANEL FRONT DOES NOT OPEN - ANCHORS INACCESSIBLE.
2201-70B RACK, ATWS INVERTER B SB 595 A.E.R.
14 7
ALL TIGHT. PANFL FRONT DOES NOT OPEN - ANCHORS INACCESSIBLE.
2202-70A RACK, ATWS DIV.1 SB 595 A.E.R.
14 7
ALL TIGHT. PANEL FRONT DOES NOT OPEN - ANCHORS INACCESSIBLE.
2202-70B RACK, ATWS DIV. 2 SB 595 A.E.R.
14 7
ALL TIGHT. PANEL FRONT DOES NOT OPEN - ANCHORS INACCESSIBLE.
901-27 PANEL SB 595 A.E.R.
6 5
ALL TIGHT. REAR OF CENTER CUBICLE DOES NOT OPEN.
901-32 PANEL SB 595 A.E.R.
4 4
ALL TIGHT.
901-33 PANEL SB 595 A.E.R.
4 4
ALL TIGHT.
901-39 PANEL SB 595 A.E.R.
4 4
3 TIGHT.1 LOOSE, WAS TIGHTENED - WILL BE REMOVED, INSPECTED AND TOROUED.
(ACTION REQUEST 960043386).
F-2
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Table F.1 Expansion Anchor Tightness Check Results 901-46 PANEL SB 595 A.E.R.
4 4
ALL TIGHT.
901-47 PANEL SB 595 A.E.R.
4 4
ALL TIGHT.
901-48 PANEL SB 595 A.E.R.
4 3
ALL TIGHT. WIRES ABOVE 4TH BOLT.
901-49 PANEL SB 595 A.E.R.
4 2
ALL TIGHT. WALL BEHIND PANEL.
901-50 PANEL SB 595 A.E.R.
4 2
1 TIGHT.1 SEVERAL TURNS LOOSE, WAS TIGHTENED - WILL BE REMOVED, INSPECTED AND TOROUED. (ACTION REQUEST 960043390). WALL BEHIND PANEL.
901-63 PANEL, UPS SB 595 A.E.R.
18 0
UPS HIGH VOLTAGE EQUIPMENT INACCESSIBLE.
902-27 PANEL, RPIS SB 595 A.E.R.
4 4
ALL TIGHT.
902-32 PANEL SB 595 A.E.R.
4 4
2 TIGHT. 2 TURNED 1 % TO 1 %
TURNS TO TIGHT - WILL BE REMOVED, INSPECTED AND TORQUED. (ACTION REQUEST 960043397).
902-33 PANEL SB 595 A.E.R.
4 4
ALL TIGHT.
902-39 PANEL SB 595 A.E.R.
4 4
ALL TIGHT.
I 902-46 PANEL SB 595 A.E.R.
4 4
ALL TIGHT.
902-47 PANEL SB 595 A.E.R.
4 4
ALL TIGHT.
F-3
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Table F.1 Expansion Anchor Tightness Check Results 902-48 PANEL SB 595 A.E.R.
4 4
ALL TIGHT, BUT WASHER LOOSE UNDER 1 - WILL BE REMOVED, INSPECTED AND TOROUED.
(ACTION REQUEST 960043400).
902-49 PANEL SB 595 A.E.R.
4 2
1 TIGHT.1 SEVERAL TURNS TO TIGHT - WILL BE REMOVED, INSPECTED AND TORQUED.
(ACTION REQUEST 960043395).
REAR INACCESSIBLE DUE TO WALL.
902-50 PANEL SB 595 A.E.R.
4 2
ALL TIGHT. REAR INACCESSIBLE DUE TO WALL.
902-63 PANEL, UPS SB 595 A.E.R.
18 0
UPS HIGH VOLTAGE EQUIPMENT INACCESSIBLE.
2201-73A RACK SB 609 C.S.R.
6 6
ALL TIGHT.
2201-73B RACK SB 609 C.S.R.
6 6
ALL TIGHT.
2202-73A RACK, ANALOG TRIP SB 609 C.S.R.
6 6
ALL TIGHT.
2202-73B RACK, ANALOG TRIP SB 609 C.S.R.
6 6
ALL TIGHT.
901-61 PNL, CONTROL SB 609 C.S.R.
16 16 ALL TIGHT.
901-62 PNL, CONTROL SB 609 C.S.R.
16 16 ALL TIGHT.
902-61 PNL, CONTROL SB 609 C.S.R.
16 16 ALL TIGHT.
902-62 PNL, CONTROL SB 609 C.S.R.
16 16 ALL TIGHT.
F-4
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Table F.1 Expansion Anchor Tightness Check Results 125 VDC BUS 2 BUS 2,125 VDC BATTERY TB 615 1-G 0
INACCESSIBLE DUE TO PANEL CONSTRUCTION.
2-8300-1 CHRGR #2,125V TB 615 1-G 6
6 ALL TIGHT 2-8300-1 A CH!<GR #2A,125V TB 615 1-G S
5 ALL TIGHT. A SIXTH LOCATION IS ABANDONED HOLE. SOUG DID NOT CONSIDER AN ANCHOR IN THIS LOCATION.
2-B350 CHRGR #2,250V TB 615 1-G 4
3 ALL TIGHT. RIGHT REAR BOLT MISSING - TO BE REINSTALLED.
SWGR 13 SWGR 13 TB 615 22-G 49 49 ALL TIGHT.
7 ALL TIGHT.
1/2-9400-100 AIR HANDLING UNIT TB 615 24-E 4
4 ALL TIGHT.
1/2-9400-102 REFRIGERATION UNIT TB 615 24-E 8
6 ALL TIGHT. 2 INACCESSIBLE DUE TO WALL.
MCC 18-4 MCC TB 615 24-E INACCESSIBLE. ANCHORS ARE UNDER EQUIPMENT.
1-8300-1 CHRGR #1,125V TB 615 24-G 6
6 ALL TIGHT.
1-8300-1 A CHRGR #1 A,125V TB 615 24-G 6
6 ALL TIGHT.
1-8350 CHRGR #1,250V TB 615 24-G 6
6 ALL TIGHT.
1/2-8350 CHRGR #1/2,250V TS 615 24-G 6
6 ALL TIGHT.
125 VDC BUS BUS, TB RES 1B/1 B-1 TB 615 24-H 13 13 ALL TIGHT.
F-5 d
L Table F.1 Expansion Anchor Tightness Check Results SWGR 23 SWGR 23 TB 615 4-G 13 13 ALL TIGHT.
9 ALL TIGHT.
24 WALL BRACE ANCHORS FOR SWGR'S 23 & 24 ALSO TESTED.
ALL TIGHT.
250 VDC BATT 2 BATTERY,250V TB 628 1-G 76 60 ALL TIGHT. WALL AT REAR OF RACKS.
125 VDC BATT 2 BATTERY,125V TB 628 2-G 39 31 ALL TIGHT.
t 125 VDC BATT 1 BATTERY,125V TB 628 24-G 66 45 ALL TIGHT. WALL AT REAR OF RACKS.
250 VDC BATT 1 BATTERY,250V TB 628 24-G 66 66 ALL TIGHT.
L SWGR 24-1 SWGR 24-1 TB 639 12-H 34 34 ALL TIGHT.
SWGR 14-1 SWGR 14-1 TB 639 16-H 32 32 ALL TIGHT.
NOTES:
A.E.R. = AUX ELECTRIC ROOM C.S.R. = CABLE SPREADING ROOM ALL TIGHT - MAY HAVE TURNED, BUT LESS THAT 1 TURN.
F-6 c
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