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Expedited Seismic Evaluation Process Report
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Issue date:	04/29/2015
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NLS2015043 Enclosure NLS2015043 ENCLOSURE EXPEDITED SEISMIC EVALUATION PROCESS REPORT FOR COOPER NUCLEAR STATION

19.F1 ATAHMENT ENGINEERING REPORT COVER SHEET & INSTRUCTIONS SHEET I OF 2 Engineering Report No.15-007 Rev 0 Page 1 of 6 Engineering Report Cover Sheet Engineering Report

Title:

Stevenson & Associates -

Expedited Seismic Evaluation Process (ESEP)Report in Response to the 50.54() Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.1: Seismic for the Cooper Nuclear Station Acceptance Engineering Report Type:

New 0 Revision El Cancclled El Superseded Superseded by:

EC No.15-014 (Admin)

(4) Report Origin: El CNS [D Vendor Vendor Document No.: 13C4215-RPT-004 Rev. 2, (5) Quality-Related: El Yes 0 No Prepared by: Stevenson & Associates Date: 4/15/2015 Responsible Engineer (Print Name/Sign)

Design Verified: N/A Date:

Design Verifr required) (Print Name/Sign) kOý Date: 41,5/2-z 15 Reviewer (Print Name/Sign)

Approvedby: 1Zrj *j If V ,L4 1of;je

[ie Date: t4-/-i6-1 Supervisor / Manager (Print Name/Sign)

1. Scope and Objective In responding to the Fukushima Near-Term Task Force Recommendation 2.1 Seismic; Cooper Nuclear Station (CNS) contracted Stevenson & Associates as a subject matter expert to develop the Expedited Seismic Evaluation Process (ESEP) in accordance with Electrical Power Research Institute (EPRI) "Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1:

Seismic" [Reference 6]. According to Reference 6; "The ESEP was developed to focus initial resources on the review of a subset of the plant equipment that can be relied upon to protect the reactorcore following beyond design basis seismic events".

This Engineering Report will review and accept the ESEP Report prepared by Stevenson &

Associates.

2. Design Inputs The design inputs are as listed:

1. NPPD Letter NLS2015017 to NRC, "Revision to Nebraska Public Power District's Response to Nuclear Regulatory Commission Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," Cooper Nuclear Station, Docket No. 50-298, DPR-46 dated February 11, 2015.

3. Assumptions No assumptions were made by CNS in the development of this Engineering Report.

4. Detailed Discussion Following the accident at the Fukushima Daiichi nuclear power plant resulting from the March 11, 2011, Great Tohoku Earthquake and subsequent tsunami, the Nuclear Regulatory Commission (NRC) established a Near Term Task Force (NTTF) to conduct a systematic review of NRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena. NTTF Recommendation 2.1 for seismic hazards, as amended by the SRMs associated with SECY-1 1-0124 and SECY-1 1-0137, instructed the NRC staff to issue requests for information to licensees pursuant to Sections 161 .c, 103.b, and 182.a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f). This information Engineering Report 15-007 Rev. 0 PAGE 2 of 6

request was for licensees under 10 CFR 50 to reevaluate the seismic hazards at their sites against present-day NRC requirements and guidance. Based upon this information, the NRC staff will determine whether additional regulatory actions are necessary (e.g., update the design basis and SSCs important to safety) to protect against the updated hazards. In developing Recommendation 2.1, the NTTF recognized that the state of knowledge of seismic hazard within the United States (U.S.) has evolved and the level of conservatism in the determination of the original seismic design bases should be reexamined.

Electric Power Research Institute (EPRI) took the responsibility of developing new Ground Motion Response Spectra (GMRS) for each site in the industry. The new GMRS that was generated for CNS [Reference 13] utilizes up-to-date models representing seismic sources for Central and Eastern United States (CEUS) Plants, ground motion equations, and site amplification.

EPRI, in conjunction with the Nuclear Energy Institute (NEI), developed the Seismic Evaluation Guidance (SPID) [Reference 4] for the Resolution of Fukushima Near-Term Task Force Recommendation 2.11: Seismic and the Template for the Seismic Hazard and Screening Reports for Central and Eastern United States (CEUS) Plants.

In Reference 5, Nebraska Public Power District (NPPD) submitted the Seismic Hazard Evaluation and Screening Report for Cooper Nuclear Station (CNS); which concluded that the expedited seismic evaluation process (ESEP) was required.

The Expedited Seismic Equipment List (ESEL) [Reference 1] is a subset of permanent plant equipment required for successful implementation of the mitigation strategies for Extended Loss of AC Power (ELAP) and Loss of Ultimate Heat Sink (LUHS) due to a beyond design-basis seismic event.

The ESEP addresses the requested information part of the 50.54(f) Letter [Reference 2] that requests "interim evaluations and actions taken or planned to address the higher seismic hazard relative to the design basis, as appropriate, prior to completion of the risk evaluation."

Stevenson & Associates Report 13C4215-RPT-004 Expedited Seismic Evaluation Process (ESEP)Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.1: Seismic for the Cooper Nuclear Station is accepted at CNS and is included as Attachment A to this Engineering Report. All comments have been resolved and no further changes are necessary.

Engineering Report 15-007 Rev. 0 PAGE 3 of 6

5. Summary of Results The results presented by Stevenson & Associates Report 13C4215-RPT-004 Expedited Seismic Evaluation Process (ESEP)Report in Response to the 50.54(o Information Request Regarding Fukushima Near-Term Task Force Recommendation 2. 1: Seismic for the Cooper Nuclear Station can be found in Attachment A. Discussion of the methodology used in the development of the ESEP Report is specifically addressed within EPRI Report 3002000704

[Reference 6] and will not be discussed in this Engineering Report. Review of ESEP Report resulted in comments that were resolved accordingly. No further review is necessary.

6. Conclusions and Recommendations

1) Stevenson & Associates Report 13C4215-RPT-004 Expedited Seismic Evaluation Process (ESEP)Report in Response to the 50.54(o Information Request Regarding Fukushima Near-Term Task Force Recommendation 2. 1: Seismic for the Cooper Nuclear Station [Attachment A] is acceptable for adoption at CNS.

Engineering Report 15-007 Rev. 0 PAGE 4 of 6

7. References

1. CNS Engineering Report 15-006 Revision 0, "Stevenson & Associates - Expedited Seismic Equipment List Walk Down Report Cooper Nuclear Station Acceptance"

2. Nuclear Regulatory Commission (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-lchi Accident,"

dated March 12, 2012

3. Nuclear Regulatory Commission Order Number EA-12-049, Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, dated March 12, 2012

4. EPRI Report 1025287 "Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1:

Seismic" Dated February 2013

5. NPPD Letter NLS2015017 to NRC, "Revision to Nebraska Public Power District's Response to Nuclear Regulatory Commission Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," dated February 11, 2015

6. EPRI Report 3002000704 "Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic" Final Report Dated May 2013

7. NEI 12-06, Rev. 0, "Diverse and Flexible Coping Strategies (FLEX) Implementation Guide", August 2012

8. NRC Letter NLS2013024 from Nebraska Public Power District (ML13070A009),

"Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design Basis External Events (Order Number EA-12-049)", February 28, 2013

9. NRC Letter NLS2012109 from Nebraska Public Power District (ML12310A200),

"Cooper Nuclear Station's First Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design-Basis External Events (Order Number EA-12-049)", August 27, 2013

10. NRC Letter NLS2014019 from Nebraska Public Power District (ML14064A201),

"Nebraska Public Power District's Second Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design-Basis External Events (Order Number EA-12-049)", February 26, 2014 Engineering Report 15-007 Rev. 0 PAGE 5 of 6

11. NRC Letter NLS2014082 from Nebraska Public Power District, "Nebraska Public Power District's Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design-Basis External Events (Order Number EA-12-049)",

August 26, 2014

12. NRC Letter NLS2015019 from Nebraska Public Power District (ML15062A040),

"Nebraska Public Power District's Fourth Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design-Basis External Events (Order Number EA-1 2- 049)", February 23, 2015

13. CNS Engineering Report 14-002 Revision 1, "LCI GMRS Report Acceptance"

8. Attachments A. Stevenson & Associates Expedited Seismic Evaluation Process (ESEP)Report in Response to the 50.54(0 Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.1: Seismic for the Cooper Nuclear Station Report 13C4215-RPT-004 Revision 2; April 15, 2015 Engineering Report 15-007 Rev. 0 PAGE 6 of 6

ENGINEERING REPORT ER 2015-007 Attachment A Stevenson & Associates "Expedited Seismic EvaluationProcess (ESEP)Report in Response to the 50.54(f)

Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.1: Seismic for the CooperNuclear Station" Report 13C4215-RPT-004 Revision 2 April 15, 2015

ER2015-007 Attachment A Page 2 of 42 Document No: 13C4215-RPT-004 Revision 2 Stevenson & Associates EngineeringSolutionsfor Nuclear Energy April 2015 Expedited Seismic Evaluation Process (ESEP) Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.1:

Seismic for the Cooper Nuclear Station Prepared for:

Nebraska Public Power District Cooper Nuclear Station Brownville, Nebraska Stevenson & Associates 1646 North Litchfield Road, Suite 250 Goodyear, AZ 85395 www.vecsa.com Phone: 781-932-9580 Fax: 781-933-4428

ER2015-007 Attachment A Page 3 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 Mi for Cooper Nuclear Station REVISION RECORD Initial Issue (Rev. 0) Date Prepared by: 4/13/2015 Samer E11-Bahey, Ph.D., P.E.

Senior Engineer Reviewed by: 4/13/2015 ostolos Karavoussianis Senior Consultant Approved by: Apostolos Karavoussianis 4/13/2015 Senior Consultant Revisions No. Prepared/Date Reviewed/Date Approved/Date Sam~er E -B 4/14/15 Apostolos Karavoussianis 4/14/15 Apostolos Karavoussianis 4/14/15 Samer EL ahey 4/15/15 Apostolos Karavoussianis 4/15/15 Apostolos Karavoussianis 4/15/15 I

ER2015-007 Attachment A Page 4 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station TABLE OF CONTENTS 1 Purpose and Objective ............................................................................................................................ 4 2 Brief Summary of the FLEX Seismic Implementation Strategies ..................................................... 5 3 Equipment Selection Process and ESEL ........................................................................................... 6 3.1 Equipment Selection Process and ESEL ................................................................................... 6 3.2 Justification for use of Equipment that is not the primary means for FLEX implementation ........ 9 4 Ground Motion Response Spectrum (GMRS) ................................................................................... 10 4.1 Plot of GMRS Submitted by the Licensee ................................................................................ 10 4 .2 C om parison to S S E ...................................................................................................................... 12 5 Review Level Ground Motion (RLGM) ............................................................................................ 13 5.1 Description of RLGM Selected ................................................................................................ 13 5.2 M ethod to E stimate ISR S ............................................................................................................. 14 6 Seismic Margin Evaluation Approach .............................................................................................. 15 6.1 Summary of Methodologies Used ............................................................................................ 15 6.2 H C L PF Screening Process ........................................................................................................... 15 6.3 Seismic Walkdown Approach .................................................................................................. 16 6.4 HCLPF Calculation Process .................................................................................................... 18 6.5 Functional Evaluation of Relays ............................................................................................. 20 6.6 Tabulated ESEL HCLPF Values (including Key failure modes) ............................................. 20 7 Inaccessible Item s ................................................................................................................................. 21 7.1 Identification of ESEL Items Inaccessible For Walkdowns .................................................... 21 7.2 Planned Walkdown / Evaluation Schedule / Close Out .......................................................... 21 8 ESEP Conclusions and Results ............................................................................................................ 22 8.1 Supporting Inform ation ................................................................................................................ 22 8.2 Summary of ESEP Identified and Planned Modifications ...................................................... 23 8.3 Modification Implementation Schedule .................................................................................. 23 8.4 Summary of Regulatory Commitments ................................................................................... 23 9 . R eferen ces ............................................................................................................................................ 24 Appendix A: Cooper Nuclear Station (CNS) ESEL (Ref. 22) 9 pages Appendix B: CNS ESEP HCLPF Values and Failure Modes Tabulation 5 pages 2

ER2015-007 Attachment A Page 5of 42 ISZA Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station LIST OF FIGURES Figure 4.1 - CNS GM RS (5% Damping) ................................................................................................... 11 Figure 4.2 - CNS GMRS vs. SSE (5% Damping) ................................................................................... 12 Figure 5.1 -Comparison of CNS RLGM, GMRS & SSE (5% Damping) ............................................. 14 LIST OF TABLES Table 3.1 - Flow Paths Credited for ESEP ................................................................................................ 7 Table 4. 1 CNS SSE (5% Dam ping) .................................................................................................. 10 Table 4.1 CNS GMRS at Control Point (5% Damping) ................................................................... 10 Table 5.1 - CN S RLG M (5% Damping) ............................................................................................... 13 Table 6.4 - HCLPF Calculation Summary ............................................................................................. 19 3

ER2015-007 Attachment A Page 6 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station I PURPOSE AND OBJECTIVE Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March 11, 2011, Great Tohoku Earthquake and subsequent tsunami, the Nuclear Regulatory Commission (NRC) established a Near Term Task Force (NTTF) to conduct a systematic review of NRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system.

The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena. Subsequently, the NRC issued a 50.54(f) letter on March 12, 2012 (Ref. 1) requesting information to assure that these recommendations are addressed by all U.S. nuclear power plants. 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. 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 Cooper Nuclear Station (CNS). The intent of the ESEP is to perform an interim action in response to the NRC's 50.54(f) letter (Ref. 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 and containment following beyond design basis seismic events.

The ESEP is implemented using the methodologies in the NRC endorsed guidance in EPRI Report 3002000704, Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic (Ref. 2).

The objective of this report is to provide summary information describing the ESEP evaluations and results for Cooper Nuclear Station. 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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ER2015-007 Attachment A Page 7 of 42 SAý-

Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station 2 BRIEF

SUMMARY

OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES*

A simplified description of the CNS Overall Integrated Plan (Ref 3.) and subsequent 6 month updates through February 2015 (Ref. 4a, 4b, 4c, and 4d) to mitigate the postulated extended loss of ac power event is that the licensee will initially remove the core decay heat by using both the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) systems concurrently. The steam-driven HPCI and RCIC pumps will initially supply water to the reactor vessel from the Emergency Condensate Storage Tanks (ECSTs) or the suppression pool (torus), depending on availability. HPCI equipment will be secured after one cycle or 10 minutes to maintain battery life and RCIC will be used as the primary make-up equipment to maintain reactor level. Steam from the reactor will be vented through the safety relief valves to the torus. The SAMG portable diesel driven generator will be made available to power 250 volt dc and 125 volt dc battery chargers. Once RCIC operation is no longer possible, a FLEX portable diesel driven pump will be placed in service providing makeup water to the reactor via the residual heat removal (RHR) low pressure coolant injection (LPCI) lines. In the long-term, additional equipment, such as 4160 volt ac diesel generators and diesel driven pumps, will be delivered from one of two Regional Response Centers established by the nuclear power industry to provide supplemental accident mitigation equipment to power an RHR pump and enter shutdown cooling.

CNS plans to use containment venting via the hardened containment vent system (HCVS) to maintain containment (torus and drywell) pressure and temperature within acceptable values. The exact timing and strategy for venting is still under evaluation. The SAMG or a FLEX portable diesel driven generator will extend battery life to support the HCVS and associated instrumentation beyond the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time required by order EA- 13-109.

The SFP will initially heat up due to the unavailability of the normal cooling system. A portable FLEX pump will be aligned and used to add water to the SFP via installed piping or hoses to maintain level as the pool boils. This will maintain a sufficient amount of water above the top of the fuel assemblies for cooling and shielding purposes. Additional equipment provided by the Regional Response Center will provide backup portable pumps and generators for SFP level instrumentation.

This section is based upon input received from Cooper Nuclear Station in Reference 21.

5

ER2015-007 Attachment A Page 8 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station 3 EQUIPMENT SELECTION PROCESS AND ESEL The selection of equipment for the ESEL followed the guidelines of EPRI Report 3002000704 (Ref. 2).

The ESEL is presented in Attachment A.

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 Cooper Nuclear Station Overall Integrated Plan (OIP) (Ref. 3), and August 2013 (Ref. 4a), February 2014 (Ref. 4b), August 2014 (Ref. 4c), and February 2015 (Ref. 4d) six month updates, in Response to the March 12, 2012, Commission Order EA-12-049 (Ref. 1). The OIP provides the Cooper Nuclear Station 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 Cooper Nuclear Station OIP (Ref. 3) including subsequent 6 month updates through August 2014 (Ref. 4a, 4b and 4c).

FLEX recovery actions are excluded from the ESEP scope per EPRI Report 3002000704 (Ref. 2). The overall list of planned FLEX modifications and the scope for consideration herein is limited to those required to support core cooling, reactor coolant inventory and sub-criticality, and containment integrity functions. Portable and pre-staged FLEX equipment (not permanently installed) are excluded from the ESEL per EPRI Report 3002000704 (Ref. 2).

The ESEL component selection followed the EPRI guidance outlined in Section 3.2 of EPRI Report 3002000704 (Ref. 2).

1. The scope of components is limited to that required to accomplish the core cooling and containment safety functions identified in Table 3-2 of EPRI Report 3002000704. The instrumentation monitoring requirements for core cooling/containment safety functions are limited to those outlined in the EPRI Report 3002000704 guidance, and are a subset of those outlined in the Cooper Nuclear Station OIP (Ref. 3) including subsequent 6 month updates through August 2014 (Ref. 4a, 4b and 4c).

2. The scope of components is limited to installed plant equipment, and FLEX connections necessary to implement the Cooper Nuclear Station OIP (Ref. 3) including subsequent 6 month updates through August 2014 (Ref. 4a, 4b and 4c) as described in Section 2.

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

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

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

6. Structures, systems, and components excluded per the EPRI Report 3002000704 (Ref. 2) guidance are:

" Structures (e.g. containment, reactor building, control building, auxiliary building, etc.)

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

" Manual valves and rupture disks.

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

6

ER2015-007 Attachment A Page 9 of 42 SA Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station

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

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

3.1.1 ESEL Development The ESEL was developed by reviewing the Cooper Nuclear Station OIP (Ref. 3), including 6 month updates through August 2014 (Ref. 4a, 4b and 4c), 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 to be used in the FLEX strategies and to identify specific components in the flow paths needed to support implementation of the FLEX strategies.

Boundaries were established at an electrical or mechanical isolation device (e.g., isolation amplifier, valve, etc.) in branch 6ircuits / branch lines off the defined electrical or fluid flow path. P&IDs were the primary reference documents used to identify mechanical components and instrumentation. The flow paths used for FLEX strategies were selected and specific components were identified using detailed equipment and instrument drawings, piping isometrics, electrical schematics and one-line diagrams, system descriptions, design basis documents, etc., as necessary.

The flow paths credited for the Cooper Nuclear Station are shown in Table 3.1 below.

Table 3.1 - Flow Paths Credited for ESEP Flow Path FLEX Drawing P&IDs Core Heat Removal using the Reactor Core Isolation Cooling (RCIC) system: Coolant from the Emergency Condensate Storage Tanks (ECSTs) to the Reactor Pressure Vessel (RPV) via the RCIC pump. Main Second Six-Month Status Steam providing motive force to the RCIC pump Report Attachment 3 2043 (Ref. 5a) turbine and exhausted to the Suppression Pool. (Ref. 4b) 2040 SH1 (Ref. 5b)

Extended core cooling strategy is to place one loop of RHR into the Shutdown Cooling (SDC) mode, using a flex pump supplying the RHR Heat Exchanger with river water via the RHR piping.

Reactor Pressure Vessel (RPV) Pressure Control using Second Six-Month Status the Automatic Depressurization System (ADS): Main Steam relieved through the ADS Safety/Relief Valves Report Attachment 3 2028 (Ref. 5c) to the Suppression Pool. (Ref. 4b) 2010 SH2 (Ref. Sd)

RPV Make Up: Coolant from the yet to be installed Second Six-Month Status on-site well to the ECSTs via the FLEX pump and Report Attachment 3 2049 SH 2 (Ref. 5e) water treatment skid. (Ref. 4b)

Hardened Containment Vent: Torus vented to N/A 2022 SH 1 (Ref. 5f) atmosphere.

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ER2015-007 Attachment A Page 10 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station 3.1.2 Power Operated Valves Page 3-3 of EPRI Report 3002000704 (Ref. 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 I equipment should be considered (e.g. RCIC/AFW trips)."

To address this concern, the following guidance is applied in the Cooper Nuclear Station ESEL for functional failure modes associated with power operated valves:

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

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

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

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

3.1.4 Termination Cabinets Termination cabinets, including cabinets necessary for FLEX Phase 2 and Phase 3 connections, provide consolidated locations for permanently connecting multiple cables. The termination cabinets and the internal connections provide a completely passive function; and the cabinets are not included in the ESEL.

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 Cooper Nuclear Station OIP (Ref. 3), including 6 month updates through August 2014 (Ref. 4a, 4b and 4c), 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 3.1 above goes on to explain that "Piping, cabling, conduit, HVAC, and their supports" are excluded from the ESEL scope in accordance with EPRI Report 3002000704 (Ref. 2).

8

ER2015-007 Attachment A Page 11 of 42

,SA Expedited Seismic Evaluation Process (ESEP) Report for Cooper Nuclear Station 13C4215-RPT-004 Rev. 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.

3.2 Justification for use of Equipment that is not the primary means for FLEX implementation No alternate equipment is used to support the "Primary Means" for FLEX implementation.

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ER2015-007 Attachment A MSi Page 12 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station

--,o 4 GROUND MOTION RESPONSE SPECTRUM (GMRS) 4.1 Plot of GMRS Submitted by the Licensee In accordance with Section 2.4.2 of the SPID (Ref. 14), the licensing design basis definition of the SSE control point for CNS is used for comparison to the GMRS. Seismic Hazard and Screening Report (Ref.

6) lists the CNS SSE PGA to be 0.2g. Horizontal SSE spectral values are taken from Table 3.1-1 of Reference 6 and shown in Tables 4.1-1 (below).

Table 4.1 CNS SSE (5% Damping)

Freq. 0.5 1 1.8 I2.5SSE for3CNS 5 9 25 33 100I SA (g) 0.13 0.19 0.41 0.5 0.53 0.42 0.34 0.26 0.2 0.2 The GMRS per the Seismic Hazard and Screening Report (Table 2.4-1 of Ref. 6,) is tabulated in Table 4.1-2 and shown in Figure 4.1 below:

Table 4.1 CNS GMRS at Control Point (5% Damping)

GMRS for CNS Freq. (Hz) GMRS (g) Freq. (Hz) GMRS (g) 100 0.241 3.5 0.364 90 0.242 3 0.294 80 0.245 2.5 0.209 70 0.249 2 0.162 60 0.258 1.5 0.116 50 0.282 1.25 0.096 40 0.321 1 0.082 35 0.342 0.9 0.076 30 0.359 0.8 0.069 25 0.386 0.7 0.063 20 0.417 0.6 0.060 15 0.463 0.5 0.055 12.5 0.486 0.4 0.044 10 0.465 0.35 0.039 9 0.449 0.3 0.033 8 0.430 0.25 0.028 7 0.417 0.2 0.022 6 0.422 0.15 0.017 5 0.454 0.125 0.014 4 0.415 0.1 0.011 10

ER2015-007A Attachment Page 13 of 42 13C4215-RPT-004 Rev. 2 Expedited Seismic Evaluation Process (ESEP) Report for Cooper Nuclear Station PSA 0.6 -,

0.5 -

_17 1

___ I

0. 4 ____

0.2 00.1 0.1 1 10 100 Spectral Frequency (Hz)

Figure 4.1 - CNS GMRS (5% Damping) 11

ER2015-007 Attachment A Page 14 of 42 SAli Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station 4.2 Comparison to SSE As identified in the Seismic Hazard and Screening Report (Ref. 6), the GMRS exceeds the SSE in portions of the 1-10 Hz range as shown in Figure 4.2 below:

0.6 ___ __ __-

0.5 - _ _ __

0.4 __ -_____

.2 0.2_____ __ _ _ _

0.1 - ______ _ SSE

-~~~~--

_- __ - - _ _ _ _IzL

a. 0.I 1___ 10 -0 Spectral Frequency (Hz)

Figure 4.2 - CNS GMRS vs. SSE (5% Damping) 12

ER2015-007 Attachment A Page 15 of 42 SA Expedited Seismic Evaluation Process (ESEP) Report for Cooper Nuclear Station 13C4215-RPT-004 Rev. 2 5 REVIEW LEVEL GROUND MOTION (RLGM) 5.1 Description of RLGM Selected The RLGM for CNS was determined in accordance with Section 4 of EPRI 30020000704 (Ref. 2) by linearly scaling the CNS SSE by the maximum GMRS/SSE ratio between the 1 and 10 hertz range.

From review of Figure 4.2-1, the maximum GMRS/SSE ration occurs at 10 Hz where the GMRS spectral acceleration is 0.465g. The SSE shape requires logarithmic interpolation between control points at 9 and 25 Hz. The SSE spectral acceleration at 10 Hz is determined as follows:

Iog(0.26 )- Iog(0 34H og O.)

0 (0 og(l OH z)- IOg(9H z))*

S..o = IO0z OI-k)-Iog19H9)) ) = 0.334g The maximum GMRS/SSE ratio between 1 - 10Hz is then calculated to be 1.39 at 10 Hz (witness 0.465 g

/ 0.334 g = 1.39).

The resulting 5% damped RLGM based on scaling the horizontal SSE by the maximum GMRS/SSE ratio of 1.39 is shown in Table 5.1 and Figure 5.1 below.

Table 5.1 - CNS RLGM (5% Damping)

RrGM for CNS Freg. 0.5 1 1.8 2.5 3 5 9 25 133 100 SA (g) 0.18 0.26 0.57 0.70 0.74 0.58 0.47 0.36 10.28 0.28 13

ER2015-007 Attachment A Page 16 of 42 SA Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station 0.8 0.7 0.6 0.5

.28 0.4

"* 0.3 -

0.2.4 0.1 0.2 0.1 1 Spectral Frequency (Hz) 10 100 Figure 5.1 - Comparison of CNS RLGM, GMRS & SSE (5% Damping) 5.2 Method to Estimate ISRS The method used to derive the ESEP in-structure response spectra (ISRS) was to uniformly scale the existing SSE-based ISRS from Reference 18 by the maximum GMRS/SSE ratio of 1.39. The scaled ISRS was determined for all buildings and elevations where ESEL items are located at CNS.

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ER2015-007 Attachment A Page 17 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station 6 SEISMIC MARGIN EVALUATION APPROACH It is necessary to demonstrate that ESEL items have sufficient seismic capacity to meet or exceed the demand characterized by the RLGM. The seismic capacity is characterized as the peak ground acceleration (PGA) for which there is a high confidence of a low probability of failure (HCLPF). The PGA is associated with a specific spectral shape, in this case the 5%-damped RLGM spectral shape. The HCLPF capacity must be equal to or greater than the RLGM PGA. The criteria for seismic capacity determination are given in Section 5 of EPRI Report 3002000704 (Ref. 2).

There are two basic approaches for developing HCLPF capacities:

1. Deterministic approach using the conservative deterministic failure margin (CDFM) methodology of EPRI NP-6041 (Ref. 7).

2. Probabilistic approach using the fragility analysis methodology of EPRI TR-103959 (Ref. 8).

For CNS, the deterministic approach using the CDFM methodology of EPRI NP-6041 (Ref. 7) was used to determine HCLPFs capacities.

6.1 Summary of Methodologies Used CNS conservatively applied the methodology of EPRI NP-6041 (Ref. 7) to all items on the ESEL. The screening walkdowns used the screening tables from Chapter 2 of EPRI NP-6041 (Ref. 7). The walkdowns were conducted by engineers who as a minimum attended the SQUG Walkdown Screening and Seismic Evaluation Training Course. The walkdowns were documented on Screening Evaluation Work Sheets from EPRI NP-6041 (Ref. 7). Anchorage capacity calculations use the CDFM criteria established within EPRI NP-6041 (Ref. 7) with CNS specific allowables and material strengths used as applicable. The input seismic demand used was the RLGM provided in Table 5.1 and Figure 5.1.

6.2 HCLPF Screening Process From Table 5.1, the spectral peak of the RLGM (amplified PGA) for CNS equals 0.74g and occurs at 3 Hz. The screening tables in EPRI NP-6041 (Ref. 7) are based on ground peak spectral accelerations of 0.8g (1St screening column) and 1.2g (2 d screening column). Accordingly, the ESEL component can be screened against the V screening column (< 0.8g) criteria of NP-6041-SL Table 2-4; however the 2 n, screening column (0.8g - 1.2g) will be used to gain additional functional seismic margin.

One ESEL component was located 40 feet above grade. For components located 40 feet above grade or more, screening based on ground peak spectral acceleration is not applicable and additional consideration is required. In accordance with Appendix B of EPRI 1019200 (Ref. 19), components that are above 40 feet from grade and have corresponding ISRS at the base of component not in exceedance of 1.8g in the component frequency range of interest may be screened using the caveats of the 2nd screening column.

The screening of anchorage for non-valve components was evaluated either by the Seismic Review Team (SRT) judgment or simple analysis. For non-valve components whose anchorage could not readily be screened by SRT judgment or simple analysis, CDFM HCLPF calculations (Ref. 9) were performed. This is documented in Attachment B.

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ER2015-007 Attachment A Page 18 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station 6.3 Seismic Walkdown Approach 6.3.1 Walkdown Approach Walkdowns for CNS were performed in accordance with the criteria provided in Section 5 of EPRI Report 3002000704 (Ref. 2), which refers to EPRI NP-6041 (Ref. 7) for the Seismic Margin Assessment process. Pages 2-26 through 2-30 of EPRI NP-6041 (Ref. 7) describe the seismic walkdown criteria, including the following key criteria:

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

If the SRT has a reasonablebasisfor assuming that the group of components are similarand are similarly anchored,then it is only necessary to inspect one component out of this group. The "similarity-basis" should be developed before the walkdown during the seismic capability preparatorywork (Step 3) by reference to drawings, calculations or specifications. The one component or each type which is selected should be thoroughly inspected which probably does mean de-energizingand opening cabinets or panelsfor this very limited sample. Generally,a spare representativecomponent can be found so as to enable the inspection to be performed while the plant is in operation. At leastfor the one component ofeach type which is selected, anchorageshould be thoroughly inspected.

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

The 100% "walk by" is to look for outliers, lack of similarity, anchoragewhich is differentfrom that shown on drawings or prescribedin criteriafor that component, potentialSI [Seismic Interaction]2 (Ref 2, page 5-4) problems, situations that are at odds with the team members 'past experience, and any other areas of seriousseismic concern. If any such concerns surface, then the limited sample size of one component of each type for thorough inspection will have to be increased. The increase in sample size which should be inspected will depend upon the number of outliers and different anchorages, etc., which are observed. It is up to the SRT to ultimately 2 EPRI Report 3002000704 (Ref 2) page 5-4 limits the ESEP seismic interaction reviews to "nearby block walls" and "piping attached to tanks" which are reviewed "to address the possibility of failures due to differential displacements." Other potential seismic interaction evaluations are "deferred to the full seismic risk evaluations performed in accordance with EPRI 1025287 (Ref. 14)."

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ER2015-007 Attachment A Page 19 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station select the sample size since they are the ones who are responsiblefor the seismic adequacy of all elements which they screenfrom the margin review. Appendix D gives guidancefor sampling selection" The CNS walkdowns included as a minimum a 100% walk-by of all items on the ESEL except as noted in Section 7.0. Any previous walkdown information that was relied upon for SRT judgment is documented in Section 6.3.2. ESEP Walkdown and Screening Report (Ref. 20) documents the walkdown results.

6.3.2 Application of Previous Walkdown Information Previous seismic walkdowns were used to support the ESEP seismic walkdowns and evaluations. Several ESEL items were previously walked down during the CNS Seismic IPEEE (Ref. 16a) and USI A-46 (Ref.

16b) program. Those previous program performed extensive walkdowns including the opening of electrical components, such as, MCCs, Switchgears, Control Cabinet, etc..., hence eliminating the need to open potentially energized equipment.

The previous walkdown observations and photographs were reviewed and steps were taken to confirm that the previous walkdowns remain valid. A walk by was performed to confirm that the equipment material condition and configuration is consistent with the previous walkdown observations and that no new significant interactions related to block walls or piping attached to tanks exist.

In general, detailed inspections were performed for ESEP and included, as a minimum, a walk-by of all the components on the ESEL by the SRT with exception of items inside the Drywell, the Steam Tunnel as listed below, and items added to the ESEL later after the walkdowns. A detailed discussion and resolution for each of the items listed below is provided in Section 7.0.

Safety Relief Valves (SRV), Located inside the Drywell.

o MS-RV-71A o MS-RV-71 B o MS-RV-71C o MS-RV-71D o MS-RV-71E o MS-RV-71F o MS-RV-71G o MS-RV-71H

" SRV Accumulators, Located inside the Drywell o IA-ACC-256A o IA-ACC-256B o IA-ACC-256C o IA-ACC-256D o IA-ACC-256E o IA-ACC-256F o IA-ACC-256G o IA-ACC-256H

Drywell Temperature Elements, Located inside the Drywell o PC-TE-505A

" PC-TE-505B 17

ER2015-007 Attachment A Page 20 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station o PC-TE-505C o PC-TE-505D o PC-TE-505E

RCIC Outboard Steam Supply Isolation Valve, RCIC-MO-16, Located inside the steam tunnel

RCIC Pump Discharge To RX Valve, RCIC-MO-2 1, Located inside the steam tunnel

Critical Motor Control Center MCC RA added to the ESEL after the walkdown was concluded

Analog Process Cabinet LRP-PNL-PL1 added to the ESEL after the walkdown was concluded 6.3.3 Significant Walkdown Findings Consistent with the guidance from NP-6041 (Ref. 7), no significant outliers or anchorage concerns were identified during the CNS seismic walkdowns.

6.4 HCLPF Calculation Process ESEL items were evaluated using the criteria in EPRI NP-6041 (Ref. 7). Those evaluations included the following steps:

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

Performing screening evaluations using the screening tables in EPRI NP-6041 (Ref. 7) as described in Section 6.2 and

" Performing HCLPF calculations considering various failure modes that include both structural failure modes (e.g. anchorage, load path etc.) and functional failure modes.

All HCLPF calculations were performed using the CDFM methodology and are documented in the HCLPF calculations (Ref. 9).

Anchorage configurations for non-valve components were evaluated either by SRT judgment, large margins in existing design basis calculations, or CDFM based HCLPF calculations (Ref. 9a, 9b, and 9c).

The results of these analysis methods are documented in Attachment B. For components beyond 40 feet above grade, Table 2-4 of NP-6041 (Ref. 7) is not directly applicable.

EPRI Report 3002000704 (Ref. 2) Section 5 references to EPRI 1019200 (Ref. 19) with respect to screening criteria beyond 40 feet above grade. This guide update allows multiplying the screening lane spectral acceleration value ranges by a factor of 1.5 in order to account for spectral accelerations at the base of the component 3. This screening level at the base of a component is compared to the ISRS demand corresponding to the RLGM. For example, by factoring the acceleration ranges for screening lane 2 of NP-6041-SL Table 2-4, the capacity at the base of a component is bounded by 1.2g* 1.5 = 1.8g. This is compared with the seismic demand presented by the ISRS (as opposed to the RLGM).

As described in the begin of Section 6, for HCLPF calculations the Conservative, Deterministic Failure Margin (CDFM) analysis criteria established in Section 6 of EPRI NP-6041 (Ref. 7) are used for a detailed analysis of components. The relevant CDFM criteria from EPRI NP-6041 (Ref. 7) are summarized in Table 6.4.

3 Page A-22 of NP-6041 (Ref. 7) also references the use of 1.5 times the bounding spectra for comparison against the floor spectra.

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ER2015-007 Attachment A Page 21 of 42 MA Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station Table 6.4 - HCLPF Calculation Summary 4

Load combination: Normal + Seismic Margin Earthquake (SME)

Ground response spectrum: Conservatively specified (84% non-exceedance probability)

Damping: Conservative estimate of median damping.

Structural model: Best estimate (median) + uncertainty variation in frequency.

Soil-structure interaction Best estimate (median) + parameter variation Material strength: Code specified minimum strength or 95% exceedance of actual strength if test data is available.

Static capacity equations: Code ultimate strength (ACI), maximum strength (AISC),

Service Level D (ASME) or functional limits. If test data is available to demonstrate excessive conservatism of code equations then use 84% exceedance of test data for capacity equations.

Inelastic energy absorption: For non-brittle failure modes and linear analysis, use 80%

of computed seismic stress in capacity evaluation to account for ductility benefits or perform nonlinear analysis and use 95% exceedance ductility levels.

In-structure (floor) spectra Use frequency shifting rather than peak broadening to generation: account for uncertainty and use median damping.

The HCLPF capacity is equal to the PGA at which the strength limit is reached. The HCLPF earthquake load is calculated as follows:

U = Normal + Ec Where:

U = Ultimate strength per Section 6 of EPRI NP-6041 (Ref. 7)

Ec = HCLPF earthquake load

Normal = Normal operating loads (dead and live load expected to be present, etc.)

For this calculation, the HCLPF earthquake load is related to a fixed reference earthquake:

Ec = SFc*Eref Where:

Eref = reference earthquake from the relevant in-structure response spectrum (ISRS)

" SFc = component-specific scale factor that satisfies U = Normal +Ec The HCLPF will be defined as the PGA produced by Ec. Because the CNS RLGM PGA is 0.28g:

HCLPF = 0.28g*SFc

" The SME pertaining to HCLPF calculations for CNS is equivalent to the RLGM.

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ER2015-007 Attachment A Page 22 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station 6.5 Functional Evaluation of Relays The CNS ESEL does not contain any relays or switches associated with the FLEX Phase I response, therefore, no evaluations were performed for relay functionality.

6.6 Tabulated ESEL HCLPF Values (including Key failure modes)

Tabulated ESEL HCLPF values including the key failure modes are included in Attachment B.

For items screened out using NP-6041 (Ref. 7) screening tables, or based on SMA analysis in the checklists and no HCLPFs were calculated, the HCLPF is listed as "> RLGM" and the failure mode is listed as "Screened."

For items where anchorage controls the HCLPF value, the anchorage HCLPF value is listed in the table and the failure mode is noted as "Anchorage".

" For items where an equipment capacity based upon the screening lane values of Table 2-4 of EPRI NP-6041 (Ref. 7) controls the HCLPF value (e.g. anchorage HCLPF capacity exceeds the equipment capacity derived from screening lanes), the screening lane HCLPF value is listed in the table and the failure mode is listed as "Equipment Capacity." Based on NP-6041 Table 2-4 lane 2, this limit is equal to 0.45g for items below 40 feet above grade.

The "Equipment Capacity" limits from above are calculated as follows:

The upper-bound spectral peak to NP-6041 Table 2-4 lane 2 is 1.2g. From Table 5.1, the RLGM spectral peak is 0.74g and the PGA is 0.28g. Thus, for equipment less than 40 feet above grade, the "Equipment Capacity" HCLPF is limited to 1.2g / 0.74g

0.28g PGA = 0.45g PGA. For equipment located greater than 40 feet above grade, if the associated ISRS spectral accelerations in the component frequency range of interest do not exceed 1.5 times the NP-6041 Table 2-4 lane 2 bounding spectrum (e.g. 1 .8g peak spectral acceleration), the "Equipment Capacity" HCLPF is conservatively limited to the RLGM PGA of 0.28g.

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ER2015-007 Attachment A Page 23 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station 7 INACCESSIBLE ITEMS 7.1 Identification of ESEL Items Inaccessible For Walkdowns Twenty four (24) ESEL items were not accessible to the SRT during the ESEP walkdowns at CNS due to plant operation, and two (2) ESEL items that were added late after the walkdowns are evaluated based on photographs provided by CNS. A description of circumstances and disposition for these items is provided below.

SRVs and their accumulators (see Section 6.3.2 for component IDs):

The SRVs were not walked down by the SRT due to Radiation Protection concerns given that the components are located within the Drywell and the station was not in outage during the available SRT walkdown window. The SRVs and their accumulators were walked down as part of the A-46 and IPEEE programs. In addition to the A-46 SEWS observations and photographs, the station provide the SRT with additional photograph and design documents. The SRT reviewed design documents, A-46 SEWS and photographs and determined to be acceptable for evaluation of the SRVs and their accumulators (including the consideration of potential seismic interaction), with no further walkdowns, in accordance with the methodology of NP-604 1.

Drywell Temperature Elements (see Section 6.3.2 for component IDs):

The Drywell temperature elements were not walked down by the SRT due to Radiation Protection concerns given that the components are located within the Drywell and the station was not in outage during the available SRT walkdown window. The temperature elements were walked down as part of the A-46 and IPEEE programs. The SRT reviewed the A-46 SEWS observation and photographs and determined to be acceptable for evaluation of the temperature elements (including the consideration of potential seismic interaction), with no further walkdowns, in accordance with the methodology of NP-6041.

RCIC-MO-16 and RCIC-MO-21:

These valve were not walked down by the SRT due to the components being located in a contaminated and high radiation area, i.e. Steam Tunnel. Station photos and or design documentation were reviewed by the SRT and determined to be acceptable for evaluation of these RCIC MOVs (including the consideration of potential seismic interaction), with no further walkdowns, in accordance with the methodology of NP-604 1.

MCC RA and LRP-PNL-PLI These items were added late after the walkdowns were performed and are screened based on photographs provided by CNS, design documentation, and previous walkdowns and reviewed by the SRT and determined to be acceptable.

7.2 Planned Walkdown / Evaluation Schedule / Close Out Since all items that were inaccessible during the ESEP were resolved by alternative means (i.e.

confirmatory photos, A-46 SEWS and design documentation) to the satisfaction of the SRT, no additional walkdowns are required.

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ER2015-007 Attachment A Page 24 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station 8 ESEP CONCLUSIONS AND RESULTS 8.1 Supporting Information CNS has performed the ESEP as an interim action in response to the NRC's 50.54(f) letter (Ref. 1). It was performed using the methodologies in the NRC endorsed guidance in EPRI Report 3002000704 (Ref.

2).

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

The ESEP is part of the overall CNS response to the NRC's 50.54(f) letter (Ref. 1). On March 12, 2014, NEI submitted to the NRC results of a study (Ref. 11) 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 re-evaluated seismic hazards. As such, the "current seismic design of operating reactors continues to provide a safety margin to withstand potential earthquakes exceeding the seismic design basis."

The NRC's May 9, 2014 NTTF 2.1 Screening and Prioritization letter (Ref. 13) concluded that the "fleetwide seismic risk estimates are consistent with the approach and results used in the GI-199 safety/risk assessment." The letter also stated that "As a result, the staff has confirmed that the conclusions reached in GI- 199 safety/risk assessment remain valid and that the plants can continue to operate while additional evaluations are conducted."

An assessment of the change in seismic risk for CNS was included in the risk evaluation submitted in the March 12, 2014 NEI Letter (Ref 11). However, due to discrepancies in Letter NLS2014027 (Ref. 23) regarding Nebraska Public Power District's seismic hazard and screening report (CEUS Sites) for CNS, dated March 31, 2014, and the NRC's May 9, 2014, letter to all power reactor licensees regarding screening and prioritization results for seismic hazard re-evaluations (Ref. 13), CNS took action to evaluate and resolve differences in GMRS models between the NRC and NEI/CNS. As a result of the evaluation CNS performed an IPEEE Adequacy Review and submitted Letter NLS2015017 (Ref. 6), a revised seismic hazard evaluation and screening report for CNS, dated February 11, 2015. Therefore, the conclusions in the pending NRC Response to NLS2015017 will govern CNS Response and Commitment.

In addition, the March 12, 2014 NEI letter (Ref. 11) provided an attached "Perspectives on the Seismic Capacity of Operating Plants," which (1) assessed a number of qualitative reasons why the design of Structures, Systems, and Components (SSCs) inherently contain margin beyond their design level, (2) discussed industrial seismic experience databases of performance of industry facility components similar to nuclear SSCs, and (3) discussed earthquake experience at operating plants.

CNS was designed using conservative practices, such that the plant has 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 structures, systems and components (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 in-structure response spectra calculations

Broadening criteria for in-structure response spectra 22

ER2015-007 Attachment A Page 25 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station

Response spectra enveloping criteria typically used in SSC analysis and testing applications

Response spectra based frequency domain analysis rather than explicit time history based time domain analysis

Bounding requirements in codes and standards

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

Bounding testing requirements

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 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 Summary of ESEP Identified and Planned Modifications The results of the CNS ESEP performed as an interim action in response to the NRC's 50.54(0 letter (Ref.

1) using the methodologies in the NRC endorsed guidance in EPRI Report 3002000704 (Ref. 2) show that all equipment evaluated are adequate in resisting the seismic loads expected to result from the site RLGM.

Therefore, no plant modifications are required as a result of the CNS ESEP.

8.3 Modification Implementation Schedule No modification implementation schedule is required because no modifications are required.

8.4 Summary of Regulatory Commitments No regulatory commitments are required.

23

ER2015-007 Attachment A Page 26 of 42

_ f Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station

9. 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(0 Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident," March 12, 2012

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

3. NRC Letter NLS2013024 from Nebraska Public Power District (MLI3070A009), "Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design Basis External Events (Order Number EA- 12-049)", February 28, 2013

4. (a) NRC Letter NLS2012109 from Nebraska Public Power District (ML12310A200), "Cooper Nuclear Station's First Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design-Basis External Events (Order Number EA- 12-049)", August 27, 2013 (b) NRC Letter NLS2014019 from Nebraska Public Power District (ML14064A201), "Nebraska Public Power District's Second Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design-Basis External Events (Order Number EA- 12-049)", February 26, 2014 (c) NRC Letter NLS2014082 from Nebraska Public Power District, "Nebraska Public Power District's Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design-Basis External Events (Order Number EA- 12-049)", August 26, 2014 (d) NRC Letter NLS2015019 from Nebraska Public Power District, "Nebraska Public Power District's Fourth Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design-Basis External Events (Order Number EA-12-049)", February 23, 2015

5. (a) Bums & Roe Drawing 2043, Rev. N55, (RET. 454003636) Cooper Nuclear Station Flow Diagram Reactor Core Isolation Coolant And Reactor Feed Systems (b) Bums & Roe Drawing 2040, SH 1, Rev. N82, (RET. 454003633) Cooper Nuclear Station Flow Diagram Residual Heat Removal System (c) Bums & Roe Drawing 2028, Rev. N52, (RET. 454003618) Cooper Nuclear Station Flow Diagram Reactor Building & Drywell Equipment Drain System (d) Bums & Roe Drawing 2010, SH 2, Rev. N95, (RET. 454003594) Cooper Nuclear Station Flow Diagram Instrument Air Reactor Building (e) Bums & Roe Drawing 2049, SH 2, Rev N37, (RET. 454003676) Cooper Nuclear Station Flow Diagram Condensate Supply System 24

ER2015-007 Attachment A Page 27 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station (f) Bums & Roe Drawing 2022, SH 1, Rev. N78, (RET. 454003610) Cooper Nuclear Station Flow Diagram Primary Containment Cooling & Nitrogen Inerting System

6. NPPD Letter NLS2015017 to NRC, "Revision to Nebraska Public Power District's Response to Nuclear Regulatory Commission Request for Information Pursuant to 10 CFR 50.54(0 Regarding the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," dated February 11, 2015

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

8. Methodology for Developing Seismic Fragilities, August 1991, EPRI, Palo Alto, CA. 1994, TR-103959

9. S&A calculations (CNS owner acceptance calculation NEDC 15-048):

a. 14C4215-CAL-001 Rev. 0, "Seismic HCLPF Capacity for Condensate Storage Tanks ECST I A and ECST IB"

b. 14C4215-CAL-002 Rev. 0, "Seismic HCLPF Capacity for Mechanical Equipment for ESEP"

c. 14C4215-CAL-003 Rev. 0, "Seismic HCLPF Capacity for Electrical Equipment for ESEP"

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

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

12. Nuclear Energy Institute (NEI), A. Pietrangelo, Letter to D. Skeen of the USNRC, "Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations", April 9, 2013

13. NRC (E Leeds) Letter to All Power Reactor Licensees et al., "Screening and Prioritization Results Regarding Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(F)

Regarding Seismic Hazard Re-Evaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights From the Fukushima Dai-Ichi Accident," May 9, 2014

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

15. NRC (E Leeds) Letter to NEI (J Pollock), "Electric Power Research Institute Final Draft Report, "Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2. 1: Seismic," as an Acceptable Alternative to the March 12, 2012, Information Request for Seismic Reevaluations," May 7, 2013

16. (a) Report NLS960143, "Individual Plant Examination for External Events (IPEEE) Report - 10 CFR 50.54(0 Cooper Nuclear Station, NRC Docket No. 50-298, License No. DPR-46" (b) Report NLS960076, "Submittal of the Unresolved Safety Issue (USI) A-46 Summary Report Cooper Nuclear Station, NRC Docket No. 50-298, License No. DPR-46" 25

ER2015-007 Attachment A Page 28 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station

17. CNS Updated Final Safety Analysis Report (UFSAR)

18. NED C87-162 Rev. 2, "CNS Frequency vs. Acceleration Response Spectra Curves"

19. EPRI Technical Report (TR) 1019200, "Seismic Fragility Applications Guide Update" December 2009.

20. S&A Report 14C4215-RPT-003 Rev. 2, (CNS Engineering Report Number ER 2015-006)

"Seismic Evaluation of Equipment at Cooper Nuclear Station for the Expedited Seismic Evaluation Process"

21. S&A Letter Received from Client, 13C4215-LRC-00 1, " Emailing: FLEX audit intro.docx",

2/25/2015

22. S&A Report 14C4215-RPT-001 Rev. 3, (CNS Engineering Report Number ER 2015-006)

"Development of Expedited Seismic Equipment List"

23. NPPD Letter NLS2014027 to Nuclear Regulatory Commission, "Nebraska Public Power District's Seismic Hazard and Screening Report (CEUS Sites) - Response to NRC Request for Information Pursuant to 10 CFR 50.54(o Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," dated March 31, 2014

24. NRC Letter NLS2014101 from Nebraska Public Power District, "Nebraska Public Power District's First Six-Month Status Report in Response to June 6, 2013 Commission Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions (Order Number EA-13-109)", December 19, 2014 26

ER2015-007 Attachment A Page 29 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station

~Al APPENDIX A: Cooper Nuclear Station (CNS) ESEL APPENDIX A: Cooper Nuclear Station (CNS) ESEL (Ref. 22)

Equipment ID Description Equipment Normal State Equipment Desired Notes State 83 13-1 RCIC PUMP Off On 84 13-2 RCIC TURBINE Off On 121 ASD-ADS/REC ADS ALTERNATE SHUTDOWN In Service In Service Temperature Indicators: PC-TI-2A, PC-TI-PANEL PANEL 2C, PC-TI-2E, PC-TI-2G 46 ECST I A EMERG COND. STOR. TK. IA N/A N/A 47 ECST I B EMERG COND. STOR. TK. IB N/A N/A 29 EE-BAT-125 IA DIV. 1 125 VDC STATION BATTERY Energized Energized 35 EE-BAT-125 lB DIV. 11 125 VDC STATION BATTERY Energized Energized 30 EE-BAT-250 IA DIV. 1 250 VDC STATION BATTERY Energized Energized 44 EE-CHR-125C 125VDC BATTERY CHARGER IC Energized Energized 45 EE-CHR-250C 250V BATTERY CHARGER IC Energized Energized 42 EE-DSC-250C FEED TO 250VDC STATION SERVICE BATTERY CHARGER IC Energized Energized 41 EE-IVTR-IA DIV I INVERTER In Service In Service 88 EE-PNL-125ASD 125 VDC ASD IPCI DISTRBUTION Energized Energized PNL inside HPCI ASD PANEL

____ PANEL 32 EE-PNL-A 125VDC DISTRIBUTION PANEL A Energized Energized 17 EE-PNL-AA2 125VDC POWER PANEL In Service In Service 99 EE-PNL-AA3 125VDC POWER PANEL Energized Energized Phase 3 37 EE-PNL-B 125VDC DISTRIBUTION PANEL B Energized Energized A-]

ER2015-007 Attachment A Page 30 of 42 Sal Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 tor Cooper Nuclear Station APPENDIX A: Cooper Nuclear Station (CNS) ESEL Equipment Equipment Equipment ID Description Notes Normal State Desired State 39 EE-PNL-BB2 125VDC POWER PANEL In Service In Service 125 EE-PNL-CCPIA CRITICAL CONTROL PANEL CCP1A Energized Energized Phase 2 CRITICAL DISTRIBUTION PANEL 127 EE-PNL-CDPIA CDPIA Energized Energized Phase 2 40 EE-PNL-NBPP 115/230 POWER PANEL In Service In Service 89 EE-STRR-125B 125 VDC STARTER RACK B Energized Energized 33 EE-STRR 125VDC RCIC STARTER RACK Energized Energized 125 RCIC 38 EE-STRR-250A 250VDC DIV I STARTER RACK In Service In Service 31 EE-SWGR-125A 125VDC SWITCHGEAR BUS IA Energized Energized 36 EE-SWGR-125B 125VDC SWITCHGEAR BUS IB Energized Energized 34 EE-SWGR-250A 250VDC SWITCHGEAR BUS IA Energized Energized 98 EE-SWGR-4160F 4160V DIV I BUS Energized Energized Phase 3 101 EE-SWGR-480F 480V SWGR CRITICAL BUS IF Energized Energized Phase 3 STATION SERVICE T'RANSFORMER 100 EE-XFMR-480F IF Energized Energized Phase 3 85 EGM RCIC GOVERNOR Energized Energized 10 HPCI ASD HPCI ALTERNATE SHUTDOWN In Service In Service Level Indicators: NBI-LI-185B, NBI-L-PANEL PANEL 19lB. PC-LI-110, CM-LI-1681B 82 HPCI-PI-I 17A ECST LEVEL In Service In Service 9 IA-ACC-256A SRV A ACCUMULATOR Closed Open/Closed 10 IA-ACC-256B SRV B ACCUMULATOR Closed Open/Closed A-2

ER2015-007 Attachment A Page 31 of 42 SA Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station APPENDIX A: Cooper Nuclear Station (CNS) ESEL Equipment ID Description Equipment Equipment Notes Normal State Desired State II IA-ACC-256C SRV C ACCUMULATOR Closed Open/Closed 12 IA-ACC-256D SRV D ACCUMULATOR Closed Open/Closed 13 IA-ACC-256E SRV E ACCUMULATOR Closed Open/Closed 14 IA-ACC-256F SRV F ACCUMULATOR Closed Open/Closed 15 IA-ACC-256G SRV G ACCUMULATOR Closed Open/Closed 16 IA-ACC-256H SRV H ACCUMULATOR Closed Open/Closed 130 LR-104 RACK-LR-104 In Service In Service Contains Indicator PC-PI-513 129 LR-139 RACK-LR-139 In Service In Service Contains Transmitter PC-PT-4A I CONTROL ROOM VERTICAL Temperature Indicators: PC-TI-505A, PC-119 LRP-PNL-H PANEL H In Service In Service TI-505B, PC-TI-505C, PC-TI-505D, PC-TI-505E Signal conditioning for pressure 132 LRP-PNL-PLI Analog Process Cabinet In Service In Service transmitters PC-PT-4A1, PC-PT-30A, PC-PT-512A 102 MCC-CA CRITICAL MCC CA Energized Energized Phase 3 103 MCC-K CRITICAL MCC K Energized Energized Phase 3 FEED TO 125VDC and 250VDC 43 MCC-LX STATION SERVICE BATTERY N/A N/A CHARGERS IC 104 MCC-Q CRITICAL MCC Q Energized Energized Phase 3 Powered via future manual transfer switch 131 MCC-RA CRITICAL MCC RA In Service In Service from future MOV-UPS for Phase I, and MCC-K for Phase 2 A-3

ER2015-007 Attachment A Page 32 of 42 pSAi Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station APPENDIX A: Cooper Nuclear Station (CNS) ESEL Equipment ID Description Equipment Equipment Notes Normal State Desired State SRV is normally closed but must be I MS-RV-71A SAFETY RELIEF VALVE A Closed Open/Closed electrically opened to control reactor pressure.

SRV is normally closed but must be 2 MS-RV-71 B SAFETY RELIEF VALVE B Closed Open/Closed electrically opened to control reactor pressure.

SRV is normally closed but must be 3 MS-RV-71C SAFETY RELIEF VALVE C Closed Open/Closed electrically opened to control reactor pressure.

SRV is normally closed but must be 4 MS-RV-71D SAFETY RELIEF VALVE D Closed Open/Closed electrically opened to control reactor pressure.

SRV is normally closed but must be 5 MS-RV-71E SAFETY RELIEF VALVE E Closed Open/Closed electrically opened to control reactor pressure.

SRV is normally closed but must be 6 MS-RV-71 F SAFETY RELIEF VALVE F Closed Open/Closed electrically opened to control reactor pressure.

SRV is normally closed but must be 7 MS-RV-71G SAFETY RELIEF VALVE G Closed Open/Closed electrically opened to control reactor pressure.

SRV is normally closed but must be 8 MS-RV-71 H SAFETY RELIEF VALVE H Closed Open/Closed electrically opened to control reactor pressure.

Identified as NBI-PIS-60A in FLEX 54 NBI-PI-60A RPV PRESSURE In Service In Service Strategy (Refs. 3 & 4), located on Rack 25-5 Identified as NBI-PIS-60B in FLEX 55 NBI-PI-60B RPV PRESSURE In Service In Service Strategy (Refs. 3 & 4), located on Rack 25-6 56 NBI-PI-61 RPV PRESSURE In Service In Service Located on Rack 25-5 I F09 Panel 9-15 Relay Panel 9-15 In Service In Service RPS Trip System A Cabinet A-4

ER2015-007 Attachment A Page 33 of 42 Sai-Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station APPENDIX A: Cooper Nuclear Station (CNS) ESEL Equipment ID Description Equipment Equipment Notes Normal State Desired State 110 Panel 9-17 Relay Panel 9-17 In Service In Service RPS Trip System B Cabinet III Panel 9-18 Relay Panel 9-18 In Service In Service Reactor Vessel Level Control Cabinet 86 PANEL 9-3 Control Panel 9-3 In Service In Service 112 Panel 9-32 Relay Panel 9-32 In Service In Service Engineered Safeguard Relay Cabinet I 113 Panel 9-33 Relay Panel 9-33 In Service In Service Engineered Safeguard Relay Cabinet 2 87 PANEL 9-4 Control Panel 9-4 In Service In Service 114 Panel 9-45 Relay Panel 9-45 In Service In Service Auto Blowdown Relay Cabinet Contains Indicators: RFC-LI-94A, RFC-LI-128 PANEL 9-5 Control Panel 9-5 In Service In Service 94B, RFC-LI-94C, RFC-PI-90A, RFC-PI-90B, RFC-PI-90C NLS2014101 Hardened Containment Vents 108 Panel P2 CONTROL ROOM PANEL P2 In Service In Service Capable of Operation Under Severe Accident Conditions (Order Number EA-13-109) (Ref. 24)

NLS2014101 Hardened Containment Vents 106 PC-AO-237 TORUS INLET OUTBOARD Closed Open Capable of Operation Under Severe ISOLATION VALVE Accident Conditions (Order Number EA-13-109) (Ref. 24)

NLS2014101 Hardened Containment Vents 107 PC-AO-32 TORUS INLET OUTBOARD Closed Open Capable of Operation Under Severe ISOLATION VALVE Accident Conditions (Order Number EA-13-109) (Ref. 24) 122 PC-LRPR-IA CONTAINMENT/TORUS RAG LEVEL EODRIn EE RECORDER WIDE Service In Serv ice Transmitters: PC-PT-4AI, P- 2 PC-PT-30A, PC-RANGE PT-512A NLS2014101 Hardened Containment Vents 105 PC-MO-233 TORUS INLET INBOARD Closed Open Capable of Operation Under Severe ISOLATION VALVE Accident Conditions (Order Number EA-1_ 1 13-109) (Ref. 24)

A-5

ER2015-007 Attachment A Page 34 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 wal for Cooper Nuclear Station APPENDIX A: Cooper Nuclear Station (CNS) ESEL Equipment ID Description Equipment Equipment Notes Normal State Desired State 59 PC-PI-20 TORUS PRESSURE In Service In Service 58 PC-PI-2104AG DRYWELL PRESSURE In Service In Service 60 PC-PI-2104BG TORUS PRESSURE In Service In Service 57 PC-PI-513 DRYWELL PRESSURE In Service In Service Located on Rack-LR-104 124 PC-PT-30A WIDE RANGE TORUS PRESSURE In Service In Service Located near MCC-K TRANSMITTER 123 PC-PT-4A I DRYWELL PRESSURE In Service In Service Located on Rack-LR-139 TRANSMITTER 66 PC-TE-IA TORUS TEMPERATURE In Service In Service TE is inside the Torus.

67 PC-TE-I B TORUS TEMPERATURE In Service In Service TE is inside the Torus.

68 PC-TE-IC TORUS TEMPERATURE In Service In Service TE is inside the Torus.

69 PC-TE-ID TORUS TEMPERATURE In Service In Service TE is inside the Torus.

70 PC-TE-I E TORUS TEMPERATURE In Service In Service TE is inside the Torus.

71 PC-TE-IF TORUS TEMPERATURE In Service In Service TE is inside the Torus.

72 PC-TE-1G TORUS TEMPERATURE In Service In Service TE is inside the Torus.

73 PC-TE-I H TORUS TEMPERATURE In Service In Service TE is inside the Torus.

74 PC-TE-2A TORUS TEMPERATURE In Service In Service TE is inside the Torus.

75 PC-TE-2B TORUS TEMPERATURE In Service In Service TE is inside the Torus.

76 PC-TE-2C TORUS TEMPERATURE In Service In Service TE is inside the Torus.

77 PC-TE-2D TORUS TEMPERATURE In Service In Service TE is inside the Torus.

A-6

ER2015-007 Attachment A Page 35 of 42 Srai Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station APPENDIX A: Cooper Nuclear Station (CNS) ESEL Equipment ID Description Equipment Equipment Notes Normal State Desired State 78 PC-TE-2E TORUS TEMPERATURE In Service In Service TE is inside the Torus.

79 PC-TE-2F TORUS TEMPERATURE In Service In Service TE is inside the Torus.

80 PC-TE-2G TORUS TEMPERATURE In Service In Service TE is inside tile Torus.

81 PC-TE-2H TORUS TEMPERATURE In Service In Service TE is inside the Torus.

61 PC-TE-505A DRYWELL TEMPERATURE In Service In Service 62 PC-TE-505B DRYWELL TEMPERATURE In Service In Service 63 PC-TE-505C DRYWELL TEMPERATURE In Service In Service 64 PC-TE-505D DRYWELL TEMPERATURE In Service In Service 65 PC-TE-505E DRYWELL TEMPERATURE In Service In Service Transmitters: NBI-LT-52A, NBI-LT-52C, NBI-PT-53A, NBI-PT-53C, PC-PT-512A &

[is Rack 25-5 INSTRUMENT RACK 25-5 In Service In Service Switches: NBI-LIS-83A, NBI-LIS-101A, NBI-LIS-I01B, NBI-LIS-57A, NBI-LIS-5713, NBI-LIS-72A, NBI-LIS-72C.

117 Rack 25-51 INSTRUMENT RACK 25-51 In Service In Service Transmitters: NBI-LITS-73A Transmitters: NBI-LITS-73 B, NBI-PIS-118 Rack 25-52 INSTRUMENT RACK 25-52 In Service In Service 52D,NBI-LT-9 IB 52D, NBI-LT-91 B Transmitters: NBI-LT-52B, NBI-LT-59B &

Switches: NBI-LIS-83B, NBI-LIS-101C, 116 Rack 25-6 INSTRUMENT RACK 25-6 In Service In Service NBI-LIS-101D, NBI-LIS-58A, NBI-LIS-5813, NBI-LIS-72B, NBI-LIS-72D, NBI-PIS-52B 27 RCIC-MO-I3i STM SUPP TO TURB VALVE Closed Open 28 RCIC-MO-132 TURB OIL COOLING WTR SUPP Closed en 28__ RC1C-MO- 132VALVE

_ Closed Open A-7

ER2015-007 Attachment A Page 36 of 42 SA Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station APPENDIX A: Cooper Nuclear Station (CNS) ESEL L2 Equipment ID Description Equipment Equipment Notes Normal State Desired State 18 RCIC-MO0-14 STEAM TRIP AND THROTTLE Open Open/Closed VALVE 19 RCIC-MO-16 OUTBOARD STM SUPP ISOL Closed Open/Closed VALVE 20 RCIC-MO-18 ECST PUMP SUCT VALVE Open Open/Closed 21 RCIC-MO-20 RCIC DISCHARGE VALVE Open Open/Closed 22 RCIC-MO-21 PUMP DISCH TO RX Closed Open/Closed 23 RCIC-MO-27 MIN FLOW BYP Closed Closed 24 RCIC-MO-30 TEST BYP TO ECST VALVE Closed Closed/Open 25 RCIC-MO-33 ECST TEST LINE SHUTOFF VALVE Closed Closed SUCTION FROM THE SUPPRESSION 26 RCIC-MO-41 CHAMBER Closed Open Transmitter NBI-LT-52A located on Rack RPV LEVEL NARROW RANGE In Service In Service 25-5. L Pnlk located on 48 RFC-LI-94A 25-5. 1-1 located on Panel 9-5 Transmitter NBI-LT-52B located on Rack 49 RFC-LI-94B RPV LEVEL NARROW RANGE In Service In Service 25-6. L located Pnl9 on 25-5.

Transmitter NBI-LT-52C located9-5 1-I located on Panel on Rack 50 RFC-LI-94C RPV LEVEL NARROW RANGE In Service In Service25.LIlctdoPae9-Transmitter 25-5. NBI-PT-53A P! located on located on Rack Panel 9-5 5I RFC-PI-90A RPV PRESSURE In Service In Service 25-5. P1 located Pnlk on Transmitter 25-6. NB1-PT-53B PI located on located Panel 9-5on Rack 52 RFC-PI-90B RPV PRESSURE In Service In Service256P1lctdoPae9-53 RFC-PI-90C RPV PRESSURE In Service In Service Transmitter NBI-PT-53C located on Rack 25-5. PI located on Panel 9-5 96 RHR-IA RHR PUMP I-A In Service In Service Phase 3 97 RHR-HX-IA RHR HEAT EXCHANGER IA In Service In Service Phase 3 90 RHR-MO-12A RHR HX-A OUTLET VALVE In Service In Service Phase 3 A-8

ER2015-007 Attachment A Page 37 of 42 SAi Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station APPENDIX A: Cooper Nuclear Station (CNS) ESEL Equipment ID Description Equipment Equipment Notes Normal State Desired State 91 RHR-MO-13A RHR PUMP A TORUS SUCTION In Service In Service Phase 3 VALVE 92 RHR-MO-15A RHR PUMP A SDC SUCTION VALVE In Service In Service Phase 3 93 RHR-MO-25A RHR INBD INJECTION VLV In Service In Service Phase 3 94 RHR-MO-27A RHR LOOP A OUTBD INJECTION In Service In Service Phase 3 VLV 95 RBR-MO-65A RHR HX-A INLET VALVE In Service In Service Phase 3 A-9

ER2015-007 Attachment A Page 38 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station APPENDIX B: CNS ESEP HCLPF Values and Failure Modes Tabulation APPENDIX B: CNS ESEP HCLPF Values and Failure Modes Tabulation Equipment ID Description Bldg. Elev. HCLPF Failure Basis g, PGA Mode 13-1 RCIC PUMP RB 860 0.31 Anchorage Evaluated per 13C4215-CAL-002 13-2 RCIC TURBINE RB 860 0.31 Anchorage Evaluated per 13C4215-CAL-002 ASD-ADS/REC ADS ALTERNATE SHUTDOWN RB 903 >RLGM Screened SRT Disposition PANEL PANEL ECST IA EMERG COND. STOR. TK. IA CB 877 0.45 Anchorage Evaluated per 13C4215-CAL-00 I ECST I B EMERG COND. STOR. TK. I B CB 877 0.45 Anchorage Evaluated per 13C4215-CAL-001 EE-BAT-125 IA DIV. 1 125 VDC STATION CB 903 0.45 Functional Evaluated per BATTERY 13C4215-CAL-003 EE-BAT-125 I B DIV. 11 125 VDC STATION CB 903 0.45 Anchorage Evaluated per BATTERY 13C4215-CAL-003 EE-BAT-250 IA DIV. 1 250 VDC STATION CB 903 0.45 Functional Evaluated per BATTERY 13C4215-CAL-003 EE-CHR-125C 125VDC BATTERY CHARGER CB 903 0.31 Anchorage Evaluated per IC 13C4215-CAL-003 EE-CHR-250C 250V BATTERY CHARGER I C CB 903 0.31 Anchorage Evaluated per 13C4215-CAL-003 EE-DSC-250C FEED TO 250VDC STATION CB 903 0.45 Functional Evaluated per SERVICE BATTERY CHARGER 13C4215-CAL-003 1C EE-IVTR-I A DIV I INVERTER CB 903 0.45 Functional Evaluated per 13C4215-CAL-003 EE-PNL-125ASD 125 VDC ASD HPCI RB 903 0.45 Functional Evaluated per DISTRBUTION PANEL 13C4215-CAL-003 EE-PNL-A 125VDC DISTRIBUTION CB 903 0.45 Functional Evaluated per PANEL A 13C4215-CAL-003 EE-PNL-AA2 125VDC POWER PANEL CB 903 0.45 Functional Evaluated per 13C4215-CAL-003 EE-PNL-AA3 125VDC POWER PANEL RB 903 0.45 Functional Evaluated per 13C4215-CAL-003 EE-PNL-B 125VDC DISTRIBUTION CB 903 0.45 Functional Evaluated per PANEL B 13C4215-CAL-003 EE-PNL-BB2 125VDC POWER PANEL CB 903 0.45 Functional Evaluated per 13C4215-CAL-003 EE-PNL-CCPIA CRITICAL CONTROL PANEL CB 918 0.45 Functional Evaluated per CCP IA 13C4215-CAL-003 EE-PNL-CDPIA CRITICAL DISTRIBUTION CB 903 0.45 Functional Evaluated per PANEL CDPIA 13C4215-CAL-003 EE-PNL-NBPP 115/230 POWER PANEL CB 918 0.45 Functional Evaluated per 13C4215-CAL-003 EE-STRR-125B 125 VDC STARTER RACK B RB 859 >RLGM Screened SRT disposition EE-STRR- 125VDC RCIC STARTER RACK RB 903 0.38 Anchorage Evaluated per 125RCIC 13C4215-CAL-003 B-1

ER2015-007 Attachment A Page 39 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station ISA APPENDIX B: CNS ESEP HCLPF Values and Failure Modes Tabulation Equipment ID Description Bldg. Elev. HCLPF Failure Basis (g, PGA). Mode EE-STRR-250A 250VDC DIV I STARTER RACK RB 903 0.45 Functional Evaluated per 13C4215-CAL-003 EE-SWGR-125A 125VDC SWITCHGEAR BUS I A CB 903 0.286 Anchorage Evaluated per I I13C4215-CAL-003 EE-SWGR-125B 125VDC SWITCHGEAR BUS 1B CB 903 0.286 Anchorage Evaluated per 13C4215-CAL-003 EE-SWGR-250A 250VDC SWITCHGEAR BUS 1A CB 903 0.286 Anchorage Evaluated per 13C4215-CAL-003 EE-SWGR-4160F 4160V DIV I BUS RB 932 0.45 Functional Evaluated per I I13C4215-CAL-003 EE-SWGR-480F 480V SWGR CRITICAL BUS IF RB 932 0.45 Functional Evaluated per 13C4215-CAL-003 EE-XFMR-480F STATION SERVICE RB 932 0.45 Functional Evaluated per TRANSFORMER IF 13C4215-CAL-003 EGM RCIC GOVERNOR RB 860 >RLGM Screened SRT Disposition HPCI ASD HPCI ALTERNATE RB 903 0.45 Functional Evaluated per PANEL SHUTDOWN PANEL 13C4215-CAL-003 HPCI-PI-I 17A ECST LEVEL CB 877 >RLGM Screened SRT disposition IA-ACC-256A SRV A ACCUMULATOR DW 921 >RLGM Screened SRT disposition IA-ACC-256B SRV B ACCUMULATOR DW 921 >RLGM Screened SRT disposition IA-ACC-256C SRV C ACCUMULATOR DW 921 >RLGM Screened SRT disposition IA-ACC-256D SRV D ACCUMULATOR DW 921 >RLGM Screened SRT disposition IA-ACC-256E SRV E ACCUMULATOR DW 921 >RLGM Screened SRT disposition IA-ACC-256F SRV F ACCUMULATOR DW 921 >RLGM Screened SRT disposition IA-ACC-256G SRV G ACCUMULATOR DW 921 >RLGM Screened SRT disposition IA-ACC-256H SRV H ACCUMULATOR DW 921 >RLGM Screened SRT disposition LR-104 RACK-LR-104 RB 903 >RLGM Screened SRT disposition LR-139 RACK-LR-139 RB 958 >RLGM Screened SRT disposition LRP-PNL-H CONTROL ROOM VERTICAL CB 932 0.45 Functional Evaluated per PANEL H 13C4215-CAL-003 LRP-PNL-PLI ANALOG PROCESS CABINET CB 918 >RLGM Screened SRT disposition MCC-CA CRITICAL MCC CA RB 932 >RLGM Screened SRT disposition MCC-K CRITICAL MCC K RB 903 >RLGM Screened SRT Disposition MCC-LX FEED TO 125VDC and 250VDC CB 903 >RLGM Screened SRT Disposition STATION SERVICE BATTERY CHARGERS IC MCC-Q CRITICAL MCC Q RB 903 >RLGM Screened SRT Disposition MCC-RA CRITICAL MCC RA RB 958 >RLGM Screened SRT Disposition MS-RV-71A SAFETY RELIEF VALVE A DW 921 >RLGM Screened SRT disposition MS-RV-71B SAFETY RELIEF VALVE B DW 921 >RLGM Screened SRT disposition MS-RV-71C SAFETY RELIEF VALVE C DW 921 >RLGM Screened SRT disposition MS-RV-71D SAFETY RELIEF VALVE D DW 921 >RLGM Screened SRT disposition MS-RV-71E SAFETY RELIEF VALVE E DW 921 >RLGM Screened SRT disposition MS-RV-71F SAFETY RELIEF VALVE F DW 921 >RLGM Screened SRT disposition B-2

ER2015-007 Attachment A Page 40 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station APPENDIX B: CNS ESEP HCLPF Values and Failure Modes Tabulation Equipment ID Description Bldg. Elev. HCLPF Failure Basis (g, PG) Mode MS-RV-71G SAFETY RELIEF VALVE G DW 921 >RLGM Screened SRT disposition MS-RV-71 H SAFETY RELIEF VALVE H DW 921 >RLGM Screened SRT disposition NBI-PI-60A RPV PRESSURE RB 931 0.45 Functional R.O.B to Rack 25-05 NBI-PI-60B RPV PRESSURE RB 931 0.45 Functional R.O.B to Rack 25-06 NBI-PI-61 RPV PRESSURE RB 903 0.45 Functional R.O.B to Rack 25-51 Panel 9-15 Relay Panel 9-15 CB 932 0.45 Functional Evaluated per 13C4215-CAL-003 Panel 9-17 Relay Panel 9-17 CB 932 0.45 Functional Evaluated per 13C4215-CAL-003 Panel 9-18 Relay Panel 9-18 CB 932 0.45 Functional Evaluated per 13C4215-CAL-003 PANEL 9-3 Control Panel 9-3 CB 932 0.45 Functional Evaluated per 13C4215-CAL-003 Panel 9-32 Relay Panel 9-32 CB 903 0.38 Anchorage Evaluated per 13C4215-CAL-003 Panel 9-33 Relay Panel 9-33 CB 903 0.38 Anchorage Evaluated per 13C4215-CAL-003 PANEL 9-4 Control Panel 9-4 CB 932 0.45 Functional Evaluated per 13C4215-CAL-003 Panel 9-45 Relay Panel 9-45 CB 903 0.38 Anchorage Evaluated per 13C4215-CAL-003 PANEL 9-5 Control Panel 9-5 CB 932 0.45 Functional Evaluated per 13C4215-CAL-003 Panel P2 CONTROL ROOM PANEL P2 CB 932 0.45 Functional Evaluated per 13C4215-CAL-003 PC-AO-237 TORUS INLET OUTBOARD RB 881 >RLGM Screened SRT Disposition ISOLATION VALVE PC-AO-32 TORUS INLET OUTBOARD RB 881 >RLGM Screened SRT Disposition ISOLATION VALVE PC-LRPR-IA CONTAINMENT/TORUS WIDE CB 932 0.45 Functional R.O.B of Panel 9-3 RANGE LEVEL RECORDER PC-MO-233 TORUS INLET INBOARD RB 881 >RLGM Screened SRT disposition ISOLATION VALVE PC-PI-20 TORUS PRESSURE RB 903 0.381 Anchorage Evaluated per 13C4215-CAL-003 PC-PI-2104AG DRYWELL PRESSURE RB 931 >RLGM Screened SRT disposition PC-PI-2104BG TORUS PRESSURE RB 903 >RLGM Screened SRT disposition PC-PI-513 DRYWELL PRESSURE RB 903 >RLGM Screened R.O.B of rack LR-104 PC-PT-30A WIDE RANGE TORUS RB 903 >RLGM Screened SRT disposition PRESSURE TRANSMITTER PC-PT-4AI DRYWELL PRESSURE RB 958 >RLGM Screened R.O.B of Rack LR-TRANSMITTER 139 PC-TE- IA TORUS TEMPERATURE RB 859 >RLGM Screened SRT disposition PC-TE-I B TORUS TEMPERATURE RB 859 >RLGM Screened SRT disposition PC-TE-IC TORUS TEMPERATURE RB 859 >RLGM Screened SRT disposition B-3

ER2015-007 Attachment A Page 41 of 42 Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station P i9&1 APPENDIX B: CNS ESEP HCLPF Values and Failure Modes Tabulation Equipment ]ED Description Bldg. Elev. HCLPF Failure Basis (g, PGA Mode ________

PC-TE-1 D TORUS TEMPERATURE RB 859 >RLGM Screened SRT disposition PC-TE-I E TORUS TEMPERATURE RB 859 >RLGM Screened SRT disposition PC-TE-IF TORUS TEMPERATURE RB 859 >RLGM Screened SRT disposition PC-TE-IG TORUS TEMPERATURE RB 859 >RLGM Screened SRT disposition PC-TE-I H TORUS TEMPERATURE RB 859 >RLGM Screened SRT disposition PC-TE-2A TORUS TEMPERATURE RB 859 >RLGM Screened SRT disposition PC-TE-2B TORUS TEMPERATURE RB 859 >RLGM Screened SRT disposition PC-TE-2C TORUS TEMPERATURE RB 859 >RLGM Screened SRT disposition PC-TE-2D TORUS TEMPERATURE RB 859 >RLGM Screened SRT disposition PC-TE-2E TORUS TEMPERATURE RB 859 >RLGM Screened SRT disposition PC-TE-2F TORUS TEMPERATURE RB 859 >RLGM Screened SRT disposition PC-TE-2G TORUS TEMPERATURE RB 859 >RLGM Screened SRT disposition PC-TE-2H TORUS TEMPERATURE RB 859 >RLGM Screened SRT disposition PC-TE-505A DRYWELL TEMPERATURE DW 921 >RLGM Screened SRT disposition PC-TE-505B DRYWELL TEMPERATURE DW 921 >RLGM Screened SRT disposition PC-TE-505C DRYWELL TEMPERATURE DW 921 >RLGM Screened SRT disposition PC-TE-505D DRYWELL TEMPERATURE DW 921 >RLGM Screened SRT disposition PC-TE-505E DRYWELL TEMPERATURE DW 921 >RLGM Screened SRT disposition Rack 25-5 INSTRUMENT RACK 25-5 RB 931 0.45 Functional Evaluated per 13C4215-CAL-003 Rack 25-51 INSTRUMENT RACK 25-51 RB 903 0.45 Functional Evaluated per 13C4215-CAL-003 Rack 25-52 INSTRUMENT RACK 25-52 RB 903 0.45 Functional Evaluated per 13C4215-CAL-003 Rack 25-6 INSTRUMENT RACK 25-6 RB 931 0.45 Functional Evaluated per 13C4215-CAL-003 RCIC-MO-131 STM SUPP TO TURB VALVE RB 860 >RLGM Screened SRT Disposition RCIC-MO-132 TURB OIL COOLING WTR RB 860 >RLGM Screened SRT Disposition SUPP VALVE RCIC-MO-14 STEAM TRIP AND THROTTLE RB 860 >RLGM Screened SRT Disposition VALVE RCIC-MO-16 OUTBOARD STM SUPP ISOL RB 903 >RLGM Screened SRT Disposition VALVE RCIC-MO-18 ECST PUMP SUCT VALVE RB 860 >RLGM Screened SRT disposition RCIC-MO-20 RCIC DISCHARGE VALVE RB 881 >RLGM Screened SRT disposition RCIC-MO-21 PUMP DISCH TO RX RB 903 >RLGM Screened SRT disposition RCIC-MO-27 MIN FLOW BYP RB 860 >RLGM Screened SRT Disposition RCIC-MO-30 TEST BYP TO ECST VALVE RB 881 >RLGM Screened SRT Disposition RCIC-MO-33 ECST TEST LINE SHUTOFF RB 881 >RLGM Screened SRT disposition VALVE RCIC-MO-41 SUCTION FROM THE RB 860 >RLGM Screened SRT disposition SUPPRESSION CHAMBER RFC-LI-94A RPV LEVEL NARROW RANGE CB 932 0.45 Functional R.O.B of Panel 9-5 RFC-LI-94B RPV LEVEL NARROW RANGE CB 932 0.45 Functional R.O.B of Panel 9-5 B-4

ER2015-007 Attachment A Page 42 of 42 SA&

Expedited Seismic Evaluation Process (ESEP) Report 13C4215-RPT-004 Rev. 2 for Cooper Nuclear Station APPENDIX B: CNS ESEP HCLPF Values and Failure Modes Tabulation Equipment ID Description Bldg. Elev. HCLPF Failure Basis (g,PGA) Mode RFC-LI-94C RPV LEVEL NARROW RANGE CB 932 0.45 Functional R.O.B of Panel 9-5 RFC-PI-90A RPV PRESSURE CB 932 0.45 Functional R.O.B of Panel 9-5 RFC-PI-90B RPV PRESSURE CB 932 0.45 Functional R.O.B of Panel 9-5 RFC-PI-90C RPV PRESSURE CB 932 0.45 Functional R.O.B of Panel 9-5 RHR-IA RHR PUMP 1-A RB 860 0.45 Functional Evaluated per 13C4215-CAL-002 RHR-HX-1A RHR HEAT EXCHANGER 1A RB 931 0.43 Anchorage Evaluated per 13C4215-CAL-002 RHR-MO-12A RHR HX-A OUTLET VALVE RB 903 >RLGM Screened SRT disposition RHR-MO-13A RHR PUMP A TORUS SUCTION RB 859 >RLGM Screened SRT disposition VALVE RHR-MO-15A RHR PUMP A SDC SUCTION RB 881 >RLGM Screened SRT disposition VALVE RHR-MO-25A RHR INBD INJECTION VLV RB 903 >RLGM Screened SRT disposition RHR-MO-27A RHR LOOP A OUTBD RB 903 >RLGM Screened SRT disposition INJECTION VLV RHR-MO-65A RHR HX-A INLET VALVE RB 931 >RLGM Screened SRT disposition B-5

v t e Letter Sequence Request
TAC:MF5235, (Approved, Closed)
Initiation Request, Request Acceptance... Administration Questions Answers 29 April 2015 Results Approval Other: ML15231A399, NLS2015019, Fourth Six-Month Status Report in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events
MONTHYEAR2015-07-01 [Table View]

NLS2015043, Expedited Seismic Evaluation Process Report

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Title:

SUMMARY

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