CNS-14-130, Expedited Seismic Evaluation Process Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident

Expedited Seismic Evaluation Process Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident

ML15002A261
Person / Time
Site:	Catawba
Issue date:	12/31/2014
From:	Henderson K Duke Energy Carolinas, Duke Energy Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CNS-14-130
Download: ML15002A261 (65)

Text

Kelvin Henderson f

DUKE Vice President ENERGYa Catawba Nuclear Station Duke Energy CN01VP 1 4800 Concord Road York, SC 29745 o: 803.701.4251 f: 803.701.3221 10 CFR 50.54(f)

CNS-14-130 December 31, 2014 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 Duke Energy Carolina, LLC (Duke Energy)

Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 Renewed License Numbers NPF-35 and NPF-52

Subject:

Catawba Nuclear Station Expedited Seismic Evaluation Process Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f)

Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident

References:

1. NRC Letter, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated March 12, 2012, ADAMS Accession No. ML12053A340

2. NEI Letter, Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations, dated April 9, 2013, ADAMS Accession No. ML13101A379

3. NRC Letter, Electric Power Research Institute Final Draft Report 3002000704, "Seismic Evaluation Guidance: Augmented Approach for the Resolution of Near-Term Task Force Recommendation 2.1: Seismic, "as an Acceptable Alternative to the March 12, 2012, Information Request for Seismic Reevaluations, dated May 7, 2013, ADAMS Accession No. ML13106A331

4. Duke Letter, Seismic Hazard and Screening Report (CEUS Sites), Response to NRC 10 CFR 50.54(f) Request for Additional 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 31, 2014, ADAMS Accession No. ML14093A052

United States Nuclear Regulatory Commission Page 2 December 31, 2014 Ladies and Gentlemen:

On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued a 50.54(f) letter to all power reactor licensees and holders of construction permits in active or deferred status. of Reference 1 requested each addressee located in the Central and Eastern United States (CEUS) to submit a Seismic Hazard Evaluation and Screening Report within 1.5 years from the date of Reference 1.

In Reference 2, the Nuclear Energy Institute (NEI) requested NRC agreement to delay submittal of the final CEUS Seismic Hazard Evaluation and Screening Reports so that an update to the Electric Power Research Institute (EPRI) ground motion attenuation model could be completed and used to develop that information. NEI proposed that descriptions of subsurface materials and properties and base case velocity profiles be submitted to the NRC by September 12, 2013, with the remaining seismic hazard and screening information submitted by March 31, 2014 (Reference 4). NRC agreed with that proposed path forward in Reference 3.

Reference 1 requested that licensees provide interim evaluations and actions taken or planned to address the higher seismic hazard relative to the design basis, as appropriate, prior to completion of the risk evaluation. In accordance with the NRC endorsed guidance in Reference 3, the attached Expedited Seismic Evaluation Process (ESEP) Report for Catawba provides the information described in Section 7 of Reference 3 in accordance with the schedule identified in Reference 2.

There are no regulatory commitments associated with this submittal.

Should you have any questions concerning this letter or require additional information, please contact Phil Barrett at (803) 701-4138.

I declare under penalty of perjury that the foregoing is true and correct. Executed on December 31, 2014.

Sincerely, Kelvin Henderson Vice President, Catawba Nuclear Station - Expedited Seismic Evaluation Process (ESEP) Report

United States Nuclear Regulatory Commission Page 3 December 31, 2014 xc:

V.M..McCree, Regional Administrator U. S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Avenue NE, Suite 1200 Atlanta, GA 30303-1257 William M. Dean, Director, Office of Nuclear Reactor Regulation US. Nuclear Regulatory Commission One White Flint North, Mailstop 13-HI6M 11555 Rockville Pike Rockville, MD 20852-2738 G. E. Miller U.S. Nuclear Regulatory Commission One White Flint North, Mailstop 8 G9A 11555 Rockville Pike Rockville, MD 20852-2738 G.A. Hutto NRC Senior Resident Catawba Nuclear Station Justin Folkwein American Nuclear Insurers 95 Glastonbury Blvd., Suite 300 Glastonbury, CT 06033-4453 EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 Renewed License Numbers NPF-35 and NPF-52

EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT December 8, 2014 Revision 1 Duke Energy Catawba Nuclear Station Page I of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 EXPEDITED SEISMIC EVALUATION PROCESS REPORT TABLE OF CONTENTS 1.0 PURPOSE AND OBJECTIVE...............................................................................................

5 2.0 BRIEF

SUMMARY

OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES...................... 5 2.1 M aintain Core Cooling and Heat Rem oval FLEX Flow Path...................................... 6 2.1.1 Steam Generators Available Phase 1..................................................................

6 2.1.2 Steam Generators Available Phase 2..................................................................

6 2.1.3 Steam Generators Available Phase 3..................................................................

6 2.1.4 Steam Generators Not Available Phase 1...........................................................

6 2.1.5 Steam Generators Not Available Phase 2...........................................................

6 2.1.6 Steam Generators Not Available Phase 3...........................................................

7 2.2 M aintain RCS Inventory FLEX Flow Path.................................................................... 7 2.2.1 Phase 1....................................................................................................................

7 2.2.2 Phase 2....................................................................................................................

7 2.2.3 Phase 3....................................................................................................................

7 2.3 M aintain Containm ent FLEX Flow Path...................................................................

8 2.3.1 Phase 1....................................................................................................................

8 2.3.2 Phase 2....................................................................................................................

8 2.3.3 Phase 3....................................................................................................................

8 3.0 EQUIPMENT SELECTION PROCESS AND EXPEDITED SEISMIC EQUIPMENT LIST (E S E L )...................................................................................................................................

8 3.1 Equipm ent Selection Process and ESEL....................................................................

8 3.1.1 ESEL Developm ent...............................................................................................

9 3.1.2 Power-Operated Valves....................................................................................

10 3.1.3 Pull Boxes..............................................................................................................

10 3.1.4 Term ination Cabinets.........................................................................................

10 3.1.5 Critical Instrum entation Indicators....................................................................

11 3.1.6 Phase 2 and Phase 3 Piping Connections..........................................................

11 3.2 Justification for Use of Equipment that is not the Primary Means for FLEX Im plem entation..........................................................................................................

11 4.0 GROUND M OTION RESPONSE SPECTRUM (GM RS)........................................................

11 4.1 Plot of GM RS Subm itted by the Licensee...............................................................

11 4.2 Com parison to Safe Shutdown Earthquake (SSE)...................................................

13 5.0 REVIEW LEVEL GROUND M OTION (RLGM )....................................................................

15 5.1 Description of RLGM Selected................................................................................

15 5.2 M ethod to Estim ate In-Structure Response Spectra (ISRS)..................................... 16 6.0 SEISM IC M ARGIN EVALUATION APPROACH.................................................................

16 6.1 Sum m ary of M ethodologies Used..........................................................................

17 6.2 HCLPF Screening Process........................................................................................

19 6.3 HCLPF Capacity Determ ination..............................................................................

20 Page 2 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 6.4 Functional Capacity Screening Using EPRI NP-6041-SL..........................................

20 6.5 Seism ic W alkdow n Approach.................................................................................

20 6.5.1 W alkdow n Approach........................................................................................

20 6.5.2 Application of Previous Walkdown Information...............................................

21 6.5.3 Significant W alkdow n Findings........................................................................

22 6.6 HCLPF Calculation Process......................................................................................

22 6.7 Functional Evaluations of Relays.............................................................................

24 6.8 Tabulated ESEL HCLPF Values (Including Key Failure Modes).................................

24 7.0 IN A CCESSIBLE ITEM S.......................................................................................................

25 7.1 Identification of ESEL Items Inaccessible for Walkdowns......................................

25 7.2 Planned Walkdown / Evaluation Schedule / Close Out...........................................

25 8.0 ESEP CONCLUSIONS AND RESULTS...............................................................................

27 8.1 Supporting Inform ation..........................................................................................

27 8.2 Identification of Planned M odifications.................................................................

28 8.3 M odification Im plem entation Schedule.................................................................

28 8.4 Sum m ary of Planned Actions..................................................................................

29 9.0 R E FER E N C ES......................................................................................................................

29 APPENDICES APPENDIX A Catawba Nuclear Station Unit 1 ESEL and HCLPF Results APPENDIX B Catawba Nuclear Station Unit 2 ESEL and HCLPF Results APPENDIX C CNS FLEX Flow Paths FIGURES Figure 4-1.

Figure 4-2.

Figure 4-3.

Figure 4-4.

Figure 5-1.

Figure 6-1.

CNS GMRS (5% Damping) - Tabular Form [4]........................................................

12 CNS GMRS (5% Damping) - Graphical Form [4]...................................................

13 CN S SSE (5% Dam ping)...........................................................................................

14 Comparison of CNS GMRS and SSE (5% Damping)...............................................

14 CNS RLGM (5% Dam ping).....................................................................................

16 Comparison of CNS SSE and RLGM vs. IPEEE RLE.................................................

19 Page 3 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 TABLES Table 4-1. CNS SSE (5% Damping) - Tabular Form [4].............................................................

13 Table 5-1. CNS RLGM (5% Dam ping)........................................................................................

15 Table 6-1. CNS IPEEE RLE (5% Dam ping).................................................................................

18 Table 6-2. Unit 1 Components that Require Further Evaluations and/or Modifications......

23 Table 6-3. Unit 2 Components that Require Further Evaluations and/or Modifications......

23 Table 7-1. Unit 2 Walkdowns & Walk-Bys Not Completed at Time of this Report.

(2 sh e e ts)...........................................................................................................................

2 5 Table 8-1. Sum m ary of Planned Actions.................................................................................

29 Page 4 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. I 1.0 Purpose and Objective Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March 11, 2011, Great Tohoku Earthquake and subsequent tsunami, the Nuclear Regulatory Commission (NRC) established a Near-Term Task Force (NTTF) to conduct a systematic review of NRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena. Subsequently, the NRC issued a 50.54(f) letter on March 12, 2012 [1], requesting information to assure that these recommendations are addressed by all U.S. nuclear power plants. The 50.54(f) letter [1]

requests that licensees and holders of construction permits under 10 CFR Part 50 reevaluate the seismic hazards at their sites against present-day NRC requirements and guidance. Depending on the comparison between the reevaluated seismic hazard and the current design basis, further risk assessment may be required. Assessment approaches acceptable to the staff include a seismic probabilistic risk assessment (SPRA), or a seismic margin assessment (SMA). Based upon the assessment results, the NRC staff will determine whether additional regulatory actions are necessary.

This report describes the Expedited Seismic Evaluation Process (ESEP) undertaken for Catawba Nuclear Station (CNS). The intent of the ESEP is to perform an interim action in response to the NRC's 50.54(f) letter [1] to demonstrate seismic margin through a review of a subset of the plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events.

The ESEP is implemented using the methodologies in the NRC-endorsed guidance in Electric Power Research Institute (EPRI) 3002000704, Seismic Evaluation Guidance:

Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic [2].

The objective of this report is to provide summary information describing the ESEP evaluations and results. The level of detail provided in the report is intended to enable NRC to understand the inputs used, the evaluations performed, and the decisions made as a result of the interim evaluations.

2.0 Brief Summary of the FLEX Seismic Implementation Strategies The CNS FLEX strategies for Reactor Core Cooling and Heat Removal, Reactor Inventory Control/Long-Term Subcriticality, and Containment Function are summarized below.

This summary is derived from the CNS Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order EA-12-049 [3] (as supplemented by subsequent six-month updates [22], [23], and [24]).

Conceptual sketches showing FLEX strategy flow paths are included in Appendix C.

Page 5 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 2.1 Maintain Core Cooling and Heat Removal FLEX Flow Path The FLEX flow path for maintaining core cooling and heat removal addresses plant operating conditions with or without steam generators available. Both scenarios are addressed below.

2.1.1 Steam Generators Available Phase 1 Phase 1 of the Maintain Core Cooling and Heat Removal Strategy relies on routing the Condenser Circulation Water (RC) inventory held in the RC piping through the steam generators. The steam generators route steam to the atmosphere via power-operated relief valves SVO0, SV07, SV13, and SV19. Other piping used by both segments includes the piping downstream of the head vent valves and Valves CA178 and CA174 in the Auxiliary Feedwater (CA) system.

2.1.2 Steam Generators Available Phase 2 Phase 2 of the Maintain Core Cooling and Heat Removal Strategy introduces cooling water from the ultimate heat sink for use with the steam generators.

The water will be provided by diesel-driven portable pumps via any of a number of connection points to either the Nuclear Service Water System (RN) or the Steam Generator Wet Layup Recirculation System (BW) which vary depending on which equipment is available after the initiating event. The RN connection points will be fed by a high-capacity diesel-driven pump and will provide an uninterrupted water supply to the Turbine-Driven Auxiliary Feed Water Pump (TDAFWP) for steam generator makeup, as long as it is operational. Steam generator overfill will be controlled by starting and stopping the TDAFWP as necessary or manually throttling the CA flow control valves. The BW connection points would be fed by a low-capacity, low-pressure diesel-driven pump and include piping isolated by Unit 1 Valves 1BW51, 1BW52, 1BW53, and 1BW54 and Unit 2 Valves 2BW44, 2BW46, 2BW48, and 2BW50. These connection points would feed water directly to the steam generators.

2.1.3 Steam Generators Available Phase 3 Phase 3 of the Maintain Core Cooling and Heat Removal Strategy continues operation as Phase 2 with the addition of providing cooling for specific components in the system. Phase 3 starts when equipment arrives from the National SAFER Response Center (NSRC) to provide indefinite coping capabilities.

2.1.4 Steam Generators Not Available Phase 1 There are no necessary actions to provide coping during Phase 1. CNS has no means of providing borated Reactor Coolant System (RCS) makeup for Phase 1.

2.1.5 Steam Generators Not Available Phase 2 A low pressure pump will provide borated makeup to the RCS if the event were to occur during a refueling outage. This pump will be the same diesel-driven Page 6 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. I low-pressure pump identified for connection to the BW system, as both of these strategies will not be performed at the same time. The suction supply for the portable pump will come from a new connection on the Refueling Water Storage Tank (FWST) supply line for the Spent Fuel Pool between valves KF-1O1B and 103A on Units 1 and 2. The discharge from the portable pump will be into a new connection on the A Train Safety Injection System (NI) pump discharge piping that feeds the RCS hot or cold legs.

If the reactor vessel head is still installed when the event occurs, the reactor head vent valves will be powered using the motor control center back-feed strategy and portable diesel generators. The RCS depressurization will be initiated from the reactor head vent valves which will provide indefinite coping for depressurization. This method allows vapor to be vented in situations where voids may develop during the RCS cooldown/depressurization phase and allows the discharge of liquid inventory if required while injecting the required borated water.

If the reactor vessel head is not installed and fuel is still in the core when the event occurs, discharge of liquid inventory while injecting the required borated water will simply overflow out of the reactor vessel into the cavity/containment, keeping the fuel covered and cooled.

2.1.6 Steam Generators Not Available Phase 3 This strategy will be the same as that used for the Phase 3 strategy when the steam generators were available with the exception that secondary side cooling will not be required.

2.2 Maintain RCS Inventory FLEX Flow Path 2.2.1 Phase 1 The CNS OIP identifies that a Phase I strategy is not required as the core is not in jeopardy of being uncovered until approximately 55 hours6.365741e-4 days <br />0.0153 hours <br />9.093915e-5 weeks <br />2.09275e-5 months <br /> after the initiating event.

2.2.2 Phase 2 The Phase 2 strategy for re-establishing reactor make-up water uses portable pumps to bypass the safety injection pumps, providing make-up water from the FWST to the safety injection pump discharge piping. The portable pump will be connected via connection points which will be installed by the end of each corresponding units' refueling outage.

2.2.3 Phase 3 Phase 3 of the Maintain RCS Inventory strategy relies on the use of a large diesel generator to power the residual heat removal system pumps. This diesel generator will also be provided by the NSRC.

Page 7 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 2.3 Maintain Containment FLEX Flow Path 2.3.1 Phase 1 The CNS strategy for maintaining containment during Phase 1 relies upon passive cooling from the ice condenser. As the system is passive and does not rely on flow, a FLEX flow path was not established for Phase 1.

2.3.2 Phase 2 The Phase 2 Maintain Containment strategy has two portions. At least one train of hydrogen igniters will be re-powered. Additionally, it is assumed that forced air circulation will be required for containment cooling based on pending analyses. This will be accomplished by operation of the H2 skimmer fans.

2.3.3 Phase 3 The Phase 3 strategy assumes that forced air circulation will be required for containment cooling based on pending analyses. This will be accomplished by operation of the VX Containment Air Return fans, H2 skimmer fans, and two of the Lower Containment Ventilation Units.

3.0 Equipment Selection Process and Expedited Seismic Equipment List (ESEL)

The complete ESELs for Unit 1 and Unit 2 are presented in Appendices A and B, respectively. These lists were developed in Augmented Approach for Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic: Seismic-- Catawba Nuclear Station Expedited Seismic Equipment List, ARES Corporation Report No.

030321.13.01-005, Duke Energy Document No. CNC-1211.00-06-0004 [17].

The selection of equipment for the ESEL followed the guidelines of EPRI 3002000704 [2].

3.1 Equipment Selection Process and ESEL The selection of equipment to be included on the ESEL was based on installed plant equipment credited in the FLEX strategies during Phase 1, 2 and 3 mitigation of a Beyond Design Basis External Event, as outlined in the CNS alP in Response to the March 12, 2012, Commission Order EA-12-049 [3], as supplemented by subsequent six-month regulatory updates [22], [23], and [24].

The OIP provides the CNS FLEX mitigation strategy and serves as the basis for equipment selected for the ESEP.

The scope of "installed plant equipment" includes equipment relied upon for the FLEX strategies to sustain the critical functions of core cooling and containment integrity consistent with the CNS aIP [3] and supplemented by subsequent six-month regulatory updates [22], [23], and [24]. FLEX recovery actions are excluded from the ESEP scope per EPRI 3002000704 [2]. The overall list of planned FLEX modifications and the scope for consideration herein is limited to those required to support core cooling, reactor coolant inventory and Page 8 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 subcriticality, and containment integrity functions. Portable and pre-staged FLEX equipment (not permanently installed) are excluded from the ESEL per EPRI 3002000704 [2].

The ESEL component selection followed the EPRI guidance outlined in Section 3.2 of EPRI 3002000704 [2].

1. The scope of components is limited to that required to accomplish the core cooling and containment safety functions identified in Table 3-2 of EPRI 3002000704 [2]. The instrumentation monitoring requirements for core cooling/containment safety functions are limited to those outlined in the EPRI 3002000704 [2] guidance, and are a subset of those outlined in the CNS OIP [3] and subsequent updates [22], [23], and [24].

2. The scope of components is limited to installed plant equipment and FLEX connections necessary to implement the CNS OIP [3] and subsequent updates [22], [23], and [24] as described in Section 2.

3. The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e., either "Primary" or "Back-up/Alternate").

4. The "Primary" FLEX success path is to be specified. Selection of the "Back-up/Alternate" FLEX success path must be justified.

5.

Phase 3 coping strategies are included in the ESEP scope, whereas recovery strategies are excluded.

6. Structures, systems, and components excluded per the EPRI 3002000704 [2]

guidance are:

Structures (e.g., containment, Reactor Building, Control Building, Auxiliary Building, etc.)

Piping, cabling, conduit, HVAC, and their supports.

Manual valves and rupture disks.

" Power-operated valves not required to change state as part of the FLEX mitigation strategies.

Nuclear steam supply system components (e.g., reactor pressure vessel and internals, reactor coolant pumps and seals, etc.)

7. For cases in which neither train was specified as a primary or back-up strategy, then only one train component (generally 'A' train) is included in the ESEL.

3.1.1 ESEL Development The ESEL was developed by reviewing the CNS OIP [3] and subsequent updates [22], [23], and [24] to determine the major equipment involved in the FLEX strategies. Further reviews of plant drawings (e.g., Process and Instrumentation Diagrams (P&IDs) and Electrical One-Line Diagrams) were Page 9 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 performed to identify the boundaries of the flow paths to be used in the FLEX strategies and to identify specific components in the flow paths needed to support implementation of the FLEX strategies. Boundaries were established at an electrical or mechanical isolation device (e.g., isolation amplifier, valve, etc.)

in branch circuits/branch lines off the defined strategy electrical or fluid flow path. P&IDs were the primary reference documents used to identify mechanical components and instrumentation. The flow paths used for FLEX strategies were selected and specific components were identified using detailed equipment and instrument drawings, piping isometrics, electrical schematics and one-line drawings, system descriptions, design basis documents, etc.

3.1.2 Power-Operated Valves Page 3-3 of EPRI 3002000704 [2] notes that power-operated valves not required to change state are excluded from the ESEL. Page 3-2 also notes that

"... functional failure modes of electrical and mechanical portions of the installed Phase 1 equipment should be considered (e.g., RCIC/AFW trips)." To address this concern, the following guidance is applied in the CNS ESEL for functional failure modes associated with power-operated valves:

" Power-operated valves that remain energized during the Extended Loss of all AC Power (ELAP) events (such as DC powered valves), were included on the ESEL.

" Power-operated valves not required to change state as part of the FLEX mitigation strategies were not included on the ESEL. The seismic event also causes the ELAP event; therefore, the valves are incapable of spurious operation as they would be de-energized.

Power-operated valves not required to change state as part of the FLEX mitigation strategies during Phase 1, and are re-energized and operated during subsequent Phase 2 and 3 strategies, were not evaluated for spurious valve operation as the seismic event that caused the ELAP has passed before the valves are re-powered.

3.1.3 Pull Boxes Pull boxes were deemed unnecessary to add to the ESELs as these components provide completely passive locations for pulling or installing cables. No breaks or connections in the cabling are included in pull boxes. Pull boxes were considered part of conduit and cabling, which are excluded in accordance with EPRI 3002000704 [2].

3.1.4 Termination Cabinets Termination cabinets, including cabinets necessary for FLEX Phase 2 and Phase 3 connections, provide consolidated locations for permanently connecting multiple cables. The termination cabinets and the internal connections provide a completely passive function; however, the cabinets are included in the ESEL to Page 10 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 ensure industry knowledge on panel/anchorage failure vulnerabilities is addressed.

3.1.5 Critical Instrumentation Indicators Critical indicators and recorders are typically physically located on panels/cabinets and are included as separate components; however, seismic evaluation of the instrument indication may be included in the panel/cabinet seismic evaluation (rule-of-the-box).

3.1.6 Phase 2 and Phase 3 Piping Connections Item 2 in Section 3.1 above notes that the scope of equipment in the ESEL includes "...

FLEX connections necessary to implement the CNS OIP [3] and subsequent updates [22], [23], and [24] as described in Section 2." Item 3 in Section 3.1 also notes that "The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e., either 'Primary' or 'Back-up/Alternate')."

Item 6 in Section 3 above goes on to explain that "Piping, cabling, conduit, HVAC, and their supports..." are excluded from the ESEL scope in accordance with EPRI 3002000704 [2].

Therefore, piping and pipe supports associated with FLEX Phase 2 arid Phase 3 connections are excluded from the scope of the ESEP evaluation. However, any active valves in FLEX Phase 2 and Phase 3 connection flow paths are included in the ESEL.

3.2 Justification for Use of Equipment that is not the Primary Means for FLEX Implementation The ESEL only uses equipment that is the primary means of implementing FLEX strategy.

4.0 Ground Motion Response Spectrum (GMRS) 4.1 Plot of GMRS Submitted by the Licensee The CNS GMRS used to select the ESEP Review Level Ground Motion (RLGM) was included in the CNS Seismic Hazard and Screening Report [4]. Digitized GMRS frequency and acceleration values from the CNS Seismic Hazard and Screening Report [4] are shown in Figure 4-1, which is Table 2.4-1 from [4]. The CNS GMRS is plotted in Figure 4-2.

Page 11 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 Table 2.4-1 UHRS and GMRS at control point for Catawba (5% of critical damping response spectra)

Freg (Hz) 1E-4 UHRS (g) 1E-5 UHRS (g)

GMRS (g) 100 2.19E-01 6.91 E-01 3.29E-01 90 2.21 E-01 7.02E-01 3.34E-01 80 2.26E-01 7.25E-01 3.45E-01 70 2.40E-01 7.81E-01 3.70E-01 60 2.75E-01 9.19E-01 4.33E-01 50 3.51E-01 1.20E+00 5.63E-01 40 4.39E-01 1.48E+00 6.98E-01 35 4.67E-01 1.56E+00 7.35E-01 30 4.82E-01 1.58E+00 7.48E-01 25 4.79E-01 1.54E+00 7.31E-01 20 4.66E-01 1.47E+00 6.99E-01 15 4.31E-01 1.32E+00 6.33E-01 12.5 4.06E-01 1.22E+00 5.89E-01 10 3.74E-01 1.11E+00 5.35E-01 9

3.52E-01 1.03E+00 4.98E-01 8

3.29E-01 9.49E-01 4.61 E-01 7

3.05E-01 8.63E-01 4.21E-01 6

2.77E-01 7.72E-01 3.77E-01 5

2.45E-01 6.67E-01 3.28E-01 4

2.03E-01 5.36E-01 2.65E-01 3.5 1.80E-01 4.67E-01 2.31E-01 3

1.56E-01 3.97E-01 1.98E-01 2.5 1.27E-01 3.16E-01 1.58E-01 2

1.19E-01 2.90E-01 1.45E-01 1.5 9.49E-02 2.26E-01 1.14E-01 1.25 8.03E-02 1.89E-01 9.55E-02 1

7.15E-02 1.64E-01 8.35E-02 0.9 6.96E-02 1.60E-01 8.14E-02 0.8 6.73E-02 1.55E-01 7.87E-02 0.7 6.36E-02 1.47E-01 7.44E-02 0.6 5.76E-02 1.33E-01 6.74E-02 0.5 4.90E-02 1.13E-01 5.74E-02 0.4 3.92E-02 9.04E-02 4.59E-02 0.35 3.43E-02 7.91 E-02 4.02E-02 0.3 2.94E-02 6.78E-02 3.44E-02 0.25 2.45E-02 5.65E-02 2.87E-02 0.2 1.96E-02 4.52E-02 2.29E-02 0.15 1.47E-02 3.39E-02 1.72E-02 0.125 1.22E-02 2.83E-02 1.43E-02 0.1 9.79E-03 2.26E-02 1.15E-02 Figure 4-1. CNS GMRS (5% Damping) -Tabular Form [4].

Page 12 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 CNS GMRS 0.8 21 1

ITT-R 0.7

-kI.

I 0.5 03

-T-

~1

-I

_____-GMRS 0.1 I

0.0 1

100 0.1 110 100 Frequency (Hz)

Figure 4-2. CNS GMRS (5% Damping) - Graphical Form [4].

The CNS Control Point is located at Elevation 544'-0", which is at the base of the mat foundation of the Reactor Buildings [4].

4.2 Comparison to Safe Shutdown Earthquake (SSE)

A description of the CNS horizontal SSE and spectral shape is included in Section 3.1 of the CNS Seismic Hazard and Screening Report [4]. The SSE is tabulated as a function of frequency in Table 4-1 and plotted in Figure 4-3.

A comparison of the CNS GMRS plotted against the SSE is shown in Figure 4-4.

Table 4-1. CNS SSE (5% Damping) - Tabular Form [4].

Frequency Spectral Acceleration (Hz)

(g) 0.33 0.06 2

0.36 6

0.36 35/PGA 0.15 Page 13 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 CNS SSE 5% Damping 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 10 100

ý

-SI 0.1 I

FrequCNSy (Hz)

Figure 4-3. CNS SSE (S% Damping).

CNS GMRS vs SSE 0.8

~~V7V

[T~___

0.7 IT T[-

ýLT I

0.6 I!

0.6 0.2

--- i--K 0.1 0.1 10 100 Frequency (Hz)

Figure 4-4. Comparison of CNS GMRS and SSE (5% Damping).

Page 14 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 5.0 Review Level Ground Motion (RLGM) 5.1 Description of RLGM Selected The procedure for determining the RLGM for the ESEP is described in Section 4 of EPRI 3002000704 [2]. The RLGM is determined by multiplying the spectral acceleration values for the 5%-damped SSE horizontal ground response spectrum by a scale factor. The scale factor is the largest ratio of spectral accelerations between the 5%-damped GMRS and the 5%-damped SSE ground response spectrum at frequencies from 1 Hz to 10 Hz, but not to exceed 2.0.

The ratio of the GMRS to the SSE over the 1 to 10 Hz frequency range is shown in Figure 4-4. The largest ratio of the GMRS to the SSE in the i to 10 Hz range is at 10 Hz. The ratio of the spectral accelerations is 0.535/0.28 = 1.91. Therefore, the RLGM is determined by multiplying the SSE ground response spectrum by 1.91. Digitized RLGM frequency and acceleration values are shown in Table 5-1.

The CNS RLGM is plotted in Figure 5-1.

Table 5-1. CNS RLGM (5% Damping).

Frequency Acceleration (Hz)

(g) 0.333 0.115 0.5 0.172 1

0.344 2

0.688 3

0.688 4

0.688 5

0.688 6

0.688 7

0.637 8

0.596 9

0.562 10 0.535 11 0.509 12 0.487 13 0.468 14 0.452 15 0.436 17.5 0.404 20 0.378 22.5 0.357 25 0.339 27.5 0.323 30 0.309 35 0.287 100 0.287 Page 15 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 CNS RLGM 0.700 0.600 0.500 oiooo

~V 0.2000 101 Frequency (Hz) 10 100 Figure 5-1. CNS RLGM (5% Damping).

5.2 Method to Estimate In-Structure Response Spectra (ISRS)

The new ISRS for the ESEP were derived by scaling the CNS design-basis SSE by the RLGM scale factor of 1.91.

6.0 Seismic Margin Evaluation Approach It is necessary to demonstrate that ESEL items have sufficient seismic capacity to meet or exceed the demand characterized by the RLGM. The seismic capacity is characterized as the peak ground acceleration (PGA) for which there is a high confidence of a low probability of failure (HCLPF). The PGA is associated with a specific spectral shape, in this case the 5%-damped RLGM spectral shape. The HCLPF capacity must be equal to or greater than the RLGM PGA. The criteria for seismic capacity determination are given in Section 5 of EPRI 3002000704 [2].

There are two basic approaches for developing HCLPF capacities:

1. Deterministic approach using the conservative deterministic failure margin (CDFM) methodology of EPRI NP-6041-SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin [7].

2. Probabilistic approach using the fragility analysis methodology of EPRI TR-103959, Methodology for Developing Seismic Fragilities [8].

Page 16 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. I 6.1 Summary of Methodologies Used Seismic capacity screening was done using information from the CNS Individual Plant Examination of External Events (IPEEE) submittal [9] and supporting documentation (CNC-1535.00-00-0005 [18], Seismic Capacity Evaluations for the IPEEE and EPRI Seismic Margins Study]).

CNS used both a SPRA [10] and a SMA to address the IPEEE. The SPRA and SMA are described in the CNS IPEEE Submittal Report [9].

The SMA conducted for Catawba is documented in EPRI NP-6359 [19], Seismic Margin Assessment of the Catawba Nuclear Station. This SMA was a trial plant review to test the EPRI Seismic Margin Methodology. It was performed prior to the publication of NRC Generic Letter 88-20, Supplement 4 [6], and NUREG-1407 [5], Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, which provided the requirements for IPEEE. Consequently, it differed in some respects from most of the SMAs that were conducted for IPEEE. The Catawba SMA was equivalent to a full-scope SMA, as defined in NUREG-1407 [5], as it included a full-scope evaluation of relays, whereas NUREG-1407 [5] placed Catawba in the focused-scope category and only required a search for low-ruggedness relays.

The SMA documented in EPRI NP-6359 [19] was for Unit 2. During the IPEEE, Duke Energy conducted an additional SMA to extend the EPRI NP-6359 [19]

results to Unit 1 and all of the items on the IPEEE Seismic Equipment List in Unit 2. This SMA is documented in CNC-1535.00-00-0005 [18].

The SMA documented in CNC-1535.00-00-0005 [18] consisted of screening walkdowns and anchorage calculations. The screening walkdowns used the screening tables from EPRI NP-6041-SL [7]. The walkdowns were conducted by registered professional engineers. Given the standards available at the time each of the two portions of the seismic review was done, the Seismic Review Team met the requirements and intent of EPRI NP-6041-SL [7] and NUREG-1407 [5]. The walkdowns were documented on Screening Evaluation Work Sheets (SEWS) from EPRI NP-6041-SL [7]. Anchorage capacity calculations utilized the CDFM criteria from EPRI NP-6041-SL [7]. Seismic demand was based on the Review Level Earthquake (RLE) selected for the Catawba Trial Plant Review, which was the Sequoyah 84th percentile site-specific spectral shape anchored to 0.3g PGA, which is similar to the NUREG/CR-0098 [11] response spectrum recommended in NUREG-1407 [5]. Frequency and acceleration values for the CNS IPEEE RLE are shown in Table 6-1. Figure 6-1 shows the EPRI NP-6359 [19] RLE compared to the SSE and RLGM response spectra.

Page 17 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 Table 6-1. CNS IPEEE RLE (5% Damping).

Freq. (Hz)

Acc. (g)

Freq. (Hz)

Acc. (g) 0.25 0.20 3.78 0.729 0.28 0.027 4.23 0.752 0.31 0.033 4.74 0.772 0.35 0.041 5.31 0.792 0.39 0.048 5.94 0.817 0.44 0.055 6.66 0.839 0.49 0.063 7.45 0.806 0.55 0.077 8.35 0.727 0.62 0.093 9.35 0.644 0.69 0.110 10.47 0.566 0.78 0.128 11.72 0.511 0.87 0.148 13.13 0.473 0.97 0.167 14.7 0.445 1.09 0.184 16.46 0.424 1.22 0.201 18.43 0.410 1.37 0.220 20.64 0.398 1.53 0.251 23.11 0.382 1.71 0.301 25.88 0.363 1.92 0.374 28.98 0.341 2.15 0.470 32.46 0.320 2.4 0.546 36.34 0.308 2.69 0.589 40.7 0.301 3.01 0.635 46 0.300 3.38 0.689 Page 18 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 CNS SSE and RLGM vs IPEEE RLE nol.

om o

• IPEEE RLEanchored at 0.30gI 0.3 14.,

0.75 SSE I

I I

I I

lI I

I i J Ž113111 IZIIIIIIIEIIIIW I

0

-ic 0.7 1 1-1 0.65 A

l 0.0 035 0.5.

0352 0.1

-w-

-- A 0.1 0.1 0.0 a

h~

TE I1.1 b.1 1

10 100 Frequency (Hz)

Figure 6-1. Comparison of CNS SSE and RLGM vs. IPEEE RLE.

6.2 HCLPF Screening Process The equipment evaluations in the IPEEE were based on plant design information, including equipment qualification test and analysis reports. Failure modes considered were functional failures, including relay chatter, and anchorage failure. The original anchorage capacities were updated as needed based on the SMA walkdowns. Seismic interactions were addressed by the SMA walkdowns.

It is seen from Figure 6-1 that the RLE envelopes the RLGM at all frequencies greater than about 3.4 Hz. The RLE is less than the RLGM at frequencies below about 3.4 Hz. This may be disregarded as there are no CNS ESEL components in this frequency range. Therefore, any components whose SMA based HCLPF exceeds the RLE can be screened out from ESEP seismic capacity determination.

The screening tables in EPRI NP-6041-SL [7] are based on ground peak spectral accelerations of 0.8g and 1.2g. These both exceed the RLGM peak spectral acceleration of 0.69g. The anchorage capacity calculations were based on SSE floor response spectra scaled to the RLE, except for equipment in the Auxiliary Building for which new floor response spectra were generated for the RLE per EPRI NP-6359 [19]. Equipment for which the screening caveats were met and for which the anchorage capacity exceeded the RLE seismic demand can be screened out from ESEP seismic capacity determination because the HCLPF capacity exceeds the RLGM.

Page 19 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. I The results of the IPEEE capacity screening are noted in Appendix A for the Unit 1 ESEL and in Appendix B for the Unit 2 ESEL. For the components that were not screened out, HCLPF capacities were determined using the deterministic EPRI NP-6041-SL [7] CDFM methodology and RLGM spectral shape and/or anchorage evaluations.

6.3 HCLPF Capacity Determination HCLPF capacities were determined by evaluating the function, anchorage, and seismic interaction failure modes. HCLPF functional capacities were determined using the screening tables in EPRI NP-6041-SL [7]. HCLPF anchorage capacities were determined using the CDFM methodology in EPRI NP-6041-SL [7]. HCLPF seismic interaction capacities were determined by walkdown screening.

6.4 Functional Capacity Screening Using EPRI NP-6041-SL The components were screened against EPRI NP-6041-SL [7], Table 2-4. ISRS were used for all components for the screening; therefore, the screening levels of EPRI NP-6041-SL [7] were increased by a factor of 1.5 (EPRI 1019200 [20],

Seismic Fragility Applications Guide Update). Thus, the accelerations for the screening levels are 1.2g and 1.8g instead of 0.8g and 1.2g.

The SSE ISRS were amplified by a factor of 1.91 throughout the frequency range and were then clipped (per EPRI 1019200 [20]), using the methodology in EPRI NP-6041-SL, Appendix 0, and the North-South and East-West clipped peaks were averaged. HCLPFs for these components are shown in Appendices A and B.

6.5 Seismic Walkdown Approach 6.5.1 Walkdown Approach Walkdowns were performed in accordance with the criteria provided in Section 5 of EPRI 3002000704 [2], which refers to EPRI NP-6041-SL [7] for the Seismic Margin Assessment process. Pages 2-26 through 2-30 of EPRI NP-6041-SL [7]

describe the seismic walkdown criteria, including the following key criteria.

"The SRT [Seismic Review Team] should "walk by" 100% of all components which are reasonably accessible and in non-radioactive or low radioactive environments. Seismic capability assessment of components which are inaccessible, in high-radioactive environments, or possibly within contaminated containment, will have to rely more on alternate means such as photographic inspection, more reliance on seismic reanalysis, and possibly, smaller inspection teams and more hurried inspections. A 100% "walk by" does not mean complete inspection of each component, nor does it mean requiring an electrician or other technician to de-energize and open cabinets or panels for detailed inspection of all components. This walkdown is not intended to be a QA or QC review or a review of the adequacy of the component at the SSE level.

Page 20 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 If the SRT has a reasonable basis for assuming that the group of components are similar and are similarly anchored, then it is only necessary to inspect one component out of this group. The "similarity-basis" should be developed before the walkdown during the seismic capability preparatory work (Step 3) by reference to drawings, calculations or specifications. The one component or each type which is selected should be thoroughly inspected which probably does mean de-energizing and opening cabinets or panels for this very limited sample. Generally, a spare representative component can be found so as to enable the inspection to be performed while the plant is in operation. At least for the one component of each type which is selected, anchorage should be thoroughly inspected.

The walkdown procedure should be performed in an ad hoc manner. For each class of components the SRT should look closely at the first items and compare the field configurations with the construction drawings and/or specifications. If a one-to-one correspondence is found, then subsequent items do not have to be inspected in as great a detail. Ultimately the walkdown becomes a "walk by" of the component class as the SRT becomes confident that the construction pattern is typical. This procedure for inspection should be repeated for each component class; although, during the actual walkdown the SRT may be inspecting several classes of components in parallel. If serious exceptions to the drawings or questionable construction practices are found then the system or component class must be inspected in closer detail until the systematic deficiency is defined.

The 100% "walk by" is to look for outliers, lack of similarity, anchorage which is different from that shown on drawings or prescribed in criteria for that component, potential SI [Seismic Interaction] problems, situations that are at odds with the team members' past experience, and any other areas of serious seismic concern. If any such concerns surface, then the limited sample size of one component of each type for thorough inspection will have to be increased. The increase in sample size which should be inspected will depend upon the number of outliers and different anchorages, etc., which are observed. It is up to the SRT to ultimately select the sample size since they are the ones who are responsible for the seismic adequacy of all elements which they screen from the margin review. Appendix D gives guidance for sampling selection."

6.5.2 Application of Previous Walkdown Information Many of the components had been walked down previously during IPEEE evaluations and have documented SEWS recording the results. Credit is given to

'EPRI 3002000704 [2] page 5-4 limits the ESEP seismic interaction reviews to "nearby block walls" and "piping attached to tanks" which are reviewed "to address the possibility of failures due to differential displacements."

Other potential seismic interaction evaluations are "deferred to the full seismic risk evaluations performed in accordance with EPRI 1025287 [15]."

Page 21 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. I these walkdowns since they were performed by qualified Seismic Review Teams.

A walk-by of these components was performed and documented, except for those Unit 2 items which will be reviewed in March 2015, as detailed in Section 7.2. The primary objective of a walk-by is to verify that the component and/or anchorage has not degraded since the original walkdown and to verify that the component is free of interaction issues that may have developed since the original walkdown.

Walkdowns were performed on all ESEL components which were not previously walked down during the IPEEE evaluations.

Masonry block walls were evaluated as part of IPEEE and shown to meet the RLE demand. Therefore, they also meet the RLGM demand. Proximity of block walls to ESEL components was noted on the SEWS forms and the presence of block walls was considered in determining a HCLPF and identification of key failure modes.

6.5.3 Significant Walkdown Findings All of the ESEL components included in the walkdowns and walk-bys completed to date were determined to have an existing capacity greater than the RLGM, with the exception of the components listed in Tables 6-2 and 6-3. These components require further evaluation and/or modification in order to have a capacity greater than the RLGM.

6.6 HCLPF Calculation Process ESEL items not included in the previous CNS IPEEE evaluations were evaluated using the criteria in EPRI NP-6041-SL [7]. The evaluations included the following steps:

Performing seismic capability walkdowns for equipment not included in previous seismic walkdowns to evaluate the equipment-installed plant conditions; Performing screening evaluations using the screening tables in EPRI NP-6041-SL [7] as described in Section 6.2; and Performing HCLPF calculations considering various failure modes that include both structural failure modes (e.g., anchorage, load path, etc.) and functional failure modes.

All HCLPF calculations were performed using the CDFM methodology and are documented in Expedited Seismic Evaluation Process for Implementation of Seismic Risk Evaluations at Catawba Nuclear Station, ARES Corporation Report 030321.13.01-003 (Duke Energy Document No. CNC-1211.00-06-0003),

Appendix D, "Calculation" [21]. HCLPF results and key failure modes are included in the ESEL tables in Appendices A and B.

Page 22 of 61

Exr)edited Seismic Evaluation Process ReDort. Catawba Nuclear Station Rev. I Endied. eim. cEvluaio Prcs R

or.

C.abaNcla

.Saio.ev.

Table 6-2. Unit 1 Components that Require Further Evaluations and/or Modifications.

ESEL EIN Description Bldg.

El.

Locati Modification/

Problem Description Action Description ID on Recommendation Including PIP Numbers 45 1CA36 Auxiliary AU 55 BB-Brace or stiffen Vertical channel which supports valve Brace/revise vertical tubing support and Feedwater to X

4 49 nearby vertical operator tubing is loose and may sway tubing to achieve seismic ruggedness.

Steam Generator channel.

significantly in a seismic event and possibly PIP# C-14-09014 1D Isolation Valve damage the tubing and/or pressure regulator and render the valve inoperable via the air-operator.

87 1PSS Primary Sample AU 54 FF-54 Anchor sink and The absence of anchorage of the sink and Inspect sink and hood following seismic Sink 1B X

3 hood.

hood will allow the sink and hood to move event. If significant primary leak is present, and possibly overturn in a seismic event, manually close valve 1NM26B to isolate This may damage the sample piping and damaged tubing.

tubing located within the cabinet/hood.

PIP# C-14-09014 197 1EATC Essential Area AU 57 FF-56 Move conduit The cabinet will likely impact a conduit Move conduit support for clearance.

12 Terminal Cabinet X

7 support.

support during a seismic event. This PIP# C-14-09014 interaction has the potential to cause relay chatter.

Table 6-3. Unit 2 Components that Require Further Evaluations and/or Modifications.

ESEL EIN Description Bldg.

El.

Locati Modification/

Problem Description Action Description ID on Recommendation Including PIP Numbers 38 2CA25 2CAPUTD AU 54 BB-Remove grating The grating is touching the valve inlet piping.

Trim grating to achieve 1" min. clearance.

7 Feedwater Safety X

0 63 stub.

The -1" grating cantilever should be removed to PIP# C-14-09014 Valve provide additional clearance.

41 2CA64 Auxiliary AU 55 BB-Brace or relocate Potential interaction between fairly lightweight, Relocate/add supports to eliminate Feedwater to X

7 64 nearby rod-hung rod-hung cable tray which may impact valve air possible interaction.

Steam Generator cable lines (soft target). Addition of simple bracing on 2A tray/support to rod-hung cable tray (near 2CA64) or relocation of PIP# C-14-09014 eliminate the support will mitigate this issue.

interaction.

44 2CA36 Auxiliary AU 55 BB-Brace or relocate Potential interaction between fairly lightweight, Relocate/add supports to eliminate Feedwater to X

4 65 nearby rod-hung rod-hung piping which may impact valve air lines possible interaction.

Steam Generator piping support to (soft target). Addition of simple bracing on rod-2D eliminate hung piping (near 2CA36) or relocation of the PIP# C-14-09014 interaction, support will mitigate this issue.

AUX = Auxiliary PIP = Pertormance Improvement Plan Page 23 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 6.7 Functional Evaluations of Relays Two types of relays (located in four cabinets) in the ESEL associated with the FLEX Phase 1 response required functional evaluations. Each relay was evaluated using the SMA relay evaluation criteria in Section 3 of EPRI NP-6041-SL [7].

Specific seismic qualification test-based capacities were available for the relays in existing plant documentation. Relay capacity to demand evaluations were performed by comparing the test-based capacities with the in-cabinet seismic demand. The in-cabinet demand was determined by scaling the ESEP ISRS by the in-cabinet amplification factors from EPRI NP-6041-SL [7]. In each case, the capacity exceeded the demand.

The ESEP relay functional evaluations are documented in Expedited Seismic Evaluation Process for Implementation of Seismic Risk Evaluations at Catawba Nuclear Station, ARES Corporation Report 030321.13.01-003 (Duke Energy Document No. CNC-1211.00-06-0003), Appendix D, "Calculation" [21].

6.8 Tabulated ESEL HCLPF Values (Including Key Failure Modes)

Tabulated ESEL HCLPF values are provided in Appendix A for Unit 1 and Appendix B for Unit 2. The following notes apply to the information in the tables:

For items screened out using the IPEEE evaluations, the HCLPF value is provided as >RLGM and the failure mode is listed as "Screened per IPEEE."

For items screened out using EPRI NP-6041-SL [7] screening tables, the screening levels are provided as >RLGM and the failure mode is listed as "Screened per EPRI NP-6041."

For items where interaction with masonry walls controls the HCLPF value, the HCLPF value is listed in the table and the failure mode is noted as "Interaction - Block Walls."

For items where anchorage controls the HCLPF value, the HCLPF value is provided as >RLGM and the failure mode is noted as "Anchorage."

For items where component function controls the HCLPF value, the HCLPF value is listed in the table and the failure mode is noted as "Functional Failure."

For items where relay function controls the HCLPF value, the HCLPF value is listed in the table and the failure mode is noted as "Relay Chatter."

Page 24 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 7.0 Inaccessible Items 7.1 Identification of ESEL Items Inaccessible for Walkdowns All ESEL items in Unit I were accessible for walkdowns except for hydrogen igniters 1EHM0035, 1EHM0039, and IEHM0045 located in the Unit 1 Reactor Building. These igniters were judged seismically adequate based on similarity to other igniters that were accessible and included in the walkdowns.

There are a total of 38 Unit 2 ESEL items that remain to be walked down as detailed in Section 7.2. All ESEL items in Unit 2, other than items discussed in Section 7.2, were accessible for walkdowns.

7.2 Planned Walkdown / Evaluation Schedule / Close Out Unit 2 components listed in Table 7-1 remain to be evaluated by walk-by or walkdown. These components are scheduled to be reviewed in March 2015.

Table 7-1. Unit 2 Walkdowns & Walk-Bys Not Completed at Time of this Report. (2 sheets)

ESEL EIN Description Bldg.

El.

Location Walkdown ID or Walk-By 7

2NCTK11 Pressurizer Relief Tank 2 CV2 554 90 Deg Walkdown 8

2NV1A NC Letdown to Regenerative Heat CV2 554 228 Deg Walkdown Exchanger Isolation Valve 38 Rad 9

2ARFD2 Air Return Fan Damper 2 CV2 595 265 Deg Walk-By 53 Rad 10 2ARFD4 Air Return Fan Damper 4 CV2 595 292 Deg Walk-By 49 Rad 11 2AVXCARF 2A Containment Air Return Fan (CARF-2A)

CV2 593 265 Deg Walk-By 12 2BVXCARF 2B Containment Air Return Fan (CARF-2B)

CV2 593 293 Deg Walk-By 13 2VXHSFA H2 Skimmer Fan A CV2 645 265 Deg Walkdown 14 2VXHSFB H2 Skimmer Fan B CV2 645 220 Deg Walkdown 15 2VX1A H2 Skimmer Fan A Damper CV2 659 263 Deg Walkdown 16 2VX2B H2 Skimmer Fan B Damper CV2 659 283 Deg Walkdown 17 2ND1B Residual Heat Removal A Train Isolation CV2 572 170 Deg Walk-By Valve 33 Rad 18 2ND2A Residual Heat Removal A Train Isolation CV2 567 176 Deg Walk-By Valve 50 Rad 30 2AVVLCVU 2A Lower Containment Ventilation Unit CV2 565 16 Deg Walkdown (LCVU-2A) 31 2DVVLCVU 2D Lower Containment Ventilation Unit CV2 565 344 Deg Walkdown (LCVU-2D) 145 2EHM0041 Group 5A Igniter Box CV2 601 216 Deg Walkdown 21 Rad 147 2EHM0045 Group 5A Igniter Box CV2 642 206 Deg Walkdown 41 Rad Page 25 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 Table 7-1. Unit 2 Walkdowns & Walk-Bys Not Completed at Time of this Report. (2 sheets)

ESEL Walkdown EIN Description Bldg.

El.

Location ID or Walk-By 212 2CFLT5610 2A Steam Generator Wide Range Level AN2 566 358 Deg Walkdown Channel #1 59 Rad 213 2CFLT5620 2B Steam Generator Wide Range Level AN2 565 149 Deg Walkdown Channel #2 59 Rad 214 2CFLT5630 2C Steam Generator Wide Range Level AN2 565 205 Deg Walkdown Channel #3 59 Rad 215 2CFLT5640 2D Steam Generator Wide Range Level AN2 567 327 Deg Walkdown Channel #4 59 Rad 218 2CFLT5490 2A Steam Generator Narrow Range Level AN2 575 002 Deg Walkdown Channel #4 59 Rad 219 2CFLT5520 2B Steam Generator Narrow Range Level AN2 565 130 Deg Walkdown Channel #4 59 Rad 220 2CFLT5550 2C Steam Generator Narrow Range Level AN2 575 205 Deg Walkdown Channel #4 59 Rad 221 2CFLT5580 2D Steam Generator Narrow Range Level AN2 568 315 Deg Walkdown Channel #4 59 Rad 222 2NCLT5171 Pressurizer Level-Low Temperature AN2 571 113 Deg Walkdown 56 Rad 230 2NCRD5850 2A NC Loop Hot Leg Wide Range CV2 567 20 Deg Walkdown Temperature 20 Rad 231 2NCRD5860 2A NC Loop Cold Leg Wide Range CV2 567 51 Deg Walkdown Temperature 28 Rad 232 2NCRD5870 2B NC Loop Hot Leg Wide Range CV2 567 160 Deg Walkdown Temperature 18 Rad 233 2NCRD5880 2B NC Loop Cold Leg Wide Range CV2 567 124 Deg Walkdown Temperature 28 Rad 234 2NCRD5900 2C NC Loop Hot Leg Wide Range CV2 567 204 Deg Walkdown Temperature 20 Rad 235 2NCRD5910 2C NC Loop Cold Leg Wide Range CV2 567 240 Deg Walkdown Temperature 29 Rad 236 2NCRD5920 2D NC Loop Hot Leg Wide Range CV2 567 340 Deg Walkdown Temperature 20 Rad 237 2NCRD5930 2D NC Loop Cold Leg Wide Range CV2 567 309 Deg Walkdown Temperature 28 Rad 240 2NILT5260 Containment Sump Level RX2 552 21 Deg Walkdown 50 Rad 241 2NILT5261 Containment Sump Level RX2 556 2 Deg Walkdown 45 Rad 242 2NILT5262 Containment Sump Level RX2 560 3 Deg Walkdown 45 Rad 243 2NILT5263 Containment Sump Level RX2 566 2 Deg Walkdown 54 Rad 244 2NILT5264 Containment Sump Level RX2 570 18 Deg Walkdown 56 Rad Page 26 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 8.0 ESEP Conclusions and Results 8.1 Supporting Information CNS has performed the ESEP as an interim action in response to the NRC's 50.54(f) letter [1]. It was performed using the methodologies in the NRC-endorsed guidance in EPRI 3002000704 [2].

The ESEP provides an important demonstration of seismic margin and expedites plant safety enhancements through evaluations and potential near-term modifications of plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events.

The ESEP is part of the overall CNS response to the NRC's 50.54(f) letter [1]. On March 12, 2014, the Nuclear Energy Institute (NEI) submitted to the NRC results of a study [12] of seismic core damage risk estimates based on updated seismic hazard information as it applies to operating nuclear reactors in the Central and Eastern United States (CEUS). The study concluded that "... site-specific seismic hazards show that there clearly has not been an overall increase in seismic risk for the fleet of U.S. plants." based on the re-evaluated seismic hazards. As such, the "... current seismic design of operating reactors continues to provide a safety margin to withstand potential earthquakes exceeding the seismic design basis... "

The NRC's May 9, 2014, NTTF 2.1 Screening and Prioritization letter [14]

concluded that the "... fleetwide seismic risk estimates are consistent with the approach and results used in the GI-199 safety/risk assessment." The letter also stated that "As a result, the staff has confirmed that the conclusions reached in GI-199 safety/risk assessment remain valid and that the plants can continue to operate while additional evaluations are conducted."

An assessment of the change in seismic risk for CNS was included in the fleet risk evaluation submitted in the March 12, 2014, NEI letter [12]; therefore, the conclusions in the NRC's May 9 letter [14] also apply to CNS.

In addition, the March 12, 2014, NEI letter [12] provided an attached "Perspectives on the Seismic Capacity of Operating Plants," which (1) assessed a number of qualitative reasons why the design of structures, systems and components (SSCs) inherently contain margin beyond their design level; (2) discussed industrial seismic experience databases of performance of industry facility components similar to nuclear SSCs; and (3) discussed earthquake experience at operating plants.

The fleet of currently operating nuclear power plants was designed using conservative practices, such that the plants have significant margin to withstand large ground motions safely. This has been borne out for those plants that have actually experienced significant earthquakes. The seismic design process has inherent (and intentional) conservatisms which result in significant seismic Page 27 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. I margins within SSCs. These conservatisms are reflected in several key aspects of the seismic design process, including:

Safety factors applied in design calculations; Damping values used in dynamic analysis of SSCs; Bounding synthetic time histories for ISRS calculations; Broadening criteria for ISRS;

" Response spectra enveloping criteria typically used in SSC analysis and testing applications; Response spectra based frequency domain analysis rather than explicit time history based time domain analysis; Bounding requirements in codes and standards; Use of minimum strength requirements of structural components (concrete and steel);

Bounding testing requirements; and Ductile behavior of the primary materials (that is, not crediting the additional capacity of materials such as steel and reinforced concrete beyond the essentially elastic range, etc.).

These design practices combine to result in margins such that the SSCs will continue to fulfill their functions at ground motions well above the SSE.

8.2 Identification of Planned Modifications Tables 6-2 and 6-3 identify items where modifications will be made to enhance the seismic capacity of the plant 8.3 Modification Implementation Schedule Plant modifications will be performed in accordance with the schedule identified in NEI letter dated April 9, 2013 [13] (endorsed by the NRC in their May 7, 2013, letter [16]), which states that plant modifications not requiring a planned refueling outage will be completed by December 2016 and modifications requiring a refueling outage will be completed within two planned refueling outages after December 31, 2014.

Page 28 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 8.4 Summary of Planned Actions The actions Listed in Table 8-1 will be performed as a result of the ESEP.

Table 8-1. Summary of Planned Actions.

Action Equipment Equipment ID Description Action Description Completion Date I

N/A N/A Perform seismic walkdowns, generate No later than the end of the HCLPF calculations and design and second planned Unit 2 implement any necessary modifications refueling outage after for Unit 2 items as detailed in Section 7.2.

December 31, 2014.

2 N/A N/A Complete evaluations/ modifications of No later than the end of the Unit I components listed in Table 6-2.

second planned Unit 1 refueling outage after December 31, 20:14.

3 N/A N/A Complete evaluations/ modifications of No later than the end of the Unit 2 components listed in Table 6-3.

second planned Unit 2 refueling outage after December 31, 2014.

4 N/A N/A Submit a letter to NRC summarizing Within 60 days following results of item 1 and confirming completion of ESEP activities, implementation of the plant modifications including items 1 through 3.

associated with items I through 3.

9.0 References

1)

Letter from E. Leeds and M. Johnson, NRC to All Power Reactor Licensees, et al.,

"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-lchi Accident," March 12, 2012.

2)

Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1 -Seismic, Electric Power Research Institute, Palo Alto, CA: May 2013, EPRI 3002000704.

3)

CNS Overall Integrated Plan, Letter from Ben Waldrop to U.S. Nuclear Regulatory Commission, "Duke Energy Carolinas, LLC (Duke Energy); Catawba Nuclear Station (CNS), Units 1 and 2, Docket Nos. 50-413 and 50-414, Renewed License Nos. NPF-35 and NPF-52; Catawba Nuclear Station Overall Integrated Plan in Response to March 12, 2012, Commission Order to Modify Licenses With Regard To Requirements for Mitigation Strategies for Beyond Design Basis External Events (Order EA-12-049)," dated February 28, 2013, Duke Energy, York, SC.

Page 29 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1

4)

Letter from Kelvin Henderson to U.S. Nuclear Regulatory Commission, "Duke Energy Carolinas, LLC (Duke Energy); Catawba Nuclear Station (CNS), Units 1 and 2, Docket Nos. 50-413 and 50-414, Renewed License Nos. NPF-35 and NPF-52; Seismic Hazard and Screening Report (CEUS Sites), Response to NRC 10 CFR 50.54(f) Request for Additional 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 31, 2014, Duke Energy, York, SC.

5)

Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, June 1991, Nuclear Regulatory Commission, NUREG-1407.

6)

USNRC Generic Letter 88-20, Supplement 4, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities-10 CFR 50.54(f),"

June 28, 1991, U.S. Nuclear Regulatory Commission, Washington, D.C.

7)

A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Rev. 1, August 1991, Electric Power Research Institute, Palo Alto, CA, EPRI NP-6041-SL.

8)

Methodology for Developing Seismic Fragilities, Electric Power Research Institute, Palo Alto, CA, July 1, 1994, EPRI TR-103959.

9)

Catawba Nuclear Station IPEEE Submittal Report, July 1994, Duke Energy, York, SC.

10)

Catawba Nuclear Station Unit 1, Probabilistic Risk Assessment, September 1992, Duke Energy, York, SC.

11)

Development of Criteria for Seismic Review of Selected Nuclear Power Plants, published May 1978, Nuclear Regulatory Commission, NUREG/CR-0098.

12)

Letter from A. Pietrangelo, NEI to D. Skeen, USNRC, "Seismic Core Damage Risk Estimates Using the Updated Seismic Hazards for the Operating Nuclear Plants in the Central and Eastern United States," March 12, 2014.

13)

Letter from A. Pietrangelo, NEI to D. Skeen, USNRC, "Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations," April 9, 2013.

14)

Letter from E. Leeds, NRC to All Power Reactor Licensees, et al., "Screening and Prioritization Results Regarding Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(F) Regarding Seismic Hazard Re-Evaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights From the Fukushima Dai-Ichi Accident," May 9, 2014.

15)

Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, Electric Power Research Institute, Palo Alto, CA, February 2013, EPRI 1025287.

Page 30 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1

16)

Letter from E. Leeds, NRC to J. Pollock, NEI, "Electric Power Research Institute Final Draft Report XXXXXX, "Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1:

Seismic," as an Acceptable Alternative to the March 12, 2012, Information Request for Seismic Reevaluations," May 7, 2013.

17)

Augmented Approach for Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic: Seismic - Catawba Nuclear Station Expedited Seismic Equipment List, Rev. 0, ARES Corporation Report No. 030321.13.01-005, Duke Energy Document No. CNC-1211.00-06-0004.

18)

Seismic Capacity Evaluations for the IPEEE and EPRI Seismic Margins Study, Revision 3, Duke Energy Document No. CNC-1535.00-00-0005.

19)

Seismic Margin Assessment of the Catawba Nuclear Station, Electric Power Research Institute, Palo Alto, CA, April 1989, EPRI NP-6359.

20)

Seismic Fragility Applications Guide Update, December 2009, Electric Power Research Institute, Palo Alto, CA, EPRI 1019200.

21)

Expedited Seismic Evaluation Process for Implementation of Seismic Risk Evaluations at Catawba Nuclear Station, Revision 1, ARES Corporation Report 030321.13.01-003, Duke Energy Document No. CNC-1211.00-06-0003.

22)

First Update to CNS Overall Integrated Plan, Letter from Kelvin Henderson to U.S. Nuclear Regulatory Commission, "Duke Energy Carolinas, LLC (Duke Energy);

Catawba Nuclear Station (CNS), Unit Nos. 1 and 2, Docket Nos. 50-413 and 50-414, Renewed License Nos. NPF-35 and NPF-52; First Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses With Regard to Requirements for Mitigation Strategies for Beyond-Design-basis External Events (Order Number EA-12-049)," dated August 28, 2013, Duke Energy, York, SC.

23)

Second Update to CNS Overall Integrated Plan, Letter from Kelvin Henderson to U.S. Nuclear Regulatory Commission, "Duke Energy Carolinas, LLC (Duke Energy);

Catawba Nuclear Station (CNS), Unit Nos. 1 and 2, Docket Nos. 50-413 and 50-414, Renewed License Nos. NPF-35 and NPF-52; "Second Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses With Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)," dated February 28, 2014, CNS-14-020, Duke Energy, York, SC.

Page 31 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1

24)

Third Update to CNS Overall Integrated Plan, Letter from Kelvin Henderson to U.S. Nuclear Regulatory Commission, "Duke Energy Carolinas, LLC (Duke Energy);

Catawba Nuclear Station (CNS), Unit Nos. 1 and 2, Docket Nos. 50-413 and 50-414, Renewed License Nos. NPF-35 and NPF-52; Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses With Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)," dated August 28, 2014, Duke Energy, York, SC.

Page 32 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. I Appendix A Catawba Nuclear Station Unit 1 ESEL and HCLPF Results Page 33 of 61

Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Catawba Nuclear Station Unit I ESEL and HCLPF Results Rev. 1 ESEL ID EIN Descriptlon 1

SNC250A Reactor Head Vent Valve 2

SNC253A Reactor Head Vent Valve 3

1NCI Pressurizer Safety Valve 4

iNC2 Pressurizer Safety Valve 5

iNC3 Pressurizer Safety Valve 6

INC34A Pressurizer Power Operated Relief Valve 7

SNCTK11 Pressurizer Relief Tank 1 8

SNVfA NC Letdown to Regenerative Heat Exchanger Isolation Valve 9

1ARFD2 Containment Air Return Fan 1A Damper 10 1AVXCARF Containment Air Return Fan 1A 11 1ARFD4 Containment Air Return Fan 1B Damper 12 SBVXCARF Containment Air Return Fan 1B 13 1VXiA Hydrogen Skimmer Fan IA Isolation 14 iVXHSFA Hydmgen Skimmer Fan 1A 15 1VX2B Hydrogen Skimmer Fan 1B Isolation Valve 16 1VXHSFB Hydmgen Skimmer Fan 1B 17 iNDIB Residual Heat Removal A Train Isolation Valve 18 1ND2A Residual Heat Removal A Train Isolation Valve 19 SNDPUA Residual Heat Removal Pump 1A 20 1NDHXAPMP Residual Heat Removal Pump Mechanical Seal Heat Exchanger 1A 21 1KCHX0040 Residual Heat Removal Pump 1A Motor Cooler 22 1NDHXA Residual Heat Removal Heat Exchanger IA 23 1KFI01B FW System/ KF System Isolation Valve 24 1FWTKOf Retueling Water Storage Tank (FWST) 25 1KCHXA Component Cooling Heat Exchanger 1A 26 1KCPUA1 Component Cooling Pump IA1 27 SKCPHXA1 Component Cooling Pump IA1 Motor Cooler 28 1KC6OSA Residual Heat Removal Heat Exchanger IA Isolation Valve 29 IRN63A Nuclear Service Water to Standby Nuclear Service Water Pond Discharge Isolation Valve 30 IRN250A Nuclear Service Water to 1CAPUTD Isolation Valve 31 IAVVLCVU 1A Lower Containment Ventilation Unit (LCVU-IA)

Bldg EL Location CV1 fof 38 Deg 25 Rad CV1 600 40 Deg 27 Fad CV1 637 102 Deg CV1 637 102 Deg CV1 637 102 Deg CV1 665 105 Deg CV1 554 90 Deg CV1 554 229 Deg38 Rad CV1 599 265 Deg 53 fad CV1 599 265 Deg 53 Rad CV1 599 293 Deg 49 Rad CV1 599 293 Deg 49 Rad cv1 658 260 Deg 45 Rad CV1 652 266 Deo 54 Rad CV1 658 285 Deg 45 Had CV1 652 272 Oeg 49 Rad CV1 568 176 Deg 25 Rad CV1 567 176 Deg 50 Rad AUX 522 GG-54 AUX 522 FF-53 AUX 522 GG-53 AUX 560 LL-51 AUX 5f4 J1-52 Yard AUX 577 HH-56 AUX 560 HH-59 AUX 560 GG-58 AUS 590 EK-50/51 AUX 581 QQ-60 Normal Operating State Closed Closed Closed Closed Closed Closed Functional Open Closed Off Closed Off Closed Off Closed Off Closed Closed Functional Functional Functional Functional Closed Intact Functional Functional Functional Closed Closed Desired Operating State Open Open Functional Functional Functional Functional Functional Closed Open On Open On Open On Open On Open Open Functional Functional Functional Functional Open Intact Functional Functional Functional Open Open Walbdown or Walk-by Screening Notes Walkdown Walkdown Walk-by Walk-by Walk-by Walk-by Walkdown Walkdown Walk-by Included in IPEEE, pg 35.

Walk-by Walk-by Included in IPEEE, pg 35.

Walk-by Walkdown Walkdown Walkdown Walkdown Walk-by Walk-by Walk-by Walk-by Rule-of-the-box with 1NDPUA Walk-by Rule-of-the-box with 1NDPUA Walk-by Walkdown Walkdown Walk-by Walk-by Walk-by Rule-of-the-box with 1KCPUA1 Walk-by Walk-by HCLPEP

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>RLGM Screened per IPEEE RXi 565 16 Deg Functional Functional Walkdown

>RLGM Screened per EPRI NP-6041 Page 34 of 61

Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Catawba Nuclear Station Unit 1 ESEL and HCLPF Results Rev. 1 ESEL ID EIN Description 32 IDVVLCVU 1D Lower Containment Ventilation Unit (LCVU-1D) 33 1BB24 Steam Generator IC Blowdown Flow Control Valve 34 1BB65 Steam Generator 1D Slowdown Flow Control Valve 35 1BB69 Steam Generator 1A Blowdown Flow Control Valve 36 IB573 Steam Generator 55 Blowdown Flow Control Valve 37 ICA174 Condenser Circulating Water System Isolation Valve 38 1CA257 SCAPUTD Feedwater Safety Valve 39 1CAIS1A Nuclear Service Water System Train 1A Isolation Valve 40 1CAPUTD Turbine Driven Auxiliary Feed Water Pump 41 SCAHXSO Turbine Driven Auxiliary Feed Water Pump Lube Oil Cooler 42 1CA64 Auxiliary Feedwater to Steam Generator 1A Isolation Valve 43 1CA52 Auxiliary Feedwater to Steam Generator IB Isolation Valve 44 1CA48 Auxiliary Feedwater to Steam Generator iC Isolation Valve 45 1CA36 Auxiliary Feedwater to Steam Generator 1D Isolation Valve 46 S5V2 Steam Generator 1D Safety Valve 47 15V3 Steam Generator 1D Safety Valve 48 ISV4 Steam Generator iS Safety Valve 49 51V5 Steam Generator 10 Safety Valve so S5V6 Steam Generator SD Safety Valve 51 1SMl Steam Generator 1D Main Steam Isolation Valve 52 1SVs Steam Generator iC Safety Valve 53 55V9 Steam Generator SC Safety Valve 54 1sVi0 Steam Generator SC Safety Valve 55 15V1l Steam Generator SC Safety Valve 56 15V12 Steam Generator 1C Safety Valve 57 15M3 Steam Generator IC Main Steam Isolation Valve 58 SV14 Steam Generator I8 Safety Valve 59 SSV15 Steam Generator 15 Safety Valve 60 1SV16 Steam Generator 55 Safety Valve 61 15V17 Steam Generator 1B Safety Valve 62 SSVS8 Steam Generator 55 Safety Valve Bldg EL Location 0X1 565 344 Deg TBl S78 IL-29 TBS 583 1L-29 TB1 583 1L-29 TB1 583 1L-29 AUX 544 CC-53 AUX 534 AA-S1 AVG 555 88.51 AUX 531 AA-5S1 AUX 551 DD-53 AUX 556 BB-50 AUX 546 CC-53 AUX 551 DD-53 AUg 554 BB-49 DH1 618 EE-44 DHS 618 EFA4 DHS 618 EE-44 DHI 618 EE-44 DHI 518 EE-44 DH1 615 DD-44 DH1 618 EE 52 DH1 618 EE-52 DHS 618 EE-52 DH1 515 EE-52 DH1 618 EE-52 0H1 615 DD-52 DHS 618 EE-53 DH1 618 EE-S3 DHS 618 EE-53 DH1 618 EE-53 DH1 618 EE-53 Normal Operating State Functional Open Open Open Open Closed Closed Closed Functional Functional Open Open Open Open Closed Closed Closed Closed Closed Open Closed Closed Closed Closed Closed Open Closed Closed Closed Closed Closed Desired Operating State Functional Closed Closed Closed Closed Open Functional Open Functional Functional Throttled Throttled Throttled Throttled Functional Functional Functional Functional Functional Closed Functional Functional Functional Functional Functional Closed Functional Functional Functional Functional Functional Walkdown or Walk-by Screening Notes Walkdown Walkdown Walkdown Walkdown Walkdown Walk-by Walkdown Walk-by Walk-by Walk-by Rule-of-the-box w Walkdown Walkdown Walkdown Walkdown Walk-by Walk-by Walk-by Watk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by ith 1CAPUD HCLPFp vRLGM

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Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Catawba Nuclear Station Unit 1 ESEL and HCLPF Results ESEL IS EIN Description Bldg 63 ISMS Steam Generator 18 Main Steam Isolation Valve DH1 64 1SV20 Steam Generator 1A Safety Valve DH1 65 1SV21 Steam Generator 1A Safety Valve DAS 66 15V22 Steam Generator 1A Safety Valve GHA 67 13V23 Steam Generator 16 Safety Valve DHI 68 I5V24 Steam Generator 1A Safety Valve DHG 69 lSM7 Steam Generator 1A Main Steam Isolation Valve DHG 70 SSVi Power Operated Relief Valve -Steam Generator 1D DHA 71 iSV7 Power Operated Relief Valve -Steam Generator IC DHG 72 ISVS3 Power Operated Relief Valve -Steam Generator 16 DHE 73 1SV19 Power Operated Relief Valve -Steam Generator 1A DHG 74 1SA2 1CAPUTD Steam Feed Isolation Valve DH1 75 1SA145 1CAPUTD Trip Throttle Valve (Stop Valve)

AUX 76 1SA144 1CAPUTD Control Valve AUX 77 INDHXB Residual Heat Removal Heat Exchanger lB AUX 78 N1IS4A Accumulator Tank 1A Isolation Valve CVG 79 1N165B Accumulator Tank 1B Isolation Valve CV1 8&

1N176A Accumulator Tank 1A Isolation Valve CVt 81 1N1888 Accumulator Tank 1A Isolation Valve CVE 82 INSHXA Containment Spray Heat Exchanger 1A AUX 83 1NSHXB Containment Spray Heat Exchanger SB AUX 84 1HLPSP Hot Leg Particulate Sample Panel AUX 85 1NMHX07 Reactor Coolant Hot Leg Sample Heat Exchanger GA AUX 86 1NMHX08 Reactor Coolant Hot Leg Sample Heat Exchanger 1B AUX 87 lPSS Primary Sample Sink lB AUX 88 IEMF46A A Train KC Radiation Monitor AUX 89 1KC2B Auxiliary Building Non-Essential Return Header Isolation Valve AUX 90 SKC3A Reactor Building Non-Essential Return Header Isolation Valve AUX 91 IKCS38 Auxiliary Building Non-Essential Return Header Isolation Valve AUX 92 IKC230A Reactor Building Noe-Essential Return Header Isolation Valve AUX 93 1KCTKA Component Cooling Surge Tank 1A AUX EL Location 615 DD-54 618 EE-44 618 EE-44 618 EE-A4 610 EEAR 618 EE-44 615 DD-43 601 FF-44 601 FF-53 601 FF-53 601 FF-44 628 FPF-53 534 AA-S5 531 AA-51 560 KK-51 560 46 Deg 47 Rad 560 137 Deg 47 Rad 560 226 Deg 47 Rad 560 312 Deg 47 Rad 577 LL-52 577 U.-52 543 FF-54 543 Rm 238 543 Rm 238 543 FF-54 577 HH-57 567 HH-S7 567 HH-57 585 JJ-55 586 HH-SS 594 NN-59 Normal Operating State Open Closed Closed losed Closed Closed Open Closed Closed Closed Closed Closed Open Open Intact Open Open Open Open Intact Intact Intact Intact Intact Intact Intact Open Open Open Open Intact Desired Operating State Closed Functional Functional Functional Functional Functional Closed Open Open Open Open Open Throttled Throttled Intact Closed Closed Closed Closed Intact Intact Intact Intact Intact Intact Intact Closed Closed Closed Closed Intact Rev. 1 Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walkdown Walkd.wn Walkdown Walkdown Walkdown Walk-by Walk-by Walk-by Walk-by Walk-by Walkdown or Walk-by Screenfing Notes Walk-by Rule-of-the-box with 1CAPUD Rule-of-the-box with 1CApUD HCLPF'

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Screened per IPEEE Screened per IPEEE Rev. 1 Page 36 of 61

Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Catawba Nuclear Station Unit I ESEL and HCLPF Results Rev. 1 ESEL ID EIN Description 94 1KFHXA Fuel Pool Cooling Heat Exchanger 1A 95 1KFHXB Fuel Pool Cooling Heat Exchanger 15 96 INVHX0O09 Let Down Heat Exchanger 97 0RN11A Nuclear Service Water Pump 1A Isolation Valve 98 2RNI1A Nuclear Service Water Pump 2A Isolation Valve 99 1RN48B Nuclear Service Water Supply Crossover Isolation Valve 100 2RN48B Nuclear Service Water Supply Crossover Isolation Valve 101 IRNSTA Nuclear Service Water Strainer 1B 102 2RNSTA Nuclear Service Water Strainer 28 103 1RN5LA Unit 1 Nuclear Service Water Non-Essential Return Header Isolation Valve 104 2RNSIA Unit 2 Nuclear Service Water Non-Essential Return Header Isolation Valve 105 1RN535 Nuclear Service Water Crossover Isolation Valve 106 1RN57A Diesel Generator Cooling Water Isolation Valve 107 1RN58B Unit I Nuclear Service Water Header B Return to SNSWP Isolation Valve 108 1RN232A 1A Diesel Generator Cooling Water Isolation Valve 109 1EMXA Essential Motor Control Center, 600 VAC 110 1EMXS Essential Motor Control Center, 600 VAC 111 1EMXC Essential Motor Control Center, 600 VAC 112 tEMXD Essential Motor Control Center, 600 VAC 113 1EMXE Essential Motor Control Center, 600 VAC 114 1EMXG Essential Motor Control Center, 600 VAC 115 1EMXI Motor Control Center, 600 VAC, single phase, normal power source for Hydrogen Igniter Group A 116 1EM)U Essential Motor Control Center, 6) VAC 117 1EMXK Essential Motor Control Center, 600 VAC 118 UEMIL Essential Motor Control Center, 600 VAC 119 1EMXM Essential Motor Control Center, 60U VAC 120 lEMXN Essential Motor Control Center, 600 VAC 121 1EMXS Motor Control Center, 4W0 VAC, single phase, Emergency power source for Hydrogen Igniter Group A 122 1ETA Essential Switchgear, 4160 VAC 123 SMKR Normal Motor Control Center, 600 VAC Bldg AUg AUX AUX RNB RNE AUX AUX RNB RNB AUX AUX AUG AUX AUg DIA AUX AUG AUX AUX DItA AUX AUX Location NN-S2 NNH52 KK-S3 RmA 52X71Y PP-S3 PP-60 Pumphouse Pumphause NN-55 MM-58 LL-56 PP-53 PP-60 EE-38 FF-54 FF-56 OB-50 BB-50 CC-39 FF-56 EE-54 Normal Operating State Intact Intact Intact Open Open Open Open Intact Intact Open Open Open Open Closed Closed Functional Functional Functional Functional Functional Functional Functional Desired Operating State Intact Intact Intact Closed Closed Closed Closed Intact Intact Closed Closed Closed Closed Closed Closed Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Walkdown or Walk-by Screening Notes Walkdowrn Walkdown Walkdown Walk-by Walk-by Walk-b Walk-by Included in IPEEE, Walk-by Walk-by Walk-by Walk-by Included in IPEEl, Walk-b, Walk-by Included in IPEEE, Walk-by Walk-by Included in IPEEE, Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by pg 720.

pg 720.

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pg 24.

AUX AUg AUX AUX AUX AUX 560 577 560 577 560 577 GG-56 BB-47 B5-47 CC-53 CC-53 BR-48 Functional Functional Functional Functional Functional Functional Walk-by Walk-by Walk-by Walkdown Walkdawn Walk-by

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>RLGM Screened per IPEEE Page 37 of 61

Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Rev. 1 Catawba Nuclear Station Unit I ESEL and HCLPF Results ESEL Normal Desired Walkdown ID EIN Description Bldg EL Location Operating State Operating State orWalk-by Screening Notes HCLPF-Key Failure Mode' 124 1MXQ Blackout Motor Control Center, 600 VAC AUX 577 BB-49 Functional Functional Walkdown

>RLGM Interaction - Block Walls 128 1MTSWO0G0 Transfer Switch (Single Pole Double Throw), Hydrogen Igniter Group A AUX 577 BE.48 Functional Functional Walkdown

>RLGM Interaction - Block Walls 129 IXFMROO3 Transformer, 480 VAC - 120 VAC, Hydrogen Igniter Group A AUX 577 BB-48 Functional Functional Walkdown

>RLGM Interaction - Block Walls 130 IVREGO013 Voltage Regulator, Hydrogen Igniter Group A AUX 577 BB-48 Functional Functional Walkdown

>RLGM Interaction - Block Walls 131 SEATC8 Essential Area Terminal Cabinet AUX 577 EE-54 Functional Functional Walkdown 0.32 Anchorage 132 IELCPO278 Hydrogen Igniter Group A Control Panel AUX 577 BB-48 De-Energized Energized Walkdown

>RLGM Interaction - Block Walls 133 SDGLSA Diesel Generator Load Sequencing Panel AUX 556 BB-43 Functional Functional Walk-by

>RLGM Screened per IPEEE 134 IELCP0357 SWO1, Hydrogen Igniter On/Off switch (Alt. Source) located on panel 1ELCP0357 AUX 577 BB-48 Functional Functional Walkdown

>RLGM Interaction - Block Walls 135 1MC7 136 BEHMOO03 137 1EHMO005 138 1EHMO007 139 EEHMO009 140 1EHM0071 141 1EHMO0i1 142 1EHM0013 143 1EHMOO15 144 1EHMDO7 145 IEHM0019 146 1EHM0021 147 1EHM0023 148 1EHM0025 149 1EHMO027 150 1EHM0029 151 1EHMO031 152 1EHM0033 Main Control Room Panel Group 2A Igniter Box Group 2A Igniter Box Group 2A Igniter Box Group 2A Igniter Box Group 2A Igniter Box Group 3A Igniter Box Group 3A Igniter Box Group 3A Igniter Box Group 3A Igniter Box Group 3A Igniter Box Group 3A Igniter Box Group 4A Igniter Box Group 4A Igniter Box Group 4A Igniter Box Group 4A Igniter Box Group 4A Igniter Box Group 4A Igniter Box AUX CV1 CVI CV1 CV1 CvS CVE CV1 CV1 CV1 CV1 CV1 CVO CVoi CVE CV1 CV1 CV1 594 BB-56 562 88 Deg 48 Rad 562 178 Deg 50 RS d 562 277 Deg 46 Rad 562 5 Deg 46 Bad 555 103 Deg 35 RBd 601 324 Deg 20 Rad 590 325 Deg 48 Rad 642 335 Deg 41 Rad 601 55 Deg 18 Rod 590 2 Deg 51 Rad 642 18 Deg 41 Rad 590 53 Deg 49 Bad 590 217 Deg 51 Rad 590 245 Deo 51 Rad 590 91 De a

51 Red 603 10 Deg 12 Rad 641 113 Deg 32 Rad Functional De-Energized De-Energized De-Energized De-Energized De-Energized De-Energized De-Energized De-Energized De-Energized De-Energized De-Energized De-Energized De-Enertized De-Energized De-Energized De-Energized De-Energized Functional Energized Energized Energized Energized Energized Energized Energized Energized Energized Energized Energized Energized Energized Energized Energized Energized Energized WaMldown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown 0.30 0.33 0.33 0.33 0.33 0.33 0.29 0.33 0.29 0.29 0.33 0.29 0.33 0.33 0.33 0.33 0.29 0.29 Interaction - Control Room Ceiling Anchorage Anchorage Anchorage Anchorage Anchorage Functional Failure Anchorage Functional Failure Functional Failure Anchorage Functional Failure Anchorage Anchorage Anchorage Anchorage Functional Failure Functional Failure 153 1EHMO35 Group SA Igniter Box CV1 601 130 Deg 30 Rad De-Energized Energized Walkdown 0.29 Functional Failure 154 1EHM0037 Group 5A Igniter Box C%1 930 14f Deg 50 Rod

-n-Fnergized Energized Walkdown 0.33 Anchorage 155 1EHM0039 Group SA Igniter Box CV1 642 161 Deo 41 Rad De-Energized Energized Walkdown 0.29 Functional Failure 156 1EHM0B41 Group 5A Igniter Box CV1 601 216 Deg 21 Bad Do-Energized Energized Walkdown 0.29 Functional Failure 157 1EHMOO43 Group 5A Igniter Box CV1 590 172 Deg 51 Rad De-Energized Energized Walkdown 0.33 Anchorage Page 38 of 61

Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Catawba Nuclear Station Unit 1 ESEL and HCLPF Results ESEL ID EIN Description Bldg 158 IEHMG045 Group 5A Igniter Box CV1 159 iEHMBW59 Group 7A Igniter Box CV1 160 IEHKMO61 Group 7A Igniter Box CV1 161 1EHMO063 Group 7A Igniter Box CV1 162 IEHMO06S Group 7A Igniter Box CVi 163 IEHMO0S3 Group 6A-1 Igniter Box CV1 164 lEHMOOSS Group 6A-1 Igniter Box CV1 165 1EHM0057 Group 6A-1 Igniter Box CVl 166 LEHMO47 Group 6A-2 Igniter Box CVi 167 1EHMO49 Group 6A-2 Igniter Box CV1 168 1EHM0051 Group 6A-2 Igniter Box CV1 169 1EHM0067 Group BA Igniter Box CV1 170 IEHMOO69 Group BA Igniter Box CV1 171 iAFWPTCP Auxiliary Feedwater Pump Turbine Control Panel (ELCP0245)

AUX 172 1EDE Power Supply for Solenoid valve 1SASVO020 and other vales, 125 VDC AUX Distribution Center, 125 VDC Distribution Center, compartments FOlA, FOIG, F011 Rev. 1 EL Location 642 204 Deg 41 Rad 714 318 Deg 24 Rad 714 49 Deg 24 Rod 714 140 Deg 24 Rad 714 21i Deg 24 Rad 666 108 Deg 46 Rad 666 157 Deg 46 Rad 666 206 Deg 46 Rad 666 321 Deg 46 Rad 666 11 Deg 46 Red 666 59 Deg 46 Red 653 216 Deg 32 Rad 653 41 Deg 32 Rad 543 CC-53 577 BB-46 Normal Operating State De-Energized De-Energized De-Energiaed De-Energized De-Energiued De-Energiued De-Energized De-Energized De-Energized De-Energized De-Energized De-Energized De-Energized Functional Functional Desired Operating State Energized Energized Energized Energized Energized Energized Energized Energized Energized Energized Energized Energized Energized Functional Functional Walkdown or Walk-by Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walk-by Walk-by Screening Notes HC.PF 5

0.29 0.29 0.29 0.29 0.29 0.29 0.29 0.29 0.29 0.29 0.29 0.29 0.29

>RLGM

>RLGM Key Failure Mode-Functional Failure Functional Failure Functional Failure Functional Failure Functional Failure Functional Failure Functional Failure Functional Failure Functional Failure Functional Failure Functional Failure Functional Failure Functional Failure Screened per IPEEE Screened per IPEEE 173 174 175 176 177 178 179 180 181 182 183 104 185 186 187 MEATC23 lSSPSA IEATC21 iMC1 1MC10 1MC2 1IMC4 1MC5 1PCC1 1PCC2 1PCC3 1PCC4 IPCC5 1PCC6 1PCC7 Essential Area Terminal Cabinet Solid State Protection System Cabinet, Control Panel, Control Cabinet Essential Area Terminal Cabinet Main Control Room Panel Main Control Room Panel Main Control Room Panel Main Control Room Panel Main Control Room Panel PLC Cabinet PLC Cabinet PLC Cabinet AUX AUX AUX AUX AUX AUX AUX AUX AUX AUX AUX AUX AUX AUX AUX 577 BB-50 594 CC-55 577 BB-E3 594 BB-53 594 BB-56 594 AA-56 594 BB.56 594 Be-56 594 BB-55 594 BB-54 594 BB-55 594 BB-54 594 CC-55 594 CC-54 594 CC-S5 Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Walkdown Walkdown Walkdown Walkdown Walkdomn Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Waikdown Walkdown Walkdown Walkdown 0.37 0.30

>RLGM 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 u.30 0.30 0.30 Anchorage Interaction -Control Room Ceiling Interaction -Block Walls Interaction -Control Room Ceiling Interaction -Control Room Ceiling interaction -Control Room Ceiling Interaction -Control Room Ceiling Interaction -Control Room Ceiling Interaction -Control Room Ceiling Interaction -Control Room Ceiling Interaction -Control Room Ceiling Interaction -Control Room Ceiling

ntcractio -Centro! nomon Ceilina Interaction -Control Room Ceiling Interaction -Control Room Ceiling PLC Cabinet PLC Cabinet PLC Cabinet PLC Cabinet Page 39 of 61

Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Catawba Nuclear Station Unit I ESEL and HCiPF Results ESEL ID EIN Description Bldg 188 EPCC8 PLC Cabinet AUX 189 BRVUS Process Cabinet AUX 190 SMCS Main Control Room Panel AUX 191 1TBOXOOI8 Control Panel Yard 192 ITBOXO599 Wide Range Neutron Flux Signal Processor Terminal Box AUX 193 1TBOX0537 Wide Range Neutron Flux Amplifier Terminal Box AUX 194 1TBOX0596 Wide Range Neutron Flux Power Supply Terminal Box AUX 195 1TBOXO587 Wide Range Neutron Flux Power Supply Terminal Box AUX 196 1EPA 125 VDC Power Panelboard AUX 197 IEATC12 Essential Area Terminal Cabinet AUX 198 ITBOX0522 Control Panel DIA 199 1StutTCI Control Panel AUX 200 1EIA Vital 120 VAC Inverter AUX Rev. 1 EL Loc-ation 594 CC-S4 577 CC-53 594 BB-S6 598 SOX46Y 577 AA-S1 577 AA-Si 577 AA-51 577 AA-51 554 DD-SS 577 FF-S6 556 BB-45 577 FF-55 554 DD-55 Normal Operating State Functional Functional Functional Functional Functional Functional Functional Functional Energized Functional Functional Functional Energized Desired Operating State Functional Functional Functional Functional Functional Functional Functional Functional Energized Functional Functional Functional Energized Walkdown or Walk-by Screening Notes Walkdowtn Walkdown Wabkdown Walkdown Walkdown Walkdowo Walkdown Walkdown Walk-by Walkdown Walkdown Walkdown Walkdown SEWS included in IPEEE, pg 392. However, inverter replaced since original evaluation.

HCLPF' 0.30

>RLGM 0.30 0.29

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM 0.29

>RLGM

>RLGM

>RLGM Key Failure Modera Interaction - Control Room Ceiling Interaction - Block Walls Interaction - Control Room Ceiling Anchorage Screened per EPRI NP-6041 Screened per EPRI NP-6041 Screened per EPRI NP-6045 Screened per EPRI NP-6041 Screened per IPEEE Modification/Investigation Screened per EPRI NP-6041 Interaction - Block Wails Interaction - Block Walls 201 LEIB Vital 120 VAC Inverter AUX 554 CC-51 Energized Energized Walkdown SEWS included in IPEEE, pg 392. However, inverter

>RLGM Interaction - Block Walls replaced since original evaluation.

202 1EIC Vital 120 VAC Inverter AUX 554 CC-55 Energized Energized Walkdeown SEWS included in IREEE, pg 392. However, inverter

>RLGM Interaction - Block Walls replaced since original evaluation.

203 1EID Vital 120 VAC Inverter AUX 554 BB-55 Energized Energized Walkdown SEWS included in IPEEE, pg 392. However, inverter BRLGM Interaction - Block Walls replaced since original evaluation.

204 EERPD 120 VAC Power Panel AUX 554 BB.55 Energized Energized Walk-by

>RLGM Screened per IPEEE 205 1ERPB 120 VAC Power Panel AUX 554 CC-55 Energized Energized Walk-by

>RLGM Screened per IPEEE 206 SERPA 12OVAC Power Panel AUX 554 DD-55 Energized Energized Walk-by

>RLGM Screened per IPEEE 207 SERPC 120 VAC Power Panel AUX 554 CC-55 Energized Energized Walk-by

>RLGM Screened per IPEEE 208 IEATC13 Essential Area Terminal Cabinet AUX 560 11-S6 Functional Functional Walkdown 0.29 Relay Chatter 209 IEDF 125 VDC Panel, 125 VDC Distribution Center, compartments FOIA, FOll AUX 560 BB-46 Energized Energized Walk-by

>RLGM Screened per IPEEE 210 1SSPSB Control Cabinet AUX 594 DD-55 Functional Functional Walkdown 0.30 Interaction - Control Room Ceiling 211 1EADA Auctioneering Diode Assembly AUX 577 BR-51 Functional Functional Walkdown

>RLGM Interaction - Block Wall 212 1EADB Auctioneering Diode Assembly AUX 560 BB-51 Functional Functional Walkdown

>RLGM Interaction - Block Wall 213 BEBA 125 VDC Battery AUX 554 DD-54 Functional Functional Walkdown SEWS included in tPEEE, pg 404. However, battery rack vRLGM Interaction - Block Walls modified since original evaluation.

214 1EBB 125 VDC Battery AUX 554 CC-55 Functional Functional Walkdown SEWS included in IPEEE, pg 404. However, battery rack vRLGM Interaction - Block Wails modified since original evaluation.

215 1EBC 125 VDC Battery AUX 554 CC-54 Functional Functional Walkdown SEWS included in IPEEE, pg 404. However, battery rack vRLGM Interaction - Block Walls modified since original evaluation.

Page 40 of 61

Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Rev. 1 Catawba Nuclear Station Unit I ESEL and HCLPF Results ESEL Normal Desired Wailkdown ID tIN Description Bldg EL Location Operating State Operating State or Walk-by Screening Notes HCLPF' Key Failure Modera 216 1EBI 125 VOC Battery AUX 554 BB-55 Functional Functional Walkdown SEWS included in IPFEE, pg 404. However, battery rack

>BLGM Interaction - Block Walls modified since original evaluation.

217 LECA 125 VDC Battery Charger AUX 554 DD-55 Functional Functional Walkdown SEWS included in IPEEE, pg 389. However, charger

>RLGM Interaction - Block Wails replaced since original evaluation.

218 1ECB 125 VDC Battery Charger AUX 554 CC-54 Functional Functional Walkdown SIWS included in IPEEE, pg 389. However, charger uRLGM Interaction - Block Walls replaced since original evaluation.

219 1ECC 125 VDC Battery Charger AUX 554 CC-55 Functional Functional Walkdown SEWS included in IFEEE, pg 389. However, charger

>RLGM Interaction - Block Walls replaced since original evaluation.

220 LECM 125 VDC Battery Charger AUX 554 BB54 Functional Functional Walldown SEWS included in IPEtE. pg 389. However, charger

>RLGM Interaction - Block Walls replaced since original evaluation.

221 1EDA 125 VDC Distribution Center, compartments FDIC, FOLD, F020, F03B, FO2A, FO3A AUX 554 DIDS5 Functional Functional Walk-by

>RLGIM Screened per IPEEt 222 LEDB 125 VOC Distribution Center, compartments FO2B,F03B, FO2A, F03A AUX 554 CC-S4 Functional Functional Walk-by tRLGM Screened per IPFEE 223 1EDC 125 VDC Distribution Center, compartments F01C, FOSD, F02B, F03B, F02A, F03A AUX 554 CC-55 Functional Functional Walk-by

>RLGM Screened per IPFEE 224 10DD 125 VOC Distribution Center, compartments FODC, FOSS, F02B, F02A, FO3A AUX 544 BB-54 Functional Functional Walk-by

>BLGM Screened per IPEEt 225 1CAFTSO40 lCAPUTI Flow Transmitter AUX 546 BB-S1 Functional Functional Walkdown

>RLGM Screened per EFPRI NP-6041 226 1CFLT5610 Steam Generator Wide Range Level Instrument AN1 568 18 Dlg 63 Rad Functional Functional Walkdown 0.86 Functional Failure 227 1CFLT5620 Steam Generator Wide Range Level Instrument AN1 567 168 Deg 59 Rad Functional Functional Walkdown 0.86 Functional Failure 228 1CFLT5630 Steam Generator Wide Range Level Instrument ANt 565 2 10 Deg 59 Rad Functional Functional Walkdown 0.86 Functional Failure 229 ICFLT5640 Steam Generator Wide Range Level Instrument ANB 568 335 Deg 59 Rad Functional Functional Walkdown 0.86 Functional Failure 230 INCPTS12O RCS Wide Range Pressure Instrument AUX 566 CC-50 Functional Functional Walkdown

>RLGM Screened per EPHI NP-6041 231 1NCPTS140 RCS Wide Range Pressure Instrument AUX 566 DD-Si Functional Functional Walkdown

>RLGM Screened per EPRI NP-tO4l 232 ICFLTS490 Steam Generator Narrow Range Level Instrument ANI 575 1 Deog 59 Rad Functional Functional Walbdown 0,86 Functional Failure 233 1CFLTS520 Steam Generator Narrow Range Level Instrument AN1 569 130 Deog 59 Rad Functional Functional Walbdown 0.86 Functional Failure 234 1CFLT5550 Steam Generator Narrow Range Level Instrument ANS 575 205 Deog 59 Rad Functional Functional Walkdown 0.86 Functional Failure 235 1CFLTS580 Steam Generator Narrow Range Level Instrument ANt 568 315 Deg 59 Rad Functional Functional Walkdown 0.86 Functiona1 Failure 236 INCLT5171 Pressudzer Level Instrument AN1 570 104 Deg 57 Rad Functional Functional Walkdown 0.29 Functional Failure 237 SSMpTSO80 Steam Generator Pressure Instrument AUX 582 DD0*0 Functional Functional WaIkdown

>8LGM Screened per EPRI NP-6041 238 1SMPTS110 Steam Generator Pressure Instrument AUX 582 0D-52 Functional Functional Walkdown

>RLGM Screened per EPHI NP-6041 239 13SMPT5140 Steam Generator Pressure Instrument AUX 582 DD-52 Functional Functional Walbdown

>RLGM Screened per EPRI NP-6041 240 ISMPT5170 Steam Generator Pressure Instrument AUS 5872 DD46 Functional Functional Walbdown

>uLGM Screened per EPRI NP-BO41 241 1NCRD5850 1A NC Loop Hot Leg Wide Range Temperature CV1 567 20 Deog 20 Rad Functional Functional Walkdown

>HLGM Suraoend per EDRI NP-6FIt 242 1NCRDS860 1A6 NC Loop Cold Leg Wide Range Temperature CVH 567 51 Deg 28 Fad Functional Functional Walkdown

>RLGM Screened per EPRI NP-6045 243 1NCRD5870 18 NC Loop Hot Leg Wide Range Temperature CVS 587 160 Deg 18 Rad Functional Functional Walkdown cRLGM Screened per EPRI NP-SO4 Page 41 of 61

Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Rev. 1 Catawba Nuclear Station Unit 1 ESEL and HCLPF Results ESEL Normal Desired Walkdown ID EIN Description Ildg EL Location Operating State Operating Sttte or Walk-by Screening Notes HC1PF Key Failure Mode-244 1NCRDS88O 18 NC Loop Cold Leg Wide Range Temperature CV1 567 124 Deo 28 Rad Functional Functional Walkdown RLGM Screened per EPRI NP-6041 245 1NCRD5900 IC NC Loop Hot Leg Wide Range Temperature CV1 567 204 Deg 20 Rad Functional Functional Walkdowr

>RLGM Screened per EPRI NP-6041 246 1NCRD5910 IC NC Loop Cold Leg Wide Range Temperature Cvi 567 240 Deg 29 Rad Functional Functional Walkdown

>RLGM Screened per EPRI NP-6041 247 1NCRDS5920 1D NC Loop Hot Leg Wide Range Temperature CV1 567 340 Deg 20 Rad Functional Functional Walkdown

>RLGM Screened per EPRI NP-6041 248 INCR05930 1D NC Loop Cold Leg Wide Range Temperature CV1 567 309 Deg 28 Rad Functional Functional Walkdown

>RLGM Screened per EFRI NP-6041 249 1EATC7 Essential Area Terminal Cabinet AUX 577 FF-55 Functional Functional Walkdown

>RLGM interaction - Block Walls 250 ITBOX0691 SR/IR N31/35 Neutron Flux Amplifier AUX 52 CC-51 Functional Functional Walkdown

>R5GM Screened per EPRI NP-6041 251 iTBOR0R9 N131/35 Neutron Flu. Amplifier Isolotion Transtonnet AUX 579 CC-51 Functinnal Functional Walkdon

>RLGM Screened per EPRI NP-6041 252 1NIS1 Outer Core Nudear Instrument Cabinet Racs 1 AUX 594 CC-56 Functional Functional Walkdon 0.30 Interaction - Control Room Ceiling 253 INCLT6390 RVLIS Plenum (Upper Range) Level Channel 1 AUX 582 AAR49 Functional Functional Walkdown

>RLGM Screened per EPRI NP-6041 254 1NCLT6400 RVLS Narrow Range Level Channel 1 AUX 582 AA-49 Functional Functional Walkdown

>RLGM Screened per EPRI NP-6041 255 IMC9 Main Control Room Panel AUX 594 CC-56 Functional Functional Walkdowcn 0.30 Interaction - Control Room Ceiling 256 ONILT5260 Containment Sump Level RX1 552 20 Deg 50 Rad Functional Functional Walkdown cRLGM Screened pen EPI NP-6041 257 lNILT5261 Containment Sump Level RXl 556 2 Deg 45 Rad Functional Functional Walkdown nRLGM Screened per EPRI NP-6041 258 INILT5262 Containment Sump Level RX1 560 4 Deg 45 Rad Functional Functional Walkdown

>RLGM Screened per EPRI NP-6041 259 1NILT5263 Containment Sump Level RX1 565 2 Deg 45 Rad Functional Functional Walkdown

>RLGM Screened per EPRI NP-6d41 260 1NILTS264 Containment Sump Level RX1 569 0 Deg 56 Rad Functional Functional Walkdown

>RLGM Screened per EPRI NP-6041 261 1NIMT5260 Containment Sump Level AUX 577 CC-47 Functional Functional Walkdown

>RLGM Screened per EPRI NP-6041

• HCLPF values of >RLGM indicate that the HCLPF eaceeds the Revsei Level Ground Motion (0.29g), but that a spevific HCOPF value mas not calculated since thecomponent mas screened *unt ftom further evaluation.

t-Key Failure Modes are defined as followns Screened per IPEEP - Indicates that the component mas evaluated In the IPEEE and therefore meets the RLGM demand.

Screened per EPRI NP-6041 - Indicates that the component meets the screening criteria of EPRI NP-6041, Table 2-4 and that neither anchorage, relay chatter, nor interactions limit the reported HCLPF.

Interaction - Block Walls - Indicates that the component is located near a block wall. The block malt mas evaluated In IPEEE and therefore the block mall meets the RLGM demand. The functional and anchorage HCOPFs exceed the reported HCOPF value.

Interaction -Control Room Ceiling - indicates that the component is located in the control room. The control room ceiling mas evaluated in this report and has a HCLPF ofO.30g. The functional and anchorage HCROFs exceed the reported HCLPF value.

Anchorage - Indicates that anchorage is the governing failure mode for the component.

Functional Failure -Indicates that functional failure Is the governing failure mode for the component.

Relay Chatter - Indicates that relay chatter is the gonereing failure mode for the component.

Modlficatnon/Investigat Ion - Indicanes that the reported HCOPF value Is conditlonal on the modification and/or further investigation as reported on the SEWS.

Total Items; 258 Page 42 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 Appendix B Catawba Nuclear Station Unit 2 ESEL and HCLPF Results Page 43 of 61

Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Rev. 1 Catawba Nuclear Station Unit 2 ESEL and HCLPF Results ESEL Normal Desired Walkdown or IS EIN Descriptlon Bldg EL Location Operating State Operating State Walk-by Screening Notes HCLPFP Key Failure Mode-1 2NC250A Reactor Head Vent Valve CV2 600 38 Deg 25 Rtd Closed Open Walkdown vRLGM Screened per EPRI NP-SO4 2

2NC253A Reactor Head Vent Valve CV2 600 43 Deg 27 Rtd Closed Open Walkdown

>RtGM Screened per EPRI NP-6041 3

2NC1 Pressurizer Safety Valve CV2 637 102 Deg 37 Rad Closed Functional Walk-by Induded in IPEEE, pg 718.

TBS TBD 4

2NC2 Safety/Relief Valve CV2 637 105 Deg 37 Rad Closed Functional Walk-by Included in IPEEE, pg 718.

TBD TBD 5

2NC3 Safety/Relief Valve CV2 637 106 Deg 37 Rod Closed Functional Walk-by Included in IPEEE, pg 718.

TBD TBD 6

2NC34A Pressurizer Power Operated Relief Valve CV2 635 105 Deg 39 Red Closed Closed Walk-by Induded in IPEEE, pg 718.

TSD TBD 7

2NCTI1 Pressurizer Relief Tank 2 CV2 554 90 Deg Functional Functional Walkdown TBD TSD 8

2NVPA NC Letdown to Regenerative Heat Exchanger Isolation Valve CV2 554 228 Oeg 38 Rad Open Closed Walkdown TBD TBD 9

2ARFD2 Air Return Fan Damper 2 CV2 595 265 Deg 53 Rad Closed Open Walk-by TBD TSD 10 2ARFD4 Air Return Fan Damper 4 CV2 595 292 Deg 49 Rad Closed Open Walk-by TBD TBD 11 2AVXCARF 2A Containment Air Return Fan (CARP-ZA)

CV2 593 265 Deg Off On Walk-by TBD TBD 12 2BVXCARF 2B Containment Air Return Fan (CARF-2B)

CV2 593 293 Deg Off On Walk-by TSD TBD 13 2VXHSFA H2 Skimmer Fan A CV2 645 265 Deg Off On Walkdown TBD TBD 14 2VXHSFB H2 Skimmer Fan S CV2 645 220 Deg Off On Walkdown TBD TBD 15 2VX1A H2 Skimmer Fan A Damper CV2 659 263 Deg Closed Open Walkdown TSD TSD 16 2VX2B H2 Skimmer Fan B Damper CV2 659 283 Deg Closed Open Walkdown TSD TBD 17 2NDPB Residual Heat Removal A Train Isolation Valve CV2 572 170 Deg 33 Red Closed Open Walk-by Included in IPEEE, pg 1316.

TSD TBD 18 2ND2A Residual Heat Removal A Train Isolation Valve CV2 567 176 Deg 50 Rad Closed Open Walk-by Induded in IPEEE, pg 1316.

TBD TBD 19 2NDPUA Residual Heat Removal Pump 2A AU4 522 FF-60 Functional Functional Walk-by

>RLGM Screened per IPEEE 20 2NDHXAPMP Residual Heat Removal Pump Mechanical Seal Heat Exchanger 2A AUX 522 FF-60 Functional Functional Walk-by Rule-of-the-box with 2NDPUA vRLGM Screened per IPEEE 21 2KCHX40 Residual Heat Removal Pump 2A Motor Cooler AUX 522 FF-60 Functional Functional Walk-by Rule-of-the-box with 2NDPUA

>RLGM Screened per IPEEE 22 2NDHXA Residual Heat Removal Heat Exchanger 1A AUX 560 LL-62 Functional Functional Walkdown

>RLGM Screened per EPRI NP-6041 23 2KFR101B FW System/ KF System Isolation Valve AUX 583 JJ-62 Closed Open Walkdown

>RLGM Screened per EPRI NP-6041 24 2FWTK01 Refueling Water Storage Tank (FWST)

Yard Intact Intact Walkdown 0.30 Anchorage 25 2KCPUA1 Component Cooling Pump 261 AUX 577 EE-58 Functional Functional Walk-by

>RLGM Screened per IPEEE 26 2KCHIAS Component Cooling Pump 2A1 Motor Cooler AUX 577 EE-58 Functional Functional Walk-by Rule-of-the-box with 2KCPUAS

>RLGM Screened per IPEEE 27 2KCHXA Component Cooling Heat Exchanger 2A AUX 577 HH-59 Functional Functional Walk-by

>RLGM Screened per IPEEl 28 2KCOS5A Residual Heat Removal Heat Exchanger 1A Isolation Valve AUX SMA U1-61 Clned Open Walk-by

>RLGM Screened per IPEEE 29 2RN250A Nuclear Service Water to ICAPUTP Isolation Valve AUX 584 KK-59 Closed Open Walk-by

>RLGM Screened per IPEEE 30 2AWLCVU 2A Lower Containment Ventilation Unit (LCVU-2A)

CV2 565 16 Deg Off On Walkdown TSD TBS 31 2DWLCVU 2D Lower Containment Ventilation Unit (LCVU.2D)

CV2 565 344 Peg Off On Walkdown TSD TBD Page 44 of 61

Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Rev. 1 Catawba Nuclear Station Unit 2 ESEL and HCLPF Results ESEL Normal Desired Walkdown or ID EIN Desorption Bldg EL Location Operating State Operating State Walk-by Screening Notes HCLPF' Key Failure Mode-32 28824 Steam Generator Blowdown Flow Control Valve TB2 581 2L-29 Open Closed Walkdown

>RLGM Screened per EPRI NPS6041 33 2BB65 Steam Generator Blowdown Flow Control Valve TB2 581 2K-29 Open Closed Walkdown

>RLGM Screened per EPRI NP-6041 34 2BB69 Steam Generator Blowdown Flow Control Valve T82 583 2L-30 Open Closed Walkdown

>RLGM Screened per EPRI NPF6041 35 2BB73 Steam Generator Blowdown Flow Control Valve TB2 583 2M-29 Open Closed Walkdown

>RLGM Screened per EPRI NP-6t041 36 2CA174 Condenser Circulating Water System Isolation Valve AUX 545 CC-61 Closed Open Walk-by

>RLGM Screened per IPEEE 37 2CA656A Nuclear Service Water System Train 2A isolation Valve AUX 555 BB-62 Closed Open Walk-by

>RLGM Screened per IPEEE 38 2CA257 2CAPUTD Feedwater Safety Valve AUX 540 BB-63 Closed Open Walkdown

>RLGM Modification/Investigation 39 2CAPUTD Turbine Driven Auxiliary Feed Water Pump AUX 531 AA 63 Functional Functional Walk-by

>RLGM Screened per IPEEE 40 2CAHX04 Turbine Driven Auxiliary Feed Water Pump Lube Oil Cooler AUX 530 AA-62 Functional Functional Walk-by Rule-of-the-box with 2 CAPUTD

>RLGM Screened per IPEEE 41 2CA64 Auxiliary Feedwater to Steam Generator 2A AUX 557 B-64 Open Throttled Walkdown

>RLGM Modification/investigation 42 2CA52 Auxiliary Feedwater to Steam Generator 2B AUX 43 2CA48 Auxiliary Feedwater to Steam Generator 2C AUX 44 2CA36 Auxiliary Feedwater to Steam Generator 2D AUX 45 2SV19 Power Operated Relief Valve -Steam Generator 2A DH2 46 2SV13 Power Operated Relief Valve -Steam Generator 28 DH2 47 2SV7 Power Operated Relief Valve -Steam Generator 2C DH2 48 2SV1 Power Operated Relief Valve -Steam Generator 2D DH2 49 2SV2 Steam Generator 2D Safety Valve DH2 50 2SV3 Steam Generator 2D Safety Valve DH2 51 2SV4 Steam Generator 2D Safety Valve DH2 52 2SV5 Steam Generator 2D Safety Valve DH2 53 25V6 Steam Generator 2D Safety Valve DH2 54 25M1 Steam Generator 2D Main Steam Isolation Valve DH2 55 2SV6 Steam Generator 2C Safety Valve D02 56 2SV9 Steam Generator 2C Safety Valve DH2 57 2SV10 Steam Generator 2C Safety Valve DH2 58 2SV11 Steam Generator 2C Safety Valve 0H2

>9 2SV12 Steam -nnerater 2C Safety Valve DH2 60 2SM3 Steam Generator ZC Main Steam Isolation Valve DH2 61 25V14 Steam Generator 2B Safety Valve DH2 62 2SV15 Steam Generator 2B Safety Valve DH2 550 OD-62 551 DD-61 554 BB-65 594 FF-71 635 FF-6O 635 FF-61 635 FF-69 615 EE-69 615 EE-69 615 EE-69 615 EE-69 615 EE-69 615 DD-69 615 EE-61 615 EE-61 615 EE-61 615 EE-61 615 EE-61 615 D0-68 615 EE-60 615 EE-60 Open Open Open COosed Closed Closed Closed Closed Closed Closed Closed Closed Open Closed Closed Closed Closed Closed Open Closed Closed Throttled Throttled Throttled Open Open Open Open Functional Functional Functional Functional Functional Closed Functional Functional Functional Functional Functional Closed Functional Functional Warkdown Walkdown Walkdown Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-bV Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Included in IPEEE, pg 721.

Included in [PEEE, pg 721.

Included in IPEEE, pg 721.

Included in IPEEE, pg 721.

Included in IPEEE, pg 721.

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>R1GM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RUGM

>RLGM

>RLGM

>RLGM Screened per EPRI NP-6041 Screened per EPRI NP-6041 Modification/Investigation Screened per PEEE Screened per IPEEE Screened per IPEEE Screened per IPEEE Screened per IPEEE Screened per IPEEE Screened ver IPEEE Screened per IPEEE Screened per IPEEE Screened per IPEE Screened per IPEEE Screened per PEEE Screened petr PEE8 Screened par P6EE Screened per IPEEE Screened per IPEEE Screened per IPEEE Screened per IPEEE Induded in IPEEE, pg 721.

Included in IPEEE, pg 72 2 Included in FPEEE.

pg 722.

Included in FPEE8, pg 722.

Included in iPEEE, pg 722.

Included in IPEEE, pg 722.

Included in IPEEE, pg 722.

Page 45 of 61

Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Catawba Nuclear Station Unit 2 ESEL and HCUPF Results ESEL ID EIN Description Bldj 63 2SV16 Steam Generator 2B Safety Valve DH2 64 2SV17 Steam Generator 2B Safety Valve DH2 65 2SV1£ Steam Generator 2£ Safety Valve DH2 6£ 25M5 Steam Generator 2B Main Steam Isolation Valve DH2 67 25V20 Steam Generator 2A Safety Valve DH2 68 2SV21 Steam Generator 2A Safety Valve DH2 69 2SV22 Steam Generator 2A Safety Valve DH2 70 2SV23 Steam Generator 2A Safety Valve DH2 71 2SV24 Steam Generator 2A Safety Valve DH2 72 25M7 Steam Generator 2A Main Steam Isolation Valve DH2 73 2SA144 2CAPUTD Control Valve AUX 74 2SA145 2CAPUTD Trip Throttle Valve (Stop Valve)

AUX 75 25A2 2CAPUTD Steam Feed Isolation Valve DH2 76 2NDHXB Residual Heat Removal Heat Exchanger 2B AUX 77 2NI54A Accumulator Tank 2A Isolation Valve CV2 78 2NI95B Accumulator Tank 2B Isolation Valve CV2 79 2NI76A Accumulator Tank 2C Isolation Valve CV2 so 2NI88B Accumulator Tank 20 Isolation Valve CV2 81 2NSHXA Containment Spray Heat Exchanger ZA AUX 82 2NSHXB Containment Spray Heat Exchanger 2B AUX

£3 2NMHX07 Reactor Coolant Hot Leg Sample Heat Excharger A AUX 84 2NMHX08 Reactor Coolant Hot Leg Sample Heat Exchanger £ AUX 85 2EMF46A ATrain KC Radiation Monitor AUX G6 2KC2B Auxiliary Building Non-Essential Return Header Isolation Valve AUX 87 2KC3A Reactor Building Non-Essential Return Header Isolation Valve AUX 88 2KC13B Auxiliary Building Non-Essential Return Header Isolation Valve AUX 89 2KC230A Reactor Building Non-Essential Return Header Isolation Valve AUX 90 2KCTKA Component Cooling Surge Tank 1A AUtX 91 2NVHX04 Let Down Heat Exchanger AUX 92 2KFHXA Fuel Pool Heat Exchanger 2A AUX 93 2KFHXB Fuel Pool Heat Exchanger 2B AUX Rev. 1 EL 615 615 615 615 615 615 615 515 615 615 531 534 628 560 560 560 560 560 560 56G 543 543 577 586 586 584 586 594 577 577 577 Location EE-60 EE-60 EE-60 DD-60 EE-70 EE-70 EE-70 EE-70 EE-70 DD-70 BB-63 BB-63 FF-61 LL-62 46 Deg 55 Rod 136 Deg 55 Red 222 Deg 45 Red 312 Deg 45 Rad LL-62 LL-62 R. 248 Rm 248 JJ-60 FF-56 FF-57 HH-59 HH-59 NN-59 KK-61 NN-61 NN-62 Normal Operating State Closed Closed Closed Open Closed Closed Closed Closed Closed Open Open Open Closed Functional Open Open Open Open Intact Intact Intact Intact Intact Open Open Open Open Intact Intact Intact Intact Desired Operating State Functional Functional Functional Closed Functional Functional Functional Functional Functional Closed Throttled Throttled Open Functional Closed Closed Closed Closed Intact Intact Intact Intact Intact Closed Closed Closed Closed Intact Intact Intact Intact Walkdoemn or Walk-by Walk-by Walk-by Walk-by Walk-by Walk by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walkdown Walk-by Walk-by Walk-by Walk-by Walkdown Walkdeown Walkdown Walkdown Walkdown Walk-by Walk-by Walk-by Walkdown Walkdown Walbdown Walkdown Walkdown Induded in IPEEE, pg 722.

Induded in IPEEE, pg 722.

Induded in IPEEE.

pg 722.

Induded in IPEEE, pg 722.

Induded in IFEEE, pg 722.

Rule-of-the-box with OCAPUD Rule-of-the-box with 1CAPUD Included in IPEEE.

pg 721.

Screening Notes Included in IPEEE, pg 722.

Induded in IPEEE, pg 722.

Induded in IPEEE, pg 722.

Included in IPEEE, pg 719.

Included in IPEEE, pg 719.

Included in IPEEE, pg 719.

Included in IPEEE, pg 719.

HC6PF'

>RLGM vRLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM 0.31

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM 0.30

>RLGM

>RLGM Key Failure Modesa Screened per 0PEEE Screened per 1PEEE Screened per 1PEEE Screened per 0PEEE Screened per IPEEE Screened per 1PEEE Screened per lPEEE Screened per 1PEEE Screened per 1PEEE Screened per IPEEE Screened per PEEE Screened per 5PEEE Screened per PEEE Screened per EPRI NP-6041 Screened per IPEEE Screened per IPEEE Screened per PEEE Screened per PEEE Screened per EPRI NP-6041 Rev. 1 Screened per EPRI NP-6041 Modification/Investigation Modification/Investigation Anchorage Screened per IPEEE Screened per IPEEE Screened per IPEEE Screened per EPRI NP-6041 Screened per EPRI NP-6041 Anchorage Screened per EPRI NP.6041 Screened per EPRI NP-6041 Page 46 of 61

Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Catawba Nuclear Station Unit 2 ESEL and HCLPF Results Rev. 1 ESEL ID EIN Description Bldg 94 2RN232A 2A Diesel Generator Cooling Water Isolation Valve D2A 95 2MC1 Main Control Room Panel AUX 96 2MC10 Main Control Room Panel AUX 97 2MC2 Main Control Room Panel AUX 98 2MC4 Main Control Room Panel AUX 99 ZMC5 Main Control Room Panel AUX 100 2PCC1 Process Control Cabinet s Protection Set I Cabinet AUX 101 2PCC2 Process Control Cabinet 2 Protecton Set 2 Cabinet AUX 102 2PCC3 Process Control Cabinet 3 Protecton Set 3 Cabinet AUX 103 2MC8 Main Control Room Panel AUX 104 2PCC4 Process Control Cabinet 4Protecoon Set 4 Cabinet AUX 105 2PCC5 Process Control Cabinet 5Protecyon Set 5 Cabinet AUX 106 2PCC6 Process Control Cabinet 6 Protection Set 6 Cabinet AUX 107 2PCC7 Process Control Cabinet 7 Protection Set 7 Cabinet AUX 108 2PCC8 Process Control Cabinet 8 Protection Set 8 Cabinet AUX 109 2RVLIS RVLS Process Control Cabinet AUX 110 2TBOXG691 SIVlR N31/N35 Neutron Flux Amplifier Terminal Box AUX i11 2TBOXO089 SR/IR N31/35 Neutron Flux Amplifier Isolation Transformer Terminal Box AUX 112 2NISG Outer Core Nudear Instrument Cabinet Rack I AUX 113 2TBOX0519 Wide Range Neutron Flux Signal Processor Terminal Box AUX 114 2TBOXG537 Wide Range Neutron Flux Amplifier Terminal Box AUX 115 2TBOX05G 6

Wide Range Neutron Flux Power Suppl Terminal Box AUX 116 2TBOX5S87 Wide Range Neutron Flux Power Supply Terminal Box AUX 117 2MTSWOO06 Transfer Switch (Single Pole Double Throw), Hydrogen Igniter Group A AUX 118 2XFMROD13 Transformer, 480 VAC - 120 VAC, Hydrogen Igniter Group A AUX 119 2VREGG1O3 Voltage Regulator, Hydrogen Igniter Group A AUX 120 2EATC8 Essential Area Temninal Cabinet AUX 121 2ELCP0278 Hydrogen Igniter Group A Control Panel AUX 122 2DGLSA Diesel Generator No. 2A Load Sequencer Relay Loading Relay AUX 123 2ELCP0357 SW01, Hydrogen Igniter On/Off switch (Alt. Source) located on panel 2ELCPO357 AUX Location DD-T7 AA-57 BB-58 AA-58 BB-S8 BB-58 BB-59 BB-60 B&-59 BS-58 88-60 CC-59 CC-60 CC-S9 CC-60 CC-61 CC-63 CC-63 CC-58 AA-63 AA-63 AA-63 AA-63 CC-65 CC-6S CC-65 EE 60 CC,55 BB-71 cc-6S Normal Operating State Closed Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Energized Energized Functional De-Energized Functional Functional Desired Operating State Closed Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Energized Energized Functional Energized Functional Functional Walbdowrn or Walk-by Walk-by Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Wolkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Screening Notes HCLPF*

>RLGM 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30

>RLGM RLGM

>RLGM 0.30 RLGM

>RLGM RtGM

>RLGM

>RLGM

>RLGM

>RLGM O.32

>RLGM

>RLGM

>RUGM Key Failure Mode 5

Screened per PEEE Interaction - Control Room Ceiling Interaction - Control Room Ceiling Interaction - Control Room Ceiling Interaction - Control Room Ceiling Interaction - Control Room Ceiling Interaction - Control Room Ceiling Interaction - Control Room Ceiling Interaction - Control Room Ceiling Interaction - Control Room Ceiling Interaction - Control Room Ceiling Interaction - Control Room Ceiling Interaction - Control Room Ceiling Interaction - Control Room Ceiling Interaction - Control Room Ceiling Interaction - Block Wall, Screened per EPRI NP-6041 Screened per EPRI NP-6041 Interaction - Control Room Ceiling Screened per EPRI NPA6041 Screened per EPRI NP-GU41 Screened per EPRI NP-604G Screened per EPRi NP-604L Interaction - Block Walls Interaction - Block Walls Interaction - Block Walls Anchorage Interaction - Block Walls Screened per EPRi NP-604A Interaction - Block Walls Functional Walbdowo GOR Interaction -Control Room Ceiling 124 2MC7 Main Control Room Panel AUX 594 BRB-56 Functional Functional Walkdown 0.30 Interaction

-Control Room Ceiling Page 47 of 61

Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Rev. 1 Catawba Nuclear Station Unit 2 ESEL and HCLPF Results ESEL Nonnal Desired Walkdown or ID EIN Description Bldg EL Location Operating State Operating State Walk-by Screening Notes HCLPFP Key Failure Mode-125 2EHMOIO3 Group 2A Igniter Box CV2 562 88 leg 48 tad De-Energized Energized Walkdown 0.33 Anchorage 126 2EHMOOS Group 2A Igniter Box CV2 562 178 Deg 51 Rad De-Energized Energized Waikdown 0.33 Anchorage 127 2EHMOD07 Group 2A Igniter Box CV2 562 277 Deg 46 Rad De-Energized Energized Walkdown 0.33 Anchorage 128 2EHM0009 Group 2A Igniter Box CV2 562 5 Deg 46 Rad De-Energized Energized Woikdown 0.33 Anchorage 129 2EHM0071 Group 2A Igniter Box CV2 555 103 Deg 35 Bad De-Energized Energized Walkdown 0.33 Anchorage 130 2EHIIMOGS Group 3A Igniter Box CV2 601 324 Deg 20 Rad De-Energized Energized Walkdown 0.29 Functional Failure 131 2EHM0013 Group 3A Igniter Box CV2 590 326 Deg 49 Rad De-Energized Energized Walkdown 0.33 Anchorage 132 2EHM0015 Group 3A Igniter Box CV2 642 335 Deg 41 Rad De-Energized Energized Walkdown 0.29 Functional Failure 133 2EHM0017 Group 3A Igniter Box CV2 601 55 Deg 18 Rod De-Energized Energized Walkdown 0.29 Functional Failure 134 2EHM0019 Group 3A Igniter Box CV2 590 2 Deg 51 Rad De-Energized Energized Walkdown 0.33 Anchorage 135 2EHM0021 Group 3A Igniter Box CV2 642 I8 Deg 41 Rad De-Energized Energized Walkdown 0.29 Functional Failure 136 2EHM0023 GrOup 4A Igniter Box CV2 590 53 Deg 50 Bad De-Energized Energized Walkdown 0.33 Anchorage 137 2EHM0025 Group 4A Igniter Box CV2 590 214 Deg 48 Rad De-Energized Energized Waldown 0.33 Anchorage 138 2EHM0027 Group 4A Igniter Box CV2 590 245 Deg 51 Rad De-Energized Energized Walkdown 0.33 Anchorage 139 2EHM0029 Group 4A igniter Box CV2 590 91 Deg 51 Rad De-Energized Energized Walkdown 0.33 Anchorage 140 2EHM0031 Group 4A Igniter Box CV2 603 10 Deg 12 Rad De-Energized Energized Walkdown 0.29 Functional Failure 141 2EHM0033 GrOup 4A Igniter Boo CV2 641 114 Deg 34 Rad De-Energized Energized Walkduwn 0.29 Functional Failure 142 2EHMOD35 Group 5A Igniter Box CV2 601 140 Deg 30 Bad De-Energized Energized Walkdown 0.29 Functional Failure 143 2EHM0037 Group 5A Igniter Box CV2 590 146 Deg 51 Red De-Energized Energized Waldown 0.33 Anchorage 144 2EHM0039 Group 5A Igniter Box CV2 642 101 Deg 41 RBd De-Energized Energized Walkdown 0.29 Functional Failure 145 2EHMO041 Group 5A Igniter Box 002 601 216 Deg 21 Bad De-Energized Energized Walkdown 0.29 Functional Failure 146 2EHMOD43 Group 5A Igniter Box CV2 590 173 Deg 51 Red De-Energized Energized Walkdown 0.33 Anchorage 147 2EHMO045 Group 5A Igniter Box CV2 642 206 Deg 41 RBd De-Energized Energized Wzlkdown 0.29 Functional Failure 148 2EHM0059 Group 7A Igniter Box CV2 714 318 Deg 24 Bad De-Energized Energized Walkdown 0.29 Functional Failure 149 2EHM0061 Group 7A Igniter Box CV2 714 49 Deg 24 Rad De-Energized Energized Walkdown 0.29 Functional Failure 150 2EHM0063 Group 7A Igniter Box CV2 714 140 Deg 24 Bed De-Energized Energized Walkdown 0.29 Functional Failure 151 2EHMg065 Group 7A Igniter Box CV2 714 218 Deg 24 Bad De-Energized Energized Walkdown 0.29 Functional Failure 152 2EHM0053 Group GA-1 Igniter Box CV2 666 108 Deg 46 Rad De-Energized Energized Walkdown 0.29 Functional Failure 153 2EHM055 Group GA-1 Igniter Box CV2 666 157 Deg 46 Rad De-Energized Energized WaIkdewn 0.29 Functional Failure 154 2EHM0057 Group 6A-1 Igniter Box CV2 666 206 Deg 46 Rad De-Energized Energized Walkdown 0.29 Functional Failure 155 2EHMOD47 Group 6A-2 Igniter Box CV2 666 321 Deg 46 Bad De-Energized Energized Walkdown 0.29 Functional Failure Page 48 of 61

Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Rev. 1 Catawba Nuclear Station Unit 2 ESEL and HCLPF Results ESEL Normal Desired Walkdown or ID EIN Description Bldg EL Location Operating State Operating State Walk-by Screening Notes HCLPF' Key Failure Mode-156 2EHM0049 Group GA-2 Igniter Box CV2 666 11 Deg 46 Rod De-Energized Energized Walkdown 0.29 Functional Failure 157 2EHMOS0 Group GA.2 Igniter Box CV2 666 59 Deg 46 Rod De-Energized Energized Walkdown 0.29 Functional Failure 158 2EHMO067 Group &A Igniter Box CV2 653 218 Deg 30 Rod De-Energized Energized Walkdown 0.29 Functional Failure 159 2EHM0069 Group 8A Igniter Box CV2 653 41 Deg 32 Rad De-Energized Energized Walkdown 0.29 Functional Failure 160

?AFWPTCP Auxiliary Feedwater Pump Turbine Control Panel IELCPB245)

AUX 543 CC-61 Functional Functional Walkdown cRLGM Interaction - Block Wall 161 2EDE 125 VDC Distribution Center, compartments FP0A, F01G, FOll AUX 577 BB-68 Functional Functional Walk-by cRLGM Screened per IPEEE 162 2EATC23 Essential Area Terminal Cabinet AUX 577 CC-65 Functional Functional Walkdown 0.37 Anchorage 163 255PSA Solid State Protection System Cabinet, Control Panel, Control Cabinet AUX 594 CC-S9 Functional Functional Walkdown 0.30 Interaction - Control Room Ceiling 164 2EATC21 Essential Area Terminal Cabinet AUX 577 BB-62 Functional Functional Walkdown cRLGM Interaction - Block Walls 165 2EPA 125 VDC Power Panelboard AUX 554 DD-55 Energized Energized Walk-by

>RLGM Screened per IPEEE 166 2EATC12 Essential Area Terminal Cabinet AUX 577 FF-59 Functional Functional Walkdown 0.29 Relay Chatter 167 2TBOX0522 Control Panel D2A 556 B8B69 Functional Functional Walkdown vRLGM Screened per EPRI NP-6041 168 2SMTC1 Control Panel AUX 577 GG-59 Functional Functional Walkdown

>RLGM Interaction - Block Walls 169 2EIA Vital 120 VAC Inverter AUX 554 0D-59 Energized Energized Walkdown SEWS induded in IPEEE, pg 1474. However, inverter

>RLGM Interaction - Block Walls replaced since original evaluation.

170 2EIB Vital 120 VAC Inverter AUX 554 CC-59 Energized Energized Walkdown SEWS included in IPEEE, pg 1475. However, inverter RLGM Interaction - Block Walls replaced since original evaluation.

171 2EIC Vital 120 VAC Inverter AUX 554 CC-59 Energized Energized Walkdown SEWS induded in IPEEE.

pg 1476. However, inverter cRLGM Interaction - Block Walls replaced since original evaluation.

172 2EID Vital 120 VAC Inverter AGO 554 BB-5B Energized Energized Walkdown SEWS induded in IPEEE, pg 1477. However, inverter cRLGM Interaction - Block Walls replaced since original evaluation.

173 2ERPD 120 VAC Power Panel AUX 554 BB-5S Energized Energized Walk-by cRLGM Screened per [PEEE 174 2ERPB 120 VAC Power Panel AUX 554 CC-59 Energized Energized Walk-by oRLGM Screened per IPEEE 175 2ERPA 120 VAC Power Panel AUX 554 DDB59 Energized Energized Walk-by

>RLGM Screened per IPEEE 176 2ERPC 120 VAC Power Panel AUX 554 CC-59 Energized Energized Walk-by vRLGM Screened per FPEEE 177 2EATC13 Essential Area Terminal Cabinet AUX 560 SI-SB Functional Functional Walkdown 0.29 Relay Chatter 178 2EDF 125 VDC Panel, 125 VDC Distribution Center, compartments FPIA, FB11 AUX 560 BR-G8 Energized Energized Walk-by

>RLGM Screened per IPEEE 179 2SSPSB Control Cabinet AUX 594 DD-59 Functional Functional Walkdewn 0.30 Interaction - Control Room Ceiling SOR 2EADA Auctioneering Diode Assembly AUX 577 BB-63 Functional Functional Walkdown oRLGM Interaction - Block Wall 181 2EADB Auctioneering Diode Assembly AUX 560 BB-63 Functional Functional Walkdown vRLGM Interaction - Block Wall 182 2EBA 125 VDC Battery AUX 554 DD-60 Functional Functional Walkdown SEWS induded in IPEEE, pg 1454. However. battery rack oRLGM Interaction - Block Walls modified since rigirnl evnaloation.

OR3 2EBB 125 VDC BGunry AGX 554 CC-S9 Functional Functional Walkdown SEWS included in IPEEE, pg 1455. However. battery rack oRLGM Interaction - Block Walls modified since original evaluation.

184 2EBC 125 VDC Battery AGX 554 CC-6B Functional Functional Walkdown SEWS indcuded in IPEEE, pg 1456. However, battery rack

>RLGM Interaction - Block Walls modified since original evaluation.

Page 49 of 61

Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Rev. 1 Catawba Nuclear Station Unit 2 ESEL and HCLPF Results ESEL Normal Desired Walkdown or ID EIN Desoription Bldg EL Location Operating State Operating State Walk-by Screening Notes HCOPF' Key Failure Mode-185 2ESD 125 VDC Battery AUX 554 BB-59 Functional Functional Walkdown SEWS induded in IPEEE, pg 1457. However, battery rack

>RLGM Interaction - Block Walils modified since original evaluation.

186 2ECA 125 VDC Battery Charger AUX 554 DD-59 Functional Functional Walbdown SEWS induded in IPEEE, pg 1464. However, charger

>RLGM Interaction - Block Walls replaced since original evaluation.

187 2ECB 125 VDC Battery Charger AUX 554 CC-60 Functional Functional Walkdown SEWS included in IPEEE, pg 1465. However, charger

>RLGM Interaction - Block Walls replaced since original evaluation.

188 2ECC 125 VDC Battery Charger AUX 554 CC-h9 Functional Functional Walkdown SEWS included in IPEEE, pg 1466. However, charger

>RLGM interaction - Block Walls replaced since original evaluation.

189 2ECD 125 VDC Battery Charger AUX 554 BB-WO Functional Functional Walkdown SEWS included in IPEEE, pg 1467. However, charger

>RLGM Interaction - Block Walls replaced since original evaluation.

190 2EDA 125 VDC Distribution Center, compartments FDIC, FOLD, FO2B, FO3B, FP2A, FO3A AUX 554 DD-59 Functional Functional Walk-by

>RLGM Screened per IPEEE 191 2EDB 125 VDC Distribution Center, compartments F02B,F03B, FO2A, FO3A AUX 554 CC-AS Functional Functional Walk-by

>RLGM Screened per PEEE 192 2EDC 125 VDC Distribution Center, compartments FOIC, FOLD, F02B, FO3B, FO2A, F03A AUX 554 CC-59 Functional Functional Walk-by

>RLGM Screened per IPEEE 193 25D5 125 VDC Distribution Center, compartments FDIC, FOSD, FO2B. FOSA, F03A AUX 544 BB-60 Functional Functional Walk-by

>RLGM Screened per IPEEE 194 2EMXI Motor Control Center, 600 VAC, single phase, normal power source for AUX 577 EE-54 Functional Functional Walk-by cREGM Screened per PEEE Hydrogen igniter Group A 195 2EMXS Motor Control Center, 480 VAC, single phase, Emergency power source for AUX 577 SB-48 Functional Functional Walk-by

>RLGM Screened per IPEEE Hydrogen Igniter Group A 196 2EMXD 197 2EMXK 198 2EMXC 199 2EMXL 200 2EMXA 201 2EMXJ 202 2EMXH 203 2EMXE 204 2EMXB 205 2EMXM 206 2EMXN 207 2ETA 208 2TBOXDOD8 209 2NSPT5370 210 2TBOXO538 211 2CAFT5040 Essential Motor Control Center, 600 VAC Essential Motor Control Center, 600 VAC Essential Motor Control Center, 600 VAC Essential Motor Control Center, 600 VAC Essential Motor Control Center, 600 VAC Essential Motor Control Center, 600 VAC Essential Motor Control Center, 600 VAC Essential Motor Control Center, 600 VAC 600 VAC Essential MCC 655 VAC Essential MCC 600 VAC Essential MCC 4160 Essential Switchgeat Terminal Box for Transmitter 2FWLTSOOO Containment Pressure Train A Terminal Box for 2CM877 & 2CM878 2CAPUTD Discharge Flow AUX AUX AUX AUX AUX AUX AUX D2A AUX AUS AUX AUX YRD AUX TO2 AUX 56O 577 577 556 577 560 594 556 560 577 560 577 598 581 574 546 BB-64 BB.67 BB-64 BB-67 FF-6O GG-58 FF-58 EE-75 FF-55 CC-61 CC-61 AA-64 50XS53Y CC-63 21-17 B-62 Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walk-by Walkdown Walkdown Walk-by Walkdown Walkdown Walkdown Walkdown

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM

>RLGM 0.32

>RLGM 0.29

>RLGM

>RLGM

>RLGM Screened perIPEEE Screened per PEEE Screened perIPEEE Screened per IPEEE Screened per IPEEE Screened per IPEEE Screened per IPEEE Screened per IPEEE Screened per IPEEE Interaction -Block Walls Anchorage Screened per IPEEE Anchorage Screened per EPRI NP-6041 Screened per EPRI NP-6041 Screened per EPRI NP-6041 Page 50 of 61

Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Catawba Nuclear Station Unit 2 ESEL and HCLPF Results ESEL ID EIN Description Bldg 212 2CFLT5610 2A Steam Generator Wide Range Level Channel #1 AN2 213 2CFLT5620 26 Steam Generator Wide Range Level Channel #2 AN2 214 2CFLT5630 2C Steam Generator Wide Range Level Channel #3 AN2 215 2CFLT5640 2D Steam Generator Wide Range Level Channel #4 AN2 216 2NCPT5120 NC Loop 2 Hot Leg Wide Range Pressure Channel #1 AUX 217 2NCPTS140 2C NC Loop Hot Leg Wide Range Pressure Channel 64 AUX 216 2CFLT5490 2A Steam Generator Narrow Range Level ChannelI4 AN2 219 2CFLT5520 2B Steam Generator Narrow Range Level ChannelI4 AN2 220 2CFLT5550 2C Steam Generator Narrow Range Level Channelad AN2 221 2CFLThSWO 2D Steam Generator Narrow Range Level Channel#4 AN2 222 2NCLT5171 Pressurizer Level-LowTemperature AN2 223 2SMPT5080 2A Steam Generator Steam Une Pressure Channel 1 AUX 224 2SMPT5110 28 Steam Generator Steam Une Pressure Channel 1 AUX 225 2SMPT5140 2C Steam Generator Steam Une Pressure Channel s AUX 226 2SMPTS170 2D Steam Generator Steam Une Pressure Channel S AUX 227 2EATC7 Essential Area Terminal Cabinet AUX 228 2M1R Normal Motor Control Center, 600 VAC AUX 229 2MXQ Blackout Motor Control Center, 60D VAC AUX 230 2NCRDS850 2A NC Loop Hot Leg Wide Range Temperature CV2 231 2NCRDSR60 2A NC Loop Cold Leg Wide Range Temperature CV2 232 2NCRRD870 2B NC Loop Hot Leg Wide Range Temperature CV2 233 2NCRRDS880 2B NC Loop Cold Leg Wide Range Temperature CV2 234 2NCRDS900 SC NC Loop Hot Leg Wide Range Temperature CV2 235 2NCRD5910 2C NC Loop Cold Leg Wide Range Temperature CV2 236 2NCRD5920 2D NC Loop Hot Leg Wide Range Temperature CV2 237 2NCRD5930 2D NC Loop Cold Leg Wide Range Temperature CV2 238 2NCLT6390 RVLIS Plenum (Upper Range) Level Channel 1 AUX 239 2NCLT6400 RVLIS Narrow Range Level Channel I AUX 240 2NILT5260 Containment Sump Level RX2 241 2NILT5261 Containment Sump Level RX2 242 2NILT5262 Containment Sump Level RX2 Rev. 1 Location 358 Deg 59 Rad 149 Deg 59 Rad 205 Deg 59 Red 327 Deg 59 Red CC-64 CC-63 042 Deg 59 Rad 130 Deg 59 Rad 205 Deg 59 Rod 315 Deg 59 Rod 113 Deg 56 Red DD-69 D0-62 DD-62 DD-69 FF-59 BB-65 B6 65 20 Deg 20 Rad 51 Deg 28 Rad 160 Deg 18 Rod 124 Deg 28 Rod 204 Deg 20 Red 240 Deg 29 Rod 340 Deg 20 Rod 309 Deg 28 Rad AA-64 AA-64 21 Deg 50 Rad 2 Deg 45 Rod 3 Deg 45 Rod Normal Operating State Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functionat Desimd Operating State Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Functional Walkdown or Walk-by Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Walkd-wn Walkdown Walkd-wn Walkdown Walkdown Walkdown Walkdown Walkdoun Walkdon Walkdown Walkdown Walkdown Walkdown Walkdown Walkdown Screening Notes HCLPF' TED TRD TED TED

>RLGMV

>RLGM TED TED TED TED TED

>RLGM RLGM RLGM

>RLGM

>RLGMV

>RLGM

>RLGM TED TED TED TED TED TED TED TED

>RLGM RLGIM TED TED TED Key Failure Mode-a TED TED TED TED Screened per EPRI NP-6041 Screened per EPRI NP-6041 -

TED TED TED TED TED Screened per EPRI NP-6041 Screened per EPRI NP-6041 Screened per EPRI NP-6041 Screened per EPRI NP-6041 Interaction - Block Wails Interaction - Block Walls Interaction - Block Walls TED TED TED TED TED TED TRD TED Screened per EPRI NP-6041 Screened per EPRI NP-6041 TED TED TED Page 51 of 61 Page 51 of 61

Expedited Seismic Evaluation Process (ESEP) Report, Catawba Nuclear Station Rev. 1 Catawba Nuclear Station Unit 2 ESEL and HCLPF Results ESEL Normal Desired Walkdown or ID EIN Description Bldg EL Location Operating State Operating State Walk-by Screening Notes HCUPF 5

Key Failure Mode-243 2NILT5263 Containment Sump Level RX2 566 2 Deg 54 Rod Functional Functional Walkdown TBD TBD 244 2NILT5264 Containment Sump Level RX2 570 18 Deg 56 Rod Functional Functional Walkdown TBD TBD 245 2NIMT5260 Containment Sump Level AUX 577 BB-67 Functional Functional Walkdowt

>RLGM Screened per FPRI NP-6041

- HCOPF saoues of >RLGM indicate that the HO.PF exceeds the Review Level Ground Motion (0.29g), but that a specific HCOPFF aue was not calculated since the component was screened out from further evaluation.

Key Failure Modes are defined as follows Screened per IPEEE

- Indicates that the component man evaluated in the IPEEE and therefore meets the RUGM demand.

Screened per EPRI NP-6041 -Indicates that the co mponent meets the screening criteria of EPRI N P-6041, Table 2.4 and that neither a nchorage, relay chatter, nor interact ions limit the reported HCLPF.

Interaction - Block Walls - Indicates that the component Is located near a block mail. The block mail mas evaluated in IPEEE and therefore the bhlck mall meets the RLGM demand. The fu nt ional and anchorage HCROFs exceed the reported HCLPF calue.

Interaction - Control Room Ceiling - Indicates that the component is located in the control room. The control room ceiling mas evaluated In this report and has a HCOPF of 0.30g. The functional and anchorage HCOPFs exceed the reported HCOPF value.

Anchorage -Indicates that anchorage is the governing failH re mode for the component.

Functional Failu re - Indicates that functional failure is the governing failu re mode for the component.

Relay Chatter -Indicates that relay chatter is the goneming failure mode for the component.

Modificatfon/Investlgation-Indicates that the reported HOFPF calue is conditional on the modification and/or further investigation as reported on the S EWS.

Total Items:

245 Page 52 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 Appendix C CNS FLEX Flow Paths Page 53 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 WATER CHEMISTRY BLUG.

LI-I MEOICAL FACILITY PARK=NG U*5 ENTIW4EE CONMORD ROAD PORTABLE PUMP SUPPLY TO ESSENTIAL SERVICE WATER HEADER / STORAGE FACILITY LOCATION Page 54 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 WATER CHEMISTRY SLOG.

MEDICAL FACILITY PARIUNG PARKING PARKING CONCORD ROAD PORTABLE PUMP SUPPLY TO SG's Page 55 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 AUXILIARY FEEDWATER (SUPPLY AND ALTERNATE SO MAKEUP)

Page 56 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 Rev. 1 IMS-0" OW-1974-MLI] I 34

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NUCLEAR SERVICE WATER (SUPPLY USING PORTABLE PUMP)

Page 57 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 PRIMARY INJECTION Page 58 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 I OMk% CmW-iWj-N[L8I a

C It Vi CaAWA NUCLEAR STATION RESIOUý HEATL REMOVL SYSEM(PAS03 Page 59 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1 UPPER CONTAINIvENT

~>REACTOR BUILDING ARF ARF Page 60 of 61

Expedited Seismic Evaluation Process Report, Catawba Nuclear Station Rev. 1

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