PNP 2014-108, Expedited Seismic Evaluation Process Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima.

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

ML14357A165
Person / Time
Site:	Palisades
Issue date:	12/18/2014
From:	Vitale A Entergy Nuclear Operations
To:	Document Control Desk, Division of Operating Reactor Licensing
References
PNP 2014-108 51-9231248-002
Download: ML14357A165 (70)

Text

Entergy Nuclear Operations, Inc.

Eiy VR0WEntergy Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043 Tel 269 764 2000 Anthony J. Vitale Site Vice President PNP 2014-108 December 18, 2014 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, MD 20852

SUBJECT:

Palisades Nuclear Plant Expedited Seismic Evaluation Process Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident Palisades Nuclear Plant Docket 50-255 License No. DPR-20

REFERENCES:

1. NRC letter, Request for Information Pursuantto Title 10 of the Code of FederalRegulations 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-ichiAccident, dated March 12, 2012 (Adams Accession No. ML12056A046).

2. NEI letter, ProposedPath Forwardfor NTTF Recommendation 2. 1:

Seismic Reevaluations, dated April 9, 2013 (ADAMS Accession No. ML13101A379).

3. NRC letter, Electric Power Research Institute Report XXXXXX, "SeismicEvaluation 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,dated May 7, 2013 (ADAMS Accession No. ML13106A331).

4. Electric Power Research Institute Report 3002000704, Seismic Evaluation Guidance:Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2. 1: Seismic, dated May 2013.

PNP 2014-108 Page 2

Dear Sir or Madam:

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

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

Reference 1 requested that licensees provide interim evaluations and actions taken or planned to address the higher seismic hazard relative to the design basis, as appropriate, prior to completion of the risk evaluation. In accordance with the NRC endorsed guidance in Reference 3, the enclosed Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant provides the information described in Section 7 of Reference 4 in accordance with the schedule identified in Reference 2.

This letter contains five new commitments, identified in the attachment.

I declare under penalty of perjury that the foregoing is true and correct. Executed on December 18, 2014.

Sincerely,

Attachment:

List of Regulatory Commitments

Enclosure:

Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant cc: Director of Office of Nuclear Regulation Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC

Attachment to PNP 2014-108 List of Regulatory Commitments

List of Regulatory Commitments The following table identifies those actions committed to by Entergy Nuclear Operations, Inc.

(ENO) in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

TYPE (Check One) SCHEDULED COMPLETION COMMITMENT ONE- DATE TIME CONTINUING (If Required)

ACTION COMPLIANCE

1. ENO will perform seismic walkdowns, X No later than the generate High Confidence of a Low end of the second Probability of Failure calculations, and planned refueling design and implement any necessary outage after modifications for inaccessible items December 31,2014.

listed in Section 7.1 of the Expedited Seismic Evaluation Process (ESEP)

Report for Palisades Nuclear Plant.

2. Modify Boric Acid Storage Tank X No later than the (T-53A) tank anchorage such end of the second that the High Confidence of a Low planned refueling Probability of Failure (HCLPF) capacity outage after is greater than the demand December 31,2014.

characterized by the Review Level Ground Motion (RLGM) Peak Ground Acceleration.

3. Modify Boric Acid Storage Tank X No later than the (T-53B) tank anchorage such end of the second that the HCLPF capacity is greater than planned refueling the demand characterized by the outage after RLGM Peak Ground Acceleration. December 31, 2014.

4. Modify Primary Makeup Storage Tank X No later than the (T-81) such that the HCLPF capacity is end of the second greater than the demand characterized planned refueling by the RLGM Peak Ground outage after Acceleration. December 31, 2014.

5. Submit a letter to NRC X Within 60 days summarizing the HCLPF calculation following results of commitments 1, 2, 3, and 4, completion of and confirming implementation of the Expedited Seismic plant modifications associated with Evaluation Process commitments 1,2, 3, and 4. activities, including commitments 1, 2, 3, and 4.

Enclosure to PNP 2014-108 Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant 54 Pages Follow

20004-021 (01/30/2014)

A AREVA AREVA Inc.

Engineering Information Record Document No.: 51 - 9231248 - 002 Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant Page 1 of 65

A 20004-021 (0113012014)

Document No.: -0923`1248-002 AREVA Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant Safety Related? YES I NO Does this document establish design or technical requirements? " YES NO Does this document contain assumptions requiring verification? [] YES NO M

Does this document contain Customer Required Format? YES. []NO Signature Block PageuiSectlone Name and PILlP, RILR, Prepared/Revlewedi Thtle/Dlselpllne Signature A-CRF, A Date Approved or Comments*

Ogden Sawyer ,j LPAl Engineering Supervisor I/../I /9 AlaaHorn* P Appendix A (Section 3.0 and P)A Engineer: /20/61//f Attachment A)

ProjectEnginecrll ( 1 2r Grant Tinsley R 12-10-14 Appendix A (Sections 3.0, and

.AB c Attaoliment A) magic Stewart R Appendix A (Sections 6.0,7.0, 8.0 and Engineer IV AttaclunentEB)

Keowl Connell , A Al Engineering Manager A JenniferButler *A-CRF / Appendix A Note: /LP designates Preparer (P), Lead Preparer (LP)

R/LR designates Reviewer (R), Lead Reviewer (LR)

A-CRF delignates Project Manager Approver of Customer Required Format (A-CRP)

A designates Approver/RTM- Verification of Reviewer Independence Page 2

A 20004-021 (01/30/2014)

Document No.: 51-9231248-002 AREVA Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant Signature Block (continued)

Project Manager Approval of Customer References (N/A if not applicable)

Name Title (printed or typed) (printed or typed) Signature Date Jennifer Butler Project Manager Page 3

A 20004-021 (01/30/2014)

Document No.: 51-9231248-002 AREVA Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant Record of Revision Revision Pages/Sections/

No. Paragraphs Changed Brief Description I Change Authorization 000 All Initial release 001 Section 2.0 Section 2.0 0 Updated References 1, 3, 4, 40, 47, 49 and 50 Appendix A - 2.0, 3.0, 3.1, 0 Added References 5, 37, 38, 51, 52 and 57 3.1.1, 3.1.3, 3.1.4, 3.1.6, 3.2, Appendix A 4.2,5.1,5. 2,6.1,6.2,6.3.1, Sections 2.0, 3.0, 3.1, 3.1.1, 3.1.3, 3.1.4, 3.1.6, 3.2, 4.2, 5.1, 5.2, 6.3.3, 6.4, 6.5, 6.6, 7.1, 7.2, 6.1, 6.2, 6.3.1, 6.3.3, 6.4, 6.5, 6.6, 7.1, 7.2, 8.1, 8.2, 8.3, 8.4 and 9.0 8.1, 8.2, 8.3, 8.4 and 9.0, were modified to incorporate Entergy comments [57] on Revision Attachment000 of the document.

Attachment B.

• Attachment A - modified to incorporate Entergy comments [57]

and to update with additional components based on new revision of supporting document.

• Attachment B - modified to incorporate Entergy comments, update HCLPF capacities based on new revisions of supporting calculations, and to uOpdate with additional componentg, based on new revision of supporting document.

002 Section 2.0 Section 2.0 0 Added Reference 58 Appendix A - 3.1.1, 6.5, 7.2, Appendix A and 8.4, Attachment A and 0 Sections 3.1.1 6.5, 7.2, and 8.4 were modified to incorporate Attachment B Entergy comments [58] on Revision 001 of the document

• Attachment A - modified to incorporate Entergy Comments [58].

" Attachment B - modified to incorporate Entergy Comments [58].

Page 4

A Document No.: 51-9231248-002 AREVA Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant Table of Contents Page SIGNATURE BLOCK ............................................................................................................................. 2 RECORD OF REVISION ....................................................................................................................... 4 1.0 DOCUMENTATION .................................................................................................................... 6

2.0 REFERENCES

........................................................................................................................... 6 APPENDIX A: EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT FOR PALISADES NUCLEAR PLANT ........................................................................ A-1 Page 5

A AR EVA Document No.: 51-9231248-002 Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant 1.0 DOCUMENTATION Appendix A to this document contains the Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant, and is presented in the customer-required format.

2.0 REFERENCES

References identified with an (*) are maintained within Palisades Nuclear Plant Records System and are not retrievable from AREVA Records Management. These are acceptable references per AREVA Administrative Procedure 0402-01, Attachment 8. See page 2 for Project Manager Approval of customer references.

1. NRC (E Leeds and M Johnson) Letter to All Power Reactor Licensees et al., "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident," March 12, 2012, NRC ADAMS Accession No. ML12053A340.

2. EPRI 30020007.04, "Seismic Evaluation Guidance, Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," May 2013.

3. Entergy Letter to U.S. NRC, letter number PNP 2013-010, "Overall Integrated Plan in Response to March 12, 2012 Commission Order to ModifyLicenses With Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049),"

February 28, 2013, NRC ADAMS Accession No. ML13246A399.

4. Entergy Letter to U.S. NRC, Letter Number PNP 2014-011, "Palisades Nuclear Plant 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 28, 2014, NRC ADAMS Accession No. ML14059A078.

5. Entergy Letter to U.S. NRC, Letter Number PNP 2014-085, "Palisades Nuclear Plant 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 28, 2014, NRC ADAMS Accession No. ML14240A278.

6. *Entergy Document, Engineering Change EC-46465, "FLEX Basis."

7. *Entergy Drawing E0001, Sheet 1, Revision 83, "Single Line Meter & Relay Diagram 480V Motor Control Center Warehouse, WD 950."

8. *Entergy Drawing E0001, Sheet 3, Revision 4, "Plant Single Line Diagram, WD 950."

9. *Entergy Drawing E0008, Sheet 1, Revision 57, "Single Line Meter & Relay Diagram 125V DC 120V Instrument & Preferred AC, System WD 950."

10. *Entergy Drawing E0008, Sheet 2, Revision 55, Single Line Meter & Relay Diagram 125V DC 120V Instrument & Preferred AC, System WD 950."

11. *Entergy Drawing E0078, Sheet 2, Revision 18, "WRSGL Schematic Diagram."

12. *Entergy Drawing E0078, Sheet 2A, Revision 6, "WRSGL Schematic Diagram."

13. *Entergy Drawing E0082, Sheet 5, Revision 13, "Schematic Diagram Wide Range Pressurizer Level Indicator/Alarm Instrumentation."

14. *Entergy Drawing E0090, Sheet 5, Revision 14, "Schematic Diagram Boric Acid System Instrumentation."

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15. *Entergy Drawing E0092, Revision 14, "Schematic Diagram Safety Injection Tank Level Indicator Alarm Instrumentation."

16. *Entergy Drawing M0201, Sheet 1, Revision 87, "Piping & Instrument Diagram, Primary Coolant System."

17. *Entergy Drawing M0201, Sheet 2, Revision 66, "Piping & Instrument Diagram, Primary Coolant System."

18. *Entergy Drawing M0202, Sheet 1, Revision 76, "Piping & Instrument Diagram, Chemical &

Volume Control System."

19. *Entergy Drawing M0202, Sheet 1A, Revision 64, "Piping & Instrument Diagram, Chemical &

Volume Control System."

20. *Entergy Drawing M0202, Sheet 1B, Revision 59, "Piping & Instrument Diagram, Chemical &

Volume Control System."

21. *Entergy Drawing M0203, Sheet 1, Revision 48, "Piping & Instrument Diagram, Safety Injection, Containment Spray, Shutdown Cooling-System."

22. *Entergy Drawing M0203, Sheet 2, Revision 27, "Piping & Instrument Diagram, Safety Injection, Containment Spray, Shutdown Cooling System."

23. *Entergy Drawing M0205, Sheet 2, Revision69, "Piping & Instrument Diagram, Main Steam and Auxiliary Turbine Systems."

24. *Entergy Drawing M0208, Sheet 1 B, Revision 37, "Piping & Instrument Diagram, Service Water System."

25. *Entergy Drawing M0220, Sheet 1, Revision 97, "Piping & Instrument Diagram, Make-up Domestic Water & Chemical Injection Systems."

26. *Entergy Drawing M0207, Sheet 1, Revision 91, "Piping & Instrument Diagram, Feedwater &

Condensate System."

27. *Entergy Drawing M0207, Sheet 2, Revision 38, "Piping & Instrument Diagram, Auxiliary Feedwater System."

28. *Entergy Procedure Palisades EOP-3.0, Revision 16, "Station Blackout."

29. *Entergy Drawing E0364, Revision 9, "Conduit and Tray Miscellaneous Plans."

30. *Entergy Drawing E0618, Sheet 569, Revision 12, "Connection Diagram Junction Box J569."

31. *Entergy Drawing M0222, Sheet 2, Revision 29 (EC-47346), "Piping & Instrument Diagram, Miscellaneous Gas Supply Systems."

32. *Entergy Drawing M0218, Sheet 2, Revision 61, "Piping & Instrument Diagram, Htg. Vent. & Air Cond. Containment Building."

33. *Entergy Drawing E0005, Sheet 5B, Revision 12, "Single Line Meter & Relay Diagram 480 Volt Motor Control Centers, System WD 950."

34. *Entergy Drawing E0099 Sh. 5, Revision 10, "Schematic Diagram, Containment Building Instrumentation (Left Channel)."

35. *Entergy Drawing E0084 Sh. 6, Revision 14, "Schematic Diagram, Pressurizer Pressure Control and Measurement Channel Instrumentation."

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A AR EVA Document No.: 51-9231248-002 Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant

36. *Entergy Drawing E0087, Sh. 6, Revision 10, "Schematic Diagram Level Indication and Alarm Indication."

37. *Entergy Drawing E0076, Sheet 4C, Revision 8, "Schematic Diagram Feedwater and Turbine Driver Instrumentation."

38. *Entergy Drawing E0226, Sheet 1B, Revision 6, "Schematic Diagram - Reactor Vessel Level Monitoring System (Left Channel)."

39. *"Palisades Nuclear Plant - Final Safety Analysis Report," Revision 31, Docket 50-255, September 2014.

40. Entergy Letter to U.S. NRC, letter number PNP 2014-033, "Palisades Nuclear Plant Seismic Hazard and Screening Report (CEUS Sites), Response to NRC 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 March 31, 2014, NRC ADAMS Accession No. ML14090A069.

41. *Entergy Technical Specification C-1 75(Q), Revision 6, "Requirements for Seismic Evaluation of Electrical and Mechanical Equipment."

42. EPRI-NP-6041-SL, "Methodology for Assessment of Nuclear Power Plant Seismic Margin,"

Revision 1, August 1991.  :

43. EPRI TR-103959, "Methodology for Developing Seismic Fragilities," July 1994.

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

45. *Entergy Document, "Palisades Nuclear Plant Individual Plant Examination of External Events (IPEEE)," Revision 1, May 1996.

46. 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.

47. 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-lchi Accident," May 9, 2014, NRC ADAMS Accession No. ML14111A147.

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

Seismic. Electric Power Research Institute," February 2013.

49. 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, NRC ADAMS Accession No. ML13101A379.

50. NRC (E Leeds) Letter to NEI (J Pollock), "Electric Power Research Institute Final Draft Report xxxxx, "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, NRC ADAMS Accession No. ML13106A331.

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A. Document No.: 51-9231248-002 AR EVA Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant

51. *Entergy Calculation EA-EC46465-05, "Anchorage Calculations for Electrical Panels EP-1 901 and EP-2001 ," Revision 0.

52. *Entergy Document, EA-EC48188-02, "Seismic Bracing of Non-Q Block Wall - South Wall of Lube Oil Storage Room," Revision 0.

53. *Entergy Document EC54011, "Fukushima Seismic 2.1 - Expedited Seismic Evaluation Program (ESEP) Report - Owner Acceptance of Report and Background Documentation," the following AREVA documents are captured in the plant document management system:

a. AREVA Document 51-9212955-007, "ESEP Expedited Seismic Equipment List (ESEL) -

Palisades Nuclear Plant."

b. AREVA Calculation 32-9227862-001, "Palisades ESEP HCLPF Calculation - Electrical Cabinet EC-02."

c. AREVA Calculation 32-9227902-001, "Palisades ESEP HCLPF Calculation - 480V Load Center, EB-1 1."

d. AREVA Calculation 32-9227941-001,-"Palisades ESEP HCLPF Calculation - Horizontal Pump, P-55C."

e. AREVA Calculation 32-9227961-001, "Palisades ESER HCLPF Calculation - 125,VDC Main Station Batteries ED-01 andED-02."

f. AREVA Calculation 32-9228097-001, "Palisades ESEP HCLPF Calculation - 480V Motor Control Center, EB-01."

g. AREVA Calculation 32-9228681-001, "Palisades ESEP HCLPF Calculation - Auxiliary Feedwater Controls, EJ-1 051."

h. AREVA Calculation 32-9229831-002, "Palisades ESEP HCLPF Calculation - Primary System Storage Tank, T-81."

i. AREVA Calculation 32-9230336-001, "Palisades ESEP Screening of Lube Oil Storage Room South Block Wall."

j. AREVA Calculation 32-9230249-003, Palisades ESEP HCLPF Calculation - Boric Acid Storage Tanks T-53A & T-53B."

k. AREVA Calculation 32-9228279-002, "Palisades ESEP HCLPF Calculation - Block Wall, C-104.11Q."

I. AREVA Calculation 32-9228841-001, "Palisades ESEP HCLPF Calculation - Block Walls C-107.16Q, C-107.17Q, and C-107.18Q."

The following references are AREVA references which were used as input for Appendix A.

54. AREVA Document 32-9223795-002, "Palisades ESEP Binning and Screening."

55. AREVA Document 51-9230420-001, "Input to Entergy ESEP Report Sections 2 and 3 for Palisades Nuclear Plant."

56. AREVA Document 51-9227749-000, "Input to Entergy ESEP Report Sections 4 and 5 for Palisades Nuclear Plant."

57. AREVA Document 51-9231028-001, "Input to Entergy ESEP Report Sections 6, 7, and 8 for Palisades Nuclear Plant."

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58. AREVA Document 38-9232220-001, "Palisades Nuclear Power Plant ESEP Report Comment Resolution Form."

59. AREVA Document 38-9233072-000, "Palisades Nuclear Power Plant ESEP Report, Revision 001, Comment Resolution Form."

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A Document No.: 51-9231248-002 AR EVA Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant APPENDIX A: EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT FOR PALISADES NUCLEAR PLANT Note: Customer requested formatting begins on the following page.

Page A-1

EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT FOR PALISADES NUCLEAR PLANT Page I

S- Palisades Nuclear Plant ESEP Report Table of Contents Page LIST OF TABLEES ................................................................. .......................................................................... 4 LIST OF FIGURES .......................................................................................................................................... 5 1.0 PURPOSE AND OBJECTIVE ........................................................................................................... 6 2.0 BRIEF

SUMMARY

OF THE FLEX SEISMIC IM PLEMENTATION STRATEGIES ................................. 6 3.0 EQUIPMENT SELECTION PROCESS AND ESEL ............................................................................... 7 3.1 Equipment Selection Process and ESEL ......................................................................... 7 3.1.1 ESEL Development .......................................................................................... 8 3.1.2 Power Operated Valves .................................................................................. 9 3.1.3 Pull Boxes ........................................................................................................ 9 3.1.4 Termination Cabinets ........................................................................................ 9 3.1.5 Critical Instrumentation Indicators ................................................................ 10 3.1.6 Phase 2 and 3 Piping Connections ............................................................... 10 3.2 Justification for Use of Equipment That is Not the Primary Means for FLEX Implementation .................... ............................................................................................ 10 4.0 GROUND MOTION RESPONSE SPECTRUM (GMRS)................................................................... 10 4.1 Plot of GM RS Submitted by the Licensee .................................................................... 10 4.2 Comparison to SSE ........................................................................................................ 12 5.0 REVIEW LEVEL GROUND MOTION (RLGM) .............................................................................. 13 5.1 Description of RLGM Selected ....................................................................................... 13 5.2 Method to Estimate In-Structure Response Spectra (ISRS) .......................................... 15 6.0 SEISMIC MARGIN EVALUATION APPROACH ................................... 15 6.1 Summary of Methodologies Used ................................................................................ 15 6.2 HCLPF Screening Process .............................................................................................. 15 6.3 Seismic W alkdown Approach ....................................................................................... 16 6.3.1 W alkdown Approach .................................................................................... 16 6.3.2 Application of Previous W alkdown Information ............................................ 17 6.3.3 Significant W alkdown Findings ....................................................................... 17 6.4 HCLPF Calculation Process ............................................................................................ 18 6.5 Functional Evaluations of Relays .................................................................................. 19 6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes) ..................................... 19 7.0 INACCESSIBLE ITEMS ..................................................................................................................... 19 7.1 Identification of ESEL Item Inaccessible for W alkdowns ............................................... 19 7.2 Planned W alkdown / Evaluation Schedule / Close Out ............................................... 21 8.0 ESEP CONCLUSIONS AND RESULTS .......................................................................................... 22 8.1 Supporting Information .................................................................................................... 22 Page 2

Palisades Nuclear Plant ESEP Report Table of Contents (continued)

Page 8.2 Identification of Planned Modifications ...................................................................... 23 8.3 Modification Implementation Schedule ....................................................................... 23 8.4 Summary of Regulatory Commitments ......................................................................... 24 9 .0 REFER EN CES .................................................................................................................................. 24 ATTACHMENT A - PALISADES NUCLEAR PLANT ESEL ......................................................................... A-1 ATTACHMENT B - ESEP HCLPF VALUES AND FAILURE MODES TABULATION ..................................... B-1 Page 3

Palisades Nuclear Plant ESEP Report List of Tables Page TABLE 4-1: GM RS FOR PALISADES NUCLEAR PLANT ......................................................................... .. 10 TABLE 4-2: SSE FOR PALISADES NUCLEAR PLANT ................................................................................ 1-2 TABLE 5-1: RLGM FOR PALISADES NUCLEAR PLANT ............................................................................ 14 Page 4

Palisades Nuclear Plant ESEP Report List of Figures Page FIGURE 4-1: GM RS FOR PALISADES NUCLEAR PLANT ......................................................................... 12 FIGURE 4-2: GMRS TO SSE COMPARISON FOR PALISADES NUCLEAR PALISADES ................................ 13 FIGURE 5-1: RLGM FOR PALISADES NUCLEAR PLANT ......................................................................... 14 Page 5

- Palisades Nuclear Plant ESEP-Report 1.0 PURPOSE AND OBJECTIVE Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March 11,.

2011, Great Tohoku Earthquake and subsequient tsunami, the Nuclear Regulatory Commission (NRC) established a Near-Term Task Force (NTTF) to conduct a systematic review of NRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena. Subsequently, the NRC issued a 50.54(f) letter on March 12, 2012 [1], requesting information to assure that these recommendations are addressed by all U.S. nuclear power plants. The 50.54(f) letter 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 Palisades Nuclear Plant. The intent of the ESEP is to perform an interim action in response to the NRC's 50.54(f) letter to demonstrate seismic margin through a review of a subset of the plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events.

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

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

2.0 BRIEF

SUMMARY

OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES The Palisades Nuclear Plant FLEX strategies for Reactor Core Cooling and Heat Removal, Reactor Inventory Control, and Containment Function are summarized below. This summary is derived from the Palisades Nuclear Plant Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order EA-12-049 [3] as augmented by the second six-month Status Report [4], the third six-month Status Report [5] and the FLEX Basis Engineering Change (EC) [6].

During Phase 1 reactor core cooling and heat removal are achieved with feedwater supply to the steam generators using the Turbine-Driven Auxiliary Feedwater (TDAFW) pump aligned to take suction from the Condensate Storage Tank (CST) and heat removal using the Atmospheric Dump Valves (ADVs).

Operation of the ADVs can be accomplished by local operation. Steam supply valves for the TDAFW pump and Auxiliary Feedwater (AFW) flow control valves to the steam generators are also required.

At approximately two (2) hours after the event starts, operators will initiate a controlled cooldown-depressurization by opening the ADVs. Decreasing the Primary Coolant System (PCS) pressure permits the Safety Injection Tanks (SITs) to provide injection of borated water to the PCS to make-up for volume shrinkage and any minor inventory losses.

Page 6

Palisades Nuclear Plant ESEP Report Prior to depletion of the CST, the primary makeup storage tank will be cross-connected to the CST for additional water. The combined capacity of these tanks provides about eight (8) hours of steam generator feed.

The Phase 2 strategy involves switching steam generator feed to an on-site portable, diesel driven FLEX pump taking water directly from Lake Michigan and pumping it to the steam generators.

Reactor Coolant System (RCS) inventory control will use the installed charging pumps, powered by a portable FLEX diesel generator, with borated water for injection from the installed Concentrated Boric Acid Storage Tanks (CBASTs). Prior to depletion of the CBASTs, boric acid batching operations will begin using the installed boric acid batching tank.

Containment function is not challenged early in the event and no actions are required during Phase 1 or Phase 2. During Phase 3, the strategy involves use of Regional Response Center (RRC) equipment to provide containment cooling via the Service Water System and the Containment Air Coolers and Fans.

Necessary electrical components are outlined in the Palisades Nuclear Plant FLEX OIP submittal, Second Six-Month Status Report and FLEX Basis, and primarily entail 480 volt AC Motor Control Centers (MCCs), 125 volt DC MCCs, vital batteries, battery chargers, and 120 volt AC distribution panels. Other supporting components include monitoring instrumentation for core cooling, reactor coolant inventory, and containment integrity. The FLEX strategy includes operators shedding unnecessary DC loads to extend the battery life into Phase 2.

The figures provided in Attachment'3 of Reference [3] provide the conceptual FLEX flow paths for Palisades Nuclear Plant Phases 1, 2 and 3.

3.0 EQUIPMENT SELECTION PROCESS AND ESEL The selection of equipment for the Expedited Seismic Equipment List (ESEL) followed the guidelines of EPRI 3002000704 [2]. The ESEL for Palisades Nuclear Plant is presented in Attachment A. Information presented in Attachment A is drawn from the following references [3], [4],[5], [6], [7], [8], [9], [10],

[11], [12], [13], [14], [15], [16], [17], [18], [19], [20], [21], [22], [23], [24], [25], [26], [27], [28], [29],

[30], [31], [32], [33], [34], [35], [36], [37], and [38].

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 Palisades Nuclear Plant OIP in Response to the March 12, 2012, Commission Order EA-12-049 [3], second six-month Status Report [4], third six-month Status Report

[5], and the FLEX Basis Engineering Change [6]. These references provide the Palisades Nuclear Plant FLEX mitigation strategy and serve 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 Palisades Nuclear Plant OIP, second six-month Status Report, and the FLEX Basis. FLEX recovery actions are excluded from the ESEP scope per EPRI 3002000704 [2]. The overall list of planned FLEX modifications and the scope for consideration herein is limited to those required to support core cooling, reactor coolant inventory, and containment integrity functions. The scope of the ESEL does not include Spent Fuel Pool (SFP) components since the SFP is excluded by EPRI 3002000704 [2]. Portable and pre-staged FLEX equipment (not permanently installed) are excluded from the ESEL per EPRI 3002000704.

Page 7

Palisades Nuclear Plant ESEP Report The ESEL component selection followed the EPRI guidance outlined in Section 3.2 of EPRI 3002000704.

1. The scope of components is limited to that required to accomplish the core cooling and containment-safety functions identified~inTable 3-2 of EPRI 3002000704. The instrumentation monitoring requirements for core cooling/containment safety functions are limited to those outlined in the EPRI 3002000704 guidance, and are a subset of those outlined in the Palisades Nuclear Plant OIP, second six-month Status Report, third six-month Status Report, and the FLEX Basis Engineering Change.

2. The scope of components is limited to installed plant equipment, and FLEX connections necessary to implement the Palisades Nuclear Plant OIP, second six-month Status Report, third six-month Status Report, and FLEX Basis Engineering Change as described in Section 2 in this report.

3. The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e.,

either "Primary" or "Back-up/Alternate").

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

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

6. Structures, systems, and components excluded per the EPRI 3002000704 guidance are:

• Structures (e.g. containment, reactor building, controlbuilding, auxiliary building, etc.)

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

• Manual valves, check valves and rupture disks

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

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

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 Palisades Nuclear Plant OIP [3], second six-month Status Report [4], third six-month Status Report [5] and FLEX Basis Engineering Change [6] to determine the major equipment involved in the FLEX strategies. Further reviews of plant drawings (e.g., Flow Diagrams 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) in branch circuits / branch lines off the defined strategy electrical or fluid flow path. Flow Diagrams were the primary reference documents used to identify mechanical components. The flow paths used for FLEX strategies were selected and specific components were identified using detailed equipment and instrument drawings, electrical schematics and one-line drawings, system descriptions, and design basis documents, as necessary.

Page 8

- Palisades Nuclear Plant ESEP Report Cabinets containing electrical equipment and instrumentation, which could be affected by earthquake motion and potentially impact the operation of equipment on the ESEL, are required to be included on the ESEL for evaluation.

For Phase 1, TDAFW is the primary path for core cooling. For Phase 2, a portable diesel-driven FLEX pump is used. For Phase 3, the RRC equipment is used to provide the power and pumps for core cooling and inventory control. An extensive relay evaluation was performed as part of the A-46 and IPEEE programs and after screening out relays for which relay chatter is not an issue, no bad actors were identified. In addition, due to the Extended Loss of AC Power (ELAP) from the BDBEE, without power, chatter of relays supporting higher voltage equipment will not result in any negative effects from energized relay seal-in or lock-out. Therefore, no relays were listed on the ESEL.

For each parameter monitored during the FLEX implementation, a single indication loop was selected for inclusion in the ESEL per EPRI 3002000704 [2). For each parameter indication, the components along the flow path from measurement to indication were included, since any failure along the path would lead to failure of that indication. Components such as flow elements were considered as part of the piping and were excluded from the ESEL.

3.1.2 Power Operated Valves Page 3-3 of EPRI 3002000704 [2] notes that power operated valves not required to change state are excluded from the ESEL. Page 3-2 also notes that "functional failure modes of electrical and mechanical portions of the installed Phase 1 equipment should be considered (e.g. AFW trips)." To address this concern, the following guidance is applied in the Palisades Nuclear Plant ESEL for functional failure modes associated with power operated valves:

• Power operated valves that remain energized during the ELAP events (such as Direct Current (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 be added to the ESEL as these components provide completely passive locations for pulling or installing cables. No breaks or connections in the cabling were included in pull boxes. Pull boxes were considered part of conduit and cabling, which were excluded in accordance with EPRI 3002000704 [21.

3.1.4 Termination Cabinets Termination cabinets necessary for FLEX Phase 2 and Phase 3 connections from FLEX components, provide consolidated locations for permanently connecting multiple cables. The termination cabinets and the internal connections provide a completely passive function; however, the cabinets are Page 9

Palisades Nuclear Plant ESEP Report included in the ESEL to ensure industry knowledge on panel/anchorage failure vulnerabilities are addressed.

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

3.1.6 Phase 2 and 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 Palisades Nuclear Plant OIP, second six-month Status Report, third six-month Status Report, and the FLEX Basis Engineering Change as described in Section 2." Item 3 in Section 3.1 also notes that "The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e., either "Primary" or "Back-up/Alternate")."

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

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

3.2 Justification for Use of Equipment That is Not the Primary Means for FLEX Implementation All equipment on the ESEL is part of the primary means of FLEX implementation. Therefore, no additional justification is required.

4.0 GROUND MOTION RESPONSE SPECTRUM (GMRS) 4.1 Plot of GMRS Submitted by the Ucensee The Safe Shutdown Earthquake (SSE) control point elevation is defined at the ground surface, in the free field [39]. Table 4-1 shows the GMRS acceleration for a range of frequencies [40]. The GMRS at the control point is shown in Figure 4-1.

Table 4-1: GMRS for Palisades Nuclear Plant Frequency GMRS (Hz) (g) 100 2.83E-01 90 2.87E-01 80 2.92E-01 70 2.99E-01 60 3.13E-01 50 3.41E-01 40 3.87E-01 35 1 4.13E-01 Page 10

Palisades Nuclear Plant ESEP Report Table 4-1: GMRS for Palisades Nuclear Plant (continued)

Frequency GMRS (Hz) (g) 30 4.46E-01 25 4.93E-01 20 5.11E-01 15 5.46E-01 12.5 5.43E-01 10 5.49E-01 9 5.59E-01 8 5.49E-01 7 5.08E-01 6 4.82E-01 5 4.66E-01 4 4.38E-01 3.5 4.16E-01 3 3.67E-01 2.5 3.07E-01 2 2.90E-01 1.5 2.25E-01 1.25 1.90E-01 1 1.49E-01 0.9 1.30E-01 0.8 1.11E-01 0.7 9.47E-02 0.6 7.97E-02 0.5 6.51E-02 0.4 5.21E-02 0.35 4.56E-02 0.3 3.91E-02 0.25 3.26E-02 0.2 2.60E-02 0.15 1.95E-02 0.125 1.63E-02 0.1 1.30E-02 Page 11

Palisades Nuclear Plant ESEP Report-GMRS at Control Point for Palisades Nuclear Plant, 5% Damping 0.60 0.1 10 100 Frequency (Hz)

Figure 4-1: GMRS for Palisades Nuclear Plant 4.2 Comparison to SSE The SSE is defined in the Final Safety Analysis Report [39] in terms of a Peak Ground Acceleration (PGA) and a design response spectrum. These spectra have been digitized and tabulated [40]. Table 4-2 shows the spectral acceleration values at selected frequencies for the 5% damped horizontal SSE.

Table 4-2: SSE for Palisades Nuclear Plant Frequency Spectral (Hz) Acceleration (g) 100 0.2 25 0.206 10 0.24 5 0.31 2.5 0.285 1 0.16 0.5 0.096 Page 12

Palisades Nuclear Plant ESEP Report GMRS to SSE Comparison for Palisades Nuclear Plant, 5% Damping 0.60 0.30 0.1 10 100 FrequencytHz)

Figure 4-2: GMRS to SSE Comparison for Palisades Nuclear Palisades The SSE envelops the GMRS in the low frequency range up to nearly 1.5 Hz. The GMRS exceeds the SSE beyond that point. As the GMRS exceeds the SSE in the I to 10 Hz range, the plant does not screen out of the ESEP according to Section 2.2 of EPRI 3002000704 [2]. The two special screening considerations as described in Section 2.2.1 of EPRI 3002000704, namely a) Low Seismic Hazard Site and b) Narrow Band Exceedances in the 1 to 10 Hz range, provide criteria for accepting specific GMRS exceedances.

However, the GMRS exceedances are not limited to the low frequency range and there are no narrow-banded exceedances. Therefore, these special screening considerations do not apply for Palisades Nuclear Plant and hence High Confidence of a Low Probability of Failure (HCLPF) evaluations are to be performed.

5.0 REVIEW LEVEL GROUND MOTION (RLGM) 5.1 Description of RLGM Selected The RLGM is selected based on Approach 1 in Section 4 of EPRI 3002000704 [2]. The RLGM is developed based on the SSE [41].

The maximum GMRS/SSE ratio between I and 10 Hz range occurs at 10 Hz where the ratio is 0.549/0.24 = 2.29. As the maximum ratio of the GMRS to the SSE over the I to 10 Hz range exceeds a value of 2, the GMRS/SSE ratio is set to the maximum scaling factor value of 2.0 for Palisades Nuclear Plant in accordance with Section 4 of EPRI 3002000704. Table 5-1 lists the horizontal ground RLGM acceleration at 5% damping at selected frequencies and the plot is shown in Figure 5-1. The RLGM is generated by plotting the digitized data on a log/linear graph paper, and connecting the points with straight lines.

Page 13

Palisades Nuclear Plant ESEP Report Table 5-1: RLGM for Palisades Nuclear Plant Frequency RLGM at 5%

(Hz) Damping (g) 100.00 0.400 33.00 0.400 15.00 0.432 10.00 0.480 6.60 0.620 3.00 0.620 1.00 0.320 0.50 0.192 0.10 0.000 Review Level Ground Motion (2xSSE) Response Spectra - Horizontal Direction 0.70 0.60 0.50 0.40 0.30 0.20 0.10 0.00 -K 0.100 1.000 10.000 100.000 Fmquency (Wz)

Figure 5-1: RLGM for Palisades Nuclear Plant Page 14

Palisades Nuclear Plant ESEP Report 5.2 Method to Estimate In-Structure Response Spectra (ISRS)

The RLGM ISRS for Palisades Nuclear Plant are generated by scaling the SSE ISRS [41]. The following steps are used to generate the RLGM ISRS:

1. Obtain the horizontal direction SSE ISRS for a particular damping value.

2. Calculate the horizontal RLGM ISRS by scaling the horizontal direction SSE ISRS by a factor of 2.0.

3. Repeat steps I and 2 to obtain RLGM ISRS for multiple damping values.

The vertical direction RLGM ISRS is obtained by scaling the vertical amplified ground response spectrum.

6.0 SEISMIC MARGIN EVALUATION APPROACH It is necessary to demonstrate that ESEL items have sufficient seismic capacity to meet or exceed the demand characterized by the RLGM. The seismic capacity is characterized as the PGA for which there is a HCLPF. The PGA is associated with a specific spectral shape, in this case the 5%-damped RLGM spectral shape. The HCLPF capacity must be equal to or greater than the RLGM PGA. The criteria for seismic capacity determination are given in Section 5 of EPRI 3002000704 [2].

There are two basic approaches for developing HCLPF capacities:

1. Deterministic approach using the Conservative Deterministic Failure Margin (CDFM) methodology of EPRI NP-6041-SL, A Methodology for Assessment of Nuclear Power.Plant Seismic Margin (Revision 1) [42].

2. Probabilistic approach using the fragility analysis methodology of EPRI TR-103959, Methodology for Developing Seismic Fragilities [43].

6.1 Summary of Methodologies Used Palisades Nuclear Plant was classified as a 0.3g focused scope in NUREG-1407 [44] and performed a Seismic Probabilistic Risk Assessment (SPRA) in accordance with the methodology of NUREG-1407 in 1996 as part of Individual Plant Examination for External Events (IPEEE) program. The SPRA is documented in [45] and consisted of screening evaluations, seismic walkdowns, and fragility analysis.

SPRA screening was performed in accordance with EPRI NP-6041-SL [42]. Seismic walkdowns took advantage of overlapping requirements between IPEEE and USI A-46 programs. Section 3.3 of [40]

established that the results of the Palisades Nuclear Plant IPEEE are not adequate to support screening of the updated seismic hazard for Palisades Nuclear Plant.

For ESEP, the screening walkdowns used the screening tables from Chapter 2 of EPRI NP-6041-SL. The walkdowns were conducted by engineers trained in EPRI NP-6041-SL and were documented on Screening Evaluation Work Sheets (SEWS) from EPRI NP-6041-SL. Anchorage capacity calculations used the CDFM criteria from EPRI NP-6041-SL. Seismic demand was based on EPRI 3002000704 [2] using an RLGM of 2.OOxSSE with a PGA of 0.40g, Figure 5-1.

6.2 HCLPF Screening Process For ESEP, the components are screened considering the RLGM (2.OOxSSE) with a 0.40g PGA. The screening tables in EPRI NP-6041-SL [42] are based on ground peak spectral accelerations of 0.8g and 1.2g. These both exceed the RLGM peak spectral acceleration.

Page 15

Palisades Nuclear Plant ESEP Report The ESEL components were prescreened based on Table 2-4 of EPRI NP-6041-SL. Additional pre-screening, specifically for anchorage, considered walkdown results and documentation from NTTF 2.3 and SEWS from IPEEE and USI A-46. Equipment anchorage was screened out in cases where previous evaluations showed large available margin against SSE. The remaining components (i.e., components that do not screen out), were identified as requiring HCLPF calculations. ESEL components were walked down and based on the equipment and anchorage conditions, prescreening decisions were confirmed and a final list of required HCLPF calculations was generated. Equipment for which the screening caveats were met and for which the anchorage capacity exceeded the RLGM seismic demand are screened out from ESEP seismic capacity determination because the HCLPF capacity exceeds the RLGM.

The Palisades Nuclear Plant ESEL contains 178 items. Of these, 47 are valves. In accordance with Table 2-4 of EPRI NP-6041-SL, active valves may be assigned a functional capacity of 0.8g peak spectral acceleration without any review other than looking for valves with large extended operators on small diameter piping, and anchorage is not a failure mode. The review of valves with large extended operators on small diameter piping is performed during walkdowns. Significant walkdown findings are summarized in Section 6.3. Therefore, valves on the ESEL are screened out from ESEP seismic capacity determination, subject to the caveat regarding large extended operators on small diameter piping.

The non-valve components in the ESEL are screened based on the SMA results. If the SMA showed that the component met the EPRI NP-6041-SL screening caveats and the CDFM capacity exceeded the RLGM demand, the components are screened out from the ESEP capacity determination.

Block Walls were identified in the proximity of ESEL equipment. The HCLPF capacity of these walls was evaluated and determined to be sufficient to meet RLGM demand.

6.3 Seismic Walkdown Approach 6.3.1 Walkdown Approach Walkdowns were performed in accordance with the criteria provided in Section 5 of EPRI 3002000704

[2], which refers to EPRI NP-6041-SL [42] for the SMA process. Pages 2-26 through 2-30 of EPRI NP-6041-SL describe the seismic walkdown criteria, including the following key criteria.

"The SRT [Seismic Review Team] should "walk by" 100% of all components which are reasonably accessible and in non-radioactiveor low radioactiveenvironments. Seismic capability assessment of components which are inaccessible,in high-radioactiveenvironments, or possibly within contaminated containment, 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 necessaryto inspect one component out of this group. The "similarity-basis"should be developed before the walkdown during the seismic capability preparatory work (Step 3) by reference to drawings,calculations or specifications. The one component or each type which is selected should be thoroughly inspected which probably does mean de-energizing and opening cabinets or panelsfor this very limited sample. Generally, a spare representativecomponent can be Page 16

Palisades Nuclear Plant ESEP Report found so as to enable the inspection to be performed while the plant is in operation. At leastfor the one component of each type which is selected, anchorageshould be thoroughly inspected.

The walkdown procedure should be performed in an ad hoc manner. Foreach class of components the SRT should look closely at the first items and compare the field configurationswith the construction drawingsand/orspecifications. If a one-to-one correspondenceis found, then subsequent items do not have to be inspected in as great a detail. Ultimately the walkdown becomes a "walk by" of the component class as the SRT becomes confident that the constructionpattern is typical. This procedure for 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 drawings or questionableconstructionpracticesare found then the system or component class must be inspected in closer detail until the systematic deficiency is defined.

The 100% "walk by" is to look for outliers, lack of similarity,anchorage which is differentfrom that shown on drawings or prescribedin criteriafor that component, potentialSI [Seismic Interaction]

problems, situations that are at odds with the team members' past experience, and any other areas of serious seismic concern. If any such concerns surface, then the limited sample size of one component of each type for thorough inspection will have to be increased. The increase in sample size which should be inspected will depend upon the number of outliers and different anchorages,etc., which are observed. It is up to the SRT to ultimately select the sample size since they are the ones who are responsiblefor the seismic adequacy of all elements which they screen from the margin review.

Appendix 6 gives guidancefor sampling selection."'

6.3.2 Application of Previous Walkdown Information Several ESEL items were previously walked down during the Palisades Nuclear Plant seismic IPEEE program, for the USI A-46 evaluation program, and NTTF Recommendation 2.3. Those walkdown results were reviewed and the following steps were taken to confirm that the previous walkdown conclusions remained valid.

" A walk by was performed to confirm that the equipment material condition and configuration is consistent with the walkdown conclusions and that no new significant interactions related to block walls or piping attached to tanks exist.

" If the ESEL item was screened out based on previous walkdowns, that screening evaluation was reviewed and reconfirmed for the ESEP.

6.3.3 Significant Walkdown Findings Consistent with the guidance from EPRI NP-6041-SL [42], no significant outliers or anchorage concerns were identified during the Palisades Nuclear Plant seismic walkdowns. Based on pre-screening of components and walkdown results, HCLPF capacity evaluations were recommended for the following ten (10) components:

9 Electrical Cabinet, EC-02

• 480V Load Center, EB-11

• Horizontal Pump, P-55C 0 125 VDC Main Station Battery, ED-01 Page 17

Palisades Nuclear Plant ESEP Report

• 125 VDC Main Station Battery, ED-02

• 480V Motor Control Center, EB-01

" Auxiliary Feedwater Controls, EJ-1051

" Primary Makeup Storage Tank, T-81

• Boric Acid Storage Tank, T-53A

• Boric Acid Storage Tank, T-53B Block walls were identified in the proximity of ESEL equipment. These block walls were assessed for their structural adequacy to withstand the seismic loads resulting from the RLGM. For any cases where the block wall represented the HCLPF failure mode for an ESEL item, it is noted in the tabulated HCLPF values described in Section 6.6. Three (3) HCLPF evaluations were performed addressing the block walls in close proximity to components ED-01, ED-02, ED-10L, ED-1OR, ED-20L, ED-20R, EB-23, EC-33, and N2 Backup Station #2.

• C-104.11Q

" C-107.16Q0 C-107.17QC and C-107.18Q

• Lube Oil Storage Room South Block Wall 6.4 HCLPF Calculation Process ESEL items identified for ESEP at Palisades Nuclear Palisades Were evaluated using the criteria in EPRI NP-6041-SL [42] and Section 5 of EPRI 3002000704 [2]. Those evaluations included the following steps:

" Performing seismic capability walkdowns for equipment not included in previous seismic walkdowns (SQUG, IPEEE, USI A-46, or NTTF 2.3) to evaluate the equipment installed plant conditions

" Performing screening evaluations using the screening tables in EPRI NP-6041-SL as described in Section 6.2

" 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. A'total of eight (8) HCLPF calculations were performed to address the ten (10) components as well as three (3) calculations to address seismic adequacy of block walls.

• Electrical Cabinet EC-02

• 480V Load Center, EB-11

• Horizontal Pump, P-55C

• 125 VDC Main Station Batteries ED-01 and ED-02

• 480V Motor Control Center, EB-01

• Auxiliary Feedwater Controls, EJ-1051 0 Primary Makeup Storage Tank, T-81 Page 18

Palisades Nuclear Plant ESEP Report

" Boric Acid Storage Tanks T-53A & T-53B

• Lube Oil Storage Room South Block Wall

" Block Wall, C-104.11Q

" Block Walls C-107.160, C-107.170, and C-107.18Q 6.5 Functional Evaluations of Relays As discussed in Section 3.1.1, no seal in/lockout type relays were identified on Palisades Nuclear Plant ESEL. Therefore, no relay evaluations were performed.

6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes)

Tabulated ESEL HCLPF values are provided in Attachment B. The following notes apply to the information in the tables.

" For items screened out using EPRI NP-6041-SL [42] screening tables, the HCLPF capacity is provided as >RLGM and the failure mode is listed as "Screened", (unless the controlling HCLPF value is governed by anchorage).

• For items where anchorage controls the HCLPF value, the HCLPF value is listed in the table and the failure mode is noted as "anchorage." For the items where the component function controls the HCLPF value, the HCLPF value is listed in the table and the failure mode is'noted as:

"functional."

After performing the HCLPF calculations, ESEL components were determined to have adequate capacity for the design basis loads and HCLPF greater than RLGM for all components except the following:

• T-53A, Boric Acid Storage Tank

• T-53B, Boric Acid Storage Tank

" T-81, Primary Makeup Storage Tank Modifications are required for the above components.

7.0 INACCESSIBLE ITEMS 7.1 Identification of ESEL Item Inaccessible for Walkdowns Thirty-six (36) components on the ESEL were inaccessible and not walked down because the walkdowns were performed with the plant on-line. These components are located in the Reactor Containment Building as well as in a locked, high radiation area within the Auxiliary Building. One component (LT-0372, Safety Injection Tank T-82C Level Transmitter) was evaluated based on review of existing photographs. The remaining thirty-five (35) components listed below require follow-up seismic walkdowns.

" CV-0861, Service Water System Return Isolation Valve from CAC Fan VHX-1

• CV-0864, Service Water System Return Isolation Valve from CAC Fan VHX-2 Page 19

Palisades Nuclear Plant ESEP Report SV-0861, Service Water System Return Isolation Valve from CAC Fan VHX-1

- SV-0864, Service Water System Return Isolation Valve from CAC Fan VHX-2

" V-1A, Containment Air Cooler Fan

" V-2A, Containment Air Cooler Fan

• V-3A, Containment Air Cooler Fan

" VHX-1, Containment Air Cooler

" VHX-2, Containment Air Cooler

• VHX-3, Containment Air Cooler

• CV-2083, CBO Isolation Valve

• CV-2099, CBO Isolation Valve

" CV-2191, CBO Relief Stop Valve

" E-56, Regenerative Heat Exchanger

" MO-3007, HPSI Injection Loop Isolation Valve

• MO-3041, SIT T-82A Isolation Valve

• MO-3045, SIT T-82B Isolation Valve

" MO-3049, SIT T-82C Isolation Valve

• MO-3052, SIT T-82D Isolation Valve

" LT-0102, PZR Level (WR) Transmitter

• LT-0365, SIT T-82A Level Transmitter

" LT-0368, SIT T-82B Level Transmitter

• LT-0374, SIT T-82D Level Transmitter

• LT-0757A, SG A WR Level Transmitter

• LT-0758A, SG B WR Level Transmitter Page 20

Palisades Nuclear Plant.ESEP Report

" PT-0105A, PZR Pressure (WR) Transmitter

• PT-0751C, SG A Pressure Transmitter

" PT-0752C, SG B Pressure Transmitter

" TE-0112CC, PCS T-cold Temperature Element

• TE-O112HC, PCS T-hot Temperature Element

• LE-0101A, Reactor Vessel Level

" T-82A, Safety Injection Tank A

" T-82B, Safety Injection Tank B

• T-82C, Safety Injection Tank C

• T-82D, Safety Injection Tank D 7.2 Planned Walkdown / Evaluation Schedule / Close Out The walkdowns of the inaccessible items identified in Section 7.1 are currently scheduled to be performed during Palisades' next refueling outage with subsequent evaluations of those components to follow. Section 8.4 summarizes regulatory commitments for the close out of inaccessible components.

Page 21

Palisades Nuclear Plant ESEP Report 8.0 ESEP CONCLUSIONS AND RESULTS 8.1 Supporting Information Palisades Nuclear Plant has performed the ESEP as an interim action in response to the NRC's 50.54(f) letter [1]. It was performed using the methodologies in the NRC endorsed guidance in EPRI 3002000704 [2].

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

The ESEP is part of the overall Palisades Nuclear Plant response to the NRC's 50.54(f) letter. On March 12, 2014, NEI submitted to the NRC results of a study [46] 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 [47] coricluded that the "fleet wide 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 Palisades Nuclear Plant was included in the fleet risk evaluation submitted in the March 12, 2014 NEI letter [46] therefore, the conclusions in the NRC's May 9 letter also apply to Palisades Nuclear Plant.

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

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

• Response spectra enveloping criteria typically used in SSC analysis and testing applications Page 22

Palisades Nuclear Plant ESEP Report

" 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, etc.)

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

The intent of the ESEP is to perform an interim action in response to the NRC's 50.54(f) letter to demonstrate seismic margin through a review of a subset of the plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events. The RLGM used for the ESEP evaluation is a scaled version of the plant's SSE rather than the actual GMRS. To more fully characterize the risk impacts of the seismic ground motion represented by the GMRS on a plant specific basis, a more detailed seismic risk assessment (SPRA or risk-based SMA) is to be performed in accordance with EPRI 1025287 [48]. As identified in the Palisades Nuclear Plant Seismic Hazard and GMRS submittal [40], Palisades Nuclear Plant screens in for a risk evaluation. The complete risk evaluation will more completely characterize the probabilistic seismic ground motion input into the plant, the plant response to that probabilistic seismic ground motion input, and the resulting plant risk characterization. Palisades Nuclear Plant will complete that evaluation in accordance with the schedule identified in NEI's letter dated April 9, 2013 [491 and endorsed by the NRC in their May 7, 2013 letter [50].

8.2 Identification of Planned Modifications Insights from the ESEP identified the following items where the HCLPF is below the RLGM and plant modifications will be made in accordance with EPRI 3002000704 [2] to enhance the seismic capacity of the plant.

" T-53A, Boric Acid Storage Tank

" T-53B, Boric Acid Storage Tank T-53B

" T-81, Primary Makeup Storage Tank 8.3 Modification Implementation Schedule Plant modifications will be performed in accordance with the schedule identified in NEI letter dated April 9, 2013 [49], which states that plant modifications not requiring a planned refueling outage will be completed by December 2016 and modifications requiring a refueling outage will be completed within two planned refueling outages after December 31, 2014.

Page 23

Palisades Nuclear Plant ESEP Report 8.4 Summary of Regulatory Commitments The following actions will be performed as a result of the ESEP.

~ Equl~mnt Act~ons # ID DesntioH Perform seismic walkdowns, No later than the generate HCLPF calculations end of the second and design and implement planned refueling N/A any necessary modifications outage after 1 N/A for inaccessible items listed in December 31, 2014.

Section 7.1 Walkdowns planned for the next refueling outage.

2 T-53A Boric Acid Modify tank anchorage such As described in Storage Tank that HCLPF>RLGM Section 8.3 3 Boric Acid Modify tank anchorage such As described in Storage Tank that HCLPF>RLGM Section 8.3 4 T-81 Primary Makeup Modify tank such that As described in Storage Tank HCLPF>RLGM Section 8.3

Submit a letter td.NRC . Within 60 days

.. summarizing the HCLPF following results of Items 1 to 4 and completion of ESEP 5 N/A N/A confirming implementation of activities, including the plant modifications items 1 to 4 associated with items 1 to 4 Page 24

Palisades Nuclear Plant ESEP Report

9.0 REFERENCES

1. NRC (E Leeds and M Johnson) Letter to All Power Reactor Licensees et al., "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident," March 12, 2012, NRC ADAMS Accession No. ML12053A340.

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

3. Entergy Letter to U.S. NRC, letter number PNP 2013-010, "Overall Integrated Plan in Response to March 12, 2012 Commission Order to Modify Licenses With Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049),"

February 28, 2013, NRC ADAMS Accession No. ML13246A399.

4. Entergy Letter to U.S. NRC, Letter Number PNP 2014-011, "Palisades Nuclear Plant 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 28, 2014, NRC ADAMS Accession No. ML14059A078.

5. Entergy Letter to U.S. NRC, Letter Number PNP 2014-085, "Palisades Nuclear Plant 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 28, 2014, NRC ADAMS Accession No. ML14240A278.

6. Entergy Document, Engineering Change EC-46465, "FLEX Basis."

7. Entergy Drawing E0001, Sheet 1, Revision 83, "Single Line Meter & Relay Diagram 480V Motor Control Center Warehouse, WD 950."

8. Entergy Drawing E0001, Sheet 3, Revision 4, "Plant Single Line Diagram, WD 950."

9. Entergy Drawing E0008, Sheet 1, Revision 57, "Single Line Meter & Relay Diagram 125V DC 120V Instrument & Preferred AC, System WD 950."

10. Entergy Drawing E0008, Sheet 2, Revision 55, "Single Line Meter & Relay Diagram 125V DC 120V Instrument & Preferred AC, System WD 950."

11. Entergy Drawing E0078, Sheet 2, Revision 18, "WRSGL Schematic Diagram."

12. Entergy Drawing E0078, Sheet 2A, Revision 6, "WRSGL Schematic Diagram."

13. Entergy Drawing E0082, Sheet 5, Revision 13, "Schematic Diagram Wide Range Pressurizer Level Indicator/Alarm Instrumentation."

14. Entergy Drawing E0090, Sheet 5, Revision 14, "Schematic Diagram Boric Acid System Instrumentation."

15. Entergy Drawing E0092, Revision 14, "Schematic Diagram Safety Injection Tank Level Indicator Alarm Instrumentation."

16. Entergy Drawing M0201, Sheet 1, Revision 87, "Piping & Instrument Diagram, Primary Coolant System."

Page 25

Palisades Nuclear Plant ESEP Report

17. Entergy Drawing M0201, Sheet 2, Revision 66, "Piping & Instrument Diagram, Primary Coolant System."

18. Entergy Drawing M0202, Sheet 1, Revision 76, "Piping & Instrument Diagram, Chemical &

Volume Control System."

19. Entergy Drawing M0202, Sheet 1A, Revision 64, "Piping & Instrument Diagram, Chemical &

Volume Control System."

20. Entergy Drawing M0202, Sheet 1B, Revision 59, "Piping & Instrument Diagram, Chemical &

Volume Control System."

21. Entergy Drawing M0203, Sheet 1, Revision 48, "Piping & Instrument Diagram, Safety Injection, Containment Spray, Shutdown Cooling System."

22. Entergy Drawing M0203, Sheet 2, Revision 27, "Piping & Instrument Diagram, Safety Injection, Containment Spray, Shutdown Cooling System."

23. Entergy Drawing M0205, Sheet 2, Revision 69, "Piping & Instrument Diagram, Main Steam and Auxiliary Turbine Systems."

24. Entergy Drawing M0208, Sheet 1B, Revision 37, "Piping & Instrument Diagram, Service Water System."

25. Entergy Drawing M0220, Sheet 1, Revision 97, "'Piping & Instrument Diagram, Make-up

. .Domestic Water & Chemical Injection Systems."

26. Entergy Drawing M0207, Sheet 1, Revision 91, "Piping & Instrument Diagram, Feedwater &

Condensate System."

27. Entergy Drawing M0207, Sheet 2, Revision 38, "Piping & Instrument Diagram, Auxiliary Feedwater System."

28. Entergy Procedure Palisades EOP-3.0, Revision 16, "Station Blackout."

29. Entergy Drawing E0364, Revision 9, "Conduit and Tray Miscellaneous Plans."

30. Entergy Drawing E0618, Sheet 569, Revision 12, "Connection Diagram Junction Box J569."

31. Entergy Drawing M0222, Sheet 2, Revision 29 (EC-47346), "Piping & Instrument Diagram, Miscellaneous Gas Supply Systems."

32. Entergy Drawing M0218, Sheet 2, Revision 61, "Piping & Instrument Diagram, Htg. Vent. & Air Cond. Containment Building."

33. Entergy Drawing E0005, Sheet 5B, Revision 12, "Single Line Meter & Relay Diagram 480 Volt Motor Control Centers, System WD 950."

34. Entergy Drawing E0099 Sh. 5, Revision 10, "Schematic Diagram, Containment Building Instrumentation (Left Channel)."

35. Entergy Drawing E0084 Sh. 6, Revision 14, "Schematic Diagram, Pressurizer Pressure Control and Measurement Channel Instrumentation."

36. Entergy Drawing E0087, Sh. 6, Revision 10, "Schematic Diagram Level Indication and Alarm Indication."

Page 26

Palisades Nuclear Plant ESEP Report

37. Entergy Drawing E0076, Sheet 4C, Revision 8, "Schematic Diagram Feedwater and Turbine Driver Instrumentation."

38. Entergy Drawing E0226, Sheet 1B, Revision 6, "Schematic Diagram - Reactor Vessel Level Monitoring System (Left Channel)."

39. "Palisades Nuclear Plant - Final Safety Analysis Report," Revision 31, Docket 50-255, September 2014.

40. Entergy Letter to U.S. NRC, letter number PNP 2014-033,."Palisades Nuclear Plant Seismic Hazard and Screening Report (CEUS Sites), Response to NRC 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 March 31, 2014, NRC ADAMS Accession No. ML14090A069.

41. Entergy Technical Specification C-175(Q), Revision 6, "Requirements for Seismic Evaluation of Electrical and Mechanical Equipment."

42. EPRI-NP-6041-SL, "Methodology for Assessment of Nuclear Power Plant Seismic Margin,"

Revision 1, August 1991.

43. EPRI TR-103959, "Methodology for Developing Seismic Fragilities," July 1994.

44. NRC NUREG71407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," June 1991.

45. Entergy Document, "Palisades Nuclear Plant Individual Plant Examination of External Events (IPEEE)," Revision 1, May 1996.

46. 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.

47. 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-lchi Accident," May 9, 2014, NRC ADAMS Accession No. ML14111A147.

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

Seismic. Electric Power Research Institute," February 2013.

49. 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, NRC ADAMS Accession No. ML13101A379.

50. NRC (E Leeds) Letter to NEI (J Pollock), "Electric Power Research Institute Final Draft Report xxxxx, "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, NRC ADAMS Accession No. ML13106A331.

Page 27

Palisades Nuclear Plant ESEP Report

51. Entergy Calculation EA-EC46465-05, "Anchorage Calculations for Electrical Panels EP-1901 and EP-2001," Revision 0.

52. Entergy Document, EA-EC48188-02, "Seismic Bracing of Non-Q Block Wall - South Wall of Lube Oil Storage Room," Revision 0.

53. Entergy Document EC54011, "Fukushima Seismic 2.1 - Expedited Seismic Evaluation Program (ESEP) Report - Owner Acceptance of Report and Background Documentation," the following AREVA documents are captured in the plant document management system:

a. AREVA Document 51-9212955-007, "ESEP Expedited Seismic Equipment List (ESEL) -

Palisades Nuclear Plant."

b. AREVA Calculation 32-9227862-001, "Palisades ESEP HCLPF Calculation - Electrical Cabinet EC-02."

c. AREVA Calculation 32-9227902-001, "Palisades ESEP HCLPF Calculation - 480V Load Center, EB-11."

d. .AREVA Calculation 32-9227941-001, "Palisades ESEP HCLPF Calculation - Horizontal Pump, P-55C."

e. AREVA Calculation 32-9227961-001, "Palisades ESEP HCLPF Calculation - 125 VDC Main Station Batteries-ED-01 and ED-02."

f. AREVA Calculation 32-9228097-001, "Palisades ESEP HCLPF Calculation - 480V Motor Control Center, EB-01."

g. AREVA Calculation 32-9228681-001, "Palisades ESEP HCLPF Calculation - Auxiliary Feedwater Controls, EJ-1051."

h. AREVA Calculation 32-9229831-002, "Palisades ESEP HCLPF Calculation - Primary Makeup Storage Tank, T-81."

i. AREVA Calculation 32-9230336-001, "Palisades ESEP Screening of Lube Oil Storage Room South Block Wall."

j. AREVA Calculation 32-9230249-003, Palisades ESEP HCLPF Calculation - Boric Acid Storage Tanks T-53A & T-53B."

k. AREVA Calculation 32-9228279-002, "Palisades ESEP HCLPF Calculation - Block Wall, C-104.11Q."

1. AREVA Calculation 32-9228841-001, "Palisades ESEP HCLPF Calculation - Block Walls C-107.160, C-107.17Q, and C-107.18Q."

Page 28

Palisades Nuclear Plant ESEP Report ATTACHMENT A - PALISADES NUCLEAR PLANT ESEL Page A-1

Palisades Nuclear Plant ESEP Report 3ervILe wdler 3y5Lemii 1 CV-0861 Return Isolation Valve Closed Open from CAC Fan VHX-1 Service Water System 2 CV-0864 Return Isolation Valve Closed Open from CAC Fan VHX-2 Service Water System Deenergize to open CV-0861 3 SV-0861 Return Isolation Valve Energized De-energized Breaker 52-1208 from CAC Fan VHX-1 Service Water System Deenergize to open CV-0864 4 SV-0864 Return Isolation Valve Energized De-energized Breaker 52-1209 from CAC Fan VHX-2 5 V-1A Containment Air Cooler Available Available Fan 6 V-2A Containment Air Cooler Available Available Fan Containment Air Cooler 7 V-3A Fan Available Available Fan 8 VHX-1 Containment Air Cooler Available Available 9 VHX-2 Containment Air Cooler Available Available 10 VHX-3 Containment Air Cooler Available Available 11 CV-0522B TDAFW pump steam Closed Open inlet valve 12 CV-0727 AFW flow control valve Closed Throttled (SG B) 13 CV-0749 AFW flow control valve Closed Throttled I___ (SG A)

Page A-2

Palisades Nuclear Plant ESEP Report

-- State. - . * .. .. teS/Com ments , - ,**-'M

.Nt~nb~

14 CV-0779 ADV (SG E-501) Closed Open as Automatically or manually opened [26]

needed CV-0780 AV(G-SBClsdOpenS as is AD

-03 lsdneeded Automatically or manually opened [26]

16 CV-0781 ADV (SG E-50A) Closed Open as Automatically or manually opened [26]

needed 17 CV-0782 ADV (SG E-50A) Closed Open as Automatically or manually opened [26]

17 C-72 AV(GE5A lsdneeded

___________ needed Automatically or manually opened [26]

18 CV-2008 T-81 Isolation Valve Closed Open [25]

19 CV-2010 T-81 Isolation Valve Closed Open [25]

20 E/P-0779 ADV (SG E-50B) On On [26]

21 E/P-0780 ADV (SG E-50B) On On [26]

22 E/P-0781 ADV (SG E-50A) On On [26]

23 E/P-0782 ADV (SG E-50A) On On [26]

24 HIC-0780A Hand Indicating Auto Auto Controller AutoAuto_[26]

25 HIC-0780B Hand Indicating Auto Auto [26]

Controller 26 HIC-0781B Hand Indicating Auto Auto [26]

Controller 27 P-8B TDAFW pump Standby Operating [27]

Page A-3

Palisades Nuclear Plant ESEP Report Eett"Lr-p*

l n 0erati c- Rf iSjtate.Ne..c.

28 OC-779 ADV(SGE-5B) Modulate as Modulate as Controls for CV-0781 shown on reference [6 28_PO_-0779_DV __SGE-50B necessary necessary Oth~ers typical of that arrangement [26]____

Modulate as Modulate as Controls for CV-0781 shown on reference

_________________ necessary necessary Others typical of that arrangement______

28 POC-0779 ADV (SG E-5013) Modulate as Modulate as 1[261 30 POC-0781 ADV (SG E-50A) ncsayeesry[26]

Modulate as Modulate.as Controls for CV-0781 shown on reference necessary necessary Others typical of that arrangement 3 SV02B TDAFW pump steam De-energized Energized [23]

32SV0522Binlet valve 33 POC-0780 ADV (SG E-5013) Energized 3nergized Energzlated Energized as Controls for CV-0781 shown on reference Others typical of that arrangement

[26]

[26]

30 SV-0779B ADV (SG E-50B) De-energized De-energized Controlstypical Others for CV-0781 of that shown arrangement on reference [6

[26]

31 POC-0782 ADV (SG E-50B) Mode-energeas neenery oDe tenerizd neenery Controls for CV-0781 shown on reference Others typical of that arrangement

[26]

[26]

36 SV-0779A ADV (SG E-5013) Energized Energized Controls for CV-0781 shown on reference Others typical of that arrangement

[26]

[26]

3 Controls for CV-0781 shown on reference [26]

7SV-0780B ADV (SG E-50B) De-energized De-energized Others typical of that arrangement 38 SV-0779C ADV (SG E-5013) De-energized De-energize'd Controls for CV-0781 Others typical of that shown on reference arrangement [26]

Energized to close. Same for all ADV SV "A" 39 SV-0781A ADV (SG E-50A) Energized Energized valves. When closed, ADVs can be operated [261 Cofrom controller.

Deenergized to open. Same for all ADV SV "B" 40 SV-0781B ADV (SG E-50A) De-energized De-energized valves. When open, ADVs can be operated [26]

from controller.

Page A-4

Palisades Nuclear Plant ESEP Report Eq wmn 'Ww rtn1~~t4 .. ... IN,..

~SEgliem z ORfrne--- -.

_____ rormalýStaife ýeedStae- -*

Deenergized to close. Same for all ADV SV "C" 41 SV-0781C ADV (SG E-SOA) De-energized De-energized valves. When closed, ADVs can be operated [26]

from controller.

42 SV-0782A ADV (SG E-50A) Energized Energized Controls for CV-0781 shown on reference [26]

42_SV-0782A DS 0e dd Others typical of that arrangement [26]

43 SV-0782B ADV (SG E-50A) De-energized De-energized Controls for CV-0781 shown on reference [26]

43Deenrze SV072D -5) De-energ Others typical of that arrangement [26]

44 SV-0782C ADV (SG E-50A) De-energized De-energized Controls for CV-0781 shown on reference [26]

Others typical of that arrangement 45 SV-2008 T-81 Isolation Valve De-energized Energized [25]

46 SV-2010 T-81 Isolation Valve De-energized Energized [25]

Primary System Make-up New FLEX valve. Required closed to prevent 47 SV-2235 Transfer Pump Isolation Energized De-energized inventory diversion.

Valve (new) Not yet installed.

Primary System Make-up New FLEX valve. Required closed to prevent 48 SV-2236 Transfer Pump Bypass Energized De-energized inventory diversion.

Isolation Valve (new) Not yet installed.

T-81 and CST combined volume > 100k gal Condensate Storage (Phase 1 - 8+ hrs); Long term water source 49 T-2 Tank Available Available (Phase 2) is Lake Michigan directly into steam [25]

generator via AFW or MFW header using a portable FLEX pump.

T-81 and CST combined volume > 100k gal (Phase 1 - 8+ hrs); Long term water source 50 T-81 Primary Makeup Storage Available Available (Phase 2) is Lake Michigan directly into steam [25]

Tank generator via AFW or MFW header using a portable FLEX pump. Screening HCLPF of 0.1 used for IPEEE.

Page A-5

Palisades Nuclear Plant ESEP Report U2 -6 Aftfl EEmltem A_ __ - - -

Ownom

le-ti ~Ddgire! State-51 CV-2083 CBO Isolation Valve Open

_____________________________ass~umption Closed Bleedoff must be isolated to support of PCP leakage of 1 gpm/pump

[8

[8 52 CV-2099 CBO Isolation Valve Open Closed Bleedoff must be isolated to support (8

__________________assumption of PCP leakage of 1 gpm/pump [8 53 CBORelif C-219 Stp Vave Oen Cosed Bleedoff must be isolated to support [8 53 CBORelif C-219 Stp Vave Oen Cosed assumption of PCP leakage of 1 gpm/pump [8 5E-7 Condensate Storage NA/A Listed per EPRI Q&A 3.20 to ensure heat [5 54 E-27 Tank Heat Exchanger NA/A exchanger pressure boundary integrity [5 55E-56 Regenerative Heat N/A N/A Listed per EPRI Q&A 3.20 to ensure heat [20]

Exchanger exchanger pressure boundary integrity 56 MO-2087 VCT Supply to Charging Open Closed- VCT volume is not credited and must be [19]

Pumps isolated in Phase 2 HPSI Injection Loop 57 MO-3007 Isolation Valve Closed Open [22]

58 MO-3041 SIT T-82A Isolation Valve Open Closed Closure to prevent N2 injection; associated [21]

____________relay in A-46 program 59SITT-82 M-304 Isoatio Vale Opn Clsed Closure to prevent N2 injection; associated [1 M-304 Isoatio Vale 59SITT-82 Opn Clsed relay in A-46 program [1 60 MO-3049 SIT T-82C Isolation Valve Open Closed. Closure to prevent N2 injection; associated [1 relay in A-46 program [1 61 MO-3052 SIT T-82D Isolation Valve Open Closed Closure to prevent relay in A-46 N2 injection; associated program [21]

62 P-55C Charging Pump P-55C Off On Deviation taken from NEI 12-06 Sect 3.2.2 (12); [20]

(Phase 2) 63 T-53A Boric Acid Storage Tank Available Available' [19]

T-53A 64 T-53B Boric Acid Storage Tank Available Available [19]

T-53B Page A-6

Palisades Nuclear Plant ESEP Report

- ntjnnnts-65 EA-11 2400V Bus 1C Available Available [7]

66 EB-01 480 V MCC 1 Available Available [7]

67 EB-11 480V Bus LCC-11 Available Available Bus supplies charging pump P55C power [7]

68 EB-19 480V Bus LCC-19 Available Available Powers MCC-1 which feeds BC 1 & 4, LCC-11, & [8]

emergency lighting panel L-04A 69 EB-21 480 V MCC 21 Available Available- Support SIT Isolation [33]

70 EB-23 480 V MCC 23 Available Available Support SIT Isolation [33]

71 EC-01 Electrical Cabinet Available Available Control room panel 72 EC-02 Electrical Cabinet Available Available Control room panel 73 EC-11 Electrical Cabinet Available Available Control room panel 74 EC-12 Electrical Cabinet Available Available Control room panel 75 EC-13 Electrical Cabinet Available Available Control room panel 76 EC-182 ADV Panel Available Available 77 EC-33 Electrical Cabinet Available Available Not required - was for valves removed from ESEL 78 ED-01 125VDC Battery #1 Available Avail-able'[9 Page A-7

Palisades Nuclear Plant ESEP Report

-C ;j faZ aatiti A",

74a7e rNWwNo,--Wtes/Comm, ent Ref

- fI W- at 79 ED-02 125VDC Battery #2 Available Available [9]

s0 ED-06 Inverter 1 Available Available [10]

81 ED-07 Inverter 2 Available Available [10]

82 ED-08 Inverter 3 Available Available [10]

83 ED-09 Inverter 4 Available Available [10]

84 ED-10OL DC Bus DIOL Available Available [9]

85 ED-10R DC Bus D1i11 Available Available [9]

86 ED-11 125VDC Panel ED-11 (-1, Available Available [0 87 ED-13 DC Bus Metering Section Available Available' [9]

88 ED-15 -Battery Charger #1 Available Available [9]

89 ED-18 Battery Charger #4 Available Available [9]

90 ED-20L DC Bus D20L Available Available [9]

91 ED-20R DC Bus D20R Available Available [9]

92 ED-21-1 125VDC Panel ED-21-1 Available Avail able [10]

Page A-8

Palisades Nuclear Plant ESEP Report

- , y ra -.....

Ing . .

Nb.i ( . Piifi

• c* * * ~ t**a~~l*-*e e . .ta.e.. .... ......o . .. *. ... : .o.*..... -hen.ts.V*.**

• bNme KMal~ M6-ps .State, 93 ED-21-2 125VDC Panel ED-21 -2 Available Available, [10]

94 ED-23 DC Bus Metering Section Available Available [9]

95 EJ-1051 Junction Box Available Available 96 EJ-569 Junction Box Available Available [30]

480V FLEX generator will connect to this new 97 EP-1901 FLEX Electrical Box Off On panel (FLEX Modification). Anchorage [51]

calculation EA-EC46465-05. Not yet installed.

98 EX-11 2400/480 V Transformer Available Available [71 99 EX-19 2400/480 V Transformer Available Available [8]

100 EY-10 Preferred 120 VAC Panel Available Available [101 1 (Y-10) 101 EY-20 Preferred 120 VAC Panel Available Available [10]

2 (Y-20) 102 EY-30 Preferred 120 VAC Panel Available Available [10]

3 (Y-30) 103 EY-40 Preferred 120 VAC Panel Available Available [10]

4 (Y-40) 104 N2 #

Backup Support AFW Flow Control Valves CV-0727, [31]

N2 Backup N2 Backup Station #1 Available Available 0749. Listed per EPRI O&A 3.22.

105 N2 #

Backup Support AFW TDAFW pump steam supply valve [31]

N2 Backup N2 Backup Station #2 Available Available CV-0522B. Listed per EPRI Q&A 3.22.

106 N2 Backup N2 Backup Station #9 Available Available Support ADVs. Listed per EPRI Q&A 3.22. FLEX [31]

Station #9 equipment not yet installed.

Page A-9

Palisades Nuclear Plant ESEP Report ESLitm-- 4 g~era~ting; tagge AFW (E-50A) Flow 107 FIC-0727 - --Indicator A-Monitor-is Controller On On [27]

108 FM-0727 AFM(-onitFow On On .. Square root extractor. [27]

109 FT-0727 AFW (E-50B) Flow On On [27]

Transmitter AFW (E-50B) Hand 110 HIC-0727 Indicator Controller Auto Auto [27]

ill I/P-0727 Signal Converter - AFW Modulate as Modulateas [27]

flow control valve (SG B) necessary necessary 112 P/S-0727 Power Supply - AFW (E- On On [37]

50B) Flow Transmitter AFW (E-50A) Flow 113 FIC-0749 Indicator Controller On On [27]

114 FM-0749 AFW (E-MoA) Flow On On Square root extractor. [27]

Monitor 115 FT-0749 AFW (E-50A) Flow On On [27]

Transmitter 116 HIC-0749 AFW (E-50A) Hand [27]

Indicator Controller Auto Auto 117 I/P-0749 Signal converter - AFW Modulate as Modulateas [27]

flow control valve (SG A) necessary necessary 118 P/S-0727A Power Supply - AFW (E- On On [37]

50A) Flow Transmitter 119 P/S-0206 Power Supply - BAST T- On On Not available until Phase 2 (powered by FLEX [14]

53B Level Indication generator) 120 LIA-0206 Boric Acid Storage Tank On 0n Not available until Phase 2 (powered by FLEX [19]

T-53B Level Indication generator)

Page A-10

Palisades Nuclear Plant ESEP Report Boric Acid Storage Tank Not available until Phase 2 121 LIT-0206 On T-53B Level Transmitter generator)

Boric Acid Storage Tank Not available until Phase 2 T-53A Level Indication generator)

LIT-0208 Boric Acid Storage Tank available until Phase 2 123 T-53A Level Transmitter erator) 124 P/S-0208 Power Supply - BAST T-53A Level Indication 125 LIA-0102A PZR Level (WR) Indicator 126 LT-0102 PZR Level (WR)

Transmitter 127 LIA-0365 SITIdcto T-82A Level Monitor to prevent N2 injection Indication 128 LM-0365 SIT T-82A Level Monitor Square root extractor 129 L-0365 SIT T-82A Level Transmitter jection 130 P/S-0365 Power Supply - SIT T-82A Level Indication jection 131 LIA-0368 SITIdcto T-82B Level Indication jection 132 LM-0368 SIT T-82B Level Monitor iuare root extractor 133 LT-0368Trnmte SIT T-82B Level jection Transmitter 134 P/S-0368 Power Supply - SIT T-82B 4 8 Level Indication Monitor to prevent N2 injection Page A-11

Palisades Nuclear Plant ESEP Report

?P _ .- J, .- et" E*E-*:tem ,

OU -

M"V- RLGM Screened Note 2 inlet valve AFW flow control valve 12 CV-0727 >RLGM Screened (SG B) 13 CV-0749 AFW flow control valve >RLGM Screened (SG A) 14 CV-0779 ADV (SG E-50B) >RLGM Screened 15 CV-0780 ADV (SG E-50B) >RLGM Screened 16 CV-0781 ADV (SG E-50A) >RLGM Screened 17 CV-0782 ADV (SG E-50A) >RLGM Screened 18 CV-2008 T-81 Isolation Valve >RLGM Screened 19 CV-2010 T-81 Isolation Valve >RLGM Screened Page B-2

Palisades Nuclear Plant ESEP Report HCLP'F()/

Equlim-' .i Equipment'escription creeig deients

______ ______ Leve'l'C m et 20 E/P-0779 ADV (SG E-50B) >RLGM Screened 21 E/P-0780 ADV (SG E-50B) >RLGM Screened 22 E/P-0781 ADV (SG E-50A) >RLGM Screened 23 E/P-0782 ADV (SG E-50A) >RLGM Screened 24 HIC-0780A Hand Indicating >RLGM Screened Controller 25 HIC-0780B Hand Indicating >RLGM Screened Controller 26 HIC-0781B Hand Indicating >RLGM Screened Controller 27 P-8B TDAFW pump >RLGM Screened Note 1 28 POC-0779 ADV (SG E-50B) >RLGM Screened 29 POC-0780 ADV (SG E-50B) >RLGM Screened 30 POC-0781 ADV (SG E-50A) >RLGM Screened 31 POC-0782 ADV (SG E-50A) >RLGM Screened 32 SV-0522B TDAFW pump steam >RLGM Screened inlet valve 33 SV-0779A ADV (SG E-50B) >RLGM Screened 34 SV-0779B ADV (SG E-50B) >RLGM Screened 35 SV-0779C ADV (SG E-50B) >RLGM Screened 36 SV-0780A ADV (SG E-50B) >RLGM Screened 37 SV-0780B ADV (SG E-50B) >RLGM Screened 38 SV-0780C ADV (SG E-50B) >RLGM Screened 39 SV-0781A ADV (SG E-50A) >RLGM Screened 40 SV-0781B ADV (SG E-50A) >RLGM Screened Page B-3

Palisades Nuclear Plant ESEP Report Item Equipmn ID Equipment t Description .Screening. FW UW:Failre, Comme.

41 SV-0781C ADV (SG E-50A) >RLGM Screened 42 SV-0782A ADV (SG E-50A) >RLGM Screened 43 SV-0782B ADV (SG E-50A) >RLGM Screened 44 SV-0782C ADV (SG E-50A) >RLGM Screened 45 SV-2008 T-81 Isolation Valve >RLGM Screened 46 SV-2010 T-81 Isolation Valve >RLGM Screened Primary System Make- Not Not New FLEX component 47 SV-2235 up Transfer Pump to be seismically Isolation Valve (new) Applicable Applicable designed.

Primary System Make-48 SV-2236 48 SV23 up Transfer Pump *Not Not to be seismically Bypass Isolation Valve Applicable Applicable designed.

(new (new)__ __

49 T-2 Condensate Storage >RLGM Screened Note 1 Tank 50 T-81 Primary Makeup 0.19 Tank Shell Modification Storage Tank Buckling Required 51 CV-2083 CBO Isolation Valve TBD TBD Note 3 52 CV-2099 CBO Isolation Valve TBD TBD Note 3 53 CV-2191 CBO Relief Stop Valve TBD TBD Note 3 54 E-27 Condensate Storage >RLGM Screened Note 2 Tank Heat Exchanger 55 E-56 Regenerative Heat TBD TBD Note 3 Exchanger 56 MO-2087 VCT Supply to Charging >RLGM Screened Pumps 57 MO-3007 HPSI Injection Loop TBD TBD Note 3 Isolation Valve 58 MO-3041 SIT T-82A Isolation TBD TBD Note 3 Valve 59 MO-3045 SIT T-82B Isolation TBD TBD Note 3 Valve Page B-4

Palisades Nuclear Plant ESEP Report Equlpmen. r Equ nDesc' plon-: Screening Commnts No.~~ LevelIod 60 MO-3049 SIT T-82C Isolation TBD TBD Note 3 Valve 61 MO-3052 SIT T-82D Isolation TBD TBD Note 3 Valve 62 P-55C Charging Pump P-55C 0.47 Anchorage Boric Acid Storage Tank HCLPF calculated with 63 T-53A T-53A 0.46 Anchorage modifications to tank supports.

Boric Acid Storage Tank HCLPF calculated with 64 T-53B T-53B 0.46 Anchorage modifications to tank supports.

65 EA-11 2400V Bus 1C >RLGM Screened Note 1 66 EB-01 480 V MCC 1 0.41 Functional 67 EB-11 480V;Bus LCC-11 0.44 Anchorage 68 EB-19 480V Bus LCC-19 >RLGM Screened Note 2 69 EB-21 480 V MCC 21 >RLGM Screened Note 1 70 EB-23 480 V MCC 23 0.40 Block Wall Note 1 71 EC-01 Electrical Cabinet >RLGM Screened Note 2 72 EC-02 Electrical Cabinet 0.50 Anchorage 73 EC-11 Electrical Cabinet >RLGM Screened Note 1 74 EC-12 Electrical Cabinet >RLGM Screened Note 1 75 EC-13 Electrical Cabinet >RLGM Screened Note 1 76 EC-182 ADV Panel >RLGM Screened Note 2 77 EC-33 Electrical Cabinet 0.42 Block Wall Note 1 78 ED-01 125VDC Battery #1 0.40 Block Wall 79 ED-02 125VDC Battery #2 0.40 Block Wall Page B-5

Palisades Nuclear Plant ESEP Report N. Equirpment 'ID' 01 dt ecrlpt on tne jee FalueComet 80 ED-06 Inverter 1 >RLGM Screened Note 1 81 ED-07 Inverter 2 >RLGM Screened Note 1 82 ED-08 Inverter 3 >RLGM Screened Note 1 83 ED-09 Inverter 4 >RLGM Screened Note 1 84 ED-10L DC Bus DIOL 0.40 Block Wall Note 1 85 ED-10R DC Bus D1OR 0.40 Block Wall Note 1 86 ED-11 125VDC Panel ED-11 >RLGM Screened 1, -2) 87 ED-13 DC Bus Metering >RLGM Screened Section 88 ED-15 Battery Charger #1 >RLGM Screened Note 1 89 ED-18 Battery Charger #4 >RLGM Screened Note 1 90 ED-20L DC Bus D20L 0.40 Block Wall Note 1 91 ED-20R. DC Bus D20R 0.40 Block Wall Note 1 92 ED-21-1 125VDC Panel ED-21-1 >RLGM Screened 93 ED-21-2 125VDC Panel ED-21 -2 >RLGM Screened 94 ED-23 DC Bus Metering >RLGM Screened Section 95 EJ-1051 Junction Box 0.46 Anchorage 96 EJ-569 Junction Box >RLGM Screened Note 2 97 EP-1901 FLEX Electrical Box >RLGM Screened Note 4 98 EX-11 2400/480V >RLGM Screened Note 1 Transformer 2400/480 V 99 EX-19 Trasfome Transformer >RLGM Screened Note 2 100 EY-10 Preferred 120 VAC Panel (Y-l0) I

>RLGM I

Screened II Page B-6

Palisades Nuclear Plant ESEP Report Itemi. .. "n* * " .. . .:" ' .. Faiiureý-" . */ , , .:- , ,

Item. Equipment: I ED-IE1ent" D riMode Comments

.No. ...- "* .. :' "" ' "  ; : ... Leb ,el o__e...._ , ___ __..__ ___ __"

101 EY-20 Preferred 120 VAC >RLGM Screened Panel 2 (Y-20) 102 EY-30 Preferred 120 VAC >RLGM Screened Panel 3 (Y-30) 103 EY-40 Preferred 120 VAC >RLGM Screened Panel 4 (Y-40) 104 N2 Backup N2 Backup Station #1 >RLGM Screened Note 2 Station #1 105 N2 Backup N2 Backup Station #2 >RLGM Block Wall Note 5 Station #2 N2 Backup Not Not New FLEX component 106 u N2 Backup Station #9 to be seismically Station #9 Applicable Applicabledein. designed.

AFW (E-50A) Flow 107 .FIC-0727 ato indicator Control "Controller >RLGM Screened Note 2 108 FM-0727 AFW (E-50B) Flow >RLGM Screened Monitor I 109 FT-0727 AFW (E-50B) Flow >RLGM Screened Transmitter AFW (E-50B) Hand 110 HIC-0727 Indicator Controller >RLGM Screened Signal Converter - AFW 111 I/P-0727 flow control valve (SG >RLGM Screened Note 2 B) 112 P/S-0727 Power Supply - AFW (E- >RLGM Screened 50B) Flow Transmitter 113 FIC-0749 AFW (E-50A) Flow >RLGM Screened Indicator Controller 114 FM-0749 AFW (E-50A) Flow >RLGM Screened Monitor 115 FT-0749 AFW (E-50A) Flow >RLGM Screened Transmitter AFW (E-50A) Hand 116 HIC-0749 ato Indicator Cont Controller >RLGM Screened Signal converter - AFW 117 I/P-0749 flow control valve (SG >RLGM Screened Note 2 A) 118 P/S-0727A Power Supply - AFW (E- >RLGM Screened 50A) Flow Transmitter 119 P/S-0206 Power Supply - BAST T- >RLGM Screened 53B Level Indication Page B-7

Palisades Nuclear Plant ESEP Report Iteqi e ..  : .... (g . .Failure

'r' --scri Epment u"" 'D pt on, reeing M 120 LIA-0206 Boric Acid Storage Tank >RLGM Screened T-53B Level Indication 121 LIT-0206 Boric Acid Storage Tank >RLGM Screened IT T-53B Level Transmitter 122 LIA-0208 Boric Acid Storage Tank >RLGM Screened T-53A Level Indication 123 LIT-0208 Boric Acid Storage Tank >RLGM Screened Note 2 T-53A Level Transmitter 124 P/S-0208 Power Supply - BAST T- >RLGM Screened 53A Level Indication 125 LIA-0102A PZR Level (WR) >RLGM Screened Indicator 126 LT-0102 PZR Level (WR) TBD TBD Note 3 Transmitter 127 LIA-0365 SIT T-82A Le6el >RLGM Screened Indication 128 LM-0365 SIT T-82A Level Monitor >RLGM Screened 129 LT-0365 SIT T-82A Level TBD TBD Note 3 Transmitter 130 P/S-0365 Power Supply - SIT T- >RLGM Screened 82A Level Indication 131 LIA-0368 SIT T-82B Level >RLGM Screened Indication 132 LM-0368 SIT T-82B Level Monitor >RLGM Screened 133 LT-0368 SIT T-82B Level TBD TBD Note 3 Transmitter 134 P/S-0368 Power Supply - SIT T- >RLGM Screened 82B Level Indication 135 LIA-0372 SIT T-82C Leve >RLGM Screened Indication 136 LM-0372 SIT T-82C Level Monitor >RLGM Screened 137 LT-0372 SITT-82C Level >RLGM Screened Transmitter 138 P/S-0372 Power Supply - SIT T- >RLGM Screened 82C Level Indication 139 LIA-0374 SIT T-82D Leve >RLGM Screened Indication 140 LM-0374 SIT T-82D Level Monitor >RLGM Screened Page B-8

Palisades Nuclear Plant ESEP Report Equipment ID Item~. 'Eqic ~ Failure.

p opment S.rning Descript Mde . Cmments No. Meoel 141 LT-0374 SIT T-82D Leve TBD TBD Note 3 Transmitter 142 P/S-0374 Power Supply - SITT- >RLGM Screened 82D Level Indication 143 P/S-0751A Power Supply - SG A NR >RLGM Screened Level Indication 144 P/S-0751C Power Supply - SG A NR >RLGM Screened Level Indication 145 LI-0757A SG A WR Level >RLGM Screened Indication SG A WR Level 146 LT-0757A TBD TBD Note 3 Transmitter 147 P/S-0757A Power Supply - SG A WR >RLGM Screened Level Transmitter 148 LI-0758A SG B WRLevel >RLGM Screened Indication "SG B WR.Level 149 LT-0758A

______Transmitter SGB Wmiee TBD TBD Note 3 150 P/S-0758A Power Supply - SG B >RLGM Screened WR Level Transmitter 151 LIA-2021 CST (T-2) Level >RLGM Screened Indication 152 LT-2021 CST (T-2) Level >RLGM Screened Transmitter 153 P/S-2021 Power Supply - CST (T-2) >RLGM Screened I_ Level Indication 154 PI-0105A PZR Pressure (WR) >RLGM Screened Indicator 155 PT-0105A PZR Pressure (WR) TBD TBD Note 3 Transmitter Wide Range Primary 156 PTR-0112 Temperature & >RLGM Screened Pressure Recorder 157 PIC-0751C SG A Pressure Indicator >RLGM Screened Controller 158 PT-0751C SG A Pressure TBD TBD Note 3 Transmitter 159 PIC-0752C SG B Pressure Indicator >RLGM Screened Controller SG B Pressure 160 PT-0752C TBD TBD Note 3 Transmitter I I I Page B-9

Palisades Nuclear Plant ESEP Report tem., P-E(g)i Failure' qupment 9teýi eo ýID Equpmn Decitin S . r SrnnCmet MOde.

Containment Sump 161 LPIR-0383 Level/Pressure >RLGM Screened Indication Power Supply -

162 P/S-1812A Containment Pressure >RLGM Screened Transmitter 163 PT-1812A Containment Pressure >RLGM Screened Note 2 Transmitter 164 TE-0112CC PCS T-cold Temperature TBD TBD Note 3 Element 165 T1-0112CC PCS T-cold Temperature >RLGM Screened Transmitter 166 TE-0112HC PCS T-hot Temperature TBD TBD Note 3 Element 167 T1"-Ol12H PCS T-hot Temperature >RLGM Screened TT-O112HC 167 Transmitter _____crene 168 LE-OIOA Reactor Vessel Level TBD TBD Note 3 Reactor Vessel Level 169 LRI-O101A1 Inidicator Recorder - >RLGM Screened Head Region Reactor Vessel Level 170 LRI-01OIA2 Inidicator Recorder - >RLGM Screened Upper Guide Structure 171 EC-l1A Post Accident Panel >RLGM Screened Note 2 172 P/S-0101AA Power Supply for LE- >RLGM Screened 0101A 173 T-82A Safety Injection Tank A TBD TBD Note 3 174 T-82B Safety Injection Tank B TBD TBD Note 3 175 T-82C Safety Injection Tank C TBD TBD Note 3 176 T-82D Safety Injection Tank D TBD TBD Note 3 177 CV-0599 TDAFW Trip/Throttle >RLGM Screened Valve 178 CV-0598 TDAFW Governor Valve >RLGM Screened Page B-10

Palisades Nuclear Plant ESEP Report Notes:

1. Anchorage screened out based on available margin during walkdown by SRT.

2. Anchorage screened out during walkdown validation by SRT.

3. Inaccessible. Per EPRI NP-6041-SLR1, Sec. 2, Seismic Capability Walkdown, Step 5 - This component was not walked down.

4. The design of anchorage for component EP-1901 is performed for seismic load of 1xSSE in Entergy Calculation EA-EC46465-05 [51]. The maximum interaction ratio for anchor bolts is 0.23. This ratio is less than 0.5, thus margin exists such that the anchorage of panel EP-1901 is adequate for 2xSSE.

5. Lube Oil Storage Room South Block Wall is reinforced for FLEX [52].

Page B-11

Entergy Nuclear Operations, Inc.

Eiy VR0WEntergy Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043 Tel 269 764 2000 Anthony J. Vitale Site Vice President PNP 2014-108 December 18, 2014 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, MD 20852

SUBJECT:

Palisades Nuclear Plant Expedited Seismic Evaluation Process Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident Palisades Nuclear Plant Docket 50-255 License No. DPR-20

REFERENCES:

1. NRC letter, Request for Information Pursuantto Title 10 of the Code of FederalRegulations 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-ichiAccident, dated March 12, 2012 (Adams Accession No. ML12056A046).

2. NEI letter, ProposedPath Forwardfor NTTF Recommendation 2. 1:

Seismic Reevaluations, dated April 9, 2013 (ADAMS Accession No. ML13101A379).

3. NRC letter, Electric Power Research Institute Report XXXXXX, "SeismicEvaluation 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,dated May 7, 2013 (ADAMS Accession No. ML13106A331).

4. Electric Power Research Institute Report 3002000704, Seismic Evaluation Guidance:Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2. 1: Seismic, dated May 2013.

PNP 2014-108 Page 2

Dear Sir or Madam:

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

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

Reference 1 requested that licensees provide interim evaluations and actions taken or planned to address the higher seismic hazard relative to the design basis, as appropriate, prior to completion of the risk evaluation. In accordance with the NRC endorsed guidance in Reference 3, the enclosed Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant provides the information described in Section 7 of Reference 4 in accordance with the schedule identified in Reference 2.

This letter contains five new commitments, identified in the attachment.

I declare under penalty of perjury that the foregoing is true and correct. Executed on December 18, 2014.

Sincerely,

Attachment:

List of Regulatory Commitments

Enclosure:

Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant cc: Director of Office of Nuclear Regulation Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC

Attachment to PNP 2014-108 List of Regulatory Commitments

List of Regulatory Commitments The following table identifies those actions committed to by Entergy Nuclear Operations, Inc.

(ENO) in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

TYPE (Check One) SCHEDULED COMPLETION COMMITMENT ONE- DATE TIME CONTINUING (If Required)

ACTION COMPLIANCE

1. ENO will perform seismic walkdowns, X No later than the generate High Confidence of a Low end of the second Probability of Failure calculations, and planned refueling design and implement any necessary outage after modifications for inaccessible items December 31,2014.

listed in Section 7.1 of the Expedited Seismic Evaluation Process (ESEP)

Report for Palisades Nuclear Plant.

2. Modify Boric Acid Storage Tank X No later than the (T-53A) tank anchorage such end of the second that the High Confidence of a Low planned refueling Probability of Failure (HCLPF) capacity outage after is greater than the demand December 31,2014.

characterized by the Review Level Ground Motion (RLGM) Peak Ground Acceleration.

3. Modify Boric Acid Storage Tank X No later than the (T-53B) tank anchorage such end of the second that the HCLPF capacity is greater than planned refueling the demand characterized by the outage after RLGM Peak Ground Acceleration. December 31, 2014.

4. Modify Primary Makeup Storage Tank X No later than the (T-81) such that the HCLPF capacity is end of the second greater than the demand characterized planned refueling by the RLGM Peak Ground outage after Acceleration. December 31, 2014.

5. Submit a letter to NRC X Within 60 days summarizing the HCLPF calculation following results of commitments 1, 2, 3, and 4, completion of and confirming implementation of the Expedited Seismic plant modifications associated with Evaluation Process commitments 1,2, 3, and 4. activities, including commitments 1, 2, 3, and 4.

Enclosure to PNP 2014-108 Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant 54 Pages Follow

20004-021 (01/30/2014)

A AREVA AREVA Inc.

Engineering Information Record Document No.: 51 - 9231248 - 002 Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant Page 1 of 65

A 20004-021 (0113012014)

Document No.: -0923`1248-002 AREVA Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant Safety Related? YES I NO Does this document establish design or technical requirements? " YES NO Does this document contain assumptions requiring verification? [] YES NO M

Does this document contain Customer Required Format? YES. []NO Signature Block PageuiSectlone Name and PILlP, RILR, Prepared/Revlewedi Thtle/Dlselpllne Signature A-CRF, A Date Approved or Comments*

Ogden Sawyer ,j LPAl Engineering Supervisor I/../I /9 AlaaHorn* P Appendix A (Section 3.0 and P)A Engineer: /20/61//f Attachment A)

ProjectEnginecrll ( 1 2r Grant Tinsley R 12-10-14 Appendix A (Sections 3.0, and

.AB c Attaoliment A) magic Stewart R Appendix A (Sections 6.0,7.0, 8.0 and Engineer IV AttaclunentEB)

Keowl Connell , A Al Engineering Manager A JenniferButler *A-CRF / Appendix A Note: /LP designates Preparer (P), Lead Preparer (LP)

R/LR designates Reviewer (R), Lead Reviewer (LR)

A-CRF delignates Project Manager Approver of Customer Required Format (A-CRP)

A designates Approver/RTM- Verification of Reviewer Independence Page 2

A 20004-021 (01/30/2014)

Document No.: 51-9231248-002 AREVA Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant Signature Block (continued)

Project Manager Approval of Customer References (N/A if not applicable)

Name Title (printed or typed) (printed or typed) Signature Date Jennifer Butler Project Manager Page 3

A 20004-021 (01/30/2014)

Document No.: 51-9231248-002 AREVA Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant Record of Revision Revision Pages/Sections/

No. Paragraphs Changed Brief Description I Change Authorization 000 All Initial release 001 Section 2.0 Section 2.0 0 Updated References 1, 3, 4, 40, 47, 49 and 50 Appendix A - 2.0, 3.0, 3.1, 0 Added References 5, 37, 38, 51, 52 and 57 3.1.1, 3.1.3, 3.1.4, 3.1.6, 3.2, Appendix A 4.2,5.1,5. 2,6.1,6.2,6.3.1, Sections 2.0, 3.0, 3.1, 3.1.1, 3.1.3, 3.1.4, 3.1.6, 3.2, 4.2, 5.1, 5.2, 6.3.3, 6.4, 6.5, 6.6, 7.1, 7.2, 6.1, 6.2, 6.3.1, 6.3.3, 6.4, 6.5, 6.6, 7.1, 7.2, 8.1, 8.2, 8.3, 8.4 and 9.0 8.1, 8.2, 8.3, 8.4 and 9.0, were modified to incorporate Entergy comments [57] on Revision Attachment000 of the document.

Attachment B.

• Attachment A - modified to incorporate Entergy comments [57]

and to update with additional components based on new revision of supporting document.

• Attachment B - modified to incorporate Entergy comments, update HCLPF capacities based on new revisions of supporting calculations, and to uOpdate with additional componentg, based on new revision of supporting document.

002 Section 2.0 Section 2.0 0 Added Reference 58 Appendix A - 3.1.1, 6.5, 7.2, Appendix A and 8.4, Attachment A and 0 Sections 3.1.1 6.5, 7.2, and 8.4 were modified to incorporate Attachment B Entergy comments [58] on Revision 001 of the document

• Attachment A - modified to incorporate Entergy Comments [58].

" Attachment B - modified to incorporate Entergy Comments [58].

Page 4

A Document No.: 51-9231248-002 AREVA Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant Table of Contents Page SIGNATURE BLOCK ............................................................................................................................. 2 RECORD OF REVISION ....................................................................................................................... 4 1.0 DOCUMENTATION .................................................................................................................... 6

2.0 REFERENCES

........................................................................................................................... 6 APPENDIX A: EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT FOR PALISADES NUCLEAR PLANT ........................................................................ A-1 Page 5

A AR EVA Document No.: 51-9231248-002 Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant 1.0 DOCUMENTATION Appendix A to this document contains the Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant, and is presented in the customer-required format.

2.0 REFERENCES

References identified with an (*) are maintained within Palisades Nuclear Plant Records System and are not retrievable from AREVA Records Management. These are acceptable references per AREVA Administrative Procedure 0402-01, Attachment 8. See page 2 for Project Manager Approval of customer references.

1. NRC (E Leeds and M Johnson) Letter to All Power Reactor Licensees et al., "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident," March 12, 2012, NRC ADAMS Accession No. ML12053A340.

2. EPRI 30020007.04, "Seismic Evaluation Guidance, Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," May 2013.

3. Entergy Letter to U.S. NRC, letter number PNP 2013-010, "Overall Integrated Plan in Response to March 12, 2012 Commission Order to ModifyLicenses With Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049),"

February 28, 2013, NRC ADAMS Accession No. ML13246A399.

4. Entergy Letter to U.S. NRC, Letter Number PNP 2014-011, "Palisades Nuclear Plant 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 28, 2014, NRC ADAMS Accession No. ML14059A078.

5. Entergy Letter to U.S. NRC, Letter Number PNP 2014-085, "Palisades Nuclear Plant 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 28, 2014, NRC ADAMS Accession No. ML14240A278.

6. *Entergy Document, Engineering Change EC-46465, "FLEX Basis."

7. *Entergy Drawing E0001, Sheet 1, Revision 83, "Single Line Meter & Relay Diagram 480V Motor Control Center Warehouse, WD 950."

8. *Entergy Drawing E0001, Sheet 3, Revision 4, "Plant Single Line Diagram, WD 950."

9. *Entergy Drawing E0008, Sheet 1, Revision 57, "Single Line Meter & Relay Diagram 125V DC 120V Instrument & Preferred AC, System WD 950."

10. *Entergy Drawing E0008, Sheet 2, Revision 55, Single Line Meter & Relay Diagram 125V DC 120V Instrument & Preferred AC, System WD 950."

11. *Entergy Drawing E0078, Sheet 2, Revision 18, "WRSGL Schematic Diagram."

12. *Entergy Drawing E0078, Sheet 2A, Revision 6, "WRSGL Schematic Diagram."

13. *Entergy Drawing E0082, Sheet 5, Revision 13, "Schematic Diagram Wide Range Pressurizer Level Indicator/Alarm Instrumentation."

14. *Entergy Drawing E0090, Sheet 5, Revision 14, "Schematic Diagram Boric Acid System Instrumentation."

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15. *Entergy Drawing E0092, Revision 14, "Schematic Diagram Safety Injection Tank Level Indicator Alarm Instrumentation."

16. *Entergy Drawing M0201, Sheet 1, Revision 87, "Piping & Instrument Diagram, Primary Coolant System."

17. *Entergy Drawing M0201, Sheet 2, Revision 66, "Piping & Instrument Diagram, Primary Coolant System."

18. *Entergy Drawing M0202, Sheet 1, Revision 76, "Piping & Instrument Diagram, Chemical &

Volume Control System."

19. *Entergy Drawing M0202, Sheet 1A, Revision 64, "Piping & Instrument Diagram, Chemical &

Volume Control System."

20. *Entergy Drawing M0202, Sheet 1B, Revision 59, "Piping & Instrument Diagram, Chemical &

Volume Control System."

21. *Entergy Drawing M0203, Sheet 1, Revision 48, "Piping & Instrument Diagram, Safety Injection, Containment Spray, Shutdown Cooling-System."

22. *Entergy Drawing M0203, Sheet 2, Revision 27, "Piping & Instrument Diagram, Safety Injection, Containment Spray, Shutdown Cooling System."

23. *Entergy Drawing M0205, Sheet 2, Revision69, "Piping & Instrument Diagram, Main Steam and Auxiliary Turbine Systems."

24. *Entergy Drawing M0208, Sheet 1 B, Revision 37, "Piping & Instrument Diagram, Service Water System."

25. *Entergy Drawing M0220, Sheet 1, Revision 97, "Piping & Instrument Diagram, Make-up Domestic Water & Chemical Injection Systems."

26. *Entergy Drawing M0207, Sheet 1, Revision 91, "Piping & Instrument Diagram, Feedwater &

Condensate System."

27. *Entergy Drawing M0207, Sheet 2, Revision 38, "Piping & Instrument Diagram, Auxiliary Feedwater System."

28. *Entergy Procedure Palisades EOP-3.0, Revision 16, "Station Blackout."

29. *Entergy Drawing E0364, Revision 9, "Conduit and Tray Miscellaneous Plans."

30. *Entergy Drawing E0618, Sheet 569, Revision 12, "Connection Diagram Junction Box J569."

31. *Entergy Drawing M0222, Sheet 2, Revision 29 (EC-47346), "Piping & Instrument Diagram, Miscellaneous Gas Supply Systems."

32. *Entergy Drawing M0218, Sheet 2, Revision 61, "Piping & Instrument Diagram, Htg. Vent. & Air Cond. Containment Building."

33. *Entergy Drawing E0005, Sheet 5B, Revision 12, "Single Line Meter & Relay Diagram 480 Volt Motor Control Centers, System WD 950."

34. *Entergy Drawing E0099 Sh. 5, Revision 10, "Schematic Diagram, Containment Building Instrumentation (Left Channel)."

35. *Entergy Drawing E0084 Sh. 6, Revision 14, "Schematic Diagram, Pressurizer Pressure Control and Measurement Channel Instrumentation."

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A AR EVA Document No.: 51-9231248-002 Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant

36. *Entergy Drawing E0087, Sh. 6, Revision 10, "Schematic Diagram Level Indication and Alarm Indication."

37. *Entergy Drawing E0076, Sheet 4C, Revision 8, "Schematic Diagram Feedwater and Turbine Driver Instrumentation."

38. *Entergy Drawing E0226, Sheet 1B, Revision 6, "Schematic Diagram - Reactor Vessel Level Monitoring System (Left Channel)."

39. *"Palisades Nuclear Plant - Final Safety Analysis Report," Revision 31, Docket 50-255, September 2014.

40. Entergy Letter to U.S. NRC, letter number PNP 2014-033, "Palisades Nuclear Plant Seismic Hazard and Screening Report (CEUS Sites), Response to NRC 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 March 31, 2014, NRC ADAMS Accession No. ML14090A069.

41. *Entergy Technical Specification C-1 75(Q), Revision 6, "Requirements for Seismic Evaluation of Electrical and Mechanical Equipment."

42. EPRI-NP-6041-SL, "Methodology for Assessment of Nuclear Power Plant Seismic Margin,"

Revision 1, August 1991.  :

43. EPRI TR-103959, "Methodology for Developing Seismic Fragilities," July 1994.

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

45. *Entergy Document, "Palisades Nuclear Plant Individual Plant Examination of External Events (IPEEE)," Revision 1, May 1996.

46. 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.

47. 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-lchi Accident," May 9, 2014, NRC ADAMS Accession No. ML14111A147.

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

Seismic. Electric Power Research Institute," February 2013.

49. 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, NRC ADAMS Accession No. ML13101A379.

50. NRC (E Leeds) Letter to NEI (J Pollock), "Electric Power Research Institute Final Draft Report xxxxx, "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, NRC ADAMS Accession No. ML13106A331.

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A. Document No.: 51-9231248-002 AR EVA Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant

51. *Entergy Calculation EA-EC46465-05, "Anchorage Calculations for Electrical Panels EP-1 901 and EP-2001 ," Revision 0.

52. *Entergy Document, EA-EC48188-02, "Seismic Bracing of Non-Q Block Wall - South Wall of Lube Oil Storage Room," Revision 0.

53. *Entergy Document EC54011, "Fukushima Seismic 2.1 - Expedited Seismic Evaluation Program (ESEP) Report - Owner Acceptance of Report and Background Documentation," the following AREVA documents are captured in the plant document management system:

a. AREVA Document 51-9212955-007, "ESEP Expedited Seismic Equipment List (ESEL) -

Palisades Nuclear Plant."

b. AREVA Calculation 32-9227862-001, "Palisades ESEP HCLPF Calculation - Electrical Cabinet EC-02."

c. AREVA Calculation 32-9227902-001, "Palisades ESEP HCLPF Calculation - 480V Load Center, EB-1 1."

d. AREVA Calculation 32-9227941-001,-"Palisades ESEP HCLPF Calculation - Horizontal Pump, P-55C."

e. AREVA Calculation 32-9227961-001, "Palisades ESER HCLPF Calculation - 125,VDC Main Station Batteries ED-01 andED-02."

f. AREVA Calculation 32-9228097-001, "Palisades ESEP HCLPF Calculation - 480V Motor Control Center, EB-01."

g. AREVA Calculation 32-9228681-001, "Palisades ESEP HCLPF Calculation - Auxiliary Feedwater Controls, EJ-1 051."

h. AREVA Calculation 32-9229831-002, "Palisades ESEP HCLPF Calculation - Primary System Storage Tank, T-81."

i. AREVA Calculation 32-9230336-001, "Palisades ESEP Screening of Lube Oil Storage Room South Block Wall."

j. AREVA Calculation 32-9230249-003, Palisades ESEP HCLPF Calculation - Boric Acid Storage Tanks T-53A & T-53B."

k. AREVA Calculation 32-9228279-002, "Palisades ESEP HCLPF Calculation - Block Wall, C-104.11Q."

I. AREVA Calculation 32-9228841-001, "Palisades ESEP HCLPF Calculation - Block Walls C-107.16Q, C-107.17Q, and C-107.18Q."

The following references are AREVA references which were used as input for Appendix A.

54. AREVA Document 32-9223795-002, "Palisades ESEP Binning and Screening."

55. AREVA Document 51-9230420-001, "Input to Entergy ESEP Report Sections 2 and 3 for Palisades Nuclear Plant."

56. AREVA Document 51-9227749-000, "Input to Entergy ESEP Report Sections 4 and 5 for Palisades Nuclear Plant."

57. AREVA Document 51-9231028-001, "Input to Entergy ESEP Report Sections 6, 7, and 8 for Palisades Nuclear Plant."

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58. AREVA Document 38-9232220-001, "Palisades Nuclear Power Plant ESEP Report Comment Resolution Form."

59. AREVA Document 38-9233072-000, "Palisades Nuclear Power Plant ESEP Report, Revision 001, Comment Resolution Form."

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A Document No.: 51-9231248-002 AR EVA Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant APPENDIX A: EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT FOR PALISADES NUCLEAR PLANT Note: Customer requested formatting begins on the following page.

Page A-1

EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT FOR PALISADES NUCLEAR PLANT Page I

S- Palisades Nuclear Plant ESEP Report Table of Contents Page LIST OF TABLEES ................................................................. .......................................................................... 4 LIST OF FIGURES .......................................................................................................................................... 5 1.0 PURPOSE AND OBJECTIVE ........................................................................................................... 6 2.0 BRIEF

SUMMARY

OF THE FLEX SEISMIC IM PLEMENTATION STRATEGIES ................................. 6 3.0 EQUIPMENT SELECTION PROCESS AND ESEL ............................................................................... 7 3.1 Equipment Selection Process and ESEL ......................................................................... 7 3.1.1 ESEL Development .......................................................................................... 8 3.1.2 Power Operated Valves .................................................................................. 9 3.1.3 Pull Boxes ........................................................................................................ 9 3.1.4 Termination Cabinets ........................................................................................ 9 3.1.5 Critical Instrumentation Indicators ................................................................ 10 3.1.6 Phase 2 and 3 Piping Connections ............................................................... 10 3.2 Justification for Use of Equipment That is Not the Primary Means for FLEX Implementation .................... ............................................................................................ 10 4.0 GROUND MOTION RESPONSE SPECTRUM (GMRS)................................................................... 10 4.1 Plot of GM RS Submitted by the Licensee .................................................................... 10 4.2 Comparison to SSE ........................................................................................................ 12 5.0 REVIEW LEVEL GROUND MOTION (RLGM) .............................................................................. 13 5.1 Description of RLGM Selected ....................................................................................... 13 5.2 Method to Estimate In-Structure Response Spectra (ISRS) .......................................... 15 6.0 SEISMIC MARGIN EVALUATION APPROACH ................................... 15 6.1 Summary of Methodologies Used ................................................................................ 15 6.2 HCLPF Screening Process .............................................................................................. 15 6.3 Seismic W alkdown Approach ....................................................................................... 16 6.3.1 W alkdown Approach .................................................................................... 16 6.3.2 Application of Previous W alkdown Information ............................................ 17 6.3.3 Significant W alkdown Findings ....................................................................... 17 6.4 HCLPF Calculation Process ............................................................................................ 18 6.5 Functional Evaluations of Relays .................................................................................. 19 6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes) ..................................... 19 7.0 INACCESSIBLE ITEMS ..................................................................................................................... 19 7.1 Identification of ESEL Item Inaccessible for W alkdowns ............................................... 19 7.2 Planned W alkdown / Evaluation Schedule / Close Out ............................................... 21 8.0 ESEP CONCLUSIONS AND RESULTS .......................................................................................... 22 8.1 Supporting Information .................................................................................................... 22 Page 2

Palisades Nuclear Plant ESEP Report Table of Contents (continued)

Page 8.2 Identification of Planned Modifications ...................................................................... 23 8.3 Modification Implementation Schedule ....................................................................... 23 8.4 Summary of Regulatory Commitments ......................................................................... 24 9 .0 REFER EN CES .................................................................................................................................. 24 ATTACHMENT A - PALISADES NUCLEAR PLANT ESEL ......................................................................... A-1 ATTACHMENT B - ESEP HCLPF VALUES AND FAILURE MODES TABULATION ..................................... B-1 Page 3

Palisades Nuclear Plant ESEP Report List of Tables Page TABLE 4-1: GM RS FOR PALISADES NUCLEAR PLANT ......................................................................... .. 10 TABLE 4-2: SSE FOR PALISADES NUCLEAR PLANT ................................................................................ 1-2 TABLE 5-1: RLGM FOR PALISADES NUCLEAR PLANT ............................................................................ 14 Page 4

Palisades Nuclear Plant ESEP Report List of Figures Page FIGURE 4-1: GM RS FOR PALISADES NUCLEAR PLANT ......................................................................... 12 FIGURE 4-2: GMRS TO SSE COMPARISON FOR PALISADES NUCLEAR PALISADES ................................ 13 FIGURE 5-1: RLGM FOR PALISADES NUCLEAR PLANT ......................................................................... 14 Page 5

- Palisades Nuclear Plant ESEP-Report 1.0 PURPOSE AND OBJECTIVE Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March 11,.

2011, Great Tohoku Earthquake and subsequient tsunami, the Nuclear Regulatory Commission (NRC) established a Near-Term Task Force (NTTF) to conduct a systematic review of NRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena. Subsequently, the NRC issued a 50.54(f) letter on March 12, 2012 [1], requesting information to assure that these recommendations are addressed by all U.S. nuclear power plants. The 50.54(f) letter 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 Palisades Nuclear Plant. The intent of the ESEP is to perform an interim action in response to the NRC's 50.54(f) letter to demonstrate seismic margin through a review of a subset of the plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events.

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

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

2.0 BRIEF

SUMMARY

OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES The Palisades Nuclear Plant FLEX strategies for Reactor Core Cooling and Heat Removal, Reactor Inventory Control, and Containment Function are summarized below. This summary is derived from the Palisades Nuclear Plant Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order EA-12-049 [3] as augmented by the second six-month Status Report [4], the third six-month Status Report [5] and the FLEX Basis Engineering Change (EC) [6].

During Phase 1 reactor core cooling and heat removal are achieved with feedwater supply to the steam generators using the Turbine-Driven Auxiliary Feedwater (TDAFW) pump aligned to take suction from the Condensate Storage Tank (CST) and heat removal using the Atmospheric Dump Valves (ADVs).

Operation of the ADVs can be accomplished by local operation. Steam supply valves for the TDAFW pump and Auxiliary Feedwater (AFW) flow control valves to the steam generators are also required.

At approximately two (2) hours after the event starts, operators will initiate a controlled cooldown-depressurization by opening the ADVs. Decreasing the Primary Coolant System (PCS) pressure permits the Safety Injection Tanks (SITs) to provide injection of borated water to the PCS to make-up for volume shrinkage and any minor inventory losses.

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Palisades Nuclear Plant ESEP Report Prior to depletion of the CST, the primary makeup storage tank will be cross-connected to the CST for additional water. The combined capacity of these tanks provides about eight (8) hours of steam generator feed.

The Phase 2 strategy involves switching steam generator feed to an on-site portable, diesel driven FLEX pump taking water directly from Lake Michigan and pumping it to the steam generators.

Reactor Coolant System (RCS) inventory control will use the installed charging pumps, powered by a portable FLEX diesel generator, with borated water for injection from the installed Concentrated Boric Acid Storage Tanks (CBASTs). Prior to depletion of the CBASTs, boric acid batching operations will begin using the installed boric acid batching tank.

Containment function is not challenged early in the event and no actions are required during Phase 1 or Phase 2. During Phase 3, the strategy involves use of Regional Response Center (RRC) equipment to provide containment cooling via the Service Water System and the Containment Air Coolers and Fans.

Necessary electrical components are outlined in the Palisades Nuclear Plant FLEX OIP submittal, Second Six-Month Status Report and FLEX Basis, and primarily entail 480 volt AC Motor Control Centers (MCCs), 125 volt DC MCCs, vital batteries, battery chargers, and 120 volt AC distribution panels. Other supporting components include monitoring instrumentation for core cooling, reactor coolant inventory, and containment integrity. The FLEX strategy includes operators shedding unnecessary DC loads to extend the battery life into Phase 2.

The figures provided in Attachment'3 of Reference [3] provide the conceptual FLEX flow paths for Palisades Nuclear Plant Phases 1, 2 and 3.

3.0 EQUIPMENT SELECTION PROCESS AND ESEL The selection of equipment for the Expedited Seismic Equipment List (ESEL) followed the guidelines of EPRI 3002000704 [2]. The ESEL for Palisades Nuclear Plant is presented in Attachment A. Information presented in Attachment A is drawn from the following references [3], [4],[5], [6], [7], [8], [9], [10],

[11], [12], [13], [14], [15], [16], [17], [18], [19], [20], [21], [22], [23], [24], [25], [26], [27], [28], [29],

[30], [31], [32], [33], [34], [35], [36], [37], and [38].

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 Palisades Nuclear Plant OIP in Response to the March 12, 2012, Commission Order EA-12-049 [3], second six-month Status Report [4], third six-month Status Report

[5], and the FLEX Basis Engineering Change [6]. These references provide the Palisades Nuclear Plant FLEX mitigation strategy and serve 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 Palisades Nuclear Plant OIP, second six-month Status Report, and the FLEX Basis. FLEX recovery actions are excluded from the ESEP scope per EPRI 3002000704 [2]. The overall list of planned FLEX modifications and the scope for consideration herein is limited to those required to support core cooling, reactor coolant inventory, and containment integrity functions. The scope of the ESEL does not include Spent Fuel Pool (SFP) components since the SFP is excluded by EPRI 3002000704 [2]. Portable and pre-staged FLEX equipment (not permanently installed) are excluded from the ESEL per EPRI 3002000704.

Page 7

Palisades Nuclear Plant ESEP Report The ESEL component selection followed the EPRI guidance outlined in Section 3.2 of EPRI 3002000704.

1. The scope of components is limited to that required to accomplish the core cooling and containment-safety functions identified~inTable 3-2 of EPRI 3002000704. The instrumentation monitoring requirements for core cooling/containment safety functions are limited to those outlined in the EPRI 3002000704 guidance, and are a subset of those outlined in the Palisades Nuclear Plant OIP, second six-month Status Report, third six-month Status Report, and the FLEX Basis Engineering Change.

2. The scope of components is limited to installed plant equipment, and FLEX connections necessary to implement the Palisades Nuclear Plant OIP, second six-month Status Report, third six-month Status Report, and FLEX Basis Engineering Change as described in Section 2 in this report.

3. The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e.,

either "Primary" or "Back-up/Alternate").

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

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

6. Structures, systems, and components excluded per the EPRI 3002000704 guidance are:

• Structures (e.g. containment, reactor building, controlbuilding, auxiliary building, etc.)

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

• Manual valves, check valves and rupture disks

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

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

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 Palisades Nuclear Plant OIP [3], second six-month Status Report [4], third six-month Status Report [5] and FLEX Basis Engineering Change [6] to determine the major equipment involved in the FLEX strategies. Further reviews of plant drawings (e.g., Flow Diagrams 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) in branch circuits / branch lines off the defined strategy electrical or fluid flow path. Flow Diagrams were the primary reference documents used to identify mechanical components. The flow paths used for FLEX strategies were selected and specific components were identified using detailed equipment and instrument drawings, electrical schematics and one-line drawings, system descriptions, and design basis documents, as necessary.

Page 8

- Palisades Nuclear Plant ESEP Report Cabinets containing electrical equipment and instrumentation, which could be affected by earthquake motion and potentially impact the operation of equipment on the ESEL, are required to be included on the ESEL for evaluation.

For Phase 1, TDAFW is the primary path for core cooling. For Phase 2, a portable diesel-driven FLEX pump is used. For Phase 3, the RRC equipment is used to provide the power and pumps for core cooling and inventory control. An extensive relay evaluation was performed as part of the A-46 and IPEEE programs and after screening out relays for which relay chatter is not an issue, no bad actors were identified. In addition, due to the Extended Loss of AC Power (ELAP) from the BDBEE, without power, chatter of relays supporting higher voltage equipment will not result in any negative effects from energized relay seal-in or lock-out. Therefore, no relays were listed on the ESEL.

For each parameter monitored during the FLEX implementation, a single indication loop was selected for inclusion in the ESEL per EPRI 3002000704 [2). For each parameter indication, the components along the flow path from measurement to indication were included, since any failure along the path would lead to failure of that indication. Components such as flow elements were considered as part of the piping and were excluded from the ESEL.

3.1.2 Power Operated Valves Page 3-3 of EPRI 3002000704 [2] notes that power operated valves not required to change state are excluded from the ESEL. Page 3-2 also notes that "functional failure modes of electrical and mechanical portions of the installed Phase 1 equipment should be considered (e.g. AFW trips)." To address this concern, the following guidance is applied in the Palisades Nuclear Plant ESEL for functional failure modes associated with power operated valves:

• Power operated valves that remain energized during the ELAP events (such as Direct Current (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 be added to the ESEL as these components provide completely passive locations for pulling or installing cables. No breaks or connections in the cabling were included in pull boxes. Pull boxes were considered part of conduit and cabling, which were excluded in accordance with EPRI 3002000704 [21.

3.1.4 Termination Cabinets Termination cabinets necessary for FLEX Phase 2 and Phase 3 connections from FLEX components, provide consolidated locations for permanently connecting multiple cables. The termination cabinets and the internal connections provide a completely passive function; however, the cabinets are Page 9

Palisades Nuclear Plant ESEP Report included in the ESEL to ensure industry knowledge on panel/anchorage failure vulnerabilities are addressed.

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

3.1.6 Phase 2 and 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 Palisades Nuclear Plant OIP, second six-month Status Report, third six-month Status Report, and the FLEX Basis Engineering Change as described in Section 2." Item 3 in Section 3.1 also notes that "The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e., either "Primary" or "Back-up/Alternate")."

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

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

3.2 Justification for Use of Equipment That is Not the Primary Means for FLEX Implementation All equipment on the ESEL is part of the primary means of FLEX implementation. Therefore, no additional justification is required.

4.0 GROUND MOTION RESPONSE SPECTRUM (GMRS) 4.1 Plot of GMRS Submitted by the Ucensee The Safe Shutdown Earthquake (SSE) control point elevation is defined at the ground surface, in the free field [39]. Table 4-1 shows the GMRS acceleration for a range of frequencies [40]. The GMRS at the control point is shown in Figure 4-1.

Table 4-1: GMRS for Palisades Nuclear Plant Frequency GMRS (Hz) (g) 100 2.83E-01 90 2.87E-01 80 2.92E-01 70 2.99E-01 60 3.13E-01 50 3.41E-01 40 3.87E-01 35 1 4.13E-01 Page 10

Palisades Nuclear Plant ESEP Report Table 4-1: GMRS for Palisades Nuclear Plant (continued)

Frequency GMRS (Hz) (g) 30 4.46E-01 25 4.93E-01 20 5.11E-01 15 5.46E-01 12.5 5.43E-01 10 5.49E-01 9 5.59E-01 8 5.49E-01 7 5.08E-01 6 4.82E-01 5 4.66E-01 4 4.38E-01 3.5 4.16E-01 3 3.67E-01 2.5 3.07E-01 2 2.90E-01 1.5 2.25E-01 1.25 1.90E-01 1 1.49E-01 0.9 1.30E-01 0.8 1.11E-01 0.7 9.47E-02 0.6 7.97E-02 0.5 6.51E-02 0.4 5.21E-02 0.35 4.56E-02 0.3 3.91E-02 0.25 3.26E-02 0.2 2.60E-02 0.15 1.95E-02 0.125 1.63E-02 0.1 1.30E-02 Page 11

Palisades Nuclear Plant ESEP Report-GMRS at Control Point for Palisades Nuclear Plant, 5% Damping 0.60 0.1 10 100 Frequency (Hz)

Figure 4-1: GMRS for Palisades Nuclear Plant 4.2 Comparison to SSE The SSE is defined in the Final Safety Analysis Report [39] in terms of a Peak Ground Acceleration (PGA) and a design response spectrum. These spectra have been digitized and tabulated [40]. Table 4-2 shows the spectral acceleration values at selected frequencies for the 5% damped horizontal SSE.

Table 4-2: SSE for Palisades Nuclear Plant Frequency Spectral (Hz) Acceleration (g) 100 0.2 25 0.206 10 0.24 5 0.31 2.5 0.285 1 0.16 0.5 0.096 Page 12

Palisades Nuclear Plant ESEP Report GMRS to SSE Comparison for Palisades Nuclear Plant, 5% Damping 0.60 0.30 0.1 10 100 FrequencytHz)

Figure 4-2: GMRS to SSE Comparison for Palisades Nuclear Palisades The SSE envelops the GMRS in the low frequency range up to nearly 1.5 Hz. The GMRS exceeds the SSE beyond that point. As the GMRS exceeds the SSE in the I to 10 Hz range, the plant does not screen out of the ESEP according to Section 2.2 of EPRI 3002000704 [2]. The two special screening considerations as described in Section 2.2.1 of EPRI 3002000704, namely a) Low Seismic Hazard Site and b) Narrow Band Exceedances in the 1 to 10 Hz range, provide criteria for accepting specific GMRS exceedances.

However, the GMRS exceedances are not limited to the low frequency range and there are no narrow-banded exceedances. Therefore, these special screening considerations do not apply for Palisades Nuclear Plant and hence High Confidence of a Low Probability of Failure (HCLPF) evaluations are to be performed.

5.0 REVIEW LEVEL GROUND MOTION (RLGM) 5.1 Description of RLGM Selected The RLGM is selected based on Approach 1 in Section 4 of EPRI 3002000704 [2]. The RLGM is developed based on the SSE [41].

The maximum GMRS/SSE ratio between I and 10 Hz range occurs at 10 Hz where the ratio is 0.549/0.24 = 2.29. As the maximum ratio of the GMRS to the SSE over the I to 10 Hz range exceeds a value of 2, the GMRS/SSE ratio is set to the maximum scaling factor value of 2.0 for Palisades Nuclear Plant in accordance with Section 4 of EPRI 3002000704. Table 5-1 lists the horizontal ground RLGM acceleration at 5% damping at selected frequencies and the plot is shown in Figure 5-1. The RLGM is generated by plotting the digitized data on a log/linear graph paper, and connecting the points with straight lines.

Page 13

Palisades Nuclear Plant ESEP Report Table 5-1: RLGM for Palisades Nuclear Plant Frequency RLGM at 5%

(Hz) Damping (g) 100.00 0.400 33.00 0.400 15.00 0.432 10.00 0.480 6.60 0.620 3.00 0.620 1.00 0.320 0.50 0.192 0.10 0.000 Review Level Ground Motion (2xSSE) Response Spectra - Horizontal Direction 0.70 0.60 0.50 0.40 0.30 0.20 0.10 0.00 -K 0.100 1.000 10.000 100.000 Fmquency (Wz)

Figure 5-1: RLGM for Palisades Nuclear Plant Page 14

Palisades Nuclear Plant ESEP Report 5.2 Method to Estimate In-Structure Response Spectra (ISRS)

The RLGM ISRS for Palisades Nuclear Plant are generated by scaling the SSE ISRS [41]. The following steps are used to generate the RLGM ISRS:

1. Obtain the horizontal direction SSE ISRS for a particular damping value.

2. Calculate the horizontal RLGM ISRS by scaling the horizontal direction SSE ISRS by a factor of 2.0.

3. Repeat steps I and 2 to obtain RLGM ISRS for multiple damping values.

The vertical direction RLGM ISRS is obtained by scaling the vertical amplified ground response spectrum.

6.0 SEISMIC MARGIN EVALUATION APPROACH It is necessary to demonstrate that ESEL items have sufficient seismic capacity to meet or exceed the demand characterized by the RLGM. The seismic capacity is characterized as the PGA for which there is a HCLPF. The PGA is associated with a specific spectral shape, in this case the 5%-damped RLGM spectral shape. The HCLPF capacity must be equal to or greater than the RLGM PGA. The criteria for seismic capacity determination are given in Section 5 of EPRI 3002000704 [2].

There are two basic approaches for developing HCLPF capacities:

1. Deterministic approach using the Conservative Deterministic Failure Margin (CDFM) methodology of EPRI NP-6041-SL, A Methodology for Assessment of Nuclear Power.Plant Seismic Margin (Revision 1) [42].

2. Probabilistic approach using the fragility analysis methodology of EPRI TR-103959, Methodology for Developing Seismic Fragilities [43].

6.1 Summary of Methodologies Used Palisades Nuclear Plant was classified as a 0.3g focused scope in NUREG-1407 [44] and performed a Seismic Probabilistic Risk Assessment (SPRA) in accordance with the methodology of NUREG-1407 in 1996 as part of Individual Plant Examination for External Events (IPEEE) program. The SPRA is documented in [45] and consisted of screening evaluations, seismic walkdowns, and fragility analysis.

SPRA screening was performed in accordance with EPRI NP-6041-SL [42]. Seismic walkdowns took advantage of overlapping requirements between IPEEE and USI A-46 programs. Section 3.3 of [40]

established that the results of the Palisades Nuclear Plant IPEEE are not adequate to support screening of the updated seismic hazard for Palisades Nuclear Plant.

For ESEP, the screening walkdowns used the screening tables from Chapter 2 of EPRI NP-6041-SL. The walkdowns were conducted by engineers trained in EPRI NP-6041-SL and were documented on Screening Evaluation Work Sheets (SEWS) from EPRI NP-6041-SL. Anchorage capacity calculations used the CDFM criteria from EPRI NP-6041-SL. Seismic demand was based on EPRI 3002000704 [2] using an RLGM of 2.OOxSSE with a PGA of 0.40g, Figure 5-1.

6.2 HCLPF Screening Process For ESEP, the components are screened considering the RLGM (2.OOxSSE) with a 0.40g PGA. The screening tables in EPRI NP-6041-SL [42] are based on ground peak spectral accelerations of 0.8g and 1.2g. These both exceed the RLGM peak spectral acceleration.

Page 15

Palisades Nuclear Plant ESEP Report The ESEL components were prescreened based on Table 2-4 of EPRI NP-6041-SL. Additional pre-screening, specifically for anchorage, considered walkdown results and documentation from NTTF 2.3 and SEWS from IPEEE and USI A-46. Equipment anchorage was screened out in cases where previous evaluations showed large available margin against SSE. The remaining components (i.e., components that do not screen out), were identified as requiring HCLPF calculations. ESEL components were walked down and based on the equipment and anchorage conditions, prescreening decisions were confirmed and a final list of required HCLPF calculations was generated. Equipment for which the screening caveats were met and for which the anchorage capacity exceeded the RLGM seismic demand are screened out from ESEP seismic capacity determination because the HCLPF capacity exceeds the RLGM.

The Palisades Nuclear Plant ESEL contains 178 items. Of these, 47 are valves. In accordance with Table 2-4 of EPRI NP-6041-SL, active valves may be assigned a functional capacity of 0.8g peak spectral acceleration without any review other than looking for valves with large extended operators on small diameter piping, and anchorage is not a failure mode. The review of valves with large extended operators on small diameter piping is performed during walkdowns. Significant walkdown findings are summarized in Section 6.3. Therefore, valves on the ESEL are screened out from ESEP seismic capacity determination, subject to the caveat regarding large extended operators on small diameter piping.

The non-valve components in the ESEL are screened based on the SMA results. If the SMA showed that the component met the EPRI NP-6041-SL screening caveats and the CDFM capacity exceeded the RLGM demand, the components are screened out from the ESEP capacity determination.

Block Walls were identified in the proximity of ESEL equipment. The HCLPF capacity of these walls was evaluated and determined to be sufficient to meet RLGM demand.

6.3 Seismic Walkdown Approach 6.3.1 Walkdown Approach Walkdowns were performed in accordance with the criteria provided in Section 5 of EPRI 3002000704

[2], which refers to EPRI NP-6041-SL [42] for the SMA process. Pages 2-26 through 2-30 of EPRI NP-6041-SL describe the seismic walkdown criteria, including the following key criteria.

"The SRT [Seismic Review Team] should "walk by" 100% of all components which are reasonably accessible and in non-radioactiveor low radioactiveenvironments. Seismic capability assessment of components which are inaccessible,in high-radioactiveenvironments, or possibly within contaminated containment, 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 necessaryto inspect one component out of this group. The "similarity-basis"should be developed before the walkdown during the seismic capability preparatory work (Step 3) by reference to drawings,calculations or specifications. The one component or each type which is selected should be thoroughly inspected which probably does mean de-energizing and opening cabinets or panelsfor this very limited sample. Generally, a spare representativecomponent can be Page 16

Palisades Nuclear Plant ESEP Report found so as to enable the inspection to be performed while the plant is in operation. At leastfor the one component of each type which is selected, anchorageshould be thoroughly inspected.

The walkdown procedure should be performed in an ad hoc manner. Foreach class of components the SRT should look closely at the first items and compare the field configurationswith the construction drawingsand/orspecifications. If a one-to-one correspondenceis found, then subsequent items do not have to be inspected in as great a detail. Ultimately the walkdown becomes a "walk by" of the component class as the SRT becomes confident that the constructionpattern is typical. This procedure for 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 drawings or questionableconstructionpracticesare found then the system or component class must be inspected in closer detail until the systematic deficiency is defined.

The 100% "walk by" is to look for outliers, lack of similarity,anchorage which is differentfrom that shown on drawings or prescribedin criteriafor that component, potentialSI [Seismic Interaction]

problems, situations that are at odds with the team members' past experience, and any other areas of serious seismic concern. If any such concerns surface, then the limited sample size of one component of each type for thorough inspection will have to be increased. The increase in sample size which should be inspected will depend upon the number of outliers and different anchorages,etc., which are observed. It is up to the SRT to ultimately select the sample size since they are the ones who are responsiblefor the seismic adequacy of all elements which they screen from the margin review.

Appendix 6 gives guidancefor sampling selection."'

6.3.2 Application of Previous Walkdown Information Several ESEL items were previously walked down during the Palisades Nuclear Plant seismic IPEEE program, for the USI A-46 evaluation program, and NTTF Recommendation 2.3. Those walkdown results were reviewed and the following steps were taken to confirm that the previous walkdown conclusions remained valid.

" A walk by was performed to confirm that the equipment material condition and configuration is consistent with the walkdown conclusions and that no new significant interactions related to block walls or piping attached to tanks exist.

" If the ESEL item was screened out based on previous walkdowns, that screening evaluation was reviewed and reconfirmed for the ESEP.

6.3.3 Significant Walkdown Findings Consistent with the guidance from EPRI NP-6041-SL [42], no significant outliers or anchorage concerns were identified during the Palisades Nuclear Plant seismic walkdowns. Based on pre-screening of components and walkdown results, HCLPF capacity evaluations were recommended for the following ten (10) components:

9 Electrical Cabinet, EC-02

• 480V Load Center, EB-11

• Horizontal Pump, P-55C 0 125 VDC Main Station Battery, ED-01 Page 17

Palisades Nuclear Plant ESEP Report

• 125 VDC Main Station Battery, ED-02

• 480V Motor Control Center, EB-01

" Auxiliary Feedwater Controls, EJ-1051

" Primary Makeup Storage Tank, T-81

• Boric Acid Storage Tank, T-53A

• Boric Acid Storage Tank, T-53B Block walls were identified in the proximity of ESEL equipment. These block walls were assessed for their structural adequacy to withstand the seismic loads resulting from the RLGM. For any cases where the block wall represented the HCLPF failure mode for an ESEL item, it is noted in the tabulated HCLPF values described in Section 6.6. Three (3) HCLPF evaluations were performed addressing the block walls in close proximity to components ED-01, ED-02, ED-10L, ED-1OR, ED-20L, ED-20R, EB-23, EC-33, and N2 Backup Station #2.

• C-104.11Q

" C-107.16Q0 C-107.17QC and C-107.18Q

• Lube Oil Storage Room South Block Wall 6.4 HCLPF Calculation Process ESEL items identified for ESEP at Palisades Nuclear Palisades Were evaluated using the criteria in EPRI NP-6041-SL [42] and Section 5 of EPRI 3002000704 [2]. Those evaluations included the following steps:

" Performing seismic capability walkdowns for equipment not included in previous seismic walkdowns (SQUG, IPEEE, USI A-46, or NTTF 2.3) to evaluate the equipment installed plant conditions

" Performing screening evaluations using the screening tables in EPRI NP-6041-SL as described in Section 6.2

" 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. A'total of eight (8) HCLPF calculations were performed to address the ten (10) components as well as three (3) calculations to address seismic adequacy of block walls.

• Electrical Cabinet EC-02

• 480V Load Center, EB-11

• Horizontal Pump, P-55C

• 125 VDC Main Station Batteries ED-01 and ED-02

• 480V Motor Control Center, EB-01

• Auxiliary Feedwater Controls, EJ-1051 0 Primary Makeup Storage Tank, T-81 Page 18

Palisades Nuclear Plant ESEP Report

" Boric Acid Storage Tanks T-53A & T-53B

• Lube Oil Storage Room South Block Wall

" Block Wall, C-104.11Q

" Block Walls C-107.160, C-107.170, and C-107.18Q 6.5 Functional Evaluations of Relays As discussed in Section 3.1.1, no seal in/lockout type relays were identified on Palisades Nuclear Plant ESEL. Therefore, no relay evaluations were performed.

6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes)

Tabulated ESEL HCLPF values are provided in Attachment B. The following notes apply to the information in the tables.

" For items screened out using EPRI NP-6041-SL [42] screening tables, the HCLPF capacity is provided as >RLGM and the failure mode is listed as "Screened", (unless the controlling HCLPF value is governed by anchorage).

• For items where anchorage controls the HCLPF value, the HCLPF value is listed in the table and the failure mode is noted as "anchorage." For the items where the component function controls the HCLPF value, the HCLPF value is listed in the table and the failure mode is'noted as:

"functional."

After performing the HCLPF calculations, ESEL components were determined to have adequate capacity for the design basis loads and HCLPF greater than RLGM for all components except the following:

• T-53A, Boric Acid Storage Tank

• T-53B, Boric Acid Storage Tank

" T-81, Primary Makeup Storage Tank Modifications are required for the above components.

7.0 INACCESSIBLE ITEMS 7.1 Identification of ESEL Item Inaccessible for Walkdowns Thirty-six (36) components on the ESEL were inaccessible and not walked down because the walkdowns were performed with the plant on-line. These components are located in the Reactor Containment Building as well as in a locked, high radiation area within the Auxiliary Building. One component (LT-0372, Safety Injection Tank T-82C Level Transmitter) was evaluated based on review of existing photographs. The remaining thirty-five (35) components listed below require follow-up seismic walkdowns.

" CV-0861, Service Water System Return Isolation Valve from CAC Fan VHX-1

• CV-0864, Service Water System Return Isolation Valve from CAC Fan VHX-2 Page 19

Palisades Nuclear Plant ESEP Report SV-0861, Service Water System Return Isolation Valve from CAC Fan VHX-1

- SV-0864, Service Water System Return Isolation Valve from CAC Fan VHX-2

" V-1A, Containment Air Cooler Fan

" V-2A, Containment Air Cooler Fan

• V-3A, Containment Air Cooler Fan

" VHX-1, Containment Air Cooler

" VHX-2, Containment Air Cooler

• VHX-3, Containment Air Cooler

• CV-2083, CBO Isolation Valve

• CV-2099, CBO Isolation Valve

" CV-2191, CBO Relief Stop Valve

" E-56, Regenerative Heat Exchanger

" MO-3007, HPSI Injection Loop Isolation Valve

• MO-3041, SIT T-82A Isolation Valve

• MO-3045, SIT T-82B Isolation Valve

" MO-3049, SIT T-82C Isolation Valve

• MO-3052, SIT T-82D Isolation Valve

" LT-0102, PZR Level (WR) Transmitter

• LT-0365, SIT T-82A Level Transmitter

" LT-0368, SIT T-82B Level Transmitter

• LT-0374, SIT T-82D Level Transmitter

• LT-0757A, SG A WR Level Transmitter

• LT-0758A, SG B WR Level Transmitter Page 20

Palisades Nuclear Plant.ESEP Report

" PT-0105A, PZR Pressure (WR) Transmitter

• PT-0751C, SG A Pressure Transmitter

" PT-0752C, SG B Pressure Transmitter

" TE-0112CC, PCS T-cold Temperature Element

• TE-O112HC, PCS T-hot Temperature Element

• LE-0101A, Reactor Vessel Level

" T-82A, Safety Injection Tank A

" T-82B, Safety Injection Tank B

• T-82C, Safety Injection Tank C

• T-82D, Safety Injection Tank D 7.2 Planned Walkdown / Evaluation Schedule / Close Out The walkdowns of the inaccessible items identified in Section 7.1 are currently scheduled to be performed during Palisades' next refueling outage with subsequent evaluations of those components to follow. Section 8.4 summarizes regulatory commitments for the close out of inaccessible components.

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Palisades Nuclear Plant ESEP Report 8.0 ESEP CONCLUSIONS AND RESULTS 8.1 Supporting Information Palisades Nuclear Plant has performed the ESEP as an interim action in response to the NRC's 50.54(f) letter [1]. It was performed using the methodologies in the NRC endorsed guidance in EPRI 3002000704 [2].

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

The ESEP is part of the overall Palisades Nuclear Plant response to the NRC's 50.54(f) letter. On March 12, 2014, NEI submitted to the NRC results of a study [46] 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 [47] coricluded that the "fleet wide 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 Palisades Nuclear Plant was included in the fleet risk evaluation submitted in the March 12, 2014 NEI letter [46] therefore, the conclusions in the NRC's May 9 letter also apply to Palisades Nuclear Plant.

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

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

• Response spectra enveloping criteria typically used in SSC analysis and testing applications Page 22

Palisades Nuclear Plant ESEP Report

" 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, etc.)

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

The intent of the ESEP is to perform an interim action in response to the NRC's 50.54(f) letter to demonstrate seismic margin through a review of a subset of the plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events. The RLGM used for the ESEP evaluation is a scaled version of the plant's SSE rather than the actual GMRS. To more fully characterize the risk impacts of the seismic ground motion represented by the GMRS on a plant specific basis, a more detailed seismic risk assessment (SPRA or risk-based SMA) is to be performed in accordance with EPRI 1025287 [48]. As identified in the Palisades Nuclear Plant Seismic Hazard and GMRS submittal [40], Palisades Nuclear Plant screens in for a risk evaluation. The complete risk evaluation will more completely characterize the probabilistic seismic ground motion input into the plant, the plant response to that probabilistic seismic ground motion input, and the resulting plant risk characterization. Palisades Nuclear Plant will complete that evaluation in accordance with the schedule identified in NEI's letter dated April 9, 2013 [491 and endorsed by the NRC in their May 7, 2013 letter [50].

8.2 Identification of Planned Modifications Insights from the ESEP identified the following items where the HCLPF is below the RLGM and plant modifications will be made in accordance with EPRI 3002000704 [2] to enhance the seismic capacity of the plant.

" T-53A, Boric Acid Storage Tank

" T-53B, Boric Acid Storage Tank T-53B

" T-81, Primary Makeup Storage Tank 8.3 Modification Implementation Schedule Plant modifications will be performed in accordance with the schedule identified in NEI letter dated April 9, 2013 [49], which states that plant modifications not requiring a planned refueling outage will be completed by December 2016 and modifications requiring a refueling outage will be completed within two planned refueling outages after December 31, 2014.

Page 23

Palisades Nuclear Plant ESEP Report 8.4 Summary of Regulatory Commitments The following actions will be performed as a result of the ESEP.

~ Equl~mnt Act~ons # ID DesntioH Perform seismic walkdowns, No later than the generate HCLPF calculations end of the second and design and implement planned refueling N/A any necessary modifications outage after 1 N/A for inaccessible items listed in December 31, 2014.

Section 7.1 Walkdowns planned for the next refueling outage.

2 T-53A Boric Acid Modify tank anchorage such As described in Storage Tank that HCLPF>RLGM Section 8.3 3 Boric Acid Modify tank anchorage such As described in Storage Tank that HCLPF>RLGM Section 8.3 4 T-81 Primary Makeup Modify tank such that As described in Storage Tank HCLPF>RLGM Section 8.3

Submit a letter td.NRC . Within 60 days

.. summarizing the HCLPF following results of Items 1 to 4 and completion of ESEP 5 N/A N/A confirming implementation of activities, including the plant modifications items 1 to 4 associated with items 1 to 4 Page 24

Palisades Nuclear Plant ESEP Report

9.0 REFERENCES

1. NRC (E Leeds and M Johnson) Letter to All Power Reactor Licensees et al., "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident," March 12, 2012, NRC ADAMS Accession No. ML12053A340.

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

3. Entergy Letter to U.S. NRC, letter number PNP 2013-010, "Overall Integrated Plan in Response to March 12, 2012 Commission Order to Modify Licenses With Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049),"

February 28, 2013, NRC ADAMS Accession No. ML13246A399.

4. Entergy Letter to U.S. NRC, Letter Number PNP 2014-011, "Palisades Nuclear Plant 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 28, 2014, NRC ADAMS Accession No. ML14059A078.

5. Entergy Letter to U.S. NRC, Letter Number PNP 2014-085, "Palisades Nuclear Plant 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 28, 2014, NRC ADAMS Accession No. ML14240A278.

6. Entergy Document, Engineering Change EC-46465, "FLEX Basis."

7. Entergy Drawing E0001, Sheet 1, Revision 83, "Single Line Meter & Relay Diagram 480V Motor Control Center Warehouse, WD 950."

8. Entergy Drawing E0001, Sheet 3, Revision 4, "Plant Single Line Diagram, WD 950."

9. Entergy Drawing E0008, Sheet 1, Revision 57, "Single Line Meter & Relay Diagram 125V DC 120V Instrument & Preferred AC, System WD 950."

10. Entergy Drawing E0008, Sheet 2, Revision 55, "Single Line Meter & Relay Diagram 125V DC 120V Instrument & Preferred AC, System WD 950."

11. Entergy Drawing E0078, Sheet 2, Revision 18, "WRSGL Schematic Diagram."

12. Entergy Drawing E0078, Sheet 2A, Revision 6, "WRSGL Schematic Diagram."

13. Entergy Drawing E0082, Sheet 5, Revision 13, "Schematic Diagram Wide Range Pressurizer Level Indicator/Alarm Instrumentation."

14. Entergy Drawing E0090, Sheet 5, Revision 14, "Schematic Diagram Boric Acid System Instrumentation."

15. Entergy Drawing E0092, Revision 14, "Schematic Diagram Safety Injection Tank Level Indicator Alarm Instrumentation."

16. Entergy Drawing M0201, Sheet 1, Revision 87, "Piping & Instrument Diagram, Primary Coolant System."

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Palisades Nuclear Plant ESEP Report

17. Entergy Drawing M0201, Sheet 2, Revision 66, "Piping & Instrument Diagram, Primary Coolant System."

18. Entergy Drawing M0202, Sheet 1, Revision 76, "Piping & Instrument Diagram, Chemical &

Volume Control System."

19. Entergy Drawing M0202, Sheet 1A, Revision 64, "Piping & Instrument Diagram, Chemical &

Volume Control System."

20. Entergy Drawing M0202, Sheet 1B, Revision 59, "Piping & Instrument Diagram, Chemical &

Volume Control System."

21. Entergy Drawing M0203, Sheet 1, Revision 48, "Piping & Instrument Diagram, Safety Injection, Containment Spray, Shutdown Cooling System."

22. Entergy Drawing M0203, Sheet 2, Revision 27, "Piping & Instrument Diagram, Safety Injection, Containment Spray, Shutdown Cooling System."

23. Entergy Drawing M0205, Sheet 2, Revision 69, "Piping & Instrument Diagram, Main Steam and Auxiliary Turbine Systems."

24. Entergy Drawing M0208, Sheet 1B, Revision 37, "Piping & Instrument Diagram, Service Water System."

25. Entergy Drawing M0220, Sheet 1, Revision 97, "'Piping & Instrument Diagram, Make-up

. .Domestic Water & Chemical Injection Systems."

26. Entergy Drawing M0207, Sheet 1, Revision 91, "Piping & Instrument Diagram, Feedwater &

Condensate System."

27. Entergy Drawing M0207, Sheet 2, Revision 38, "Piping & Instrument Diagram, Auxiliary Feedwater System."

28. Entergy Procedure Palisades EOP-3.0, Revision 16, "Station Blackout."

29. Entergy Drawing E0364, Revision 9, "Conduit and Tray Miscellaneous Plans."

30. Entergy Drawing E0618, Sheet 569, Revision 12, "Connection Diagram Junction Box J569."

31. Entergy Drawing M0222, Sheet 2, Revision 29 (EC-47346), "Piping & Instrument Diagram, Miscellaneous Gas Supply Systems."

32. Entergy Drawing M0218, Sheet 2, Revision 61, "Piping & Instrument Diagram, Htg. Vent. & Air Cond. Containment Building."

33. Entergy Drawing E0005, Sheet 5B, Revision 12, "Single Line Meter & Relay Diagram 480 Volt Motor Control Centers, System WD 950."

34. Entergy Drawing E0099 Sh. 5, Revision 10, "Schematic Diagram, Containment Building Instrumentation (Left Channel)."

35. Entergy Drawing E0084 Sh. 6, Revision 14, "Schematic Diagram, Pressurizer Pressure Control and Measurement Channel Instrumentation."

36. Entergy Drawing E0087, Sh. 6, Revision 10, "Schematic Diagram Level Indication and Alarm Indication."

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Palisades Nuclear Plant ESEP Report

37. Entergy Drawing E0076, Sheet 4C, Revision 8, "Schematic Diagram Feedwater and Turbine Driver Instrumentation."

38. Entergy Drawing E0226, Sheet 1B, Revision 6, "Schematic Diagram - Reactor Vessel Level Monitoring System (Left Channel)."

39. "Palisades Nuclear Plant - Final Safety Analysis Report," Revision 31, Docket 50-255, September 2014.

40. Entergy Letter to U.S. NRC, letter number PNP 2014-033,."Palisades Nuclear Plant Seismic Hazard and Screening Report (CEUS Sites), Response to NRC 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 March 31, 2014, NRC ADAMS Accession No. ML14090A069.

41. Entergy Technical Specification C-175(Q), Revision 6, "Requirements for Seismic Evaluation of Electrical and Mechanical Equipment."

42. EPRI-NP-6041-SL, "Methodology for Assessment of Nuclear Power Plant Seismic Margin,"

Revision 1, August 1991.

43. EPRI TR-103959, "Methodology for Developing Seismic Fragilities," July 1994.

44. NRC NUREG71407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," June 1991.

45. Entergy Document, "Palisades Nuclear Plant Individual Plant Examination of External Events (IPEEE)," Revision 1, May 1996.

46. 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.

47. 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-lchi Accident," May 9, 2014, NRC ADAMS Accession No. ML14111A147.

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

Seismic. Electric Power Research Institute," February 2013.

49. 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, NRC ADAMS Accession No. ML13101A379.

50. NRC (E Leeds) Letter to NEI (J Pollock), "Electric Power Research Institute Final Draft Report xxxxx, "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, NRC ADAMS Accession No. ML13106A331.

Page 27

Palisades Nuclear Plant ESEP Report

51. Entergy Calculation EA-EC46465-05, "Anchorage Calculations for Electrical Panels EP-1901 and EP-2001," Revision 0.

52. Entergy Document, EA-EC48188-02, "Seismic Bracing of Non-Q Block Wall - South Wall of Lube Oil Storage Room," Revision 0.

53. Entergy Document EC54011, "Fukushima Seismic 2.1 - Expedited Seismic Evaluation Program (ESEP) Report - Owner Acceptance of Report and Background Documentation," the following AREVA documents are captured in the plant document management system:

a. AREVA Document 51-9212955-007, "ESEP Expedited Seismic Equipment List (ESEL) -

Palisades Nuclear Plant."

b. AREVA Calculation 32-9227862-001, "Palisades ESEP HCLPF Calculation - Electrical Cabinet EC-02."

c. AREVA Calculation 32-9227902-001, "Palisades ESEP HCLPF Calculation - 480V Load Center, EB-11."

d. .AREVA Calculation 32-9227941-001, "Palisades ESEP HCLPF Calculation - Horizontal Pump, P-55C."

e. AREVA Calculation 32-9227961-001, "Palisades ESEP HCLPF Calculation - 125 VDC Main Station Batteries-ED-01 and ED-02."

f. AREVA Calculation 32-9228097-001, "Palisades ESEP HCLPF Calculation - 480V Motor Control Center, EB-01."

g. AREVA Calculation 32-9228681-001, "Palisades ESEP HCLPF Calculation - Auxiliary Feedwater Controls, EJ-1051."

h. AREVA Calculation 32-9229831-002, "Palisades ESEP HCLPF Calculation - Primary Makeup Storage Tank, T-81."

i. AREVA Calculation 32-9230336-001, "Palisades ESEP Screening of Lube Oil Storage Room South Block Wall."

j. AREVA Calculation 32-9230249-003, Palisades ESEP HCLPF Calculation - Boric Acid Storage Tanks T-53A & T-53B."

k. AREVA Calculation 32-9228279-002, "Palisades ESEP HCLPF Calculation - Block Wall, C-104.11Q."

1. AREVA Calculation 32-9228841-001, "Palisades ESEP HCLPF Calculation - Block Walls C-107.160, C-107.17Q, and C-107.18Q."

Page 28

Palisades Nuclear Plant ESEP Report ATTACHMENT A - PALISADES NUCLEAR PLANT ESEL Page A-1

Palisades Nuclear Plant ESEP Report 3ervILe wdler 3y5Lemii 1 CV-0861 Return Isolation Valve Closed Open from CAC Fan VHX-1 Service Water System 2 CV-0864 Return Isolation Valve Closed Open from CAC Fan VHX-2 Service Water System Deenergize to open CV-0861 3 SV-0861 Return Isolation Valve Energized De-energized Breaker 52-1208 from CAC Fan VHX-1 Service Water System Deenergize to open CV-0864 4 SV-0864 Return Isolation Valve Energized De-energized Breaker 52-1209 from CAC Fan VHX-2 5 V-1A Containment Air Cooler Available Available Fan 6 V-2A Containment Air Cooler Available Available Fan Containment Air Cooler 7 V-3A Fan Available Available Fan 8 VHX-1 Containment Air Cooler Available Available 9 VHX-2 Containment Air Cooler Available Available 10 VHX-3 Containment Air Cooler Available Available 11 CV-0522B TDAFW pump steam Closed Open inlet valve 12 CV-0727 AFW flow control valve Closed Throttled (SG B) 13 CV-0749 AFW flow control valve Closed Throttled I___ (SG A)

Page A-2

Palisades Nuclear Plant ESEP Report

-- State. - . * .. .. teS/Com ments , - ,**-'M

.Nt~nb~

14 CV-0779 ADV (SG E-501) Closed Open as Automatically or manually opened [26]

needed CV-0780 AV(G-SBClsdOpenS as is AD

-03 lsdneeded Automatically or manually opened [26]

16 CV-0781 ADV (SG E-50A) Closed Open as Automatically or manually opened [26]

needed 17 CV-0782 ADV (SG E-50A) Closed Open as Automatically or manually opened [26]

17 C-72 AV(GE5A lsdneeded

___________ needed Automatically or manually opened [26]

18 CV-2008 T-81 Isolation Valve Closed Open [25]

19 CV-2010 T-81 Isolation Valve Closed Open [25]

20 E/P-0779 ADV (SG E-50B) On On [26]

21 E/P-0780 ADV (SG E-50B) On On [26]

22 E/P-0781 ADV (SG E-50A) On On [26]

23 E/P-0782 ADV (SG E-50A) On On [26]

24 HIC-0780A Hand Indicating Auto Auto Controller AutoAuto_[26]

25 HIC-0780B Hand Indicating Auto Auto [26]

Controller 26 HIC-0781B Hand Indicating Auto Auto [26]

Controller 27 P-8B TDAFW pump Standby Operating [27]

Page A-3

Palisades Nuclear Plant ESEP Report Eett"Lr-p*

l n 0erati c- Rf iSjtate.Ne..c.

28 OC-779 ADV(SGE-5B) Modulate as Modulate as Controls for CV-0781 shown on reference [6 28_PO_-0779_DV __SGE-50B necessary necessary Oth~ers typical of that arrangement [26]____

Modulate as Modulate as Controls for CV-0781 shown on reference

_________________ necessary necessary Others typical of that arrangement______

28 POC-0779 ADV (SG E-5013) Modulate as Modulate as 1[261 30 POC-0781 ADV (SG E-50A) ncsayeesry[26]

Modulate as Modulate.as Controls for CV-0781 shown on reference necessary necessary Others typical of that arrangement 3 SV02B TDAFW pump steam De-energized Energized [23]

32SV0522Binlet valve 33 POC-0780 ADV (SG E-5013) Energized 3nergized Energzlated Energized as Controls for CV-0781 shown on reference Others typical of that arrangement

[26]

[26]

30 SV-0779B ADV (SG E-50B) De-energized De-energized Controlstypical Others for CV-0781 of that shown arrangement on reference [6

[26]

31 POC-0782 ADV (SG E-50B) Mode-energeas neenery oDe tenerizd neenery Controls for CV-0781 shown on reference Others typical of that arrangement

[26]

[26]

36 SV-0779A ADV (SG E-5013) Energized Energized Controls for CV-0781 shown on reference Others typical of that arrangement

[26]

[26]

3 Controls for CV-0781 shown on reference [26]

7SV-0780B ADV (SG E-50B) De-energized De-energized Others typical of that arrangement 38 SV-0779C ADV (SG E-5013) De-energized De-energize'd Controls for CV-0781 Others typical of that shown on reference arrangement [26]

Energized to close. Same for all ADV SV "A" 39 SV-0781A ADV (SG E-50A) Energized Energized valves. When closed, ADVs can be operated [261 Cofrom controller.

Deenergized to open. Same for all ADV SV "B" 40 SV-0781B ADV (SG E-50A) De-energized De-energized valves. When open, ADVs can be operated [26]

from controller.

Page A-4

Palisades Nuclear Plant ESEP Report Eq wmn 'Ww rtn1~~t4 .. ... IN,..

~SEgliem z ORfrne--- -.

_____ rormalýStaife ýeedStae- -*

Deenergized to close. Same for all ADV SV "C" 41 SV-0781C ADV (SG E-SOA) De-energized De-energized valves. When closed, ADVs can be operated [26]

from controller.

42 SV-0782A ADV (SG E-50A) Energized Energized Controls for CV-0781 shown on reference [26]

42_SV-0782A DS 0e dd Others typical of that arrangement [26]

43 SV-0782B ADV (SG E-50A) De-energized De-energized Controls for CV-0781 shown on reference [26]

43Deenrze SV072D -5) De-energ Others typical of that arrangement [26]

44 SV-0782C ADV (SG E-50A) De-energized De-energized Controls for CV-0781 shown on reference [26]

Others typical of that arrangement 45 SV-2008 T-81 Isolation Valve De-energized Energized [25]

46 SV-2010 T-81 Isolation Valve De-energized Energized [25]

Primary System Make-up New FLEX valve. Required closed to prevent 47 SV-2235 Transfer Pump Isolation Energized De-energized inventory diversion.

Valve (new) Not yet installed.

Primary System Make-up New FLEX valve. Required closed to prevent 48 SV-2236 Transfer Pump Bypass Energized De-energized inventory diversion.

Isolation Valve (new) Not yet installed.

T-81 and CST combined volume > 100k gal Condensate Storage (Phase 1 - 8+ hrs); Long term water source 49 T-2 Tank Available Available (Phase 2) is Lake Michigan directly into steam [25]

generator via AFW or MFW header using a portable FLEX pump.

T-81 and CST combined volume > 100k gal (Phase 1 - 8+ hrs); Long term water source 50 T-81 Primary Makeup Storage Available Available (Phase 2) is Lake Michigan directly into steam [25]

Tank generator via AFW or MFW header using a portable FLEX pump. Screening HCLPF of 0.1 used for IPEEE.

Page A-5

Palisades Nuclear Plant ESEP Report U2 -6 Aftfl EEmltem A_ __ - - -

Ownom

le-ti ~Ddgire! State-51 CV-2083 CBO Isolation Valve Open

_____________________________ass~umption Closed Bleedoff must be isolated to support of PCP leakage of 1 gpm/pump

[8

[8 52 CV-2099 CBO Isolation Valve Open Closed Bleedoff must be isolated to support (8

__________________assumption of PCP leakage of 1 gpm/pump [8 53 CBORelif C-219 Stp Vave Oen Cosed Bleedoff must be isolated to support [8 53 CBORelif C-219 Stp Vave Oen Cosed assumption of PCP leakage of 1 gpm/pump [8 5E-7 Condensate Storage NA/A Listed per EPRI Q&A 3.20 to ensure heat [5 54 E-27 Tank Heat Exchanger NA/A exchanger pressure boundary integrity [5 55E-56 Regenerative Heat N/A N/A Listed per EPRI Q&A 3.20 to ensure heat [20]

Exchanger exchanger pressure boundary integrity 56 MO-2087 VCT Supply to Charging Open Closed- VCT volume is not credited and must be [19]

Pumps isolated in Phase 2 HPSI Injection Loop 57 MO-3007 Isolation Valve Closed Open [22]

58 MO-3041 SIT T-82A Isolation Valve Open Closed Closure to prevent N2 injection; associated [21]

____________relay in A-46 program 59SITT-82 M-304 Isoatio Vale Opn Clsed Closure to prevent N2 injection; associated [1 M-304 Isoatio Vale 59SITT-82 Opn Clsed relay in A-46 program [1 60 MO-3049 SIT T-82C Isolation Valve Open Closed. Closure to prevent N2 injection; associated [1 relay in A-46 program [1 61 MO-3052 SIT T-82D Isolation Valve Open Closed Closure to prevent relay in A-46 N2 injection; associated program [21]

62 P-55C Charging Pump P-55C Off On Deviation taken from NEI 12-06 Sect 3.2.2 (12); [20]

(Phase 2) 63 T-53A Boric Acid Storage Tank Available Available' [19]

T-53A 64 T-53B Boric Acid Storage Tank Available Available [19]

T-53B Page A-6

Palisades Nuclear Plant ESEP Report

- ntjnnnts-65 EA-11 2400V Bus 1C Available Available [7]

66 EB-01 480 V MCC 1 Available Available [7]

67 EB-11 480V Bus LCC-11 Available Available Bus supplies charging pump P55C power [7]

68 EB-19 480V Bus LCC-19 Available Available Powers MCC-1 which feeds BC 1 & 4, LCC-11, & [8]

emergency lighting panel L-04A 69 EB-21 480 V MCC 21 Available Available- Support SIT Isolation [33]

70 EB-23 480 V MCC 23 Available Available Support SIT Isolation [33]

71 EC-01 Electrical Cabinet Available Available Control room panel 72 EC-02 Electrical Cabinet Available Available Control room panel 73 EC-11 Electrical Cabinet Available Available Control room panel 74 EC-12 Electrical Cabinet Available Available Control room panel 75 EC-13 Electrical Cabinet Available Available Control room panel 76 EC-182 ADV Panel Available Available 77 EC-33 Electrical Cabinet Available Available Not required - was for valves removed from ESEL 78 ED-01 125VDC Battery #1 Available Avail-able'[9 Page A-7

Palisades Nuclear Plant ESEP Report

-C ;j faZ aatiti A",

74a7e rNWwNo,--Wtes/Comm, ent Ref

- fI W- at 79 ED-02 125VDC Battery #2 Available Available [9]

s0 ED-06 Inverter 1 Available Available [10]

81 ED-07 Inverter 2 Available Available [10]

82 ED-08 Inverter 3 Available Available [10]

83 ED-09 Inverter 4 Available Available [10]

84 ED-10OL DC Bus DIOL Available Available [9]

85 ED-10R DC Bus D1i11 Available Available [9]

86 ED-11 125VDC Panel ED-11 (-1, Available Available [0 87 ED-13 DC Bus Metering Section Available Available' [9]

88 ED-15 -Battery Charger #1 Available Available [9]

89 ED-18 Battery Charger #4 Available Available [9]

90 ED-20L DC Bus D20L Available Available [9]

91 ED-20R DC Bus D20R Available Available [9]

92 ED-21-1 125VDC Panel ED-21-1 Available Avail able [10]

Page A-8

Palisades Nuclear Plant ESEP Report

- , y ra -.....

Ing . .

Nb.i ( . Piifi

• c* * * ~ t**a~~l*-*e e . .ta.e.. .... ......o . .. *. ... : .o.*..... -hen.ts.V*.**

• bNme KMal~ M6-ps .State, 93 ED-21-2 125VDC Panel ED-21 -2 Available Available, [10]

94 ED-23 DC Bus Metering Section Available Available [9]

95 EJ-1051 Junction Box Available Available 96 EJ-569 Junction Box Available Available [30]

480V FLEX generator will connect to this new 97 EP-1901 FLEX Electrical Box Off On panel (FLEX Modification). Anchorage [51]

calculation EA-EC46465-05. Not yet installed.

98 EX-11 2400/480 V Transformer Available Available [71 99 EX-19 2400/480 V Transformer Available Available [8]

100 EY-10 Preferred 120 VAC Panel Available Available [101 1 (Y-10) 101 EY-20 Preferred 120 VAC Panel Available Available [10]

2 (Y-20) 102 EY-30 Preferred 120 VAC Panel Available Available [10]

3 (Y-30) 103 EY-40 Preferred 120 VAC Panel Available Available [10]

4 (Y-40) 104 N2 #

Backup Support AFW Flow Control Valves CV-0727, [31]

N2 Backup N2 Backup Station #1 Available Available 0749. Listed per EPRI O&A 3.22.

105 N2 #

Backup Support AFW TDAFW pump steam supply valve [31]

N2 Backup N2 Backup Station #2 Available Available CV-0522B. Listed per EPRI Q&A 3.22.

106 N2 Backup N2 Backup Station #9 Available Available Support ADVs. Listed per EPRI Q&A 3.22. FLEX [31]

Station #9 equipment not yet installed.

Page A-9

Palisades Nuclear Plant ESEP Report ESLitm-- 4 g~era~ting; tagge AFW (E-50A) Flow 107 FIC-0727 - --Indicator A-Monitor-is Controller On On [27]

108 FM-0727 AFM(-onitFow On On .. Square root extractor. [27]

109 FT-0727 AFW (E-50B) Flow On On [27]

Transmitter AFW (E-50B) Hand 110 HIC-0727 Indicator Controller Auto Auto [27]

ill I/P-0727 Signal Converter - AFW Modulate as Modulateas [27]

flow control valve (SG B) necessary necessary 112 P/S-0727 Power Supply - AFW (E- On On [37]

50B) Flow Transmitter AFW (E-50A) Flow 113 FIC-0749 Indicator Controller On On [27]

114 FM-0749 AFW (E-MoA) Flow On On Square root extractor. [27]

Monitor 115 FT-0749 AFW (E-50A) Flow On On [27]

Transmitter 116 HIC-0749 AFW (E-50A) Hand [27]

Indicator Controller Auto Auto 117 I/P-0749 Signal converter - AFW Modulate as Modulateas [27]

flow control valve (SG A) necessary necessary 118 P/S-0727A Power Supply - AFW (E- On On [37]

50A) Flow Transmitter 119 P/S-0206 Power Supply - BAST T- On On Not available until Phase 2 (powered by FLEX [14]

53B Level Indication generator) 120 LIA-0206 Boric Acid Storage Tank On 0n Not available until Phase 2 (powered by FLEX [19]

T-53B Level Indication generator)

Page A-10

Palisades Nuclear Plant ESEP Report Boric Acid Storage Tank Not available until Phase 2 121 LIT-0206 On T-53B Level Transmitter generator)

Boric Acid Storage Tank Not available until Phase 2 T-53A Level Indication generator)

LIT-0208 Boric Acid Storage Tank available until Phase 2 123 T-53A Level Transmitter erator) 124 P/S-0208 Power Supply - BAST T-53A Level Indication 125 LIA-0102A PZR Level (WR) Indicator 126 LT-0102 PZR Level (WR)

Transmitter 127 LIA-0365 SITIdcto T-82A Level Monitor to prevent N2 injection Indication 128 LM-0365 SIT T-82A Level Monitor Square root extractor 129 L-0365 SIT T-82A Level Transmitter jection 130 P/S-0365 Power Supply - SIT T-82A Level Indication jection 131 LIA-0368 SITIdcto T-82B Level Indication jection 132 LM-0368 SIT T-82B Level Monitor iuare root extractor 133 LT-0368Trnmte SIT T-82B Level jection Transmitter 134 P/S-0368 Power Supply - SIT T-82B 4 8 Level Indication Monitor to prevent N2 injection Page A-11

Palisades Nuclear Plant ESEP Report

?P _ .- J, .- et" E*E-*:tem ,

OU -

M"V- RLGM Screened Note 2 inlet valve AFW flow control valve 12 CV-0727 >RLGM Screened (SG B) 13 CV-0749 AFW flow control valve >RLGM Screened (SG A) 14 CV-0779 ADV (SG E-50B) >RLGM Screened 15 CV-0780 ADV (SG E-50B) >RLGM Screened 16 CV-0781 ADV (SG E-50A) >RLGM Screened 17 CV-0782 ADV (SG E-50A) >RLGM Screened 18 CV-2008 T-81 Isolation Valve >RLGM Screened 19 CV-2010 T-81 Isolation Valve >RLGM Screened Page B-2

Palisades Nuclear Plant ESEP Report HCLP'F()/

Equlim-' .i Equipment'escription creeig deients

______ ______ Leve'l'C m et 20 E/P-0779 ADV (SG E-50B) >RLGM Screened 21 E/P-0780 ADV (SG E-50B) >RLGM Screened 22 E/P-0781 ADV (SG E-50A) >RLGM Screened 23 E/P-0782 ADV (SG E-50A) >RLGM Screened 24 HIC-0780A Hand Indicating >RLGM Screened Controller 25 HIC-0780B Hand Indicating >RLGM Screened Controller 26 HIC-0781B Hand Indicating >RLGM Screened Controller 27 P-8B TDAFW pump >RLGM Screened Note 1 28 POC-0779 ADV (SG E-50B) >RLGM Screened 29 POC-0780 ADV (SG E-50B) >RLGM Screened 30 POC-0781 ADV (SG E-50A) >RLGM Screened 31 POC-0782 ADV (SG E-50A) >RLGM Screened 32 SV-0522B TDAFW pump steam >RLGM Screened inlet valve 33 SV-0779A ADV (SG E-50B) >RLGM Screened 34 SV-0779B ADV (SG E-50B) >RLGM Screened 35 SV-0779C ADV (SG E-50B) >RLGM Screened 36 SV-0780A ADV (SG E-50B) >RLGM Screened 37 SV-0780B ADV (SG E-50B) >RLGM Screened 38 SV-0780C ADV (SG E-50B) >RLGM Screened 39 SV-0781A ADV (SG E-50A) >RLGM Screened 40 SV-0781B ADV (SG E-50A) >RLGM Screened Page B-3

Palisades Nuclear Plant ESEP Report Item Equipmn ID Equipment t Description .Screening. FW UW:Failre, Comme.

41 SV-0781C ADV (SG E-50A) >RLGM Screened 42 SV-0782A ADV (SG E-50A) >RLGM Screened 43 SV-0782B ADV (SG E-50A) >RLGM Screened 44 SV-0782C ADV (SG E-50A) >RLGM Screened 45 SV-2008 T-81 Isolation Valve >RLGM Screened 46 SV-2010 T-81 Isolation Valve >RLGM Screened Primary System Make- Not Not New FLEX component 47 SV-2235 up Transfer Pump to be seismically Isolation Valve (new) Applicable Applicable designed.

Primary System Make-48 SV-2236 48 SV23 up Transfer Pump *Not Not to be seismically Bypass Isolation Valve Applicable Applicable designed.

(new (new)__ __

49 T-2 Condensate Storage >RLGM Screened Note 1 Tank 50 T-81 Primary Makeup 0.19 Tank Shell Modification Storage Tank Buckling Required 51 CV-2083 CBO Isolation Valve TBD TBD Note 3 52 CV-2099 CBO Isolation Valve TBD TBD Note 3 53 CV-2191 CBO Relief Stop Valve TBD TBD Note 3 54 E-27 Condensate Storage >RLGM Screened Note 2 Tank Heat Exchanger 55 E-56 Regenerative Heat TBD TBD Note 3 Exchanger 56 MO-2087 VCT Supply to Charging >RLGM Screened Pumps 57 MO-3007 HPSI Injection Loop TBD TBD Note 3 Isolation Valve 58 MO-3041 SIT T-82A Isolation TBD TBD Note 3 Valve 59 MO-3045 SIT T-82B Isolation TBD TBD Note 3 Valve Page B-4

Palisades Nuclear Plant ESEP Report Equlpmen. r Equ nDesc' plon-: Screening Commnts No.~~ LevelIod 60 MO-3049 SIT T-82C Isolation TBD TBD Note 3 Valve 61 MO-3052 SIT T-82D Isolation TBD TBD Note 3 Valve 62 P-55C Charging Pump P-55C 0.47 Anchorage Boric Acid Storage Tank HCLPF calculated with 63 T-53A T-53A 0.46 Anchorage modifications to tank supports.

Boric Acid Storage Tank HCLPF calculated with 64 T-53B T-53B 0.46 Anchorage modifications to tank supports.

65 EA-11 2400V Bus 1C >RLGM Screened Note 1 66 EB-01 480 V MCC 1 0.41 Functional 67 EB-11 480V;Bus LCC-11 0.44 Anchorage 68 EB-19 480V Bus LCC-19 >RLGM Screened Note 2 69 EB-21 480 V MCC 21 >RLGM Screened Note 1 70 EB-23 480 V MCC 23 0.40 Block Wall Note 1 71 EC-01 Electrical Cabinet >RLGM Screened Note 2 72 EC-02 Electrical Cabinet 0.50 Anchorage 73 EC-11 Electrical Cabinet >RLGM Screened Note 1 74 EC-12 Electrical Cabinet >RLGM Screened Note 1 75 EC-13 Electrical Cabinet >RLGM Screened Note 1 76 EC-182 ADV Panel >RLGM Screened Note 2 77 EC-33 Electrical Cabinet 0.42 Block Wall Note 1 78 ED-01 125VDC Battery #1 0.40 Block Wall 79 ED-02 125VDC Battery #2 0.40 Block Wall Page B-5

Palisades Nuclear Plant ESEP Report N. Equirpment 'ID' 01 dt ecrlpt on tne jee FalueComet 80 ED-06 Inverter 1 >RLGM Screened Note 1 81 ED-07 Inverter 2 >RLGM Screened Note 1 82 ED-08 Inverter 3 >RLGM Screened Note 1 83 ED-09 Inverter 4 >RLGM Screened Note 1 84 ED-10L DC Bus DIOL 0.40 Block Wall Note 1 85 ED-10R DC Bus D1OR 0.40 Block Wall Note 1 86 ED-11 125VDC Panel ED-11 >RLGM Screened 1, -2) 87 ED-13 DC Bus Metering >RLGM Screened Section 88 ED-15 Battery Charger #1 >RLGM Screened Note 1 89 ED-18 Battery Charger #4 >RLGM Screened Note 1 90 ED-20L DC Bus D20L 0.40 Block Wall Note 1 91 ED-20R. DC Bus D20R 0.40 Block Wall Note 1 92 ED-21-1 125VDC Panel ED-21-1 >RLGM Screened 93 ED-21-2 125VDC Panel ED-21 -2 >RLGM Screened 94 ED-23 DC Bus Metering >RLGM Screened Section 95 EJ-1051 Junction Box 0.46 Anchorage 96 EJ-569 Junction Box >RLGM Screened Note 2 97 EP-1901 FLEX Electrical Box >RLGM Screened Note 4 98 EX-11 2400/480V >RLGM Screened Note 1 Transformer 2400/480 V 99 EX-19 Trasfome Transformer >RLGM Screened Note 2 100 EY-10 Preferred 120 VAC Panel (Y-l0) I

>RLGM I

Screened II Page B-6

Palisades Nuclear Plant ESEP Report Itemi. .. "n* * " .. . .:" ' .. Faiiureý-" . */ , , .:- , ,

Item. Equipment: I ED-IE1ent" D riMode Comments

.No. ...- "* .. :' "" ' "  ; : ... Leb ,el o__e...._ , ___ __..__ ___ __"

101 EY-20 Preferred 120 VAC >RLGM Screened Panel 2 (Y-20) 102 EY-30 Preferred 120 VAC >RLGM Screened Panel 3 (Y-30) 103 EY-40 Preferred 120 VAC >RLGM Screened Panel 4 (Y-40) 104 N2 Backup N2 Backup Station #1 >RLGM Screened Note 2 Station #1 105 N2 Backup N2 Backup Station #2 >RLGM Block Wall Note 5 Station #2 N2 Backup Not Not New FLEX component 106 u N2 Backup Station #9 to be seismically Station #9 Applicable Applicabledein. designed.

AFW (E-50A) Flow 107 .FIC-0727 ato indicator Control "Controller >RLGM Screened Note 2 108 FM-0727 AFW (E-50B) Flow >RLGM Screened Monitor I 109 FT-0727 AFW (E-50B) Flow >RLGM Screened Transmitter AFW (E-50B) Hand 110 HIC-0727 Indicator Controller >RLGM Screened Signal Converter - AFW 111 I/P-0727 flow control valve (SG >RLGM Screened Note 2 B) 112 P/S-0727 Power Supply - AFW (E- >RLGM Screened 50B) Flow Transmitter 113 FIC-0749 AFW (E-50A) Flow >RLGM Screened Indicator Controller 114 FM-0749 AFW (E-50A) Flow >RLGM Screened Monitor 115 FT-0749 AFW (E-50A) Flow >RLGM Screened Transmitter AFW (E-50A) Hand 116 HIC-0749 ato Indicator Cont Controller >RLGM Screened Signal converter - AFW 117 I/P-0749 flow control valve (SG >RLGM Screened Note 2 A) 118 P/S-0727A Power Supply - AFW (E- >RLGM Screened 50A) Flow Transmitter 119 P/S-0206 Power Supply - BAST T- >RLGM Screened 53B Level Indication Page B-7

Palisades Nuclear Plant ESEP Report Iteqi e ..  : .... (g . .Failure

'r' --scri Epment u"" 'D pt on, reeing M 120 LIA-0206 Boric Acid Storage Tank >RLGM Screened T-53B Level Indication 121 LIT-0206 Boric Acid Storage Tank >RLGM Screened IT T-53B Level Transmitter 122 LIA-0208 Boric Acid Storage Tank >RLGM Screened T-53A Level Indication 123 LIT-0208 Boric Acid Storage Tank >RLGM Screened Note 2 T-53A Level Transmitter 124 P/S-0208 Power Supply - BAST T- >RLGM Screened 53A Level Indication 125 LIA-0102A PZR Level (WR) >RLGM Screened Indicator 126 LT-0102 PZR Level (WR) TBD TBD Note 3 Transmitter 127 LIA-0365 SIT T-82A Le6el >RLGM Screened Indication 128 LM-0365 SIT T-82A Level Monitor >RLGM Screened 129 LT-0365 SIT T-82A Level TBD TBD Note 3 Transmitter 130 P/S-0365 Power Supply - SIT T- >RLGM Screened 82A Level Indication 131 LIA-0368 SIT T-82B Level >RLGM Screened Indication 132 LM-0368 SIT T-82B Level Monitor >RLGM Screened 133 LT-0368 SIT T-82B Level TBD TBD Note 3 Transmitter 134 P/S-0368 Power Supply - SIT T- >RLGM Screened 82B Level Indication 135 LIA-0372 SIT T-82C Leve >RLGM Screened Indication 136 LM-0372 SIT T-82C Level Monitor >RLGM Screened 137 LT-0372 SITT-82C Level >RLGM Screened Transmitter 138 P/S-0372 Power Supply - SIT T- >RLGM Screened 82C Level Indication 139 LIA-0374 SIT T-82D Leve >RLGM Screened Indication 140 LM-0374 SIT T-82D Level Monitor >RLGM Screened Page B-8

Palisades Nuclear Plant ESEP Report Equipment ID Item~. 'Eqic ~ Failure.

p opment S.rning Descript Mde . Cmments No. Meoel 141 LT-0374 SIT T-82D Leve TBD TBD Note 3 Transmitter 142 P/S-0374 Power Supply - SITT- >RLGM Screened 82D Level Indication 143 P/S-0751A Power Supply - SG A NR >RLGM Screened Level Indication 144 P/S-0751C Power Supply - SG A NR >RLGM Screened Level Indication 145 LI-0757A SG A WR Level >RLGM Screened Indication SG A WR Level 146 LT-0757A TBD TBD Note 3 Transmitter 147 P/S-0757A Power Supply - SG A WR >RLGM Screened Level Transmitter 148 LI-0758A SG B WRLevel >RLGM Screened Indication "SG B WR.Level 149 LT-0758A

______Transmitter SGB Wmiee TBD TBD Note 3 150 P/S-0758A Power Supply - SG B >RLGM Screened WR Level Transmitter 151 LIA-2021 CST (T-2) Level >RLGM Screened Indication 152 LT-2021 CST (T-2) Level >RLGM Screened Transmitter 153 P/S-2021 Power Supply - CST (T-2) >RLGM Screened I_ Level Indication 154 PI-0105A PZR Pressure (WR) >RLGM Screened Indicator 155 PT-0105A PZR Pressure (WR) TBD TBD Note 3 Transmitter Wide Range Primary 156 PTR-0112 Temperature & >RLGM Screened Pressure Recorder 157 PIC-0751C SG A Pressure Indicator >RLGM Screened Controller 158 PT-0751C SG A Pressure TBD TBD Note 3 Transmitter 159 PIC-0752C SG B Pressure Indicator >RLGM Screened Controller SG B Pressure 160 PT-0752C TBD TBD Note 3 Transmitter I I I Page B-9

Palisades Nuclear Plant ESEP Report tem., P-E(g)i Failure' qupment 9teýi eo ýID Equpmn Decitin S . r SrnnCmet MOde.

Containment Sump 161 LPIR-0383 Level/Pressure >RLGM Screened Indication Power Supply -

162 P/S-1812A Containment Pressure >RLGM Screened Transmitter 163 PT-1812A Containment Pressure >RLGM Screened Note 2 Transmitter 164 TE-0112CC PCS T-cold Temperature TBD TBD Note 3 Element 165 T1-0112CC PCS T-cold Temperature >RLGM Screened Transmitter 166 TE-0112HC PCS T-hot Temperature TBD TBD Note 3 Element 167 T1"-Ol12H PCS T-hot Temperature >RLGM Screened TT-O112HC 167 Transmitter _____crene 168 LE-OIOA Reactor Vessel Level TBD TBD Note 3 Reactor Vessel Level 169 LRI-O101A1 Inidicator Recorder - >RLGM Screened Head Region Reactor Vessel Level 170 LRI-01OIA2 Inidicator Recorder - >RLGM Screened Upper Guide Structure 171 EC-l1A Post Accident Panel >RLGM Screened Note 2 172 P/S-0101AA Power Supply for LE- >RLGM Screened 0101A 173 T-82A Safety Injection Tank A TBD TBD Note 3 174 T-82B Safety Injection Tank B TBD TBD Note 3 175 T-82C Safety Injection Tank C TBD TBD Note 3 176 T-82D Safety Injection Tank D TBD TBD Note 3 177 CV-0599 TDAFW Trip/Throttle >RLGM Screened Valve 178 CV-0598 TDAFW Governor Valve >RLGM Screened Page B-10

Palisades Nuclear Plant ESEP Report Notes:

1. Anchorage screened out based on available margin during walkdown by SRT.

2. Anchorage screened out during walkdown validation by SRT.

3. Inaccessible. Per EPRI NP-6041-SLR1, Sec. 2, Seismic Capability Walkdown, Step 5 - This component was not walked down.

4. The design of anchorage for component EP-1901 is performed for seismic load of 1xSSE in Entergy Calculation EA-EC46465-05 [51]. The maximum interaction ratio for anchor bolts is 0.23. This ratio is less than 0.5, thus margin exists such that the anchorage of panel EP-1901 is adequate for 2xSSE.

5. Lube Oil Storage Room South Block Wall is reinforced for FLEX [52].

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