ML12333A277

WF3-CS-12-00003, Rev. 0, Waterford Steam Electric Station Unit 3 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic. Part 1

ML12333A277
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Site:	Waterford
Issue date:	11/16/2012
From:	Patel D Entergy Nuclear Operations
To:	Office of Nuclear Reactor Regulation
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ML123330366	List: ML12333A272 ML12333A274 ML12333A275 ML12333A276 ML12333A277 W3F1-2012-0100, WF3-CS-12-00003, Rev. 0, Waterford Steam Electric Station Unit 3 Seismic Walkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic. Part 2
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ENCLOSURE to W3F1-2012-0100 WATERFORD 3 SEISMIC WALKDOWN REPORT

Engineering Report No.\VF3-CS-12-t)Ot)03 Rev 0 Page 1 of 35 ENTERGY NUCLEAR Engineering Report Cover Sheet Engineering Report Title: Waterford Steam Electric Station Unit 3 Seismic VaIkdown Report for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic EC No.40510 Engineering Report Type: Cancelled E Superseded Superseded by: Applicable Site(s)Report Origin: Entergy Vendor Vendor Document No.: N/A No Prepared by: Reviewed by: Re e ed b\: Approved by: Approved by: Dinesh Patel-7--7 (BrianPace Peer Review Team Leader (Benamin Koshab)Desien aanaeer Robert Porter>Design Supervisor Mark Adarns Date: 11-16-2012 Date: 11-16-2012 Date: I I162012 Date:________Date:______New Revision LI LI IP1 LI 1P2 LI 1P3 LI JAF Li PNPS LI vy LI wo LI ANO1 LI ANO2 LI ECH LI GGNS LI RBS LI WF3 PLP LI Quality-Related:

LI Yes Engineering Report No.WF3-CS-12-00003 Rev.0 Page 2 of 35 __________________________________________________________________________ Section Title Page 2.1 SAFE SHUTDOWN EARTHQUAKE (SSE) ............................................................................................ 4 2.2 DESIGN CODES, STANDARDS, AND METHODS ................................................................................ 4 4.1 EQUIPMENT SELECTION PERSONNEL ............................................................................................

12 4.2 SEISMIC WALKDOWN ENGINEERS ...................................................................................................

12 4.3 LICENSING BASIS REVIEWERS .........................................................................................................

12 4.4 IPEEE REVIEWERS .............................................................................................................................

12 4.5 PEER REVIEW TEAM ..........................................................................................................................

13 6.1 SAMPLE OF REQUIRED ITEMS FOR THE FIVE SAFETY FUNCTIONS ...........................................

16 6.2 SPENT FUEL POOL ITEMS .................................................................................................................

19 6.3 DEFERRED INACCESSIBLE ITEMS on SWEL ...................................................................................

20 7.1 SEISMIC WALKDOWNS .......................................................................................................................

23 7.2 AREA WALK-BYS ................................................................................................................................. 24 CONDITON IDENTIFICATION .........................................................................................................................

26 CONDITION RESOLUTION .............................................................................................................................

26 8.1 Licensing Basis Evaluation ....................................................................................................................

27 8.2 Corrective Action Program Entries ........................................................................................................

27 8.3 Plant Changes .......................................................................................................................................

28 9.1 PEER REVIEW PROCESS ...................................................................................................................

29 9.2 PEER REVIEW RESULTS

SUMMARY

................................................................................................

29 ATTACHMENT A IPEEE VULNERABILITIES TABLE ..................................................................................

36 ATTACHMENT B SEISMIC WALKDOWN EQUIPMENT LISTS ...................................................................

38 ATTACHMENT C SEISMIC WALKDOWN CHECKLISTS (SWCs)

...............................................................

76 ATTACHMENT D AREA WALK-BY CHECKLISTS (AWCs) .......................................................................

592 ATTACHMENT E POTENTIALLY ADVERSE SEISMIC CONDITIONS .....................................................

802 ATTACHMENT F LICENSING BASIS EVALUATION FORMS ...................................................................

809 ATTACHMENT G PEER REVIEW CHECKLIST FOR SWEL .....................................................................

817 ATTACHMENT H PEER REVIEW COMMENT FORM ...............................................................................

821 ATTACHMENT J SEISMIC WALKDOWN ENGINEER TRAINING CERTIFICATES ..................................

835 __________________________________________________________________________

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 3 of 35 The Great Tohoku Earthquake of March 11, 2011 and the resulting tsunami caused an accident at the Fukushima Dai-ichi nuclear power plant in Japan. In response to this accident, the Nuclear Regulatory Commission (NRC) established the Near-Term Task Force (NTTF). The NTTF was tasked with conducting a systematic and methodical review of NRC processes and regulations and determining if the agency should make additional improvements to its regulatory system. On March 12, 2012 the NRC issued a 10CFR50.54(f) Letter [Ref.

1], which request ed information evaluation of several of the NTTF recommendations. To support NTTF Recommendation 2.3, Enclosure 3 to the 50.54(f) Letter requested that all licensees perform seismic walkdowns to gather and report information from the plant related to degraded, non-conforming, or unanalyzed conditions with respect to its current seismic licensing basis. The Electric Power Research Institute (EPRI), with support and direction from the Nuclear Energy Institute (NEI), published industry guidance for conducting and documenting the seismic walkdowns which represented the results of extensive interaction between NRC, NEI, and other stakeholders. This industry guidance document, EPRI Report 1025286 [Ref.

2],

Entergy Waterford Steam Electric Station Unit 3 has committed to using this NRC-endorsed guidance as the basis for conducting and documenting seismic walkdowns for resolution of NTTF Recommendation 2.3: Seismic. The objective of this report is to document the results of the seismic walkdown effort undertaken for resolution of NTTF Recommendation 2.3: Seismic in accordance with the Guidance, and provide the information necessary for responding to Enclosure 3 to the 50.54(f).

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 4 of 35 Waterford Steam Electric Station Unit 3 (WSES-3) is located on the west (right descending) bank of the Mississippi River in St. Charles Parish, near the town of Taft, Louisiana. The Nuclear Steam Supply System (NSSS) is a pressurized water reactor (PWR) designed by Combustion Engineering Incorporated. The W SES-3 Facility Operating License was issued on March 16, 1985, and is currently rated at 3716 MWt power [Ref.

3]. This section summarizes the seismic licensing basis of structures, systems and components (SSCs) at

W SES-3 which bound the context of the NTTF 2.3 Seismic Walkdown program. 2.1 SAFE SHUTDOWN EARTHQUAKE (SSE) The selection of the SSE is based on a hypothetical earthquake with an epicentral intensity of VI MM occurring adjacent to the site. According to the most recent and acceptable intensity-acceleration relationship by Trifunac-Brady the intensity VI MM corresponds to a horizontal surface acceleration of 0.06g. In order to comply with the minimum accepted acceleration as stipulated by 10CFR100, Appendix A , WSES-3 was designed for a maximum horizontal ground surface acceleration of 0.10g. This very conservative surface acceleration is double the maximum acceleration appropriate for the maximum earthquake which has occurred in ears. The peak vertical acceleration for the postulated SSE is 2/3 peak horizontal acceleration or 0.067g. 2.2 DESIGN CODES, STANDARDS, AND METHODS Principle structures, systems, and components (SSCs) which may either serve to prevent accidents or to mitigate their consequences are designed and are erected in accordance with applicable codes to withstand any deleterious natural phenomena which could be reasonably assumed to occur at the site during the lifetime of the plant. Redundancy is provided in the reactor protective and safety feature systems so that no single failure of an active component of the system would prevent action necessary to avoid an unsafe condition. Seismic Category I defines SSCs as those components (1) whose failure could cause uncontrolled release of radioactivity, (2) that are essential for safe reactor shutdown and the immediate and long-term operation following a Design Basis Accident, or (3) that are essential for a safe and orderly shutdown of the Nuclear Steam Supply System. Response Spectra The design response spectra used in the plant design differ from the design response spectra recommended in NRC Regulatory Guide 1.60, Design Response Spectra for Seismic Design of Nuclear Power Plants, Revision 1 December 1973. The regulatory guide response spectra have slightly higher values in general. Use of Regulatory Guide 1.60 permits utilization of Engineering Report No.WF3-CS-12-00003 Rev.0 Page 5 of 35 damping values indicated in Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants, October 1973. These damping values are equal or greater than the values utilized for WSES-3 plant design. By utilizing lower damping values in the W SES-3 design, as compared to the damping values of Regulatory Guide 1.61, the analysis and design of W SES-3 compensates for any differences. Structures The seismic Category I structures consist of the following: a) Reactor Building (comprising a free standing steel containment vessel, a containment internal structure and a reinforced concrete Shield). b) Reactor Auxiliary Building c) Fuel Handling Building d) Component Cooling Water System Structure Subsystems and Their Supports The following list comes from WSES-FSAR-UNIT-3 Table 3.2-1. All systems that have components classified as Seismic Category I will be listed here. For a more detailed version of specific components classified as Seismic Category I, see Table 3.2-

1. Reactor Coolant System Safety Injection System Shutdown Cooling System Refueling Water Level Indicating System Chemical and Volume Control System Containment Spray System Waste Management System Component Cooling Water System Sampling System Containment Cooling System Essential Services Chilled Water System Fuel Handling System Spent Fuel Pool System Main Steam and Feedwater System Emergency Feedwater System Compressed Air Systems Containment Isolation System Emergency Diesel Generator System Control Room Air Conditioning System RAB Cable Vault and Switchgear Areas Ventilation System Engineering Report No.WF3-CS-12-00003 Rev.0 Page 6 of 35 RAB H&V Equipment Room Ventilation System FHB Ventilation System Containment Atmospheric Release System Shield Building Ventilation System Controlled Ventilation Area System Reactor Cavity Cooling System Miscellaneous HVAC Equipment Combustible Gas Control Containment Vacuum Relief Actuation System Containment Pressure Indication System Containment Water Level Indication System Electrical Systems and Equipment Radiation Monitoring Accident Radiation Monitors Inadequate Core Cooling Instrumentation Miscellaneous Codes and Industry Standards Seismic Class I structures are generally proportioned to maintain elastic behavior when subjected to various combinations of dead loads, thermal loads, accident loads, seismic and tornado loads. Safety-related structural steel is designed in accordance with American Institute of Steel Construction (AISC), Manual of Steel Construction, 7th Edition. Safety-related concrete is designed in accordance with American Concrete Institute (ACI-308-63), Building Code Requirements for Reinforced Concrete with the exception that ACI 318-71 is used for design of reinforcing steel splices. Safety-related welds are designed in accordance with American Welding Society (AWS) D1.1-72, AWS Structural Welding Steel. IEEE-323-1971, General Guide for Type Test of Class I Electric Equipment for Nuclear Power Generating Stations. IEEE Standard 344-1971, IEEE Recommended Practice for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations, was used in qualifying electrical equipment. Some equipment was qualified in accordance with IEEE 344-1975. Piping systems, pumps, valves, heat exchangers and pressure vessels are designed to the following codes and industrial standards. Note that various pieces of equipment were Engineering Report No.WF3-CS-12-00003 Rev.0 Page 7 of 35 designed by different code years at WSES-3. Items were purchased to code years and addenda as specified in W SES-3 specifications ASME Boiler and Pressure Vessel Code,Section II, "Material Specifications, including the latest published addenda in force on the date of purchase and /or design. ASME Boiler and Pressure Vessel Code,Section III, "Nuclear Vessels, including the latest published addenda in force on the date of purchase and/or design. ASME Boiler and Pressure Vessel Code,Section VIII, "Unfired Pressure Vessels," including the latest published addenda in force on the date of purchase and/or design. ASME Boiler and Pressure Vessel Code,Section IX, " ANSI B31.1

.0-1967, Power Piping Code ANSI B31.7-1969, Nuclear Piping Code Engineering Report No.WF3-CS-12-00003 Rev.0 Page 8 of 35 Entergy WSES-3 has committed to conduct and document seismic walkdowns for resolution of NTTF Recommendation 2.3: Seismic in accordance with the EPRI Seismic Walkdown Guidance [Ref.

2]. The approach provided in the Guidance for addressing the actions and information requested in Enclosure 3 to the 50.54(f) Letter includes the following activities, the results of which are presented in the sections shown in parenthesis: Assignment of appropriately qualified personnel (Section 4.0) Reporting of actions taken to reduce or eliminate the seismic vulnerabilities identified by the Individual Plant Examination of External Events (IPEEE) program (Section 5.0) Selection of SSCs to be evaluated (Section 6.0) Performance of the seismic walkdowns and area walk-bys (Section 7.0) Evaluation and treatment of potentially adverse seismic conditions with respect to the seismic licensing basis of the plant (Section 8.0) Performance of peer reviews (Section 9.0) The coordination and conduct of these activities was initiated and tracked by Entergy corporate leadership, which provided guidance to each Entergy site throughout the seismic walkdown program, including WSES-3. Entergy contracted with an outside nuclear services company to provide engineering and project management resources to supplement and assist each individual site. Each site had dedicated engineering contractors, supported by

their own project management and technical oversight, who worked closely with plant personnel.

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 9 of 35 The NTTF 2.3 Seismic Walkdown program involved the participation of numerous personnel with various responsibilities. This section identifies the project team members and their project responsibilities, and provides brief experience summaries for each. For organizational purposes, personnel are presented as being primarily involved wit h either the walkdown effort or the peer review.

Training certificates of those qualified as Seismic Walkdown Engineers are included in Attachment H. Table 4-1 summarizes the names and responsibilities of personne l used to conduct the seismic walkdowns. Experience summaries of each person follow. Table 4-1 Greg Ferguson, DE-Civil X X X X Mar c McCloskey, DE-Mechanical X David Constance, Operations X 2 Ricky Tran , DE-Electrical X John Meibaum, SE-Electrical X James Jamison, SE-Mechanical X Bill Hardin, Licensing X Stephen Picard, DE

-Civil X X J. McDonald, PSA Engineer X Dinesh Patel (ENERCON) X 1 X X Brian Pace (ENERCON) X X Natalie George (ENERCON) X Chu-Chieh Jay Lin (SC Solutions

) X Notes: 1. Designated lead SWE

2. Plant operations representative Engineering Report No.WF3-CS-12-00003 Rev.0 Page 10 of 35 Mr. Ferguson is a Registered Professional Engineer with over 35 years of experience currently assigned to Design Engineering Group (Civil) at Waterford 3. Mr. Ferguson has significant experience dealing with design and modification of seismic structures. Mr.

Ferguson was involved with the Waterford 3 IPEEE seismic walkdowns. Mr. Ferguson

completed the NTTF 2.3 Seismic Walkdown Training Course in June of 2012.

Mr. McCloskey is a Mechanical engineer with over five years of experience currently assigned to Design Engineering Group (Mechanical) at Waterford 3 Mr. Constance is an Operations training instructor with over 30 years of experience currently assigned to Operations group at Waterford 3. He was also involved in operations group at Waterford 3. Mr. Constance also held a Senior Reactor Operator License from 2000 to 2008 and is also a Certified Shift Technical Advisor Mr. Tran is an Electrical engineer with over 20 years of experience currently assigned to Design Engineering Group (Electrical) at Waterford 3. Mr. Tran was also involved with Procurement Engineering Group at Waterford 3 from 1992 to 1996 Mr. Meibaum is an Electrical engineer with over 23 years of experience currently assigned to Systems Engineering Group (Electrical) at Waterford 3 Mr. Jamison is a Mechanical Engineer with over four years of experience currently assigned to Systems Engineering Group (Mechanical) at Waterford 3

Mr. Hardin is a Senior Licensing Specialist with over 40 years of experience currently assigned to Licensing group at Waterford 3. Prior to this Mr. Hardin was a Senior Reactor Operator working as Control Room Supervisor and Senior Operations Training Instructor.

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 11 of 35 Mr. Picard is a Civil Engineer with three years of experience in Design Engineering (Civil). Mr. Picard completed the NTTF 2.3 Seismic Walkdown Training Course in July of 2012.

Mr. McDonald is an Electrical Engineer with ten years of experience in electrical, I&C, and PRA Design Engineering.

Mr. Patel is a Lead Engineer in ENERCON Services Kennesaw, GA Office in the Civil/Structural Engineering department. Mr. Patel has a BS degree in Civil/Architectural Engineering with over 30 years of experience. Mr. Patel has extensive concrete, building design, piping and pipe support design experience. Mr. Patel was also Lead/Responsible Engineer for Main and Auxiliary Transformer Replacements, Vacuum Pump Replacement, Diesel Governor Replacement, Power Uprate related modifications and Emergency Sump Strainers design and installations at various nuclear sites. Mr. Patel has significant seismic experience including the design and modification of nuclear structures and distribution system, selecting equipment, developing specifications, witnessing seismic testing, and equipment supports. Mr. Patel also has extensive experience performing security upgrades at Entergy Sites including Grand Gulf Nuclear Station, River Bend Station, Arkansas Nuclear O ne and W aterford 3 as well as at the Progress Energy sites (Crystal River, Brunswick, Harris, Robinson), Florida Power Sites (St. Lucie and Turkey Point) and the Southern Company Sites (Farley, Hatch and Vogtle). These upgrades included modifications to security buildings, ISFSI installations, VBS installations as well as protected area changes. For the Entergy Sites, Mr. Patel supported the 2002 ICM Security upgrades, the 2004 Revised DBT Security upgrades, and the PA expansion associated with the ISFSI installation at GGNS. Mr. Patel completed the NTTF 2.3 Seismic Walkdown Training Course in July of 2012. Rouge, LA. He is a degreed civil engineer from Louisiana State University with experience in several nuclear projects for Entergy. Mr. Pace was involved with Service Water System Modifications at Arkansas Nuclear One (ANO), where he helped design safety related tie-back restraints for Service Water piping at ANO. Engineering Change and Work Management process. He has design experience for River Bend Station as well as A NO. Mr. Pace completed the NTTF 2.3 Seismic Walkdown Training Course in August of 2012.

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 12 of 35 Ms. George is for piping design and analysis, support qualification, modification development, and various other mechanical/piping design tasks. Ms. George has approximately two years of support qualification and pipe stress analysis experience. She has performed ANSI B31.1 and ASME pipe stress analysis for safety related piping at several plants including Wolf Creek, Turkey Point, and Brunswick while assigned to the Kennesaw office. Her computer analysis code experience includes ME101, AUTOPIPE, and PipeStress2000. Ms. George completed the NTTF 2.3 Seismic Walkdown Training Course in August of 2012.

Dr. Lin is a Senior Engineer in SC Solutions Walnut Creek, CA office. He has over 15 years of experience in Seismic and Safety assessment of various structures, soil structural integration analyses and design evaluation, finite/discrete element simulation of different types of structures and materials, analyzing and designing industrial and urban steel and reinforced concrete structures, structural evaluation, field investigation, data acquisition, structural health monitoring of bridges and experimental modal analysis, performance analysis of vibration, and seismic design. Dr. Lin completed the NTTF 2.3 Seismic Walkdown Training Course in July of 2012.

4.1 EQUIPMENT SELECTION PERSONNEL A total of nine individuals served as Equipment Selection Personnel see Table 4-1. 4.2 SEISMIC WALKDOWN ENGINEERS A total of six individuals served as Seismic Walkdown Engineers see Table 4-1. 4.3 LICENSING BASIS REVIEWERS A total of three individuals served as Licensing Basis Reviewers see Table 4-1. 4.4 IPEEE REVIEWERS A total of two individuals served as IPEEE Reviewers see Table 4-1.

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 13 of 35 4.5 PEER REVIEW TEAM Table 4-2 summarizes the names and responsibilities of personnel who conduct ed peer reviews of the seismic walkdown program. Experience summaries of each person follow. Table 4-2: Peer Review Team Benjamin Kosbab (ENERCON) X 2 X 2 X 1 ,2 Heidi Graf (ENERCON)

X 2 Sada Dhingra (ENERCON)

X Matthew Wilkinson (ENERCON) X Greg Ferguson X X Notes: 1. Peer Review Team Leader

2. Lead peer reviewer of particular activity Dr. Kosbab is a civil/structural engineer with ENERCON specializing in seismic engineering of nuclear power plant structures, systems, and components. He has earned Master of Science and Ph.D. degrees in civil/structural engineering from the Georgia Institute of Technology with a focus on probabilistic seismic response and fragility analysis of industrial structures. In the nuclear industry, Dr. Kosbab has been involved with seismic time-history and response spectra development, seismic equipment qualification, design of seismic supports, walkdowns, seismic fragility screening, dynamic structural analysis, seismic instrumentation analysis, and soil-structure interaction analysis for plant modifications at numerous nuclear facilities. Dr. Kosbab maintains active involvement with the Nuclear Energy Institute (NEI) Seismic Task Force, and completed the EPRI NTTF 2.3: Seismic Walkdown Training in July, 2012. Ms. Graf is a mechanical engineer in the Power Generation Group out of the Kennesaw, Georgia office of ENERCON. Ms. Graf has over 17 years of commercial nuclear power engineering experience including 7 years in Nuclear Plant Design and Support with the Southern Nuclear Company (SNC) Farley Nuclear Plant. Ms. Graf has completed multiple modification projects for various nuclear plants on numerous systems. She has knowledge of Engineering Report No.WF3-CS-12-00003 Rev.0 Page 14 of 35 plant documentation. Ms. Graf has completed several training courses on plant operations and has an understanding on many systems. She has spent many hours studying the IPEEE and the USI A-46 programs and their impacts on the industry Mr. Wilkinson is a Civil Engineer with over 5 years of experience. He has a B.S. in Civil Engineering. Mr. Wilkinson is curreoffice in Kennesaw, GA. As a civil engineer, he is responsible for the development of engineering packages, calculations, analyses, drawings, and reports. Mr. Wilkinson has significant design experience with Florida Power and Light, primarily providing his services for Turkey Point Nuclear Station (PTN) on several modification packages and calculations. Mr. Wilkinson has significant site support experience at PTN, McGuire Nuclear Station and River Bend Nuclear Station. Moreover, Mr. Wilkinson worked directly at PTN for the majority of 2010 to support the Independent Spent Fuel Storage Installation (ISFSI) construction and 2011 to 2012 to support the Extended Power Uprate (EPU) project design phase. Mr. Wilkinson performed seismic walkdowns at Vogtle Nuclear Station.

over 30 years of experience in the design, construction, start up and operation of HVAC and Mechanical Systems for nuclear power plants. Prior to joining ENERCON, Mr. Dhingra was a Senior Lead Consultant for D.P. Engineering and Senior Mechanical Engineer for Entergy Operations, Inc. providing engineering services to Entergy Nuclears Waterford and River Bend stations.

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 15 of 35 During the IPEEE program in response to NRC Generic Letter 88-20 [Ref.

4], plant-specific seismic vulnerabilities were identified at many plants. In this conconditions found during the IPEEE program related to seismic anomalies, outliers, or other findings. IPEEE Reviewers (see Section 4.4) reviewed the IPEEE final report [Ref.

5] and supporting documentation to identify items determined to present a seismic vulnerability by the IPEEE program. IPEEE Reviewers then reviewed additional plant documentation to identify the eventual resolutions to those seismic vulnerabilities not resolved via the completion of the IPEEE program. The seismic vulnerabilities identified for WSES-3 during the IPEEE program are reported in Attachment A. A total of 2 seismic vulnerabilities were identified by the WSES-3 IPEEE program. For each identified seismic vulnerability, the table in Attachment A includes three pieces of information requested by Enclosure 3 of the 50.54(f) Letter: a description of the action taken to eliminate or reduce the seismic vulnerability; whether the configuration management program has maintained the IPEEE action (including procedural changes) such that the vulnerability continues to be addressed; when the resolution actions were completed. The list of IPEEE vulnerabilities provided in Attachment A was used to ensure that some equipment enhanced as a result of the IPEEE program were included in SWEL1 (see Section 6.1.2). Documents describing these equipment enhancements and other modifications initiated by identification of IPEEE vulnerabilities were available and provided to the Seismic Walkdown Engineers (SWEs) during the NTTF 2.3 Seismic Walkdowns.

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 16 of 35 This section summarizes the process used to select the SSCs that were included in the Seismic Walkdown Equipment List (SWEL) in accordance with Section 3 of the Guidance. A team of equipment selection personnel with extensive knowledge of plant systems and components was selected to develop the SWEL. The SWEL is comprised of two groups of items: SW EL 1 consists of a sample of equipment required for safe shutdown of the reactor and to maintain containment integrity (i.e. supporting the five safety functions)

SWEL 2 consists of items related to the spent fuel pool The final SWEL is the combination of SWEL1 and SWEL2. The development of these two groups is described in the following sections. 6.1 SAMPLE OF REQUIRED ITEMS FOR THE FIVE SAFETY FUNCTIONS Safe shutdown of the reactor involves four safety functions: Reactor reactivity control (RRC) Reactor coolant pressure control (RCPC) Reactor coolant inventory control (RCIC) Decay heat removal (DHR) Maintaining containment integrity is the fifth safety function: Containment function (CF) The overall process for developing a sample of equipment to support these five safety functions is summarized in Figure 1-1 of the Guidance. Figure 1-1 of the Guidance provides a screening method for selecting SSCs, starting with all of the SSCs for the plant and reducing the number based on certain screening criteria referenced in Section 3 of the Guidance. The list of equipment coming out of Screen #3 and entering Screen #4 is defined as Base List 1. The list of equipment coming out of Screen #4 is the first Seismic Walkdown Equipment List, or SWEL 1. Development of these lists is described separately in the following sections.

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 17 of 35 6.1.1 Base List 1 Based on Figure 1-1 and Section 3 of the Guidance, Base List 1 should represent a set of Seismic Category (SC) I equipment or systems that support the five safety functions. The IPEEE program was intended to address the seismic margin of SSCs associated with each of the five safety functions. At WSES-3, the EPRI Seismic Margin Assessment (EPRI SMA) method was used to complete the seismic IPEEE program, based on EPRI Report NP-6041 tiAssessment of (Ref. 6). As described in Section 3 of the WSES-3 IPEEE report [Ref.

5], an equipment list was developed representing the Sachieving and maintaining a safe shutdown condition for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following a SSE event. This equipment list of SSCs on the success paths is consistent with the requirements of Screens #1 through #3 of the Guidance. Therefore, the IPEEE equipment list of SSCs on the success paths is used as the starting point for the NTTF 2.3 Seismic Walkdown Base List 1. Each component was then checked in the Entergy Electronic Database to verify its safety classification, preventative maintenance, environment, etc. Plant personnel were consulted to find any additional components that were added or replaced in the past 15 years (since the IPEEE report). The resulting list represents Base List 1. Base List 1 is presented as Table 1 in Attachment B, and has 624 total items. 6.1.2 SWEL 1 Based on Figure 1-1 and Section 3 of the Guidance, SWEL 1 should represent a diverse population of items on Base List 1 including representative items from some o f the variations within each of five sample selection attributes. Additionally, the selection of SWEL 1 items includes consideration of the importance of the contribution to risk for the SSCs.

Equipment Selection Personnel (see Section 4.1) developed SWEL 1 using an iterative process. The following paragraphs describe how the equipment selected for inclusion on the final SWEL 1 are representative with respect to each of the five sample selection attributes while also considering risk significance.

In general, preference for inclusion on SWEL 1 was given to items that are accessible and have visible anchorage while still maintaining the sample selection attributes. SWEL 1 is presented as Table 2 in Attachment B, and has 97 total items. Variety of Types of Systems Items were selected from Base List 1 ensuring that each of the five safety functions was well represented. Additionally, components from a variety of frontline and support Engineering Report No.WF3-CS-12-00003 Rev.0 Page 18 of 35 systems, as listed in Appendix E of the Guidance, were selected. The system type of each item on SWEL 1 is listed on Table 2 of Attachment B. Major New and Replacement Equipment With assistance from plant Operations and Engineering, Equipment Selection Personnel identified items on Base List 1 which are either major new or replacement equipment installed within the past 15 years or have been modified or upgraded recently. These items are designated as such on Base List 1 on Table 1 of Attachment B. A robust sampling of these items is represented on SWEL 1. Variety of Equipment Types According to Appendix B of the Guidance, there are 22 classes of mechanical and electrical equipment. The items on Base List 1 were classified accordingly and the total number of items from each class was determined. Items were then selected from Base List 1 ensuring that each of the equipment classes there was also represented on SWEL 1 in approximately the same ratios. The equipment class of each item on SWEL 1 is listed in Table 2 of Attachment B. Note that SWEL 1 does not include components from Class

13. W SES-3 has no Seismic Class I components that are Class 13, and therefore would not be represented on Base List 1 or SWEL 1.

Variety of Environments Items were selected from Base List 1 located in a variety of buildings, rooms, and elevations. These item locations included environments that were both inside and outside, as well as having high temperature and/or elevated humidity and also within the containment building. Items that were part of borated systems were included as well. The location and environment of each item on SWEL 1 is listed on Table 2 of Attachment B. IPEEE Enhancements With assistance from IPEEE Reviewers, Equipment Selection Personnel identified items on Base List 1 which were enhanced as a result of seismic vulnerabilities identified during the IPEEE program (see Section 5.0). These items are designated as such on Base List 1 on Table 1 of Attachment B. These items are represented on SWEL 1. Risk Significance Information from the plant Probabilistic Risk Analysis (PRA) model was used to determine whether items were risk significant. Where otherwise comparable items could be chosen relative to the sample selection attributes, the item with higher risk significance was chosen.

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 19 of 35 6.2 SPENT FUEL POOL ITEMS The overall process for developing a sample of SSCs associated with the spent fuel pool (SFP) is similar to that of the screening process for SWEL1 and is summarized in Figure 1-2 of the Guidance. The equipment coming out of Screen #2 and entering Screen #3 is defined as Base List 2. The equipment coming out of Screen #4 are the items that could potentially cause the SFP to drain rapidly. The items coming out of either Screen #3 or Screen #4 are the second Seismic Walkdown Equipment List, or SWEL 2. Development of these lists is described separately in the following sections. 6.2.1 Base List 2 Based on Figure 1-2 and Section 3 of the Guidance, Base List 2 should represent the Seismic Category I equipment or systems associated with the SFP. To develop Base List 2, Equipment Selection Personnel (see Section 4.1) reviewed plant design and licensing basis documentation and plant drawings for the SFP and its associated cooling system. Base List 2 is presented as Table 3 in Attachment B, and has 75 total items. 6.2.2 Rapid Drain-Down Rapid drain-down is defined as unintentionally lowering the water level to the top of the fuel assemblies within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after an earthquake. Consistent with the Guidance, the Equipment Selection Personnel (see Section 4.1) identified SSCs that could cause the SFP to drain rapidly by first reviewing the SFP documentation to identify penetrations below about 10 ft. above the top of the fuel assemblies. This review assessed the hydraulic lines and connected equipment of each such penetration for potentially seismically-induced failure modes that could lead to rapid drain down.

The list of SSCs that could cause rapid drain-down is presented as Tabl e 4 in Attachment B which includes the specific basis for determining which SSCs could or could not cause rapid drain-down. The rapid drain-down list is presented as Table 4 in Attachment B, and has a total of 5 items that could potentially cause rapid drain down. 6.2.3 SWEL 2 Based on Figure 1-2 and Section 3 of the Guidance, SWEL 2 is a broad population of items on Base List 2 including representative items from some of the variations within each of the four sample selection attributes (using a sample process similar to SWEL 1), as well as each item that could potentially cause rapid-drain down of the SFP. Due Engineering Report No.WF3-CS-12-00003 Rev.0 Page 20 of 35 to the population of items on Base List 2 being much smaller than Base List 1, the sampling attributes are satisfied differently for SWEL 2 than for SWEL 1. The following paragraphs describe how the equipment selected from Base List 2 for inclusion on SWEL 2 are representative with respect to each of the four sample selection attributes. SWEL 2 is presented as Table 5 in Attachment B, and has 26 total items; of these, 21 items are selected from Base List 2, and 5 are from the rapid drain-down list. Variety of Types of Systems There are two systems associated with SFP cooling. Both of these systems are well-represented on SWEL

2. Major New and Replacement Equipment There have been no major new or replacement equipment installations within the past 15 years associated with the SFP. Therefore, this sampling attribute is not applicable. Variety of Equipment Types There are 6 different equipment classes represented on Base List 2: 0, 2, 5, 7, 8, 14, and 21. Each of these equipment classes is represented on SWEL 2. Variety of Environments All SFP components are located nearby each other, but are in two different environments. The SFP equipment is inside, but some equipment is part of a borated system while the remainder is not. The location and environment of each item on SWEL 2 is listed on Table 5 of Attachment B. 6.3 DEFERRED INACCESSIBLE ITEMS on SWEL Each item on the SWEL shall be walked down as part of the NTTF 2.3 Seismic Walkdown program. In order to perform the seismic walkdowns of these items, it is necessary to have access to them and to be able to view their anchorage. In some cases, it was not feasible to gain access to the equipment or view its anchorage because WSES-3 was at power until October 17, 2012. For these cases, walkdowns of these items have been deferred until the next available refueling outage (RFO) and/or specific equipment outages.

The inaccessible items and some items within cabinets that are available will be walked down during the current outage (RFO 18). The results of these walkdowns will be incorporated into the first updated submittal of the report. The walkdown of the remaining items will be completed during specific system outage windows or during the next scheduled RFO (19). A second

update to the report will be submitted after RFO 19. WSES-3 will provide two updated submittal reports incorporating these deferred walkdowns. The first update will be provided two months after the end of Refuel Outage 18, and the second update will be provided two months after the end of Refuel Outage 19, tentatively scheduled for the Spring of 2014.

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 21 of 35 Deferred items are summarized in the table below.

The reason for deferral is identified as either ACC (indicating that the item is in an inaccessible item while the plant is at power or other work must be done for access), CAB (indicating that the item requires opening cabinet/panel doors which was not permitted by plant Operations personnel during the walkdown period, due to being energized or otherwise), or INS (indicated that insulation must be removed in order to complete the walkdown of that component). A total of 26 items are deferred. Of these, 12 were inaccessible at the time of the initial walkdowns, 13 are in cabinets/panels that have to be opened, and one needs insulation removed. Table 6-1: Deferred Items SWEL1-002 (AWC-018*) SSDEMCC311B Motor Control Center 311B RAB +21, Switchgear B Room CAB SWEL1-004 (AWC-019) CEDEBKR3918

-B Reactor Trip Switchgear Breaker TCB

-2 Compartment 2C RAB +21, Switchgear B Room CAB SWEL1-005 (AWC-026) SSDESWGR31AB Switchgear 31AB RAB +21, Switchgear AB Room CAB SWEL1-006 (AWC-016*) 4KVESWGR3A Switchgear 3A RAB +21, Switchgear A Room CAB SWEL1-018 (AWC NOT ASSIGNED) BM MVAAA109 Reactor Drain Tank Outlet Inside Containment Isolation RCB -11, COL. 14 ACC SWEL1-019 (AWC NOT ASSIGNED) CAPMVAAA103 Containment Purge Inlet Inside Annulus ANN +21, PEN. P

-10 ACC SWEL1-026 (AWC-031) CVCMVAAA209 Charging Header Isolation RB +21 ACC SWEL1-029 (AWC-038*) HVCMVAAA205

-A Control Room Emergency Filtration A Inlet Damper RAB +46, 8A

-L INS SWEL1-037 (AWC NOT ASSIGNED) SI MVAAA307

-A Safety Injection Tank 1A Fill/Drain RCB +35, COL. 17 ACC SWEL1-038 (AWC NOT ASSIGNED) SI MVAAA405

-B RC Loop 1 SDC Suction Inside Containment Isolation RCB +21, COL. 17 ACC SWEL1-047 (AWC NOT ASSIGNED) RC ISV1014 Reactor Vessel Vent to Quench Tank Isolation RCB +46, AZM 196 ACC SWEL1-048 (AWC NOT ASSIGNED) RC ISV3184 Pressurizer Vent to Quench Tank RCB +46, AZM 196 ACC Engineering Report No.WF3-CS-12-00003 Rev.0 Page 22 of 35 SWEL1-059 (AWC NOT ASSIGNED) ACCMFAN0002

-B Wet Cooling Tower B Fan 2

-SB CTB -35, 12A-Q1 ACC SWEL1-075 (AWC NOT ASSIGNED) IC ICDC1-C Instrument Cabinet C

-1C RCB +21, COL. 16 ACC SWEL1-077 (AWC NOT ASSIGNED) IC ICDC9 Instrument Cabinet C

-9 RCB +21, COL. 6 ACC SWEL1-079 (AWC NOT ASSIGNED) SG ILT1115

-A Steam Generator 1 Level IXMITR (Wide Range) RCB -4, COL. 18 ACC SWEL1-080 (AWC NOT ASSIGNED) RC ITE0122

-HA1 Reactor Coolant Loop 2 Hot Leg Temperature RCB -4 ACC SWEL1-084 (AWC-032*) IC ECP08 Engineered Safeguard Control Panel CP8 RAB +46, Control Room CAB SWEL1-097 (AWC-018*) 4KVESWGR3B Switchgear 3B RAB +21, Switchgear B Room CAB SWEL2-001 (AWC-054) FS EBKR314A

-5D Fuel Pool Purification Pump Circuit Breaker FHB +1, 1FH

-V CAB SWEL2-002 (AWC-054) FS EBKR314A

-5M Fuel Pool Pump A Circuit Breaker FHB +1, 1FH

-V CAB SWEL2-003 (AWC-054) FS EBKR314B

-6F Refueling Canal Drain Pump Circuit Breake r FHB +1, 2FH

-W CAB SWEL2-006 (AWC-054) FS EBKR314B

-5M Fuel Pool Pump B Circuit Breaker FHB +1, 2FH

-W CAB SWEL2-007 (AWC-054) FS EOL314A

-5M Fuel Pool Pump A TOL FHB +1, 1FH

-V CAB SWEL2-008 (AWC-054) FS EOL314B

-5M Fuel Pool Pump B TOL FHB +1, 2FH

-W CAB SWEL2-009 (AWC-054) FS EOL314B

-6F Refueling Canal Drain Pump TOL FHB +1, 2FH

-W CAB

• Denotes that the AWC is submitted with this Final Report. Although its associated SWEL item has been deferred, the AWC was conducted due to other SWEL items being in the surrounding area.

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 23 of 35 The NTTF 2.3 Seismic Walkdown program conducted in accordance with the Guidance involves two primary walkdown activities: Seismic Walkdowns and Area Walk-Bys. These activities were conducted at WSES-3 by teams of two trained and qualified Seismic Walkdown Engineers (SWEs) (see Section 4.2). Each team included one engineer with several years of experience in seismic design and the qualification of nuclear power plant SSCs. The second engineer had less experience, but sufficient experience to properly perform the tasks. A total of two SWE teams were used. In certain instances, the teams

-check consistency between the SWES and to ensure that lessons learned were being shared. SWE teams were periodically accompanied into the field by WSES-3 design engineering and operations personnel to open cabinets and answer questions. The seismic walkdowns and area walk-bys were conducted over the course of 3 weeks during October of 2012. Each morning, a pre-job brief with all personnel involved was conducted. This pre-job brief was used to outline the components and areas that would be walked down that day, to ensure consistency between the teams, to reinforce expectations to identifying potential personal safety issues specific to that day, and to allow team members to ask questions and share lessons learned in the field. The SWE teams brought cameras (regular and pole mounted with remote monitor), tape measures, and flash lights into the field to assist with the seismic walkdowns and area walk-bys. 7.1 SEISMIC WALKDOWNS Seismic walkdowns were performed in accordance with Section 4 of the Guidance for all items on the SWEL (SWEL 1 plus SWEL 2), except for those determined to be inaccessible and deferred (see Section 6.3). To document the results of the walkdown, a separate Seismic Walkdown Checklist (SWC) with the same content as that included in Appendix C of the Guidance was created for each item. Additionally, photographs were taken of each item, and included on the corresponding SWC. Prior to performance of the walkdowns, documentation packages were developed that contained the pre-filled SWC and other pertinent information including the location drawings, equipment drawings, response spectra information, previous IPEEE seismic walkdown documentation, current operability evaluations, and anchorage drawings where applicable. These documentation packages were brought with the SWE teams into the plant during the seismic walkdowns. Walkdown inspections focused not only on anchorage conditions and seismic spatial interactions, but also included inspections for other potentially adverse seismic conditions.

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 24 of 35 Anchorage, in all cases, was considered to specifically mean anchorage of the component to the structure. This included anchor bolts to concrete walls or floors, structural bolts to structural steel and welds to structural steel or embedded plates. For welds, the walkdown team looked for cracks and corrosion in the weld and base metal. Other bolts or connections , such as flange bolts on in-line components were not considered as equipment anchorage.

These bolts and connections were evaluated by the SWEs and any potential adverse seismic rather than under . Thus, components with no attachments to the structure are considered as not having anchorage. Nevertheless, the attachment of these components to other equipment was evaluated and inspected for potentially adverse seismic conditions. Cabinets/panels on the SWEL that could be reasonably opened without presenting safety or operational hazards were opened during the walkdown. This allowed visual observation of (if it could be observed without breaking the plane of the equipment opening). Where opening the cabinet/panel exhibit ed undue safety or operational hazards, it was considered inaccessible and the completion of the walkdown of that item was deferred to a later time (see Section 6.3). In addition to the general inspection requirements, at least 50% of the SWEL items having anchorage required confirmation that the anchorage configuration was consistent with plant documentation. Not considering deferred items, there were a total of 97 SWEL1 and SWEL2 items. Of the 97 SWEL items , 56 were considered to have anchorage (i.e., removing in-line/line-mounted components). Of these 56 anchored components, the walkdowns of 31 SWEL items included anchorage configuration verification, which is greater than 50%. Whe n an anchorage configuration verification was conducted, the specific plant documentation used for comparison to the as-found conditions was referenced on the SWC. A total of 123 SWCs are attached, 97 26 with walkdown was initiated, but whose completion was ultimately deferred because the cabinet/panel could not be opened during the walkdown period. Therefore, the 97 completed SWCs represent the completed walkdowns of each SWEL item accessible during the walkdown period. 7.2 AREA WALK-BYS Seismic area walk-bys were performed in accordance with Section 4 of the Guidance for all plant areas containing items on the SWEL (SWEL 1 plus SWEL 2); except for those SWEL items located in plant areas inaccessible during the walkdown period (see Section 6.3). Area Engineering Report No.WF3-CS-12-00003 Rev.0 Page 25 of 35 walk-bys were not deferred where components were deferred simply to open cabinets/panels. A separate Area Walk-By Checklist (AWC) with the same content as that included in Appendix C of the Guidance was used to document the results of each area walk-by performed. Photographs were taken of each area, and included on the corresponding AWC. Where possible, area walk-bys were conducted once for plant areas containing more than one SWEL item. In cases where the room or area containing a component was very large, the extent of the area encompassed by the area walk-by was limited to a radius of approximately 35 ft. around the subject equipment. The extent of the areas included in the area walk-bys is described on the AWC for that area. Because certain areas contained more than one SWEL item, there are fewer total area walk-bys conducted than seismic walkdowns. A total of 46 area walk-bys was necessary to cover all plant areas containing at least one accessible SWEL item. The AWC for each area walk-by completed is included in Attachment D. A total of 46 AWCs are attached, which represent all of the areas containing a SWEL item that were accessible during the walkdown period. Note that the AWCs number up to AWC-053, but some numbers were not used. These unused numbers may be used for AWCs for deferred items.

The walkdown team will select additional AWCs for walkdowns of SWEL items inside containment as well as other deferred items (see Section 6.3).

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 26 of 35 During the course of the seismic walkdowns and area walk-bys, the objective of the SWE teams was to identify existing degraded, non-conforming, or unanalyzed plant conditions with respect to its current seismic licensing basis. This section summarizes the process used to handle conditions identified, what conditions were found, and how they were treated for eventual resolution. CONDITON IDENTIFICATION When an unusual condition was observed by a SWE team in the field, the condition was noted on the SWC or AWC form and briefly discussed between the two SWEs to agree upon whether it was a potentially adverse seismic condition. These initial conclusions were based on conservative engineering judgment and the training required for SWE qualification. For conditions that were reasonably judged as insignificant to seismic response, the disposition was included on the SWC or AWC checklist and the appropriate question was However, some unusual or uncertain conditions (i.e. mild surface corrosion) were reported to site personnel through the Corrective Action Program (CAP) for tracking purposes (see Section 8.2). Not all observations were reported through the CAP. Often times, only a Work Request (WR) was written, or the observation was deemed too insignificant to write a WR or report through the CAP. A total of 72 seismically insignificant conditions were identified and were either reported through the CAP or had a Work Request written for them. These conditions were generally related to either housekeeping (5), missing bolts or screws that posed no seismic concern (7) or mild surface corrosion (60). For conditions that were judged as potentially significant to seismic response, then the condition was photographed and the appropriate question on the SWC or AWC was marked

then immediately reported to site personnel for further resolution (see Section 8.2) and documented for reporting in Attachment E.

A total of 19 potentially adverse seismic conditions were identified. These conditions were generally related to missing or loose anchorage (5), seismically significant housekeeping issues (5), seismically significant corrosion (7), or concrete cracks (2). CONDITION RESOLUTION Conditions observed during the seismic walkdowns and area walk-bys determined to be potentially adverse seismic conditions are summarized in Attachment E, including how each condition has been addressed and its current status.

Each potentially adverse seismic condition is addressed either with a Licensing Basis Evaluation (LBE) to determine whether it Engineering Report No.WF3-CS-12-00003 Rev.0 Page 27 of 35 requires entry into the CAP, or by entering it into the CAP directly.

The decision to conduct a LBE or enter the condition directly into the CAP was made on a case-by-case basis, based on the perceived efficiency of each process for eventual resolution of each specific condition. Some unusual conditions that were not seismically significant were entered into the CAP directly. Other unusual observed conditions either had a WR written for them or were deemed insignificant to report. Further resolution of these conditions is not tracked or reported as part of the NTTF 2.3 Seismic Walkdown program, except by noting the CR and / or Work Request (WR) numbers generated on the applicable SWCs and AWCs. 8.1 Licensing Basis Evaluation Potentially adverse seismic conditions identified as part of the NTTF 2.3 Seismic Walkdown program may be evaluated by comparison to the current licensing basis of the plant as it relates to the seismic adequacy of the equipment in question, as is described in Section 5 of the Guidance. If the identified condition is consistent with existing seismic documentation associated with that item, then no further action is required. If the identified condition cannot easily be shown to be consistent with existing seismic documentation, or no seismic documentation exists, then the condition is entered into the CAP. Of the 19 identified potentially adverse seismic conditions, 7 LBEs were performed. Each LBE performed is documented consistently, and included in Attachment F. The results of these LBEs with respect to the associated potentially adverse seismic conditions are summarized in Attachment E. A total of 7 potentially adverse seismic conditions evaluated using a LBE were dispositioned and required no further action, whereas 0 required CAP entry. 8.2 Corrective Action Program Entries Conditions identified during the seismic walkdowns and area walk-bys that required in for an eventual disposition. Conditions entered into the CAP included three types of unusual conditions identified: Seismically insignificant unusual conditions Potentially adverse seismic condition that does not pass a LBE Potentially adverse seismic condition that bypasses a LBE A total of 34 Condition Reports (CRs) were generated in the CAP as a result of the NTTF 2.3 Seismic Walkdown program. A total of 14 identified conditions already had Engineering Report No.WF3-CS-12-00003 Rev.0 Page 28 of 35 CRs written for them. Of these, the majority (36) were from seismically insignificant unusual conditions. A total of 12 CRs were written relative to potentially adverse seismic conditions. The CR numbers, current status, and resolution (where applicable and available) are summarized for these potentially adverse seismic conditions in Attachment E. 8.3 Plant Changes The CAP entries (CRs) generated by the NTTF 2.3 Seismic Walkdown program are being resolved in accordance with the plant CAP process, including operability evaluations, extent of condition evaluations, and root cause analysis (where applicable). There was one item that required immediate field work as a result of the NTTF 2.3 Seismic Walkdown program. A temporary enclosure was found to be installed in the the floor plug is removed from a contaminated pipe chase below the floor. No calculation or drawings could be found for the enclosure. Since the enclosure is unanchored, there is a distinct possibility that it could move across the floor and strike two different safety related panels during a seismic event. One Local Control Panel (PAC LCP-63), and the Instrument Cabinet (C-3B).

CR-WF3-2012-05172 was initiated. As a result, the enclosure was braced at the top and bottom to existing structural members to ensure the rigidity of the enclosure. This bracing was installed using existing site procedures. EC40448 was issued to document the acceptability of the modified enclosure. Final and complete resolutions of the CRs for seismically insignificant unusual conditions and potential ly adverse seismic conditions will determine if future modifications to the plant are required. Current status and resolutions (where applicable and available) for CRs related to potentially adverse seismic conditions are provided in Attachment E.

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 29 of 35 9.1 PEER REVIEW PROCESS The peer review for the Near Term Task Force (NTTF) Recommendation 2.3 Seismic Walkdowns was performed in accordance with Section 6 of the Guidance. The peer review included an evaluation of the following activities: review of the selection of the structures, systems, and components (SSCs) that are included in the Seismic Walkdown Equipment List (SWEL); review of a 25% sample of the checklists prepared for the Seismic Walkdowns and area walk

-bys; review of licensing basis evaluations and decisions for entering the potentially review of the final submittal report. At least two members of the peer review team (see Section 4.5) were involved in the peer review of each activity. The team member with the most relevant knowledge and experience

t ook the lead for that particular peer review activity. A designated overall Peer Review Team Leader provided oversight related to the process and technical aspects of the peer review, paying special attention to the interface between peer review activities involving different members of the peer review team. 9.2 PEER REVIEW RESULTS

SUMMARY

The following sections summarize the process and results of each peer review activity.

9.2.1 Seismic Walkdown Equipment List Development Peer review of the selection of SSCs for SWEL development was conducted by two peer reviewers. These peer reviewers both have extensive knowledge and experience related to nuclear power plant design, operations, documentation, and SSCs.

The peer review was conducted prior to the seismic walkdowns occurring, and was performed as follows: The draft of SWEL 1 and SWEL 2 were provided to the peer reviewers, along with the corresponding base lists (Base List 1, Base List 2, and SFP rapid drain-down list). the peer reviewers were also provided a written description from the equipment selection personnel of how the SWEL 1 and SWEL 2 were developed.

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 30 of 35 Each peer reviewer independently reviewed the equipment selection process and the resulting SWEL in terms of the equipment selection process presented in Section 3 of the Guidance. The peer reviewers discussed their findings and generated consolidated comments. General comments on the overall list and how it represents adequate diversity were documented on a peer review checklist based on Appendix F of the Guidance. Specific comments on documentation of the various lists and individual item selection decisions were documented on formal comment forms following utility procedure. Comments were provided to the Equipment Selection Personnel (see Section 4.1). Comment resolutions were provided to the peer reviewers to confirm acceptable resolution of all comments. The peer review team reviewed the initial SWEL 1 and SWEL 2 and provided comments and suggestions for modification of the SWEL. Comments included suggesti ng to add components associated with IPEEE vulnerabilities identified at W SES-3 to SWEL1. All of the peer review comments were addressed by the Equipment Selection Personnel. The resolutions were reviewed by the peer review team and it was determined that all comments were adequately addressed.

Based on completion of the SWEL peer review activities described, the peer review team concludes that the Equipment Selection Personnel developed a SWEL that adequately reflects the selection and screening process outlined in the Guidance. The peer reviewers confirmed that all SSCs in the SWEL1 and SWEL2 are Seismic Category I components that do not undergo regular inspections and represent a diverse blend of different component types from critical systems and safety-related functions. The list contains major new and replacement items

. Risk significance w as considered in the component selection. Additionally

, SFP items were appropriately addressed. Specific considerations for how the SWEL adequately represents the sample selection attributes described in Section 3 of the Guidance are provided on the peer review checklist of the SWEL

. The peer review checklist of the SWEL is provided in Attachment G.

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 31 of 35 9.2.2 Seismic Walkdowns and Area Walk

-Bys Peer review of the seismic walkdowns and area walk-bys was conducted by two peer reviewers, each of whom is a qualified SWE and has broad knowledge of seismic engineering applied to nuclear power plants. One of the peer reviewers participated in the seismic walkdown program for a different utility, and the other is engaged with the industry team which developed the Guidance (see Section 4.2). The peer reviews were conducted at the W SES-3 site concurrent with the walkdowns at approximately 50% completion. The peer review was performed as follows: The peer review team reviewed the walkdown packages (including checklists, photos, drawings, etc.) for SWEL items already completed to ensure that the checklists were completed in accordance with the Guidance. A total of 27 SWC and 14 AWC forms were reviewed, each representing approximately 25% of their respective totals. In the context of the Guidance, the peer review team considered the number of walkdown packages reviewed to be appropriate. The packages reviewed represent a variety of equipment types in various plant areas. Specific SWC forms reviewed were SWEL1-003, 006, 007, 008, 010, 011, 012, 013, 016, 043, 044, 046, 051, 054, 056, 060, 066, 069, 070, 071, 072, 082, 087, 089, 090, 092, and 094. Specific AWC forms reviewed were AWC-001, 002, 004, 005, 006, 007, 010, 014, 016, 021, 023, 025, 028, and 042. While reviewing the walkdown packages, the peer reviewers conducted informal interviews of the SWEs and asked clarifying questions to verify that they were conducting walkdowns and area walk-bys in accordance with the Guidance. The peer review team held a meeting with the SWE teams to provide feedback on the walkdown and walk-by packages reviewed and the informal interviews, and discuss potential modifications to the documentation packages in the context of the Guidance. Each peer reviewer accompanied each SWE team into the field and observed them perform a walkdown of a SWEL component and its associated area walk-by. During these observations, the peer reviewers asked clarifying questions to verify the walkdown and walk-by process being followed was in accordance with the Guidance. The items walked down under the observation of a peer reviewer were SWEL1-042, 061, 062, and 064. The associated area walk-bys performed under the observation of a peer reviewer are AWC-038 and -039. The peer review team held a meeting with the SWE teams to provide feedback on the walkdown and walk-by observations, and discuss how lessons learned Engineering Report No.WF3-CS-12-00003 Rev.0 Page 32 of 35 from review of the walkdown packages had been incorporated into the walkdown process. As a result of the peer review activities, the SWE teams modified their documentation process to include additional clarifying details, particularly related to checklist

insignificant. The peer review team felt these modifications would be of benefit for future reviews of checklists incorporated into the final report. These modifications were recommended following review of the walkdown and area walk-by packages, and the observation walkdowns and area walk-bys demonstrated that the SWEs understood the recommendations and were incorporating them into the walkdown and area walk-by process. Previously completed checklists were revised to reflect lessons learned from the peer review process. Based on completion of the walkdown and walk-by peer review activities described, the peer review team concluded that the SWE teams are familiar with and followed the process for conducting seismic walkdowns and area walk-bys in accordance with the Guidance. The SWE teams adequately demonstrated their ability to identify potentially adverse seismic conditions such as adverse anchorage, adverse spatial interaction, and other adverse conditions related to anchorage, and perform anchorage configuration verifications, where applicable. The SWEs also demonstrated the ability to identify seismically-induced flooding interactions and seismically-induced fire interactions such as the examples described in Section 4 of the Guidance. The SWEs demonstrated appropriate use of self checks and peer checks. They discussed their observations with questioning attitude , and documented the results of the seismic walkdowns and area walk-bys on the appropriate checklists. 9.2.3 Licensing Basis Evaluations Licensing Basis Evaluations (LBEs) were developed by members of the walkdown engineering team to document the disposition of those potentially adverse seismic conditions identified which did not require entry into the CAP. Each LBE was independently reviewed for technical content and CAP entry decisions by the lead LBE peer reviewer. A second peer reviewer reviewed the set of all LBEs to ensure the process and decisions made were in compliance with Section 5 of the Guidance. Based on these reviews, the peer review team concludes that the LBEs properly evaluate field conditions relative to the specific plant licensing basis documents and make appropriate decisions for entering the potentially adverse seismic conditions into

-level peer review comments are documented in Attachment H.

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 33 of 35 9.2.4 Submittal Report The peer review team was provided with an early draft of this submittal report for peer review. The peer review team verified that the submittal report met the objectives and requirements of Enclosure 3 to the 50.54(f) Letter, and documented the NTTF 2.3 Seismic Walkdown program performed was in accordance with the Guidance. The peer review team provided the results of review activities to the SWE team for consideration. The SWE team satisfactorily addressed all peer review comments in the final version of the submittal report. The signature of the Peer Review Team Leader

provides documentation that all elements of the peer review as described in Section 6 of the Guidance were completed.

Engineering Report No.WF3-CS-12-00003 Rev.0 Page 34 of 35 1. 10CFR50.54(f) Letter, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident, dated March 12, 2012

2. EPRI 1025286, Seismic Walkdown Guidance for Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic, June 2012

3. Waterford Steam Electric Station Unit 3, Final Safety Analysis Report (FSAR), Revision 306 4. Generic Letter No. 88-20, Supplement 4, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities

5. Waterford 3 Individual Plant Examination of External Events (IPEEE) Reduced Scope Seismic Margin Assessment (SMA). Report No. WF3-CS-12-00001 , 02-07-2012/Revision, 0

6. EPRI Report NP-6041-Engineering Report No.WF3-CS-12-00003 Rev.0 Page 35 of 35 ATTACHMENT A IPEEE VULNERABILITIES TABLE ATTACHMENT B SEISMIC WALKDOWN EQUIPMENT LISTS ATTACHMENT C SEISMIC WALKDOWN CHECKLISTS (SWCs) ATTACHMENT D AREA WALK-BY CHECKLISTS (AWCs) ATTACHMENT E POTENTIALLY ADVERSE SEISMIC CONDITIONS ATTACHMENT F LICENSING BASIS EVALUATION FORMS ATTACHMENT G PEER REVIEW CHECKLIST FOR SWEL ATTACHMENT H PEER REVIEW COMMENT FORM ATTACHMENT J SEISMIC WALKDOWN ENGINEER TRAINING CERTIFICATES

1. IPEEE VULNERABILITY COMMITMENT RESOLUTION CMP RESOLVED V01 Several Seismic interaction issues were found during IPEEE walkdown in the Control Room. Panels not bolted together Personal storage lockers book cases, storage cabinets, lockers, copy machine were behind the panels and were not secured Breathing air cabinet can interact with CP-08 Resolve the Seismic interaction issue in the Control Room

CR-94-1019 was issued to document loose items in the Control Room Modifications were implemented and all issues were resolved -Panels were bolted together -Personal storage lockers, book cases and storage cabinets were either bolted or relocated

-Breathing air cabinet was secured such that it would not interact with CP-08 Y 02-15-1995 V02 Station Air Pipe is close to Switchgear 4KVESWGR3B and can interact with the Switchgear Resolve the Seismic interaction issue with the Switchgear CR-94-1111 was issued to document the station air pipe which is adjacent to 4KVESWGR3B Station Air pipe was rerouted to provide adequate clearance between the switchgear and pipe.

Y Note 1 03-30-1995

Prepared by: Dinesh Patel Date: 10/27/2012 Note 1: Plant documents were modified to incorporate vulnerability resolution.

SWEL1-001 MS MVAAA106 A VALVE MAIN STEAM LINE 1 SAFETY

1. 1 RB +46 R1 MS 0 O, H N 5817-4718 SWEL1-002 SSDEMCC311B PANEL MOTOR CONTROL CENTER 311B RAB +21 212 SSD 1 I N SQ-E-2 SWEL1-003 SSDEMCC315B PANEL MOTOR CONTROL CENTER 315B CTB -35 B59A SSD 1 O, H N 1564-2118 1564-2119 SWEL1-004 CEDEBKR3918 B CKTBRK REACTOR TRIP SWGR BREAKER TCB

-2 COMPARTMENT 2C RAB +21 212 CED 2 I N SQ-NSSS-ICE-3 SWEL1-005 SSDESWGR31AB PANEL SWITCHGEAR 31AB RAB +21 212B SSD 2 I N SQ-E-8 SWEL1-006 4KVESWGR3A PANEL SWITCHGEAR 3A RAB +21 212A 4KV 3 I N 1564-4017 1564-4018 1564-4019 SWEL1-007 ID EMTMD B TRANSF SUPS INVERTER M D AC INPUT MAIN TRANSFORMER RAB +21 212A B ID 4 I Y 1564-1829 SWEL1-008 SSDEMT315B TRANSF (4160-480/277V XFMR) STA SERVICE XFMR

-3B315-S RAB +21 B59A SSD 4 O, H Y 5817-117 SWEL1-009 CC MPMP0001 B PUMP COMPONENT COOLING WATER PUMP B RAB +21 233 B CC 5 I Y 1564-1347 SWEL1-010 EFWMPMP0001 AB PUMP EMERGENCY FEEDWTR PUMP AB RAB -35 B49 AB EFW 5 I Y 1564-1493 SWEL1-011 EGFMPMP0001 A PUMP DIESEL OIL TRANSFER PUMP A RB -35 B52 A EGF 5 O, H Y 1564-1224 SWEL1-012 SI MPMP0002 B PUMP HIGH PRESSURE SAFETY INJECTION PUMP B RAB -35 B16 B SI 5 I, B Y 1564-72 SWEL1-013 SI MPMP0001 A PUMP LOW PRESSURE SAFETY INJECTION PUMP A RAB -35 B15 A SI 6 I, B N 1564-85 1564-86 1564-87 SWEL1-014 ACCMVAAA126 B VALVE ACC HEADER B CCW HX OUTL TEMPERATURE CONTROL VALVE RAB +21 236 B ACC 7 I N 5817-2158 5817-2159 SWEL1-015 ACCMVAAA138 A VALVE ACC WET COOLING TOWER A CROSS-CONNECT ISOLATION CTA -35 A ACC 7 O, H N 1564-9924 1564-9925 SWEL1-016 BAMMVAAA126 A VALVE BORIC ACID MAKEUP PUMP A RECIRC VALVE RAB -35 B38 A BAM 7 I, B N 1564-150 SWEL1-017 BD MVAAA103 A VALVE S/G 1 BLOWDOWN OUTSIDE CONTAINMENT ISOLATION RB -4 B100 A BD 7 I N 5817-11962 SWEL1-018 BM MVAAA109 VALVE REACTOR DRAIN TANK OUTLET INSIDE CONTAINMENT ISOL RCB -11 421 BM 7 I, T, H, B N 1564-4667 SWEL1-019 CAPMVAAA103 VALVE CONTAINMENT PURGE INLET INSIDE ANNULUS ANN +21 420 CAP 7 I, T, H N 1564-4379 SWEL1-020 CC MVAAA135 B VALVE DRY COOLING TOWER B CCW INLET ISOLATION CTB -35 B60A B CC 7 O, H N 1564-8424 1564-8431 SWEL1-021 CC MVAAA322 B VALVE CCW HEADER B RETURN FROM ESSENTIAL CHILLERS ISOL RAB +21 236 B CC 7 I N 1564-4045 SWEL1-022 CC MVAAA835 A VALVE CNTMT FAN COOLERS TRAIN A TEMPERATURE CONTROL RB -4 B100 A CC 7 I N 1564-2514 1564-4390 SWEL1-023 CC MVAAA963 A VALVE SHUTDOWN HEAT EXCHANGER A CCW FLOW CONTROL RAB -35 B17 A CC 7 I N 1564-4628 1564-4629 SWEL1-024 CMUISV0407 B ICNTRL SV FOR CMU

-407B CTB -35 B59A B CMU 7 O, H N 5817-4614 SWEL1-025 CS MVAAA125 B VALVE CONTAINMENT SPRAY HDR B ISOLATION RB -35 B53 B CS 7 I, B N 1564-3133 SWEL1-026 CVCMVAAA209 VALVE CHARGING HEADER ISOLATION RB +21 225B CVC 7 I, B N 5817-1776 SWEL1-027 EFWMVAAA223 B VALVE EMERGENCY FEEDWATER HDR B TO SG2 BACKUP FLOW CNTRL RB +46 R2 B EFW 7 O, H N 5817-3745 SWEL1-028 EFWMVAAA229 B VALVE EMERGENCY FEEDWATER TO SG2 BACKUP ISOLATION RB +46 R2 A EFW 7 O, H N 5817-3571 5817-3744 SWEL1-029 HVCMVAAA205 A VALVE CONTROL ROOM EMERG FLTR A INLET DAMPER RAB +46 314 A HVC 7 I N 1564-6628 SWEL1-030 HVRMVAAA107 VALVE RAB NORMAL SUPPLY TO CVAS DOWNSTREAM ISOLATION RAB -35 B17 HVR 7 I N 1564-8744 5817-3595 SWEL1-031 HVRMVAAA303 A VALVE CVAS FILTER TRAIN A MINIMUM FLOW INLET RAB +46 299 A HVR 7 I N SQ-HV-42 SWEL1-032 HVRMVAAA502 A VALVE EG A ROOM EXHAUST FAN VARIABLE PITCH BLADE RAB +46 304 A HVR 7 I N SQ-HV-11 SWEL1-033 IA MVAAA909 VALVE IA ISOL TO CONTAINMENT @ PEN #9 RB -4 B100 IA 7 I N 1564-3559 SWEL1-034 MS MVAAA116 A VALVE STEAM GENERATOR 1 ATMOSPHERE DUMP VALVE RB +46 R1 A MS 7 O, H N 5817-232 5817-423 5817-6268 5817-12128 SWEL1-035 MS MVAAA124 B VALVE MAIN STEAM ISOLATION VALVE 2 RB +46 R2 B MS 7 O, H N 1564-3707 SWEL1-036 SI MVAAA129 B VALVE LPSI PUMP B DISCHARGE FLOW CONTROL RAB -35 B16 B SI 7 I, B N 1564-284 SWEL1-037 SI MVAAA307 A VALVE SAFETY INJECTION TANK 1A FILL/DRAIN RCB +35 421 A SI 7 I, T, H, B N 1564-145 SWEL1-038 SI MVAAA405 B VALVE RC LOOP 1 SDC SUCTION INSIDE CONTAINMENT ISOL RCB +21 421 B SI 7 I, T, H, B N 1564-1269 SWEL1-039 SVSMVAAA201 B VALVE AH-30 SB INLET DAMPER D

-50(SB) RAB +7 B SVS 7 I, B N 1564-5763 1564-6028 1564-6628 SWEL1-040 BAMMVAAA113 A VALVE BORIC ACID MAKEUP TANK A GRAVITY FEED VALVE RAB -35 B38 A BAM 8 I, B N SQ-NSSS-PE-24 SWEL1-041 CARMVAAA204 A VALVE CAR EXHAUST HEADER A DISCHARGE RAB +46 299 A CAR 8 I N 1564-4386 SWEL1-042 CHWMVAAA900 VALVE SWGR MAIN SVSMAHU0001

-B CHW OUTLET FCV RAB +46 323 B CHW 8 I N 5817-2710 SWEL1-043 EGAISV0411 B VALVE EG B EMERGENCY MODE FUEL CONTROL #2 RAB +21 222 EGA 8 I N 1564-2080 1564-2081 1564-2082 SWEL1-044 MS MVAAA120 A VALVE MSIV 1 UPSTREAM DRIP POT NORMAL DRAIN RB +46 R1 A MS 8 O, H N 1564-1548 SWEL1-045 MS MVAAA401 A VALVE EFW PUMP AB TURBINE STEAM SUPPLY FROM S/G 1 RB +46 300 A MS 8 O, H N 5817-5815 5817-6914 5817-7250 5817-8170 SWEL1-046 NG ISV0809 VALVE NITROGEN ACCUMULATOR #5 OUTLET STOP RB +46 R1 NG 8 O, H N 5817-5363 5817-5366 5817-5367 5817-6400 SWEL1-047 RC ISV1014 VALVE REACTOR VESSEL VENT TO QUENCH TANK ISOLATION RCB +46 421 B RC 8 I, T, H, B N 5817-5362 5817-5369 SWEL1-048 RC ISV3184 VALVE PRESSURIZER VENT TO QUENCH TANK RCB +46 421 RC 8 I, T, H, B N 5817-5362 5817-5369 SWEL1-049 SBVMVAAA110 A VALVE SBV EXHAUST FAN A SUCTION ISOLATION RAB +46 299 A SBV 8 I N 1564-4384 SWEL1-050 SBVMVAAA112 B VALVE SBV EXHAUST FAN B RECIRC CHECK RAB +46 - B SBV 8 I N 1564-4084 SWEL1-051 SI ISV1161 A VALVE LPSI PUMP A MINIMUM FLOW RECIRC RAB -35 B15 SI 8 I, B N 5817-6401 SWEL1-052 SI MVAAA121 A VALVE SI RECIRCULATING HDR A TO RWSP DOWNSTREAM ISOLATION RB -35 B53 A SI 8 I, B N 1564-6618 SWEL1-053 SI MVAAA138 B VALVE LPSI HEADER TO RC LOOP 1B FLOW CONTROL RB -35 B53 SI 8 I, B N 1564-142 SWEL1-054 SI MVAAA219 A VALVE HPSI DISCHARGE HEADER A ORIFICE BYPASS RB -35 B53 A SI 8 I, B N 1564-6615 SWEL1-055 SI MVAAA227 B VALVE HPSI HDR B TO RC LOOP 2A FLOW CONTROL RB -35 B53 B SI 8 I, B N 5817-11904 5817-11905 SWEL1-056 SI MVAAA415 B VALVE SHUTDOWN COOLING HX B TEMPERATURE CONTROL RAB -35 B16 B SI 8 I, B N 5817-1419 5817-3594 SWEL1-057 SI MVAAA502 A VALVE RC LOOP 1 HOT LEG INJ ISOLATION RB -35 B53 SI 8 I, B N 1564-6617 SWEL1-058 SI MVAAA602 B VALVE SAFETY INJECTINO SUMP OUTLET HEADER B ISOLATION RB -35 B53 B SI 8 I, B N 5817-10933 SWEL1-059 ACCMFAN0002 B BLOWER WET COOLING TOWER B FAN 2-SB CTB -35 B60A ACC 9 O, H N SQ-MN-203 SWEL1-060 CC MFAN0003 A BLOWER DRY COOLING TOWER A FAN 3-SA CTA -35 B59A A CC 9 O, H N 1564-1435 5817-11268 SWEL1-061 HVCMAHU0001 A BLOWER CONTROL ROOM AIR HANDLING UNIT AH

-12A RAB +46 A HVC 9 I N 1564-7545 1564-9361 1564-9362 1564-9363 SWEL1-062 HVCMFAN0010 B BLOWER CONTROL ROOM EMER FILTRATION UNIT B RAB +46 314 B HVC 9 I N 1564-5052 SWEL1-063 HVRMFAN0025 A BLOWER EDG ROOM A EXHAUST FAN E-28A RAB +46 299 A HVR 9 I N 1564-4560 SWEL1-064 HVCMAHU0009 A BLOWER CONTROL ROOM EMER FILTRATION UNIT A RAB +46 A HVC 10 I Y 1564-5051 SWEL1-065 HVRMAHU0028 A BLOWER CCW PUMP ROOM A AIR HANDLING UNIT AH

-10A RAB +21 235 A HVR 10 I N 1564-4586 1564-8953 SWEL1-066 HVRMAHU0032 B BLOWER SHUTDOWN COOLING HX B AIR HANDLING UNIT AH

-3B RAB -35 B20 B HVR 10 I N 1564-4585 SWEL1-067 RFRMCHL0001 B HTEXCH ESSENTIAL CHILLER B RAB +46 299 B RFR 11 I Y 5817-10990 SWEL1-068 DC EPDPB DC PANEL PDP B-DC RAB +21 212B B DC 14 I Y SQ-E-1 SWEL1-069 ID EPDPMD PANEL PDP M D NUCLEAR INST CHANNEL D RAB +21 212 ID 14 I Y 5817-4678 SWEL1-070 DC EBATB BATTRY 125V BAT & RACK 125VDC BATTERY B-S 60 CELLS RAB +21 213 B DC 15 I Y SQ-E-5 SWEL1-071 DC EBC1A BATTRY BATTERY CHARGER A1 RAB +21 212 A DC 16 I Y 1564-1146 1564-1157 SWEL1-072 EG MDSG0001 B GENERA EMERGENCY DIESEL GENERATOR B RAB +21 222 B EG 17 I Y 1564-1999 1564-2040 1564-2041 1564-2042 1564-2043 1564-2055 1564-2057 1564-2058 SWEL1-073 CC ILS7011 B IBISSW CCW SURGE TANK B SIDE LO/LO LEVEL SWITCH RAB +91 417 B CC 18 I N 1564-1489 SWEL1-074 EGFILT6903 A IXMITR DIESEL OIL FEED TNK A LVL TRANSMITTER RAB +46 328B EGF 18 O, H Y 1564-7973 SWEL1-075 IC ICDC1 C PANEL INSTRUMENT CABINET C

-1C RCB +21 421 IC 18 I, T, H N 1564-9155 SWEL1-076 IC ICDC12 PANEL INSTRUMENT CABINET C

-12 RB -4 B145 IC 18 I N 1564-9210 SWEL1-077 IC ICDC9 PANEL INSTRUMENT CABINET C

-9 RCB +21 421 IC 18 I, T, H N 1564-9158 SWEL1-078 NG IPIS0945 B IBISSW NITROGEN HEADER 6 PRESS INDICATING SWITCH RB +46 NG 18 O, H N 5817-5815 5817-7250 5817-8270 SWEL1-079 SG ILT1115 A IXMITR STEAM GENERATOR 1 LEVEL IXMITR (WIDE RANGE)

RCB -4 421 SG 18 I, T, H N 5817-5727 SWEL1-080 RC ITE0122 HA1 IXMITR REACTOR COOLANT LOOP 2 HOT LEG TEMPERATURE RCB -4 423 RC 19 I, T, H N SQ-IC-6 SWEL1-081 EG ECP6850 B PANEL DIESEL GEN B ENGINE CONTROL PANEL RAB +21 222 B EG 20 I N 1564-2169 SWEL1-082 ESFECP0001 A PANEL ESFAS CABINETS RAB +21 212A ESF 20 I Y 1564-6395 SWEL1-083 IC EAUX2 PANEL AUXILIARY PANEL 2(SB)

RAB +35 262 B IC 20 I N SQ-IC-36 SWEL1-084 IC ECP08 PANEL ENGINEERED SAFEGUARD CONTRL PANEL CP8 RAB +46 304 ESF 20 I N 1564-9337 SWEL1-085 IC ECP22 PANEL CORE PROTECTION CALCULATORS RAB +46 304 CPC 20 I N 1564-5421 1564-5422 1564-5423 1564-5424 SWEL1-086 IC ECP26 PANEL PROCESS ANALOG CONTROL PROTECTIVE CHANNEL B RAB +46 304 PAC 20 I N 1564-2554 SWEL1-087 IC ELCP61 PANEL LOCAL CONTROL PANEL PAC LCP-61 RAB +21 212A PAC 20 I Y 5817-5884 SWEL1-088 RFRECP3311 3B PANEL WATER CHILLER COMPRESSOR CONTROL PANEL 3B RAB +46 299 B RFR 20 I Y SQ-HV-6A SWEL1-089 BAMMTNK0001 A ACCUMU BORIC ACID MAKEUP TANK A RAB -35 A BAM 21 I, B N 1564-427 SWEL1-090 CC MHX0001 A HTEXCH COMPONENT COOLING WATER HEAT EXCHANGER A RAB +21 220 A CC 21 I Y 1564-1466 SWEL1-091 CC MTNK0001 C ACCUMU COMPONENT COOLING WATER SURGE TANK RAB +91 417 CC 21 O, H Y 1564-4554 SWEL1-092 CS MHX0001 A HTEXCH SHUTDOWN COOLING HEAT EXCHANGER A RAB -35 B48 A CS 21 I, B Y SQ-NSSS-PE-39 SWEL1-093 EGFMTNK0001 A ACCUMU FUEL OIL STORAGE TANK A RB -35 A EGF 21 O, H Y 1564-2525 1564-2526 1564-2527 SWEL1-094 NG MACC0003 ACCUMU NITROGEN ACCUMULATOR

1. 3 RAB +21 NG 21 I Y SQ-MN-245 SWEL1-095 EGAMCMP0002 A BLOWER EG A AIR COMPRESSOR A2 RAB +21 221 A EGA 12 I N 1564-2060 5817-9388 SWEL1-096 EGAMRCR0002 A ACCUMU (AIR RECEIVER) EGA AIR RECEIVER A2 RAB +21 221 EGA 12 I Y 1564-2060 5817-9388 SWEL1-097 4KVESWGR3B PANEL SWITCHGEAR 3B RAB +21 212 B 4KV 3 I N 1564-4013 1564-4014 1564-4331