VR0WEntergy Eiy Entergy Nuclear Operations, Inc.

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 Pursuant to Title 10 of the

Code of Federal Regulations 50.54(f) Regarding

Recommendations 2.1, 2.3, and 9.3 of the Near-Term Task Force

Review of Insights from the Fukushima Dai-ichi Accident, dated

March 12, 2012 (Adams Accession No. ML12056A046).

2. NEI letter, Proposed Path Forward for NTTF Recommendation 2. 1:

Seismic Reevaluations, dated April 9, 2013 (ADAMS Accession

No. ML13101A379).

3. NRC letter, Electric Power Research Institute Report XXXXXX,

"Seismic Evaluation Guidance: Augmented Approach for the

Resolution of Fukushima Near-Term Task Force Recommendation

2.1: Seismic," as an Acceptable Alternative to the March 12, 2012,

Information Request for Seismic Reevaluations, 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
A

AREVA

20004-021 (01/30/2014)

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

AREVA

20004-021 (0113012014)

Document No.: -0923`1248-002

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 NOM 

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

AREVA

20004-021 (01/30/2014)

Document No.: 51-9231248-002

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

AREVA

20004-021 (01/30/2014)

Document No.: 51-9231248-002

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
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AREVA Document No.: 51-9231248-002

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
AR AEVA 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 SixMonth 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."

Page 6
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AR EVA Document No.: 51-9231248-002

Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant

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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AR AEVA 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 1 B, 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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AR EVA Document No.: 51-9231248-002

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, C104.11Q."

I. AREVA Calculation 32-9228841-001, "Palisades ESEP HCLPF Calculation - Block Walls C107.16Q, 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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Expedited Seismic Evaluation Process (ESEP) Report for Palisades Nuclear Plant

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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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

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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 R EFER EN C ES .................................................................................................................................. 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 cooldowndepressurization 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.

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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.

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- 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

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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

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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

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Palisades Nuclear Plant ESEP ReportGMRS 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

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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 narrowbanded 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.

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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

Fmquency (Wz)

100.000

Figure 5-1: RLGM for Palisades Nuclear Plant

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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.

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Palisades Nuclear Plant ESEP Report

The ESEL components were prescreened based on Table 2-4 of EPRI NP-6041-SL. Additional prescreening, 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 NP6041-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-radioactive or low radioactive environments. Seismic capability assessment of

components which are inaccessible, in high-radioactive environments, or possibly within contaminated

containment, will have to rely more on alternate means such as photographic inspection, more reliance

on seismic reanalysis, and possibly, smaller inspection teams and more hurried inspections. A 100%

"walk by" does not mean complete inspection of each component, nor does it mean requiring an

electrician or other technician to de-energize and open cabinets or panels for detailed inspection of all

components. This walkdown is not intended to be a QA or QC review or a review of the adequacy of the

component at the SSE level.

If the SRT has a reasonable basis for assuming that the group of components are similar and are

similarly anchored, then it is only necessary to inspect one component out of this group. The

"similarity-basis" should be developed before the walkdown during the seismic capability preparatory

work (Step 3) by reference to drawings, calculations or specifications. The one component or each type

which is selected should be thoroughly inspected which probably does mean de-energizing and opening

cabinets or panels for this very limited sample. Generally, a spare representative component can be

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Palisades Nuclear Plant ESEP Report

found so as to enable the inspection to be performed while the plant is in operation. At least for the

one component of each type which is selected, anchorage should be thoroughly inspected.

The walkdown procedure should be performed in an ad hoc manner. For each class of components the

SRT should look closely at the first items and compare the field configurations with the construction

drawings and/or specifications. If a one-to-one correspondence is found, then subsequent items do not

have to be inspected in as great a detail. Ultimately the walkdown becomes a "walk by" of the

component class as the SRT becomes confident that the construction pattern is typical. This procedure

for inspection should be repeated for each component class; although, during the actual walkdown the

SRT may be inspecting several classes of components in parallel. If serious exceptions to the drawings

or questionable construction practices are found then the system or component class must be inspected

in closer detail until the systematic deficiency is defined.

The 100% "walk by" is to look for outliers, lack of similarity, anchorage which is different from that

shown on drawings or prescribed in criteria for that component, potential SI [Seismic Interaction ]

problems, situations that are at odds with the team members' past experience, and any other areas of

serious seismic concern. If any such concerns surface, then the limited sample size of one component of

each type for thorough inspection will have to be increased. The increase in sample size which should

be inspected will depend upon the number of outliers and different anchorages, etc., which are

observed. It is up to the SRT to ultimately select the sample size since they are the ones who are

responsible for the seismic adequacy of all elements which they screen from the margin review.

Appendix 6 gives guidance for 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

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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

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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.

Act~ons ~ Equl~mnt # ID DesntioH

Perform seismic walkdowns, No later than the

generate HCLPF calculations end of the second

and design and implement planned refueling

1 N/A N/A any necessary modifications outage after

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 SixMonth 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 SixMonth 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, C104.11Q."

1. AREVA Calculation 32-9228841-001, "Palisades ESEP HCLPF Calculation - Block Walls C107.160, 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

1 CV-0861

3ervILe wdler 3y5Lemii

Return Isolation Valve

from CAC Fan VHX-1

Closed Open

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

.Nt~nb~ -- State. - . • .. .. teS/Com m ents , - ,••-'M

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

is CV-0780 AV(G-SBClsdOpen as AD -03S 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]

___________ needed 17 C-72 AV(GE5A lsdneeded 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 Controller [26]

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.N e..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______

Modulate as Modulate as

30 POC-0781 ADV (SG E-50A) ncsayeesry[26]

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

28 POC-0779 ADV (SG E-5013) 1[261

necessary necessary Others typical of that arrangement

3 SV02B TDAFW pump steam De-energized Energized [23] 32 SV0522Binlet valve

33 POC-0780 ADV (SG E-5013) Energized Energzlated as Controls for CV-0781 shown on reference [26] 3nergized Energized Others typical of that arrangement [26]

30 SV-0779B ADV (SG E-50B) De-energized De-energized Controls Others typical for CV-0781 of that shown arrangement on reference [6[26]

31 POC-0782 ADV (SG E-50B) Mode-energeas oDe tenerizd Controls for CV-0781 shown on reference [26] neenery neenery Others typical of that arrangement [26]

36 SV-0779A ADV (SG E-5013) Energized Energized Controls for CV-0781 shown on reference [26] Others typical of that arrangement [26]

3 Controls for CV-0781 shown on reference [26] 7 SV-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 shown on reference [26] Others typical of that arrangement

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--- -.

_____ rormal ýStaife 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] 43 Deenrze 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! State51 CV-2083 CBO Isolation Valve Open Closed Bleedoff must be isolated to support [8 ________ _____________________________ass~umption of PCP leakage of 1 gpm/pump [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

59 SITT-82 M304 Isoatio Vale Opn Clsed Closure to prevent N2 injection; associated [1 59 SITT-82 M-304 Isoatio Vale 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 N2 injection; associated [21] relay in A-46 program

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

- ntjnnnts65 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 K Mal~ 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 Controller On On [27]

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

- -- A-Monitor-is

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

121 LIT-0206 Boric Acid Storage Tank

T-53B Level Transmitter On Not available until Phase 2

generator)

Boric Acid Storage Tank T-53A Level Indication

123 LIT-0208 Boric Acid Storage Tank

T-53A Level Transmitter

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 SIT T-82A Level Idcto

Indication

128 LM-0365 SIT T-82A Level Monitor

129 L-0365 SIT T-82A Level

Transmitter

130 P/S-0365 Power Supply - SIT T-82A Level Indication

SIT T-82B Level 131 LIA-0368 Idcto

Indication

132 LM-0368 SIT T-82B Level Monitor

133 LT-0368Trnmte SIT T-82B Level

Transmitter

134 P/S-0368 Power Supply - SIT T-82B 4 8 Level Indication

Not available until Phase 2

generator)

available until Phase 2

erator)

Monitor to prevent N2 injection

Square root extractor

jection

jection

jection

iuare root extractor

jection

Monitor to prevent N2 injection

Page A-11
Palisades Nuclear Plant ESEP Report

?P _ .- J, .- et"

E•E-•:tem , : '-•-_________ :;•,- •: - ••.. -. .otes~o mi • :-t..s•;-•.. .f: er~ees• OU M"V- - <Q - -fgSIT T-8C Leve

135 LIA-0372 SITnT-8cato LvlOn On Monitor to prevent N2 injection [21] Indication Ste

136 LM-0372 SIT T-82C Level Monitor On On Square root extractor [21]

137 LT-0372 SIT T-82C Level On On Monitor to prevent N2 injection [21]

Transmitter

138 P/-0372 Power Supply - SIT T-82C

138 PIa-0372 P u IT On On Monitor to prevent N2 injection [15]

139 LIA-0374 SIT T-82D Level On On Monitor to prevent N2 injection [21] Indication

140 LM-0374 SIT T-82D Level Monitor On On Square root extractor (21]

141 LT-0374 SITT-82D Transmitter Level On On Monitor to prevent N2 injection [21]

142 P/S-0374 Power Supply - SIT T-82D On On Monitor to prevent N2 injection [15]

Level Indication

SG E-50A&B Pressure

143 P/S-0751A Level Channel A Power On On Power supply for LT-0102. [13]

Supply

SG E-50A&B Pressure

144 P/S-0751C Level Channel A Power On On

Supply

145 LI-0757A SG A WR Level, Indication On On 2 WR/SG [26]

146 LT-0757A SG A WR Level On On 2 WR/SG [26] 6 LTransmitter

147 P/S-0757A Power Supply- SG A WR On On [11] Level Transmitter

Page A-12
Palisades Nuclear Plant ESEP Report

fESEL~tMMOM A & .... . . .gences .NtsC en. . Re. 

vpto.• -D i , •D esi~ed•Stat~e. .. -..-.. .• - " . - .-.-..-. . ,.: ..,.,,. .. q.

148 LI-0758A SG B WR Level Indication On On 2 WR/SG [26]

149 LT-0758A SG B WR Level [26] Transmitter On On 2 WR/SG

150 P/S-0758A Power Supply - SG B WR On On [12] Level Transmitter

151 LIA-2021 CST (T-2) Level On On. [25] Indication

152 LT-2021 CST (T-2) Level On On [25] Transmitter

Power Supply - CST (T-2) On On [36] 153 P/S-2021 Level Indication

154 PI-0105A PZR Pressure (WR) On On [17] Indicator

155 PT-0105A PZR Pressure (WR) On On [17]

Transmitter

Wide Range Primary

156 PTR-0112 Temperature & Pressure On On For PT-0105A, TE-0112CC and TE-O112HC [35]

Recorder

5SG A Pressure Indicator 157 PIC-0751C Controller On On 2 WR/SG [261

158 PT-0751C SG A Pressure On On 2WR/SG [26] - 1 Transmitter

5SG B Pressure Indicator 159 PIC-0752C Controller On On 2 WR/SG [26]

160 PT-0752C SG B Pressure On On 2 WR/SG [26] Transmitter

Page A-13
Palisades Nuclear Plant ESEP Report

Containment Sump

161 LPIR-0383 Level/Pressure On On [34]

Indication

Power Supply -

162 P/S-1812A Containment Pressure On On Power supply for PT-1812A [34]

Transmitter

Containment Pressure 161 LPIR-083 sTransmitter On On [34]

PCS T-cold Temperature 162 TE-0112C ontainElement On On [34]

165 PT-18Oh12AC PCS T-cold Temperature On On [1 Transmitter

166 TTE- PCS T-hot Temperature On On [16] O112HCElement

16 0112HC Element O n[6

167 TT- PCS T-hot Temperature On On [16] 0112HC Transmitter

168 LE-0101A Reactor vessel level Available Available For Modes 5 and 6 [38]

LRI- Reactor vessel level

169 0101A1 indicator recorder- Available Available -For Modes 5 and 6 [38] Head Region

LRI- Reactor vessel level

170 indicator recorder - Available Available For Modes 5 and 6 [38] 0101A2 Upper Guide Structure

171 EC-11A Post Accident Panel Available Available For Modes 5 and 6 [38]

172 P/S- Power supply for LE- Available Available For Modes 5 and 6 [38] OOIAA OOIA

173 T-82A Safety Injection Tank A Available Available' [21]

Page A-14
Palisades Nuclear Plant ESEP Report

A" T. . . . - " - "

'Number ~ ' i~cri~i . 2;~h.Dje-ie~tt

174 T-82B Safety Injection Tank B Available Available [21]

175 T-82C Safety Injection Tank C Available Available. [21]

176 T-82D Safety Injection Tank D Available Available [21]

177 C-0599 TDAFW Trip/throttle 17 C-99valve Open Open [23]

178 CV-0598 TDAFW Governor valve Open Cycle [23]

Page A-o5
Palisades Nuclear Plant ESEP Report

ATTACHMENT B - ESEP HCLPF VALUES AND FAILURE MODES TABULATION

Page B-1
Palisades Nuclear Plant ESEP Report

Itm Equipmen ID E ip"'g) D- tComments Faiur ~ • ..Equiment~lD :: qi':gu~s ent Descrip~idn;••, Sil r; hi~ngi.• .. .•;.• . ::.:;••.•

No. Moe! - -Level

Service Water System

1 CV-0861 Return Isolation Valve TBD TBD Note 3

from CAC Fan VHX-1

Service Water System

2 CV-0864 Return Isolation Valve TBD TBD Note 3

from CAC Fan VHX-2

Service Water System

3 SV-0861 Return Isolation Valve TBD TBD Note 3

from CAC Fan VHX-1

Service Water System

4 SV-0864 Return Isolation Valve TBD TBD Note 3

from CAC Fan VHX-2

5 V-1A Containment Air Cooler TBD TBD Note 3

V-lA Fan 6V-2A Cotimn i olr TBD TBD Note 3 Fan Containment Air Cooler 6 V-2A TBD TBD Note 3

7 . -A Containment Air Cooler TBD TBD Nlote 3

Fan

8 VHX-1 Containment Air Cooler TBD TBD Note 3

9 VHX-2 Containment Air Cooler TBD TBD Note 3

10 VHX-3 Containment Air Cooler TBD TBD Note 3

11 CV-0522B TDAFW pump steam >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 FW UW:Failre, . . Equipmn ID Equipment t Description .Screening. 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 Make48 48 SV-2236 SV23 up Transfer Pump •Not Not to be seismically

Bypass Isolation Valve Applicable Applicable (new designed. (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 tne dt ecrlpt on 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 >RLGM Screened

Panel _ (Y-l0) I I 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 Control >RLGM Screened Note 2 indicator "Controller

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 Indicator ato 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

Ite qi e .. : .... (g . .Failure

'r' Epment ''--scri 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

Item~. ~ Failure. Equipment ID 'Eqic Descript p opment S.rning 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 ______Transmitter LT-0758A 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

P-E(g)i tem., Failure' 9teýi qupment ýID Equpmn Decitin SrnnCmet eo S . r 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

167 TT-O112HC 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