Text

R02 15-5462, Revision a, Expedited Seismic Evaluation Process (ESEP) Report
	ML16028A319
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Site:	Columbia
Issue date:	10/27/2015
From:	Guerra E, Lee L; Rizzo Associates
To:	; Office of Nuclear Reactor Regulation
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ML16028A316	List: GO2-16-021, R02 15-5462, Revision a, Expedited Seismic Evaluation Process (ESEP) Report; ML16028A318;
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EXPEDITED SEISMIC EVALUATION PROCESS (ESEP)

REPORT COLUMBIA GENERATING STATION 76 NORTH POWER PLANT LOOP RICHLAND, WASHINGTON REPORT NO. R02 15-5462 REVISION A OCTOBER 27, 2015 RIZZO ASSOCIATES 500 PENN CENTER BOULEVARD PENN CENTER EAST BUILDING 5, SUITE 100 PITTSBURGH, PA 15235 TELEPHONE: (412) 856-9700 TELEFAX: (412) 856-9749 WWW.RIZZOASSOC.COM R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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APPROVALS Project No.:

Report Name:

15-5462 Expedited Seismic Evaluation Process (ESEP) Report Columbia Generating Station Date:

December 17, 2015 Revision No.:

0 Approval by the responsible manager signifies that the document is complete, all required reviews are complete, and the document is released for use.

Originators:

Eddie M. Guerra, Senior

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Project Engineer, RIZZO

""*"'*"Associates Eddie M. Guerra, P.E. (RIZZO)

Lawrence K. Lee (ERIN)

December 17.7 2015 Date December 17. 2015 Date Reviewers:

Habib Shtaih (Energy Northwest)

Steve Sheahan (Energy Northwest)

)J ~4~Va4 Drgzt1Iysdgned by Nshlkant H.

Valoye V.P AIdvLancdEgnelag*

Date,: 201$,12.16 16'38.40 -05080' Nishikant R. Vaidya, Ph.D., P.E. (RIZZ.O)

Vincent Andersen (ERIN)

December 17. 2015 Date December 1 7. 2015 Date December 17. 2015 Date December i17. 2015 Date Shannon Kinnunen (Energy Northwest)

Approved by:

December 17, 2015 Date R02 Columbia Generating Station ESEP 155462/15, Rev. 0 (December 17, 2015)

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CHANGE MANAGEMENT RECORD Project No.:

Report Name:

Revision No.:

15-5462 Expedited Seismic Evaluation Process (ESEP) Report Columbia Generating Station Richland, Washington A

REVISIONI N.DATE DESCRIPTIONS OF CHANGES/AFFECTED PAGES A

JOctober 27, 2015 JFor Review 4

1 4

4 4

4 R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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TABLE OF CONTENTS PAGE LIST OF TABLES.............................................................................. 7 LIST OF FIGURES.............................................................................

8 LIST OF ACRONYMS........................................................................

9 1.0 PURPOSE AND OBJECTIVE........................................................ 12 2.0

SUMMARY

OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES.......................................................................... 14 3.0 EQUIPMENT SECTION PROCESS AND EXPEDITED SEISMIC EQUIPMENT LIST.................................................................... 16 3.1 EQUIPMENT SELECTION PROCESS AND ESEL................................ 16 3.1.1 ESEL Development................................................. 17 3.1.2 Power-Operated Valves............................................. 18 3.1.3 Pull Boxes............................................................ 18 3.1.4 Termination Cabinets............................................... 19 3.1.5 Critical Instrumentation Indicators.........."........................ 19 3.1.6 Phase 2 and Phase 3 Piping Connections.......................... 19 3.1.7 Relays................................................................ 19 3.2 JUSTIFICATION FOR USE OF EQUIPMENT THAT IS NOT THE PRIMARY MEANS FOR FLEX IMPLEMENTATION........................................... 20 4.0 GROUND MOTION RESPONSE SPECTRUM (GMRS)......................... 21 4.1 PLOT OF GMRS SUBMVITTED BY THE LICENSEE................................ 21 4.2 COMPARISON TO SSE................,.......................................... 22 5.0 REVIEW LEVEL GROUND MOTION (RLGM).................................. 24

5.1 DESCRIPTION

OF RLGM SELECTED.............................................. 24 5.2 METHOD TO ESTIMATE IN-STRUCTURE RESPONSE SPECTRA................. 25 5.3 REVIEW OF EXISTING MODELS................................................... 25 6.0 SEISMIC MARGIN EVALUATION APPROACH................................ 27 6.1]

SUMMARY

OF METHODOLOGIES USED....................................

[...... 27 6.2 HCLPF SCREENING PROCESS.................................................... 28 6.3 SEISMIC WALKDOWN APPROACH................................................ 30 6.3.1 Walkdown Approach................................................ 30 R02 Columbia Generating Station ESEP

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TABLE OF CONTENTS (CONTINUED)

PAGE 6.3.2 Seismic Review Team Qualifications............................. 31 6.3.3 Application of Previous Walkdown Information................. 33 6.3.4 Summary of Walkdown Findings.................................. 33 6.4 HCLPF CALCULATION PROCESS................................................. 34 6.4.1 CDFM Approach.................................................... 35 6.4.2 Component Structural Capacity.................................... 36 6.4.3 Functional Evaluations.............................................. 36 6.5 FUNCTIONAL EVALUATIONS OF RELAYS........................................ 37 6.6 RESOLUTION OF WALKDOWN FINDINGS...................................... 38 6.7 TABULATED ESEL HCLPF VALUES (INCLUDING KEY FAILURE MODES)......................................................................... 40 7.0 INACCESSIBLE COMPONENTS................................................... 41 7.1 IDENTIFICATION OF ESEL ITEMS INACCESSIBLE FOR WALKDOWNS........ 41 7.2 PLANNED WALKDOWN / EVALUATION SCHEDULE / CLOSE OUT............ 43 8.0 ESEP CONCLUSIONS AND RESULTS............................................ 44 8.1 SUPPORTING INFORMATION....................................................... 44 8.2 IDENTIFICATION OF PLANNED MODIFICATIONS.................................. 45 8.3 MODIFICATION IMPLEMENTATION SCHEDULE................................... 45 8.4

SUMMARY

OF REGULATORY COMMITMENTS.................................... 46

9.0 REFERENCES

......................................................................... 47 ATTACHMENTS:

ATTACHMENT A:

EXPEDITED SEISMIC EQUIPMENT LIST (ESEL)

ATTACHMENT B:

SUMMARY

OF ESEL HCLPF VALUES ATTACHMENT C:

WALKDOWN TEAM QUALIFICATIONS R02 Columbia Generating Station ESEP

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LIST OF TABLES TABLE NO.

TABLE 4-1 TABLE 4-2 TABLE 6-1 TABLE 6-2 TABLE 6-3 TABLE 7-1 TITLE PAGE 5% DAMPED UHRS FOR 10O4 ANi0-O AZR LEVELS AND GMRS AT CONTROL POINT FOR THE CGS SITE............................................................. 22 SSE (5% DAMPING) FOR CGS.................................... 23

SUMMARY

OF SCREENED-OUT COMPONENTS............ 29

SUMMARY

OF CONSERVATIVE DETERMINISTIC FAILURE MARGIN APPROACH..................................35

SUMMARY

.OF ESEP WALKDOWN FINDING RESOLUTIONS...................................................... 38 CGS ESEL ITEMS INACCESSIBLE DURING WALKDOWNS....................................................... 41 R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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LIST OF FIGURES FIGURE NO.

FIGURE 4-1 FIGURE 4-2 FIGURE 5-1 TITLE PAGE 5% DAMPED UHRS FOR 10-4AN 10-5 HAZARD LEVELS AND GMRS AT CONTROL POINT FOR THE CGS SITE............................................................... 21 GMRS VS SSE FOR CGS SITE...................................... 23 RLGM VS 2X SSE FOR CGS ESEP EVALUATION............. 24 R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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LIST OF ACRONYMS ACRONYM AC ACI ADS AFW AISC AOV ASCE ASME BE BDBEE CDFM CGS CIP CST DC DG EL ELAP EPRI ESEL ESEP FLEX FSAR ft g

GERS GIP GMRS HCLPF TITLE Alternate Current American Concrete Institute Automatic Depressurization System Auxiliary Feed Water American Institute of Steel Construction Air-Operated Valve American Society of Civil Engineers American Society of Mechanical Engineers Best Estimate Beyond Design-Basis External Event Conservative Deterministic Failure Margin Columbia Generating Station Cast-In-Place Condensate Storage Tank Direct Current Diesel Generator Elevation Extended Loss of all AC Power Electric Power Research Institute Expedited Seismic Equipment List Expedited Seismic Evaluation Process Diverse and Flexible Coping Strategies Final Safety Analysis Report Feet Gravity Generic Equipment Ruggedness Data

Generic Implementation Procedure Ground MotionResponse Spectra High Confidence of Low Probability of Failure R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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LIST OF ACRONYMS (CONTINUED)

ACRONYM HSS HVAC Hz IEEE IPEEE ISRS LPCI MAFE MCC MOV NEI NPP NRC NTTF GIP PGA PRA PSA psi P&IDs RCIC RHR RLGM RPV SEI SEWS SI SMA SME SPRA SQUG TITLE Hollow Steel Section Heating, Ventilation, and Air Conditioning Hertz Institute of Electrical and Electronics Engineers Individual Plant Evaluation of External Events In-Structure Response Spectra Low Pressure Coolant Injection Mean Annual Frequency of Exceedance Motor Control Center Motor-Operated Valve Nuclear Energy Institute Nuclear Power Plant Nuclear Regulatory Commission Near-Term Task Force Overall Integrated Plan Peak Ground Acceleration Probabilistic Risk Assessment Probabilistic Safety Assessment Pound per Square Inch Process and Instrumentation Diagrams Reactor Core Isolation Cooling Residual Heat Removal Review Level Ground Motion Reactor Pressure Vessel Structural Engineering Institute Seismic Screening and Evaluation Work Sheets Seismic Interaction Seismic Margin Assessment Seismic Margin Earthquake Seismic Probabilistic Risk Assessment Seismic Qualification Utilities Group R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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LIST OF ACRONYMS (CONTINUED)

ACRONYM SRT SRV SSC SSE SSHAC SW TRS UHRS U.S.

V Vs TITLE Seismic Review Team Safety/Relief Valve Structures, Systems, or Components Safe Shutdown Earthquake Senior Seismic Hazard Analysis Committee Service Water Test Response Spectrum Unifonrm Hazard Response Spectra United States of America Volts Shear Wave Velocity R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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EXPEDITED SEISMIC EVALUATION PROCESS (ESEP)

COLUMBIA GENERATION STATION RICILLAND, WASHINGTON 1.0 PURPOSE AND OBJECTIVE Following the accident at the Fukushima Dai-ichi Nuclear Power Plant (NPP) 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 the 50.54(f) letter on March 12, 2012 [1], requesting information to assure that these recommendations are addressed by all United States (U.S.) NPPs. The 50.54(f) letter 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. Risk 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 the Columbia Generating Station (CGS). 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 NTTF Recommendation 2.1: Seismic [2].

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The objective of this report is to provide summarY infor-mation 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.

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2.0

SUMMARY

OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES The CGS FLEX strategies to maintain Core Cooling and to maintain Containment are summarized below. The Spent Fuel Pool Cooling strategies are not described because they are not included within the scope of the ESEP process. The CGS FLEX strategy summary is derived from the Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order BA-12-049 [3] including the required 6 month updates that have been provided since the OIP was submitted.

The Phase 1 FLEX strategy relies on installed plant equipment. Reactor Core Cooling is achieved via operation of the Reactor Core Isolation Cooling (RCIC) system with injection from the suppression pool water source to the Reactor vessel. Although RCIC suction is normally aligned to Condensate Storage Tank (CST), the OIP assumes that the non-seismically Category 1 CST would be unavailable due to the initiating event (e.g., beyond design-basis seismic event) and the RCIC pump suction is automatically realigned to the suppression pool based on instrumentation sensing low CST level. Reactor vessel cooldown is achieved using RCIC and manual operation of one (1) Safety/Relief Valve (SRV) from the Automatic Depressurization System (ADS) that discharges to the suppression pool. Containment Control during Phase 1 will be maintained by the to-be-installed hardened containment vent. The design of the hardened containment vent system will be described in the Integrated Plan required by NRC Order EA 109. Anticipatory wetwell venting (i.e., "early" containment venting) will be proceduralized using the hardened containment vent to relieve pressure, control suppression pool water temperature, and enable continued cooling of the Reactor via RCIC operation. Key Reactor and Containment parameters required to support Core Cooling and Containment are monitored via DC powered instrumentation. A DC load shed strategy is implemented to extend battery life.

The Phase 2 FLEX strategy relies on installed plant equipment an portable on-site equipment.

The strategy for Phase 2 Core Cooling assumes that the RCIC system is maintained available.

The Reactor Pressure Vessel (RPV) pressure is reduced to 175-3 00 psig to maintain RCIC operation. RCIC will continue to operate to provide RPV injection during Phase 2. Given the proposed FLEX strategy for early containment venting, suppression pool inventory will decrease as steam from the suppression pool vents to the atmosphere. Thermal hydraulic calculations support that suppression pool makeup would be required by approximately 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months after the initial event. Suppression pool makeup can be supplied via one of two portable FLEX pumps R02 Columbia Generating Station ESEP I*

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with suction from the Service Water (SW) spray ponds aligned for discharge to the residual heat removal (RTIR) system piping. In addition, given that steam motive force for RPV makeup from RCIC is expected to significantly decrease after approximately 2-3 days, alternate RPV makeup will be performed via aligning a FLEX pump to the RHR C train. The same FLEX pump can be utilized to support both suppression pooi makeup and RPV makeup because the makeup requirements will be relatively small (e.g., few hundred gpm) at the delayed time when makeup is required.

Necessary electrical components are outlined in the CGS FLEX OIP submittal [3], including subsequent 6 month updates through August 2015, and primarily entail 480 VAC essential motor control centers, vital batteries, equipment installed to support FLEX electrical connections, and monitoring instrumentation required for Core Cooling, Reactor Coolant Inventory, and Containment Integrity. Given success of the DC load shed procedure, the CGS 125 VDC and 250 VDC station batteries are available for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months without recharging. CGS maintains a portable, trailer-mounted FLEX 480V 400 kW diesel generator (i.e., DG4) that is normally staged in the yard next to the Diesel Generator Building. Permanently installed connections on the outside of the Diesel Generator Building to support either the Division 1 or 2 of 480 VAC distribution systems allow an expedited transition to the Phase 2 mitigation strategy. An additional FLEX Diesel Generator (i.e., DG5) is available for alignment to either the Division 1 or 2 of 480 VAC distribution systems as a back-up to DG4 using the same permanently installed connections on the outside of the Diesel Generator Building.

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3.0 EQUIPMENT SECTION PROCESS AND EXPEDITED SEISMIC EQUIPMENT LIST The selection of equipment for the Expedited Seismic Equipment List (ESEL) followed the guidelines of EPRI 3002000704 [2]. The ESEL for CGS is presented in AttachmentA.

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 (BDBEE), as Outlined in the CGS OIP in Response to the March 12, 2012, Commission Order EA-12-049 submitted in February, 2013 [3] including subsequent 6 month updates through August 2015. The OIP provides the CGS 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 CGS OIP [1] including subsequent 6 month updates through August 2015. 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, 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] as follows:

1.

The scope of components is limited to that required to accomplish the core cooling and containment safety functions identified in Table 3-1 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 CGS OIP [3] including subsequent 6 month updates through August 2015.

2.

The scope of components is limited to installed plant equipment and FLEX connections necessary to implement the CGS OIP [1] including subsequent 6 month updates through August 2015 as described in Section 2.0O.

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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 check 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.)

3.1.1 ESEL Development The ESEL was developed by reviewing the CGS OIP [3], including subsequent 6 month updates through August 2015, 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 performed to identify the boundaries of the flow paths 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., circuit breaker, valve, etc.) in branch circuits / branch lines off the defined strategy electrical or fluid flowpath.

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., as necessary.

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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 CGS 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.

Based on a Nuclear Energy Institute (NET) report [4] that compiled pertinent "Questions and Answers" related to the EPRI 3002000704 [2] methodology, manually Operated MOVs and AOVs required for success of the FLEX strategies in Tables 3-1 of EPRI 3002000704 [2] "should be included on the ESEL to ensure they receive the proper interaction reviews during the walkdowns. Some MOV and AOVs have had failures during earthquakes linked to seismic interactions. Several failures occurred where the valve yoke broke due to valve operator impact with walls or beams or other more rugged components during the quake. Extended cast iron operators may be particularly susceptible to these interaction concerns." Therefore, MOVs and AOVs that are locally, manually operated for success of the FLEX strategies are included on the ESEL.

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].

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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 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 CGS OIP [3] including subsequent 6 month updates through March 2015 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. 0 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 and 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 path are included in the ESEL.

3.1.7 Relays As discussed in Section 3.1.2 above, Page 3-2 of EPRI 3002000704 [2] 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 from an electrical component R02 Columbia Generating Station ESEP L

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perspective, the following guidance is applied in the CGS ESEL for functional failure modes associated with electrical components:

Based on an NEI report [4], "power operated valves not required to change state are excluded from the ESEL is because past analyses have consistently showed that they are inherently seismically robust. This is not necessarily the case for electrical components, so they should be included on the ESEL."

Therefore, circuit breakers required to support supply power, whether or not they are required to change state, are included on the ESEL.

Include on the ESEL all relays associated with SRV solenoid valves that could chatter and potentially cause inadvertent SRV actuation to sufficiently depressurize the RPV and preclude adequate RCIC injection.

Relays associated with automatic initiation of the RCIC system are not included on the ESEL since only manual start is credited. The CGS operators are trained to either manually initiate RCIC before the auto-initiation signal occurs on Low-Low RPV water level, or the operators will manually initiate RCIC if the auto-initiation signal occurred, but RCIC did not auto-start. In addition, the operators will be able to override any spurious actuation due to potential relay chatter early in the event (and any such override would be included as part of the manual start action).

3.2 JUSTIFICATION FOR USE OF EQUIPMENT THAT IS NOT THE PRIMARY MEANS FOR FLEX IMPLEMENTATION The complete ESEL for CGS is presented in Attachment A. The format of the ESEL summary table for CGS is consistent with the example ESEL format provided in EPRI 3002000704 [2]1.

There are 165 individual line items on the CGS ESEL. The ESEL is based on the primary means for FLEX implementation.

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4.0 GROUND MOTION RESPONSE SPECTRUM (GMRS) 4.1 PLOT OF GMRS SUBMITTED BY THE LICENSEE The CGS major structures are founded on compacted granular structural backfill at foundation elevations varying between 406 ft-3 inches for the Reactor Building to 441 ft-0 inches for the Diesel Generator Building. The design-basis analysis applies the Safe Shutdown Earthquake (SSE) ground motion at the respective building foundations. Based on the Seismic Hazard and Screening Report for CGS [5], the control point elevation for the Ground Motion Response Spectra (GMRS) is at the surface of the finished grade (El. 441 ft).

Figure 4-1 presents the Uniform Hazard Response Spectra (UHRS) and GMRS at the control point (EL 441 ft) developed in [5]. Table 4-1 presents the spectral accelerations at selected frequencies for the UHRS for 10-4 and 1 0. hazard levels and GMRS.

2.5 41.5

,I-0.5 6UHRS~10-5!

UHRS 10-4 0.1 1

10 100 Frequency [Hz]

FIGURE 4-1 5% DAMPED UHRS FOR 1 0 -4 AND 10-s HAZARD LEVELS AND GMRS AT CONTROL POINT FOR THE CGS SITE R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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TABLE 4-1 5% DAMPED UHRS FOR 10-4 AND 10-s HAZARD LEVELS AND GMRS AT CONTROL POINT FOR THE CGS SITE HORIZONTAL SPECTRAL ACCELERATION (g) AT THE CONTROL FREQUENCY POINT FOR CGS (Hz) lxl04 MAFE UHRS 1Xl0"5 MAFE UHRS GMRS 100.000 0.2484 0.4288 0.2484 50.000 0.295 1 0.5057 0.295 1 33.333 0.3471 0.6242 0.3471 25.000 0.3916 0.7238 0.3916 20.000 0.3595 0.6537 0.3595 13.333 0.4341

0. 8088 0.4341 10.000 0.4978 0.9638

0. 5067 6.667 0.7427 1.4240 0.7501 5.000 1.2160 2.4340 1.2711 3.333 1.3236 2.8030 1.4474 2.500 0.7958 1.7767 0.9078 2.000 0.7360 1.7620 0.8878 1.333 0.5313 1.3565 0.6748 1.000 0.3781 0.9234 0.4634 0.667 0.3089 0.7 104 0.3609 0.500

0. 1851 0.4552 0.2281 0.333 0.0837

0. 1917 0.0974 0.200 0.0435 0.0912 0.0472

0. 133 0.0262 0.0540 0.0280 0.10O0 0.0196 0.0397 0.0207 Note:

MAFE = mean annual frequency of exceedance.

4.2 COMPARISON TO SSE Figure 4-2 compares the GMRS [5] with the site SSE at the control point elevation. (Table 4-2 provides the CGS SSE in tabular form for selected points.) The SSE horizontal spectrum is characterized by a Peak Ground Acceleration (PGA) of 0.25 acceleration of gravity (g) and a shape that conforms to the Newmark-Hall Spectrum Shape. Figure 4-2 illustrates that the maximum GMRS/SSE ratio between 1 and 10 Hz range is about 2.4 and occurs at 3.33 Hz.

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TABLE 4-2 SSE (5% DAMPING) FOR CGS FREQUENCY SPECTRAL (Hz)

ACCELERATION (g) 0.40

0. 12 2.05 0.60 6.10 0.60 18.90 0.25 100.00 0.25 1.6 1.4 0.8 ii 0.6 0.2 0

0.1 1

10 100 Frequency [Hz]

FIGURE 4-2 GMRS VS SSE FOR CGS SITE R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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5.0 REVIEW LEVEL GROUND MOTION (RLGM)

5.1 DESCRIPTION

OF RLGM SELECTED The Review Level Ground Motion (RLGM) is a spectrum representing the seismic demand level for which the ESEP margin evaluation is conducted. The RLGM is selected based on Criteria 1 of Section 4 of the EPRI Seismic Evaluation Guidance: "Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1 - Seismic" [2] which recommends to derive the RLGM by linearly scaling the SSE by a maximum ratio of the GMRS/SSE between 1 and 10 Hertz (Hz) range (not to exceed 2 x SSE). The SSE-based In-Structure Response Spectra (ISRS) at the level of the ESEL equipment will also be scaled by the same factor.

The maximum GMRS/SSE ratio between I and 10 Hz range is 2.4 and occurs at 3.33 Hz. Since this ratio exceeds the maximum value of 2, the RLGM will be derived by linearly scaling the SSE by a factor of 2 as shown on Figure 5-1. Note that the RLGM PGA is 0.50g. Seismic capacities for ESEP equipment will be compared against the PGA of the RLGM.

1.6 1.4

-* 1.2 W0.8 0.6 0.2 0

1 10 100 0.1 Frequency [Hz]

FIGURE 5-1 RLGM VS 2X SSE FOR CGS ESEP EVALUATION R02 Columbia Generating Station ESEP 155462/1 5, Rev. A (October 27, 2015)

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5.2 METHOD TO ESTIMATE IN-STRUCTURE RESPONSE SPECTRA The SSE-based ISRS are linearly scaled by a maximum scale factor of 2.0. Scaling of ISRS was performed for CGS building elevations where representative ESEL components are located.

5.3 REVIEW OF EXISTING MODELS The structural models representing safety-related structures should be capable of capturing the overall structural response for both the horizontal and vertical components of ground motion. A review of existing lumped mass models (stick models) and finite element models was performed for CGS structures housing ESEL components.

According to CGS Final Safety Analysis Report [6] and the Reactor Building model calculation

[7], the Reactor Building and other Seismic Category I structures were modeled as a 3-D system of lumped masses and springs idealizing both the inertia and the stiffness properties of the structure. The model included lumped masses at each floor and at points considered of critical interest (i.e., at supports/anchors for equipment and systems). The lumped masses contain the weight of walls, floor slabs, water, heavy equipment, and systems that are mounted on or hanging from the floors. These lumped masses were connected by weightless linear elastic springs which account for the axial, flexural and shear stress effects of the structure by including structural properties such as cross-sectional areas, effective shear areas, and moments of inertia.

Stiffness parameters were modified accordingly to account for openings on walls and floors.

Torsional effects were also considered by applying a twisting moment about the center of rigidity of the floor under consideration.

Another aspect to evaluate in the lump mass model is the cracking of the concrete. ASCE 4-13

[8] states significant cracking of the concrete occurs when the average shear stress state exceeds 3x *1(fc). According to the Reactor Building model calculation [7] the average applied shear stress, scaled for a 2x SSE demand, is 197.3 pounds per square inch (psi). This value is compared to 3x */4000psi =189.7 psi. Since the applied shear stress is greater than 3x */(fc), for 2x SSE loading the concrete sections are probably cracked and it is reasonable to consider a Response Level 2. Therefore, a damping ratio of 7% can be adopted (as in the original analyses).

It is likely that the stiffness of some concrete walls should be reduced to about 50% of the uncracked value. In the extreme case that all the member stiffness is reduced, the frequency will be correspondingly reduced to about 70%. The fundamental frequency would reduce from 3.3 R02 Columbia Generating Station ESEP

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Hz to about 2.3 Hz. Since the shape of the SSE spectrum shows the same spectral value for both frequencies, no substantial changes will occur in the seismic demand.

Typical lumped mass systems are known for the rigid slab assumption where the model ignores the floor's in-plane and out-of-plane deformations. In lieu of updating the existing models to assess the effect of vertical flexibility of floors/diaphragms, a review of the modeling assumptions for floor slabs supporting ESEL components will be conducted on a case by case basis. Potential vertical amplification will be assessed via conservative assumptions in the margin calculation.

The level of conservatism typically employed in design-basis analyses may exceed the criteria for seismic margin assessments. For instance, parameter variations in design-basis models are generally incorporated by enveloping the full range of possible ISRS. Section 3.7.2.9 of the FSAR [6] states that parameter variations in design-basis floor spectra is considered by broadening the resonant peaks by +/- 15%. Such enveloping creates a broad frequency content envelope in-structure spectra which contains more power than could possibly be produced by a RLGM earthquake and introduces considerable conservatism which is not desirable for a SMA.

For seismic margin evaluations, it is preferable to shift the resonant peaks between 10% and 15%

rather than peak broadening. Potential cases of excessive conservatism induced by broadening rather than peak shifting will be assessed on a case by case basis when performing subsequent margin evaluations.

A higher RLGM will result in soil degradation for which the existing design-basis SSI models need to be reviewed to ensure linear scaling will adequately characterize a higher seismic response. Although it is expected that damping and shear wave velocity from a higher RLGM input should fall within the lower and upper bound soil profiles, a case by case evaluation may be performed to assess potential high soil non-linearity. In case it is deemed necessary to assess the effects of considerable soil degradation for a particular structure, the strain compatible shear wave velocity (Vs) and damping will be estimated by using the strain dependent shear modulus degradation and damping curves for cohesionless soils presented in [9].

In summary, a general review of the existing design-basis Reactor Building model shows that the effects of concrete cracking and soil degradation resulting from the RLGM should be encompassed by the parameter variation specified in the design-basis criteria. However, further review may be performed for cases where non-linear effects due to a higher RLGM are judged to fall outside the property variation limits.

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6.0 SEISMIC MARGIN EVALUATION APPROACH 6.1

SUMMARY

OF METHODOLOGIES USED The seismic margins for components on the ESEL are developed following the EPRI guidelines described in EPRI NP-6041 [10], EPRI TR-103959 (Methodology for Developing Seismic Fragilities) [11] and EPRI 1002988 (Seismic Fragility Application Guide) [12]. Additionally, EPRI 1019200 (Seismic Fragility Application Guide Update) [ 13] is used to develop margins using the Conservative Deterministic Failure Margin (CDFM) approach.

The ESEL is first grouped to identify similar components relative to equipment classes (e.g.,

Motor Control Centers, Horizontal Pumps, Control & Instrumentation Panels, etc.) and then sampled for representative items based on type of equipment, manufacturer, location, and anchorage, etc. Representative samples in each equipment group are then evaluated to obtain the seismic margins using the EPRI guidelines.

The overall strategy for developing seismic margins for the various structures, systems, or components (SSC) is as follows:

1.

Perform screening verification Walkdown to document that caveats associated with generic high confidence of low probability of failure (HCLPF) capacities are met and perform anchorage calculations.

2.

Develop the HCLPF capacities based on available experience data, published generic ruggedness spectra, design criteria documents, and design analysis.

3.

Perform improved analysis of selected equipment if necessary.

A number of components on the ESEL are breakers and switches that are housed in a "parent" component, such as a motor control center (MCC) or control panel. Walkdowns are performed for each of these housed components to confirm the mounting location. For the purposes of this evaluation, calculations will not be explicitly performed for these housed components. Instead, their HCLPF is assigned based on the parent component. An exception to this is relays mounted inside control panels, as relays will be evaluated separately.

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Seismic walkdowns, as described in EPRI NP-6041 [10], were performed for all components on the ESEL located outside Primary Containment. The components were walked down from August 31, 2015 to September 3, 2015.

Subsequent to the seismic walkdowns, thirteen components (13) were added to the ESEL (i.e.,

Items 153 through 165 in Attachment A). These components consist of override switches mounted on Control Room cabinets, pressure switches mounted on instrument racks, pressure switches mounted on the wall, and a MCC breaker.

CGS personnel provided detailed photographs of each of the additional components, and the photographs confirm that all additional components are either mounted on a component that was walked down by the Seismic Review Team (SRT) or located in the same area as components walked down by the SRT. Therefore these components are grouped by similarity with other ESEL components for the HCLPF evaluations.

HCLPF calculations are performed for all "parent" components to determine their structural and functional seismic capacity relative to the RLGM for CGS.

6.2 HCLPF SCREENING PROCESS A total of nineteen (19) components on the CGS ESEL were screened out during the seismic walkdowns. These components are judged to have a seismic capacity that is well above the RLGM, so no specific seismic capacity evaluation will be performed for them.

The justification for screening is documented in the individual Seismic Screening and Evaluation Work Sheets (SEWS) for each component. The screened components fall into one of three types of components, as described below. Table 6-1 lists each of the screened components and identifies the reason it is screened out from specific evaluation.

Wall-Mounted Instruments - These components are typically lightweight transmitters, switches, or emitters that are ruggedly mounted to the wall. The natural frequency of these mounting configurations is high (i.e., >33 Hz),

therefore no dynamic amplification is expected relative to the floor acceleration. In addition, GERS for switches and transmitters provide lower bound functional capacities at ZPA about 1.3g and 4g respectively. Based on the light weight and the relative low seismic demand, seismic capacity is judged to be well above RLGM.

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Passive wall-mounted electrical boxes - These components are typically lightweight splice boxes or terminal boxes that are rigidly mounted to the wall. These boxes are included on the ESEL because wiring is routed through them, but they do not perform any active electrical functions. Therefore their seismic HCLPF capacity is controlled by failure of the anchorage, which is judged to be above the RLGM level.

Motor-Operated Valves (MOVs) - Walkdowns indicate that most MOVs are mounted on well-supported, large-diameter lines, have robust yokes, and valve operators fall within the caveats for weight and eccentricity for the earthquake experience database. MOVs that meet these criteria are screened out from further evaluation, as EPRI NIP-6041 [10] and previous SPRA experience indicate that MOVs have very high seismic capacities. As identified in Section 6.3.4, two MOVs on the CGS ESEL do not fall within the earthquake experience database and are, therefore, not screened out.

TABLE 6-1 OF SCREENED-OUT COMPONENTS

SUMMARY

ESEL ID COMPONENT ID DESCRIPTION IREASON SCREENED OUT 55Wetwell Wide Range Level Wall-Mounted CSL6A Monitor Instrument 56 SPTM-TE-1A Suppression Pool Temperature Wall-Mounted

______________Element Instrument RCIC P-i Suction Pressure Wall-Mounted 17RI-T5 Transmitter Instrument RCIC High Exhaust Pressure Wall-Mounted 15 CCP-A switch RCIC-PS-9A Instrument 159RCI-PS9B RCIC High Exhaust Pressure Wall-Mounted

_________switch RCIC-PS-9B Instrument Division 1 FLEX DG Passive wall-mounted 65 B-1 40 Connection Point (480 VAC) electrical box Passive wall-mounted 116 E-DISC-S11iD1 Fusible Safety Switch eetia o

Terminal Box for Battery E-Passive wall-mounted 10-TB1l B 1-1 electrical box E-T-B/l Terminal Box for Battery E-Passive wall-mounted 11B2-1 electrical box Terminal Box outside Passive wall-mounted 19TR22Penetration E-X-105C electrical box 10T-52Terminal Box inside Passive wall-mounted Penetration E-X-105C electrical box 151 SB-W020 Splice Box Passive wall-mounted R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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SUMMARY

TABLE 6-1 OF SCREENED-OUT COMPONENTS (CONTINUED)

ESEL ID COMPONENT ID DESCRIPTION

[REASON SCREENED OUT electrical box RCIC Turbine Trip Throttle Well-supported motor-18RI--1 Valve operated valve RCIC Turbine Steam Stop Well-supported motor-19RI--5 Valve operated valve

CCV.RCIC Head Spray Line Well-supported motor-213 Injection Valve operated valve

24RCI-V-10 RIC ST uctin Vlve Well-supported motor-24 RIC-V 10 RCICCST ucton Vlve operated valve RCIC Suppression Pool Well-supported motor-2RCCV31 Suction Valve operated valve 35*

RHR-V-42C LPCI Loop C Injection Valve Well-supported motor-operated valve LPCI C Suppression Pool Test Well-supported motor-36RRV21 Return Line operated valve Note:

Although these valves are located 40 ft above effective grade of the Reactor Building, they possess a robust seismic configuration. Hence their seismic HCLPF capacity is determined to be above 0.50g PGA.

6.3 SEISMIC WALKDOWN APPROACH 6.3.1 Walkdown Approach The seismic walkdowns of CGS were performed in accordance with the criteria provided in Section 5 of EPRI 3002000704 [2], which refers to EPRI NP-6041 [10] for the SMA process.

The procedures used for different equipment categories are summarized below.

The SRT reviewed equipment on the equipment walkdown list that were reasonably accessible and in nonradioactive or moderately radioactive environments. For components in high radioactive environments, a smaller team and more expedited reviews were employed. For components that were not accessible, the equipment inspection relied on alternate means, such as photographs and plant qualification documents.

For each component, the SRT perfonned a thorough inspection and recorded information related to anchorage, load path configuration, and any potential seismic vulnerability associated to the component seismic capacity. These details recorded in SEWS were subsequently used to verify R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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as-built conditions and determine seismic HCLPF capacities. The 100 percent walkdown of all reasonably accessible components was performed to look for outliers, anchorage which is different from that shown on drawings or prescribed in criteria for that component, potential Seismic Interaction (SI) problems, situations that were at odds with the team members' past experience, and any other areas of serious seismic concern.

Walk-bys also served to provide the SRT with the sufficient degree of confidence in relation to plant maintenance and construction practices. This is especially used to reinforce the engineering judgment applied for the capacity assessment of inaccessible components.

As a result of the walkdowns performed for the CGS ESEL components, the SRT did not identify any significant seismic vulnerability concerns. Furthermore, the SRT observed that, in general, the areas containing ESEL equipment were well maintained and organized. This observation served as a basis for supporting subsequent assessments of inaccessible components.

For each item on the equipment walkdown list, a specific SEWS was prepared covering the different caveats. Each SEWS consists of:

General description of the equipment: Equipment ID, name, equipment category, and building/floor/room

Equipment evaluation caveats

Equipment anchorage

SI issues A database of SEWS was developed in an electronic format using iPad tablets to facilitate entry of the information collected during the walkdowns. The database includes the record of equipment qualifications, walkdown observations, and photographs. A separate walkdown notebook was developed to document the information from the database of SEWS.

6.3.2 Seismic Review Team Qualifications Walkdowns were completed by Mr. Adam [Ielffrich, P.E., Mr. Brian Lucarelli, E.I.T., and Mr.

Lawrence Lee, with assistance from CGS personnel. After walkdowns were complete, walkdown findings and capacity screening decisions were reviewed with Dr. Nish Vaidya, P.E.

A SUlmmary of the walkdown team qualifications is provided below.

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Mr. Helffrich has more than five years of experience in the nuclear field and is a licensed professional engineer in the state of Pennsylvania. Mr. Helffrich completed the five day Seismic Qualification Utilities Group (SQUG) Walkdown Screening and Seismic Evaluation training course. Mr. Helffrich completed walkdowns in support of ESEP evaluations and SPRA at three other NPP sites. Mr. Helffrich also completed walkdowns in support of NTTF Recommendation 2.3 Seismic at two NPP sites.

Mr. Lucarelli has more than five years of experience in the nuclear field and has completed the five day SQUG Walkdown Screening and Seismic Evaluation training course. Mr. Lucarelli completed walkdowns in support of ESEP evaluations and SPRA at four other NPP sites. Mr.

Lucarelli also completed walkdowns in support of NTTF Recommendation 2.3 Seismic at three NPP sites.

Mr. Lee has over twenty years of experience in the field of Probabilistic Safety Assessment (PSA) for nuclear power plants and other complex systems and facilities. Mr. Lee has experience in leading Level 1 and Level 2 PSA updates (internal and external events), shutdown safety assessment, and utility response to NRC compliance using PSA techniques. Mr. Lee has performed plant walkdowns, system fault tree analysis, accident sequence analysis, and probabilistic fragility analysis to support initial seismic PRA models for numerous nuclear power plants. He has also supported projects for the risk informed prioritization to perform plant-specific fragility calculations. Mr. Lee has been an instructor for the EPRI Seismic PRA training class. In addition, he has provided technical oversight for the update of the EPRI Seismic PRA Implementation Guide.

Dr. Vaidya has over 40 years of experience on a variety of civil and geotechnical engineering consulting projects. He is a recognized expert in the field of seismic isolation for nuclear power facilities. Dr. Vaidya also has experience with SSHAC Level 3 and 4 analysis. Dr. Vaidya has participated in the standards development activities of the AC, AISC, and ASME. Dr. Vaidya has performed numerous analyses related to the seismic and dynamic response of building structures as well as mechanical and electrical equipment including response spectrum and time history analysis, computation of floor response sPectra, equipment qualification to IEEE 344 standard, qualification of systems and components, evaluation of equipment supports and evaluation of anchorage and fragility analysis.

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6.3.3 Application of Previous Walkdown Information Documentation for two previous seismic walkdowns at CGS was reviewed: the Individual Plant Evaluation for External Events (IPEEE) walkdowns and the NTTF 2.3 Seismic Walkdowns.

The IPEEE final report shows that failure of the base connection of important MCCs represented the major contributing sequence to seismic risk at CGS. As a result, strengthening of the anchorage of several MCCs was recommended as part of the IPEEE improvements. The SRT, therefore, paid particular attention to the connection of MCCs included in the ESEL, and their base panels were opened to inspect the welds, verify IPEEE drawings of the connections, and create additional weld maps.

A portion of the NTTF 2.3 walkdowns were completed while the plant was in an outage; as such, numerous recent photographs (August 2013) are available of the interior of electrical cabinets that cannot normally be opened while the plant is in operation. The ESEP SRT used these photographs to confirm that interior components are well-mounted and in good condition, and that adjacent cabinets are bolted together to prevent pounding during a seismic event.

6.3.4 Summary of Walkdown Findings Consistent with the guidance from EPRI NP-6041 [10], no significant outliers or anchorage concerns were identified during the CGS seismic walkdowns. The following observations were noted during the seismic walkdowns:

A crack in the concrete floor was observed to run through one anchor bolt location for the Remote Shutdown Panel (E-CP-C6 1/P00 1). The anchorage evaluation for this component should neglect the capacity of this anchor and assume the load is taken by the remaining anchors, which were observed to be in good condition.

An installation issue was observed at one of the welds connecting E-MC-7A to its base channel. This issue was discussed with plant personnel during the walkdowns and Condition Report A/R 00337190 was issued to address this concern. The HCLPF anchorage evaluation for this component should neglect the capacity of this weld and assume the load is taken by the remaining welds, which were observed to be in good condition.

Multiple hollow steel section (HSS) supports for a cable tray run along the rear faces of MCC E-MC-7A and E-MC-Sl-lD. The gap between the HSS supports and the MCCs is approximately 1 inch. The capacity evaluation should evaluate this as a potential SI.

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Instrument rack E-IR-P0 17 is mounted near a concrete wall, and the gap to the wall is approximately 1 inch. The capacity evaluation should evaluate this as a potential SI.

Two valves (RCIC-V-l19 and RCIC-V-46) were identified not to meet the actuator weight-eccentricity caveats for the use of the 1.5 x bounding spectrum capacity. Specifically, these two valves are mounted on small (2 inch diameter) lines and require specific evaluation.

The RCIC Vacuum Tank (RCIC-TK-1) is covered in insulation and the load path could not be verified during walkdowns. Further evaluation of the load path for this tank is required.

Vital Battery Charger E-C2-1 is observed to be of an older vintage. Review of photographs of the component internals from the NTTF 2.3 walkdowns confirms that the battery charger is not a solid-state type. Use of generic seismic experience data is not appropriate for this component, so the seismic capacity analysis should consider plant-specific equipment qualifications for its functional capacity.

The gap between Battery Charger E-C2-1 and an adjacent HVAC support member is observed to be approximately 3/8 inch. The seismic capacity evaluation should evaluate this as a potential SI.

Each of the aforementioned observations has been fully addressed as part of the seismic HCLPF calculations. A summary of the approach taken to resolve each case is provided in Table 6-3 of Section 6. 6.

6.4 IICLPF CALCULATION PROCESS ESEL items for CGS were evaluated using the criteria in EPRI NP-604 1 [10]. Those evaluations included the following steps:

Performing seismic capability walkdowns for equipment to verify the installed plant conditions

Performing screening evaluations using the screening tables in EPRI NP-604 1

[10] as described in Section 6.2

Performing HCLPF calculations considering various failure modes that include both structural failure modes (e.g., anchorage, and load path, etc.) and functional failure modes All H-CLPF calculations were performed using the CDFM methodology and are documented in Reference [14].

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6.4.1 CDFM Approach HCLPF values for functionality and anchorage are calculated for each representative component selected from the ESEL. Representative components are selected and evaluated as a bounding case for each equipment group and their HCLPFs are assigned to all components within the group. The functional HCLPF for equipment is based on experience data, Generic Equipment Ruggedness Data (GERS), test response data, and design criteria. The functional evaluation is supplemented with the verification of the equipment anchorage following SQUG/GIP procedures. The seismic demand on the equipment is based on the floor response spectra near the equipment support location, and the component damping values as recommended in EPRI NP-6041 [ 10].

The CDFM approach described in EPRI 1019200 [ 13] is utilized to obtain the component HCLPF values. The HCLPF capacities are stated in terms of a selected ground motion PGA.

The CDFM approach is consistent with EPRI NP-6041I-SL [ 10], updated to accommodate the parameters presented in Table 6-2.

TABLE 6-2

SUMMARY

OF CONSERVATIVE DETERMINISTIC FAILURE MARGIN APPROACH (EPRI 1019200 [13], TABLE A.1)

TECHNICAL ISSUE RECOMMENDED METHOD Load Combination Normal + SME.

Groud Repons Spetrum Anchor CDFM Capacity to defined response spectrum shape without

__rondResponseSpetru_

consideration of spectral shape variability.

Perform seismic demand analysis in accordance with latest version of Seisic DmandAmerican Society of Civil Engineers (ASCE) 4.

Damping Conservative estimate of median damping.

Structural Model BE (Median) + Uncertainty Variation in Frequency.

Soil Structure Interaction BE (Median) + Parameter Variation.

In-Structure (Floor) Spectra Use frequency shifting rather than peak broadening to account for Generation uncertainty plus use conservative estimate of median damping.

Code specified minimum strength or 95% exceedance actual strength if Material Strength test data are available.

Code ultimate strength (ACI), maximum strength (AISC), Service Level D (ASME), or functional limits. If test data are available to Static Strength Equations demonstrate excessive conservatism of code equation then use 84%

exceedance of test data for strength equation.

For non-brittle failure modes and linear analysis, use appropriate Absortioninelastic energy absorption factor from ASCE/SEI 43-05 to account for Inelastic Energy Abopinductility benefits, or perform non-linear analysis and go to 95%

exceedance ductility levels.

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6.4.2 Component Structural Capacity In general, the CDFM approach:

1.

Develops the elastic seismic response for the structures and components for the ground motion.

2.

Develops strength margin factor using component capacities as described in Table 6-2.

3.

Develops inelastic energy absorption factor based on ASCE 43-05 or at about the 95 percent exceedance probability of ductility levels.

4.

Calculates the CDFM capacity as:

HCLPFcDFM = Fs " *"PGA (Equation 6-1)

where, Fs Strength margin factor, F = Inelastic energy absorption factor The strength margin factor is defined as:

Fs

=

-Dns(Equation 6-2)

where, S =

Strength of the structural element Ds=Non-seismic demand (normal operating loads)

Ds=Seismic demand 6.4.3 Functional Evaluations The HCLPF capacities for functionality are based on the comparison of the demand (ISRS) with EPRI NP-6041 [ 10] screening level HCLPFs, existing analysis, GERS, or test response spectra.

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The screening level HCLPF values provided in EPRI NP-6041 [10] Table 2-4 are presented in terms of the 5 Hz spectral acceleration at the foundation level. In accordance with EPRI 1019200 [13], these values are used to develop mounting level capacity assuming a median structure amplification factor of 1.5. The RLGM ISRS are compared with this mounting level capacity to develop HCLPF associated with 2x SSE. Anchorage checks are performed based on the spectral accelerations at the estimated equipment frequencies.

Available plant-specific seismric qualifications tests are generally biaxial and all of the published GERS are constructed on the basis of the results of previous biaxial tests of similar types of equipment. These tests apply table input motion in one-horizontal direction and in the vertical direction. For most equipment, for which GERS are available, the vertical test response spectrum (TRS) are at least equal to the horizontal TRS. The published GERS define the horizontal component of the table motion, which is, therefore, taken to represent the capacity stated either in terms of the vertical or horizontal input.

The seismic demand on equipment, on the other hand, is typically defined by JSRS in three orthogonal directions, two horizontal and one vertical. The procedure used to develop the functional capacity compares the resultant horizontal and the vertical ISRS separately with the GERS or TRS. The minimum seismic margin is taken to obtain the functional HCLPF capacity.

6.5 FUNCTIONAL EVALUATIONS OF RELAYS The functional evaluation of relays is performed based on the CDFM methodology described in EPRI NP-6041 [10] and EPRI 1019200 [13]. Since manual actuation is credited, only the non-operational, normally open state is needed for these relays. This is associated with the de-energized state and is the limiting state for the function during condition for the relay.

The relays evaluated for the ESEP are those that can cause inadvertent actuation of any one (1) of the 18 SRV's. Relay chatter can cause inadvertent operation of the SRV's, which can lead to premature depressurization of the RPV. Premature depressurization of the RPV is assumed to preclude long term RPV makeup from RCIC due to loss of adequate steam motive force to the RCIC turbine. This assumption is judged to be conservative because after the relays stop chattering, the RPV would re-pressurize and support reestablishing RCIC flow. All SRV relays that can cause inadvertent SRV operation are included on the ESEL for evaluation. The SRV relays are either GE 12HMA24A2F or GE 12HFA151 A2F model relays. In addition, successful operation of one (1) of the seven (7) ADS SRVs is sufficient for subsequent controlled RPV R02 Columbia Generating Station ESEP

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depressurization to support the credited FLEX strategy. Long term operation of a single ADS SRV for controlled RPV depressurization requires availability of the ADS SRV itself along with support from the associated solenoid valve, accumulator tank, and back-up nitrogen bottles.

It is worth noting that similar vintage plants also evaluate the relays associated with the automatic start and alignment of the RCIC system as part of the ESEP. For CGS, only manual RCIC start is credited in support of the FLEX mitigation strategy for the ESEP, and no relays are required to properly implement the RCIC system. It is possible for relay chaffer to cause inadvertent stop or start of the pump/turbine, or spurious actuation of a valve; however, it has been determined that Control Room operators are properly trained to manually initiate or realign the RCIC system as part of the emergency operating procedures. It is therefore determined that even with relay chatter of the RCIC system, human action is credited to correct this almost immediately after the seismic event and no evaluation of the RCIC relays is necessary.

6.6 RESOLUTION OF WALKDOWN FINDINGS This Section provides a summary of how the walkdown findings noted in Section 6.3.4 were resolved as part of the HCLPF margin evaluation. A total of eight (8) specific observations were identified by the ESEP SRT with the intent to further evaluate their significance and credibility given the RLGM demand level at the component's respective building location. All eight observations were addressed and resolved as summarized in Table 6-3 below.

TABLE 6-3

SUMMARY

OF ESEP WALKDOWN FINDING RESOLUTIONS ESL CMOET WALKDOWN FINDING

SUMMARY

OF RESOLUTION ID ID Anchorage capacity for E-CP-C61/P001 neglects the contribution of this specific anchor and develops a E-CP Flor rackruning margin HCLPF capacity following GTP criteria.

63 ECP-loorcrac runing Resulting anchorage HCLPF is 0.89g which is slightly 63C61/P001 through one anchor bolt higher than the resulting functional HCLPF of 0.84g.

Anchorage capacity does not control over the panel's functional capacity.

Welded connection capacity for E-MC-7A does not credit the southwest welded corner in the margin Instllaton isuecalculation. The margin calculation takes advantage of 69EM-A observed at weld the capacity of the base angle to accommodate inelastic 69

-MC7A connecting E-MC-7A deformation and the availability from operators to to is bae chnnel manually reset the MCC. Welding capacity is shown

________________not to control over the MCC's functional capacity.

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TABLE 6-3

SUMMARY

OF ESEP WALKDOWN FINDIING RESOLUTIONS (CONTINUED)

ESL CMOET WALKDOWN FINDING [

SUMMARY

OF RESOLUTION ID JIDI The spectral acceleration needed to close the gap between MCC and wall was estimated to be about 4g.

MCC E-MC-Potential SI with Further review of the RLGM demand at the Radwaste 69, 78 7A and E-adjacent 11SS cable tray Building (El 467) for components with fundamental MC-S 1/iD support frequencies above 8 Hz shows spectral accelerations below 1.8g. Impact between MCCs and concrete wall is

___________deemed not credible.

The gap between rack and wall was observed to be about 33 EIR-017 otetialSI ith1.5".

Given the RLGM demand at the location of the 33 EIR-017 adjacent concrete wall rack, the maximum displacement of the rack is estimated to be about 0.4". Therefore impact is deemed not

_____________credible.

Calculations were performed to check the valve displacements and pipe stress levels given the RCICV-1 Vales re munte on unsupported span, the 2-in pipe diameter, and the 2,20 and RCIC-V-small 2 inch diameter expected demand level at EL 422 of the Reactor 22,lnesBuilding.

Results show that displacements and torsional 46 linesand bending loads will remain within pipe allowables.

Therefore the HCLPF capacity is dictated by the

___________functional capacity of the MOV.

The HCLPF seismic capacity of the RCIC vacuum tank was evaluated based on results from the seismic qualification stress analysis (Ref: QID 094003-01). This Load path not visible report shows that the tank is anchored to the floor via 29 RIC-K-l due to insulation cover four 1/22" diameter CIP bolts. Given this anchorage detail and the available seismic margin for potential failure modes, the HCLPF capacity is demonstrated to exceed

___________the 0.50g PGA margin level.

The component-specific Test Response Spectra (TRS) is Battry harer s nt a used for HCLPF calculation. HCLPF is shown to solid-state type exceed the 0.50g PGA RLGM margin level.

A more refined calculation was performned to justify a higher fundamental frequency in the side-to-side 85 E-C2-1 direction. By considering the internal framing stififness Potential SI with and the lower center of gravity of the charger, it is adjacent HVAC justified that the charger possesses a side-to-side fundamental frequency greater than 10Hz. This in turn leads to an estimated maximum displacement within the

___________allowable clearance of 3/8".

R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015) r° RIZZO Page 39 of 99

6.7 TABULATED ESEL HCLPF VALUES (INCLUDING KEY FAILURE MODES)

Attachment B tabulates the HCLPF values for all components on the ESEL. All HCLPF values exceed the RLGM of 0.50g PGA. The table in Attachment B also identifies the method used to develop the HCLPF values and the controlling failure mode. Most of the controlling failure modes are either anchorage failure or loss of functionality and do not involve structural integrity.

In several cases the HCLPF value is stated as "> 0.50". This corresponds to components that were screened out either during the seismic walkdowns or after further review of the component's load path and RLGM-specific demand level.

R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 40 of 99 F \\RIZZO

7.0 INACCESSIBLE COMPONENTS 7.1 IDENTIFICATION OF ESEL ITEMS INACCESSIBLE FOR WALKDOWNS A total of twenty three (23) items in the ESEL were inaccessible during walkdowns due to their location within the Primary Containment. Table 7-1 provides the description of the 23 inaccessible components, the reason for their inaccessibility, and the criteria implemented to confirm the installed condition and, therefore, evaluate their seismic capacity. The criteria implemented to confirm the installed condition follows EPRI NP-604 1 [10] where a number of ways of confirming the installed condition of equipment, including follow up walkdowns, photographic, or other confirmatory evidence is provided. As can be seen under "Method of Evaluation" in Table 7-1 below, most inaccessible components were assessed using photographs dating from a June 2013 plant outage and plant design drawings.

TABLE 7-1 CGS ESEL ITEMS INACCESSIBLE DURING WALKDOWNS ES EL [COMPONENT f E~PINREASON FOR METHOD OF ID j

ID

[

ECRPTO INACCESSIBLEO)~

EVALUATION MS-R-3DADS RV D IsidePriary Photographs from 1 MSRV-3 ADSRV3DonsidaPinmary recent outage; design Contanment drawings 2 MSRV-4 AD SRV4A Isid Priary Photographs from 3SR-AAD R

AInside Primary recent outage; design MS-RV-4B DS SRV 4BContainment daig Photographs from 4

MS-RV-4C ADS SRV 4C Inside Primary recent outage; design Containment drawings MS-R-4CADS RV C IsidePriary Photographs from 5 MSRV-4 AD SV4DonsidaPinmary recent outage; design Contanment drawings MS-R-4DADS RV D IsidePriary Photographs from 6 MSRV-AD SRVSBonsidaPinmary recent outage; design Contanment drawings Insie Prmary Photographs from 7

MS-RV-5C ADS SRV 5C IonsidaPimaryt recent outage; design Contanment drawings R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 41 of 99 E~ziNl[

TABLE 7-1

SUMMARY

OF INACCESSIBLE ITEMS IN CGS ESEL (CONTINUED)

ESEL COMPONENT ITREASON FOR METHOD OF ID J

ID I

DESCRIPTION jINACCESSIBLE(I)

EVALUATION 8 S-P-3A Div. 1 Solenoid A for Inside Primary recntoutage;adesigno ADS SRV 3D Containment drawintoags

dsg 9

MS-SPV-4AA Div. 1 Solenoid A for Inside Primary Photographs from ADS RV 4

Cotaiment recent outage; design ADS RV 4

Conainiient drawings 10 MS-SPV-4BA Div. 1 Solenoid A for Inside Primary Photographs from ADS RV 4

Cotainent recent outage; design ADS RV 4

Conainmnt drawings 11 MS-SPV-4CA Div. 1 Solenoid A for Inside Primary Photographs from ADS RV 4

Cotaiment recent outage; design ADS RV 4

Cotainent drawings 12 MS-SPV-4DA Div. 1 Solenoid A for Inside Primary Photographs from ADS SRV 4D Containment recent outage; design drawings 13 MS-SPV-5BA Di. 1 Solenoid A for Inside Primary Poorpsfo ADiv SR Cnanet recent outage; design ADS RV S

Cotainent drawings 14 MS-SPV-5CA Div. 1 Solenoid A for Inside Primary Photographs from ADS SRV 5C Containment recent outage; design drawings Design drawings; Inside Primary located in pipe; seismic 55 CMS-LE-6A Wetwell Wide Range Containment, capacity based on LeelMoitrunderwater distribution system capacity Inside Primary Design documentation; 56 SPTM-TE-1A Suppression Pool Cnanet iiaiyt te Temertue Eemnt underwater components Accumulator Tank Photographs from 100 MS-TK-3V for Div. 1 Solenoid A Inside Primary rentoag;dsn Containment drawintoags

dsg for ADS SRV 3D daig Accumulator Tank IniePiay Photographs from 101 MS-TK-35S for Div. 1 Solenoid A ConsidaPinmary recent outage; design for ADS SRV 4A daig Accumulator Tank IniePiay Photographs from 102 MS-TK-3R for Div. 1 Solenoid A IonsidaPinmary recent outage; design for ADS SRV 4B daig Accumulator Tank IniePiay Photographs from 103 MS-TK-3M for Div. 1 Solenoid A IonsidaPinmary recent outage; design foContainmentCdrawings R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 42 of 99

~IN

'~RI Z ZO

TABLE 7-1

SUMMARY

OF INACCESSIBLE ITEMS IN CGS ESEL (CONTINUED)

ESEL[COMIPONENT [1REASON FOR METHOD OF IDI ID I

DESCRIPTION IINACCESSIBLE(I EVALUATION Accumulator Tank IniePiay Photographs from 104 MS-TK-3P for Div. 1 Solenoid A IonstdaPinmary recent outage; design for ADS SRV 4D daig Accumulator Tank IniePiay Photographs from 105 MS-TK-3U for Div. 1 Solenoid A ConsaidePnmary recent outage; design for ADS SRV 5B daig AccumlatorTankPhotographs from 106 MS-K3 AfrcDv.muSlenoid Tan Inside Primary recent outage; design 106 MSTK3N fo Di.

Sleoi A Containment aig for ADS SRV 5C drawings________

Note:

(1) COS was in power operation during the ESEP walkdowns and the CGS primary containment is inerted with nitrogen during power operation (typical or GE Mark I and II containments).

7.2 PLANNED WALKDOWN / EVALUATION SCHEDULE I CLOSE OUT No additional walkdowns are planned, and no closeout issues are required as a result of the evaluations performed as described in this report.

R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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[) !zz

8.0 ESEP CONCLUSIONS AND RESULTS 8.1 SUPPORTING INFORMATION CGS 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 CGS response to the NRC's 50.54(f) letter [1]. On March 12, 2014, NEl submitted to the NRC results of a study [16] 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 [...] has not been an overall increase in seismic risk for the fleet of U.S.

plants" based on the reevaluated 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."

Since CGS is a western U.S. plant, results from the Reference [16] study are not applicable.

However, CGS's Seismic Hazard and Screening Report [5] provides an estimate of the plant-level HCLPF capacity equal to 0.395g PGA. This estimate is based on GI-199 Appendix A methodology and from CGS's most recent SCDF value of 4.9x1 06/year. It is therein concluded that (1) CGS's plant-level HCLPF capacity spectrum well exceeds its seismic design-basis spectrum and (2) CGS continues to have a very low seismic risk using the latest PSHA results.

This assessment of the change in seismic risk included in CGS's screening submittal [5] is in accordance with the interim evaluations presented in the NRC's May 13, 2015 letter [ 18].

Furthermore, the March 12, 2014 NEI letter [ 15] provided an attached "Perspectives on the Seismic Capacity of Operating Plants," [ 17] which (1) assessed a number of qualitative reasons why the design of SSCs inherently contain margin beyond their design level, (2) discussed industrial seismic experience databases of perforniance of industry facility components similar to nuclear SSCs, and (3) discussed earthquake experience at operating plants.

R02 Columbia Generating Station ESEP

©

-E JM*-uv 155462/15, Rev. A (October 27, 2015)

Page 44 of 99

!%RIZZO

The fleet of currently operating NPPs 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 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 SSCs analysis and testing applications

Bounding requirements in codes and standards

Use of minimum strength requirements of structural components (concrete and steel)

Bounding testing requirements

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 Insights from the ESEP identified no items where the HCLPF is below the RLGM in terms of PGA. Accordingly, no plant modifications are necessary in accordance with EPRI 3002000704

[2] to enhance the seismic capacity of the plant.

8.3 MODIFICATION IMPLEMENTATION SCHEDULE No modification implementation schedule is required because no plant modifications are required.

R02 Columbia Generating Station ESEP ERI*.

155462/15, Rev. A (October 27, 2015)

Page 45 of 99 SRIzz

8.4

SUMMARY

OF REGULATORY COMMITMENTS No regulatory commitments are required as a result of the ESEP.

R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 46 of 99 r*

RIZZ

9.0 REFERENCES

1.

NRC (E Leeds and M Johnson) Letter 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-Ichi Accident," March 12, 2012. (ML12053A340)

2.

Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1 - Seismic. EPRI, Palo Alto, California:

May 2013, 3002000704, 2013. (ML13101A379) 3*

Letter G02-13-034 dated February 28, 2013, from A. L. Javorik (Energy Northwest) to NRC, "Energy Northwest's Response to NRC Order EA-12-049 - Overall Integrated Plan for Mitigating Strategies"

4.

EPRI 3002000704: NTTF 2.1 Seismic Augmented Approach Guideline, Questions and Answers. Nuclear Energy Institute, August 6, 2014.

5.

Columbia Generating Station, Docket No. 50-397 Seismic Hazard and Screening Report, Response to NRC Request for Informaation 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, March 12, 2015.

6.

FSAR, 1998, "Columbia Generating Station Final Safety Analysis Report," Amendment 53, November 1998.

7.

EQE, 1994, "Reactor Building Model," Calculation 59037-C-005 Rev. 0, EQE Engineering Consultants, February 1994.

8.

ASCE, 4-13, "Seismic Analysis of Safety-Related Nuclear Structures and Commentary,"

American Society of Civil Engineers, July 2013.

9.

EPRI, 1993 "Guidelines for Determining Design Basis Ground Motions, Volume 1:

Method and Guidelines for Estimating Earthquakes Ground Motion in Eastern North America", EPRI Report TR-102293.

10.

Electric Power Research Institute, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," EPRI NP-6041, Revision 1, Palo Alto, California, USA, August 1991.

11.

Electric Power Research Institute, "Methodology for Developing Seismic Fragilities,"

EPRI TR-103959, June 1994.

12.

Electric Power Research Institute, "Seismic Fragility Application Guide," EPRI 1002988, December 2002.

R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 47 of 99

- s

13.

Electric Power Research Institute, "Seismic Fragility Application Guide Update," EPRI 10 19200, December 2009.

14.

RIZZO Associates, 2015, 15-5462-F-i "CGS ESEP HCLPF Calculations," Revision 0, October 23, 2015

15.

Nuclear Energy Institute, A. Pietrangelo, Letter to E. Leeds of the USNRC, "Seismic Risk Evaluations for Plants in the Central and Eastern United States," March 12, 2014.

ADAMS Accession No. ML14083A584.

16.

Electric Power Research Institute, "Fleet Seismic Core Damage Frequency Estimates for Central and Eastern U.S. Nuclear Power Plants Using New Site-Specific Seismic Hazard Estimates" March 11, 2014. ADAMS Accession No. ML14083A586.

17.

Electric Power Research Institute, "Perspective on the Seismic Capacity of Operating Plants" March 11, 2014. ADAMS Accession No. ML14080A590.

18.

NRC (W M Dean) Letter to WUS Reactor Sites, "Screening and Prioritization Results for the Western United States Sites 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 13, 2015. ADAMS Accession No. ML15113B344.

19.

RIZZO Associates, 2015, L02 155462 "Transmittal of Walkdown Documentation and Walkdown and Screening Interim Report," Revision 1, October 7, 2015.

R02 Columbia Generating Station ESEP 155462115, Rev. A (October 27, 2015)

Page 48 of 99 E~W

ATTACHMENT A EXPEDITED SEISMIC EQUIPMENT LIST R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 49 of 99 ERW F'~uRIZZO

EXPEDITED SEISMIC EQUIPMENT LIST ITEM

[rEQUIPMENT

[EQUIPMENT N.

EQUIPMENT NUMBER DESCRIPTION NORMAL DESIRED NOTES N.((

STATE ISTATE Although only 1 ADS valve 1 MS-RV-3D ADS SRV 3D In Service In Service is required for success, all 7 ADS valves are identified for completeness.

2 MS-RV-4A ADS SRV 4A In Service In Service 3

MS-RV-4B ADS SRV 4B In Service In Service 4

MS-RV-4C ADS SRV 4C In Service In Service 5

MS-RV-4D ADS SRV 4D In Service In Service 6

MS-RV-5B ADS SRV 5B In Service In Service 7

MS-RV-5C ADS SRV 5C In Service In Service Although only 1 ADS solenoid valve for the 8 SSV3ADiv.

1 Solenoid A for ADS ISevc Inerie respective SRV is required SRV 3D for success, all 7 Div. 1 ADS solenoid valves are

______________________identified for completeness.

9

~~~~~~Div.

1 Solenoid A for ADS InSrie nSrvc 10 MS-SPV-ABA Dv1SoeodAfrDS In Service In Service SRV 4B 11 MS-SPV.4CA

~Div. 1 Solenoid A for ADS InSrie nSrvc 12 MS-SPV-4BA Dv1SoeodAfrDS In Service In Service SRV 4D 13 MS-SPV-4BA Dv1SoeodAfrDS In Service In Service SRV SB______

14 MS-SPV-4DA Div. 1 Solenoid A for ADS In Service In Service SRV SC DiOnly1aSsingledADSfmanual ADRVV peaton DHoweverotheoentireobottl 15 CIA-SPV-5AADBoteRc In Service In Service rakithaporaeSC tOinclud on thnge EDSmnul.

ASbottle israckre CtK-upor ASupply opr aDSions.A

4Bwve, and SB.

r otl Noreayks are assrocriatedSS 16 RCIC-DT-2 1ACCDrivtte TRbine Stndbyvi In Service cnimdta prtr woinllubdalen tohvrie anyL 2Aspuriouts actumationdet replay chttr eArly in the 17 RCI-P-1 RIC PUP Stanby Indervic 18 RCICV.4 RCC Turbne TrwiThrotleeInCerviceysInmSrvicee 16 RIC-D-1RCIC DieTurbine SteaatondSeviey In Service cofre-ha prtr 19 RCIC-V-41 Valve CloPSeadb Opn Service_______

R02 Columbia Generating Station ESEP 1 55462/15, Rev. A (October 27, 2015)

Page 50 of 99

~' E~iN RIZZO J

EXPEDITED SEISMIC EQUIPMENT LIST (CONTINUED)

ITEMEQUIPMENT EQUIPMENT ITO.

EQUIPMENT NUMBER DESCRIPTION NORMAL DESIRED NOTES N.JSTATE STATE RCIC Turbine Lube Oil In Service -

In Service -

20 RI--6Cooling Valve Closed Open RCIC Head Spray Line In Service -

In Service -

21 RI--3Injection Valve Closed Open Min flow valve is normally open, but would need to 2 RCCV19RCIC Minimumn Flow Valve In Service -

InSrie close during RCIC to Wetwell Closed operation to prevent diverting RCIC flow from

____________the RPV to the wetwell.

2 RCV.2RCIC Flow Control In Service -

In Service -

Governor Valve Open Open RCIC suction normally aligned to CST, but ESEP In Srvie In Srvie -guidance assumes that CST 24 RCIC-V-10 RCIC CST Suction Valve ISevc-Inerie would be unavailable due to Open losed initiating event (e.g.,

beyond design basis seismic event).

RCIC suction nornally aligned to CST, but ESEP guidance assumes that CST would be unavailable due to 2 RCCV31RCIC Suppression Pool In Service -

In Service -

initiating event (e.g.,

Suction Valve Closed Open beyond design basis seismic event). Therefore, the RCIC Suppression Pool suction valve is required to

______open.

26 RCIC-HX-2 RCIC Lube Oil Cooler In Service In Service RCIC Vacuum Pump (for Sady I

evc 27 RCIC-P-2

~~vacuum tank)

Sady I

evc RCIC Condensate Pump (for Sady I

evc 28 RCIC-P-4

~~vacuum tank)

Sady I

evc 29 RCIC-TK-1 RCIC Vacuum Tank Standby In Service East of Turbine 30 RCIC-HX-l Barometric Condenser Standby In Service 31 RCIC-FIC-600 RCIC Flow Controller In Service In Service 32 RCIC-FT-3 RCIC Flow transmitter In Service In Service 33 E-IR-P0 17 RCIC Instrument Rack In Service In Service 34

~~~~~~RCIC Woodward Governor InSrie nSrvc EG-MEG-RTurbine Supports long term RPV makeup using FLEX pump for flow and LPCI C 35 RRV4CL ILopCIjtinVle In Service -

In Service -

injection path. This valve RIR--2CLCILopC necio ale Closed Open is operated manually; therefore no associated breaker is necessary to

________support FLEX.

R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 51 of 99 IN RIZZO

EXPEDITED SEISMIC EQUIPMENT LIST (CONTINUED)

EQUIPMENT EQUIPMENT NTO.

EQUIPMENT NUMBER DESCRIPTION NORMAL DESIRED NOTES N.STATE STATE Supports long term Suppression Pool level makeup using FLEX pump LPCIC Spprssin Pol n Sevic n Srvie -for flow and LPCI C test 36 RHR-V-21 OpeCSppeso Po nSevc nSeve-return path. This valve is Test Return Line MOV Closed Opn operated manually; therefore no associated breaker is necessary to support FLEX.

POST ACCIDENT MON 37 MS-LRJPR-623A INSTR LOOP A PRESS &

In Service In Service LEVEL RECORDER 38 MS-LT-26A RP EE IERNE In Service In Service

______________XMTR (LOOP A) 39 MSS-NTRUSGNALIn Service In Service MS-SU-IARESIST UNIT 120VAC/24VDC POST-40 MS-E/S-613A ACCIDENT POWER In Service In Service SUPPLY 41 E-PP-.7AA Critical Inverter Fed Ups In Service In Service Power Panel 42 MS-PT-51A MS RPV PRESSURE In Service In Service RPV Water Level and 43 E-IR-P004 PrsueIsrmn ak In Service In Service 44 E-CP-H13/P601 ManCnrlRo ae -

In Service In Service CP-H 13/P601 45

~~~~~~Main Control Room Panel E-InSrie nSrvc E-CPH13/612CP-H13/P612 46 CMS-PR-3 WE ELPESIn Service In Service RECORDER (DIV 1) 47 SIGALRESSTR9UIT In Service In Service E-SRU-95(POWERED)

Control Room Panel E-CP-48 E-CP-H13/P841 H1/81In Service In Service 49 SUMRE-CPAB-PESS In Service In Service CMS-PT-3MONITOR 50 E-IR-66 RXBD NT AK In Service In Service (rack for CMS-PT-3)

CLASS 1 120VAC/24VDC 51 E-E/S-99 POWER SUPPLY (Power In Service In Service Supply for Panel 841 and 831) 52 CMS-LR-3 SUPPRESS POOL LEVEL In Service In Service RECORDER (DIV 1) 53

~~~~WET WELL LEVEL WIDE InSrie nSrvc 53 CMS-LT-6A RANGEervice__InService 54 Contol-Rom3Pnel83CP-In Service In Service E-CPH13/831H13/P831 55 CMS-LE-6A WEWL IERNE In Service In Service

_____________LEVEL MONITOR___________

56 SPTM-TE-1A Suppression Pool nSrie nSrvc

_________________Temperature ElementInSrie nSrvc R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 52 of 99 F' ~RIZZO

EXPEDITED SEISMIC EQUIPMENT LIST (CONTINUED)

ITEM EQUIPMENT EQUIPMENT NO.

EQUIPMENT NUMBER DESCRIPTION NORMAL DESIRED NOTES ISTATE STATE 57 SPTM-MVII-1A Voltage/Current Converter In Service In Service 58 SPTM-SRU-15 Signal Resistor Unit In Service In Service 59 SPTM-SUM-1 TemperatureInSrie nSrve Summer/AveragerInSrie nSrvc 60 SPTM-TI-.5 Suppression Pool MCR In Service In Service Indicator_____________

Needed because RCIC 61 E-CP-H13/P621 RCIC Control Panel In Service In Service circuitry wires pass through this cabinet Needed because RCIC and 62 E-CP-H13/P684 Control System Div 1 In Service In Service ADS circuitry wires pass Termination Cabinet through this cabinet Needed because RCIC 63 E-CP-C61/P00 1 Remote Shutdown Panel In Service In Service circuitry wires pass through this cabinet Needed because ADS 64 E-CP-H131P628 ADS Control Panel (Division In Service In Service circuitry wires pass through

1) this cabinet.

Power to E-TB-D1 140 may be supplied by either FLEX Diesel Generator DG4 or DG5. Both DG4 and DG5 are portable 480 VAC generators that are out of Division 1 FLEX DG the scope of the ESEL.

65 TB-D1 140 CnetoPot(4VA)

In Service In Service DG4 and DG5 include Connctin Pint 480AC)approximately 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months of fuel prior to the need for refueling. Equipment required to refuel the portable FLEX DGs is also out of the scope of the ESEL.

66 E-CBDG4/7AADG4 OUTPUT BREAKER Lce pn Coe 66 E-CBDG4/7AATO E-MC-7AA Lce pn Coe Motor Control Center E-MC-67 E-MC-7AA 7AIn Service In Service 68 E.-CB-7A6A EM-A FedrBakr In Service In Service

_____(RPS Rm 1) 69 E-MC-7A Moo oto etrEM-In Service In Service

_____________7A 70 E-B-1-VSTTONBTTR In Service In Service E-Bl1-1 Breaker E-CB-7A2BL from 71 E-CB-7A2BL E-MC-7A to Battery Charger In Service In Service E-C1-I A Breaker E-CB-C 1/lA/1 from 72 E-CB-C1/1A/1 E-MC-7A to Battery Charger In Service In Service E-CI1-lA 73 E-B-lVITL-ATTRY In Service In Service E-Cl-lA

~CHARGER E-C 1-1A R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 53 of 99

©,

EXPEDITED SEISMIC EQUIPMENT LIST (CONTINUED)

ITEM 11EQUIPMENT EQUIPMENT NO.

EQUIPMENT NUMBER DESCRIPTION NORMAL DESIRED NOTES I.STATE STATE Breaker E-CB -Cl1/1A/2 from 74 E-BC/A2Battery Charger E-CI-1A to In Service In Service E-CB-C/1A12Div. 1 125VDC PANEL E-PNL-CI/1 75 E-PNL-C1/1 125VDC PANEL E-PNL-In Service In Service Cl/i Breaker E-CB-C1I/1A from Div. 1 125VDC PANEL E-76 E-CB-C 1/lA PNL-Cl/l to Div. 1 l25VDC In Service In Service MAIN DIST PANEL E-DP-77 125DDC-AIN/1STIn Service In Service E-DPSl/1PANEL E-DP-Sl/1 125 VDC Bus MCC E-MC-Supplies motive power for 78 E-MC-S1I/ID1/D In Service In Service valves RCIC-V-10, RCIC-S___1/1D___V-3 1, and RCIC-V-46.

12VCBsPnlE-DP-Supplies control power for 79 E-DP-S 1/1D 15VCBsPnlIn Service In Service RCIC valves and vacuum S1/1D tank purnps.

i25VDC Bus Panel E-DP-Supplies motive and control 80 E-DP-S1/1A S1/lA In Service In Service power for ADS SRV

_____________solenoid valves.

48VDC for RCIC-V-2 81 RCIC-E/S-603 Control Power (in panel E-In Service In Service CP-H113/P612)

DIV 1 CRITICAL POWER 82 E-IN-3A SUPPLY INVERTER A E-In Service In Service

_______IN-3A 250V STATION BATTERY 83 E-B2-1

-B-In Service In Service 84 CB 1 BekrC1frmEM-A In Service In Service

______to Battery Charger E-C2-1 85 E-C2-1

-2lVTA ATR In Service In Service

____________CHARGER E-C2-1 Breaker CB2 from Battery 86 CB2 Charger E-C2-1 to Div. 1 In Service In Service 250VDC MAIN DIST

_____PANEL E-DP-S2/1 87 E-MC-S2/1A-A 25VCBsMCEM-In Service In Service

_____S2/lA-A 88 E-DISC-7A3B 120 VAC E-PP-7A Supply In Service In Service 89 E-DISC-PP7A/5L E-PP-7AF Supply In Service In Service 90 RCIC-RMS-V/10 Coto wthfrRI--

In Service In Service 10 91 RCIC-RMS-V/31 Coto wthfrRI--

In Service In Service 92 RCIC-RMS-S36 RCCMna ntainIn Service In Service Pushbutton 93 MSRSR/DC Control Switch for MS-RV-In Service In Service 94

~~~~~~Control Switch for MS-RV-InSrie nSrvc 94 MS-RMS-RV/4A/C 4AnService_

InService R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 54 of 99 ERIFd

~~pizzo isJ ~ o

EXPEDITED SEISMIC EQUIPMENT LIST (CONTINUED)

ITE EQUIPMENT TEQUIPMENT NTEo. EQUIPMENT NUMBER DESCRIPTION NORMAL DESIRED NOTES No]STATE j

STATE 95 MS-RMS-RV/4B/C Control Switch for MS-RV-InSrie nSrvc 96 MS-RMS-RV/4C/C CnrlStcfoM-R-In Service In Service 4C 97

~~~~~~Control Switch for MS-RV-InSrie nSrvc 96 MS-RMS-RV/4C/C 4DSric nSevc 98 MS-RMS-RV/5B/C Control Switch for MS-RV-InSrie nSrvc SB 99 MS-RMS-R V/SC/C CnrlSicfoM-R-In Service In Service 5C Although only 1 ADS accumulator tank for the Accumulator Tank for Div. 1 respective solenoid is 100 MS-TK-3V Solenoid A for ADS SRV 3D In Service In Service required for success, all 7~

Div. 1 ADS accumulator tanks are identified for completeness.

101 MS-TK-3 S AcmltrTnfoDi.1 In Service In Service Solenoid A for ADS SRV 4A Accumulator Tank for Div. 1 102 MS-TK-3R In Service In Service Solenoid A for ADS SRV 4B Accumulator Tank for Div. 1 103 MS-TK-3M In Service In Service Solenoid A for ADS SRV 4C Accumulator Tank for Div. 1 104 MS-TK-3P In Service In Service Solenoid A for ADS SRV 4D Accumulator Tank for Div. 1 105 MS-TK-3U In Service In Service Solenoid A for ADS SRV SB Accumulator Tank for Div. 1 106 MS-TK-3N In Service In Service Solenoid A for ADS SRV SC Pressure transmitter for RCIC suction line. If RCIC 17 CCPTSRCIC P-i Suction Pressure ISevc Inerie suction valves inadvertently Transmitter close, RCIC will trip on low suction and not damage

_____________________itself.

108 E-CP-H13/P894 Termination Cabinet TCGI In Service In Service Process Instrument 109 E-CP-H13/P682 TriaonCbetIn Service In Service 110 E-PNL-I~N/3 Di ttcTase wth In Service In Service Panel 111 E-DISC-7A4A Disconnect E-DISC-7A4A In Service In Service 15KVA Regulating XFMR 112 E-TR-7AJ2 for E-PP-7AF (Remote In Service In Service Shutdown) 113 E-TR-7A 7KEX RfoCrtcl In Service In Service Power PNL E-PP-7A 114 E-PP-7AF 12/4v25 eoeIn Service In Service Shutdown Power Panel 115 E-DISC PP7AF/9 Disconnect E-DISC PP7AF/9 In Service In Service 116 E-DISC-S11D1 Fusible Safety Switch In Service In Service 250VDC Main Distribution 117 E-DP-S2/1 ae -P5/

In Service In Service R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 55 of 99

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" RIZZO J

EXPEDITED SEISMIC EQUIPMENT LIST (CONTINUED)

'No.M EQUIPMENT EQUIPMENT TEM EQUIPMENT NUMBER DESCRIPTION NORMAL.

DESIRED NOTES STATE STATE 118~ E-TR-IN/3 InetrAtraePwrIn Service In Service Supply 119 E-PP-7A RdatCrtclPwrIn Service In Service Panel 121 E-TB-B1/1 Terminal Box for Battery E-In Service In Service B2-1 122 E-MC-S2/lA-B 25VCBsMCEM-In Service In Service S2/IA-B Included all SRV relays that could chatter and potentially cause 12

-SRYAK8 2-MS-RLY-ADK38 Relay nSeic Inerce inadvertent MSRV 12

-SRYAK8 for MS-RV-1A (Solenoid C) iSevc Inerce actuation to sufficiently depressurize the RPV and preclude adequate RCIC injection 124 2-MS-RLY-ADK30 2-MS-RLY-ADK3O Relay In Service In Service

_________________for MS-RV-lB (Solenoid C) 125 2-MS-RLY-ADK42 2-MS-RLY-ADK42 Relay In Service In Service

_________________for MS-RV-1C (Solenoid C) 126 2-MS-RLY-ADK4O 2-MS-RLY-ADK40 Relay In Service In Service for MS-RV-1D (Solenoid C) 127 2-MS-RLY-ADK22 2-MS-RLY-ADK22 Relay In Service In Service for MS-RV-2A (Solenoid C) 128 2-MS-RLY-ADK24 2-MS-RLY-ADK24 Relay In Service In Service for MS-RV-2B (Solenoid C) 129 2-MS-RLY-ADK36 2-MS-RLY-ADK36 Relay In Service In Service for MS-RV-2C (Solenoid C) 130 2-MS-RLY-ADK34 2.-MS-RLY-ADK34 Relay In Service In Service for MS-RV-2D (Solenoid C) 131 2-MS-RLY-ADK28 2-MS-RLY-ADK28 Relay In Service In Service for MS-RV-3A (Solenoid C) 132 2-MS-RLY-ADK32 2-MS-RLY-ADK32 Relay In Service In Service

_________________for MS-RV-3B (Solenoid C) 133 2-MS-RLY-ADK26 2-MS-RLY-ADK26 Relay In Service In Service for MS-RV-3C (Solenoid C) 134 2-MS-RLY-ADK100 2-MS-RLY-ADKI00 Relay In Service In Service for MS-RV-3D (Solenoid C)_____________

.135 2-MS-RLY-ADK93 2-MS-RLY-ADK93 Relay In Service In Service

~~~for MS-RV-4A (Solenoid C) 136 2-MS-RLY-ADK87 2-MS-RLY-ADK87 Relay In Service In Service for MS-RV-4B (Solenoid C) 137 2-MS-RLY-ADK98 2-MS-RLY-ADK98 Relay In Service In Service for MS-RV-4C (Solenoid C) 138 2-MS-RLY-ADK91 2-MS-RLY-ADK9l Relay In Service In Service for MS-RV-4D (Solenoid C) 13

-SRYAK5 2-MS-RLY-ADK95 Relay In Service In Service

__39

_2-M_-RLY-

__DK

_5 for MS-RV-5B (Solenoid C) 140 2-MS-RLY-ADK89 2-MS-RLY-ADK89 Relay In Service In Service

_________________for MS-RV-5C (Solenoid C)

R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 56 of 99

EXPEDITED SEISMIC EQUIPMENT LIST (CONTINUED)

NTo.

EQUIPMENT NUMBER J DESCRIPTION 141 2MS-RL-ADK1 2-MS-RLY-ADK1A Relay 141

-MS-LY-AK1A for MS-RV-3D (Solenoid A)

I2-MS-RLY-ADK8A Relay 142 2MS-RY-AD8A jfor MS-RV-3D (Solenoid A)

EQUIPMENT NOR~MAL NOTES 143 2-MS-RLY-ADK6A 2-MS-RLY-ADK6A Relay for ADS SRVs (Solenoid A 144 2-MS-RLY-ADK7A 2-MS-RLY-ADK7A Relay for ADS SRVs (Solenoid A 2-MS-RLY-ADK1A 145 2-MS-RLY-ADK1B 146 2-MS-RLY-ADK8B 147 2-MS-RLY-ADK6B 148 2-MS-RLY-ADK7B 2-MS-RLY-ADK7B Rela:

149 TB-R322 Terminal Box outside Penetration E-X-1 05C In Service Tenminal Box is a passive penetration mounted directly to the wall structure. It was added for completeness since it contains wiring to actuate the ADS System, but is excluded from the ESEP Analysis since it is considered part of the structure.

150 TB-C522 Peerto 0CIn Service 151 SB-W020 Splice Box In Service Terminal Box is a passive penetration mounted directly to the wall structure. It was added for completeness since it contains wiring to actuate the ADS System, but is excluded from the ESEP Analysis since it is considered part of the structure.

The Splice Box is a passive component adjoining two series of cable along the electrical distribution system. It was added to the list for completeness since it contains wiring to actuate the ADS System, but is excluded from the ESEP analysis since it is considered part of a distribution system.

T R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 57 of 99

EXPEDITED SEISMIC EQUIPMENT LIST (CONTINUED)

ITEMEQUIPMENT EQUIPMENT ITEM EQUIPMENT NUMBER DESCRIPTION NORMAL DESIRED NOTES STATE STATE Relays associated with Division 2 ADS SRV 152 E-CP-H13/P631 ADS Control Panel (Division In Service In Service solenoids could cause relay

2) chatter and inadvertent MSRV actuation.

153 RCIC-RMS-PISl~

RCIC Low DischargeInSrie nSrvc Pressure Override SwitchInSrie nSrvc RCIC High Area 154 LD-RMS-S2A Temperature Override Test In Service In Service Switch LD-RMS-S2A RCIC High Area 155 LD-RMS-S2B Temperature Override Test In Service In Service Switch LD-RMS-S2B___________

156 E-CPH13/P632 Main Control Room Panel E-InSrie nSrvc 157 E-CP-H13/P632 ManCnrlRo ae -

In Service In Service CP-H13/P642 RCJCHig ExaustPresur InServce ypased Support override of RCIC 158 CICPS-A RIC ighExhust resure In ervce ypased High Exhaust Pressure Trip 15 CCP-Aswitch RCIC-PS-9A (25 psig).

RCICHig ExhustPresureSupport override of RCIC 159 RC*C-S-9 RIC ighExaus Pessre In Service Bypassed High Exhaust Pressure Trip 15 ~CP-Bswitch RCIC-PS-9B (25 psig).

Support override of RCIC 160 RCIC-PS-12A RCIC High Exhaust Pressure In Service Bypassed High Exhaust Pressure Trip switch RCIC-PS-1 2A (10 psig).

Support override of RCIC 161 RCIC-PS-1213 RCIC High Exhaust Pressure In Service Bypassed High Exhaust Pressure Trip switch RCIC-PS-12B (10 psig).

RCICHig ExhustPresureSupport override of RCIC 162 RCI -P

-12 R IC ighEx aus P essre In Service Bypassed High Exhaust Pressure Trip 12 RI-S1Cswitch RCIC-PS-12C (10 psig).

RCICHig ExhustPresureSupport override of RCIC 163 RCI -P

-12 R IC igh Ex aus P ess re In Service Bypassed High Exhaust Pressure Trip 13 RI-S1Dswitch RCIC-PS-12D (10 psig).

164 E-IR-P029 RCIC Instrument Rack In Service In Service Breaker RCIC-42-$21A5C 165 RCIC-42-S21A5C for power to RCIC Minimum In Service Off Ruppor minabimum flowevalve

__________Flow Valve RCIC-V-19 RCICminimumflowvalve.

R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 58 of 99 FRIZZO

ATTACHMENT B

SUMMARY

OF ESEL HCLPF VALUES R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 59 of 99

¶1~RIZZO

SUMMARY

OF ESEL HICLPF VALUES ELEV ITEM# EQUIPMENT ID DESCRIPTION JBLDG 1

(F)

GROUP HCLPF (g)

FAILURE MODE EVALUATION METHOD 1

MS-RV-3D ADS SRV 3D C

547 ADS SRVs

> 0.50 Functional Screened > 0.50g PGA 2

MS-RV-4A ADS SRV 4A C

547 ADS SRVs

> 0.50 Functional Screened > 0.50g PGA 3

MS-RV-4B ADS SRV 4B C

547 ADS SRVs

> 0.50 Functional Screened > 0.50g PGA 4

MS-RV-4C ADS SRV 4C C

547 ADS SRVs

> 0.50 Functional Screened > 0.50g PGA 5

MS-RV-4D ADS SRV 4D C

547 ADS SRVs

> 0.50 Functional Screened > 0.50g PGA 6

MS-RV-5B ADS SRV 5B C

547 ADS SRVs

> 0.50 Functional Screened > 0.50g PGA 7

MS-RV-5C ADS SRV SC C

547 ADS SRVs

> 0.50 Functional Screened > 0.50g PGA 8

MS-SPV-3DA Div. 1 Solenoid A for ADS SRV C

547 ADS SRVs

> 0.50 Functional Screened > 0.50g PGA 3D__

9 MS-SPV-4AA Div 1SlniAfoADSV

C.547 ADS SRVs

> 0.50 Functional Screened > 0.50g PGA 10 MS-PV-4BA Div. 1 Solenoid A for ADS SRV C

57ASS~

.0 Fntoa cend

.0 G

10 MS-PV-4BA 4BC 54 ADSRs

>05Fucinl Sred>0.gPA 11 MSSPV-4CA Div. 1 Solenoid A for ADS SRV C

57ASS~

.0 Fntoa cend

.~

G 11 MS-PV-4CA 4CC 57 ASSRs

>.5Futinl Sred>05gPA 12 MS-PV-4DA Div. 1 Solenoid A for ADS SRV C

57ASS~

.0 Fntoa cend

.~

G 12 M-SPV4DA 4DC 54 ASSRs

>.5Fucinl Sred>05gPA 13 MS-PV-SBA Div. 1 Solenoid A for ADS SRV C

54 ADSRs

>00 Fucinl cred>0.gPG 13 MS-SV-5BA SBC 54 ADSRs

>05Futinl Sred>0.gPG 14 MS-PV-SCA Div. 1 Solenoid A for ADS SRV C

57ASS~

.0 Fntoa cend

.~

G 14 MS-SV-5CA SCC 54 ADSRs

>05Fucinl Sred>0.gPG 15 CIA-TK-2A ADS Bottle Rack R

441 ADS Bottle

> 0.50 Functional Screened > 0.50g PGA 16 RCIC-DT-1 RCIC Drive Turbine R

422 RCIC Pump 1.01 Functional Earthquake Experience and Turbine Data 17 RCIC-P-1 RCIC PUMP R

422 RCIC Pump 1.01 Functional Earthquake Experience and Turbine Data MOVs - Low 18 RCIC-V-1 RCIC Turbine Trip Throttle R

422 Elevation

> 0.50 Functional Screened during Valve Meeting walkdowns

______Caveats_______

R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 60 of 99 r! RlZZO9

SUMMARY

OF ESEL HCLPF VALUES (CONTINUED)

ITEM# [EQUIPMENT ID }DESCRIPTION

[BLDG1]

(FT)'

GROUP JHCLPF (g)]

FAILURE MODE EVALUATION METHOD MOVs - Low 19 RCIC-V-45 RCIC Turbine Steam Stop Valve R

422 Elevaton Sceeneddurin Meeting

>05Fucinl walkdowns Caveats____________

MOVs -

20 RCIC-V-46 RCIC Turbine Lube Oil Cooling R

422 Small 1.01 Functional Earthquake Experience Valve Diameter Data Lines MOVs - High 21 RCIC-V-13 RCIC Head Spray Line Injection R

548 Elevation

> 0.50 Functional Screened during Valve Meeting walkdowns Caveats____________

MOVs -

22 RCIC-V-19 RCIC Minimum Flow Valve to R

422 Small 1.01 Functional Earthquake Experience Wetwell Diameter Data Lines Assigned based on rule 2 RCCV2 RCIC Flow Control Governor R

42 RCIC Pump 10Fucinl of the box. Parent 2 RCCV2 Valve R

42 and Turbine 10Fucinl component: RCIC-DT-1 MOVs - Low Elevation

>05Fucinl Screened during 24 RCIC-V-10 RCIC CST Suction Valve R

422 MeetingntinaMeetingwalkdowns Caveats MOVs - Low 25 RCIC-V-31 Valve Suppression Pool Suction R

422 Elevation

> 0.50 Functional Screened during VveMeeting walkdowns

______Caveats Assigned based on rule 26 RI-X2 RI ub i

olrR 42 RCIC Pump 11Fucinl of the box. Parent 2 RCCH2 RCCLbOiColrR 42 and Turbine 10Fucinl component: RCIC-DT-1 R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 61 of 99 E~IN r~pizz~

0

£

SUMMARY

OF ESEL HICLPF VALUES (CONTINUED)

Ir I1 ITE] QUIMET I [DESRITIO BDG1 FET)

GROUP

[HCLPF (g)

FAILURE MODE JEVALUATION METHOD RCIC 27 RCIC-P-2 RCIC Vacuum Pump (for R

422 Vacuum 0.86 Functional Scaling from previous vacuum tank)

Tank seismic analysis RCICSclnfrmpeiu 28 RCIC-P-4 RCIC Condensate Pump (for R

422 Vacuum 0.86 Functional Seimcaalys fo peios vacuum tank)

Tank simcaayi RCIC 29 RCIC-TK-1 RCIC Vacuum Tank R

422 Vacuum 0.86 Functional Scaling from previous Tank seismic analysis RCIC 30 RCIC-HX-1 Barometric Condenser R

422 Barometric

> 0.50 Structural Screened > 0.50g PGA Condenser Control Assigned based on rule 31 RCIC-FIC-600 RCIC Flow Controller W

501 Room 0.66 Functional of the box. Parent Benchboard component: E-CP-H 13/P601 Instrument Generic Equipment 32 CI-F-3 RCC lo trnsiterR 71 Raks0.79 Functional Ruggedness Spectra 32 RIC-T-3 RCICFlo trnsmtterR 41 Rcks(GERS)

InstrmentGeneric Equipment 33 EI-01 CCIstuetRakR 41Racks 0.79 Functional Ruggedness Spectra

____________(GERS)

Assigned based on rule 34 E-iGR RCIC Woodward Governor R

422 andI Turbin 1.01 Fucinl of the box. Parent EMEGR Turbine Pn umpbuntina component: RCIC-DT-1 MOVs - High 35Elevation Screened during 35 RJR-V-42C LPCI Loop C Injection Valve R

522 Meig>

0.50 Functional wadon

_____MCaeatsn walkdowns__________

R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 62 of 99 T RIZZO

SUMMARY

OF ESEL IICLPF VALUES (CONTINUED) 1IELEV f I

ITM D ES~n(FT)LD 1

GROUP HCLPF (g)

FAILURE MODE EVALUATION METHOD MOVs - High 36 RHR-V-21 LPCI C Suppression Pool Test R

441 Elevation

> 0.50 Functional Screened during Return Line MOV Meeting walkdowns Caveats POSTACCDEN MONConrolAssigned based on rule M3L7PR INSTR LOOP A PRESS &

W 501 Room 0.66 Functional o h o.Prn 623ALEVL REORDR Bechbardcomponent:

E-CP-623 LEVL RCORDR BnchbardH 13/P601 RPV LEVEL WIDE RANGE Instrument Generic Equipment 38 MS-LT-26A MT(OPA)R 522 Rcs0.79 Functional Ruggedness Spectra XMTR(LOO A) acks(GERS)

Control Assigned based on rule 39 NQDTA MS INSTR SIGNAL RESIST 51Room 06Fucinl of the box. Parent UNIT Vertical component: E-CP-Cabinets H 13/P612 Control Assigned based on rule 400VMC/E4DC-6OSA-Room 06Fucinl of the box. Parent 120SE/-63 VC2VCPS-W 5010.6Fntoa ACCIDENT POWER SUPPLY Vertical component: E-CP-Cabinets H13/P612 WallGeeiEqimn 41 E-PP-7AA Critical Inverter Fed Ups Power W

51Mounted 06Fcinl Rggednessc Squpmetr Panel W

51Distribution (GE uctonl RugdeS) etr

_______Panel (GERS)

Instrument Generic Equipment 42 MS-PT-51lA MS RPV PRESSURE R

522 Racks 0.79 Functional Ruggedness Spectra (GERS)

RPV ate Levl ad Prssue IntruentGeneric Equipment 43E-RP04Instrument RakR 52Rcs0.79 Functional Ruggedness Spectra E-IRP00 RP Waer LvelandPresure R

52 Rcks(GERS) 44E-CP-Main Control Room Panel E-W 501 Contol

.6Fntoa Earthquake Experience H13/P601 CP-H13/P601 Rohom r

06Fucinl Data R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 63 of 99 E~iN RIzz0

SUMMARY

OF ESEL HCLPF VALUES (CONTINUED)

ITEM# EQUIPMENT ID [

DESCRIPTION BLDG1 (FT)V GROUP HCLPF (g) [FAILURE MODE JEVALUAT METHOD Control 45E-CP-Main Control Room Panel E-RoomEatqkexprnc H13/P612 CP-H13/P612 W,

501 Vertical 0.66 Functional EarthqakeEpec Cabinets WTELPESControl Assigned based on rule 46 CMS-PR-3 WTELPSSW 501 Room 0.66 Functional o h o.Prn RECORDER (DIV 1)

Benchboard component: E-CP-H13/P601 Control Assigned based on rule SIGNAL RESISTOR UNIT Room of the box. Parent 4 E-R-5 (POWERED)

W 51Vertical 06Fucinl component: E-CP-Cabinets H 13/P841 Control 48E-CP-Control Room Panel E-CP-W 51Room0.6Fntoa Erhqkexpinc 48 H13/P841 H13/P841 Wer501alData Cabinets Instrment 0.79Generic Equipment 49 CMS-PT-3 MONITRE.HMPRSIntuet 07 Functional Ruggedness Spectra SUPPTRECHM PESR 501 Racks (GERS)

Generic Equipment 50 E-IR-66 RX BLDG INSTR RACK (rack R

501 Instrument 07 ucinl Rgens pcr for CMS-PT-3)

Racks 0(Guntonl RugdeSpe)r CLS 2VC2VCControl Assigned based on rule Room of the box. Parent 51 E-E/S-99 POWER SUPPLY (Power W

501 Vetcl0.66 Functional cmoet

-P Supply for Panel 841 and 831)

CaiertscHi 3/mpon4t 1-Control Assigned based on rule 5 CM-R3 SUPPRESS POOL LEVEL W

0 om06 ucinl of the box. Parent 5 CMLR3 RECORDER (DIV 1)

WB501 hRom r

06Fucinl component: E-CP-BenchoardH13/P601 R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 64 of 99 ERIN

~ ~

u~:~pi~zo

SUMMARY

OF ESEL HCLPF VALUES (CONTINUED)

ITE

[QUIMET I

DSCRPTON LD 1

T)

GROUP HCLPF (g) JFAILURE MODE EVALUATION METHOD Control Assigned based on rule WETWELL LEVEL WIDE Room of the box. Parent 53 CMS-LT-6A RNEW 501 Vetcl0.66 Functional cmoet

-P Cabinets H 13/P831 Control 54 E-P oto omPnlEC-W 501 VRtica 0.66 Functional Earthquake Experience H13/P831 H13/P831 VetclData Cabinets 55 CSL-A WETWELL WIDE RANGE C

47 Underwater

> 0.50 Functional Screened during 55 CSL-A LEVEL MONITOR C

47 Instruments walkdowns 56 SPTM-TE-1A Supeso olTmeaue C

46 Underwater

> 0.50 Functional Screened during

_______Element Isrmnswalkdowns Control Assigned based on rule 57 SPTM-MV/I-VlaeuretCnrerW 501 Rom0.66 Functional o h o.Prn 1A VlaeCretCnetrVertical component: E-CP'-

________Cabinets H 13/P831 Control Assigned based on rule Room,of the box. Parent 58 SPTM-SRU-15 Signal Resistor Unit W

501 Vetcl0.66 Functional cmoet

-P Cabinets H 13/P831 Control Assigned based on rule Room of the box. Parent 59 SPTM-SUM-1 Temperature Summer/Averager W

501 Vetcl0.66 Functional cmoet

-P

____Cabinets H 13/P831 Control Assigned based on rule of the box. Parent 60 SPTM-TI-5 Suppression Pool MCR Indicator W

501 Room 0.66 Functional cmoet

-P BenchoardH13/P601 Control 61

-P-RCIC Control Panel W

501 Rom0.66 Functional Earthquake Experience H13/P621 Vertical Data

______Cabinets R02 Columbia Generating Station ESEP 15*5462/15, Rev. A (October 27, 2015)

Page 65 of 99 P)IZZO

SUMMARY

OF ESEL HCLPF VALUES (CONTINUED)

ITEM [EQUIPMENT ID]

DESCRIPTION

[BLDG11E(FT)

GROUP HCLPF (g)

FAILURE MODE

[EVALUATION METHOD Control 62 E-CP-Control System Div 1 W

501 Room0.6Fntoa Erhqkexpinc H13/P684 Termination Cabinet Termination 0.6Fntoa DarthqakeEprec Cabinets E-CP-

~~~~~Rem oteEat q a e xp r nc 63 EC61P-O Remote Shutdown Panel W

467 Shutdown 0.84 Functional DarthakeEprec C6 l/P001 Panel Dt Control E-CP-Room Earthquake Experience 64 H13/P628 ADS Control Panel (Division 1)

W 501 Vertical 0.66 Functional Data Cabinets 65 TB-D 1140 Division 1 FLEX DG D

441 D FLEX

> 0.50 Functional Screened during Connection Point (480VAC)

Connection walkdowns E-CB-DG4 OUTPUT BREAKER TODWalGeriEqpmn 66 DG/A

-C7AD 455 Mounted 0.84 Functional Ruggedness Spectra DG/A

-C7ABreaker (GERS)

Moto CotrolCener EMC-Generic Equipment 67 E-MC-7AA Moo7 CnrlAetrA-C D

441 D MCC 0.56 Functional Ruggedness Spectra (GERS)

E-MC7AA eede BrekerAssigned based on rule 68 E-CB-7A6A E

C-AFedrBakrW 467 W MCCs 0.58 Functional of the box. Parent (RPPS Rmf 1)

____component:

E-MC-7A 69 E-MC-7A Motor Control Center E-MC-7A W

467 W MCCs 0.58 Functional Earthquake Experience Data 125V STATION BATTERY E-Generic Equipment 70 E-B 1-1 B 1-1 W

467 Batteries 0.73 Functional Ruggedness Spectra

________(GERS)

Breaker E-CB-7A2BL from E-Assigned based on rule 71 E-CB-7A2BL MC-7A to Battery Charger E-W 467 W MCCs 0.58 Functional of the box. Parent C1-IA

___component:

E-MC-7A Breaker E-CB-C1/1A/1 from E-Solid State Assigned based on rule 72 E-CB-C1/IA/1 MC-7A to Battery Charger E-W 467 Charger and 0.59 Functional of the box. Parent

__________Cl1-1A

_______Inverter

______________component:

E-C1-lA R02 Columbia Generating Station ESEP 155462/1 5, R~ev. A (October 27, 2015)

Page 66 of 99 o

IZ q

SUMMARY

OF ESEL HCLPF VALUES (CONTINUED)

ITEM# EQUIPMENT ID DESCRIPTION BLDG1 (FT GROUP HCLPF (g)] FAILURE MODE EVALUATION METHOD BB11VTLBTEYSolid State Generic Equipment 73 B-Cl-IACAGREl-A W

467 Charger and 0.59 Functional Ruggedness Spectra CHRE

-AInverter (GERS)

Brae -BCll/

rmSolid State Assigned based on rule 7 E-BC/A2 Battery Charger E-CI-1A to W

467 Charger and 0.59 Functional of the box. Parent 7 E-BC/A2 Div. 1 125VDC PANEL E-PNL-Ivre opnn:EC1l Frame-75 E-PNL-C1/1 125VDC PANEL E-PNL-C1/1 W

467 Mounted 0.77 Functional Earthquake Experience

_______Panel Data Breaker E-GB-Cl1/lA from Div.

Frame-Assigned based on rule 76 VCEPAEL-EPNL-l/1 W

467 Mounted 0.77 Functional cmoet

-B 76 BC-

/A to Div. 1 l25VDC MAIN DIST Panel cmoet

-B PANEL E-DP-Sl1/1 Cl/lA Distribution Generic Equipment 125VDC MAIN DIST PANEL W

47 Panels -

77 ED-1/

-PSI1W 47 Wall/Floor 1.17 Functional Ruggedness Spectra 77 BD-1l ED-llMounted (GERS) 78 E-MC-S 1/1D 125 VDC Bus MCC E-MC-W 467 W MCCs 0.58 Functional Earthquake Experience S1/iD

______Data Wallte Generic Equipment 79 E-PS11D 15VC u Pne -D-S/D 47 Distribution 0.64 Functional Ruggedness Spectra Panel (GERS)

Wallte Generic Equipment 80 E-PS11A 15VC u Pne

-D-S/A 51 Distribution 0.64 Functional Ruggedness Spectra Panel (GERS) 48D o CCV2CnrlControl Assigned based on rule Room of the box. Parent 81 RCIC-E/S-603 Power (in panel E-CP-W 501 Vetcl0.66 Functional cmoet

-P H/61)Cabinets

_HI 13/P612 R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 67 of 99

'rIpZZO

SUMMARY

OF ESEL HCLPF VALUES (CONTINUED)

ITEM# EQUIPMENT ID DESCRIPTION BLDG (F)J GROUP HCLPF (g)

FAILURE MODE ]EVALUATI METHOD DIV 1 CRITICAL POWER Solid State Generic Equipment 82 E-IN-3A SUPPLY INVERTER A E-IN-W 467 Charger and 0.59 Functional Ruggedness Spectra 3A Inverter (GERS) 250V STATION BATTERY E-Generic Equipment 83 E-B2-1 B2-1 W

467 Batteries 0.73 Functional Ruggedness Spectra (GERS)

Assigned based on rule 84 CB1 Breaker CB1 from E-MC-7A to W

467 Battery 0.84 Functional of the box. Parent Battery Charger E-C2-1 Charger component: E-C2-1 85 E-C2-1 E-B2-1 VITAL BATTERY W

467 Battery 0.84 Functional Seismic Qualification CHARGER E-C2-1 Charger

_____Report Breaker CB2 from Battery Asge ae nrl 86 CB2 Charger E-C2-1 to Div. 1 Ch47 argery 0.4Fnto a

ssignted basdon.

P ruen 250 VDC MAIN DIST PANEL WC47 hattery08Fucinl othbx.Pet E-DP-S2/1 component: E-C2-1 250VDC Bus MCC E-MC-Generic Equipment 87 E-MC-S2/1A-A S/AR 471 R MCCs 0.65 Functional Ruggedness Spectra 52/lA(GERS)

Assigned based on rule 88 E-DISC-7A3B 120 VAC E-PP-7A Supply W

467 W MCCs 0.58 Functional of the box. Parent component: E-MC-7A Distribution E-DISC-Paes-Assigned based on rule 89 P7A5 -P-A Spl W

47 Wall/Floor 1.17 Functional of the box. Parent Mounted component: E-PP-7A ControlAssigned based on rule 90 Control Switch for RCIC-V-10 W

501 Room 0.66 Functional o h o.Prn RC/10 MS Benhboad component: E-CP-encoarH13/P601 ControlAssigned based on rule 91 C-M-otrlof the box. Parent RC91RS-Control Switch for RCIC-V-3 1 W

501 Room 0.66 Functional cmoet

-P V/31BehbadcmoetE-P BenchoardH13/P601 R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 68 of 99

~i!iV

" RIZZO

SUMMARY

OF ESEL HCLPF VALUES (CONTINUED)

ITEM T I

ELEV TE# EQUIPMENT ID IDESCRIPTION BLD]i (FT)

GROUP

]HCLPF (g)

FAILURE MODE EVALUATION METHOD 92 RCIC-RMS-RCIC Manual Initiation WC0 ontom 0.6sFuctinal of the box.baeparent~nrl 92 6

Pushbutton.6Funchbna component: E-CP-

$6PsbtoBenchboard H 13/P601 M-M-Control Assigned based on rule MS-RMS-of the box. Parent 93

/D/CControl Switch for MS-RV-3D W

501 Room 0.66 Functional component: E-CP-R/DCBenchboard H 13/P601 ControlAssigned based on rule MS-RMS-Control wthfrM-V4 0

om06 ucinl of the box. Parent 94CotolSith4oAM-V-AC 01 Rom0hbducioa component: E-CP-RV4/

enchboard H 13/P601 MS-RMS-Control Assigned based on rule 95 Control Switch for MS-RV-4B W

501 Room 0.66 Functional o h o.Prn RV/4B/C Benchboard component: E-CP-H13/P601 ControlAssigned based on rule 96 RM-Cotrlof the box. Parent 96

-MS-Control Switch for MS-RV-4C W

501 Room 0.66 Functional cmoet

-P RV/4C/C Benchboard cmoet

-P

___________H13/P601 ControlAssigned based on rule 97 RM-Cotrlof the box. Parent M97MS-Control Switch for MS-RV-4D W

501 Room 0.66 Functional cmoet

-P RV/4D/CBecbadomoetE-P BenchoadH 13/P601 ControlAssigned based on rule 98 M-M-Control Switch for MS-RV-5B W

501 Room 0.66 Functional o h o.Prn RV/5B/C Benhboad component: E-CP-encoarH13/P601 ControlAssigned based on rule MS-RMS-oto of the box. Parent 99

/C/CControl Switch for MS-RV-5C W

501 Room 0.66 Functional cmoet

-P BenchoardH13/P601 R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 69 of 99 r,~RIzzo

SUMMARY

OF ESEL IICLPF VALUES (CONTINUED)

ITEM#J EQUIPMENT ID f DESCRIPTION BLDG1 (LE)

GROUP HCLPF (g) [FAILURE MODE JEVALUAT METHOD Accumulator Tank for Div. 1AD 10 M-K-V oenidAfo ASSR 3 58Accumulators

> 0.50 Structural Screened > 0.50g PGA Accumulator Tank for Div. 1AS 101 MS-TK-3S Solenoid A for ADS SRV 4A C

548 ADSmltr

.0Srcurl Sree

.0 G

10 M-K-R oenidAfo ASSR 4 58Accumulators

> 0.50 Structural Screened > 0.50g PGA Accumulator Tank for Div. 1AD 102 MS-TK-3M Solenoid A for ADS SRV 4B C

548 ADSmltr

.0Srcurl Sree

.0 G

10 M-K-P oenidAfo ASSR 4 58 Accumulators

> 0.50 Structural Screened > 0.50g PGA Accumulator Tank for Div. 1AS 10 M-K-U oenidAfo ASSR 5 58Accumulators

> 0.50 Structural Screened > 0.50g PGA Accumulator Tank for Div. 1AS 10 M-K-N oenidAfo ASSR 5 58 Accumulators

> 0.50 Structural Screened > 0.50g PGA 10 RI-P-RI P1SutAccusur 42 Moulators

> 0.50 Stuncturnal Screened> duringPG Accmitr ntumulatos wa.0ltucual Srend>owns gP Control EC-RoomEatqaeEprec 108 P-Termination Cabinet TCG1 W

501 0.66 Functional Eatqkexprnc 18H13/P894 Termination Data Cabinets Control 109 E-CP-Process Instrument Termination W

501 Room 0.66 Functional Earthquake Experience H 13/P682 Cabinet Termination Data Cabinets 110 E-PNL-IN/3 Div 1 Static Transfer Switch Fra67 Mone-

.7Fntoa Earthquake Experience Panel

~Panel Dt Assigned based on rule 111 E-DISC-7A4A Disconnect E-DISC-7A4A W

467 W MCCs 0.58 Functional of the box. Parent

_________cornponent:

E-MC-7A R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 70 of 99 EiMu RIzz 0'~1

SUMMARY

OF ESEL HICLPF VALUES (CONTINUED)

ITE fEQIPENTIDT ESRIPIO

[LD '(FT)

GROUP ]HCLPF (g)

FAILURE MODE EVALUATION METHOD Wall-112 E-TR-7A12 1 5KVA Regulating XFMR for W

467 Mounted 0.77 Functional Earthquake Experience E-PP-7AF (Remote Shutdown Trnfomr Data 7KEXM foCrtclFloor Generic Equipment 113 E-TR-7A Power PNL E-PP-7A W

467 Mounted 0.63 Functional Ruggedness Spectra

___Transformer

______(GERS)

WallGeeiEqimn 1 EP AF 120/241 v 225a Remote MountedGeriEqpmn 11

-P-7A ShtonPwrPnlW 47 Distribution 0.64 Functional Ruggedness Spectra Panel (GERS)

Wall B-DISC MutdAssigned based on rule 115 F/9Disconnect B-DISC PP7AF/9 W

467 Mountedonl ofte ox arn 115 F/9Distribution 06 ucinl o h o.Prn Panel component: E-PP-7AF Unistrut MountedScendurg 116 E-DISC-S 11D1 Fusible Safety Switch W

467 Screenedunduring Distribution

>05Fucinl walkdowns

_____Panel Distribution Generic Equipment 250VDC Main Distribution W

47 Panels -

117 B-DP-S2/1 Panel E-DP-S2/1 W

47 Wall/Floor 1.17 Functional Ruggedness Spectra Mounted (GERS)

Wall-118 E-TR-IN/3 Inverter Alternate Power Supply W

467 Mounted 0.77 Functional Earthquake Experience Transformers Data DistributionGeeiEqpmn 19 -P-ARwatCrtclPwrPnl W

47 Panels -GeeiEqpmn 11 EP-7 Rdase rtialPwe Pnl 47 Wall/Floor 1.17 Functional Ruggedness Spectra Mounted (GERS) 120 B-TB-B 1/1 Terminal Box for Battery B-B 1-W 467 Terminal

> 0.50 Functional Screened during 1

Boxes

______________walkdowns 121 E-TB-B2/1 Terminal Box for Battery E-B2-W 467 Terminal

> 0.50 Functional Screened during

_______1

____Boxes

______________walkdowns R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 71 of 99

'5R5Z 0

[

SUMMARY

OF ESEL HCLPF VALUES (CONTINUED)

ITEM# EQUIPMENT ID [DESCRIPTION BLDG1 (FT)

GROUP HCLPF (g) [FAILURE MODE JEVALUATI METHOD 250VDC Bus MCC E-MC-Generic Equipment 122 E-MC-S2/1A-B 2/AR 471 R MCCs 0.65 Functional Ruggedness Spectra 52/lA(GERS) 13 2-MS-RLY-2-MS-RLY-ADK38 Relay forSesiQuiiato ADK38 MS-RV-1A (Solenoid C)

W 501 ADS Relays 0.62 Relay Chatter Report 124 2-MS-RLY-2-MS-RLY-ADK30 Relay for Seismic Qualification ADK30 MS-RV-1B (Solenoid C)

W 501 ADS Relays 0.62 Relay Chatter Report 125 2-MS-RLY-2-MS-RLY-ADK42 Relay for Seismic Qualification ADK42 MS-RV-1C (Solenoid C)

W 501 ADS Relays 0.62 Relay Chatter Report 16 2-MS-RLY-2-MS-RLY-ADK40 Relay for Seismic Qualification 126_

ADK40 MS-RV-1D (Solenoid C)

W 501 ADS~ Relays 0.62 Relay Chatter Report 17 2-MS-RLY-2-MS-RLY-ADK22 Relay for W

51 ASRly062eayCter Seismic Qualification 17 ADK22 MS-RV-2A (Solenoid C)

W 51ASRly062eayhter Report 18 2-MS-RLY-2-MS-RLY-ADK24 Relay for W

51 ASRly062eayCter Seismic Qualification 128_

ADK24 MS-RV-2B (Solenoid C)

W 51ASRly062eayCter Report 19 2-MS-ELY-2-MS-RLY-ADK36 Relay for Seismic Qualification 129_

ADK36 MS-RV-2C (Solenoid C)

W 501 ADS Relays 0.62 Relay Chatter Report 10 2-MS-ELY-2-MS-RLY-ADK34 Relay for Seismic Qualification 130_

ADK34 MS-RV-2D (Solenoid C)

W 501 ADS Relays 0.62 Relay Chatter Report 11 2-MS-RLY-2-MS-RLY-ADK28 Relay for W

51 ASRly0.2eayCter Seismic Qualification 11ADK28 MS-RV-3A (Solenoid C)

W 51ASRly062eayCter Report 12 2-MS-RLY-2-MS-RLY-ADK32 Relay for Seismic Qualification 32 ADK32 MS-RV-3B (Solenoid C)

W 501 ADS Relays 0.62 Relay Chatter Report 13 2-MS-ELY-2-MS-RLY-ADK26 Relay for Seismic Qualification 133_

ADK26 MS-RV-3C (Solenoid C)

W 501 ADS Relays 0.62 Relay Chatter Report 134 2-MS-ELY-2-MS-RLY-ADK100 Relay for Seismic Qualification

___ADK100 MS-RV-3D (Solenoid C)

W 501 ADS Relays 0.62 Relay Chatter Report 135 2-MS-ELY-2-MS-RLY-ADK93 Relay for Seismic Qualification

___ALDK93 MS-RV-4A (Solenoid C)

W 501 ADS Relays 0.62 Relay Chatter Report 16 2-MS-ELY-2-MS-RLY-ADK87 Relay for Seismic Qualification

__3_

ADK87 MS-RV-4B (Solenoid C)

W 501 ADS Relays 0.62 Relay Chatter Report 2-MS-ELY-2-MS-RLY-ADK98 Relay for Seismic Qualification 137 D98M-V4(SlniC)W 501 ADS Relays 0.62 Relay Chatter Report R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 72 of 99 E~iN F'~ RIZZO

SUMMARY

OF ESEL HCLPF VALUES (CONTINUED)

ITEM T

I' BLDG1 ELEVT TE# EQUIPMENT ID DESCRIPTION BLGIFT)j GROUP jHCLPF (g)

FAILURE MODE EVALUATION METHOD 2-MS-RLY-2-MS-RLY-ADK91 Relay for Seismic Qualification 138 AK1M--4(SeniC)W 501 ADS Relays 0.62 Relay Chatter Report 392-.MS-RLY-2-MS-RLY-ADK95 Relay for Seismic Qualification 139 S-V5 (oeoi

)W 501 ADS Relays 0.62 Relay Chatter Report 2-MS-RLY-2-MS-RLY-ADK89 Relay for Seismic Qualification 140 A

89M-V5(SlniC)W 501 ADS Relays 0.62 Relay Chatter Report 112-MS-RLY-2-MS-RLY-ADK1A Relay for Seismic Qualification 141 S-V3 (oeoi

)W 501 ADS Relays 0.56 Relay Chatter~ Report

____ADK8A MS-RV-3D (Solenoid A)

W_51__SReays_.6_elyCater Reor 132-MS-RLY-2-MS-RLY-ADK6A Relay for Seismic Qualification 142 DSS~ (oeoi

)W 501 ADS Relays 0.56 Relay Chatter Report 143 2-MS-RLY-2-MS-RLY.-ADK7A Relay for W

51ASRly 0.6eayCter Seismic Qualification ADK6A ADS SRVs (Solenoid A)

W 51ASRly 0.6eayCter Report 452-MS-RLY-2-MS-RLY-ADK1A Relay for Seismic Qualification 144 S-V3 (oeoi

)W 501 ADS Relays 0.56 Relay Chatter Report 16ADK7A ADS-RVs3 (Solenoid A)

W 51ASRly

.6 RlyCatr Rpr 2-MS-RLY-2-MS-RLY-ADK1B Relay for Seismic Qualification 145 AD6BASSRVs3 (Solenoid B)

W 501 ADS Relays 0.56 Relay Chatter Report 182-MS-RLY-2-MS-RLY-ADK7B Relay for Seismic Qualification 147 DSS~ (oeoi

)W 501 ADS Relays 0.56 Relay Chatter Report 149 TB-R322 Terminal Box outside R

Terminal

> 0.50 Functional Screened during Penetration E-X-105C Boxes walkdowns 150 TB-C522 Terminal Box inside Penetration C

Terminal

> 0.50 Functional Screened during E-X-105C Boxes walkdowns Screened during 151 SB-W020 Splice Box W

Splice Box

> 0.50 Functional wados Control E-CP-Room0.6Fntoa EatqaeEprnc 152 H1/61 ADS Control Panel (Division 2)

W 501 Vertical 0.6Fntoa DarthqakeEprec

-~~ Cabinets R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 73 of 99 E~iN

~RI~Z9

SUMMARY

OF ESEL HCLPF VALUES (CONTINUED) 1 ELEV11 DESCRIPTION BLDG 1 (FT GROUP HCLPF (g)

FAILURE MODE EVALUATION METHOD Control Assigned based on rule RCIC Low Discharge Pressure W

51 Room 06Fucinl of the box. Parent Override Switch W

51 Vertical 06Fucinl component: E-CP-Cabinets H 13/P621 RCIC High Area Temperature Control Assigned based on rule Room of the box. Parent Override Test Switch LD-RMS-W 501 Vetcl0.66 Functional cmoet

-P S2A Cabinets H 13/P632 RCCHg raTmeaueControl Assigned based on rule RCCHg raTmeaueRoom of the box. Parent Override Test Switch LD-RiMS-W 501 Vetcl0.66 Functional cmoet -P S2B CabinetsH1364 Control Main Control Room Panel E-W 501 Room 0.66 Functional Earthquake Experience CP-H1 3/P632 Vertical Data Cabinets Control Main Control Room Panel E-W 501 Room 0.66 Functional Earthquake Experience CP-H 13/P642 Vertical Data Cabinets Wall-Scenddrg RCIC High Exhaust Pressure R

422 Mounted

> 0.50 Functional Screendodurns switch RCIC-PS-9A Instrumentswakwn RCIC High Exhaust Pressure Wall-oute

>05Fucinl Screened during switch RCIC-PS-9B RI2 onstruedt 05Fucinl walkdowns swith RCC-PS12AIns~ent

.79Generic Equipment RCI Hih Ehaut resure R

41 nstumet 079Functional Ruggedness Spectra sCCwigch ExhaustPressre1247 Racks (GERS) swith RCC-PS12BIns~ent

.79Generic Equipment RCI Hih Ehaut resure R

41 nstumet 079Functional Ruggedness Spectra RCCwigch ExhaustPressre1247 Racks (GERS)

R02 Columbia Generating Station ESEP 155462/1 5, Rev. A (October 27, 2015)

Page 74 of 99 i~iN pizza

SUMMARY

OF ESEL HCLPF VALUES (CONTINUED)

ITEM EQUIPMENT ID DESCRIPTION f

1f GROUP HCLPF (g)

FAILURE MODE

[EVALUATION METHOD 162 CIC-S-1C Intruent

.79Generic Equipment sw CCP-1C RI itch R

Exhaust PresB r R

471 nsrmt

-0.9Functional Ruggedness Spectra Rsw igch ExhaustPrssure2 Racks (GERS)

RCICHig Exaus Presur IntruentGeneric Equipment 163 RCIC-PS-12D RCCHg xas rsue R

471 Intuet0.79 Functional Ruggedness Spectra switch RCIC-PS-1 2D Racks (GERS)

Instrument Generic Equipment 164 E-R-02 RICIntnmen RckR 71Raks0.79 Functional Ruggedness Spectra 164

-JRP02 RCL IntruentRackR 41 Rcks(GERS)

Breaer cIC-2-S1A5CforAssigned based on rule 165 C42-of the box. Parent 15 S2lA5C power to RCIC Minimum Flow R

471 R MCCs 0.65 Functional component: E-MC-Valve RCIC-V-19

$2/lA-B Note:

R=Reactor Building, C =Primary Containment, W = Radwaste Building, D = Diesel Generator Building R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 75 of 99

"'*RIZZO otrs

ATTACHMENT C WALKDOWN TEAM QUALIFICATIONS R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

Page 76 of 99 r' RIZZO

F'R I ZZO Adam Helifrich, P.E.,,,,,

Rim*

mBlm Skill Areas:

Civil Engineering Seismic Walk downs Finite Element Analysis Computer Programming Mr. Adam Helifrich, is a Project Engineer with RIZZO Associates (RIZZO).

He recently received his Bachelor of Science in Civil Engineering from the University of Pittsburgh. Since joining RIZZO, Adam has emerged as a strong member in the field of Finite Element Modeling and Analysis. He is also qualified to perform Seismic walkdown reviews of existing NPP's who are considering fragility analyses. To supplement seismic walkdown experience, Mr. Helffrich has completed the SQUG Walkdown Screening and Seismic Evaluation Training Course, along with the NTTF 2.3 Seismic Walkdown Training.

Mr. Helifrich is currently enrolled in the University of Pittsburgh's Civil and Environmental Engineering department as he pursues his Master of Engineering degree in Structural Engineering.

Watts Bar Nuclear Power Plant 1 & 2 SPRA and ESEP Evaluation URS I Spring City, Tennessee Mr. H-elffrich, as a project engineer, is currently performing tasks in support of Watts Bar's SPRA Evaluation, as well as development of the ESEP Evaluation for submittal to the NRC. He has performed seismic walkdowns of the plant and is currently overseeing the SSI analysis of onsite buildings, fragility analyses of all plant components used in the SPRA, and development of calculations and reports for the submittal of the plants ESEP evaluation required by the NRC.

FERMI 2 Seismic Fragility Evaluation URS I Monroe, Michigan Mr. Helffrich, as a Project Engineer, is currently engaged in performing seismic evaluations of plant structures and components in support of developing seismic fragilities and the seismic probabilistic risk assessment (SPRA). This includes the development of Finite Element Models of several buildings in order to develop seismic response of these buildings at all elevations. As part of this effort, he has also performed the NTTF 2.1 and SPRA walkdowns to ensure plant operability during and after a seismic event. He is currently performing the fragility calculations associated with updating the SPRA of the plant.

Page 77 of 99

F'%

Adam Helifrich, P.E.

CSRF SSl Analysis Department of the Navy I United States Mr. Helifrich is an Engineer, and has completed a qualification process for the CSRF facility owned by the United States Navy. He has developed an accurate SSI model to analyze the CSRF structure itself, along with several adjacent storage slabs. In Structure Response Spectra (ISRS) were developed at various locations within the structure and on the adjacent slabs. This information was used as the input to a non-linear analysis of a rocking and sliding of the storage containers to ensure no impact or tipping occurs in the facility.

Beaver Valley Unit I NPP Seismic PRA ABS Consulting I FirstEnergy Nuclear Operating Company I Shippingport, Pennsylvania Mr. Helffrich was engaged in performing seismic evaluations of plant structures and components in support of developing seismic fragilities and the seismic probabilistic risk assessment (SPRA). As part of this effort, he has also performed the NTTF 2.3 walkdowns and has completed SQUG training for qualification. In addition, he also participated in the NRC audit of the NTTF 2.3 evaluation at the plant.

Beaver Valley Unit 2 NPP Seismic PRA ABS Consulting I FirstEnergy Nuclear Operating Company j Shippingport, Pennsylvania Mr. Helffrich was engaged in performing seismic evaluations of plant structures and components in support of developing seismic fragilities and the seismic probabilistic risk assessment (SPRA). As part of this effort, he has also performed the NTTF 2.3 walkdowns and has completed SQUG training for qualification.

In addition, he also developed some building models for analysis at the plant. These models have since been used in SSI analysis and fragility evaluation.

Davis-Besse NPP Seismic PRA ABS Consulting I FirstEnergy Nuclear Operating Company I Oak Harbor, Ohio Mr. Helffrich was engaged in performing seismic evaluations of plant structures and components in support of developing seismic fragilities and the seismic probabilistic risk assessment (SPRA). As part of this effort, he has also performed the NTTF 2.3 walkdowns and has completed SQUG training for qualification.

DNFSB Work for Various Sites DNFSB I United States In response to DNFSB request, Mr. Helifrich conducted a non-linear cracking analysis of a concrete roof section.

Adam investigated the cracking potential of the roof as it relates to various depths of roof, and rebar sizes.

KKL Leibstadt NPP Fragility Analysis KKLIAXPO I Switzerland As part of the fragility analysis for the Leibstadt Nuclear Power Plant in Switzerland, As in Engineer, Mr. Helffrich has developed full FEA models of major buildings, and has completed the seismic Soil Structure Interaction analysis. The ISRS are being utilized in the safety evaluation of equipment housed in the buildings. Additionally, Mr. Helifrich has performed walkdowns of the several systems such as the filtered containment vent system and the fuel pool cooling and cleanup system.

Page 78 of 99

Ip Adam Helifrich, P.E.

%,J SSM Seismic Fan Qualifications SSM I United States As an Engineer, Mr. Helffrich has constructed and analyzed several fan models for qualification purposes in various seismically controlled environments such as, nuclear power plants, chemical facilities, and water treatment plants.

Finite Element Models were used to simulate seismic conditions and ensure adequacy of the fan units under varying seismic demands and code standards.

SSM VBS Seismic Analysis Duct Analysis SSM I United States As an Engineer, Mr. Helffrich developed the FEA model of the Duct system and ran the analysis to assure the client that the duct system is safe for operation under specified seismic loads.

UAE Site A (Alternate) NPP Site Selection/Site CharacterizationlPSAR and EIA ENEClKEPCO E&C I United Arab Emirates:

As an intern, Mr. Helffrich developed and reviewed boring logs for both sites; constructed drawings of cross-sections for a site; and performed several checks and modifications to figures and slides for presentation purposes.

Calvert Cliffs NPP Unit 3 UniStar I Chesapeake Bay, Maryland As an intern, Mr. Helffrich was responsible for cutting several cross-sections of the sub surface for analysis purposes based on the boring logs that were taken of the site.

PREVIOUS EXPERIENCE:

PennDOT Clearfield, Pennsylvania Conducted STAMPP program for roadway safety Worked independently and unsupervised through several counties Studied technical diagrams of roadways and foundations Applied gathered knowledge in roadway safety reports Page 79 of 99

O0Q 0

00

',0

~R I ZZO Brian IA. Lucarelli, EI.T.

A,

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A-El Skill Areas:

Seismic Fragility Evaluations Seismic Walkdown Inspection Soil Mechanics Roller Compacted Concrete Construction Materials Testing Quality Assurance Mr. Lucarelli has experience in seismic walkdown inspections of operating nuclear plants and seismic fragility evaluations of structures, systems, and components. He has attended the 5-day SQUG Walkdown Screening and Seismic Evaluation Training Course and has also provided support during peer reviews to the ASME/ANS PRA Standard.

Mr. Lucarelli also has experience in geotechnical modeling, structural modeling, and quality control in support of applications for proposed nuclear plants.

Watts Barr NPP Seismic Scoping Study URS Consulting I TVA I Rhea County, Tennessee As an Engineering Associate, Mr. Lucarelli has been engaged in performing seismic evaluations of plant structures and components in support of developing seismic fragilities for the seismic PRA. As part of this effort, Mr. Lucarelli was part of the Seismic Walkdown Team. He was responsible to perform the NTTF 2.1 Seismic Walkdown and Equipment Screening and to perform walkdowns in support of the Expedited Seismic Evaluation Process (ESEP).

Mr. Lucarelli also developed seismic fragilities for miscellaneous components such as the Polar Crane, Steel Containment Vessel Penetrations, and Control Room Ceiling.

Perry NPP Seismic PRA ABS Consulting j FirstEnergy Nuclear Operating Company I Perry, Ohio As an Engineering Associate, Mr. Lucarelli has been engaged in performing seismic evaluations of plant structures and components in support of developing seismic fragilities for the seismic PRA. As part of this effort, Mr. Lucarelli was part of the Seismic Walkdown Team. He was responsible to perform the NTTF 2.1 Seismic Walkdown and Equipment Screening. He was also responsible to perform the NTTF 2.3 Seismic Walkdown and walkdowns in support of the Expedited Seismic Evaluation Process (ESEP). Mr. Lucarelli managed the development of equipment fragilities for PNPP and acted as the point of contact between the team of fragility analysts and the PRA analyst developing the logic model.

Mr. Lucarelli participated in the Peer Review of the PNPP Seismic PRA in support of the work related to walkdowns and equipment fragilities. As part of the PNPP Peer Review, Mr. Lucarelli engaged in the direct response of comments from peer reviewers as well as technical discussions regarding compliance with the ASME Standard.

Page 81 of 99

Ip Brian A. Lucarelli, E.I.T.

J' Beaver Valley Unit I NPP Seismic PRA ABS Consulting I FirstEnergy Nuclear Operating Company I Shippingport, Pennsylvania As an Engineering Associate, Mr. Lucarelli has been engaged in performing seismic evaluations of plant structures and components in support of developing seismic fragilities for the seismic PRA. As part of this effort, Mr. Lucarelli was part of the Seismic Walkdown Team and was responsible to perform the NTTF 2.1 Seismic Walkdown and Equipment Screening. Mr. Lucarelli performed walkdowns in support of the Expedited Seismic Evaluation Process (ESEP).

Beaver Valley Unit 2 NPP Seismic PRA ABS Consulting I FirstEnergy Nuclear Operating Company I Shippingport, Pennsylvania As an Engineering Associate, Mr. Lucarelli has been engaged in performing seismic evaluations of plant structures and components in support of developing seismic fragilities for the seismic PRA. As part of this effort, Mr. Lucarelli was part of the Seismic Walkdown Team. He was responsible to perform the NTTF 2.1 Seismic Walkdown and Equipment Screening. He was also responsible to perform the NTTF 2.3 Seismic Walkdown. Mr. Lucarelli performed walkdowns in support of the Expedited Seismic Evaluation Process (ESEP).

Davis-Besse NPP Seismic PRA ABS Consulting I FirstEnergy Nuclear Operating Company I Oak Harbor, Ohio As an Engineering Associate, Mr. Lucarelli has been engaged in performing seismic evaluations of plant structures and components in support of developing seismic fragilities for the seismic PRA. As part of this effort, Mr. Lucarelli was part of the Seismic Walkdown Team. He was responsible to perform the NTTF 2.1 Seismic Walkdown and Equipment Screening. He was also responsible to perform the NTTF 2.3 Seismic Walkdown. Mr. Lucarelli performed walkdowns in support of the Expedited Seismic Evaluation Process (ESEP).

Visaginas NPP Units 3 and 4 Visagino Atomine Elektrine UAB I Villnius, Lithuania As an Engineering Associate, Mr. Lucarelli Evaluated cone penetration test (OPT) data to evaluate site uniformity, provide recommended elastic modulus values for geologic layers, and evaluate dissipation test results to determine the coefficient of consolidation for geologic layers.

Vogtle NPP Geotechnical Investigation Westinghouse Electric Company I Burke County, Georgia RIZZO conducted a settlement analysis to predict the total and differential settlements expected during construction of the Vogtle Units. Mr. Lucarelli was responsible for reviewing on-site heave and settlement data and the excavation sequence to calibrate the material properties in the settlement model.

He was also responsible for creating a settlement model that implemented the expected AP1000 construction sequence and presenting the results in a report.

Levy County NPP Foundation Considerations Sargent & LundylProgress Energy I Crystal River, Florida Mr. Lucarelli has been extensively involved in the design and specification of the Roller Compacted Concrete (RCC) Bridging Mat that will support the Nuclear Island foundation. He authored numerous calculations and reports related to the work for this project, including responding to Requests for Additional Information from the NRC. He performed finite element analyses of the stresses within the Bridging Mat under static and dynamic loading conditions, evaluation of whether the stresses in the Bridging Mat met the applicable requirements of ACl 349 and A~l 318, and the determination of long-term settlement. As part of laboratory testing program for RCC, Mr. Lucarelli assisted in the evaluation, selection, and testing specification for the concrete materials to ensure they met the applicable ASTM material standards. He also authored the Work Plan and served as on-Page 82 of 99

pI' Brian A. Lucarelli, E.LT site quality control during laboratory testing of RCC block samples in direct tension and biaxial direct shear. His responsibilities included inspection of the testing being performed, control of documentation related to testing activities, and ensuring subcontractors fulfilled the requirements of RIZZO's NQA-1 Quality Assurance Program.

Blue Ridge Dam Rehab Tennessee Valley Authority I Fannin County, Georgia RIZZO conducted a deformation analysis of the downstream side of the Blue Ridge Dam to assess the observed movement in the Mechanically Stabilized Earth (MSE) wall. Mr. Lucarelli prepared a two dimensional finite element model of the dam, which included reviewing construction documentation and instrument readings to determine cross sectional dimensions and material properties.

Akkuyu NPP Site Investigation WorleyParsons I Mersin Province, Turkey RIZZO conducted a geotechnical and hydrogeological investigation of the proposed site for four Russian WER-1200 reactors. This investigation entailed geotechnical and hydrogeological drilling and sampling, geophysical testing, and geologic mapping.

Mr. Lucarelli served as on-site quality control for this project.

His responsibilities included controlling all records generated on site, interfacing with TAEK (Turkish Regulatory Agency) auditors, and tracking nonconformance observed during the field investigation in accordance with RIZZO's NQA-1 Quality Assurance Program.

Mr. Lucarelli also assisted in the preparation of the report summarizing the findings of the field investigation.

Calvert Cliffs NPP Unit 3 Unistar ICalvert County, Maryland RIZZO completed a COLA-level design of the Ultimate Heat Sink Makeup Water Intake Structure at the Calvert Cliffs site. Mr. Lucarelli authored and checked calculations to determine the design loads, as prescribed by ASCE 7, to be used in a Finite Element model of the structure. Mr. Lucarelli was also responsible for ensuring that the design met the requirements of the Design Control Document.

Mr. Lucarelli also performed a settlement analysis for the Makeup Water Intake Structure.

Areva RAI Support Services for U.S. EPR Design Certification AREVA Mr. Lucarelli assisted in the calculation of the subgrade modulus distribution for the foundation of the Nuclear Auxiliary Building (NAB) for the U.S. Evolutionary Power Reactor (U.S. EPR). This iterative process included modeling subsurface profiles in DAPSET to obtain a soil spring distribution under the basemat. The soil spring distribution was then modeled in GTSTRUDL as the basemat support.

C.W. Bill Young Regional Reservoir Forensic Investigation Confidential Client I Tampa, Florida RIZZO conducted a forensic investigation into the cause of soil-cement cracking on the reservoir's upstream slope. This investigation involved a thorough review of construction testing results and documentation to determine inputs for seepage and slope stability analyses. Mr. Lucarelli reviewed construction documentation and conducted quality control checks on the data used for the analyses. Mr. Lucarelli also prepared a number of drawings and figures that presented the results of the forensic investigation.

PREVIOUS EXPERIENCE Page 83 of 99

Ip Brian A. Lucarelli, E.I.T.

%-4 Aquaculture Development Makili I Mali, Africa As the project coordinator, his primary responsibilities included maintaining a project schedule, developing a budget for project implementation, and coordinating technical reviews of project documentation with a Technical Advisory Committee.

The University Of Pittsburgh Chapter Of Engineers Without Borders designed and constructed an aquaculture pond in rural Mali, Africa with a capacity of 3.6 million gallons. This pond is designed to maintain enough water through a prolonged dry season to allow for year-round cultivation of tilapia. As the project technical lead, Mr.

Lucarelli was involved in developing conceptual design alternatives and planning two site assessment trips.

These scope of these site assessment trips included topographic surveying, the installation of climate monitoring instrumentation, soil sampling and characterization, and laboratory soils testing.

Southwestern Pennsylvania Commission Pittsburgh, Pennsylvania As a transportation intern, Mr. Lucarelli analyzed data in support of various studies dealing with traffic forecasting, transit use, and highway use. He also completed fieldwork to assess the utilization of regional park-and-ride facilities.

Page 84 of 99

0 O0 Presents this Certfifca te ofA Achi'evemen t Tob Certify 'That J

has Compfeted~the SQUq Walkdown Screening andSeismic E&valuation 'Training Course flfeCd?!uust 2 0-24, 2012 SQUG Instn~ctor S5ljJ Injctru I IIII I I I I IJIIIIII I I II II II I I I

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LAWRENCE LEE Manager II Probabilistic Safety Assessment and Reliability AREAS OF EXPERTISE Probabilistic Safety Assessment for Internal and External Hazards Fault Tree Analysis Event Tree Analysis Shutdown Safety On-line Maintenance Severe Accident Analysis Maintenance Rule Risk In formed Regulatory Compliance Equipment Survivability EDUCATION B.S., Mechanical Engineering, University of California, Berkeley SECURITY CLEARANCE U.S.. Citizen WORK EXPERIENCE

SUMMARY

Mr. Lee is employed as a Manager With ERIN Engineering and Research, Inc. He has over twenty years of experience in the nuclear field specializing in Probabilistic Safety Assessment.

Mr. Lee has experience in leading Level 1 and Level 2 PSA updates (internal and external events), Maintenance Rule implementation, shutdown safety assessment, On-line Maintenance, In-Service inspection of

piping, MOV prioritization, AOV prioritization, and utility response to NRC compliance using PSA techniques.

WORK EXPERIENCE Mr. Lee holds a Bachelor of Science degree in Mechanical Engineering from the University of California, Berkeley.

He is responsible for leading Probabilistic Safety Assessment (PSA),

Maintenance Rule Implementation, Shutdown Safety Assessment, and On-line Maintenance projects.

His areas of PSA expertise include fault tree and event tree analysis, thermal-hydraulic evaluations using the Modular Accident Analysis Program (MAAP) code, and containment safety studies during severe accident conditions.

His recent experience includes external hazard PSA, including seismic PSA development.

Mr. Lee has performed plant walkdowns, system fault tree analysis, accident sequence analysis, and probabilistic fragility analysis to support initial seismic PRA models for Limerick, Dresden, Quad Cities, LaSalle, and Oyster Creek.

He has also supported projects for the risk-informed prioritization to perform plant specific fragility calculations.

He has also used the EPRI ACUBE software to quantify the seismic CAFTA models for the initial Exelon SPRAs as well as for the KKM SPRA.

Mr. Lee has been an instructor for the EPRI Seismic PRA training class.

In addition, he has provided technical oversight for the update of the EPRI Seismic PRA Implementation Guide.

Mr. Lee was a member of the Palo Verde Seismic PRA peer review team using the ASME/ANS Combined PRA Standard guidelines as supported by the NEI 12-13 External Hazards PRA Peer Review Process Guidelines.

Mr. Lee was the lead for the Seismic Plant Response technical element.

To support other SPRA experience, Mr. Lee participated in the update of the Columbia Generating Station (CGS) Seismic PRA (SPRA) in 2004 using EQESRA to perform a seismic convolution to calculate the frequency of discrete seismic damage states.

Risk insights from the SPRA were used to support. the CGS License Renewal project.

In addition, Mr. Lee supported -the incorporation of Hope Creek seismic IPEEE sequences into the Hope Creek Level 1 and Level 2 PRA models to support risk insights for PRA applications.

Mr. Lee was the project manager for an EPRI pilot study to perform a Probabilistic Flood Hazard Assessment (PFHA).

The use of enhanced PFHA methodologies to calculate more defensible external flood initiating event frequencies will serve to support the development of external flood PRA models and applications.

Page 86 of 99 0/91 o*/i 9/I 5

LAWRENCE LEE Page 2 PROFESSIONAL ORGANIZATIONS American Society of Mechanical Engineers American Nuclear Society Mr. Lee served as lead investigator for the EPRI BWR and PWR Pilot Spent Fuel Pool (SFP)-Reactor risk assessment studies.

The pilot project developed a methodology to evaluate SFP risk and the potential interaction between severe SFP and reactor accident scenarios.

The methodology was implemented for a pilot plant to quantify the SFP-Reactor risk for 1) internal and external events hazards, 2) at-power and shutdown operating modes, and 3) Level 1 and Level 2 end states.

Mr. Lee has been the technical lead in the updates of the Level 1 and Level 2 PSAs for Quad Cities, Dresden, LaSalle, Clinton, Oyster Creek, Hope Creek, and Columbia Generating Station.

These projects included update and documentation of system models, accident sequence analysis, and system notebooks to incorporate plant specific and BWR design basis data.

His PSA experience also includes contributions to the Peach Bottom, Limerick, Nine Mile Point Units 1 and 2, Vermont Yankee, and Duane Arnold Level 2 IPE projects.

Mr. Lee led the development and evaluation of Station Blackout accident scenarios, thermal-hydraulic calculations, and coping times for Duane Arnold and Oyster Creek in in responding to INPO IER Li-11-4 in response to the Fukushima Daiichi event.

Mr. Lee serves as the technical lead for providing risk assessment information to the Quad Cities and Dresden exte'rnal flood Integrated Assessments (IA) in response to the NRC's March 12, 2012, 50.54(f) request for information.

Mr. Lee has participated in developing Internal Flooding PSA models for Quad Cities, Dresden, LaSalle, Clinton, Oyster Creek, Hope Creek, Fermi-2, and Duane Arnold.

Mr. Lee has experience in developing Severe Accident Management Alternatives (SAMA) risk evaluations to support the licensing extension submittals, including those for Quad Cities, Dresden, Oyster Creek, and Columbia Generating Station.

Mr. Lee participated in the development of the STP 3 and 4 ABWR risk evaluation to support the Combined Construction and Operating License Application (COLA) submittal to the NRC.

He has experience in applying the EPRI methodology for risk-informed in-service inspection evaluation of piping systems at the Quad Cities, Dresden, LaSalle and Clinton stations. These projects included using PRA techniques and insights to identify risk important piping segments, defining the elements that are to be inspected within this risk important piping, evaluating the risk impacts of proposed changes to the inspection

program, and identifying appropriate inspection methods.

Mr. Lee has used plant specific PSA models to evaluate the risk impact of implementing Extended Power Uprate (EPU) for Quad Cities, Dresden, Clinton, Fermi, Brunswick, Hope Creek, Monticello, and Grand Gulf.

He also has evaluated the risk impact of implementing the Maximum Extended Load Line Limit Analysis+ (MELLLA+)

for Brunswick.

Mr. Lee has experience in the NET PSA Peer Review process. He was a member of the Clinton PSA Peer Review Certification team and he also participated in the Bruce B Station (Ontario Power Generation)

PSA peer review using the NET guidelines.

Mr. Lee was a member of the Columbia Generating Station (CGS) and Limerick Generating Station (LGS) PSA Peer Review Teams using ASME PRA Standard guidelines.

Page 87 of 990.//1 0.1/19/i5

Mr. Lee has extensive experience in using PSA techniques to comply LAWRENCE LEE with NRC requirements.

He hsmodified adquantified PSA models Pae3 to support utility response for MSPI, SDPs, NOEDs, LARs, exigent Page technical specifications, and Management Directive 8.3 evaluations.

In addition, he has modified plant specific PSA models in support of utility response to GL 89-10 MOV prioritization, AOV prioritization, the In-Service Testing Program, and the Maintenance Rule.

Mr. Lee has experience in using PSA techniques to support On-line Maintenance safety evaluations for the Duane Arnold, Columbia Generation Station, and Fitzpatrick On-line Maintenance Programs. In

addition, he has converted the fault tree/event tree based PSA models for the Columbia Generation Station into large fault tree models to facilitate rapid solution times for supporting On-line Maintenance safety evaluations.

He has experience in Probabilistic Shutdown Safety Assessment (PSSA).

Mr. Lee developed fault tree and event tree models for the safety analysis of Duane Arnold refueling outage RFO 12.

Mr. Lee also developed fault tree and event tree models for the Lungmen ABWR PSAR shutdown safety assessment.

In addition, he has experience using the Outage Risk Assessment and Management (ORAM) Software for the Nine Mile Point Unit 2, LaSalle, Duane

Arnold, Quad Cities, and Fermi 2 Shutdown Safety Assessment projects.

Mr. Lee has also performed an independent review of the Pickering A Risk Assessment (PARA) for the Canadian Atomic Energy Control Board (AECB).

This review included an evaluation of the PARA quantification methodology, which used the SETS and CAFTA codes to calculate the risk of fuel damage for the Pickering A CANDU reactor design.

He has extensive experience in reviewing plant operating procedures as part of various IPE, IPEEE, EPU, ORAM-SENTINEL and PARAGON projects.

As a result of these reviews, Mr. Lee has provided input to improvements in plant procedures, technical specifications and supplementary training plans.

The ORAM-SEN~TINEL and the PARAGON software model development is used to support both the probabilistic and defense-in-depth evaluation required by the Maintenance Rule.

Page 88 of 99 0/!91 oz!ikV):/

FR I ZZO Nishikant R. Vaidya, Ph.D., PE Hh1A us~,.

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Geotechnical Engineering Earthquake Engineering Civil Engineering Seismic Hazard Analysis Seismic Isolation ASME NQA-1 Structural Engineering Foundation Engineering Soil Dynamics Soil Remediation Seismic Walkdowns Dr. Vaidya has over 40 years of experience on a variety of civil/geotechnical engineering consulting projects.

His experience includes fossil fuel and nuclear power generation facilities, commercial and industrial structures, offshore structures, and equipment and piping installations. He is a recognized expert in the field of seismic isolation for nuclear power facilities. Dr. Vaidya has participated in the standards development activities of the Adl, AISE, and ASME. His fields of specialization include the analysis and design of hardened structures, soil dynamics and stability of foundation materials, stress analysis, safety evaluation, earthquake engineering, shock and vibration analysis, design of structural repair as well as retrofit, and computer applications in civil engineering. Dr. Vaidya has performed numerous analyses related to the seismic and dynamic response of building structures as well as mechanical and electrical equipment including response spectrum and time history analysis, computation of floor response spectra, equipment qualification to IEEE 344 standard, qualification of systems and components, evaluation of equipment supports and evaluation of anchorage and fragility analysis.

Dr. Vaidya has recently completed EPRI sponsored Seismic Walkdown Training. He has effectively disseminated the walkdown training to others on RIZZO's staff.

NUCLEAR PLANT PROJECTS Charleston Naval Weapons Station I Update FSAR Chapter 2

KAPL I Charleston, South Carolina USA Dr. Vaidya is currently directing the update of the SAR Sections 2.4 and 2.5 for the CNWS in accordance with NRC and DOE guidelines. This work is being performed for the Knolls Atomic Power Laboratory (KAPL-RSE) in support of KAPL-RSE plan (including, timing and resources) for updating technically significant environmental factors (e.g. seismology, meteorology) affecting the safety of the S6G Moored Training Ship.

RlZZO staff will update the hydrology, geology seismology and geotechnical information in Sections 2.4 and 2.5 to (a) include new data since 1985, and (b) reflect improved approaches and methods generated since 1985 to evaluate the new information. Dr. Vaidya is overseeing the implementation of the methodologies recommended in U.S. Nuclear Regulatory Commission {NRC) Regulatory Guides to develop site seismic hazard and the Ground Motion Response Spectra (GMRS) using current up to date seismo-tectonic information. This effort uses available site specific geotechnical information for the subgrade materials.

Page 89 of 99

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'l*O*"*l I*° FERMI 2 NPP I Seismic PRA I Fragility Evaluation URS I DTE Energy Newport, Michigan Dr. Vaidya is currently engaged in performing seismic evaluations of plant structures and components in support of developing seismic fragilities and the seismic PRA. This includes developing the PSHA and GMRS, development of Finite Element Models of several buildings, calculation of seismic response and in-structure response spectra, performing seismic fragility analyses, developing the seismic equipment list and organizing the the seismic walkdowns, and supporting the plant response models and quantification of core damage frequencies.

Gosgen NPP I Probabilistic Seismic Analysis and Walkdowns Gosgen, Switzerland At Goesgen Dr. Vaidya completed a major effort to re-evaluate the seismic fragilities of structures, equipment and distribution systems in support of the plant PSA. This effort included review and assessment of the site seismic hazard, development of scenario events and the respective response spectra, ground motion time histories, site response analysis, soil structure interaction analysis and fragility calculations.

Selected structures and components are being evaluated using nonlinear response analysis to improve their respective fragility estimates.

Perry NPP I Seismic PRA ABS Consulting FirstEnergy Nuclear Operating Company I Perry, Ohio Dr. Vaidya is engaged in performing seismic evaluations of plant structures and components in support of developing seismic fragilities and the seismic PRA. As part of this effort, he supported is part of the Seismic Walkdown Team responsible to perform the NTTF 2.3 Walkdowns for the Perry Power Station. In addition, Dr. Vaidya is part of the team responsible for the SPRA Walkdowns to be performed in compliance with the ASME ANS RA-Sa-2009 Standard and the NTTF 2.1 Recommendations.

Beaver Valley Unit I NPP Seismic PRA ABS Consulting I FirstEnergy Nuclear Operating Company Shippingport, Pennsylvania Dr. Vaidya supported seismic evaluations of plant Structures and Components in support of developing seismic fragilities and the seismic PRA. As part of this effort, he supported is part of the Seismic Walkdown Team responsible to perform the NTTF 2.3 Walkdowns for the Perry Power Station. In addition, Dr. Vaidya is part of the team responsible for the SPRA Walkdowns to be performed in compliance with the ASME ANS RA-Sa-2009 Standard and the NTTF 2.1 Recommendations.

Beaver Valley Unit 2 NPP Seismic PRA ABS Consulting I FirstEnergy Nuclear Operating Company I Shippingport, Pennsylvania Dr. Vaidya supported seismic evaluations of plant Structures and Components in support of developing seismic fragilities and the seismic PRA. As part of this effort, he supported is part of the Seismic Walkdown Page 90 of 99

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Team responsible to perform the NTTF 2.3 Walkdowns for the Perry Power Station. In addition, Dr. Vaidya is part of the team responsible for the SPRA Walkdowns to be performed in compliance with the ASME ANS RA-Sa-2009 Standard and the NTTF 2.1 Recommendations.

Davis-Besse NPP Seismic PRA ABS Consulting I FirstEnergy Nuclear Operating Company I Oak Harbor, Ohio:

Dr. Vaidya is engaged in performing seismic evaluations of plant Structures and Components in support of developing seismic fragilities and the seismic PRA. As part of this effort, Dr. Vaidya is part of the Seismic Walkdown Team responsible to perform the NTTF 2.3 Walkdowns for the Davis-Besse Power Station. In addition, Dr. Vaidya is part of the team responsible for validating the adequacy of Stick Models implemented for the determination of the dynamic response of Nuclear Safety Related Structures. Results arising from this study will validate the use of the existing IPEEE Stick Models for the seismic re-evaluation of plant structures to support the SPRA and the NTTF 2.1 submittals.

AP1000 Hydrodynamic Load Testing on Valves Dr. Vaidya has performed Pipe Stress and ANSYS analyses of various piping systems of the AP1000 plant to investigate the effects of simultaneous high frequency hydrodynamic loads on the structural integrity and functionality of piping components and in-line valves. The analyses included modal superposition and direct integration using forcing function time histories including nonlinear gap elements at the supports. The results of the analysis were utilized to justify a cut-off frequency for extraction of modes and to justify the current testing and qualification program for valves.

High Frequency Cut off Criteria Analysis for Seismic and Suppression Pool Loads US-ABWR Toshiba Corporation Dr. Vaidya has developed finite element models of Feedwater (FW) and Safety/Relief Value Discharge Line (SRVDL) piping systems; and generic electrical cabinet and actuator models in STAAD.Pro to determine the cut-off frequency criteria for high frequency seismic and suppression pool loads. He assessed the application of broadened spectra in the context of modal analysis, and missing mass effects in accordance with Reg. Guide 1.92 and ASCE 4-98. This evaluation utilized the concept of strain energy to justify a cut-off frequency in the dynamic analysis of the systems and equipment.

AP1 000 VCS Duct System Engineering Analysis and HVAC Design SSM Industries I Various Locations Worldwide Dr. Vaidya provided seismic design support for the VCS Duct System for AP1000 Containment. Dr. Vaidya reviewed several analytical models to determine the reaction loads on different containment modules due to the duct runs associated with the VCS System inside the AP1000 Containment. Dr. Vaidya also reviewed the calculation of the composite fundamental frequency of specific duct systems.

KKL Leibstadt NPP Fragility Study PSA Walkdowns Kernkraftwerk I Leibstadt, Switzerland Dr. Vaidya directed the development of fragility evaluation of the plant structures, systems, and components.

Representative ANSYS models for the building structures were developed including structural and foundation Page 91 of 99

Ip Nishikant R. Vaidya, Ph.D., P.E.

%J4 elements, soil structure interaction, and major equipment housed in the buildings. The project evaluated mesh sensitivity of the finite element models by comparing the 1g-push and mode-frequency analysis results from fine and coarse mesh models. The modeling effort involved the use of various structural analysis and preprocessor software including ANSYS, Staad.Pro, SAP2000, Revit Structural, AutoCAD and Alibre Design. The models were converted to SASSI format for soil structure interaction analysis. This effort subsequently supported the development of the seismic, wind and tornado fragilities for the KKL Structures, Systems, and Components.

The SSI analysis extracted seismic forces and moments on structural elements and In-Structure Response Spectra for use in fragility calculations. Floor displacement and accelerations were examined to determine critical structural elements. Fragility analysis for the plant SSCs were performed using the CDFM approach as well as the separation of variables method.

Dr. Vaidya provided insights on the wind hazard evaluations and the impacts on the wind fragilities of plant SSCs. This included the review of the site wind hazard analysis focusing on the occurrence of extreme wind and tornado events, the methodology and models developed to characterize them probabilistically. The results of the wind hazard analysis were subsequently implemented in high wind fragility evaluations.

Dr. Vaidya performed extensive walkdowns in support of the seismic fragility analysis of plant equipment including the FPCCU System, the Containment Isolation System (ClS).

The site walkdowns were also performed to obtain tornado missile inventory and identify vulnerable SSCs and consequences.

He also developed numerous reports related with the structural seismic analysis of KKL NPP structures, conservative deterministic safety margins analysis, and fragility analysis.

PEGASOS j Probabilistic Procedures Swiss Nuclear I Switzerland Dr. Vaidya reviewed the results of the PEGASOS project, a large scale project that applied the most advanced probabilistic procedures to develop the seismic hazard at four nuclear power plant sites in Switzerland. He participated in a Roundtable Workshop to evaluate the results of PEGASOS and subsequently developed a position paper for implementation of the PEGASOS hazard in the Leibstadt plant SPRA.

Genkai NPP Units 3 & 4 I Tornado Analysis URS I Kyushu Electric Power Company I Genkai, Japan Dr. Vaidya provided consultant services for the tornado-borne missile analysis of Genkai Nuclear Power Plant in Japan. The overall project was conducted to support KyOshO Electric Power Company, Inc (Kyasho EPCO) for the purpose of meeting the Japan Nuclear Regulation Authority's (NRA) requirements for restarting the currently shutdown reactors at its Genkai Units 3 and 4 (Genkai NPP) in an efficient and cost-effective manner. Dr.

Vaidya's responsibilities included technical support and recommendations for tornado missile protection strategies of critical and safety-related targets.

Palo Verde NPP j Seismic PRA Peer Review Westinghouse Tonopah, Arizona USA As a key member of the Seismic PRA Peer Review Team, Dr. Vaidya assisted in the review of materials related to the PVNGS Nuclear Site. His focus was to lead the review in the Seismic Fragility Analysis Technical Element (SFR) as well as to participate in the review of the following: SHA and SPR technical elements and the overall Seismic PRA evaluation. Dr. Vaiyda checked conformance to the Standard, ASME/ANS RA-Sa-2009 "ASME PRA Standard Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications" PVNGS Seismic PRA.

The reviewer had access to a copy of the ASME PRA Code since the Code is copyrighted and cannot be copied and distributed. The peer review was documented in an MS Access Database provided by Westinghouse and subsequently in a Peer Review Report. The report represents the consensus of the peer review team and will be signed by each of the reviewers.

Page 92 of 99

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Nishikant R. Vaidya, Ph.D., P.E.

Watts Bar Unit 2 I Seismic PRA URS I TVA I Southern Tennessee USA Dr. Vaidya was responsible for the development of a Scoping Study Plan for the Watts Bar Nuclear Plant. He reviewed available FSAR information on existing building models and assessed if these models are adequate to develop structure response and ISRS in support of fragility analysis and a Regulatory Guide He also reviewed the existing soil structure interaction (SSI) models and analysis including the foundation compliance used in the design analysis. Dr. Vaidya developed a preliminary report summarizing the reviewed documents and methods and assumptions.

New Plant Activities, USA Westinghouse I Various USA locations Dr. Vaidya has been involved in the review and recommendations *for detailing field investigations for site characterization (e.g. seismic, geotechnical, determination of foundation conditions) for several sites to support construction of a Design Certified Standard Plant in a Combined Operating License Application/Early Site Permit approach. Dr. Vaidya is co-author of the Foundation Interface Criteria document.

Expended Core Facility Bettis Atomic Power Laboratory /Idaho National Engineering Laboratory I Idaho USA The Expended Core Facility receives, examines, and prepares Navy Nuclear Fuel for storage and further processing. In support of a reassessment, Dr. Vaidya performed engineering calculations related to the building capacity, including the building steel, crane support structure, and the respective building foundations. His study also examined the response of below-grade concrete pits, which store spent fuel in the wet. Dr. Vaidya directed the seismic reassessment, reviewed the seismic criteria used in the design analysis, the current site geologic and seismological information, as well as, established spectrum to be used as the evaluation basis seismic criteria.

Defense Nuclear Facilities Safety Board (DNFSB) I Seismic Consulting Services u.s. Department of Energy j Washington, D.C. USA In this ongoing project for the DNFSB, Dr. Vaidya participates in consulting activities for the board. He develops draft positions on issues related to DOE's seismic design bases and earthquake engineering of structures systems and components for the DOE nuclear facilities at the Savannah River Site, South Carolina; Rocky Flats, Colorado; Los Alamos National Laboratory, New Mexico; Hanford, Washington; Oak Ridge, Tennessee; and Pantex, Texas. Dr. Vaidya's consulting services include reviews of seismicity and seismic potential, deterministic and probabilistic seismic hazard analyses, vibratory ground motion, site response analysis, soil-structure interaction analysis, and equipment qualification.

AP1000 I Foundation Interface Conditions Report (FICR)

Westinghouse Electric Company, Various Clients Various Worldwide Sites RIZZ0 has supported Westinghouse in development, detailed design, site-specific layout issues, and licensing of these passive de'signs. RIZZO has had major roles in the Plant Parameter Envelop (PPE) development, especially those parameters tied to site specific seismic hazards, geologic hazards, geotechnical, and foundation conditions.

Dr. Vaidya performed soil structure interaction analysis as well as detailed settlement and bearing capacity analyses of plant structures. He has reviewed constructability in support of the Design Certification submittal to USNRC. He is also involved in the review of the analysis, design, and qualification of the Auto-Depressurization Systems' valving, piping and support systems.

Dr. Vaidya evaluated the standard design nuclear island structure and foundation for the construction induced settlements and the resulting forces and moments in the structural elements. This analysis incorporated a unique nonlinear iterative approach, which reflected the progressive loading as well as stiffness buildup of the foundation mat.

Page 93 of 99

FP Nishikant R. Vaidya, Ph.D., P.E.

Cernavoda NPP j Probabilistic Seismic Hazard Analysis (PSHA)

I Cernavoda Nuclear Power Plant EnergoNuclear I Cernavoda, Romania:

Dr. Vaidya provided a lead role to perform a Deterministic and Probabilistic Seismic Hazard Analysis. As part of the Level 1 seismic PSA for Cemavoda, this project developed the site seismic hazard in terms of annual exceedance probabilities for various ground motion measures. Dr. Vaidya established the seismotectonic model for the region to explain recorded earthquakes and to define the potential for future seismic activity. In addition, he defined the seismic sources and the source characteristic in terms of recurrence parameters and the uncertainties in these parameters. Historic seismicity as well as the structural geology was examined to identify the sources. Attenuation relationships were developed and the associated uncertainties for use in the PSHA.

The PSHA calculations were based on standard methods advanced by Cornell and the USGS. Modeling uncertainties were propagated in the PSHA through logic tree formalism.

Crystal River NPP Progress Energy I Crystal River, Florida USA Dr. Vaidya developed seismic input for the nonlinear time history analysis of the spent fuel racks to replace existing fuel racks at the Crystal River NPP. Dr. Vaidya developed an approach to demonstrate that the resulting time histories satisfied the Power Spectra Density requirement in NCR's Standard Review Plan.

Callaway Unit 2 NPP I EPR Design and COL Application UniStar I Callaway, Missouri USA Dr. Vaidya prepared the site related SAR Chapter 2.5 regarding the Seismic hazards plus related Seismic Analysis to demonstrate fit of the standard plant as sited to the generic approved siting envelop. This was done for Ameren Missouri's COLA submittal. He was also involved in verification of the foundation, settlement and bearing capacity of in-situ soils.

US EPR AREVA I Various Worldwide Sites Dr. Vaidya is a member of the Structural Review Board and Foundation Interface Conditions Report (FICR) providing input in support of AREVA's Design Control Document for the EPR. As part of this effort, he has directed several projects related to static and dynamic soil structure interaction analysis for the standard plant design, analysis of short-term and long-term settlements associated with a range of site conditions and constructions sequences. He is currently responsible for developing a site interface document FICR, which defines the site interface conditions required to fully support AREVA's standard design.

AP600 and AP1000 Westinghouse I Various Worldwide Sites RIZZO has supported Westinghouse in development, detailed design, site-specific layout issues, and licensing of these passive designs. RIZZO has had major roles in the Plant Parameter Envelop (PPE) development, especially those parameters tied to site specific seismic hazards, geologic hazards, geotechnical and foundation conditions.

Dr. Vaidya performed soil structure interaction analysis as well as detailed settlement and bearing capacity analyses of plant structures. He has reviewed constructability in support of the Design Certification submittal to USNRC. He is also involved in the review of the analysis, design and qualification of the Auto-Depressurization Systems' valving, piping and support systems.

Dr. Vaidya evaluated the standard design nuclear island structure and foundation for the construction induced settlements and the resulting forces and moments in the structural elements. This analysis incorporated a unique nonlinear iterative approach, which reflected the progressive loading as well as stiffness buildup of the foundation mat.

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IF' Nishikant R. Vaidya, Ph.D., P.E.

Id Dry Storage Facility I Seismic Motion Analysis Idaho National Engineering Laboratory (INEL) I Naval Reactor Complex, Idaho USA This project stored Navy-spent fuel in multi-purpose canisters, which will be housed in concrete overpacks. The Overpacks are freestanding on a concrete slab on grade. Dr. Vaidya performed the soil-structure interaction analysis to calculate the seismic motions at the top of the storage slab, and developed the nonlinear slip and rocking response of the overpacks to the seismic motions. His nonlinear analysis demonstrated a significant margin to overpack kinematic instability. Dr. Vaidya used SASSI and LS-DYNA for this purpose.

DC Cook NPP I Auxiliary Building Equipment Seismic Qualification SSM Industries I Michigan USA Dr. Vaidya performed the seismic qualification to support SSM Industries' seismic qualification upgrade of HVAC equipment. He conducted the seismic qualification in accordance with IEEE-344 Standard following the static evaluation procedures. Dr. Vaidya performed a dynamic analysis to calculate the fundamental frequency to develop the static seismic coefficient.

DC Cook NPP ISeismic Fragility Assessment Westinghouse I Michigan USA In support of Westinghouse's seismic fragility assessment for the Donald C. Cook Nuclear Plant, Dr. Vaidlya participated in the seismic hazard analysis, developed the seismic input, performed soil structure interaction analysis of the Auxiliary Building and developed the floor response spectra. He also supported the review of the equipment fragility's and the seismic margin evaluations. He conducted the seismic qualification in accordance with IEEE-344 Standard following the static evaluation procedures. Dr. Vaidya performed a dynamic analysis to calculate the fundamental frequency to develop the static seismic coefficient.

DAM PROJECTS Saluda Dam Remediation South Carolina Electric and Gas Company I Columbia, South Carolina RIZZO conducted a field geotechnical and subsequent stability analysis for the earth dam. As Project Consultant and Principal Structural Engineer, Dr. Vaidya performed the seismic evaluation for the replacement of the 211-feet high and 7,800-feet long rockfill embankment and Roller Compacted Concrete Berm.

Santee Cooper Project East Dam and East Dam Extension Santee Cooper I Moncks Corner, South Carolina As Project Consultant and Principal Structural Engineer, Dr. Vaidya provided technical expertise on seismic criteria and evaluation for this dam. He developed the design for the South Dam Approach including a three-span concrete bridge supported on drilled caissons. The project consisted of engineering analyses, development of repair and modification schemes, and preparation of plans and specifications for remediation of the bridge structure and its foundation. His project responsibilities included the static, pseudo-static and dynamic evaluation, liquefaction analysis and remediation design. He was involved in developing responses to the Federal Energy Regulatory Commission.

Mirant Energy, Swinging Bridge Dam Sullivan County, New York As Principal Structural Engineer, Dr. Vaidya assisted in the investigation of this "puddle-type" hydraulic fill dam to assess the potential for settlement-induced buckling of the Penstock through the dam. Dr. Vaidya performed the slope stability analysis, settlement and deformation analysis, and development of alternate remediation schemes.

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Nishikant R. Vaidya, Ph.D., P.E.

Jd Eastvale Dam Beaver Falls Municipal Authority I Beaver Falls, Pennsylvania RIZZO converted a timber crib dam to an anchored concrete dam gravity section. As Project Consultant and Principal Structural Engineer, Dr. Vaidya supervised the safety evaluation and developed the remedial design for the dam.

Youghiogheny Hydroelectric Plant DIR Hydro I Confluence, Pennsylvania This project involved the design, finance, and construction management of a complete two-unit, 12 MW Plant with vertical Francis units. Work for the project included major pump-over diversion, lining and grouting an 18-foot diameter rock tunnel, a penstock bifurcation, gate structure, river cofferdam, road construction, and seven miles of transmission line. Dr. Vaidya's project responsibilities included engineering and design of the major structures, support of project financial analysis, licensing with the FERC, and permitting with all state agencies and the Corps of Engineers.

Remmel Dam Entergy f Hot Springs, Arkansas Dr. Vaidya developed the remedial design of this buttress Ambursen dam. As Principal Structural Engineer, his responsibilities included engineering analysis and design, evaluation of sliding and overturning stability, definition of the seismic criteria, and development of alternate remediation schemes.

George B. Stevenson Dam Pennsylvania Department of Environmental Resources I Pennsylvania RIZZO performed a feasibility study directed at installing a hydroelectric plant at the existing dam. As Principal Structural Engineer, Dr. Vaidya modeled the effects of raising the lake level to increase energy output.

Columbia Dam South Carolina Electric & Gas Company I Columbia, South Carolina As the Principal Structural Engineer, Dr. Vaidya developed conceptual design for the remediation and structural upgrades of converting the timber crib dam to a gravity dam. He also directed the engineering analysis in support of the design and developed project documentation for FERC review.

Rio Dam Mirant Energy I Sullivan County, New York RIZZO stabilized an arch gravity dam with anchors and stabilizing a hydraulic fill dam subject to liquefaction. As Principal Structural Engineer, Dr. Vaidya provided direction for the design development and preparation of drawings and specifications.

SPECIALTY STRUCTURES Charleston County Courthouse County of Charleston JCharleston, South Carolina The Charleston County Courthouse is an historic three-story, unreinforced masonry structure, over 200 years old. As part of the renovation of this building, Dr. Vaidya provided engineering services to the exterior stabilization, design of the structural and foundation components. He designed a new foundation system to accommodate the loads that will minimize loss of the existing historic fabric. Dr. Vaidya used his investigation and studies to develop recommendations for the remediation, temporary bracing, and construction sequence.

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IF Nishikant R. Vaidya, Ph.D., P.E.

98 Broad Street County of Charleston i Charleston, South Carolina The scope of this project consisted of demolishing the rear portion of the building while preserving the functionality, as well as, the historic fabric of 98 Broad Street. Serving as Principal-in-Charge, Dr. Vaidya developed the demolition and stabilization package for the structure.

Meyers-Peace House (8 Courthouse Square)

County of Charleston I Charleston, South Carolina Dr. Vaidya assisted in this unique historic restoration. As part of the site preparation for the construction of a new Judicial Center Complex, the historic 8 Courthouse Square was separated from the 1950's construction which adjoins it, stabilized, and moved to a new location about 500 feet from its present location. Dr. Vaidya assisted in the development for the selective demolition and stabilization alternatives for the building, helped design the moving procedure, and a new pile foundation for the new location.

Public Science Building North Charleston, South Carolina As part of the Building Seismic Safety Council's (BSSC) Case Study Project, Dr. Vaidya performed a seismic evaluation of a new building design using FEMA 273 "Guidelines for the Seismic Rehabilitation of Buildings,"

and compared the results to the record design developed on the basis of the 1997 Standard Building Code (SBC). The building evaluated is to be built at a site approximately 30 km from the postulated epicenter of the 1886 earthquake. In addition to use as offices and County operations, the PSB will also serve as the County's Emergency Command Center and is designated for immediate occupancy. Its current design reflects the Group Ill seismic criteria of the 1997 SBC.

Edgewood Country Club Building Renovations Edgewood Country Club Pittsburgh, Pennsylvania Dr. Vaidya assessed the inspection of the facility and reviewed the documentation for the extent of the damage to the building. He also supervised the excavation of a test pit to determine bedrock degradation, and subsequently, developed repair and retrofit concepts.

Fagan's Restaurant Weavertown Environmental Group This project consists of partial demolition of the existing structure and stabilization of an interior wall, which is now exposed to the elements. Dr. Vaidya supervised the demolition to assure minimum disruption to the business.

Sludge Treatment Facility Elirya, Ohio The facility treats sludges prior to disposal in a landfill. Sludges were dredged from an impoundment at the site and pumped into two 30-foot diameter treatment tanks where they are calcinated and bound in fly ash. Dr.

Vaidya provided the engineering for the installation of the facility's material handling system. He developed the engineering drawings and specifications, reviewed and evaluated the bids for the construction of the civil structures, and the installation of the equipment.

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Nishikant R. Vaidya, Ph.D., P.E.

%J ISPAT I Mexicana Plant Trimark, Mexico Dr. Vaidya participated in the development of seismic design criteria for the structure and systems of a direct reduction facility. He supervised the review of the structural design codes; developed a seismic design basis; supervised the input, steps and procedures for the static seismic design of the remaining buildings; and derived the seismic criteria and analysis methodologies for systems attached to the plant.

Rail Transfer Facility Capels Resources, Inc. I West Virginia Dr. Vaidya participated in the investigation of the feasibility of a rail transfer facility as part of a 5,000-acre development in West Virginia. He assisted in the design of a reinforced concrete storage pit for the transfer station.

Recycling Facility Browning Ferris Industries, Inc.

Dr. Vaidya assisted in the design and development of BFI's first materials recovery facility. He directed the structural review of the prefabricated steel building and foundation design.

Fort Thornpkins Adaptive Reuse Study Russo & Sonder Architects I New York The U.S. Navy requested a study of the adaptive reuse of the Fort in support of its Surface Acron Group Homeport facility. Potential use of the facility needed to be compatible with requirements for historical preservation, in accordance with the Advisory Council on Historic Preservation and the New York State Historic Preservation Office. Dr. Vaidya was retained to evaluate the existing structure and the site, in accordance with NAVFAC design criteria.

Fairfield Pumped Storage Project South Carolina Electric & Gas Co. I South Carolina Dr. Vaidya reviewed the inspection of the intake structure and the four gates, and developed recommendations for the repair of the intake structure Control of Vibrations and Noise Combustion Engineering For the scope of this project, Dr. Vaidya directed the design review and evaluation of the structure. He established acceptable levels of vibrations, and supervised the design of a field program to measure the vibrations under different operational conditions of the plant. Dr. Vaidya proposed conceptual solutions in order to reduce the vibration levels. He supervised the engineering evaluation and design to assess the impact of potential modifications.

Building Addition and Upgrade - Greens and Building Shenango, Inc. tPittsburgh, Pennsylvania Dr. Vaidya assisted in providing structural and foundation engineering services for the planned addition and upgrading of Greensand Building at Shenango's Sharpsville facility. He reviewed the structural and foundation analysis, and designed the modifications to the existing structure.

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IF Nishikant R. Vaidya, Ph.D., P.E.

Portable Production Studio Unitel Mobile Video, Inc. I Pennsylvania Dr. Vaidya performed the structural analysis of a portable television production studio. He examined earlier conclusions and assumptions made in a structural analysis, and supervised independent calculations to evaluate structural performance and to estimate the remaining useful life. Dr. Vaidya established concepts to reinforce the trailer structure to improve its performance.

Design Analysis of MLS Antenna Towers C-K Composites Dr. Vaidya directed the structural analysis, design evaluation, modification, and design optimization for several self-supporting trussed towers for Microwave Landing System Antennas for the Federal Aviation Administration.

Structural Upgrade of Boiler House Penntech Papers Dr. Vaidya participated in the design and construction supervision for the upgrade of an 80-year-old paper mill building. Dr. Vaidya developed the concept of a reinforced concrete loading wall.

Gianna and Brenda Fields AGIP, S.P.A. IItaly Dr. Vaidya assisted in the seismic hazard analysis for proposed production platforms located in the Adriatic Sea offshore from Ancona, Italy. He reviewed the geology and tectonic structure of the site region, as well as, the available instrumental and historical earthquake data in order to define Seismotectonics provinces.

Yanbu Crude Oil Terminal Snamprogetti, S.P. I Saudi Arabia Dr. Vaidya assisted in the evaluation of the seismic design criteria for the Yanbu Crude Oil Terminal in Saudi Arabia. He evaluated the seismic hazard and defined the strength level earthquake. He also developed the structural design guidelines consistent with the design basis earthquakes.

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EXPEDITED SEISMIC EVALUATION PROCESS (ESEP)

REPORT COLUMBIA GENERATING STATION 76 NORTH POWER PLANT LOOP RICHLAND, WASHINGTON REPORT NO. R02 15-5462 REVISION A OCTOBER 27, 2015 RIZZO ASSOCIATES 500 PENN CENTER BOULEVARD PENN CENTER EAST BUILDING 5, SUITE 100 PITTSBURGH, PA 15235 TELEPHONE: (412) 856-9700 TELEFAX: (412) 856-9749 WWW.RIZZOASSOC.COM R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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APPROVALS Project No.:

Report Name:

15-5462 Expedited Seismic Evaluation Process (ESEP) Report Columbia Generating Station Date:

December 17, 2015 Revision No.:

0 Approval by the responsible manager signifies that the document is complete, all required reviews are complete, and the document is released for use.

Originators:

Eddie M. Guerra, Senior

--*2

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Project Engineer, RIZZO

""*"'*"Associates Eddie M. Guerra, P.E. (RIZZO)

Lawrence K. Lee (ERIN)

December 17.7 2015 Date December 17. 2015 Date Reviewers:

Habib Shtaih (Energy Northwest)

Steve Sheahan (Energy Northwest)

)J ~4~Va4 Drgzt1Iysdgned by Nshlkant H.

Valoye V.P AIdvLancdEgnelag*

Date,: 201$,12.16 16'38.40 -05080' Nishikant R. Vaidya, Ph.D., P.E. (RIZZ.O)

Vincent Andersen (ERIN)

December 17. 2015 Date December 1 7. 2015 Date December 17. 2015 Date December i17. 2015 Date Shannon Kinnunen (Energy Northwest)

Approved by:

December 17, 2015 Date R02 Columbia Generating Station ESEP 155462/15, Rev. 0 (December 17, 2015)

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,'3ERIz, I

CHANGE MANAGEMENT RECORD Project No.:

Report Name:

Revision No.:

15-5462 Expedited Seismic Evaluation Process (ESEP) Report Columbia Generating Station Richland, Washington A

REVISIONI N.DATE DESCRIPTIONS OF CHANGES/AFFECTED PAGES A

JOctober 27, 2015 JFor Review 4

1 4

4 4

4 R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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TABLE OF CONTENTS PAGE LIST OF TABLES.............................................................................. 7 LIST OF FIGURES.............................................................................

8 LIST OF ACRONYMS........................................................................

9 1.0 PURPOSE AND OBJECTIVE........................................................ 12 2.0

SUMMARY

OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES.......................................................................... 14 3.0 EQUIPMENT SECTION PROCESS AND EXPEDITED SEISMIC EQUIPMENT LIST.................................................................... 16 3.1 EQUIPMENT SELECTION PROCESS AND ESEL................................ 16 3.1.1 ESEL Development................................................. 17 3.1.2 Power-Operated Valves............................................. 18 3.1.3 Pull Boxes............................................................ 18 3.1.4 Termination Cabinets............................................... 19 3.1.5 Critical Instrumentation Indicators.........."........................ 19 3.1.6 Phase 2 and Phase 3 Piping Connections.......................... 19 3.1.7 Relays................................................................ 19 3.2 JUSTIFICATION FOR USE OF EQUIPMENT THAT IS NOT THE PRIMARY MEANS FOR FLEX IMPLEMENTATION........................................... 20 4.0 GROUND MOTION RESPONSE SPECTRUM (GMRS)......................... 21 4.1 PLOT OF GMRS SUBMVITTED BY THE LICENSEE................................ 21 4.2 COMPARISON TO SSE................,.......................................... 22 5.0 REVIEW LEVEL GROUND MOTION (RLGM).................................. 24

5.1 DESCRIPTION

OF RLGM SELECTED.............................................. 24 5.2 METHOD TO ESTIMATE IN-STRUCTURE RESPONSE SPECTRA................. 25 5.3 REVIEW OF EXISTING MODELS................................................... 25 6.0 SEISMIC MARGIN EVALUATION APPROACH................................ 27 6.1]

SUMMARY

OF METHODOLOGIES USED....................................

[...... 27 6.2 HCLPF SCREENING PROCESS.................................................... 28 6.3 SEISMIC WALKDOWN APPROACH................................................ 30 6.3.1 Walkdown Approach................................................ 30 R02 Columbia Generating Station ESEP

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TABLE OF CONTENTS (CONTINUED)

PAGE 6.3.2 Seismic Review Team Qualifications............................. 31 6.3.3 Application of Previous Walkdown Information................. 33 6.3.4 Summary of Walkdown Findings.................................. 33 6.4 HCLPF CALCULATION PROCESS................................................. 34 6.4.1 CDFM Approach.................................................... 35 6.4.2 Component Structural Capacity.................................... 36 6.4.3 Functional Evaluations.............................................. 36 6.5 FUNCTIONAL EVALUATIONS OF RELAYS........................................ 37 6.6 RESOLUTION OF WALKDOWN FINDINGS...................................... 38 6.7 TABULATED ESEL HCLPF VALUES (INCLUDING KEY FAILURE MODES)......................................................................... 40 7.0 INACCESSIBLE COMPONENTS................................................... 41 7.1 IDENTIFICATION OF ESEL ITEMS INACCESSIBLE FOR WALKDOWNS........ 41 7.2 PLANNED WALKDOWN / EVALUATION SCHEDULE / CLOSE OUT............ 43 8.0 ESEP CONCLUSIONS AND RESULTS............................................ 44 8.1 SUPPORTING INFORMATION....................................................... 44 8.2 IDENTIFICATION OF PLANNED MODIFICATIONS.................................. 45 8.3 MODIFICATION IMPLEMENTATION SCHEDULE................................... 45 8.4

SUMMARY

OF REGULATORY COMMITMENTS.................................... 46

9.0 REFERENCES

......................................................................... 47 ATTACHMENTS:

ATTACHMENT A:

EXPEDITED SEISMIC EQUIPMENT LIST (ESEL)

ATTACHMENT B:

SUMMARY

OF ESEL HCLPF VALUES ATTACHMENT C:

WALKDOWN TEAM QUALIFICATIONS R02 Columbia Generating Station ESEP

& flM1W 155462115, Rev. A (October 27, 2015)

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LIST OF TABLES TABLE NO.

TABLE 4-1 TABLE 4-2 TABLE 6-1 TABLE 6-2 TABLE 6-3 TABLE 7-1 TITLE PAGE 5% DAMPED UHRS FOR 10O4 ANi0-O AZR LEVELS AND GMRS AT CONTROL POINT FOR THE CGS SITE............................................................. 22 SSE (5% DAMPING) FOR CGS.................................... 23

SUMMARY

OF SCREENED-OUT COMPONENTS............ 29

SUMMARY

OF CONSERVATIVE DETERMINISTIC FAILURE MARGIN APPROACH..................................35

SUMMARY

.OF ESEP WALKDOWN FINDING RESOLUTIONS...................................................... 38 CGS ESEL ITEMS INACCESSIBLE DURING WALKDOWNS....................................................... 41 R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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LIST OF FIGURES FIGURE NO.

FIGURE 4-1 FIGURE 4-2 FIGURE 5-1 TITLE PAGE 5% DAMPED UHRS FOR 10-4AN 10-5 HAZARD LEVELS AND GMRS AT CONTROL POINT FOR THE CGS SITE............................................................... 21 GMRS VS SSE FOR CGS SITE...................................... 23 RLGM VS 2X SSE FOR CGS ESEP EVALUATION............. 24 R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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'" RIZZO

LIST OF ACRONYMS ACRONYM AC ACI ADS AFW AISC AOV ASCE ASME BE BDBEE CDFM CGS CIP CST DC DG EL ELAP EPRI ESEL ESEP FLEX FSAR ft g

GERS GIP GMRS HCLPF TITLE Alternate Current American Concrete Institute Automatic Depressurization System Auxiliary Feed Water American Institute of Steel Construction Air-Operated Valve American Society of Civil Engineers American Society of Mechanical Engineers Best Estimate Beyond Design-Basis External Event Conservative Deterministic Failure Margin Columbia Generating Station Cast-In-Place Condensate Storage Tank Direct Current Diesel Generator Elevation Extended Loss of all AC Power Electric Power Research Institute Expedited Seismic Equipment List Expedited Seismic Evaluation Process Diverse and Flexible Coping Strategies Final Safety Analysis Report Feet Gravity Generic Equipment Ruggedness Data

Generic Implementation Procedure Ground MotionResponse Spectra High Confidence of Low Probability of Failure R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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LIST OF ACRONYMS (CONTINUED)

ACRONYM HSS HVAC Hz IEEE IPEEE ISRS LPCI MAFE MCC MOV NEI NPP NRC NTTF GIP PGA PRA PSA psi P&IDs RCIC RHR RLGM RPV SEI SEWS SI SMA SME SPRA SQUG TITLE Hollow Steel Section Heating, Ventilation, and Air Conditioning Hertz Institute of Electrical and Electronics Engineers Individual Plant Evaluation of External Events In-Structure Response Spectra Low Pressure Coolant Injection Mean Annual Frequency of Exceedance Motor Control Center Motor-Operated Valve Nuclear Energy Institute Nuclear Power Plant Nuclear Regulatory Commission Near-Term Task Force Overall Integrated Plan Peak Ground Acceleration Probabilistic Risk Assessment Probabilistic Safety Assessment Pound per Square Inch Process and Instrumentation Diagrams Reactor Core Isolation Cooling Residual Heat Removal Review Level Ground Motion Reactor Pressure Vessel Structural Engineering Institute Seismic Screening and Evaluation Work Sheets Seismic Interaction Seismic Margin Assessment Seismic Margin Earthquake Seismic Probabilistic Risk Assessment Seismic Qualification Utilities Group R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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LIST OF ACRONYMS (CONTINUED)

ACRONYM SRT SRV SSC SSE SSHAC SW TRS UHRS U.S.

V Vs TITLE Seismic Review Team Safety/Relief Valve Structures, Systems, or Components Safe Shutdown Earthquake Senior Seismic Hazard Analysis Committee Service Water Test Response Spectrum Unifonrm Hazard Response Spectra United States of America Volts Shear Wave Velocity R02 Columbia Generating Station ESEP 155462/15, Rev. A (October 27, 2015)

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EXPEDITED SEISMIC EVALUATION PROCESS (ESEP)

COLUMBIA GENERATION STATION RICILLAND, WASHINGTON 1.0 PURPOSE AND OBJECTIVE Following the accident at the Fukushima Dai-ichi Nuclear Power Plant (NPP) 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 the 50.54(f) letter on March 12, 2012 [1], requesting information to assure that these recommendations are addressed by all United States (U.S.) NPPs. The 50.54(f) letter 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. Risk 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 the Columbia Generating Station (CGS). 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 NTTF Recommendation 2.1: Seismic [2].

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The objective of this report is to provide summarY infor-mation 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.

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2.0

SUMMARY

OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES The CGS FLEX strategies to maintain Core Cooling and to maintain Containment are summarized below. The Spent Fuel Pool Cooling strategies are not described because they are not included within the scope of the ESEP process. The CGS FLEX strategy summary is derived from the Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order BA-12-049 [3] including the required 6 month updates that have been provided since the OIP was submitted.

The Phase 1 FLEX strategy relies on installed plant equipment. Reactor Core Cooling is achieved via operation of the Reactor Core Isolation Cooling (RCIC) system with injection from the suppression pool water source to the Reactor vessel. Although RCIC suction is normally aligned to Condensate Storage Tank (CST), the OIP assumes that the non-seismically Category 1 CST would be unavailable due to the initiating event (e.g., beyond design-basis seismic event) and the RCIC pump suction is automatically realigned to the suppression pool based on instrumentation sensing low CST level. Reactor vessel cooldown is achieved using RCIC and manual operation of one (1) Safety/Relief Valve (SRV) from the Automatic Depressurization System (ADS) that discharges to the suppression pool. Containment Control during Phase 1 will be maintained by the to-be-installed hardened containment vent. The design of the hardened containment vent system will be described in the Integrated Plan required by NRC Order EA 109. Anticipatory wetwell venting (i.e., "early" containment venting) will be proceduralized using the hardened containment vent to relieve pressure, control suppression pool water temperature, and enable continued cooling of the Reactor via RCIC operation. Key Reactor and Containment parameters required to support Core Cooling and Containment are monitored via DC powered instrumentation. A DC load shed strategy is implemented to extend battery life.

The Phase 2 FLEX strategy relies on installed plant equipment an portable on-site equipment.

The strategy for Phase 2 Core Cooling assumes that the RCIC system is maintained available.

The Reactor Pressure Vessel (RPV) pressure is reduced to 175-3 00 psig to maintain RCIC operation. RCIC will continue to operate to provide RPV injection during Phase 2. Given the proposed FLEX strategy for early containment venting, suppression pool inventory will decrease as steam from the suppression pool vents to the atmosphere. Thermal hydraulic calculations support that suppression pool makeup would be required by approximately 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months after the initial event. Suppression pool makeup can be supplied via one of two portable FLEX pumps R02 Columbia Generating Station ESEP I*

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with suction from the Service Water (SW) spray ponds aligned for discharge to the residual heat removal (RTIR) system piping. In addition, given that steam motive force for RPV makeup from RCIC is expected to significantly decrease after approximately 2-3 days, alternate RPV makeup will be performed via aligning a FLEX pump to the RHR C train. The same FLEX pump can be utilized to support both suppression pooi makeup and RPV makeup because the makeup requirements will be relatively small (e.g., few hundred gpm) at the delayed time when makeup is required.

Necessary electrical components are outlined in the CGS FLEX OIP submittal [3], including subsequent 6 month updates through August 2015, and primarily entail 480 VAC essential motor control centers, vital batteries, equipment installed to support FLEX electrical connections, and monitoring instrumentation required for Core Cooling, Reactor Coolant Inventory, and Containment Integrity. Given success of the DC load shed procedure, the CGS 125 VDC and 250 VDC station batteries are available for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months without recharging. CGS maintains a portable, trailer-mounted FLEX 480V 400 kW diesel generator (i.e., DG4) that is normally staged in the yard next to the Diesel Generator Building. Permanently installed connections on the outside of the Diesel Generator Building to support either the Division 1 or 2 of 480 VAC distribution systems allow an expedited transition to the Phase 2 mitigation strategy. An additional FLEX Diesel Generator (i.e., DG5) is available for alignment to either the Division 1 or 2 of 480 VAC distribution systems as a back-up to DG4 using the same permanently installed connections on the outside of the Diesel Generator Building.

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3.0 EQUIPMENT SECTION PROCESS AND EXPEDITED SEISMIC EQUIPMENT LIST The selection of equipment for the Expedited Seismic Equipment List (ESEL) followed the guidelines of EPRI 3002000704 [2]. The ESEL for CGS is presented in AttachmentA.

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 (BDBEE), as Outlined in the CGS OIP in Response to the March 12, 2012, Commission Order EA-12-049 submitted in February, 2013 [3] including subsequent 6 month updates through August 2015. The OIP provides the CGS 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 CGS OIP [1] including subsequent 6 month updates through August 2015. 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, 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] as follows:

1.

The scope of components is limited to that required to accomplish the core cooling and containment safety functions identified in Table 3-1 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 CGS OIP [3] including subsequent 6 month updates through August 2015.

2.

The scope of components is limited to installed plant equipment and FLEX connections necessary to implement the CGS OIP [1] including subsequent 6 month updates through August 2015 as described in Section 2.0O.

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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 check 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.)

3.1.1 ESEL Development The ESEL was developed by reviewing the CGS OIP [3], including subsequent 6 month updates through August 2015, 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 performed to identify the boundaries of the flow paths 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., circuit breaker, valve, etc.) in branch circuits / branch lines off the defined strategy electrical or fluid flowpath.

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., as necessary.

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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 CGS 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.

Based on a Nuclear Energy Institute (NET) report [4] that compiled pertinent "Questions and Answers" related to the EPRI 3002000704 [2] methodology, manually Operated MOVs and AOVs required for success of the FLEX strategies in Tables 3-1 of EPRI 3002000704 [2] "should be included on the ESEL to ensure they receive the proper interaction reviews during the walkdowns. Some MOV and AOVs have had failures during earthquakes linked to seismic interactions. Several failures occurred where the valve yoke broke due to valve operator impact with walls or beams or other more rugged components during the quake. Extended cast iron operators may be particularly susceptible to these interaction concerns." Therefore, MOVs and AOVs that are locally, manually operated for success of the FLEX strategies are included on the ESEL.

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].

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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 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 CGS OIP [3] including subsequent 6 month updates through March 2015 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. 0 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 and 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 path are included in the ESEL.

3.1.7 Relays As discussed in Section 3.1.2 above, Page 3-2 of EPRI 3002000704 [2] 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 from an electrical component R02 Columbia Generating Station ESEP L

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perspective, the following guidance is applied in the CGS ESEL for functional failure modes associated with electrical components:

Based on an NEI report [4], "power operated valves not required to change state are excluded from the ESEL is because past analyses have consistently showed that they are inherently seismically robust. This is not necessarily the case for electrical components, so they should be included on the ESEL."

Therefore, circuit breakers required to support supply power, whether or not they are required to change state, are included on the ESEL.

Include on the ESEL all relays associated with SRV solenoid valves that could chatter and potentially cause inadvertent SRV actuation to sufficiently depressurize the RPV and preclude adequate RCIC injection.

Relays associated with automatic initiation of the RCIC system are not included on the ESEL since only manual start is credited. The CGS operators are trained to either manually initiate RCIC before the auto-initiation signal occurs on Low-Low RPV water level, or the operators will manually initiate RCIC if the auto-initiation signal occurred, but RCIC did not auto-start. In addition, the operators will be able to override any spurious actuation due to potential relay chatter early in the event (and any such override would be included as part of the manual start action).