3F1112-06, Response to Recommendation 2.3, Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident

Response to Recommendation 2.3, Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident

ML12335A085
Person / Time
Site:	Crystal River
Issue date:	11/21/2012
From:	Franke J Duke Energy Corp, Florida Power Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
3F1112-06
Download: ML12335A085 (87)

Text

PDuke E Energy Crystal River Unit 3 Nuclear Generating Plant Docket No. 50-302 Operating License No. DPR-72 SECURITY-RELATED INFORMATION -WITHHOLD UNDER 10CFR2.390 Ref: 10CFR50.54(f)

November 21, 2012 3F1 112-06 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - Response to Recommendation 2.3, Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident

Reference:

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 (Accession No. ML12053A340)

Dear Sir:

By letter dated March 12, 2012, the Nuclear Regulatory Commission (NRC) issued a Request for Information (Reference) requesting Licensees to provide information regarding Recommendation 2.3 (Seismic) to support the evaluation of the NRC staff recommendations for the Near-Term Task Force (NTTF) review of the accident at the Fukushima Dai-lchi nuclear facility.

By this letter, Florida Power Corporation (FPC) submits the Crystal River Unit 3 (CR-3) response regarding the performance of seismic walkdowns to identify and address degraded, non-conforming or unanalyzed conditions and to verify the current plant configuration with the current seismic licensing basis. Enclosure 1 to this letter provides the requested information and is consistent with the guidance provided by the Electric Power Research Institute's (EPRI) 2012 Technical Report 1025286, "Seismic Walkdown Guidance For Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic." contains CR-3 management signatures documenting the site review of Enclosure 1. contains a new regulatory commitment contained in this submittal.

FPC requests that Attachments 6 and 7 of Enclosure 1 to this letter, which contain security-related information, be withheld from public disclosure in accordance with 10CFR2.390.

ATTACHMENTS 6 AND 7 OF ENCLOSURE I OF THIS LETTER CONTAIN SECURITY-RELATED INFORMATION.

UPON REMOVAL OF ATTACHMENTS 6 AND 7 FROM ENCLOSURE 1, THIS LETTER IS DECONTROLLED.

SECURITY-RELATED INFORMATION -WITHHOLD UNDER 10CFR2.390 Crystal River Nuclear Plant 15760 W. Power Line Street Crystal River, FL 34428

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U. S. Nuclear Regulatory Commission Page 2 of 3 3F1 112-06 SECURITY-RELATED INFORMATION -WITHHOLD UNDER 10CFR2.390 If you have any questions regarding this submittal, please contact, Mr. Daniel Westcott, Superintendent, Licensing and Regulatory Programs, at (352) 563-4796.

Sincerely, Jon A. Fr k6 Vice Pr sident Crystal River Nuclear Plant JAF/dwh

Enclosures:

1. Response to Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident

2. Report Review by Site Management

3. List of Regulatory Commitments xc:

NRR Project Manager Regional Administrator, Region II Senior Resident Inspector ATTACHMENTS 6 AND 7 OF ENCLOSURE 1 OF THIS LETTER CONTAIN SECURITY-RELATED INFORMATION.

UPON REMOVAL OF ATTACHMENTS 6 AND 7 FROM ENCLOSURE 1, THIS LETTER IS DECONTROLLED.

SECURITY-RELATED INFORMATION -WITHHOLD UNDER 10CFR2.390

U. S. Nuclear Regulatory Commission Page 3 of 3 3F1 112-06 STATE OF FLORIDA COUNTY OF CITRUS Jon A. Franke states that he is the Vice President, Crystal River Nuclear Unit 3 Nuclear Generating Plant for Florida Power Corporation, that he is authorized on the part of said company to sign and file with the Nuclear Repulatory Commission the information attached hereto; and that all such statements mr and att forth therein are true and correct to the best of his knowledge, informatio

'an b on A. Franke Vice President Crstal River Nuclear Unit 3 Nuclear Generating Plant The foregoing document was acknowledged before me this 51I day of 6~T01 dao 2, by Jon A. Franke.

Signature of Nota4 State of Florid SHARON i. LAYTON NOTARY PUBLIC STATE OF FLORIDA (Print, type, or s mp Name of Notary Public)

ENCLOSURE 1 FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT DOCKET NO. 50-302 LICENSE NO. DPR-72 RESPONSE TO RECOMMENDATION 2.3 SEISMIC WALKDOWN OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM THE FUKUSHIMA DAI-ICHI ACCIDENT

U. S. Nuclear Regulatory Commission 3F1 112-06 Page 1 of 16 RESPONSE TO RECOMMENDATION 2.3 SEISMIC WALKDOWN OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM THE FUKUSHIMA DAI-ICHI ACCIDENT 1.0 Introduction...................................................................................................

2 2.0 Seismic Licensing Basis.................................................................................

3 3.0 Personnel Qualifications.................................................................................

4 3.1 Equipment Selection Personnel....................................................................

5 3.2 Seismic Walkdown Engineers.....................................................................

6 3.3 Licensing Basis Reviewers..........................................................................

7 3.4 IP E E E R eview ers.......................................................................................

.. 7 3.5 Peer Review Team Members........................................................................

7 4.0 Selection of SSCs.............................................................................................

7 4.1 SWEL 1 Development...................................................................................

7 4.2 SWEL 2 Development.................................................................................

11 5.0 Seismic Walkdowns and Area Walk-Bys.......................................................

12 5.1 Seismic Walkdown Methodology.................................................................

12 5.2 Area Walk-By Methodology........................................................................

14 5.3 R e s u lts............................................................................................................

1 5 5.4 Maintenance Assessment..........................................................................

15 5.5 Planned or Newly Installed Changes...........................................................

16 5.6 Inaccessible Ite m s.....................................................................................

.. 16 6.0 Licensing Basis Evaluations..........................................................................

16 7.0 IPEEE Vulnerabilities Resolution Report....................................................

16 8.0 Peer Review Report.......................................................................................

16 :

Base List 1 :

SWEL 1 :

Base List 2 :

Rapid Drain-Down List :

SWEL 2 :

Seismic Walkdown Checklists :

Area Walk-By Checklists :

Peer Review Report

U. S. Nuclear Regulatory Commission 3F1 112-06 Page 2 of 16 1

INTRODUCTION The Nuclear Regulatory Commission (NRC) has issued a Request for Information pursuant to Title 10 of the Code of Federal Regulations 50.54(f) (hereafter, 50.54(f) letter) regarding, "Recommendations 2.1, 2.3, and 9.3 of the Near-Term Task Force (NTTF) review of insights from the Fukushima Dai-Ichi Accident," resulting from the Great Tohoku Earthquake and subsequent tsunami.

This submittal report, pursuant to the NRC's request for information, addresses the scope associated with the 50.54(f) letter Enclosure 3, NTTF Recommendation 2.3 Seismic, only.

Specifically, this report provides information for the Crystal River Unit 3 Nuclear Generating Plant (CR-3) regarding the performance of seismic walkdowns to identify and address degraded, non-conforming or unanalyzed conditions and to verify the current plant configuration with the current seismic licensing basis. The information provided herein and the activities described in this report are consistent with the guidance provided by the Electric Power Research Institute's (EPRI) 2012 Technical Report 1025286, "Seismic Walkdown Guidance: For Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic."

The NRC, in its letter dated May 31, 2012 (Accession No. ML12145A529), endorsed the EPRI guidance document.

Per EPRI 1025286, the 2.3 Seismic Walkdown inspections performed were non-intrusive visual inspections of primarily plant Seismic Category I structure, systems and components (SSCs).

During the inspections, observed degraded, nonconforming, or unanalyzed conditions were identified and addressed through the CR-3 Corrective Action Program (CAP). Based on the EPRI guidance document, the list of structures, systems and components (SSCs) for inspection were obtained through a systematic selection process to establish a broad, diverse and representative Seismic Walkdown Equipment List (SWEL). The SWEL was made up of two separate lists: SWEL 1 included 134 SSCs from various locations throughout the plant and SWEL 2 included a shorter specific list of six Spent Fuel Pool (SFP) SSCs.

The selection process for the SSCs combined with the inspection checklist attributes assessed design basis seismic capabilities of the plant.

These attributes pertain to SSC anchorage, interaction and other considerations based on NRC and industry insights of the Fukushima Dai-Ichi Accident.

Similar past seismic efforts include the Individual Plant Examination for External Events (IPEEE) and Unresolved Safety Issue (USI) A-46, "Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactor," programs. Many of the same SSCs inspected for the IPEEE were re-inspected for the current 2.3 Seismic Walkdowns. Most of the SWEL items originated from the USI A-46 and IPEEE Safe Shutdown Equipment Lists (SSEL).

These programs occurred in the 1990s. The USI A-46 program reviewed equipment in older nuclear plants and assessed their seismic capability related to experience based data and calculations.

Where needed, equipment modifications were made to meet the required seismic capabilities.

The IPEEE program used Seismic Margin Assessment (SMA) programs to assess the plants capabilities to perform properly to a larger Review Level Earthquake (RLE). Modifications were also performed as a result, if necessary.

The 2.3 Seismic Walkdown Inspections were performed to visually check the condition of the SSCs and its anchorage to meet its seismic design basis. Also inspected were the surrounding equipment and area for interactions with other SSCs, fire hazards, water spray, and housekeeping issues that may interact with the SSCs. Conditions found were recorded on the developed checklists and evaluated. Any condition that was assessed to be a potential adverse

U. S. Nuclear Regulatory Commission 3F1 112-06 Page 3 of 16 seismic condition (PASC) was to be further evaluated for its ability to meet its seismic design basis requirements and put into the plant CAP, as necessary.

For CR-3, no PASCs were identified, as discussed in Section 5.3, Results. In addition to checking the SSCs with respect to their design basis, this report discusses the general adequacy of licensee monitoring and maintenance procedure by reviewing walkdown observations. No SSCs initially assessed to be a PASC were identified as a result of the seismic walkdowns to be a PASC, as they were evaluated and determined by the seismic walkdown engineers (SWE) and/or by the CAP to meet their seismic design basis requirements and be able to perform their intended safety function.

2 SEISMIC LICENSING BASIS CR-3 is located on gently southwesterly dipping biogenic carbonate (limestone and dolomite) rocks which have been differentially dissolved along the most pervious zones of the rock, resulting in a network of general vertically oriented dissolved zones (solution channels). The bedrock solution process was studied, and the natural solution process was not considered to present any future threat to the soundness of the rock system. The site is located in an area where groundwater exists under water table conditions, at a depth of approximately 10.0 feet below the ground surface. The nearest known faulting occurs three miles east of the plant.

Florida is an area which is considered seismically inactive.

In a 300-year history, only eight earth-quakes of Intensity IV (Modified Mercalli) or greater have had their epicenters located within the state. No earthquake is known to have occurred within 50 miles of the plant site. The closest area of significant seismic activity is Charleston, South Carolina.

Attenuation data available for this area indicates that the site experiences an observed intensity no higher than Intensity V (Modified Mercalli). That maximum ground motion at the site due to this earthquake probably did not exceed 0.025g. For design purposes, the maximum ground acceleration was assumed to be 0.05g.

Response spectra were developed for the site normalized for a maximum ground motion of 0.05g and based upon a large earthquake in the Charleston, South Carolina region and moderate earthquakes in Florida. From the results of site subsurface and regional tectonic investigations, it has been concluded that the foundation mass system can competently support the plant and that inactive regional tectonic elements present no threat to the structural integrity of the installation.

Since CR-3 is a pre-Regulatory Guide 1.29 plant, Seismic Category I as regards seismic licensing basis per the EPRI guidance is discussed in terms of seismic design class of SSC for CR-3. For CR-3, Class I SSCs are defined as those whose failure might cause or increase the severity of a Loss-of-Coolant Accident or result in an uncontrolled release of radioactivity, and those structures and components which are vital to safe shutdown and isolation of the reactor.

Class II SSCs are defined as those SSCs which are important to reactor operation but not essential to safe shutdown and isolation of the reactor, and whose failure would not result in the release of substantial amounts of radioactivity. The balance of SSCs are designated Class Ill.

The seismic design of Class I structures are based on the response to ground acceleration such that:

Primary steady state stresses, when combined with the seismic stress resulting from the response to a ground acceleration of 0.05g acting horizontally and 0.033g acting vertically and occurring simultaneously, have been maintained within the allowable working stress limits accepted as good practice and, where applicable, set forth in the appropriate design standards.

U. S. Nuclear Regulatory Commission 3F1 112-06 Page 4 of 16 Primary steady state stresses, when combined with the seismic stress resulting from the response to a ground acceleration of 0.10g acting horizontally and 0.067g acting vertically and occurring simultaneously, have been limited so that the function of the structure is not impaired so as to prevent a safe and orderly shutdown of the plant.

The seismic design of Class II SSCs is based on the first item in the criteria above or in accordance with the recommendations of the Uniform Building Code. The seismic design of Class 1* structures are based on the above criteria to prevent fall down only. Class I and II SSCs have been designed to meet these requirements under applicable codes including:

Building Code Requirements for Reinforced Concrete, ACI 318-63 Specifications for Structural Concrete for Buildings, ACI 301-66 Specification for the Design and Erection of Structural Steel for Buildings, 1963, AISC

" ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels;Section VIII, Unfired Pressure Vessels;Section IX, Welding Qualifications (applicable portions)

Specification for the Design and Construction of Reinforced Concrete Chimneys, ACI 505-54 Code Requirements for Nuclear Safety Related Concrete Structures, ACI 349-97 Standard Specification for the Design Fabrication and Erection of Structural Steel for Buildings November, 1983, AISC Manual of Steel Construction, "Allowable Stress Design," 9th Edition, including the Specification for Structural Steel Buildings, June, 1989, AISC A variety of seismic analysis methods were used for various SSCs as detailed in Chapter 5 of the CR-3 Final Safety Analysis Report (FSAR). A general description is provided below:

The Containment Building was analyzed by using the general bending theories of a thin shell of revolution.

Class I structures, other than containment, were analyzed by employing a response spectrum method.

Piping was analyzed using uniform response spectra methodologies and independent support motion methodologies to combine the vertical and horizontal inputs.

Class I mechanical, electrical, and instrument components, seismic design criteria was designed, tested or analyzed to ensure that the equipment will function if subjected to seismic accident loadings and vibratory loadings.

For equipment qualified by testing and for equipment qualified by analysis, the effect of the building and floor amplification was included.

U. S. Nuclear Regulatory Commission 3F1112-06 Page 5 of 16

• Equipment such as safety feature valves, tanks, and heat exchangers were stress analyzed using the equivalent static load methods for normal and abnormal conditions.

Earthquake experience data methodology using seismic qualification utility group (SQUG) methodologies was used where appropriate as a basis for seismic technical evaluation of commercial grade replacement items.

3 PERSONNEL QUALIFICATIONS 3.1 Equipment Selection Personnel 3.1.1 Billy Alumbaugh Billy R. Alumbaugh is a Registered Professional Engineer and has over 30 years of engineering experience, including 16 years nuclear experience with site experience working for a utility and as a consultant. Progressive experience in civil engineering ranges from individual contribution to supervisory and project management.

Mr. Alumbaugh supervised multiple engineers at an operating nuclear facility and was involved in several projects including: Control Room expansion, Equipment obsolescence, Dry Fuel Storage, and Containment redesign/design pressure uprate. Training received includes Auxiliary Operator, Waste Control Operator, Systems Training, 10CFR50.59 certification, and modification/change control. As a consultant, he served as the Civil/Structural Engineering Design Lead for the new plant Design Certification and Combined Operating License projects providing a technical review of civil based licensing responses to clients or the NRC and project management.

More recently he served as the Civil-Structural-Architect Discipline Manager for the detailed design phase of the US-APWR including all aspects of the design including the site specific and DCD seismic evaluations.

Mr. Alumbaugh has a MS and a BS degree in Civil Engineering.

3.1.2 Harold Bamberger Harold Bamberger has over 40 years of experience in both field and office functions required for designing, analyzing, and installing piping and pipe supports for metallic and non-metallic systems in major power, chemical, and pharmaceutical facilities.

Mr. Bamberger has worked for various nuclear power plants in design and review of piping, piping supports and other nuclear structure using ASME Section III, ASME/ANSI B31.1 and B31.3, and applicable nuclear plant procedures.

Mr. Bamberger is a Registered Professional Engineer and holds an AD, Mechanical Engineering Technology degree with additional classes in Mechanical Engineering and Technology.

3.1.3 Leon Gagne Leon Gagne is a licensed Senior Reactor Operator and has almost 30 years of operations experience at CR-3; ranging from entry level operator to Shift manager and is currently performing the duties of an Operations Specialist.

Along with his extensive plant operating experience, he has provided key interface with all other site departments.

Some of Mr. Gagne's roles for the

U. S. Nuclear Regulatory Commission 3F1 112-06 Page 6 of 16 Operations Department have included performing and leading major plant outages, self assessments, benchmarking trips, Operating Experience and Corrective Action Program coordination, and special site projects such as the Reactor Coolant System leakage identification and elimination.

3.1.4 Nazir Sheikh Nazir Sheikh is a Registered Professional Engineer and has over 35 years engineering experience with over 30 years nuclear experience. Mr. Sheikh has been associated with nuclear design in nuclear piping design, concrete and steel structures using ACI 318, ACI 349, AISC, mechanical electrical equipment qualification and testing in accordance with IEEE-344. Mr. Sheikh has a BS in Civil Engineering and is currently studying towards a MS in Mechanical Engineering.

3.2 Seismic Walkdown Engineers 3.2.1 Nazir Sheikh - see above 3.2.2 Doyle Adams Doyle G. Adams has over 36 years engineering experience in both design and construction. This includes over 22 years of nuclear experience at an operating nuclear facility.

Nuclear experience includes design engineer, design engineering supervisor, SQUG/IPEEE qualifications and walkdown engineer, seismic equipment testing and qualification. Mr. Adams was the lead responsible civil design engineer for the Reactor Building design pressure uprate, Steam Generator and Reactor Head Replacement projects.

Mr. Adams has a BS Architectural Engineering in structures.

3.2.3 Gayuruddin Ahmed Gayuruddin Ahmed has over 30 years of experience in civil engineering. He has experience in design of nuclear power plant structures and components with extensive use of finite element methods of frame analysis.

He also has experience in ACI, AWS, and AISC code requirements and related design procedures of power plants for seismic conditions.

3.2.4 Martin Foster Martin P. Foster has over 35 years of engineering experience most of which is for working at or for various nuclear plants around the US. His experience includes Seismic equipment qualification and testing for such equipment as control panels, instrument and battery racks, chargers and direct replacement of valves and instruments, seismic design of structures and equipment, seismic piping stress analysis and support design. Mr. Foster has a BS in Civil Engineering.

U. S. Nuclear Regulatory Commission 3F1112-06 Page 7 of 16 3.2.5 Primo Novero Primo Novero has over 46 years of engineering experience, with 40 years in structural design, four in construction, and two in environmental.

He has 34 years of nuclear structural design experience in various structures, systems and components involving different materials, and in diverse topical matters including seismic.

Mr. Novero has a BS in Civil Engineering and Environmental Engineering, and is a Registered Professional Engineer in the field of Civil, Structural and Environmental.

3.3 Licensing Basis Reviewers 3.3.1 Primo Novero - see above 3.3.2 Nazir Sheikh - see above 3.4 IPEEE Reviewers 3.4.1 Doyle Adams - see above 3.5 Peer Review Team Members 3.5.1 Billy Alumbaugh - see above 3.5.1 Louis Wade Louis Wade has over 30 years experience in Quality Assurance/Quality Control (QA/QC), Project Management, and QA/QC consulting. Mr. Wade has over 15 years in management positions associated with construction, maintenance, modifications, including work package control, and operation of DOE and NRC regulated facilities such as Nuclear Power Plants, Vitrification Facilities, Radioactive Waste Facilities, Gaseous Diffusion Facilities, and TRU Waste Characterization and Disposal.

Mr. Wade is an ASQ Certified Quality Auditor 10600, Lead Auditor per ANSI N45.2.23, and Lead Auditor per ASME-NQA 1.

4 SELECTION OF SSCS 4.1 SWEL 1 Development The selection of SSCs included in SWEL 1 for CR-3 was based on the EPRI guidance document, Section 3. This selection process was conducted by Equipment Selection Personnel selecting SSCs based on selection criteria.

During the process, plant operations staff assisted the Equipment Selection Personnel. The process, as described in the EPRI guidance document, involves the use of screening "selection criteria."

These screens are listed as follows:

0 Screen #1: Seismic Category I 0

Screen #2: Equipment or systems NOT regularly inspected

U. S. Nuclear Regulatory Commission 3F1 112-06 Page 8 of 16 Screen #3: Supports five safety functions Reactor reactivity control Reactor coolant pressure control o

Reactor coolant inventory control o

Decay heat removal o

Containment function Screen #4: Sample considerations (systems, major new/replacement, equipment types, environments, IPEEE enhancements)

The list of equipment resulting from Screen #3 is Base List 1. At CR-3, the Base List 1 was created, as suggested by the EPRI guidance document, through the use of previous equipment lists from implementation of the IPEEE and USI A-46 program. The EPRI guidance document suggested the use of previous SSELs, as such, from the IPEEE or USI A-46 program.

Per EPRI 1025286, the first screen narrows the list to SSCs classified as Seismic Category I items because only those have a defined seismic licensing basis against which to evaluate the as-installed configuration.

The second screen further narrows the list by selecting only those remaining items that do not have regular inspections to confirm their configuration is consistent with the licensing basis.

The third screen ensures that those remaining items are associated with at least one of the five safety functions.

The CR-3 SSEL submitted to the NRC for the USI A-46 program meets the criteria for Screens #1, #2, and #3, and thus using this USI A-46 SSEL was appropriate. Other SSCs added to this original equipment list were modified plant equipment verified to meet the criteria for Screens #1, #2, and #3.

The original equipment (USI A-46 SSEL) list with these additional items served as Base List 1, included as Attachment 1.

Once Base List 1 was established, Screen #4 was applied to ensure the inspections encompassed a broad and varying array of equipment. Screen #4 included selection considerations compiled from the EPRI guidance document and from the 50.54(f) letter.

This resulted in the creation of SWEL 1, included as Attachment 2.

Considerations made for the creation of SWEL 1 are detailed in the sections below.

4.1.1 Equipment types/classes A breakdown of the inspected items into the various equipment classes is provided in the following table.

B, e LiSt 0

Other 88 11 Motor Control Centers and Wall-Mounted 11 1

1 Contactors 2

Low Voltage Switchgear and Breaker Panels 17 5

U. S. Nuclear Regulatory Commission 3F1 112-06 Page 9 of 16

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~

=

~Base; List I Nb~*p#

Ttl Selected 3

Medium Voltage Metal-Clad Switchgear 5

2 4

Transformers 14 2

5 Horizontal Pumps 42 8

6 Vertical Pumps 7

1 7

Pneumatic-Operated Valves 87 11 8

Motor-Operated and Solenoid Operated Valves 130 20 9

Fans 12 2

10 Air Handlers 48 4

11 Chillers 2

1 12 Air Compressors 14 2

13 Motor Generators 0

0 Distribution Panels and Automatic Transfer 39 14 Switches 15 Battery Racks 4

3 16 Battery Chargers and Inverters 15 4

17 Engine Generators 2

1 18 Instrument Racks 108 15 19 Temperature Sensors 28 4

20 Instrument and Control Panels 138 19 21 Tanks and Heat Exchangers 88 13 Total 899 134 4.1.2 Five Safety Functions The appropriate distribution of SSCs serving each of the five safety functions was maintained in the selection of SSCs to achieve a sufficient representation of the 5 safety functions for the SWEL 1 as follows:

U. S. Nuclear Regulatory Commission 3F1112-06 Page 10 of 16 bSWEL..

Reactor reactivity control 57 Reactor coolant pressure control 56 Reactor coolant inventory control 56 Decay heat removal 108 Containment function 61 This table demonstrates full coverage of the five safety functions for the selected SSCs.

4.1.3 Locations Although not required by the guidance, SSCs in a variety of plant locations were considered for inclusion on SWEL 1, including the Control Complex, Reactor Building, Auxiliary Building, Diesel Generator Building, Emergency Feedwater Pump (EFP) 3 Building, Intermediate Building, Turbine Building, EFW Tank Building, and the West Berm. SWEL 1 in Attachment 2 includes the building of each SSC.

4.1.4 Environments SSCs from a variety of environments including dry and hot, wet and cold, mild and harsh, and inside and outside buildings were included for inspection in the SWEL 1. SWEL 1 in Attachment 2 includes the environment of each SSC.

4.1.5 Systems During the SWEL 1 selection process, consideration was given to equipment of varying systems including the Reactor Coolant, Feedwater, Main Steam, and Decay Heat Closed Cycle Cooling Systems, among others.

Table B-1 of Appendix E, "Safety Function-System Matrix for PWRs," of the EPRI guidance was consulted to ensure systems to support safety functions were included.

Additionally, equipment in the Auxiliary Building (Sea Water Room) and associated with the Raw Water (RW)/Service Water (SW) System that comprises the Nuclear Services Cooling Water function for the Ultimate Heat Sink was included in SWEL 1. SWEL 1 in Attachment 2 includes the system of each SSC.

4.1.6 Risk The contribution of individual items to overall risk was considered in the selection of the SSCs from the SSEL list for items to include in the SWEL 1. The selection team was able to readily identify items that posed a higher risk ranking due to their knowledge and experience of nuclear plant operations and those SSCs that contribute to nuclear plant risk profiles. An element of the team's experience

U. S. Nuclear Regulatory Commission 3F1112-06 Page 11 of 16 included knowledge of seismic PRA and other risk lists that comprise SSCs and conditions that combine probability and consequences of an event. Items, such as emergency diesels, station batteries, core cooling systems, emergency cooling water systems, and 1 E electrical switchgear, are identified as critical equipment that have a higher risk profile. These equipments were included while maintaining a balance with the other requirements of SWEL equipment selection.

4.1.7 IPEEE vulnerabilities No seismic vulnerabilities were identified at CR-3; however, the USI A-46 program did address several outliers which were considered in the selection process. SWEL 1 in Attachment 2 identifies the USI A-46 outliers inspected in the walkdowns.

4.1.8 Modified, replacement, and new equipment A review of the plant modifications from 1995 (the initiation time of IPEEE) to the present was performed to determine significant modifications to the plant.

Fourteen significant modifications were identified for CR-3: three pertained to installing new equipment and 11 were equipment replacements. Plant support personnel identified the modifications to be included in the walkdowns. SWEL 1 in Attachment 2 identifies the modified SSCs inspected in the walkdowns 4.1.9 Accessibility Before and during the walkdowns, some SSCs were determined to be inaccessible due to a variety of reasons, such as the item was in a high radiation area, blocked by sensitive instruments or were overhead and required scaffolding to access. When an item was removed from SWEL 1, a review of Base List 1 was completed to determine if similar equipment was accessible and a substitution was made. Items that did not have an acceptable substitute are to be inspected at a later date and are discussed in Section 5.6.

4.2 SWEL 2 Development Equipment Selection Personnel along with plant operations and systems personnel developed the CR-3 Base List 2, Rapid Drain-Down List and SWEL 2 from the CR-3 Spent Fuel Pool (SFP) System and based on the EPRI guidance document which presents screening criteria to identify specific equipment that is unique to the SFP SSCs.

Screen #1 and #2 limit SFP SSCs to those which have a Seismic Class I licensing basis and are capable of being visually reviewed in the plant. This list is determined to be Base List 2 and is included as Attachment 3.

Six (6) items were selected for the SWEL 2 from Base List 2 and include the Borated recirculating pump motor SFP-2, the control switch for pump SFP-2, the SFP coolant pump motor SFP-1B, valve SFV-90 and heat exchangers SFHE-1B and SFHE-1A, as representative of the equipment in the SFP System.

The Rapid Drain-Down List, included as Attachment 4, identifies items that have the possibility of providing a hydraulic pathway for a rapid drain-down of the SFP within 72

U. S. Nuclear Regulatory Commission 3F1 112-06 Page 12 of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after an earthquake to a level approximately ten feet above the spent fuel stored in the pool.

The Spent Fuel Pool Cooling System is designed Seismic Class I, which includes piping, valves, supports, pumps and motors, and heat exchanger.

The following items were evaluated to determine if it provided a rapid drain down path:

The suction and return piping penetrate the SFPs at elevation 154 ft. 6 in., which is 21 ft. 6 in. above the spent fuel and, thus, do not create a rapid drain down possibility.

The weir gates are considered part of the SFP structure and, therefore, do not create a credible failure mode.

The fuel transfer tube is a Seismic Class 1 SSC and has a gate valve on the SFP end and a blind flange on the Reactor Building end. The blind flange has double seals with a pressure leak detection channel between.

A line penetrates into the cask area at elevation 146 ft.10 in. and proceeds down to elevation 138 ft. 4 in. and is used for lowering the water in the cask area for cask loading. This line contains a pipe connected to the outside of the cask area wall that provides a siphon break. A postulated break in the exterior line from seismic activity would lower the water level in the SFP to a minimum of 13 ft. 10 in. above the top of the fuel and would not pose a rapid drain-down threat.

The evaluation found no drain-down paths that would meet the Rapid Drain-Down criteria. Therefore, the SWEL 2 list (Attachment 5), includes the six (6) selected items prior to this evaluation.

5 SEISMIC WALKDOWNS AND AREA WALK-BYS The methodology used to complete the walkdowns and area walk-bys complies with the EPRI guidance.

The walkdowns and area walk-bys were performed by the Seismic Walkdown Engineers (SWEs) listed in Section 3.2 in groups of at least two. The SWEs used engineering judgment, based on their experience and training, to identify PASCs. After active discussion of observations and judgments, all issues that were not resolved by consensus of the SWEs were further evaluated as described in Section 5.0 of the EPRI guidance document.

Walkdown results, including observations and PASCs, are documented on the Seismic Walkdown Checklists, and area walk-bys on Area Walk-By Checklists. These Checklists are provided as Attachments 6 and 7, respectively.

5.1 Seismic Walkdown Methodology The SWEL 1 and SWEL 2 lists were combined into one to develop the individual walkdown packages.

Working with the site personnel, the walkdown packages were grouped based on elevation, location and the expected number of SSCs that could be walked down during the scheduled time and date. Two separate inspection teams were utilized; each team consisted of two SWEs, a seismic support engineer and a plant representative. A pre-job brief was performed prior to each day's walkdown activities to ensure team members could perform the task safely and effectively.

U. S. Nuclear Regulatory Commission 3F1 112-06 Page 13 of 16 Seismic walkdowns were performed on each SWEL 1 and 2 item that was accessible at the time of the walkdown effort for CR-3. When SWEL items were inaccessible and an appropriate substitute was not available, the item was documented to be inspected at a future date as detailed in Section 5.6.

The seismic walkdowns focused on identifying PASCs for the SSCs listed on the SWEL using the following criteria for adverse anchorage conditions, adverse seismic spatial interactions, or other adverse seismic conditions:

5.1.1 Adverse Anchorage Conditions Lack of anchorage or inadequate anchorage has been the primary cause for malfunction and failure of equipment during an earthquake. During the walkdown inspection, the anchorage was inspected against specific design details for approximately 50% of the SWEL items that include anchorage.

For all SWEL items with anchorage, a general visual inspection of anchorage was performed to determine if the SSC had indications of the following:

Bent, broken, missing, or loose hardware Corrosion that is more than mild surface oxidation Visible cracks within 1 OD of an anchor Gaps that may exist at the visible parts of the equipment foundation Other potential adverse concerns In cases where the anchorage was inaccessible and a substitution was not possible, an alternate method was used to assess potential degraded, non-conforming, or unanalyzed conditions which included:

A review of previous walkdown packages to validate prior inspection attributes for adequacy A determination whether the local environment could cause the degradation of anchorage or its installation, (e.g., adverse environment conditions):

o Evidence of moisture or relatively high humidity, o

Evidence of corrosion on other nearby components and o Anchorage, and/or indication of vibration that could loosen the fasteners.

A check whether the equipment and its anchorage have been subjected to maintenance or modified since it was last walked down The SWEs used engineering judgment to assess whether the anchorage is potentially vulnerable to seismic failure or malfunction (i.e., PASC). The basis for any judgment used in the assessment was documented in the seismic walkdown checklists.

U. S. Nuclear Regulatory Commission 3F1 112-06 Page 14 of 16 5.1.2 Adverse Seismic Spatial Interactions Seismic spatial interaction is the physical interaction between the SWEL item and a nearby component caused by relative motion between the two during an earthquake.

The walkdown included an inspection of the adjacent and surrounding areas to each SWEL item for adverse seismic interaction conditions which could occur that would affect the capability of the item to perform its intended safety-related functions. The three types of seismic spatial interaction effects considered were: proximity to an item, failure of an SSC and falling on an item, and flexibility of attached lines impacting an item.

5.1.3 Other Adverse Seismic Conditions In addition to adverse anchorage and spatial interaction conditions, other potentially adverse seismic conditions that could challenge the adequacy of SWEL items were also identified when present, such as:

Degraded conditions Loose or missing fasteners that secure internal or external components to equipment Large, heavy components mounted on a cabinet that are not typically included by the original manufacturer Cabinet doors or panels that are not latched or fastened 5.2 Area Walk-By Methodology The focus of the area walk-bys was to identify potentially adverse seismic conditions associated with other SSCs located in the vicinity of the SWEL item (either within the room or for large rooms within approximately 35 ft from the item). The key examination factors that were considered included: anchorage conditions, significantly degraded equipment in the area, a visual assessment of cable/conduit raceways and HVAC ducting, housekeeping items that could cause adverse seismic interaction, seismically induced fire and flooding/spray interactions as described below.

5.2.1 Seismically Induced Fire Interactions The occurrence of a seismic event that could create fire in multiple locations, simultaneously degrade fire suppression capabilities, and as a result prevent mitigation of fire damage to multiple safety-related functions.

During the seismic walkdown, the engineers visually assessed any potential sources of fire (e.g., compressed flammable gas bottles, fuel tanks, other combustible material, etc.) located in the vicinity of the SWEL item to ensure it was adequately restrained.

Additionally, potential interactions were assessed to determine if relative motion of high voltage equipment and adjacent support structures that have different foundations can cause high voltage busbars to short out against the grounded bus duct and cause a fire.

U. S. Nuclear Regulatory Commission 3F1 112-06 Page 15 of 16 5.2.2 Seismically Induced Flood/ Spray Interactions Seismically induced flooding events can potentially cause multiple failures of safety-related systems. Two examples of potential flooding sources are rupture of piping and vessels. Instances of concern include threaded fire protection piping, sprinkler head impact, flexible headers and stiff branch pipes, non-ductile mechanical couplings, seismic anchor motion and failed supports.

As the SWEs performed the walkdowns, they visually assessed the potential sources of water located in the vicinity of the subject SSC to ensure they had adequate support and, therefore, were not likely to be a source of flooding or spray that could adversely affect the subject item.

The items that were identified as potential conditions were documented. Any assessment and disposition of the effects were documented with the subject item. During the walkdown and walk-bys, spray nozzle clearance with nearby lighting was inspected. It was determined that adequate clearances existed.

5.3 Results When conditions were observed during the inspection that were not readily identified as acceptable, they were documented along with an evaluation of the condition using available design information and based on the SWEs' experience. SSCs were initially assessed to be a PASC at the time of the inspection and noted as such on the checklist, or the condition was further discussed before determining if it was a PASC. Non-PASC conditions found during the inspections are those evaluated and determined to not affect the ability of the item to perform its intended safety function during or after design basis ground motion as noted in the Current Licensing Basis. Items not readily evaluated to meet this non-PASC criterion were entered into the CAP for resolution. No SSCs initially assessed to be a PASC were identified as a result of the seismic walkdowns to be a PASC, as they were evaluated and determined by the SWEs and/or by the CAP to meet their seismic design basis requirements and be able to perform their intended safety function.

The inspection of eight SWEL items was not fully completed at the time of the walkdowns because the equipment was energized and/or otherwise secured as detailed in Section 5.6 below.

5.4 Maintenance Assessment The maintenance assessment, as required as part of the 50.54(f) response, was completed by analyzing the number of housekeeping and maintenance issues identified during the walkdowns and area walk-bys. During the walkdowns, a number of relatively minor housekeeping issues were noted. These housekeeping items have either been documented in the Work Management System and/or have been addressed and corrected. These conditions suggest that monitoring and maintenance processes and procedures are adequate, but reflect a temporary condition with the plant being in an extended outage. Assurance of proper housekeeping will be addressed as part of the plant re-start program. None of the identified housekeeping items would have prevented important to safety systems from fulfilling their intended safety function.

U. S. Nuclear Regulatory Commission 3F1 112-06 Page 16 of 16 5.5 Planned or Newly Installed Changes There were no planned or newly installed changes to the plant as a result of implementing the seismic walkdown.

5.6 Inaccessible Items There were eight items that could not be fully inspected during the walkdowns and are listed in the table below. All of these features are due to be inspected by the end of 2014. The features are comprised of electrical equipment with doors or panels that were not opened during the walkdowns because they were either locked and/or represented a danger or sensitivity risk to affect the plant status.

ature.dExpected

-.. ildi.g Completion Battery Charger DPBC-1B Control 01/14/2013 Battery Charger DPBC-1J EFP-3 01/28/2013 480V ES MCC 3AB Auxiliary 02/2014 4160V ES BUS 3A Control Fall 2014 4160V ES BUS 3B Control 02/2013 480V ES MCC 3A Control Fall 2014 480V Unit Substation Control Fall 2014 Engineered Safeguard Auxiliary Relay Rack Control 02/2013 RR2AB The remaining inspections will be completed and the updated report submitted to the NRC by December 31, 2014.

6 LICENSING BASIS EVALUATIONS As no PASCs were identified, the SSCs inspected would have been capable of fulfilling their intended safety function.

7 IPEEE VULNERABILITIES RESOLUTION REPORT No seismic vulnerabilities were identified at CR-3. However, there were several identified plant outliers within the scope of the previously conducted USI A-46 walkdown. The USI A-46 outliers were reported to be resolved in a letter to the NRC dated January 11, 2001.

8 PEER REVIEW REPORT The Peer Review Report is included in Attachment 8.

RESPONSE TO RECOMMENDATION 2.3 SEISMIC WALKDOWN OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM THE FUKUSHIMA DAI-ICHI ACCIDENT : Base List I

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 1 of 35 This list identifies SSCs that come out of Screen #3 and enter Screen #4. Therefore the equipment on this list satisfies the first three screens as defined in the EPRI Guidance. This list is mostly comprised of the USI A-46 SSEL (including IPEEE items), which automatically satisfies the first three screens. Also included on this list are 5 SSCs that were not on the SSEL and were plant suggested SSCs as well as 3 modifications.

These were verified to pass through the first three screens. This list's numbering system is the same as the 899 SSEL with the 5 additional numbers labeled 950-954, therefore there is a total of 907 items on this list.

BASE LIST I FOR CR3 BASE u)

EQUIPMENT ID DESCRIPTION LIST NO 1

ASV-050 EFTB-1 TRIP & THROTTLE 0

2 CEILING CONTROL ROOM CEILING 0

3 MU-012-FT MAKE-UP FLOW TRANSMITTER 0

4 AHF-15A DECAY HEAT CLSD CYCLE PUMP 0

AIR HANDLING UNIT 5

AHF-15B DECAY HEAT CLSD CYCLE PUMP 0

AIR HANDLING UNIT 6

ASV-079 MSDT-20 ISOLATION 0

7 ASV-080 MSDT-19 ISOLATION 0

8 ASV-089 ASDT-12 ISOLATION 0

9 ASV-090 ASDT-13 ISOLATION 0

10 ASV-091 ASDT-14 ISOLATION 0

11 ASV-092 ASDT-15 ISOLATION 0

12 ASV-157 ASDT-16 ISOLATION 0

13 ASV-160 ASDT-18 ISOLATION 0

14 ASV-163 ASDT-17 ISOLATION 0

15 ASV-169 ASDT-13 ISOLATION 0

16 CDV-102 CDT-1 LOWER ISOLATION 0

17 CDV-103 CONDENSATE STORAGE TANK TO 0

EFW PUMPS ISOLATION 18 CDV-290 CDT-1 TO EFP SUCTION 0

19 DFP-2A EDG-1A ENGINE DRIVEN FUEL PUMP 0

20 DFP-2B EDG-1B ENGINE DRIVEN FUEL PUMP 0

21 DFP-3A EDG-1A DC MOTOR DRIVEN FUEL 5

PUMP 22 DFP-3B EDG-1B DC MOTOR DRIVEN FUEL 0

PUMP 23 DHV-095 MAKE-UP TO PZR AUXILIARY SPRAY 0

ISOLATION VALVE 24 DHV-126 MAKE-UP TO PZR AUXILIARY SPRAY 0

ISOLATION VALVE 25 DJHE-1 EMERG DIESEL GEN EGDG-1A 0

STANDBY HTR

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 2 of 35 BASE LIST I FOR CR3 U)

BASE uo EQUIPMENT ID DESCRIPTION LIST NO 26 DJHE-10 EMERG DIESEL GEN EGDG-1A 0

ENGINE DRIVEN WATER PUMP 27 DJHE-2 EMERG DIESEL GEN EGDG-1B 0

STANDBY HTR 28 DJHE-3 EMERG DIESEL GEN EGDG-1A 0

COOLING WATER RADIATOR 29 DJHE-4 EMERG DIESEL GEN EGDG-1 B 0

COOLING WATER RADIATOR 30 DJHE-5 EMERG DIESEL GEN EGDG-1 B 0

COOLING WATER RADIATOR 31 DJHE-6 EMERG DIESEL GEN EGDG-1A AIR 0

COOLANT RADIATOR 32 DJHE-7 EMERG DIESEL GEN EGDG-1A AIR 0

COOLANT RADIATOR 33 DJHE-8 EMERG DIESEL GEN EGDG-1B AIR 0

COOLANT RADIATOR 34 DJHE-9 EMERG DIESEL GEN EGDG-1A AIR 0

COOLANT RADIATOR 35 DJP-1 EMERG DIESEL GEN EGDG-1A 0

ENGINE DRIVEN WTR PP 36 DJP-2 EMERG DIESEL GEN EGDG-1 B 0

ENGINE DRIVEN CLG WTR 37 DJP-5 EMERG DIESEL GEN EGDG-1A 0

ENGINE DRIVEN AIR COOLANT 38 DJP-6 EMERG DIESEL GEN EGDG-1 B 0

ENGINE DRIVEN AIR COOLANT 39 DJT-1 EMERG DIESEL GEN EGDG-1A 0

EXPANSION TANK 40 DJT-2 EMERG DIESEL GEN EGDG-1B 0

EXPANSION TANK 41 DJV-67 THERMOSTATIC 3-WAY BY-PASS 0

VLV FOR EGDG-1A 42 DJV-68 THERMOSTATIC 3-WAY BY-PASS 0

VLV FOR EGDG-1 B 43 DJV-69 THERMOSTATIC 3-WAY BY-PASS 0

VLV FOR EGDG-1A 44 DJV-70 THERMOSTATIC 3-WAY BY-PASS 0

VLV FOR EGDG-1 B 45 EFTB-1 EFP-2 TURBINE DRIVE 0

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 3 of 35 BASE LIST I FOR CR3 cn BASE Ci EQUIPMENT ID DESCRIPTION

<J LIST NO 46 EFV-36 HOTWELL SHUTOFF TO EFP-1 AND 0

EFP-2 47 EGX-1A EMERGENCY DIESEL GENERATOR 0

EGDG-1A EXHAUST MUFFLER 48 EGX-1B EMERGENCY DIESEL GENERATOR 0

EGDG-1A EXHAUST MUFFLER 49 IAV-010 AIR SUPPLY TIE IA TO SA 0

50 IAV-031 AIR SUPPLY BYPASS FOR IADR-1 0

51 IAV-036 IADT-2 ISOLATION 0

52 IAV-037 IADT-1 ISOLATION 0

53 IAV-048 IADT-9 ISOLATION 0

54 IAV-049 IADT-10 ISOLATION 0

55 IAV-050 IADT-8 ISOLATION 0

56 IAV-051 IADT-7 ISOLATION 0

57 IAV-068 IADT-8 ISOLATION 0

58 IAV-069 IADT-7 ISOLATION 0

59 IAV-388 IADT-8 ISOLATION 0

60 IAV-393 IADT-7'ISOLATION 0

61 MSV-297 MSDT-22 ISOLATION VALVE 0

62 MSV-299 MSDT-23 ISOLATION VALVE 0

63 MSV-301 MSDT-24 ISOLATION VALVE 0

64 MSV-303 MSDT-25 ISOLATION VALVE 0

65 MUV-014 MAKEUP PUMPS ISOLATION VALVE 0

TO RCP SEALS 66 MUV-015 MAKEUP PUMPS ISOLATION VALVE 0

TO RCP SEALS 67 MUV-017 MUV-16 BYPASS VALVE 0

68 MUV-034 BYPASS ISOLATION VALVE TO RCP 0

SEALS 69 MUV-035 BYPASS ISOLATION VALVE TO RCP 0

SEALS 70 MUV-273 MU SYSTEM TO PZR AUX. SPRAY 0

ISOLATION VALVE 71 MUV-453 SEAL INJECTION MUFL-3A INLET 0

ISOLATION VALVE 72 MUV-454 SEAL INJECTION MUFL-3A INLET 0

ISOLATION VALVE 73 MUV-455 SEAL INJECTION MUFL-3A OUTLET 0

ISOLATION VALVE

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 4 of 35 BASE LIST 1 FOR CR3 BASE uC EQUIPMENT ID DESCRIPTION

-J LIST NO 74 MUV-456 SEAL INJECTION MUFL-3A OUTLET 0

ISOLATION VALVE 75 MUV-457 SEAL INJECTION MUFL-3B INLET 0

ISOLATION VALVE 76 MUV-458 SEAL INJECTION MUFL-3B INLET 0

ISOLATION VALVE 77 MUV-459 SEAL INJECTION MUFL03B OUTLET 0

ISOLATION VALVE 78 MUV-460 SEAL INJECTION MUFL-3B OUTLET 0

ISOLATION VALVE 79 NGV-312 ADV NITROGEN BACKUP ISOLATION 0

VALVE 80 NGV-324 NITROGEN HEADER VENT VALVE 0

81 SAV-5 AIR SUPPLY ISOLATION BETWEEN 0

SA AND IA 82 SCV-530 SC FROM SW ISOLATION VAVLE 0

83 SCV-534 SW FROM AIR COMPRESSORS 0

ISOLATION VALVE 84 SCV-535 SC TO AIR COMPRESSORS 0

ISOLATION VALVE 85 SCV-536 SC FROM AIR COMPRESSORS 0

ISOLATION VALVE 86 SCV-537 SW FROM AIR COMPRESSORS 0

ISOLATION VALVE 87 SWV-103 SC FROM SW ISOLATION VALVE 0

88 SWV-104 SW TO SC ISOLATION 0

89 MTMC-03 480V ES MCC 3A1 1

90 MTMC-04 480V ES MCC 3A2 1

91 MTMC-05 480V ES MCC 3B1 1

92 MTMC-06 480V ES MCC 3B2 1

93 MTMC-07 480V ES MCC 3AB 1

94 MTMC-08 480V PRESSURIZER HEATER MCC 3A 1

95 MTMC-09 480V PRESSURIZER HEATER MCC 3B 1

96 MTMC-12 480V TURBINE MCC 3A 1

97 MTMC-18 480V REACTOR MCC-3A2 1

98 MTMC-21 480V ES MCC 3A3 1

99 MTMC-22 480V ES MCC 3B3 1

100 DPXS-1 MAN XFER SWITCH FOR POWER TO 2

MU.P-3B AND MUP-5B 101 MTSW-3A 480V TURBINE AUXILIARY BUS A 2

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 5 of 35 BASE LIST 1 FOR CR3 C,,

BASE u)

EQUIPMENT ID DESCRIPTION LIST NO 10I 102 MTSW-3C 480V REACTOR AUXILIARY BUS A 2

103 MTSW-3D 480V REACTOR AUXILIARY BUS B 2

104 MTSW-3F 480V ES BUS 3A 2

105 MTSW-3G 480V ES BUS 3B 2

106 MTSW-3J 480V PLANT AUXILIARY BUS 2

107 MTXS-1 ES MCC 3AB INPUT POWER 2

TRANSFER SWITCH 108 VBXS-1A VITAL BUS TRANSFER SWITCH A 2

109 VBXS-1 B VITAL BUS TRANSFER SWITCH B 2

110 VBXS-1C VITAL BUS TRANSFER SWITCH C 2

111 VBXS-1 D VITAL BUS TRANSFER SWITCH D 2

112 VBXS-1 E MANUAL TRANSFER SWITCH 2

113 VBXS-3A EFIC VITAL BUS TRANSFER SWITCH A 2

114 VBXS-3B EFIC VITAL BUS TRANSFER SWITCH B 2

115 VBXS-3C EFIC VITAL BUS TRANSFER SWITCH C 2

116 VBXS-3D EFIC VITAL BUS TRANSFER SWITCH D 2

117 MTSW-2C 4160V ES 3A (NORTH) 3 118 MTSW-2D 4160V ES 3A (SOUTH) 3 119 MTSW-2E 4160V ES 3B (NORTH) 3 120 MTSW-2F 4160V ES 3B (SOUTH) 3 121 MUXS-1 4160V ISOLATION SWITCH 3

122 ACDP-51; TRN CONTROL COMPLEX DISTRIBUTION 4

PANEL A TRANSFORMER 123 ACDP-52; TRN CONTROL COMPLEX DISTRIBUTION 4

PANEL B TRANSFORMER 124 ACDP-68; TRN ES DISTRIBUTION PANEL 3AB 4

TRANSFORMER 125 MTSW-3F; TRN 4160/480V ES BUS 3A 4

TRANSFORMER 126 MTSW-3G; TRN 4160/480V ES BUS 3B 4

TRANSFORMER 127 MTSW-3J: TRN 4160/480V PLANT AUXILIARY BUS 4

TRANSFORMER 128 VBTR-1A CONSTANT VOLTAGE 4

TRANSFORMER A 129 VBTR-1 B CONSTANT VOLTAGE 4

TRANSFORMER B 130 VBTR-2E 480/120V REDUNDANT POWER 4

SUPPLY TRANSFORMER 3E

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 6 of 35 BASE LIST 1 FOR CR3 BASE Cn EQUIPMENT ID DESCRIPTION LIST NO 131 VBTR-3E REDUNDANT POWER SUPPLY 4

SOLATRON 3E 132 VBTR-4A REGULATING TRANSFORMER A 4

133 VBTR-4B REGULATING TRANSFORMER B 4

134 VBTR-4C REGULATING TRANSFORMER C 4

135 VBTR-4D REGULATING TRANSFORMER D 4

136 BSP-1A REACTOR BUILDING SPRAY PUMP A 5

137 BSP-1B REACTOR BUILDING SPRAY PUMP B 5

138 CAP-1A BORIC ACID PUMP A 5

139 CAP-1B BORIC ACID PUMP B 5

140 CHP-1A CHILLED WATER PUMP A 5

141 CHP-1B CHILLED WATER PUMP B 5

142 DCP-1A DECAY HEAT CLOSED CYCLE 5

COOLING PUMP A 143 DCP-1B DECAY HEAT CLOSED CYCLE 5

COOLING PUMP B 144 DFP-1A AC MOTOR DRIVEN FUEL OIL 5

TRANSFER PUMP A 145 DFP-1B AC MOTOR DRIVEN FUEL OIL 5

TRANSFER PUMP B 146 DFP-1C DC MOTOR DRIVEN FUEL OIL 5

TRANSFER PUMP C 147 DFP-1 D DC DRIVEN FUEL OIL TRANSFER 5

PUMP D 148 DHP-1A DECAY HEAT PUMP A 5

149 DHP-1B DECAY HEAT PUMP B 5

150 EFP-1 MOTOR DRIVEN EMERGENCY 5

FEEDWATER PUMP 151 EFP-2 TURBINE DRIVEN EMERGENCY 5

FEEDWATER PUMP 152 FWP-7 AUXILIARY FEEDWATER PUMP 7 5

153 GWP-1A CONDENSATE INJECTION PUMP 3A 5

154 GWP-1B CONDENSATE INJECTION PUMP 3B 5

155 MUP-1A MAKE-UP AND PURIFICATION PUMP 3A 5

156 MUP-1B MAKE-UP AND PURIFICATION PUMP 3B 5

157 MUP-1C MAKE-UP AND PURIFICATION PUMP 3C 5

158 MUP-2A MUP-1A MAIN OIL PUMP 5

159 MUP-2B MUP-1B MAIN OIL PUMP 5

160 MUP-2C MUP-1C MAIN OIL PUMP 5

161 MUP-4A MUP-1A GEAR OIL PUMP 5

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 7 of 35 BASE LIST 1 FOR CR3 BASE U)

EQUIPMENT ID DESCRIPTION

-J LIST NO 162 MUP-4B MUP-1B GEAR OIL PUMP 5

163 MUP-4C MUP-1C GEAR OIL PUMP 5

164 SCP-3 SECONDARY SERCIVES CLOSED 5

CYCLE COOLING BOOSTER PUMP 165 SFP-1A SPENT FUEL COOLANT PUMP A 5

166 SFP-1B SPENT FUEL COOLANT PUMP B 5

167 SSP-4A CONDENSER A HOTWELL SAMPLE 5

EXTRACTION PUMP 168 SSP-4B CONDENSER B HOTWELL SAMPLE 5

EXTRACTION PUMP 169 SSP-4C CONDENSER AB HOTWELL 5

CROSSOVER SAMPLE EXTRACTION PUMP 170 SSP-4D LP HEATER SAMPLE EXTRACTION 5

PUMP 171 SWP-1A EMERGENCY NUCLEAR SERVICE 5

CCC PUMP 3A 172 SWP-1B EMERGENCY NUCLEAR SERVICE 5

CCC PUMP 3B 173 SWP-1C NORMAL NUCLEAR SERVICE 5

CLOSED CYCLE COOLING PUMP 174 SWP-2A NUCLEAR SERVICE BOOSTER PUMP A 5

175 SWP-2B NUCLEAR SERVICE BOOSTER PUMP B 5

176 CDP-1A CONDENSATE PUMP A 6

177 CDP-1B CONDENSATE PUMP B 6

178 RWP-1 NORMAL NUCLEAR SERVICES SEA 6

WATER PUMP MOTOR COLLER 179 RWP-2A NUCLEAR SERVICE SEAWATER 6

PUMP 3A 180 RWP-2B NUCLEAR SERVICE SEAWATER 6

PUMP 3B 181 RWP-3A DECAY HEAT SERVICE SEA WATER 6

PUMP 3A 182 RWP-3B DECAY HEAT SERVICE SEA WATER 6

PUMP 3B 183 CAV-057 BORIC ACID PUMP TO MAKE-UP 7

TANK 184 CAV-060 BORIC ACID PUMP TO MAKE-UP 7

TANK 185 CDV-100 EMERGENCY MAKEUP TO HOTWELL 7

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 8 of 35 BASE LIST 1 FOR CR3 BASE CO EQUIPMENT ID DESCRIPTION LIST NO 186 CHV-068 SERVICE WATER TO CHHE-1A 7

187 CHV-068-POS CHV-68 CONTROL 7

188 CHV-069 SERVICE WATER TO CHHE-1 B 7

CONTROL VALVE 189 CHV-069-POS CHV-69 CONTROL 7

190 CHV-100 AHHE-44 (EFIC ROOMS) CONTROL 7

VALVE 191 CHV-100-POS CHV-100 CONTROL 7

192 CHV-113 AHHE-43 (EFIC ROOMS) CONTROL 7

VALVE 193 CHV-113-POS CHV-113 CONTROL 7

194 DCV-186 NITROGEN SUPPLY TO DCT-1A 7

195 DCV-188 NITROGEN SUPPLY TO DCT-1B 7

196 DCV-190 DCT-1A OVERPRESSURE CONTROL 7

VALVE 197 DCV-191 DCT-1B OVERPRESSURE CONTROL 7

VAVLE 198 EGV-52 EDG A AIR START 3-WAY VALVE 7

199 EGV-53 EDG A AIR START 3-WAY VALVE 7

200 EGV-54 EDG B AIR START 3-WAY VALVE 7

201 EGV-55 EDG B AIR START 3-WAY VALVE 7

202 GWV-196 GWP-1A CONTROL VALVE 7

203 MSV-025 ATMOSPHERIC DUMP VALVE A 7

204 MSV-026 ATMOSPHERIC DUMP VALVE B 7

205 MSV-033 MAIN STEAM LINE A-2 SAFETY 7

VALVE 206 MSV-034 MAIN STEAM LINE A-1 SAFETY 7

VALVE 207 MSV-035 MAIN STEAM LINE B-1 SAFETY 7

VALVE 208 MSV-036 MAIN STEAM LINE B-2 SAFETY 7

VALVE 209 MSV-037 MAIN STEAM LINE A-2 SAFETY 7

VALVE 210 MSV-038 MAIN STEAM LINE A-1 SAFETY 7

VALVE 211 MSV-039 MAIN STEAM LINE B-1 SAFETY 7

VALVE 212 MSV-040 MAIN STEAM LINE A-1 SAFETY 7

VALVE

U. S. Nuclear Regulatory Commission 3F1112-06, Attachment 1 Page 9 of 35 BASE LIST 1 FOR CR3 BASEu)

EQUIPMENT ID DESCRIPTION

<)

LIST NO 213 MSV-041 MAIN STEAM LINE B-2 SAFETY 7

VALVE 214 MSV-042 MAIN STEAM LINE A-2 SAFETY 7

VALVE 215 MSV-043 MAIN STEAM LINE A-1 SAFETY 7

VALVE 216 MSV-044 MAIN STEAM LINE B-1 SAFETY 7

VALVE 217 MSV-045 MAIN STEAM LINE B-2 SAFETY 7

VALVE 218 MSV-046 MAIN STEAM LINE A-2 SAFETY 7

VALVE 219 MSV-047 MAIN STEAM LINE B-1 SAFETY 7

VALVE 220 MSV-048 MAIN STEAM LINE B-2 SAFETY 7

VALVE 221 MSV-411 MAIN STEAM LINE A-2 ISOLATION 7

VALVE 222 MSV-412 MAIN STEAM LINE A-1 ISOLATION 7

VALVE 223 MSV-413 MAIN STEAM LINE B-1 ISOLATION 7

VALVE 224 MSV-414 MAIN STEAM LINE B-2 ISOLATION 7

VALVE 225 MU-003-POC MUV-51 CONTROL 7

226 MU-012-POC MUV-108 CONTROL 7

227 MU-015-POC MUV-16 CONTROL 7

228 MU-025-POC MUV-31 CONTROL 7

229 MUV-016 RCP SEAL INJECTION FLOW 7

CONTROL VALVE 230 MUV-031 MAKE-UP FLOW CONTROL VALVE 7

231 MUV-049 LET-DOWN FLOW ISOLATION VALVE 7

232 MUV-050 LET-DOWN BLOCK OROFICE 7

ISOLATION VALVE 233 MUV-051 LET-DOWN FLOW CONTROL VALVE 7

234 MUV-090 LETDOWN FILTER MUFL-1 B TO MUT-7 1 ISOLATION VALVE 235 MUV-091 LETDOWN FILTED MUFL-1A TO MUT-7 1 ISOLATION VALVE

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 10 of 35 BASE LIST 1 FOR CR3 BASE Cn EQUIPMENT ID DESCRIPTION LIST NO C..)

236 MUV-096 LETDOWN FILTER MUFL-1A INLET 7

ISOLATION VALVE 237 MUV-097 LETDOWN FILTER MUFL-1B INLET 7

ISOLATION VALVE 238 MUV-103 BORIC ACID PUMP TO MAKE-UP 7

TANK ISOLATION VALVE 239 MUV-108 BORIC ACID PUMP TO MUT-1 FLOW 7

CONTROL VALVE 240 MUV-116 DEMINERALIZER MUDM-1A 7

ISOLATION VALVE TO LETDOWN FILTER 241 MUV-117 DEMINARALIZER MUDM-1B 7

ISOLATION VALVE TO LETDOWN FILTER 242 MUV-124 ISOLATION VALVE TO PURIFICATION 7

DEMINERALIZER MUDM-1A 243 MUV-133 ISOLATION VALVE TO 7

DEMINERALIZER MUDM-1B 244 MUV-200 LETDOWN ISOLATION VALVE TO 7

DEMINERALIZER MUDM-1A 245 MUV-201 LETDOWN ISOLATION VALVE TO 7

DEMINERALIZER MUDM-1B 246 MUV-242 PREFILTER MUFL-2A INLET 7

ISOLATION VALVE 247 MUV-243 PREFILTER MUFL-2A DISCHARGE 7

ISOLATION VALVE 248 MUV-244 PREFILTER MUFL-2B DISCHARGE 7

ISOLATION VALVE 249 MUV-245 PREFILTER MUFL-2B INLET 7

ISOLATION VALVE 250 MUV-253 REACTOR COOLANT PUMP SEAL 7

BLEEDOFF ISOLATION VALVE 251 RWV-150 RW RECIRCULATION FLOW 7

CONTROL VALVE 252 SCV-095 IAP-1A & IAHE-1A COOLING WATER 7

CONTROL VALVE 253 SCV-096 lAP-1 B & IAHE-1 B COOLING WATER 7

CONTROL VALVE 254 SWV-012 SEAL RETURN COOLERS SUPPLY 7

ISOLATION VALVE

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 11 of 35 BASE LIST 1 FOR CR3 BASE u

EQUIPMENT ID DESCRIPTION LIST NO 255 SWV-035 INLET ISOLATION FOR AHHE-31A 7

AND AHHE-32A 256 SWV-037 INLET ISOLATION FOR AHHE-31 B 7

AND AHHE-32B 257 SWV-039 INLET ISOLATION FOR AHHE-31C 7

AND AHHE-32C 258 SWV-041 ISOLATION FROM OUTLETS OF 7

AHHE-31A AND AHHE-32A 259 SWV-043 OUTLET ISOLATION FOR AHHE-31 B 7

AND AHHE-32B 260 SWV-045 OUTLET ISOLATION FOR AHHE-31 C 7

AND AHHE-32C 261 SWV-047 ISOLATION TO MUHE-1 B AND WDHE-1 7

262 SWV-048 LETDOWN COOLER 3B INLET 7

ISOLATION VALVE 263 SWV-049 LETDOWN COOLER 3B DISCHARGE 7

ISOLATION VALVE 264 SWV-050 OUTLET ISOLATION FOR MUHE-1A 7

AND WDHE-1 265 SWV-151 IC FROM RB FAN COOLERS 7

ISOLATION VALVE 266 SWV-152 IC TO RB RAN COOLERS ISOLATION 7

VALVE 267 SWV-353 SW TO RB FAN COOLERS 7

ISOLATION VALVE 268 SWV-354 SW FROM RB FAN COOLERS 7

ISOLATION VALVE 269 SWV-355 IC FROM RB FAN COOLERS 7

ISOLATION VALVE 270 ARV-48 VACUUM BREAKER RELIEF FOR 8

CDHE-4A 271 ARV-49 VACUUM BREAKER RELIEF FOR 8

CDHE-4B 272 ASV-005 EFTB-1 STEAM ADMISSION 8

273 ASV-204 EFTB-1 STEAM ADMISSION 8

274 CAV-126 RC LETDOWN SAMPLE INSIDE 8

PENETRATION ISOLATION VALVE 275 CFV-5 CORE FLOOD TANK A DISCHARGE 8

ISOLATION

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 12 of 35 BASE LIST I FOR CR3 BASE u)

EQUIPMENT ID DESCRIPTION LIST NO 276 CFV-6 CORE FLOOD TANK B DISCHARGE 8

ISOLATION 277 EFV-01 HOTWELL ISOLATION TO TURBINE 8

DRIVEN EFP-2 278 EFV-02 HOTWELL ISOLATION TO MOTOR 8

DRIVEN EFP-1 279 EFV-03 CDT-1 AND EFT-2 ISOLATION TO EFP-1 8

280 EFV-04 CDT-1 AND SFT-2 ISOLATION TO EFP-2 8

281 FWV-14 FEEDWATER PUMP A SUCTION 8

ISOLATION VALVE 282 FWV-15 FEEDWATER PUMP B SUCTION 8

ISOLATION VALVE 283 FWV-28 FEEDWATER PUMPS DISCHARGE 8

CROSSTIE ISOLATION VALVE 284 FWV-29 OTSG B MAIN BLOCK VALVE 8

285 FWV-30 OTSG A MAIN BLOCK VALVE 8

286 FWV-31 OTSG A LOW LOAD BLOCK VALVE 8

287 FWV-32 OTSG B LOW LOAD BLOCK VALVE 288 FWV-33 OTSG B STARTUP BLOCK VALVE 8

289 FWV-36 OTSG A STARTUP BLOCK VALVE 8

290 MSV-055 OTSG A TO EFTB-1 STOP CHECK 8

291 MSV-056 OTSG B TO EFTB-1 STOP CHECK 8

292 MUV-018 RCP SEAL ISOLATION VALVE 8

293 MUV-023 HP INJ CONTROL VALVE TO RCS 8

INLET LINES LOOP A 294 MUV-024 HP INJ CONTROL VALVE TO RCX 8

INLET LINES LOOP A 295 MUV-025 HP INJ CONTROL VALVE TO RCS 8

INLET LINES LOOP B 296 MUV-026 HP INJ CONTROL VALVE TO RCS 8

INLET LINES LOOP B 297 MUV-027 HPI CONTROL VALVE TO RCS LOOP A 8

298 MUV-058 HI PRESS INJECTION SUCTION 8

FROM BWST 299 MUV-073 BWST TO MUP-1A & MUP-1B 8

ISOLATION VALVE 300 MUV-100 MUFL-1A/MUFL-1 B BYPASS VALVE 8

301 MUV-1 12 LETDOWN TO MUT-1 OR WD 8

DIVERTER 302 MUV-194 MUFL-2A/MUFL-2B BYPASS VALVE 8

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 13 of 35 BASE LIST I FOR CR3 BASE EQUIPMENT ID DESCRIPTION LIST NO 303 RCV-1 1 PRESSURIZER BLOCK VALVE 8

304 RCV-13 PRESSURIZER INLET AUXILIARY 8

SPRAY ISOLATION VALVE 305 RCV-53 DH TO PRESSURIZER AUXILIARY 8

SPRAY ISOLATION VALVE 306 AH-033-ASV AHD-1, AHD-1 D, AHD-2, & AHD-3 8

CONTROL 307 AH-033-SV AHD-1, AHD-1 D, AHD-2, & AHD-3 8

CONTROL 308 AH-194-SV AHD-13 & AHD-14 CONTROL 8

309 AH-195-SV AHD-15 & AHD-16 CONTROL 8

310 AH-196-SV AHD-1, AHD-1 D, AHD-2, & AHD-3 8

CONTROL 311 AH-199-SV AHD-4 & AHD-5 CONTROL 8

312 AH-200-SV AHD-6 & AHD-7 CONTROL 8

313 AH-246-SV AHD-12 CONTROL 8

314 AH-250-SV AHD-17 & AHD-22 POST-ACCIDENT 8

CONTROL 315 AH-310-SV AHD-87 CONTROL 8

316 AH-311-SV AHD-88 CONTROL 8

317 AH-355-SV AHD-63 CONTROL 8

318 AH-356-SV AHD-63 CONTROL 8

319 AH-357-SV AHD-61 CONTROL 8

320 AH-358-SV AHD-62 CONTROL 8

321 AH-365-SV AHD-60 CONTROL 8

322 AH-366-SV AHD-60 CONTROL 8

323 AH-367-SV AHD-58 CONTROL 8

324 AH-368-SV AHD-59 CONTROL 8

325 AH-381-ASV AHD-1, AHD-1 D, AHD-2, & AHD-3 8

CONTROL 326 AH-381-SV AHD-1, AHD-1 D, AHD-2, & AHD-3 8

CONTROL 327 AH-517-SV A-SOLENOID VALVE 8

328 AH-518-SV B-SOLENOID VALVE 8

329 AH-648-ASV AHD-1, AHD-1 D, AHD-2, & AHD-3 8

CONTROL 330 AH-648-SV AHD-1, AHD-1 D, AHD-2, & AHD-3 8

CONTROL 331 AH-649-ASV AHD-1, AHD-1 D, AHD-2, & AHD-3 8

CONTROL

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 14 of 35 BASE LIST 1 FOR CR3 BASE U)

EQUIPMENT ID DESCRIPTION LIST NO 332 AH-953-SV AHD-99 CONTROL 8

333 AH-966-SV AHD-1 CONTROL 8

334 AH-967-SV AHD-1 & AHD-1 D CONTROL 8

335 AH-968-SV AHD-2 CONTROL 8

336 AH-969-SV AHD-3 CONTROL 8

337 AH-970-SV AHD-99 CONTROL 8

338 AH-971-SV AHD-12 CONTROL 8

339 CAV-002 PZR AND LETDOWN OUTSIDE 8

PENETRATION SAMPLE ISOLATION 340 CAV-057-SV CAV-57 CONTROL 8

341 CAV-060-SV CAV-60 CONTROL 8

342 CHV-090 NORMAL COOLING TO AHHE-44 8

ISOLATION VALVE 343 CHV-097 APPX. R COOLING FROM AHHE-44 8

ISOLATION VALVE 344 CHV-100-SV CHV-100 CONTROL 8

345 CHV-101 NORMAL COOLING FROM AHHE-44 8

ISOLATION VALVE 346 CHV-108 NORMAL COOLING TO AHHE-43 8

ISOLATION VALVE 347 CHV-113-SV CHV-113 CONTROL 8

348 EFV-55 EFP-2 TO OTSG B CONTROL VALVE 8

349 EFV-56 EFP-2 TO OTSG A CONTROL VALVE 8

350 EFV-57 EFP-1 TO OTSG B CONTROL VALVE 8

351 EFV-58 EFP-1 TO OTSG A CONTROL VALVE 8

352 EGV-36 EDG A AIR START SOLENOID VALVE 8

353 EGV-37 EDG A AIR START SOLENOID VALVE 8

354 EGV-40 EDG B AIR START SOLENOID VALVE 8

355 EGV-41 EDG B AIR START SOLENOID VALVE 8

356 EGV-56 EDG A AIR START VALVE 8

357 EGV-57 EDG A AIR START VALVE 8

358 EGV-58 EDG B AIR START VALVE 8

359 EGV-59 EDG B AIR START VALVE 8

360 IAV-188 IAP-1A LOADER/UNLOADER VALVE 8

361 IAV-189 lAP-1 B LOADER/UNLOADER VALVE 8

362 MSV-411/412-SV5 MSV-411 & MSV-412 CONTROL 8

363 MSV-411/412-SV6 MSV-411 & MSV-412 CONTROL 8

364 MSV-411-SV1 MSV-411 CONTROL 8

365 MSV-411-SV2 MSV-411 CONTROL 8

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 15 of 35 BASE LIST I FOR CR3 uB EQUIPMENT ID DESCRIPTION LIST NO 366 MSV-411-SV4 MSV-411 CONTROL 8

367 MSV-412-SV1 MSV-412 CONTROL 8

368 MSV-412-SV2 MSV-412 CONTROL 8

369 MSV-412-SV4 MSV-412 CONTROL 8

370 MSV-413/414-SV5 MSV-413 & MSV-414 CONTROL 8

371 MSV-413/414-SV6 MSV-413 & MSV-414 CONTROL 8

372 MSV-413-SV1 MSV-413 CONTROL 8

373 MSV-413-SV2 MSV-413 CONTROL 8

374 MSV-413-SV4 MSV-413 CONTROL 8

375 MSV-414-SV1 MSV-414 CONTROL 8

376 MSV-414-SV2 MSV-414 CONTROL 8

377 MSV-414-SV4 MSV-414 CONTROL 8

378 MU-003-SV MUV-51 AIR FAIL LOCK 8

379 MU-015-SV MUV-16 AIR FAIL LOCK 8

380 MU-025-SV MUV-31 AIR FAIL LOCK 8

381 MUV-049-SV MUV-49 CONTROL 8

382 MUV-050-SV MUV-50 CONTROL 8

383 MUV-103-SV MUV-103 CONTROL 8

384 RCV-10 PRESSURIZER POWER OPERATED 8

RELIEF VALVE 385 SCV-099 IAP-1A & IAHE-1A COOLING WATER 8

ISOLATION VALVE 386 SCV-100 IAP-1B &IAHE-1C COOLING WATER 8

ISOLATION VALVE 387 SWV-012-SV1 SWV-12 CONTROL 8

388 SWV-012-SV2 SWV-12 CONTROL 8

389 SWV-151-SV SWV-151 CONTROL 8

390 SWV-1 52-SV1 SWV-152 CONTROL 8

391 SWV-1 52-SV2 SWV-1 52 CONTROL 8

392 SWV-353-SV1 SWV-353 CONTROL 8

393 SWV-353-SV2 SWV-353 CONTROL 8

394 SWV-354-SV1 SWV-354 CONTROL 8

395 SWV-354-SV2 SWV-354 CONTROL 8

396 SWV-355-SV SWV-355 CONTROL 8

397 AHF-01A REACTOR BUILDING AIR HANDLING 9

FAN A 398 AHF-01B REACTOR BUILDING AIR HANDLING 9

FAN B 399 AHF-01C REACTOR BUILDING AIR HANDLING 9

FAN C

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 16 of 35 BASE LIST I FOR CR3 BASE uw EQUIPMENT ID DESCRIPTION LIST NO.

400 AHF-17A CONTROL COMPLEX NORMAL 9

SUPPLY FAN A 401 AHF-17B CONTROL COMPLEX NORMAL 9

SUPPLY FAN B 402 AHF-19A CONTROL COMPLEX RETURN FAN A 9

403 AHF-19B CONTROL COMPLEX RETURN FAN B 9

404 AHF-22A DIESEL GENERATOR ROOM A 9

SUPPLY 405 AHF-22B DIESEL GENERATOR ROOM A 9

SUPPLY 406 AHF-22C DIESEL GENERATOR ROOM B 9

SUPPLY 407 AHF-22D DIESEL GENERATOR ROOM B 9

SUPPLY 408 AH-196-POS1 AHD-1 CONTROL 10 409 AH-196-POS2 AHD-2 CONTROL 10 410 AH-196-POS3 AHD-3 CONTROL 10 411 AH-196-POS4 AHD-1D CONTROL 10 412 AHD-01 CONTROL COMPLEX MAKE-UP AIR 10 413 AHD-01D CONTROL COMPLEX MAKE-UP AIR 10 414 AHD-02 PNEUMATIC RELIEF TO 10 ATMOSPHERE 415 AHD-03 CONTROL COMPLEX FANS INTAKE 10 416 AHD-04 AHF-17A INTAKE 10 417 AHD-05 AHF-17A DISCHARGE 10 418 AHD-06 AHF-17B INTAKE 10 419 AHD-07 AHF-17B DISCHARGE 10 420 AHD-12 SUPPLY TO CHEMLAB 10 421 AHD-13 AHF-19A INTAKE 10 422 AHD-14 AHF-19A DISCHARGE 10 423 AHD-15 AHF-19B INTAKE 10 424 AHD-16 AHF-19B DISCHARGE 10 425 AHD-17 95' ELEVATION RETURN 10 426 AHD-22 SECONTARY PLANT LAB HOOD 10 SUPPLY 427 AHD-58 AHF-22C DISCHARGE 10 428 AHD-59 AHF-22D DISCHARGE 10 429 AHD-60 EGDG-1B INTER-ROOM 10 430 AHD-61 AHF-22A DISCHARGE 10 431 AHD-62 AHF-22B DISCHARGE 10

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 17 of 35 BASE LIST I FOR CR3 BASE EQUIPMENT ID DESCRIPTION

<U LIST NO 432 AHD-63 EGDG-1A INTER-ROOM 10 433 AHD-87 AHF-8A DISCHARGE 10 434 AHD-88 AHF-8B DISCHARGE 10 435 AHD-99 VENTILLATION EQUIPMENT ROOM 10 SUPPLY AIR 436 AHF-08A SPENT FUEL COOLANT PUMP A AIR 10 HANDLING 437 AHF-08B SPENT FUEL COOLANT PUMP B AIR 10 HANDLING 438 AHF-54A EFIC ROOMS COOLING A 9

439 AHF-54B EFIC ROOMS COOLING B 10 440 AHHE-04A HEATING UNIT A FOR CONTROL 10 COMPLEX 441 AHHE-04B HEATING UNIT B FOR CONTROL 10 COMPLEX 442 AHHE-05A COOLING UNIT A FOR CONTROL 10 COMPLEX 443 AHHE-05B COOLING UNIT B FOR CONTROL 10 COMPLEX 444 AHHE-13A COOLING COILS FOR REACTOR 10 BUILDING PENETRATIONS A, 445 AHHE-13B COOLING COILS FOR REACTOR 10 BUIDLING PENETRATION B 446 AHHE-29A COOLING UNIT FOR SFP-1A 10 447 AHHE-29B COOLING UNIT FOR SFP-1B 10 448 AHHE-30A COOLING UNIT FOR DCP-1A 10 449 AHHE-30B COOLING UNIT FOR DCP-1B 10 450 AHHE-31A COOLING UNIT A FOR REACTOR 10 BUILDING 451 AHHE-31B COOLING UNIT B FOR REACTOR 10 BUILDING 452 AHHE-31C COOLING UNIT C FOR REACTOR 21 BUILDING 453 AHHE-32A MOTOR COOLER FOR AHF-1A 10 454 AHHE-32B MOTOR COOLER FOR AHF-1B 10 455 AHHE-32C MOTOR COOLER FOR AHF-1C 10 456 AHHE-43 COOLING UNIT A FOR EFIC ROOMS 10 457 AHHE-44 COOLING UNIT B FOR EFIC ROOMS 10 458 CHHE-1A CONTROL COMPLEX CHILLER A 11 459 CHHE-1B CONTROL COMPLEX CHILLER B 11

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 18 of 35 BASE LIST I FOR CR3 U,

BASE u

EQUIPMENT ID DESCRIPTION LIST NO

-J 4&

B 460 AH-506-PS AHP-1A & AHP-1 B CONTROL 12 461 AH-508-PS AHP-1C & AHP-1D CONTROL 12 462 AHP-01A CONTROL COMPLEX HVAC AIR 12 COMPRESSOR A 463 AHP-01B CONTROL COMPLEX HVAC AIR 12 COMPRESSOR B 464 AHP-01C CONTROL COMPLEX HVAC AIR 12 COMPRESSOR 465 AHP-01D CONTROL COMPLEX HVAC AIR 12 COMPRESSOR D 466 IAP-1A INSTRUMENT AIR COMPRESSOR A 12 467 lAP-1B INSTRUMENT AIR COMPRESSOR B 12 468 SAP-1A STATION AIR COMPRESSOR 1A 12 469 SAP-1B STATION AIR COMPRESSOR 1B 12 470 WDP-1A WASTE GAS COMPRESSOR A 12 471 WDP-1B WASTE GAS COMPRESSOR B 12 472 ACDP-05 DIESEL ROOM 480 VOLT 14 DISTRIBUTION PANEL 3A 473 ACDP-06 DIESEL ROOM 480 VOLT 14 DISTRIBUTION PANEL 3B 474 ACDP-51 CONTROL COMPLEX DISTRIBUTION 14 PANEL A 475 ACDP-52 CONTROL COMPLEX DISTRIBUTION 14 PANEL B 476 ACDP-68 ES DISTRIBUTION PANEL 3AB 14 477 DPDP-1A 250/125V DC MAIN PANEL 3A 14 478 DPDP-1B 250/125V MAIN PANEL 3B 14 479 DPDP-1C 250/125V DC MAIN PANEL 3C 14 480 DPDP-3A 250/125V DC TURBINE BUILDING 14 PANEL A 481 DPDP-3B 250/125V DC TURBNIE BUILDING 14 PANEL B 482 DPDP-4B CONTROL COMPLEX DC PANEL 3B 14 483 DPDP-5A 250/125V DC ES PANEL A 14 484 DPDP-5B 250/125V DC ES PANEL B 14 485 DPDP-6A 250/125V DC ES DIESEL 14 GENERATOR PANEL A 486 DPDP-6B 250/125V DC ES DIESEL 14 GENERATOR PANEL B

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 19 of 35 BASE LIST I FOR CR3 Cn BASE un EQUIPMENT ID DESCRIPTION LIST NO 487 DPDP-8A 250/125V DC ESSENTIAL SERVICES 14 PANEL A 488 DPDP-8B 250/125V DC ESSENTIAL SERVICES 14 PANEL B 489 DPDP-8C 250/125 VOLT DC EFIC PANEL D 14 490 DPDP-8D 250/125 VOLT DC EFIC PANEL D 14 491 DPDS-1A BATTERY 3A DISCONNECT SWITCH 14 492 DPDS-1B BATTERY 3B DISCONNECT SWITCH 14 493 DPDS-1C BATTERY 3C DISCONNECT SWITCH 14 494 DPXS-1C DPBC-I INPUT POWER TRANSFER 14 SWITCH 495 VBDP-01 REGULATED INSTRUMENT BUS 3A 14 496 VBDP-02 REGULATED INSTRUMENT BUS 3B 14 497 VBDP-03 VITAL BUS A 14 498 VBDP-04 VITAL BUS B 14 499 VBDP-05 VITAL BUS C 14 500 VBDP-06 VITAL BUS D 14 501 VBDP-07 COMPUTER 120 VAC DISTRIBUTION 14 PANEL 502 VBDP-08 EFIC VITAL BUS A 14 503 VBDP-09 EFIC VITAL BUS C 14 504 VBDP-10 EFIC VITAL BUS B 14 505 VBDP-1 1 EFIC VITAL BUS D 14 506 VBDP-12 120 VOLT REGULATED 14 DISTRIBUTION PANEL 507 VBDP-13 120 VOLT REGULATED 14 DISTRIBUTION PANEL 508 VBDP-14 120 VOLT REGULATED 14 DISTRIBUTION PANEL 509 VBDP-15 120 VOLT REGULATED 14 DISTRIBUTION PANEL 510 DPBA-1A 250/125V BATTERY A 15 511 DPBA-1B 250/125V BATTERY B 15 512 DPBA-1C 250/125V BATTERY C 15 513 DPBC-1A BATTERY CHARGER A 16 514 DPBC-1B BATTERY CHARGER B 16 515 DPBC-1C BATTERY CHARGER C 16 516 DPBC-1D BATTERY CHARGER D 16 517 DPBC-1E BATTERY CHARGER E 16 518 DPBC-1F BATTERY CHARGER F 16

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 20 of 35 BASE LIST I FOR CR3 BASE u)

EQUIPMENT ID DESCRIPTION

-J LIST NO-519 DPBC-1G BATTERY CHARGER G 16 520 DPBC-1H BATTERY CHARGER H 16 521 DPBC-1I BATTERY CHARGER I 16 522 VBIT-1A DUAL INPUT INVERTER 3A 16 523 VBIT-1B DUAL INPUT INVERTER 3B 16 524 VBIT-1C DUAL INPUT INVERTER 3C 16 525 VBIT-1D DUAL INPUT INVERTER 3D 16 526 VBIT-1E DUAL INPUT INVERTER 3E 16 527 EGDG-1A DIESEL GENERATOR A 17 528 EGDG-1B DIESEL GENERATOR B 17 529 CA-i 1-LT BORIC ACID STORAGE TANK A 18 530 CA-13-LT BORIC ACID STORAGE TANK B 18 531 CD-067-LT1 CONDENSATE STORAGE TANK 18 LEVEL 532 CD-100-LT CONDENSER CDHE-4A HOTWELL 18 LEVEL 533 CD-101-LT CONDENSER CDHE-4B HOTWELL 18 LEVEL 534 CH-378-PT CHV-68 CONTROL OF CHHE-1A 18 535 CH-379-PT CHV-69 CONTROL OF CHHE-1B 18 536 DC-05-PT DCP-1A DISCHARGE 18 537 DC-06-PT DCP-1B DISCHARGE PRESSURE 18 538 DC-50-LT DC SURGE TANK DCT-1A LEVEL 18 539 DC-54-LT DC SURGE TANK DCT-1B LEVEL 18 540 DF-1-LS DIESEL GENERATOR FUEL OIL DAY 18 TANK A LEVEL SWITCH 541 DF-2-LS DIESEL GENERATOR FUEL OIL DAY 18 TANK A LEVEL SWITCH 542 DF-3-LS DIESEL GENERATOR FUEL OIL DAY 18 TANK B LEVEL SWITCH 543 DF-4-LS DIESEL GENERATOR FUEL OIL DAY 18 TANK B LEVEL SWITCH 544 DH-07-LT BORATED WATER STORAGE TANK 18 LEVEL 545 DH-07-LT1 BORATED WATER STORAGE TANK 18 LEVEL 546 DH-37-LT BORATED WATER STORAGE TANK 18 LEVEL 547 EF-23-FT EFP-1 TO OTSG-B FLOW 18 548 EF-24-FT EFP-2 TO OTSG B FLOW 18

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 21 of 35 BASE LIST I FOR CR3 BASE EQUIPMENT ID DESCRIPTION LIST NO-549 EF-25-FT EFP-1 TO OTSG A FLOW 18 550 EF-26-FT EFP-2 TO OTSG A FLOW 18 551 EF-98-LT EMERGENCY FEEDWATER TANK 18 LEVEL 552 EF-99-LT EMERGENCY FEEDWATER TANK 18 LEVEL 553 IA-04-PT INSTRUMENT AIR HEADER 18 PRESSURE 554 IA-12-PS INSTRUMENT AIR COMPRESSOR 18 LOAD/UNLOAD PRESS SWITCH 555 MS-106-PT MAIN STEAM LINE A-2 PRESSURE 18 TRANSMITTER 556 MS-107-PT MAIN STEAM LINE A-1 PRESSURE 18 TRANSMITTER 557 MS-108-PT MAIN STEAM LINE A-2 PRESSURE 18 TRANSMITTER 558 MS-109-PT MAIN STEAM LINE A-1 PRESSURE 18 TRANSMITTER 559 MS-110-PT MAIN STEAM LINE B-2 PRESSURE 18 TRANSMITTER 560 MS-111-PT MAIN STEAM LINE B-1 PRESSURE 18 TRANSMITTER 561 MS-112-PT MAIN STEAM LINE B-2 PRESSURE 18 TRANSMITTER 562 MS-113-PT MAIN STEAM LINE B-1 PRESSURE 18 TRANSMITTER 563 MU-002-PT PRESSURE TO RCP SEALS 18 564 MU-004-DPT LET-DOWN FLOW TRANSMITTER 18 565 MU-004-DPT1 LET-DOWN FLOW TRANSMITTER 18 566 MU-007-DPT1 RCP 3A1 SEAL INJECTION FLOW 18 TRANSMITTER 567 MU-007-DPT2 REP 3A2 SEAL INJECTION FLOW 18 TRANSMITTER 568 MU-007-DPT3 RCP 3B1 SEAL INJECTION FLOW 18 TRANSMITTER 569 MU-007-DPT4 RCP 3B2 SEAL INJECTION FLOW 18 TRANSMITTER 570 MU-014-LT1 MAKE-UP TANK LEVEL 18 TRANSMITTER

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 22 of 35 BASE LIST 1 FOR CR3 C,,

BASE un EQUIPMENT ID DESCRIPTION LIST NO 571 MU-014-LT2 MAKE-UP TANK LEVEL 18 TRANSMITTER 572 MU-017-PT MAKE-UP TANK PRESSURE 18 TRANSMITTER 573 MU-018-DPT LET-DOWN FILTER DELTA-P 18 TRANSMITTER 574 MU-023-DPT1 HI PRESS INJECTION FLOW LOOP B-1 18 575 MU-023-DPT2 HI PRESS INJECTION FLOW LOOP A-2 18 576 MU-023-DPT3 HI PRESS INJECTION FLOW LOOP B-2 18 577 MU-023-DPT4 HI PRESS INJECTION FLOW LOOP A-1 18 578 MU-023-DPT5 HI PRESS INJECTION FLOW LOOP B 18 579 MU-023-DPT6 HI PRESS INJECTION FLOW LOOP A 18 580 MU-023-DPT7 HI PRESS INJECTION FLOW LOOP B 18 581 MU-023-DPT8 HI PRESS INJECTION FLOW LOOP A 18 582 MU-024-DPT MAKE-UP FLOW TRANSMITTER 18 583 MU-024-DPT2 MAKE-UP FLOW TRANSMITTER 18 584 MU-027-DPT RCP TOTAL SEAL INJECTION FLOW 18 TRANSMITTER 585 MU-031-FT1 RCP 3A1 SEAL RETURN FLOW 18 TRANSMITTER 586 MU-031-FT2 RCP 2A2 SEAL RETURN FLOW 18 TRANSMITTER 587 MU-031-FT3 RCP 3B1 SEAL RETURN FLOW 18 TRANSMITTER 588 MU-031-FT4 RCP 3B2 SEAL RETURN FLOW 18 TRANSMITTER 589 MU-081-DPT LET-DOWN PRE-FILTER DELTA-P 18 TRANSMITTER 590 MU-102-DPI RCP SEAL INJECTION FILTER DELTA-18 P INDICATOR 591 NGV-299 ADV BACKUP NITROGEN SUPPLY 18 VALVE 592 NGV-308 ADV BACKUP NITROGEN SUPPLY 18 VALVE 593 RC-001-LT1 PRESSURIZER LEVEL TRANSMITTER 18 594 RC-001-LT2 PRESSURIZER LEVEL TRANSMITTER 18 595 RC-001-LT3 PRESSURIZER LEVEL TRANSMITTER 18 596 RC-003A-PT1 REACTOR COOLANT SYSTEM 18 PRESSURE TRANSMITTER

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 23 of 35 BASE LIST I FOR CR3 BSC EQUIPMENT ID DESCRIPTION

<C LIST NO j

597 RC-003A-PT2 REACTOR COOLANT SYSTEM 18 PRESSURE TRANSMITTER 598 RC-003A-PT3 REACTOR COOLANT SYSTEM 18 PRESSURE TRANSMITTER 599 RC-003A-PT4 REACTOR COOLANT SYSTEM 18 PRESSURE TRANSMITTER 600 RC-003B-PT1 REACTOR COOLANT SYSTEM 18 PRESSURE TRANSMITTER 601 RC-003B-PT2 REACTOR COOLANT SYSTEM 18 PRESSURE TRANSMITTER 602 RC-003B-PT3 REACTOR COOLANT SYSTEM 18 PRESSURE TRANSMITTER 603 RC-014A-DPT1 REACTOR COOLANT SYSTEM HOT 18 LEG A FLOW 604 RC-014A-DPT2 REACTOR COOLANT SYSTEM HOT 18 LEG A FLOW 605 RC-014A-DPT3 REACTOR COOLANT SYSTEM HOT 18 LEG A FLOW 606 RC-014A-DPT4 REACTOR COOLANT SYSTEM HOT 18 LEG A FLOW 607 RC-014B-DPT1 REACTOR COOLANT SYSTEM HOT 18 LEG B FLOW 608 RC-014B-DPT2 REACTOR COOLANT SYSTEM HOT 18 LEG B FLOW 609 RC-014B-DPT3 REACTOR COOLANT SYSTEM HOT 18 LEG B FLOW 610 RC-014B-DPT4 REACTOR COOLANT SYSTEM HOT 18 LEG B FLOW 611 RC-131-PT REACTOR COOLANT SYSTEM 18 PRESSURE 612 RC-131-PT1 REACTOR COOLANT SYSTEM 18 PRESSURE 613 RW-08-PT DECAY HEAT SEAWATER PUMP B 18 DISCHARGE PRESSURE 614 RW-09-PT DECAY HEAT SEA WATER PUMP A 18 DISCHARGE PRESSURE 615 RW-23-PT NUCLEAR SERVICE SEAWATER 18 PUMPS DISCHARGE PRESSURE 616 SF-9-FIT SPENT FUEL COOLANT FLOW 18 TRANSMITTER

U. S. Nuclear Regulatory Commission 3F1112-06, Attachment 1 Page 24 of 35 BASE LIST 1 FOR CR3 U,

BASE un EQUIPMENT ID DESCRIPTION C

LIST NO 6SO 617 SP-17-LT STEAM GENERATOR A LEVEL 18 618 SP-18-LT STEAM GENERATOR A LEVEL 18 619 SP-19-LT STEAM GENERATOR A LEVEL 18 620 SP-20-LT STEAM GENERATOR A LEVEL 18 621 SP-21-LT STEAM GENERATOR B LEVEL 18 622 SP-22-LT STEAM GENERATOR B LEVEL 18 623 SP-23-LT STEAM GENERATOR B LEVEL 18 624 SP-24-LT STEAM GENERATOR B LEVEL 18 625 SP-25-LT STEAM GENERATOR A LEVEL 18 626 SP-26-LT STEAM GENERATOR A LEVEL 18 627 SP-27-LT STEAM GENERATOR A LEVEL 18 628 SP-28-LT STEAM GENERATOR A LEVEL 18 629 SP-29-LT STEAM GENERATOR B LEVEL 18 630 SP-30-LT STEAM GENERATOR B LEVEL 18 631 SP-31-LT STEAM GENERATOR B LEVEL 18 632 SP-32-LT STEAM GENERATOR B LEVEL 18 633 SW-002-PT NUCLEAR SERVICE CCC PUMPS 18 DISCHARGE PRESSURE 634 SW-135-PT NUCLEAR SERVICE COOLING 18 WATER SURGE TANK PRESSURE 635 SW-139-LT NUCLEAR SERVICE COOLING 18 WATER SURGE TANK LEVEL 636 CA-10-TE BOTTOM BORIC ACID STORAGE 19 TANK CAT-5A 637 CA-12-TE BOTTOM BORIC ACID STORAGE 19 TANK CAT-5B 638 MU-005-TE LETDOWN LINE TEMPERATURE 19 639 RC-002-TE1 PRESSURIZER TEMPERATURE 19 640 RC-002-TE2 PRESSURIZER TEMPERATURE 19 641 RC-004A-TE1 REACTOR COOLANT SYSTEM HOT 19 LEG A TEMPERATURE 642 RC-004A-TE2 REACTOR COOLANT SYSTEM HOT 19 LEG A TEMPERATURE 643 RC-004A-TE3 REACTOR COOLANT SYSTEM HOT 19 LEG A TEMPERATURE 644 RC-004A-TE4 REACTOR COOLANT SYSTEM HOT 19 LEG A TEMPERATURE 645 RC-004B-TE1 REACTOR COOLANT SYSTEM HOT 19 LEG B TEMPERATURE

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 25 of 35 BASE LIST I FOR CR3 BASE u,

EQUIPMENT ID DESCRIPTION

-J LIST NO.-

646 RC-004B-TE2 REACTOR COOLANT SYSTEM HOT 19 LEG B TEMPERATURE 647 RC-004B-TE3 REACTOR COOLANT SYSTEM HOT 19 LEG B TEMPERATURE 648 RC-004B-TE4 REACTOR COOLANT SYSTEM HOT 19 LEG B TEMPERATURE 649 RC-005A-TE1 REACTOR COOLANT SYSTEM COLD 19 LEG A TEMPERATURE 650 RC-005A-TE2 REACTOR COOLANT SYSTEM COLD 19 LEG B TEMPERATURE 651 RC-005A-TE3 REACTOR COOLANT SYSTEM COLD 19 LEG B TEMPERATURE 652 RC-005A-TE4 REACTOR COOLANT SYSTEM COLD 19 LEG A TEMPERATURE 653 RC-005B-TE1 REACTOR COOLANT SYSTEM COLD 19 LEG B TEMPERATURE 654 RC-005B-TE2 REACTOR COOLANT SYSTEM COLD 19 LEG B TEMPERATURE 655 RC-005B-TE3 REACTOR COOLANT SYSTEM COLD 19 LEG B TEMPERATURE 656 RC-005B-TE4 REACTOR COOLANT COLD LEG B 19 TEMPERATURE 657 RW-12-TE DH SERVICE SEA WATER PUMP RWP 19 3B DISCHARGE TEMPERATURE 658 RW-13-TE DECAY HEAT SEA WATER PUMP RWP 19 3A DISCHARGE TEMPERATURE 659 RW-19-TE NUCLEAR SERVICE SEA WATER 19 PUMP RWP-2B DISCHARGE TEMP 660 RW-32-TE DECAY HEAT COOLER DCHE-1B SEA 19 WATER OUTLET 661 RW-33-TE DECAY HEAT COOLER DCHE-1A SEA 19 WATER OUTLET 662 RW-43-TE SEA WATER OUTLET NUCLEAR 19 SERVICE COOLERS 663 AHCP-4 SCR CABINET FOR AHHE-4A AND 20 AHHE-4B 664 AHPL-1 PANEL 20 665 ATCP-1 ANTICIPATED TRANSIENT WITHOUT 20 SCRAM LOGIC CABINET

U. S. Nuclear Regulatory Commission 3F17112-06, Attachment 1 Page 26 of 35 BASE LIST I FOR CR3 U)

BASE uf EQUIPMENT ID DESCRIPTION LIST NO 666 CCBT CENTRAL CONTROL BOARD 20 TERMINATION CABINET 667 DPTP-5A ENGINEERED SAFEGUARDS DC 20 TEST PANEL 668 DPTP-5B ENGINEERED SAFEGUARDS DC 20 TEST PANEL 669 DPTP-6A ENGINEERED SAFEGUARDS DIESEL 20 GENERATOR DC TEST PANEL 670 DPTP-6B ENGINEERED SAFEGUARDS DIESEL 20 GENERATOR DC TEST PANEL 671 DPTP-8A ENGINEERED SAFEGUARDS DC 20 TEST PANEL 672 DPTP-8B ENGINEERED SAFEGUARDS DC 20 TEST PANEL 673 DRRD-02 CRD DC BREAKER CABINET 20 674 DRRD-10 CRD AC BREAKER CABINET A - UNIT 10 20 675 DRRD-1 1 CRD AC BREAKER CABINET B - UNIT 11 20 676 EFIC-A EMERGENCY FEEDWATER INITIATION 20 AND CONTROL CABINET A 677 EFIC-B EMERGENCY FEEDWATER 20 INITIATION AND CONTROL CABINET B 678 EFIC-C EMERGENCY FEEDWATER 20 INITIATION AND CONTROL CABINET C 679 EFIC-D EMERGENCY FEEDWATER 20 INITIATION AND CONTROL CABINET D 680 EFIC-EC-A EMERGENCY FEEDWATER AUXILIARY 20 EQUIPMENT CABINET A 681 EFIC-EC-B EMERGENCY FEEDWATER AUXILIARY 20 EQUIPMENT CABINET B 682 EFIC-RC-1C EMERGENCY FEEDWATER 20 AUXILIARY RELAY CABINET 1C 683 EFIC-RC-1 D EMERGENCY FEEDWATER 20 AUXILIARY RELAY CABINET 1D 684 EGCP-1A EMERGENCY DIESEL GENERATOR A 20 CONTROL PANEL 685 EGCP-11B EMERGENCY DIESEL GENERATOR B 20 CONTROL PANEL 686 EGCP-2A EMERGENCY DIESEL GEN A 20 ELECTRICAL EQUIPMENT CABINET

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 27 of 35 BASE LIST 1 FOR CR3 U,

BASE u,

EQUIPMENT ID DESCRIPTION LIST NO 687 EGCP-2B EMERGENCY DIESEL GEN B 20 ELECTRICAL EQUIPMENT CABINET 688 EGCP-3A DIESEL GENERATOR A CONTROL 20 POWER RELAY CABINET 689 EGCP-3B DIESEL GENERATOR B CONTROL 20 POWER RELAY CABINET 690 EGCP-4A DIESEL GENERATOR A POWER 20 CIRCUIT BREAKER PANEL 691 EGCP-4B DIESEL GENERATOR B POWER 20 CIRCUIT BREAKER PANEL 692 ESCC-1 ES SYSTEM CHANNEL TEST 20 CABINET 1 693 ESCC-1A ES SYSTEM CHANNEL CABINET 1A 20 694 ESCC-1B ES SYSTEM CHANNEL CABINET 1B 20 695 ESCC-2 ES SYSTEM CHANNEL TEST 20 CABINET 2 696 ESCC-2A ES SYSTEM CHANNEL CABINET 2A 20 697 ESCC-2B ES SYSTEM CHANNEL CABINET 2B 20 698 ESCC-3 ES SYSTEM CHANNEL TEST 20 CABINET 3 699 ESCC-3A ES SYSTEM CHANNEL CABINET 3A 20 700 ESCC-3B ES SYSTEM CHANNEL CABINET 3B 20 701 ESCP-4A ENGINEERED SAFEGUARDS 20 ACTUATION RELAY CABINET 4A 702 ESCP-4B ENGINEERED SAFEGUARDS 20 ACTUATION RELAY CABINET 4B 703 ESCP-4C ENGINEERED SAFEGUARDS 20 ACTUATION RELAY CABINET 4C 704 ESCP-4D ENGINEERED SAFEGUARDS 20 ACTUATION RELAY CABINET 4D 705 ESCP-5A ENGINEERED SAFEGUARDS 20 ACTUATION RELAY CABINET 5A 706 ESCP-5B ENGINEERED SAFEGUARDS 20 ACTUATION RELAY CABINET 5B 707 ESCP-5C ENGINEERED SAFEGUARDS 20 ACTUATION RELAY CABINET 5C 708 ESCP-5D ENGINEERED SAFEGUARDS 20 ACTUATION RELAY CABINET 5D 709 ESPSC-3A1 ENGINEERED SAFEGUARDS 20 PRESSURE SWITCH CABINET 3A1

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 28 of 35 BASE LIST 1 FOR CR3 U,

BASE co EQUIPMENT ID DESCRIPTION LIST NO 710 ESPSC-3A2 ENGINEERED SAFEGUARDS 20 PRESSURE SWITCH CABINET 3A2 711 ESPSC-3A3 ENGINEERED SAFEGUARDS 20 PRESSURE SWITCH CABINET 3A3 712 ESPSC-3A4 ENGINEERED SAFEGUARDS 20 PRESSURE SWITCH CABINET 3A4 713 ESPSC-3B1 ENGINEERED SAFEGUARDS 20 PRESSURE SWITCH CABINET 3B1 714 ESPSC-3B2 ENGINEERED SAFEGUARDS 20 PRESSURE SWITCH CABINET 3B2 715 ESPSC-3B3 ENGINEERED SAFEGUARDS 20 PRESSURE SWITCH CABINET 3B3 716 FSCP-1 CONTROL PANEL WITH BATTERY 20 RACK AND MODULES 717 FSCP-2 LEASED LINE ANNUNCIATOR 20 718 FSCP-3 LEASED LINE ANNUNCIATOR 20 719 HTCP-2 EDG-B EMERGENCY LOAD 20 SHEDDING - HEAT TRACING 720 HTCP-3 EDG-B EMERGENCY LOAD 20 SHEDDING - HEAT TRACING 721 HTCP-4 EDG-A EMERGENCY LOAD 20 SHEDDING - HEAT TRACING 722 HTCP-5 EDG-A EMERGENCY LOAD 20 SHEDDING - HEAT TRACING 723 HVAC-16A HVAC CONTROL CABINET 16A 20 724 HVAC-16B HVAC CONTROL CABINET 16B 20 725 HVAC-17A HVAC CONTROL CABINET 17A 20 726 HVAC-17B HVAC CONTROL CABINET 17B 20 727 HVAC-19 HVAC CONTROL CABINET 19 20 728 HVAC-20 HVAC CONTROL CABINET 20 20 729 ICS MAIN CONTROL BOARD 20 730 ICS-1 INTEGRATED CONTROL SYSTEM 20 CABINET 1 731 ICS-2 INTEGRATED CONTROL SYSTEM 20 CABINET 2 732 ICS-3 INTEGRATED CONTROL SYSTEM 20 CABINET 3 733 ICS-4 INTEGRATED CONTROL SYSTEM 20 CABINET 4

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 29 of 35 BASE LIST 1 FOR CR3 BASE co EQUIPMENT ID DESCRIPTION LIST NO 734 ICS-5 INTEGRATED CONTROL SYSTEM 20 CABINET 5 735 MTCP-1A UNDERVOLTAGE TEST CABINET A 20 736 MTCP-1 B UNDERVOLTAGE TEST CABINET B 20 737 NI&P-A1 NI&P SYSTEM SUBASSEMBLY A 20 CABINET 1 738 NI&P-A2 NI&P SYSTEM SUBASSEMBLY A 20 CABINET 2 739 NI&P-B1 NI&P SYSTEM SUBASSEMBLY B 20 CABINET 1 740 NI&P-B2 NI&P SYSTEM SUBASSEMBLY B 20 CABINET 2 741 NI&P-C1 NI&P SYSTEM SUBASSEMBLY C 20 CABINET 1 742 NI&P-C2 NI&P SYSTEM SUBASSEMBLY C 20 CABINET 2 743 NI&P-D1 NI&P SYSTEM SUBASSEMBLY D 20 CABINET 1 744 NI&P-D2 NI&P SYSTEM SUBASSEMBLY D 20 CABINET 2 745 NNI-1 AUXILIARY CONTROL SYSTEM 20 CABINET 1 746 NNI-2 AUXILIARY CONTROL SYSTEM 20 CABINET 2 747 NNI-3 AUXILIARY CONTROL SYSTEM 20 CABINET 3 748 NNI-4 AUXILIARY CONTROL SYSTEM 20 CABINET 4 749 NNI-5 AUXILIARY CONTROL SYSTEM 20 CABINET 5 750 NNI-6 AUXILIARY CONTROL SYSTEM 20 CABINET 6 751 NNI-7 AUXILIARY CONTROL SYSTEM 20 CABINET 7 752 NNI-8 AUXILIARY CONTROL SYSTEM 20 CABINET 8 753 NSP NUCLEAR SAMPLE PANEL 20 754 PORV/TEMP PORV & TEMPERATURE 20 SATURATION CABINET

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 30 of 35 BASE LIST 1 FOR CR3 BASE n

EQUIPMENT ID DESCRIPTION

<C LIST NO 755 RCCP-1 PRESSURIZER HEATER SCR 20 CONTROL PANEL 756 RCITS-A REACTOR COOLANT INVENTORY 20 TRACKING SYSTEM CABINET A 757 RCITS-B REACTOR COOLANT INVENTORY 20 TRACKING SYSTEM CABINET B 758 RCITS-C REACTOR COOLANT INVENTORY 20 TRACKING SYSTEM CABINET C 759 RCPM-3A REACTOR COOLANT PUMP POWER 20 MONITORING CABINET A 760 RCPM-3B REACTOR COOLANT PUMP POWER 20 MONITORING CABINET B 761 RCTR-1 RCP-1C TRANSFORMER CABINET 20 762 RCTR-2 RCP-1 D TRANSFORMER CABINET 20 763 RCTR-3 RCP-1A TRANSFORMER CABINET 20 764 RCTR-4 RCP-1B TRANSFORMER CABINET 20 765 RNR RECORDER NEST RACK 20 766 RR1 AUXILIARY RELAY RACK 20 767 RR1A ENGINEERED SAFEGUARD 20 AUXILIARY RELAY RACK RR1A 768 RR1AB ENGINEERED SAFEGUARD 20 AUXILIARY RELAY RACK RR1AB 769 RR1B ENGINEERED SAFEGUARD 20 AUXILIARY RELAY RACK RR1B 770 RR2 AUXILIARY RELAY RACK 20 771 RR2A ENGINEERED SAFEGUARD 20 AUXILIARY RELAY RACK RR2A 772 RR2AB ENGINEERED SAFEGUARD 20 AUXILIARY RELAY RACK RR2AB 773 RR2B ENGINEERED SAFEGUARD 20 AUXILIARY RELAY RACK RR2B 774 RR3 AUXILIARY RELAY RACK 20 775 RR3A ENGINEERED SAFEGUARD 20 AUXILIARY RELAY RACK RR3A 776 RR3B ENGINEERED SAFEGUARD 20 AUXILIARY RELAY RACK RR3B 777 RR4A POST-ACCIDENT MONITORING 20 PANEL 4A 778 RR4B POST-ACCIDENT MONITORING 20 I

_PANEL 4B N

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 31 of 35 BASE LIST I FOR CR3 BASE CA EQUIPMENT ID DESCRIPTION LIST NO 779 RR5B1 ENGINEERED SAFEGUARD 20 AUXILIARY RELAY RACK RR5B1 780 RR5B2 ENGINEERED SAFEGUARD 20 AUXILIARY RELAY RACK RR5B2 781 RR-HV AUXILIARY RELAY RACK HEATING 20 AND VENTILATION 782 RR-PSA AUXILIARY RELAY RACK 20 783 RSA REMOTE SHUTDOWN RELAY 20 CABINET A 784 RSA-1 REMOTE SHUTDOWN RELAY 20 CABINET A-1 785 RSACA REMOTE SHUTDOWN AUXILIARY 20 CABINET A 786 RSACB REMOTE SHUTDOWN AUXILIARY 20 CABINET B 787 RSB REMOTE SHUTDOWN RELAY 20 CABINET B 788 RSB-1 REMOTE SHUTDOWN RELAY 20 CABINET B-1 789 RSPA REMOTE SHUTDOWN PANEL -

20 SECTION A 790 RSPAB REMOTE SHUTDOWN PANEL -

20 SECTION AB 791 RSPB REMOTE SHUTDOWN PANEL -

20 SECTION B 792 TPC TRANSMITTER POWER SUPPLY 20 CABINETS A & B 793 VBDP-05-SIP 120 VOLT VITAL BUS 3C STATUS 20 INDICATION PANEL 794 VBDP-06-SIP 120 VOLT VITAL BUS 3D STATUS 20 INDICATION PANEL 795 VBDP-08/12/13-SIP VITAL BUS STATUS INDICATION 20 PANEL A FOR VBDP-8/12/13 796 VBDP-09-SIP VITAL BUS STATUS INDICATION 20 PANEL C FOR VBDP-9 797 VBDP-10/14/15-SIP VITAL BUS STATUS INDICATION 20 PANEL B FOR VBDP-10/14/15 798 VBDP-11-SIP VITAL BUS STATUS INDICATION 20 PANEL D FOR VBDP-1 1 799 VBXS-2A AUTO TRANSFER SWITCH FOR NNI 20

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 32 of 35 BASE LIST I FOR CR3 u,

BASE

'f EQUIPMENT ID DESCRIPTION

<J LIST NO 800 VBXS-2B AUTO TRANSFER SWITCH FOR ICS 20 801 AHDR-1A CONTROL COMPLEX HVAC 21 COMPRESSED AIR DRYER A 802 AHDR-1B CONTROL COMPLEX HVAC 21 COMPRESSED AIR DRYER B 03 CAHE-1 PRESSURIZER SAMPLE COOLER 21 804 CAHE-2A STEAM GENERATOR A SAMPLE 21 COOLER 805 CAHE-2B STEAM GENERATOR B SAMPLE 21 COOLER 806 CAHE-5 PASS RC SAMPLE COOLER 21 807 CAHE-6 PASS DECAY HEAT SAMPLE 21 COOLER 808 CAHE-8 PASS RC SAMPLE PRE-COOLER 21 809 CAT-5A BORIC ACID STORAGE A 21 810 CAT-5B BORIC ACID STORAGE TANK B 21 811 CDHE-4A MAIN CONDENSER A 21 812 CDHE-4B MAIN CONDENSER B 21 813 CDT-1 CONDENSATE STORAGE TANK 21 814 CHT-1 CHILLED WATER EXPANSION TANK 21 815 DCHE-1A DECAY HEAT CLOSED CYCLE 21 COOLING A 816 DCHE-1B DECAY HEAT CLOSED CYCLE 21 COOLING B 817 DCT-1A DECAY HEAT CCC SURGE TANK A 21 818 DCT-1B DECAY HEAT CCC SURGE TANK B 21 819 DFT-1A DIESEL GENERATOR FUEL OIL 21 STORAGE TANK A 820 DFT-1B DIESEL GENERATOR FUEL OIL 21 STORAGE TANK B 821 DFT-3A DIESEL GENERATOR FUEL OIL DAY 21 TANK A 822 DFT-3B DIESEL GENERATOR FUEL OIL DAY 21 TANK B 823 DHHE-1A DECAY HEAT REMOVAL HEAT 21 EXCHANGER A 824 DHHE-1 B DECAY HEAT REMOVAL HEAT 21 EXCHANGER B 825 DHT-1 BORATED WATER STORAGE TANK 21

U. S. Nuclear Regulatory Commission 3F17112-06, Attachment 1 Page 33 of 35 BASE LIST 1 FOR CR3 BS EQUIPMENT ID DESCRIPTION LIST NO 826 DLHE-1A DIESEL GENERATOR LUBE OIL 21 COOLER 1A 827 DLHE-1B DIESEL GENERATOR LUBE OIL 21 COOLER 1B 828 DLHE-2A DIESEL GENERATOR LUBE OIL 21 COOLER 2A 829 DLHE-2B DIESEL GENERATOR LUBE OIL 21 COOLER 2B 830 EFT-2 EMERGENCY FEEDWATER TANK 21 831 EGT-1A EDG A AIR RECEIVER 1A 21 832 EGT-1 B EDG A AIR RECEIVER 1 B 21 833 EGT-2A EDG B AIR RECEIVER 2A 21 834 EGT-2B EDG B AIR RECEIVER 2B 21 835 IADR-1 INSTRUMENT AIR DRYER 1 21 836 IAHE-1A INSTRUMENT AIR AFTER-COOLER A 21 837 IAHE-1B INSTRUMENT AIR AFTER-COOLER B 21 838 IAT-1A INSTRUMENT AIR RECEIVER A 21 839 IAT-1B INSTRUMENT AIR RECEIVER B 21 840 MSV-411-AR1 MSV-411 AIR RESERVOIR 1 21 841 MSV-411-AR2 MSV-411 AIR RESERVOIR 2 21 842 MSV-411-AR3 MSV-411 AIR RESERVOIR 3 21 843 MSV-412-AR1 MSV-412 AIR RESERVOIR 1 21 844 MSV-412-AR2 MSV-412 AIR RESERVOIR 2 21 845 MSV-412-AR3 MSV-412 AIR RESERVOIR 3 21 846 MSV-413-AR1 MSV-413 AIR RESERVOIR 1 21 847 MSV-413-AR2 MSV-413 AIR RESERVOIR 2 21 848 MSV-413-AR3 MSV-413 AIR RESERVOIR 3 21 849 MSV-414-AR1 MSV-414 AIR RESERVOIR 1 21 850 MSV-414-AR2 MSV-414 AIR RESERVOIR 2 21 851 MSV-414-AR3 MSV-414 AIR RESERVOIR 3 21 852 MUDM-1A MAKE-UP AND PURIFICATION 21 DEMINERALIZER 1A 853 MUDM-1B MAKE-UP AND PURIFICATION 21 DEMINERALIZER 1B 854 MUHE-1A LET-DOWN COOLER 3A 21 855 MUHE-1 B LET-DOWN COOLER 3B 21 856 MUHE-1C LET-DOWN COOLER 3C 21 857 MUHE-2A RCP SEAL RETURN COOLER 3A 21 858 MUHE-2B RCP SEAL RETURN COOLER 3B 21 859 MUT-1 MAKE-UP TANK 21

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 1 Page 34 of 35 BASE LIST 1 FOR CR3 co BASE co EQUIPMENT ID DESCRIPTION LIST NO 860 NGT-XX ADV BACKUP NITROGEN SUPPLY 21 TANKS (10) 861 RWSP-1A CYCLONE SEPARATOR A 21 862 RWSP-1B CYCLONE SEPARATOR B 21 863 SAHE-1A STATION AIR COMPRESSOR AFTER-21 COOLER A 864 SAHE-1 B STATION AIRE COMPRESSOR 21 AFTER-COOLER B 865 SFDM-1 SPENT FUEL COOLANT 21 DEMINERALIZER 866 SFHE-1A SPENT FUEL COOLER A 21 867 SFHE-1B SPENT FUEL COOLER B 21 868 SWHE-1A NUCLEAR SERVICE CCC HEAT 21 EXCHANGER 3A 869 SWHE-1B NUCLEAR SERVICE CCC HEAT 21 EXCHANGER 3B 870 SWHE-1C NUCLEAR SERVICE CCC HEAT 21 EXCHANGER 3C 871 SWHE-1D NUCLEAR SERVICE CCC HEAT 21 EXCHANGER 3D 872 SWHE-2 SW SUPPLY TO PASS SAMPLE 21 COOLERS 873 SWT-1 NUCLEAR SERVICE CLOSED CYCLE 21 SURGE TANK 874 SWV-012-AR SWV-12 CONTROL 21 875 SWV-151-AR AIR RESERVOIR FOR SWV-151 21 876 SWV-152-AR AIR RESERVOIR FOR SWV-152 21 877 SWV-353-AR AIR RESERVOIR FOR SWV-353 21 878 SWV-354-AR AIR RESERVOIR FOR SWV-354 21 879 SWV-355-AR AIR RESERVOIR FOR SWV-355 21 880 WDHE-1 REACTOR COOLANT DRAIN TANK 21 COOLER 881 WDT-1A WASTE GAS DECAY TANK 1A 21 882 WDT-1B WASTE GAS DECAY TANK 1B 21 883 WDT-1C WASTE GAS DECAY TANK 1C 21 884 WDT-3A RC BLEED TANK 3A 21 885 WDT-3B RC BLEED TANK 3B 21 886 WDT-3C RC BLEED TANK 3C 21 887 WDT-5 REACTOR COOLANT DRAIN TANK 21 888 DPDP-1D 125VDC DISTRIBUTION PANEL 14

U. S. Nuclear Regulatory Commission 3F17112-06, Attachment 1 Page 35 of 35 BASE LIST I FOR CR3 BASE un EQUIPMENT ID DESCRIPTION LIST NO,-

889 DAP-1 AIR COMPRESSOR 12 890 AH-1047-TS EFPB DIESEL ROOM TEMPERATURE 19 SWITCH 891 DFP-4 DIESEL FUEL FILTRATION PUMP 5

892 DEFL-4 RECIRCULATION LINE FILTER 0

SEPERATOR 893 EF-65-FI EFP-3 FULL FLOW RECIRC FLOW 18 INDICATOR 894 SAP-5 AIR COMPRESSOR 12 895 DPBC-1J BATTERY CHARGER 16 896 EFV-146 EFP-3 DISCHARGE STOP CHECK 8

897 EFV-154 EFP-3 RECIRC ISOLATION 8

898 EFV-144 CDT-1 TO EFP-3 ISOLATION 8

899 DPBA-1 D BATTERY BANK 15 950 AHFL-22A EGDG-1A COMBUSTION AIR FILTER 0

951 AHFL-5A AHF-22A & AHF-22B OUTSIDE AIR 0

PREFILTER 952 MTSW-3F 480V ES BUS 3A 2

953 MTSW-3F-T 480V UNIT SUBSTATION 4

ENGINEERED SAFEGUARDS BUS 3A 954 SWV-355-AR1 SWV-355 CLOSURE ASSIST AIR 21 RESERVOIR MOD 1 IADR-2 INSTRUMENT AIR DRYER 2 21 MOD 2 IAV-676 ADV BACKUP AIR SUPPLY 18 VALVE MOD 3 IAP-3B INSTRUMENT AIR COMPRESSOR 12

RESPONSE TO RECOMMENDATION 2.3 SEISMIC WALKDOWN OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM THE FUKUSHIMA DAI-ICHI ACCIDENT : SWEL 1

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 2 Page 1 of 13 Feature N=New R=Modification or Replacement A= USI A-46 Outlier 0

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x EGDG-1A INTER-ROOM EFIC ROOMS COOLING A EGDG-1A COMBUSTION AIR FILTER x

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x x

ASDT-14 ISOLATION

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 2 Page 2 of 13 Feature N=New R=Modification or Replacement A= USI A-46 Outlier PASS RC SAMPLE PRE-COOLER 0

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CH CH CH DF DIESEL GENERATOR FUEL OIL DAY TANK A LEVEL SWITCH X

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U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 2 Page 3 of 13 Feature N=New R=Modification or Replacement A= USI A-46 Outlier x

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x BATTERY CHARGER B BATTERY CHARGER D x

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x DJ DL DP DP DP DP DP DP DP DP BATTERY CHARGER 250/125V DC MAIN PANEL 3A 125VDC DISTRIBUTION PANEL 250/125V-DC EFIC PNL DPDP-8C

U. S. Nuclear Regulatory Commission 3F1112-06, Attachment 2 Page 4 of 13 Feature N=New R=Modification or Replacement A= USI A-46 Outlier

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x EF EF EFP-3 FULL FLOW RECIRC FLOW INDICATOR EMERGENCY FEEDWATER TANK LEVEL X

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EF EMERGENCY FEEDWATER INITIATION AND CONTROL CABINET D EMERGENCY FEEDWATER AUXILIARY EQUIPMENT CABINET A X

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X I IEF MOTOR DRIVEN EMERGENCY FEEDWATER PUMP X

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U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 2 Page 5 of 13 Feature N=New R=Modification or Replacement A= USI A-46 Outlier 7-0 0.

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0 S

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x EF EF EFP-3 DISCHARGE STOP CHECK EFP-3 RECIRC ISOLATION EGDG-1 B AIR START VALVE x

x x

DIESEL GENERATOR B CONTROL POWER RELAY CABINET DIESEL GENERATOR A EMERG DIESEL GEN AIR RECEIVER x

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U. S. Nuclear Regulatory Commission, Attachment 2 3F1112-06 Page 10 of 13 Feature N=New R=Modification or Replacement A= USI A-46 Outlier%

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U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 2 Page 11 of 13 Feature N=New R=Modification or Replacement A= USI A-46 Outlier 0

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U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 2 Page 12 of 13 Feature N=New R=Modification or Replacement A= USI A-46 Outlier I-0 0.

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RESPONSE TO RECOMMENDATION 2.3 SEISMIC WALKDOWN OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM THE FUKUSHIMA DAI-ICHI ACCIDENT : Base List 2

U. S. Nuclear Regulatory Commission 3F1112-06, Attachment 3 Page 1 of 1 Feature Building SWITCH, CONTROL BORATED WATER AB RECIRCULATION PP RM-L4 INLET SAMPLE FLOW AB SFPS DISCHARGE FLOW TO SFFL-1A/1B AND SFDM-1 AB SFFL-1A DIFFERENTIAL PRESSURE AB SFFL-1B DIFFERENTIAL PRESSURE AB SPENT FUEL COOLANT PUMP 1B AB SPENT FUEL COOLANT PP MOTOR 1A AB SPENT FUEL COOLANT PP MOTOR 1B AB BORATED WATER RECIRCULATION PUMP MOTOR AB SFFL-1A OULET TO HEADER ISOLATION AB SPENT FUEL CASK AREA ISOLATION AB

.SPENT FUEL CASK AREA LOOP SEAL VENT AB SPENT FUEL COOLER A AB SPENT FUEL COOLER B AB

RESPONSE TO RECOMMENDATION 2.3 SEISMIC WALKDOWN OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM THE FUKUSHIMA DAI-ICHI ACCIDENT : Rapid Drain-Down List

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 4 Page 1 of 1 There were no items on the Rapid Drain-Down List for CR-3.

RESPONSE TO RECOMMENDATION 2.3 SEISMIC WALKDOWN OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM THE FUKUSHIMA DAI-ICHI ACCIDENT : SWEL 2

U. S. Nuclear Regulatory Commission 3F1 112-06, Attachment 5 Page 1 of 1 Feature Building SWITCH, CONTROL BORATED WATER AB RECIRCULATION PP SPENT FUEL COOLANT PP AB BORATED WATER RECIRCULATION PUMP AB SPENT FUEL CASK AREA LOOP SEAL VENT AB SPENT FUEL COOLER B AB SPENT FUEL COOLER A AB

RESPONSE TO RECOMMENDATION 2.3 SEISMIC WALKDOWN OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM THE FUKUSHIMA DAI-ICHI ACCIDENT SECURITY-RELATED INFORMATION -WITHHOLD UNDER 10CFR2.390 : Seismic Walkdown Checklists SECURITY-RELATED INFORMATION -WITHHOLD UNDER 10CFR2.390

RESPONSE TO RECOMMENDATION 2.3 SEISMIC WALKDOWN OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM THE FUKUSHIMA DAI-ICHI ACCIDENT SECURITY-RELATED INFORMATION -WITHHOLD UNDER 10CFR2.390 : Area Walk-by Checklists SECURITY-RELATED INFORMATION -WITHHOLD UNDER 10CFR2.390

RESPONSE TO RECOMMENDATION 2.3 SEISMIC WALKDOWN OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM THE FUKUSHIMA DAI-ICHI ACCIDENT : Peer Review Report

U. S. Nuclear Regulatory Commission, Attachment 8 3F1 112-06 Page 1 of 4 CRYSTAL RIVER UNIT 3 SEISMIC WALKDOWN PEER REVIEW REPORT Peer Review activities were performed on the Seismic Walkdown Program in addition to the Programmatic Controls / Oversight that were established for the project. A brief description of the Programmatic Controls / Oversight and Peer Review findings is provided below:

Programmatic Controls / Oversight Programmatic Controls / Oversight were developed for the 2.3 Seismic Walkdowns and implemented at Crystal River Unit 3 (CR-3). A specification based on the EPRI guidance was established to control SWEL development and walkdown requirements. A specification was developed since EPRI 1025286 was written as guidance, whereas, the specification provided definitive criteria and control to avoid interpretation and promote consistency. The specification was inclusive of the EPRI guidance. A Quality Assurance (QA) person was present at the site during the inspection to assure form and specification compliance. Technical oversight was performed by the Project Manager (PM). The PM was onsite during the SWEL development, and intermittently during the walkdowns and report generation phases of the project. An in-process review of work was performed during those intervals. Inspections at the four Progress Energy sites were being performed concurrently and lessons learned were relayed to the inspection teams at the other sites to determine if commonality was present within the fleet.

These in-process reviews were performed through all phases of the project with the intent of meeting the intent of the EPRI guidance.

Peer Review Separate from the programmatic controls implemented at the sites, Peer Review activities were performed on the seismic walkdown program that spanned from the development of the specification and SWEL through the physical walkdowns and ultimately to the report preparation and review. The Peer Review team concluded that the inspection program was performed in accordance with the guidance provided in EPRI 1025286. The Peer Review found the effort at CR-3 was performed in a competent manner and a very broad spectrum of components located throughout the power block was included in the program. The results were documented in a Duke (legacy Progress Energy) engineering change package. The aspects of the program that were reviewed by the Peer Review justifying this statement are provided as follows:

Inspection Team The Peer Review found Seismic Walkdown Engineers (SWE) performing the inspections were very experienced with a background in design engineering including seismic design at nuclear facilities dating back to the design of the first generation nuclear power plants. SWEs had prior seismic walkdown experience at operating nuclear power plants, Department of Energy facilities, and other pertinent applications.

Training consistent with the EPRI training was provided to all SWEs before any inspections were performed. The resumes of the SWEs were reviewed and it was determined that the SWEs were found to have qualifications that were consistent with the requirements of the regulatory guidance.

U. S. Nuclear Regulatory Commission, Attachment 8 3F1 112-06 Page 2 of 4 Selection of SWEL Items The Peer Review concluded the process used to select SWEL items included both selected and diverse aspects.

The list of equipment was obtained from the USI A-46 Safe Shutdown Equipment List (SSEL) and the appropriate screening filters identified in the EPRI guidance were applied. The number of items included in the SWEL represented an appropriate number of items in each equipment class when compared to the total number of items on the SSEL.

The items that were individually selected typically were items that would have the most severe consequence in the event that the target item were to fail during the seismic event and resulted in components associated with the Emergency Diesel Generators, vital power, and heat removal systems being well represented.

Other conditions given additional consideration included environmental and distribution into diverse structures, while items that are included in other programmatic inspections, (e.g., AOV, MOV, Appendix R, ASME Section Xl Subsection IWE/IWL), were minimized. The process used to determine the SWEL items was in accordance with the EPRI guidance and adequately represents a diverse sample of the equipment required to perform the five safety functions.

The Peer Review confirmed site Operations experience was included in the review of the components to assure a representative distribution of equipment was included in the SWEL.

Operations also performed preliminary walkdowns to determine if the components could be safely accessed.

A selection/substitution criterion was established before the items were assessed and if items were judged inaccessible, then the substitution criteria were used. The Peer Review interviewed the personnel making the equipment selections and Operations personnel to confirm an acceptable approach was used in selecting the equipment for sampling.

A sample of modifications performed at the site since the last IPEEE/USI A-46 inspection, previous USI A-46 outliers, and upgrades were reflected in the SWEL.

The SWEL contained 134 components in SWEL-1 and an additional 6 items in SWEL-2 totaling 140 selected items for the combined SWEL. The number of items on SWEL-1 exceeds the recommended range of 90-120 items. The SWEL was taken from the USI A-46 SSEL which contained both safety-related and non-seismic components.

The SWEL was expanded to assure that a minimum of 90 seismic components were inspected without the inclusion of representative non-seismic items from the USI A-46 SSEL. This decision was reviewed as being a conservative interpretation of the EPRI guidance.

The process used to select the SWEL items, inclusion of Operations personnel into the selection of the items, USI A-46 outliers and modifications were represented and the number and distribution of items was in accordance with the EPRI guidance and confirmed by the Peer Review utilizing the Peer Review Checklist for the SWEL.

Pre-Inspection Preparation Peer Review was performed on the pre-inspection prepared walkdown packages which consisted of general configuration and structural drawings, anchorage detailing, and seismic demand on the anchorage and it was confirmed that these packages were available in the field during the inspection. The inspection packages were reviewed for thoroughness to the criteria and samples were selected to determine appropriateness of the information.

At random intervals during the walkdown phase of the project, the SWEs were questioned to determine if they had been adequately prepared and specifically, they were questioned to determine if they

U. S. Nuclear Regulatory Commission, Attachment 8 3F1 112-06 Page 3 of 4 knew the vertical and horizontal strong motion demand in the areas that they would be working.

The SWEs demonstrated that they had adequately prepared for the inspections prior to entering the field.

Conduct of Inspections The Peer Review concluded the SWEs conducted field inspections with the walkdown packages "in-hand." The Seismic Walkdown Checklist (SWC) and Area Walk-By Checklist (AWC) were physically used in the field and place keeping practices were employed. The SWEL items were inspected; the forms were filled out in the field, and were reviewed by the SWEs before they left the area. As a result of conversations with the SWEs and Peer Review observations during the inspections, it was concluded that pertinent and thorough conversations occurred between the SWEs in the field to generally reach a consensus on a real time basis in the field.

The inspections were performed in accordance with the EPRI guidance.

Review of Walkdown and Area Walk-By Checklist The peer reviewers discussed the inspections with the SWEs prior to field implementation and sampled field reports during the inspections to determine adequacy of the inspection.

The SWEs were asked to describe the encountered field conditions and the forms were reviewed to determine if the information was representative. The checklist was used predominately with hand written notes being used judiciously to reflect conditions. Good practices were noted on the forms, (e.g., calculation and drawing references, condition identification, and area housekeeping conditions).

The final documents (i.e., package including checklist, photographs, drawings, notes) were compared to the field notes with the QA representative reviewing approximately 100% of the forms and the Peer Review reviewing over 50% of the forms. As a result of the Peer Review, there were some instances that required the SWE to obtain and/or delineate additional information in the walkdown packages. Once incorporated, the information presented on the forms was consistent with expectations and are judged representative of the field conditions.

Decisions for Entering Potential Adverse Seismic Conditions (PASCs) into CAP Process The Peer Review concluded that the identification of potential SSCs was placed into the CR-3 CAP process before a full assessment was performed to determine if the condition identified was a PASC. The CR-3 in-process review board/engineering reviewed the CAP items. Some items were classified as simple work requests and some identified conditions were justified by locating specific documentation demonstrating that the component was qualified as-is requiring no field actions to be performed. The original USI A-46/IPEEE inspection results were reviewed to ascertain the status of the equipment at the time of the previous inspections. The process used by the site personnel to assess the PASC items was logical and intrusive. Ultimately there were no PASC items at CR-3.

The SWEs were questioned and they revealed that the equipment at CR-3 was common to other sites that had higher Peak Ground Accelerations (PGA) and some of the equipment at other sites was at higher elevations subject to in-structure seismic amplification. The 0.10g Safe Shutdown Earthquake (SSE) PGA at the site provided excess margin which reduced the number of PASCs at this site. Review of USI A-46/IPEEE documentation revealed this conclusion was consistent with the past inspections. The Peer Review reviewed the items that were addressed by work orders and CAP items, reviewed the

U. S. Nuclear Regulatory Commission, Attachment 8 3F1 112-06 Page 4 of 4 supporting documentation, and assessed the SWE's input and concluded that the identified items were not PASCs and that the conclusion of "no adverse conditions" is justifiable.

Review of Licensing Basis A Peer Review for the developed licensing basis evaluations, including the decisions for entering potentially adverse seismic conditions in the CR-3 CAP was performed and found to be acceptable.

Review of Submittal Report The Peer Review reviewed the submittal report and it was found to be consistent with the information provided in the inspection reports and the supporting documentation and met the objectives and requirements of the 50.54(f) letter.

Summary The Peer Review concluded the program was controlled and performed in accordance with the guidance outlined in EPRI 1025286. The number of items in the SWEL met and exceeded the minimum requirements and was distributed appropriately among the various criteria. The types of issues encountered were appropriate for the seismic demand for the site. Several significant modifications have been made at the site and these improvements were included in the component sampling.

Several housekeeping items were encountered resulting in a number of work requests and CAP items. The site addressed most of the items during the inspections.

These items did not present an operability issue because the unit was in shutdown, the fuel removed from the vessel and in the Spent Fuel Pool at the time of the inspection. A general impression of the SWEs was maintenance was being performed at the site and the site was taking advantage of the fact that the fuel was offloaded. The inspections were performed as if the unit was in power operation and were evaluated accordingly.

In conclusion, the Peer Review found the personnel involved in the inspections had a thorough knowledge of the site before the inspections and inspected the SWEL items in accordance with provided guidance. The conditions encountered and the degree of severity of the conditions indicates that CR-3 is conducting its maintenance and modification programs with consideration of seismic requirements.

The performed inspections and assessments were conducted in accordance with the guidance provided in EPRI 1025286. The results were assessed to be reasonable and consistent with seismic demand for the region and age of the unit.

ENCLOSURE 2 REPORT REVIEW BY SITE MANAGEMENT FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT DOCKET NO. 50-302 LICENSE NO. DPR-72 REPORT REVIEW BY SITE MANAGEMENT

U. S. Nuclear Regulatory Commission 3F1 112-06 Page 1 of 1 REPORT REVIEW BY SITE MANAGEMENT This submittal report is provided to the Nuclear Regulatory Commission in response to its request for information.

Specifically, by letter dated March 12, 2012, the Staff requested licensees to provide information regarding recommendation 2.3 (Seismic) of the Near-Term Task Force Review of insights from the Fukushima Dai-lchi Accident.

The report provides information for the Crystal River Unit 3 Nuclear Generating Plant regarding the performance of seismic walkdowns to identify and address degraded, non-conforming or unanalyzed conditions and to verify the current plant configuration with the current seismic licensing basis.

The information provided herein and the activities described in this report are consistent with the guidance provided by the Electric Power Research Institute's (EPRI) 2012 Technical Report 1025286 titled "Seismic Walkdown Guidance For Resolution of Fukushima Near-Term Task Force Recommendation 2.3: Seismic."

The signatures below document site management review of this document:

Signatures Date Site Fukus ima Program Manager

,,esi (RE "or Seismic EC)

I ~/1-2/ - / 2..

Sit~

ngineer MIVagement (R Superv 'or of higher) 7__

I/ JA Additionally, the Walkdown Report is ýubmitted under cover letter signed by management.

senior site

ENCLOSURE 3 REPORT REVIEW BY SITE MANAGEMENT FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT DOCKET NO. 50-302 LICENSE NO. DPR-72 LIST OF REGULATORY COMMITMENTS

U. S. Nuclear Regulatory Commission 3F1 112-06 Page 1 of 1 LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by Florida Power Corporation (FPC) in this document. Any other actions discussed in the submittal represent intended or planned actions by FPC.

They are described to the NRC for the NRC's information and are not regulatory commitments.

Please notify the CR-3 Superintendent, Licensing and Regulatory Programs of any questions regarding this document or any associated regulatory commitments.

Commitment Due Date The remaining seismic inspections (8) will be completed December 31, 2014 and the updated report submitted to the NRC