ML15008A086: Difference between revisions

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#REDIRECT [[NL-14-152, Entergy'S 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...]]
| number = ML15008A086
| issue date = 12/22/2014
| title = Entergy'S 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...
| author name = Coyle L
| author affiliation = Entergy Nuclear Northeast
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000247, 05000286
| license number = DPR-026, DPR-064
| contact person =
| case reference number = NL-14-152
| document type = Letter, Report, Miscellaneous
| page count = 146
| project =
| stage = Response to RAI
}}
 
=Text=
{{#Wiki_filter:w Entergy,---
Enterav Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-024 Tel 914 254 6700 Lawrence Coyle Site Vice President NL-14-152 December 22, 2014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk 11545 Rockville Pike, TWFN-2F1 Rockville, MD 20852-2738
 
==SUBJECT:==
 
==REFERENCES:==
 
Entergy's 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 Indian Point Unit Numbers 2 and 3 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and 64 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 (Accession No. ML12053A340)
: 2. NEI Letter, Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations, dated April 9, 2013, (Accession No.ML13101A345)
: 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, (Accession No. ML13106A331) 4, Electric Power Research Institute Final Report 3002000704, "Seismic Evaluation Guidance:
Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," dated May 2013
 
==Dear Sir / 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.
NL-14-152 Docket Nos. 50-247 and 50-286 Page 2 of 2 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 proposedthat 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 attached Expedited Seismic Evaluation Process Report for Indian Point Units 2 and 3 provides the information described in Section 7 of Reference 4 in accordance with the schedule identified in Reference 2.This letter contains regulatory commitments in the Attachment.
If you have any questions regarding this report, please contact Mr. Robert Walpole, Manager, Regulatory Assurance at (914)254-6710.I declare under penalty of perjury that the foregoing is true and correct. Executed on December L42HAI)/,4 e /A l4?
 
==Enclosures:==
 
1 Expedited Seismic Evaluation Process Report for Indian Point Unit 2 2 Expedited Seismic Evaluation Process Report for Indian Point Unit 3
 
==Attachment:==
 
List of Regulatory Commitments cc: Mr. Douglas Pickett, Senior Project Manager, NRC NRR DORL Mr. John Boska, Senior Project Manager, NRC NRR DJLL Mr. Daniel H. Dorman, Regional Administrator, NRC Region 1 NRC Resident Inspector Mr. John B. Rhodes, President and CEO, NYSERDA Ms. Bridget Frymire, New York State Dept. of Public Service ENCLOSURE 1 TO NL-14-152 EXPEDITED SEISMIC EVALUATION PROCESS REPORT FOR INDIAN POINT UNIT 2 ENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 DOCKET NOS. 50-247 AND 50-286 ATTACHMENT
 
===9.1 SHEET===
1 OF 2 ENGINEERING REPORT COVER SHEET & INSTRUCTIONS EN-DC-147 REV 6 or u~Y~2:~:~.r1n K A 20004-021 (01/30/2014)
AREVA AREVA Inc.Engineering Information Record Document No.: 51 -9225674 -001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 2 Page 1 of 68 A AREVA-20004-021 (01/102014).ocurnent No.:. ai-022.r674-001 Expedited aelsmid-Evaluati nPrmcess (ESEP) Report fbr Indian Point UWni.2.SdafelrRsatd?
El' S 0 I~NO Does lhisdocument astablils delsign or tealmisat reqLremiontaT EI-YBS O ~NO Do~es this doctument contlai assumptons requiing vqrlflcation?
L]"ES NNO Does this documejitco~nfetiCustorer Reqpird Format? 19 IYES. LINO0 Signature Baock NOWr -P/L;"sgnatg rep~r (P).Ldad PrepizektX)
FAROS deignates Renviewer (K)I Lead-Rey!awor (LR)A-c1RF designates ProjecatManager AppiovetafCustomer-RequfredFormat (A-CRMF)A designeatsAlppcverakTM-.VericationitfRaviewerThde'pendence.
'Page 2 A AREVA 20004-021 (01/30/2014)
Document No.: 51-9225674-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 2 Signature Block (continued)
Project Manager Approval of Customer References (N/A if not applicable)
Name Title (printed or typed) (printed or typed) Signature Date Mike Terrell Project Manager TO te Fr JTnf" U1 Page 3 or Im'A AREVA 20004-021 (01/30/2014)
Document No.: 51-9225674-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 2 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* References were added due to additional components based on new Appendix A -Section 2.0, revision of supporting document.3.0, 3.1, 3.1.3, 4.2, 5.1, 6.1, Appendix A 6.2, 6.3.3, 6.4, 6.6, 7.1, 7.2,
* Section 2.0, 3.0, 3.1, 3.1.3, 4.2, 5.1, 6.1, 6.2, 6.3.3, 6.4, 6.6, 7.1, 8.1, 8.2, 8.4, 9.0, Attachment 7.2, 8.1, 8.2, 8.4, and 9.0 were modified to incorporate Entergy A, and Attachment B Comments [91] on Revision 0 of the document and updated with additional components based on new revision of supporting document.* Attachment A modified to incorporate Entergy Comments [91] and to update with additional components based on new revision of supporting document.* Attachment B -modified to incorporate Entergy Comments [91]and to update with additional components based on new revision of supporting document.Page 4 A AR EVA Document No.: 51-9225674-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 2 Table of Contents Page S IG N A T U R E B LO C K ................................................................................................................................
2 R E C O R D O F R E V IS IO N ..........................................................................................................................
4 1.0 D O C U M E N T A T IO N ......................................................................................................................
6 2 .0 R E F E R E N C E S ..............................................................................................................................
6 APPENDIX A: EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT FOR INDIAN POINT UNIT 2 .......................................................................................
A-I Page 5 ro.A AR EVA Document No.: 51-9225674-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 2 1.0 DOCUMENTATION This document contains the Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 2. This document is located in Appendix A and is presented in the customer requested format.
 
==2.0 REFERENCES==
 
References identified with an (*) are maintained within Indian Point Unit 2 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-Ichi Accident," March 12, 2012.2. 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 NL-13-042 "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. ML13079A348.
: 4. *Entergy Letter to U.S. NRC, letter number NL-14-031, "Indian Point Energy Center's Second Six-Month Status Report for the Implementation of Order Number EA-12-049 Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," February 27, 2014, NRC ADAMS Accession No. ML14070A365.
: 5. *Entergy Letter to U.S. NRC, letter number NL-14-1 10, "Indian Point Energy Center's Third Six-Month Status Report for the Implementation of Order EA-12-049 Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," August 27, 2014, NRC ADAMS Accession No. ML14251A227.
: 6. *Entergy Engineering Evaluation, EC No. 45874, Revision 1, "FLEX-Beyond Design Basis External Event Phases 1, 11, and III Strategy Development Evaluation." 7. *Entergy Drawing 9321-F-2017, Revision 84, "Flow Diagram -Main Steam." 8. *Entergy Drawing 9321-F-2019, Revision 116, "Flow Diagram -Boiler Feedwater." 9. *Entergy Drawing 251132, Revision 6, "AFW: Aux. Boiler Feed Pump #22 Flow Control Loop No.s 1188, 1213, 1261 & 1264." 10. *Entergy Drawing 251123, Revision 6, "AFW Flow to Steam Gen's. #21 & 22 Loop No's. 405, 406, 1200 & 1201." 11. *Entergy Drawing D251129, Revision 5, "AFW Flow to Steam Gen's. #23 & 24 Loop No's 405, 406,1202 & 1203." 12. *Entergy Drawing A207567, Revision 9, "Wiring for Transmitter Rack No. 5, 9, & 20." 13. *Entergy Drawing A241172, Revision 23, "Control Room Panel SC (JB1)." Page 6 Dr Im C-11 A AR EVA Document No.: 51-9225674-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 2 14. *Entergy Drawing 262727, Revision 0, "BLR FD. Water Sys. -Stm. Gen. #22 LID Level Ind./Recording LOOP-LT-427D." 15. *Entergy Drawing D260512, Revision 4, "Loop Diag. F.W. S.G. #23 Wide Range Level Loop Number: 437." 16. *Entergy Drawing 9321-F-70513, Revision 17, "Transmitter Racks Piping Arrangement-Sheet No. 4 Instrumentation." 17. *Entergy Drawing D252556, Revision 4, "MS Flow & Pressure Channel I (SG #22) Loop Number 429." 18. *Entergy Drawing D252557, Revision 4, "MS Flow & Pressure Channel I (SG #23) Loop Number 439." 19. *Entergy Drawing A241185, Revision 15, "Control Room Panel FB (JA2)." 20. *Entergy Drawing B225317, Revision 4, "Rack A3 Layout Reactor Protection System." 21. *Entergy Drawing 9321-F-2018, Revision 146, "Flow Diagram Condensate
& Boiler Feed Pump Suction -UFSAR Figure No. 10.2-5 (Sht. 1)." 22. *Entergy Drawing D262603, Revision 4, "Cond. Storage Tank Level Loops 1102, 1128." 23. *Entergy Drawing IP2--S-000313, Revision 1, "Condensate Storage Tank Level Indicator LI-1128." 24. *Entergy Drawing A235296, Revision 71, "Flow Diagram Safety Injection System UFSAR Figure No. 6.2-1 (Sht. 2)." 25. *Entergy Drawing 9321-F-3006, Revision 97, "Single Line Diagram 480V MCC 26A and 26B." 26. *Entergy Document 9P32.AA/12, "Individual Plant Examination for Indian Point Unit No. 2 External Events and Verification of Seismic Adequacy of Mechanical and Electrical Equipment Project No. 4342/9P32, USI A-46 Summary Report, Task AA: USI A-46," November 4, 1996.27. *Entergy Drawing 9321-F-2736, Revision 129, "Flow Diagram Chemical & Volume Control System -UFSAR Figure No 9.2-1 (Sht. 1)." 28. *Entergy System Design Description 1.0, Revision 15, "Indian Point Energy Center Unit No. 2 System Description No. 1 Reactor Coolant System." 29. *Entergy Drawing 9321-F-2735, Revision 141, "Flow Diagram Safety Injection System -UFSAR Figure No. 6.2-1 (Sht. 1)." 30. *Entergy Drawing 208093, Revision 13, "Piping Arrangement Remote Reactor Head Vent." 31. *Entergy Drawing IP2--S-000253, Revision 4, "Reactor Head Vent MOV HCV3101." 32. *Entergy System Design Description 30.0, Revision 5, "Indian Point Station Unit No. 2 System Description 30.0 Electric Heat Trace." 33. *Entergy Drawing 9321-F-3005, Revision 110, "One Line Diagram 480V Motor Control Center 27 & 27A." 34. *Entergy Drawing 226076, Revision 8, "I&C Loop Diagram Safety Injection System Containment Pressure Sensing." 35. *Entergy Drawing 241170, Revision 22, "Control Room Panel SB-1 (JB8)." Page 7
* jn.~~~~C Yr ZinZ~nI A AR EVA Document No.: 51-9225674-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 2 36. *Entergy Drawing A209762, Revision 72, "Flow Diagram Service Wtr Sys Nuclear Stm Supply Plant Sh 2 of 2 -UFSAR Figure No. 9.6-1 (Sht. 2)." 37. *Entergy Drawing A241169, Revision 18, "Control Room Panel SA (JB6)." 38. *Entergy Drawing 9321-F-2030, Revision 40, "Flow Diagram Fuel Oil to Diesel Generators." 39. *Entergy Drawing 250907, Revision 34, "Electrical Distribution and Transmission System -UFSAR Fig. No. 8.2-1 & 8.2-2." 40. *Entergy Drawing 9321-F-3008, Revision 92, "Single Line Diagram D.C. Power Panels 21, 22, 23, and 24 -UFSAR Figure No. 8.2-16." 41. *Entergy Drawing A208503, Revision 36, "Schem Dia of 118 VAC Inst Buses 21A, 22A, 23A and 24A (Located in CCR)." 42. *Entergy Drawing A225098, Revision 4, "Logic Diagram Nuclear Instrumentation Trip Signals (Sheet #5) UFSAR Figure No. 7.2-5." 43. *Entergy Drawing A241186, Revision 13, "Control Room Panel FC (JA3)." 44. *Entergy Drawing A241187, Revision 22, "Control Room Panel FD (JA4)." 45. *Entergy Drawing 9321-F-2738, Revision 122, "Flow Diagram Reactor Coolant System -UFSAR Figure No. 4.2-1." 46. *Entergy Drawing D260430, Revision 4, "Loop Diagram R.C.S Pressurizer Level Control Loop Number:459." 47. *Entergy Drawing B235537, Revision 2, "Reactor Vessel Level Instrumentation System Tag L-1311 & L-1312." 48. *Entergy Drawing 208538, Revision 11, "Diag. of External Conn's for Reactor Vessel System Process Cabinet "LCI" (Side Bays)." 49. *Entergy Drawing A242186, Revision 08, "Wiring Diagram Core Exit Thermocouple Monitoring System (Rack D8) Loc. in CCR." 50. *Entergy Design Basis Document IP2-MS DBD, Revision 02, "Design Basis Document for Main Steam System." 51. *Entergy Drawing A241171, Revision 29, "Control Room Panel SB-2 (JC3)." 52. *Entergy Drawing 23523, Revision 8, EC-29868, "Atmospheric Steam Dump Panel." 53. *Entergy System Design Description 3, Revision 12, "Indian Point Energy Center Unit No. 2 System Description No. 3 Chemical and Volume Control System." 54. *Entergy Drawing A208800 Revision 5, "Reactor Level Instr. Rack & Piping Details." 55. *Entergy Drawing A208726 Revision 7, "Installation of Reactor Vessel Level System Conduit Layout." 56. *Entergy Drawing D260510 Revision 7, "Loop Diag. F.W. S.G. #21 Wide Range Level Loop Number: 417, 5001." 57. *Entergy Drawing D260513 Revision 4,"Loop Diag. F.W. S.G. #22 Wide Range Level Loop Number: 447." 58. *Entergy Drawing A227551 Revision 64, "Fire Protection System Diagram Details Sheet #1." Page 8
*y*i.rrr a~f A AR EVA Document No.: 51-9225674-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 2 59. *Entergy Plant Equipment Database for Indian Point Unit 2.60. *Entergy Drawing 313113, Revision 0, "RCS Wide Range Press. Transmitter Schematic Wiring Diagram for Ind. Lights for Hydr. Isolator PIS-402 Electrical." 61. *Entergy Drawing A208537 Revision 06, "D/C Misc Field Mounted Equipment for Reactor Vessel Level System." 62. *Entergy System Design Description 10.1, Revision 16, "Indian Point Energy Center Unit No. 2 System Description No. 10.1 Safety Injection System." 63. *Entergy Drawing 9321-F-7015 Revision 34, "Instrument Piping Schematics Sheet No. 6 Instrumentation." 64. *Entergy Drawing 9-9239 DWG 1, Revision 4, "General Plan of 30'-0" Diam x 35'-3" High Dome Roof Tank." 65. *Entergy Drawing D260209, Revision 3, "LOOP Diagram RCS Pressure Loop #1, Loop Number 402." 66. *Entergy Drawing B238709 Revision 2, "Control Schematic
-Pressure Reducing Valve PCV-1139 for Auxiliary Boiler Feed Pump #22." 67. *Entergy Drawing A226980 Revision 2, "Aux. Boiler Feed Pump Control Station PT1 Wiring." 68. *Entergy Drawing 208538 Revision 11, "Diag of External Conn's for Reactor Vessel System Process Cabinet "LCI"." 69. *Entergy Drawing 300825 Revision 0, "Steam Gen. 22 Wide Range Level Transmitter LT-427D." 70. *Entergy Drawing A206649, Revision 26, "Conduit Layout Control Building Elevation 33'-0"." 71. *Entergy Drawing A208728 Revision 6, "Supplementary DWG Reactor Vessel Level Sys.Schematic
& Tray Schedule." 72. *Entergy Plant Electrical Cable and Conduit Database for Indian Point Unit 2.73. 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 2013.74. EPRI, "Indian Point Seismic Hazard and Screening Report," Revision 1, October 2013.75. Entergy Letter to U.S. NRC, letter number NL-14-042, "Entergy Seismic Hazard and Screening 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 ," March 31, 2014, NRC ADAMS Accession No. ML14099A110.
: 76. *"Indian Point Energy Center Unit 2 Updated Final Safety Analysis Report," Revision 25, Docket No. 50-247, 2014.77. *"Indian Point Energy Center Unit 3 Updated Final Safety Analysis Report," Revision 5, Docket No. 50-286, 2013.78. EPRI-NP-6041-SL, "Methodology for Assessment of Nuclear Power Plant Seismic Margin," Revision 1, August 1991.79. EPRI TR-103959, "Methodology for Developing Seismic Fragilities," July 1994.Page 9 A AR EVA Document No.: 51-9225674-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 2 80. NRC NUREG-1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," June 1991.81. *Entergy Document, "Individual Plant Examination of External Events for Indian Point Unit No. 2 Nuclear Generating Station," December 1995.82. SQUG, "Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment, Seismic Qualification Utility Group," Revision 3A, December 2001.83. NRC (E. Leeds) Letter to All Power Reactor Licensees et al., "Screening and Prioritization Results Regarding Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(F)Regarding Seismic Hazard Re-Evaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights From the Fukushima Dai-Ichi Accident," May 9, 2014.84. 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, 2014.85. 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.
: 86. 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, 2013.87. *Entergy Document EC54070, "ESEP Reports," the following AREVA documents are captured in the plant document management system: a. AREVA Document 51-9212950-008, "ESEP Expedited Seismic Equipment List (ESEL) -Indian Point Unit 2." b. AREVA Calculation 32-9227343-001, "Indian Point Unit 2 ESEP HCLPF Calculation
-BUS 2A, 3A, 5A, and 6A." c. AREVA Calculation 32-9227577-000, "Indian Point Unit 2 ESEP HCLPF Calculation
-Battery Chargers 21-24." d. AREVA Calculation 32-9227719-000, "Indian Point Unit 2 ESEP HCLPF Calculation
-125 VDC Power Panels EPB3, EPA9, PC1, and PC2." e. AREVA Calculation 32-9229683-000, "Indian Point Unit 2 ESEP HCLPF Calculation
-Battery Room and MCC-24A and 29A Blockwalls." f. AREVA Calculation 32-9230350-000, "Indian Point Unit 2 ESEP HCLPF Calculation
-Condensate Storage Tank." g. AREVA Calculation 32-9230411-000, "Indian Point Unit 2 ESEP HCLPF Calculation
-Refueling Water Storage Tank, 0021RWST." h. AREVA Document 32-9232862-000, "Indian Point Unit 2 ESEP Calculation
-Fire Protection Water Storage Tank." AREVA Document 32-9232844-000, "Indian Point Unit 2 ESEP Calculation
-Primary Water Storage Tank." The following references are AREVA references which were used as input for Appendix A.Page 10 1- 0 r -ri rd 4. 1 C)A AR EVA Document No.: 51-9225674-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 2 8 7.. AREVA Document 51-9229367-001, "Input to Entergy ESEP Report Sections 2 and 3 for Indian Point Unit 2." 88. AREVA Document 51-9227335-000, "Input to Entergy ESEP Report Sections 4 and 5 for Indian Point Unit 2." 89. AREVA Document 51-9230378-000, "Input to Entergy ESEP Report Sections 6, 7, and 8 for Indian Point Unit 2.90. AREVA Document 32-9224077-003, "Indian Point Unit 2 ESEP Binning and Screening." 91. AREVA Document 38-9232222-000, "Indian Point Unit 2 ESEP Report Comment Resolution Form." Page 11 A AR EVA Document No.: 51-9225674-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 2 APPENDIX A: EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT FOR INDIAN POINT UNIT 2 Note: Customer requested formatting begins on the following page.Page A-1 EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT FOR INDIAN POINT UNIT 2 (IP2)Page 1 Indian Point Unit 2 ESEP Report Table of Contents Page LIST O F TA B LES ............................................................................................................................................
4 LIST O F FIG U R ES ..........................................................................................................................................
5 1.0 PURPOSE AND OBJECTIVE
...............................................................................................................
6 2.0 BRIEF
 
==SUMMARY==
OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES
..................................
6 3.0 EQUIPMENT SELECTION PROCESS AND ESEL ..............................................................................
7 3.1 Equipment Selection Process and ESEL ...........................................................................
8 3.1.1 ESEL Development
..........................................................................................
9 3.1.2 Power Operated Valves ...................................................................................
9 3.1.3 Pull Boxes ......................................................................................................
9 3.1.4 Termination Cabinets ....................................................................................
10 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 GMRS Submitted by the Licensee ....................................................................
10 4.2 Comparison to SSE ............................................................................................................
12 5.0 REVIEW LEVEL GROUND MOTION (RLGM) ...............................................................................
14 5.1 Description of RLGM Selected .....................................................................................
14 5.2 Method to Estimate In-Structure Response Spectra (ISRS) ..........................................
16 6.0 SEISM IC MARGIN EVALUATION APPROACH .............................................................................
16 6.1 Summary of Methodologies Used ...............................................................................
17 6.2 HCLPF Screening Process ..............................................................................................
17 6.3 Seismic W alkdown Approach .......................................................................................
18 6.3.1 W alkdown Approach ......................................................................................
18 6.3.2 Application of Previous W alkdown Information
..........................................
19 6.3.3 Significant W alkdown Findings ......................................................................
19 6.4 HCLPF Calculation Process ............................................................................................
20 6.5 Functional Evaluations of Relays .................................................................................
21 6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes) ......................................
21 7.0 INACCESSIBLE ITEMS .....................................................................................................................
21 7.1 Identification of ESEL Item Inaccessible for Walkdowns
........................
21 7.2 Planned W alkdown / Evaluation Schedule / Close Out ...............................................
23 8.0 ESEP CONCLUSIONS AND RESULTS ..........................................................................................
23 8.1 Supporting Information 2...................................................................................................
23 Page 2
.... .orr w on Indian Point Unit 2 ESEP Report Table of Contents (continued)
Page 8.2 Identification of Planned M odifications
......................................................................
25 8.3 M odification Im plementation Schedule ......................................................................
25 8.4 Summary of Regulatory Commitments
........................................................................
25 9 .0 R EFER EN C ES ..................................................................................................................................
26 ATTACHM ENT A -INDIAN POINT UNIT 2 ESEL ....................................................................................
A-1 ATTACHMENT B -ESEP HCLPF VALUES AND FAILURE MODES TABULATION
.....................................
B-1 Page 3 or1 Lntoir on k.. y Indian Point Unit 2 ESEP Report List of Tables Page TABLE 4-1: GM RS FOR INDIAN POINT UNIT 2 ......................................................................................
11 TABLE 4-2: SSE FOR INDIAN POINT UNIT 2 ..........................................................................................
13 TABLE 5-1: RLGM FOR INDIAN POINT UNIT 2 .....................................................................................
14 Page 4 F c r mcal 0- I t Indian Point Unit 2 ESEP Report List of Figures Page FIGURE 4-1: GM RS FOR INDIAN POINT UNIT 2 ....................................................................................
12 FIGURE 4-2: GMRS TO SSE COMPARISON FOR INDIAN POINT UNIT 2 .................................................
13 FIGURE 5-1: RLGM FOR INDIAN POINT UNIT 2 ...................................................................................
16 Page 5 or, Indian Point Unit 2 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 subsequent tsunami, the Nuclear Regulatory Commission (NRC)established a Near-Term Task Force (NTTF) to conduct a systematic review of NRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena.
Subsequently, the NRC issued a 50.54(f) letter on March 12, 2012 [1], requesting information to assure that these recommendations are addressed by all U.S. nuclear power plants. The 50.54(f) letter 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 Indian Point Unit 2. 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 Indian Point Unit 2 FLEX strategies for Reactor Core Cooling and Heat Removal, Reactor Inventory Control/Long Term Subcriticality, and Containment Function are summarized below. This summary is derived from the Indian Point Energy Center Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order EA-12-049
[3], and is consistent with the second and third six-month status reports [4][5] and supplemented by supporting FLEX engineering calculations
[6].Core Cooling and Heat Removal The Phase 1 FLEX strategy at Indian Point Unit 2 for this function is to use Atmospheric Dump Valves (ADVs) and Main Steam Safety Valves (MSSVs) to remove heat, with the steam generator being fed by the turbine-driven Auxiliary Feedwater (AFW) pump. Backup nitrogen cylinders are available to support cycling the ADVs. Suction for the AFW pump is from the Condensate Storage Tank (CST), the Primary Water Storage Tank (PWST), and the Fire Water Storage Tank (FWST).During Phase 2 of the FLEX strategy, portable diesel-driven pumps will be staged to provide makeup to the CST or to the steam generator feedwater pump suction. The diesel-driven steam generator FLEX feed pump will be staged to provide feedwater to steam generators when the turbine-driven AFW pump becomes unavailable.
Diesel fuel for FLEX equipment will be provided from existing onsite Emergency Diesel Generator (EDG) Fuel Oil Storage Tanks.Page 6 ni n5 Indian Point Unit 2 ESEP Report The key parameters to be monitored are: steam generator level, steam generator pressure, CST level, Reactor Coolant System (RCS) pressure, and RCS temperature.
RCS Inventory Control For At Power modes In Phase 1, plant cooldown and depressurization will occur. RCS Inventory control is achieved via the accumulators.
During Phase 2, to avoid adverse effects on the RCS natural circulation flow, the cold-leg accumulator isolation valves are electrically closed during the cooldown to prevent nitrogen injection into the RCS.A FLEX pump will be used to provide RCS makeup with borated water supplied by the Refueling Water Storage Tank (RWST). To allow borated water injection into the RCS, the reactor head vent can be opened, if necessary, to provide a letdown path.If an Extended Loss of AC Power (ELAP) event occurs during cold weather months when freezing of the RWST could possibly occur, a FLEX diesel generator can be used to repower the Electric Heat Trace (EHT) system.For Shutdown modes In Phase 1, if the refueling canal is full then RCS makeup can be supplied by gravity feed from the RWST.During Phase 2, a FLEX pump will be used to provide RCS makeup from the RWST in the same manner as for the At Power modes.Additional key parameters to be monitored are pressurizer level, reactor vessel level, and nuclear instrumentation.
Containment Function Containment function is not expected to be challenged during Phase 1 or Phase 2 for an ELAP event occurring when the plant is in Mode 1-4. Therefore, no FLEX strategy beyond monitoring containment pressure and temperature was developed to support containment function.Supporting Systems Necessary electrical components are outlined in the Indian Point Unit 2 0IP and primarily entail station batteries, Direct Current (DC) buses, distribution panels, inverters, battery chargers, and instrument buses.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 Indian Point Unit 2 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], [321, [33], [34], [35], [36], [37], [381, [391, [40], [41], [421, [43], [441, [451, [461, [47], [481, [49], [501,[51], [52], [53], [54], [55], [56], [57], [58], [59], [60], [61], [62], [63], [64], [65], [66], [67], [68], [69], [70],[71], and [72].Page 7 c f " , J Indian Point Unit 2 ESEP Report 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 Indian Point Unit 2 0IP in Response to the March 12, 2012, Commission Order EA-12-049
[3], and is consistent with the second and third six-month status reports issued to the NRC [4] [5]. The OIP provides the Indian Point Unit 2 FLEX mitigation strategy and serves as the basis for equipment selected for the ESEP.The scope of "installed plant equipment" includes equipment relied upon for the FLEX strategies to sustain the critical functions of core cooling and containment integrity consistent with the Indian Point Unit 2 0IP. 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 subcriticality, and containment integrity functions.
Portable and pre-staged FLEX equipment (not permanently installed) are excluded from the ESEL per EPRI 3002000704.
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 in Table 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 Indian Point Unit 2 0IP.2. The scope of components is limited to installed plant equipment, and FLEX connections necessary to implement the Indian Point Unit 2 OIP as described in Section 2 of 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, control building, auxiliary building, etc.).* Piping, cabling, conduit, HVAC, and their supports.* Manual valves, check valves and rupture disks.* Power-operated valves not required to change state as part of the FLEX mitigation strategies.
* Nuclear steam supply system components (e.g., RPV 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.Page 8
'ioý- OnIv Indian Point Unit 2 ESEP Report 3.1.1 ESEL Development The ESEL was developed by reviewing the Indian Point OIP [3], second and third six-month status reports [4] [5], and supporting FLEX engineering calculations
[6] to determine the major equipment involved in the FLEX strategies.
Further reviews of plant drawings (e.g., Piping and Instrumentation Diagrams (P&IDs) and Electrical One Line Diagrams) were performed to identify the boundaries of the flowpaths to be used in the FLEX strategies and to identify specific components in the flowpaths needed to support implementation of the FLEX strategies.
Boundaries were established at an electrical or mechanical isolation device (e.g., isolation amplifier, valve, etc.) in branch circuits / branch lines off the defined strategy electrical or fluid flowpath.
P&IDs were the primary reference documents used to identify mechanical components and instrumentation.
The flow paths used for FLEX strategies were selected and specific components were identified using detailed equipment and instrument drawings, piping isometrics, electrical schematics and one-line drawings, system descriptions, design basis documents, as necessary.
Cabinets and equipment controls containing relays, contactors, switches, potentiometers, circuit breakers and other electrical and instrumentation that could be affected by high-frequency earthquake motions and that impact the operation of equipment in the ESEL are required to be on the ESEL. These cabinets and components were identified in the ESEL.For each parameter monitored during the FLEX implementation, a single indication was selected for inclusion in the ESEL. 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 not included in 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 as part of the FLEX mitigation strategies 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 Indian Point Unit 2 ESEL for functional failure modes associated with power operated valves:* Power operated valves that remain energized during the ELAP events (such as DC powered valves), were included on the ESEL." Power operated valves not required to change state as part of the FLEX mitigation strategies were not included on the ESEL. The seismic event also causes the ELAP event; therefore, the valves are incapable of spurious operation as they would be de-energized.
* Power operated valves not required to change state as part of the FLEX mitigation strategies during Phase 1, and are re-energized and operated during subsequent Phase 2 and 3 strategies, were not evaluated for spurious valve operation as the seismic event that caused the ELAP has passed before the valves are re-powered.
3.1.3 Pull Boxes Pull boxes were deemed unnecessary to 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
[2].Page 9 Indian Point Unit 2 ESEP Report 3.1.4 Termination Cabinets Termination cabinets, including cabinets necessary for FLEX Phase 2 and Phase 3 connections, provide consolidated locations for permanently connecting multiple cables. The termination cabinets and the internal connections provide a completely passive function; however, the cabinets are included in the ESEL to ensure industry knowledge on panel/anchorage failure vulnerabilities is addressed.
 
====3.1.5 Critical====
Instrumentation Indicators Critical indicators and recorders are typically physically located on panels/cabinets and are included as separate components; however, seismic evaluation of the instrument indication may be included in the panel/cabinet seismic evaluation (rule-of-the-box).
 
====3.1.6 Phase====
2 and 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 Indian Point Unit 2 OIP as described in Section 2." Item 3 in Section 3.1 also notes that "The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e., either "Primary" or "Back-up/Alternate")." Item 6 in Section 3.1 above goes on to explain that "Piping, cabling, conduit, HVAC, and their supports" are excluded from the ESEL scope in accordance with EPRI 3002000704
[2].Therefore, piping and pipe supports associated with FLEX Phase 2 and Phase 3 connections are excluded from the scope of the ESEP evaluation.
However, any active valves in FLEX Phase 2 and Phase 3 connection flow path are included in the ESEL.3.2 Justification for Use of Equipment That is Not the Primary Means for FLEX Implementation The Indian Point Unit 2 ESEL is based on the primary means of implementing the FLEX strategy.Therefore, no additional justification is required.4.0 GROUND MOTION RESPONSE SPECTRUM (GMRS)4.1 Plot of GMRS Submitted by the Licensee In accordance with the guidance provided in Section 2.4.2 of the Screening, Prioritization and Implementation Details (SPID) [73] for rock sites, the SSE control point elevation is defined at the top of hard-rock and is applicable at grade in the free field as well as the various foundations elevations
[74].Table 4-1 shows the GMRS acceleration for a range of spectral frequencies
[75]. The GMRS at the control point is shown in Figure 4-1.Page 10 itorrn V~V~ u~Indian Point Unit 2 ESEP Report Table.4-1:
GMRS for Indian Point Unit 2 Frequency GMRS (Hz) (g)100 4.12E-01 90 4.46E-01 80 5.04E-01 70 5.94E-01 60 7.04E-01 50 8.06E-01 45 8.42E-01 40 8.66E-01 35 8.77E-01 30 8.75E-01 25 8.58E-01 20 8.28E-01 15 7.67E-01 12.5 7.17E-01 10 6.48E-01 9 6.04E-01 8 5.55E-01 7 5.02E-01 6 4.46E-01 5 3.85E-01 4 3.14E-01 3 2.36E-01 2.5 1.94E-01 2 1.59E-01 1.5 1.17E-01 1.25 9.42E-02 1 7.04E-02 0.9 6.40E-02 0.8 5.71E-02 0.7 4.99E-02 0.6 4.25E-02 0.5 3.48E-02 Page 11 For niocn~+~tion
~Indian Point Unit 2 ESEP Report Table 4-1: GMRS for Indian Point Unit 2 (continued)
Frequency GMRS (Hz) (g)0.4 2.78E-02 0.3 2.09E-02 0.2 1.39E-02 0.167 1.16E-02 0.125 8.69E-03 0.1 6.95E-03 GMRS at Control Point for Indian Point Unit 2, 5% Damping 1.00 0.90 0.80 0.70 0.60 0.50 0.40 0.30 0.20 0.10 0.00 0.1 1 10 100 Frequency (Hz)Figure 4-1: GMRS for Indian Point Unit 2 4.2 Comparison to SSE The SSE corresponds to a horizontal acceleration of 0.15g [76]. The SSE is defined in the FSAR in terms of a PGA and a design response spectrum.
These spectra have been digitized and tabulated
[75]. Table 4-2 shows these spectral acceleration values at selected frequencies for the 5% damped horizontal SSE.Page 12 Indian Point Unit 2 ESEP Report Table 4-2: SSE for Indian Point Unit 2 Frequency Spectral Acceleration (Hz) (g)100 0.15 25 0.15 10 0.168 5 0.228 2.5 0.234 1 0.127 0.5 0.075 GMRS to SSE Comparison for Indian Point Unit 2, 5% Damping 0.90 080 0.60 0.40 0.30 0.10 0.10 0.1 10 1(X Frequncy (Hz)Figure 4-2: GMRS to SSE Comparison for Indian Point Unit 2 The SSE envelops the GMRS for lower frequencies up to nearly 3Hz. The GMRS exceeds the SSE beyond that point. As the GMRS exceeds the SSE in the 1 to 1OHz 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-frequency GMRS exceedances at Low Seismic Hazard Sites and b) Narrow Band Exceedances in the 1 to 10Hz range, provide criteria for accepting specific GMRS exceedances.
However, the GMRS exceedances occur in the frequency range of interest and cannot be characterized as narrow-banded exceedances.
Therefore, these special screening considerations do not apply for Indian Point Unit 2 and High Confidence of a Low Probability of Failure (HCLPF) evaluations are to be performed.
Page 13 or In io 0 Indian Point Unit 2 ESEP Report 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. The maximum GMRS/SSE ratio between 1 and 10Hz range occurs at 10Hz where the ratio is 0.648/0.168
= 3.86. As the maximum ratio of the GMRS to the SSE over the i to 10Hz range exceeds a value of 2, the GMRS/SSE ratio is set to the maximum scaling factor value of 2.0 for Indian Point Unit 2 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.Table 5-1: RLGM for Indian Point Unit 2 Frequency RLGM at 5% Damping (Hz) (g)100.00 0.30 33.00 0.30 15.78 0.31 13.65 0.31 12.20 0.32 10.09 0.34 8.43 0.36 7.23 0.38 6.55 0.40 5.89 0.43 5.34 0.44 4.94 0.46 4.59 0.47 4.17 0.48 3.83 0.49 3.67 0.49 3.41 0.50 3.15 0.49 2.89 0.49 2.65 0.48 2.55 0.47 2.24 0.45 2.02 0.42 Page 14 T::::, -.1 ... vý--,,-nadon On!1 Indian Point Unit 2 ESEP Report Table 5-1: RLGM for Indian Point Unit 2 (continued)
Frequency RLGM at 5% Damping (Hz) (g)1.78 0.40 1.64 0.38 1.57 0.36 1.46 0.34 1.34 0.32 1.23 0.30 1.16 0.28 1.09 0.27 1.01 0.26 0.96 0.24 0.91 0.23 0.86 0.22 0.81 0.21 0.76 0.21 0.74 0.20 0.71 0.19 0.68 0.19 0.64 0.18 0.60 0.17 0.56 0.16 0.53 0.15 0.10 0.14 Page 15
~- or ~ *ii ~d Indian Point Unit 2 ESEP Report Review Level Ground Motion (2xSSE) Response Spectra -Horizontal Direction 0.60 0.50 0.40~0.30-0.20 0.10 0.00 0.10 1.00 10.00 Frequency (Hz)Figure 5-1: RLGM for Indian Point Unit 2 100.00 5.2 Method to Estimate In-Structure Response Spectra (ISRS)The RLGM ISRS for Indian Point Unit 2 are generated by scaling the SSE ISRS [76] [77]. 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 1 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) [78].Page 16 Indian Point Unit 2 ESEP Report 2. Probabilistic approach using the fragility analysis methodology of EPRI TR-103959, Methodology for Developing Seismic Fragilities
[79].6.1 Summary of Methodologies Used Indian Point Unit 2 was classified as a 0.3g full scope plant in NUREG-1407
[80] and performed a SPRA as part of Individual Plant Examination for External Events (IPEEE) program. The SPRA is documented in [81]. Indian Point Unit 2 IPEEE program followed the NUREG-1407 methodology for seismic evaluation with plant seismic walkdowns using EPRI NP-6041-SL
[78] and Generic Implementation Procedure
[82]. Walkdown efforts were coordinated for evaluations pertaining to the IPEEE and Unresolved Safety Issue (USI) A-46. Section 3.3 of [75] established that the results of the Indian Point Unit 2 SPRA performed as part of IPEEE will not be used as a basis for Indian Point Unit 2 to screen-out of further risk assessment.
For ESEP, the SMA consisted of screening walkdowns and HCLPF calculations.
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 2xSSE with a PGA of 0.3g as shown on Figure 5-1.6.2 HCLPF Screening Process For ESEP, the components are screened at RLGM (2xSSE) with a 0.3g PGA. The screening tables in EPRI NP-6041-SL
[781 are based on ground peak spectral accelerations of 0.8g and 1.2g. These both exceed the RLGM peak spectral acceleration.
The ESEL components were prescreened based on Table 2-4 of EPRI NP-6041-SL.
Additional pre-screening, specifically for anchorage, considered walkdown results and documentation from NTTF 2.3 and SEWS from IPEEE and USI A-46. Equipment anchorage was screened out in cases where previous evaluations showed large available margin against SSE. The remaining components (i.e., components that do not screen out), were identified as requiring HCLPF calculations.
ESEL components were walked down and based on the equipment and anchorage conditions, prescreening decisions were confirmed and a final list of required HCLPF calculations was generated.
Equipment for which the screening caveats were met and for which the anchorage capacity exceeded the RLGM seismic demand are screened out from ESEP seismic capacity determination because the HCLPF capacity exceeds the RLGM.The Indian Point Unit 2 ESEL contains 181 items. Of these, 30 are valves. In accordance with Table 2-4 of EPRI NP-6041-SL, both active and passive valves may be assigned a functional capacity of 0.8g peak spectral acceleration withoutany review other than looking for valves with large extended operators on small diameter piping, and anchorage is not a failure mode. 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.
Page 17 Fof h-o rrra ' in f- l Indian Point Unit 2 ESEP Report Block walls in close proximity to Battery Room 21, 22, 23 and 24 as well as MCC-24A and MCC-29A were identified in the proximity of ESEL equipment.
These block walls were assessed for potential seismic interaction impact resulting from the RLGM by reviewing the existing plant documents and/or by generating new analyses and found to be acceptable.
 
===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
[78] for the Seismic Margin Assessment process. Pages 2-26 through 2-30 of EPRI NP-6041-SL describe the seismic walkdown criteria, including the following key criteria."The SRT[Seismic Review Team] should "walk by" 100% of all components which are reasonably accessible and in non-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 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 Page 18
'ju Indian Point Unit 2 ESEP Report sample size of one component of each type for thorough inspection will have to be increased.
The increase in sample size which should be inspected will depend upon the number of outliers and different anchorages, etc., which are observed.
It is up to the SRT to ultimately select the sample size since they are the ones who are responsible for the seismic adequacy of all elements which they screen from the margin review. Appendix D gives guidance for sampling selection." 6.3.2 Application of Previous Walkdown Information Several ESEL items were previously walked down during the Indian Point Unit 2 seismic IPEEE program, which was performed in accordance with 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 the previous walkdown, 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
[78], no significant outliers or anchorage concerns were identified during the Indian Point Unit 2 seismic walkdowns.
Based on walkdown results, HCLPF capacity evaluations were recommended for the following sixteen (16) components:
* BATTCHG21, Battery Charger 21* BATTCHG22, Battery Charger 22" BATTCHG23, Battery Charger 23* BATTCHG24, Battery Charger 24" BUS 2A, 480VAC Bus 2A* BUS 3A, 480VAC Bus 3A* BUS 5A, 480VAC Bus 5A* BUS 6A, 480VAC BUS 6A" PC1, 125 VDC Power Panel 21* PC2, 125 VDC Power Panel 22* EPB3, 125 VDC Power Panel 23* EPA9, 125 VDC Power Panel 24* 0021RWST, 21 Refueling Water Storage Tank* CST, Condensate Storage Tank* 300KFPT, Fire Water Storage Tank" PWST, Primary Water Storage Tank Page 19
-r Indian Point Unit 2 ESEP Report Several 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. One (1) HCLPF evaluation was performed addressing the block walls in close proximity to battery rooms 21, 22, and 24 as well as MCC-24A and MCC-29A.The following is the list of the block walls and the wall type.* Four (4) Walls in Battery Room No. 23 -Partially grouted" Wall ID 4-033-21 -Brick* Wall ID 4-033-23 -Brick* Wall ID 4-033-25 -Brick" Wall ID 4-033-08 -Hollow Block* Wall ID 4-033-10 -Hollow Block* Wall ID 4-033-12 -Hollow Block* Wall ID 4-033-14 -Hollow Block* Wall ID 4-033-16 -Hollow Block" Wall ID 4-033-18 -Hollow Block* Wall ID 4-033-33 -Hollow Block" Wall ID 4-033-35 -Hollow Block* Wall ID 4-033-37 -Hollow Block* Turbine Building North Block Wall -Partially Grouted" Turbine Building East Block Wall -Partially Grouted 6.4 HCLPF Calculation Process ESEL items identified for ESEP at Indian Point Unit 2 were evaluated using the criteria in EPRI NP-6041-SL [78] 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, or N1TF 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.
Eight (8) HCLPF calculations were performed to address the sixteen (16) components as well as the seismic adequacy of block walls.* Calculation "Battery Chargers 21-24" addressing four (4) components:
BATTCHG21, BATTCHG22, BATTCHG23 and BA1TCHG24 Page 20 F r rn z-i n y Indian Point Unit 2 ESEP Report* Calculation "BUS 2A, 3A, 5A, and 6A" addressing four (4) components:
BUS 2A, BUS 3A, BUS 5A and BUS 6A" Calculation "125 VDC Power Panels EPB3, EPA9, PC1, and PC2" addressing four (4) components:
PC1, PC2, EPB3 and EPA9* Calculation "Refueling Water Storage Tank, 0021RWST" addressing one (1) component 0021RWST" Calculation, "Condensate Storage Tank" addressing one (1) component CST* Calculation "Battery Room and MCC-24A and MCC-29A Blockwalls" addressing eighteen (18)block walls* Calculation for "300KFPT, Fire Water Storage Tank" addressing one (1) component* Calculation for "PWST, Primary Water Storage Tank" addressing one (1) component 6.5 Functional Evaluations of Relays No seal in/lockout type relays were identified on Indian Point Unit 2 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
[78] 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:
* 0021RWST, 21 Refueling Water Storage Tank" 300KFPT, Fire Water Storage Tank Modifications are planned for the above component.
 
===7.0 INACCESSIBLE===
 
ITEMS 7.1 Identification of ESEL Item Inaccessible for Walkdowns Forty-one (41) components on the ESEL were inaccessible and not walked down since they are located in the Primary Containment Building in a locked high radiation area or other areas that were inaccessible due to contamination and radiation concerns at the time of the walkdowns and there were no alternate means of evaluating these items: Page 21 or.Indian Point Unit 2 ESEP Report* ACCUM 21, Accumulator Tank 21* ACCUM 22, Accumulator Tank 22" ACCUM 23, Accumulator Tank 23" ACCUM 24, Accumulator Tank 24* CH-HCV-133, RHR Purification Line Control Valve* EXC6, Terminal Box* EWV8, Terminal Box" HCV-3100, Remote Reactor Head Vent MOV* HCV-3101, Remote Reactor Head Vent MOV* INST RK 21, Instrument Rack 21" LCV-459, Letdown Isolation Valve" LIS-1311, Hydraulic Isolator" LIS-1312, Hydraulic Isolator" LT-417D, SG 21 WR Level Transmitter
* LT-447D, SG 24 WR Level Transmitter
* LT-427D, Steam Generator 22 WR Level Transmitter
* LT-437D, Steam Generator 23 WR Level Transmitter
* LT-459, Pressurizer Level Transmitter
* MOV-882, RHR Pump Suction Isolation Valve* MOV-894A, Boronated Water Injection Valve" MOV-894B, Boronated Water Injection Valve* MOV-894C, Boronated Water Injection Valve" MOV-894D, Boronated Water Injection Valve* NC-31D, Neutron Source Range Detector* NC-32D, Neutron Source Range Detector" NC-41D, Neutron Source Range Detector* NC-42D, Neutron Source Range Detector* PT-402, RCP WR Pressure Transmitter and RVLIS Train A" EXG7, Reactor Level Hydraulic Isolator Rack* TE-411, Temperature Element for RVLIS-Train A* TE-1203-1, Fan Coil Unit 21 Temperature Element" TE-1203-2, Fan Coil Unit 22 Temperature Element Page 22 rn'o~u,'oni r 1 ()', Indian Point Unit 2 ESEP Report" TE-1203-3, Fan Coil Unit 23 Temperature Element" TE-1203-4, Fan Coil Unit 24 Temperature Element" TE-1203-5, Fan Coil Unit 25 Temperature Element* TE-1313, Reactor Vessel Upper Compensation Temp* TE-1314, Reactor Vessel Upper Compensation Temp* TE-1317, Reactor Vessel Upper Compensation Temp" TE-1318, Reactor Vessel Upper Compensation Temp* TE-1319, Reactor Vessel Lower Tap* LT-1312, Reactor Vessel Level Transmitter WR In addition, two (2) heat exchangers listed below were not walked down due to their inaccessibility.
These two components were evaluated based review of existing calculations, NTTF 2.3 walkdowns and associated photographs, and a recent scan performed in the area and screened out.* 0021RHRHX, RHR Heat Exchanger
#21* O022RHRHX, RHR Heat Exchanger
#22 7.2 Planned Walkdown / Evaluation Schedule / Close Out The walkdowns of the inaccessible items identified in Section 7.1 are scheduled to be performed no later than the second planned refueling outage after December 31, 2014.8.0 ESEP CONCLUSIONS AND RESULTS 8.1 Supporting Information Indian Point Unit 2 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 Indian Point Unit 2 response to the NRC's 50.54(f) letter. On March 12, 2014, NEI submitted to the NRC results of a study [84] 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 [83] concluded 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." Page 23 4 ! " '. ..0 '- 0111041: Indian Point Unit 2 ESEP Report An assessment of the change in seismic risk for Indian Point Unit 2 was included in the fleet risk evaluation submitted in the March 12, 2014 NEI letter [84]; therefore, the conclusions in the NRC's May 9 letter also apply to Indian Point Unit 2.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 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" 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 [73]. As identified in the Indian Point Unit 2 Seismic Hazard and GMRS submittal
[75], Indian Point Unit 2 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.
Indian Point Unit 2 will complete that evaluation in accordance with the schedule identified in NEI's letter dated April 9, 2013 [85] and endorsed by the NRC in their May 7, 2013 letter[86].Page 24 Indian Point Unit 2 ESEP Report 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.* 0021RWST, 21 Refueling Water Storage Tank* 300KFPT, Fire Water 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 [85], 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, 2014.8.4 Summary of Regulatory Commitments The following actions will be performed as a result of the ESEP.Equipment.
Action ,Equipment ID Description ActionmDescription Completion hDate Perform seismic walkdowns, No later than the generate HCLPF calculations and end of the second 1 N/A N/A design and implement any planned refueling necessary modifications for plane rfuel inaccessible items listed in oeafter Section 7.1 December 31, 2014.21 Refueling Modify tank and anchorage such As described in 2 0021RWST Water Stor that HCLPF>RLGM Section 8.3.Storage Tank 3 30OKFPT Fire Water Modify tank and anchorage such As described in Storage Tank that HCLPF>RLGM Section 8.3 Submit a letter to NRC Within 60 days summarizing the HCLPF results of following 4 N/A N/A Items 1 to 3 and confirming completion of ESEP implementation of the plant activities, including modifications associated with items 1 to 3.items 1 to 3.Page 25 Indian Point Unit 2 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-Ichi Accident," March 12, 2012.2. 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 NL-13-042 "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. ML13079A348.
: 4. Entergy Letter to U.S. NRC, letter number NL-14-031, "Indian Point Energy Center's Second Six-Month Status Report for the Implementation of Order Number EA-12-049 Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," February 27, 2014, NRC ADAMS Accession No. ML14070A365.
: 5. Entergy Letter to U.S. NRC, letter number NL-14-110, "Indian Point Energy Center's Third Six-Month Status Report for the Implementation of Order EA-12-049 Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," August 27, 2014, NRC ADAMS Accession No. ML14251A227.
: 6. Entergy Engineering Evaluation, EC No. 45874, Revision 1, "FLEX-Beyond Design Basis External Event Phases I, II, and III Strategy Development Evaluation." 7. Entergy Drawing 9321-F-2017, Revision 84, "Flow Diagram -Main Steam." 8. Entergy Drawing 9321-F-2019, Revision 116, "Flow Diagram -Boiler Feedwater." 9. Entergy Drawing 251132, Revision 6, "AFW: Aux. Boiler Feed Pump #22 Flow Control Loop No.s 1188, 1213, 1261 & 1264." 10. Entergy Drawing 251123, Revision 6, "AFW Flow to Steam Gen's. #21 & 22 Loop No's. 405, 406, 1200 & 1201." 11. Entergy Drawing D251129, Revision 5, "AFW Flow to Steam Gen's. #23 & 24 Loop No's 405, 406, 1202 & 1203." 12. Entergy Drawing A207567, Revision 9, "Wiring for Transmitter Rack No. 5, 9, & 20." 13. Entergy Drawing A241172, Revision 23, "Control Room Panel SC (JB1)." 14. Entergy Drawing 262727, Revision 0, "BLR FD. Water Sys. -Stm. Gen. #22 L/D Level Ind./Recording LOOP-LT-427D." 15. Entergy Drawing D260512, Revision 4, "Loop Diag. F.W. S.G. #23 Wide Range Level Loop Number: 437." 16. Entergy Drawing 9321-F-70513, Revision 17, "Transmitter Racks Piping Arrangement
-Sheet No. 4 Instrumentation." 17. Entergy Drawing D252556, Revision 4, "MS Flow & Pressure Channel I (SG #22) Loop Number 429." Page 26 4 il J Indian Point Unit 2 ESEP Report 18. Entergy Drawing D252557, Revision 4, "MS Flow & Pressure Channel I (SG #23) Loop Number 439." 19. Entergy Drawing A241185, Revision 15, "Control Room Panel FB (JA2)." 20. Entergy Drawing B225317, Revision 4, "Rack A3 Layout Reactor Protection System." 21. Entergy Drawing 9321-F-2018, Revision 146, "Flow Diagram Condensate
& Boiler Feed Pump Suction -UFSAR Figure No. 10.2-5 (Sht. 1)." 22. Entergy Drawing D262603, Revision 4, "Cond. Storage Tank Level Loops 1102, 1128." 23. Entergy Drawing IP2--S-000313, Revision 1, "Condensate Storage Tank Level Indicator LI-1128." 24. Entergy Drawing A235296, Revision 71, "Flow Diagram Safety Injection System UFSAR Figure No. 6.2-1 (Sht. 2)." 25. Entergy Drawing 9321-F-3006, Revision 97, "Single Line Diagram 480V MCC 26A and 26B." 26. Entergy Document 9P32.AA/12, "Individual Plant Examination for Indian Point Unit No. 2 External Events and Verification of Seismic Adequacy of Mechanical and Electrical Equipment Project No. 4342/9P32, USI A-46 Summary Report, Task AA: USI A-46," November 4, 1996.27. Entergy Drawing 9321-F-2736, Revision 129, "Flow Diagram Chemical & Volume Control System-UFSAR Figure No 9.2-1 (Sht. 1)." 28. Entergy System Design Description 1.0, Revision 15, "Indian Point Energy Center Unit No. 2 System Description No. 1 Reactor Coolant System." 29. Entergy Drawing 9321-F-2735, Revision 141, "Flow Diagram Safety Injection System -UFSAR Figure No. 6.2-1 (Sht. 1)." 30. Entergy Drawing 208093, Revision 13, "Piping Arrangement Remote Reactor Head Vent." 31. Entergy Drawing IP2--S-000253, Revision 4, "Reactor Head Vent MOV HCV3101." 32. Entergy System Design Description 30.0, Revision 5, "Indian Point Station Unit No. 2 System Description 30.0 Electric Heat Trace." 33. Entergy Drawing 9321-F-3005, Revision 110, "One Line Diagram 480V Motor Control Center 27& 27A." 34. Entergy Drawing 226076, Revision 8, "I&C Loop Diagram Safety Injection System Containment Pressure Sensing." 35. Entergy Drawing 241170, Revision 22, "Control Room Panel SB-1 (JB8)." 36. Entergy Drawing A209762, Revision 72, "Flow Diagram Service Wtr Sys Nuclear Stm Supply Plant Sh 2 of 2 -UFSAR Figure No. 9.6-1 (Sht. 2)." 37. Entergy Drawing A241169, Revision 18, "Control Room Panel SA (JB6)." 38. Entergy Drawing 9321-F-2030, Revision 40, "Flow Diagram Fuel Oil to Diesel Generators." 39. Entergy Drawing 250907, Revision 34, "Electrical Distribution and Transmission System -UFSAR Fig. No. 8.2-1 & 8.2-2." Page 27
'ný 1 Gý ýOn Indian Point Unit 2 ESEP Report 40. Entergy Drawing 9321-F-3008, Revision 92, "Single Line Diagram D.C. Power Panels 21, 22, 23, and 24 -UFSAR Figure No. 8.2-16." 41. Entergy Drawing A208503, Revision 36, "Schem Dia of 118 VAC Inst Buses 21A, 22A, 23A and 24A (Located in CCR)." 42. Entergy Drawing A225098, Revision 4, "Logic Diagram Nuclear Instrumentation Trip Signals (Sheet #5) UFSAR Figure No. 7.2-5." 43. Entergy Drawing A241186, Revision 13, "Control Room Panel FC (JA3)." 44. Entergy Drawing A241187, Revision 22, "Control Room Panel FD (JA4)." 45. Entergy Drawing 9321-F-2738, Revision 122, "Flow Diagram Reactor Coolant System -UFSAR Figure No. 4.2-1." 46. Entergy Drawing D260430, Revision 4, "Loop Diagram R.C.S Pressurizer Level Control Loop Number:459." 47. Entergy Drawing B235537, Revision 2, "Reactor Vessel Level Instrumentation System Tag L-1311& L-1312." 48. Entergy Drawing 208538, Revision 11, "Diag. of External Conn's for Reactor Vessel System Process Cabinet "LC1" (Side Bays)." 49. Entergy Drawing A242186, Revision 08, "Wiring Diagram Core Exit Thermocouple Monitoring System (Rack D8) Loc. in CCR." 50. Entergy Design Basis Document IP2-MS DBD, Revision 02, "Design Basis Document for Main Steam System." 51. Entergy Drawing A241171, Revision 29, "Control Room Panel SB-2 (JC3)." 52. Entergy Drawing 23523, Revision 8, EC-29868, "Atmospheric Steam Dump Panel." 53. Entergy System Design Description 3, Revision 12, "Indian Point Energy Center Unit No. 2 System Description No. 3 Chemical and Volume Control System." 54. Entergy Drawing A208800 Revision 5, "Reactor Level Instr. Rack & Piping Details." 55. Entergy Drawing A208726 Revision 7, "Installation of Reactor Vessel Level System Conduit Layout." 56. Entergy Drawing D260510 Revision 7, "Loop Diag. F.W. S.G. #21 Wide Range Level Loop Number: 417, 5001." 57. Entergy Drawing D260513 Revision 4,"Loop Diag. F.W. S.G. #22 Wide Range Level Loop Number: 447." 58. Entergy Drawing A227551 Revision 64, "Fire Protection System Diagram Details Sheet #1." 59. Entergy Plant Equipment Database for Indian Point Unit 2.60. Entergy Drawing 313113, Revision 0, "RCS Wide Range Press. Transmitter Schematic Wiring Diagram for Ind. Lights for Hydr. Isolator PIS-402 Electrical." 61. Entergy Drawing A208537 Revision 06, "D/C Misc Field Mounted Equipment for Reactor Vessel Level System." Page 28 Indian Point Unit 2 ESEP Report 62. Entergy System Design Description 10.1, Revision 16, "Indian Point Energy Center Unit No. 2 System Description No. 10.1 Safety Injection System." 63. Entergy Drawing 9321-F-7015 Revision 34, "Instrument Piping Schematics Sheet No. 6 Instrumentation." 64. Entergy Drawing 9-9239 DWG 1, Revision 4, "General Plan of 30'-0" Diam x 35'-3" High Dome Roof Tank." 65. Entergy Drawing D260209, Revision 3, "LOOP Diagram RCS Pressure Loop #1, Loop Number 402." 66. Entergy Drawing B238709 Revision 2, "Control Schematic
-Pressure Reducing Valve PCV-1139 for Auxiliary Boiler Feed Pump #22." 67. Entergy Drawing A226980 Revision 2, "Aux. Boiler Feed Pump Control Station PT1 Wiring." 68. Entergy Drawing 208538 Revision 11, "Diag of External Conn's for Reactor Vessel System Process Cabinet "LCi"." 69. Entergy Drawing 300825 Revision 0, "Steam Gen. 22 Wide Range Level Transmitter LT-427D." 70. Entergy Drawing A206649, Revision 26, "Conduit Layout Control Building Elevation 33'-0"." 71. Entergy Drawing A208728 Revision 6, "Supplementary DWG Reactor Vessel Level Sys.Schematic
& Tray Schedule." 72. Entergy Plant Electrical Cable and Conduit Database for Indian Point Unit 2.73. 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 2013.74. EPRI, "Indian Point Seismic Hazard and Screening Report," Revision 1, October 2013.75. Entergy Letter to U.S. NRC, letter number NL-14-042, "Entergy Seismic Hazard and Screening 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 ," March 31, 2014, NRC ADAMS Accession No. ML14099A110.
: 76. "Indian Point Energy Center Unit 2 Updated Final Safety Analysis Report," Revision 25, Docket No. 50-247, 2014.77. "Indian Point Energy Center Unit 3 Updated Final Safety Analysis Report," Revision 5, Docket No. 50-286, 2013.78. EPRI-NP-6041-SL, "Methodology for Assessment of Nuclear Power Plant Seismic Margin," Revision 1, August 1991.79. EPRI TR-103959, "Methodology for Developing Seismic Fragilities," July 1994.80. NRC NUREG-1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," June 1991.81. Entergy Document, "Individual Plant Examination of External Events for Indian Point Unit No. 2 Nuclear Generating Station," December 1995.Page 29 Indian Point Unit 2 ESEP Report 82. SQUG, "Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment, Seismic Qualification Utility Group," Revision 3A, December 2001.83. 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, 2014.84. 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, 2014.85. 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.
: 86. 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, 2013.87. Entergy Document EC54070, "ESEP Reports," the following AREVA documents are captured in the plant document management system: a. AREVA Document 51-9212950-008, "ESEP Expedited Seismic Equipment List (ESEL) -Indian Point Unit 2." b. AREVA Calculation 32-9227343-001, "Indian Point Unit 2 ESEP HCLPF Calculation
-BUS 2A, 3A, 5A, and 6A." c. AREVA Calculation 32-9227577-000, "Indian Point Unit 2 ESEP HCLPF Calculation
-Battery Chargers 21-24." d. AREVA Calculation 32-9227719-000, "Indian Point Unit 2 ESEP HCLPF Calculation
-125 VDC Power Panels EPB3, EPA9, PC1, and PC2." e. AREVA Calculation 32-9229683-000, "Indian Point Unit 2 ESEP HCLPF Calculation
-Battery Room and MCC-24A and 29A Blockwalls." f. AREVA Calculation 32-9230350-000, "Indian Point Unit 2 ESEP HCLPF Calculation
-Condensate Storage Tank." g. AREVA Calculation 32-9230411-000, "Indian Point Unit 2 ESEP HCLPF Calculation
-Refueling Water Storage Tank, O021RWST." h. AREVA Document 32-9232862-000, "Indian Point Unit 2 ESEP Calculation
-Fire Protection Water Storage Tank." AREVA Document 32-9232844-000, "Indian Point Unit 2 ESEP Calculation
-Primary Water Storage Tank." Page 30 ur ~ T Indian Point Unit 2 ESEP Report ATTACHMENT A -INDIAN POINT UNIT 2 ESEL Page A-1 Indian Point Unit 2 ESEP Report ESEL Equipment Operating"State Item Notes/Comments References Number .ID .Description..
Norm alStat'eDesired State 1 MS-45B Steam Generator 22 Safety Closed Open 1 needed for success [7]Relief Valve 2 MS45C Steam Generator 23 Safety Closed Open I needed for success [71 Relief Valve Steam Generator 22 Manual operation with compressed 3 PCV-1135 Atmospheric Dump Valve Closed Open nitrogen bottles [7]4 P Steam Generator 23 Closed Open Manual operation with compressed
[7)Atmospheric Dump Valve nitrogen bottles 5 0022AFP Turbine Driven Auxiliary Feed Standby Operating Auto start [8][9]Pump 22 Manual operation 6 HCV-1118 22 AFW Pump Speed Control Closed Open 118 V Bus 21 [8]Valve 7 PCV-1139 22 AFW Pump Steam Supply Closed Open Panel 21 [7]8 PCV-1213 22 AFW Pump Bearing Cooling Closed Open Part of 22AFW pump skid, fails open [8][9]9 PCV-1310A 22 AFW Pump Steam Isolation Open Open Per Loss of All AC Power procedure
[7]manually open valve 10 PCV-1310B 22 AFW Pump Steam Isolation Open Open Per Loss of All AC Power procedure
[7]manually open valve 11 FT-1201 AFW to Steam Generator 22 Operating Operating
[81[10](12]
Flow Transmitter 12 FT. 1202 AFW to Steam Generator 23 Operating Operating
[8][11][121 Flow Transmitter 13 FJ-1201 AFW to Steam Generator 22 Operating Operating
[10]Flow Indication 14 FlI-1202 AFW to Steam Generator 23 Operating Operating
[ill Flow Indication I I Page A-2 h"!1,rnu'd-_,, Cnrlv Indian Point Unit 2 ESEP Report ESEL Equipment Operating State Item, Notes/Comments References Number, ID Description', Normal State Desired-State 15 INST RK 5 Instrument Rack S Operating Operating
[12][26]Condnse & Fedwter[10][111]
16 PNLSC Condenser
& Feedwater Operating Operating
[Supervisory Panel [13]17 FCV-405B Auxiliary Feed Flow Control Closed Open Emergency procedures include manual [8][10]Valve SG 22 operation with hand-wheel 18 FCV-405C Auxiliary Feed Flow Control Closed Open Emergency procedures include manual [8][111 Valve SG 23 operation with hand-wheel 19 LT-427D Steam Generator 22 WR Level Operating Operating
[8][14][16]
Transmitter Steam Generator 23 WR Level 20 LT-437D transmitt r Operating Operating
[8](15](16]
Transmitter 21 LI-427D Steam Generator 22 WR Level Operating Operating
[141 Indicator 22 LI-437D Steam Generator 23 WR Level Operating Operating
[15]Indicator 23 INST RK 21 Instrument Rack 21 Operating Operating
[14][15][16 1 [26]24 PC-429 Steam Generator 22 Steam Operating Operating
[50]Pressure Controller 25 PC-439 Steam Generator 23 Steam Operating Operating
[501 Pressure Controller 26 PT-429A Steam Generator 22 Steam Operating Operating
[71[121(17]
Pressure Transmitter 27 PT-439A Steam Generator 23 Steam Operating Operating
[71[12181 Pressure Transmitter Steam Generator 22 Steam 28 PI-429A Pressure Indicator Operating Operating
[17 Page A-3 For linfotrmaton On-l~y Indian Point Unit 2 ESEP Report ESEL Equipment Operating State ... .Item' Notes/Comments References, Number ID Description Normal State Desired State " , 29 P 1-439A Steam Generator 23 Steam Operating Operating Pressure Indicator 30 INST RK 9 Instrument Rack 9 Operating Operating
[12][191[26]31 PNL FB Flight Panel FB Operating Operating
[19]32 CST Condensate Storage Tank Available Available
[21][22]33 LT-1128 CST Level Transmitter Operating Operating
[13][21][22][23]34 LI-1128 CST Level indicator Operating Operating
[13][21][22][23]35 MOV-894A Boronated Water Injection Open Open Closed for Phase 2 [241[251 Valve 36 MOV-894B Boronated Water Injection Open Open Closed for Phase 2 [24][25]Valve 37 MOV-894C Boronated Water Injection Open Open Closed for Phase 2 [24][25]valve 38 MOV-894D Boronated Water Injection Open Open Closed for Phase 2 [24][25]Valve 39 MCC-26A 480V Motor Control Center Operating Operating Power to accumulator isolation valve [24][25]MCC-26A 26A 40 MCC-26B 480V Motor Control Center Operating Operating Power to accumulator isolation valve [24][25]26B 41 CH-MOV-222 Seal Water Return Isolation Open Closed Manual operated valve [26][27]ValIve CCW Return from RCP ThermalI 42 FCV-625 Barrier Open Closed Manual operated valve [28]Page A-4 i o n on j Indian Point Unit 2 ESEP Report." .Equipment .,- Operating State Item .Notes/Comments References rNumber ID Desciption Normal State, Desired State 43 0021RWST 21 Refueling Water Storage Available Available
[29]Tank 44 0021RHRHX RHR Heat Exchanger
#21 Available Available Gravity feed path through heat [24]Sexchanger
[24]45 OO22RHRHX RHR Heat Exchanger
#22 Available Available Gravity feed path through heat exchanger
[4 46 HCV-3100 Remote Reactor Head Vent Available Available
[25][301 MOV 47 Remote Reactor Head Vent Available Available
[25][30]MOV [31]Control for remote reactor head vent [31][37]8 PNL SA Panel SA Available Available MOVs 49 EHT PNL 24 Electric Heat Trace Panel 24 Available Available RWST line freeze protection (32]50 MCC-27 480V Motor Control Center 27 Operating Operating Power to EHT PNL 24 [33][39]51Containment Pressure Operating Operating 126]134]Transmitter 52 PI-948A Containment Pressure Operating Operating
[34][351 53 TE-1203-1 Fan Coil Unit 21 Temperature Operating Operating
[26][36]Element 54 TE-1203-2 Fan Coil Unit 22 Temperature Operating Operating
[261f36]Element Fan Coil Unit 23 Temperature 55 TE-1203-3 Elem nt Operating Operating
[26][36]Element 56 TE-1203-4 Fan Coil Unit 24 Temperature Operating Operating
[26][36]Element II_ __Page A-S For Information Only Indian Point Unit 2 ESEP Report ESEL,. .Equipment Operating State Item Notes/Comrnments.,:
:. .'References Number ID Description Normal State Desired State*~~~~~~~~~....
.:. .... .
57Fan Coil Unit 25 Temperature Operating Operating
[26][36]Element Temperature Controller 58 TIC-1203 Containment Average Operating Operating
[26][361 Temperature Containment T-Ave 59 TT-1203 temert Trane Operating Operating
[26][36]Temperature Transmitter 60 TI-1203A Containment Temperature Operating Operating
[36][37]Indicator 61 INST RK 24 Instrument Rack 24 Operating Operating
[26]62 21FOST EDG Fuel Oil Storage Tank 21 Available Available
[381 63 22FOST EDG Fuel Oil Storage Tank 22 Available Available
[381 64 23FOST EDG Fuel Oil Storage Tank 23 Available Available
[38]65 BATT21 Battery Bank Operating Operating
[39][40]66 BATT22 Battery Bank Operating Operating
[39][40]67 BATT23 Battery Bank Operating Operating
[39][40]68 BATT24 Battery Bank Operating Operating
[39][40]69 BATTCHG21 Battery Charger 21 Operating Operating
[39][40]70 BATTCHG22 Battery Charger 22 Operating Operating
[39][40]Page A-6 For itformation, Onldv Indian Point Unit 2 ESEP Report ESEL Equipment Operating State Item Notes/Comments" References Number ID Description Normal State Desired State.71 BATTCHG23 Battery Charger 23 Operating Operating
[391[401 72 BATTCHG24 Battery Charger 24 Operating Operating
[39][40]73 BUS 2A 480VAC Bus 2A Operating Operating
[39][401 74 BUS 3A 480VAC Bus 3A Operating Operating
[391[401 75 BUS 5A 480VAC Bus SA Operating Operating
[39][40]76 BUS 6A 480VAC Bus 6A Operating Operating
[391[401 77 EPA9 125 VDC Power Panel 24 Operating Operating MCC-27A, distribution panel 24AA [39][40]78 EPB3 125 VDC Power Panel 23 Operating Operating MCC-26C, distribution panel 23AA [39][401 79 EPE1 118 VAC Instrument Bus 21A Operating Operating Power for CET [40][41]80 EPE2 118 VAC Instrument Bus 22A Operating Operating
-[40][41]81 EPE3 118 VAC Instrument Bus 23A Operating Operating Power for RVLIS [40][41]82 EPF1 125 VDC Distribution Panel Operating Operating
[26][39]21AA 83 EPF3 125 VDC Distribution Panel Operating Operating
[26][39]84 MCCA M t23AA Operating Operating
[6[39]84 MCC-24A Motor Control Center 24A Operating Operating
[39]Page A-7 Indian Point Unit 2 ESEP Report..ESEL Eq"uipment Operating State Item,*, Notes/Comments References"N~umber.
ID , ' Description NoNrmal State Desired State 85 MCC-29A Motor Control Center 29A Operating Operating
[39]86 PCi 125 VDC Power Panel 21 Operating Operating MCC-29A [26](391 1______ (40](261(39]87 PC2 125 VDC Power Panel 22 Operating Operating MCC-24A [40]88 PC3 125 VDC Distribution Panel 21 Operating Operating
-[26([39]89 PC4 125 VDC Distribution Panel 22 Operating Operating
[26][39]90 PE-6 118 VAC Instrument Bus 24 Operating Operating
[426]39][40]91 PE-7 118 VAC Instrument Bus 23 Operating Operating
[26][39](40]92 PE-8 118 VAC Instrument Bus 21 Operating Operating
[26][39][40]93 PE-9 118 VAC Instrument Bus 22 Operating Operating
[426]39]94 NC-31D Neutron Source Range Operating Operating
[421 Detector 95 NC-32D Neutron Source Range Operating Operating
[421 Detector 96 NC-41D Neutron Source Range Operating Operating
[261 Detector 97 NC-42D Neutron Source Range Operating Operating
[261 Detector 98 NI-31 Source Range Indication Operating Operating
[26]Page A-8
~for~~t~o*n Indian Point Unit 2 ESEP Report ESEL Equipment Operating State Item Notes/Comments References, Number ID ...0 Description Normal State Desired State : A " 99 NI-32 Source Range Indication Operating Operating
-126]100 NI-35 Source Range Indication Operating Operating
[26]101 NI-36 Source Range Indication Operating Operating
[26]102 PNL FC Flight Panel FC Operating Operating
[43]103 NI-41 Source Range Indication Operating Operating
[26]104 NI-42 Source Range Indication Operating Operating
[261 105 PNL FD Flight Panel FD Operating Operating
[44]106 LT-459 Pressurizer Level Transmitter Operating Operating
[45][46]107 LI-459 Pressurizer Level Indicator Operating Operating
[46]108 LT-1311 Reactor Vessel Level Operating Operating RVLIS input [47]Transmitter Narrow RCP WR Pressure Transmitter 109 PT-402 and RVLIS Train A Operating Operating
[261148]110 TE-411A Temperature Element for Operating Operating
[48]RVLIS-Train A RVLIS Cai net RVLIS Cabinet Operating Operating
[48]Cabinet EPH8 112 PNL-D8 Core Exit Thermocouple Operating Operating
[49]Monitoring (D-8) Opraigpeain_[9 Page A-9 For Ofr~it:; rf Indian Point Unit 2 ESEP Report ESEL Equipment Operating State..Item .., * ..... Notes/Comments References Number ID Description Normal Desired State' .113 PM-948A Containment Pressure Operating Operating
[341 Repeater RACK B9 MERLIN ID: Reactor Protection CH IV 114 IP2-CB Instrument Logic Rack B9 Operating Operating
[26][34]CCR Rack B9 115 LI-1311 Reactor Vessel Level Indicator Operating Operating
[47]Narrow PNL A#3 MERLIN ID: 116 IP2-CB Assessment Panel #3 Operating Operating
[26][47]CCR AAS PNL 3 117 PM-429A Steam Generator 22 Steam Operating Operating
[17]Pressure Isolator RACKA3 MERLIN ID: Reactor Protection CH I Operating Operating
[17][20]IP2-CB Instrument Logic Rack A3 [26]CCR Rack A3 Steam Generator 23 Steam 119 PM-439A Pressure Operating Operating
[18]RACK A2 MERLIN ID: Reactor Protection CH I IP2-CB Instrument Logic Rack A2 CCR Rack A2 121 LQM-427D Steam Generator 22 WR Level Operating Operating
[14]RACK B5 122 MERLIN ID: Instrument Rack B5 Operating Operating
[14]P2-C B-53-_____CCR Rack B5 ___________________
Page A-iO Kr~~1 ~t raon Only Indian Point Unit 2 ESEP Report ESEL .Equipment O'perating State.Item ' ." Notes/Comments References Number. ID DescrIption.. " Norimal State Desired State .123 LQM-437D Steam Generator 23 WR Level Operating Operating
[15]RACK BiO MERLIN ID: Reactor Protection CH IV 124 PC ItrmnLoiRakBG Operating Operating
-[15][26]IP2-CB Instrument Logic Rack B10 CCR Rack BiO 125 LM-459A Pressurizer Level Module Operating Operating
[46]RACKA4 MERLIN ID: Reactor Protection CH I IP2-CB Instrument Logic Rack A4 CCR Rack A4 PNLSB-1 MERLIN ID: 127 MERLIN-5D:
Supervisory Panel SB-1 Operating Operating
[26][35]IP2-C B-53-CCR PNL SB-1 PNL SB-2 128 MERLIN ID: Supervisory Panel SB-2 Operating Operating
[26][51]IP2-CB CCR PNL SB-2 129 MS-45-A Steam Generator 21 Safety Closed Open 1 needed for success [7]Relief Valve 130 MS45D Steam Generator 24 Safety Closed Open 1 needed for success [7]Relief Valve Steam Generator 21 131 PCV-1134 Atmospheric Dm Closed Open Manual operation
[6] [7]Atmospheric Dump Valve Steam Generator 24 132 PCV-1137 Atmospheric Dm Closed Open Manual operation
[6] [7]Atmospheric Dump Valve 133 PNL#1 Atmospheric Steam Dump Operating Operating
[26]Panel Page A-11 Indian Point Unit 2 ESEP Report.......................................
... ESEL Equipment Operating State ".: Item Notes/Comrnes
 
==References:==
* Number ID Description Normal State Desired State :, 134 PNL#2 Atmospheric Steam Dump Operating Operating
[26]Panel 135 N2 Tanks Backup Nitrogen Cylinders Intact Intact [52]136 N-854 Nitrogen Pressure Regulator Closed Open [52]137 N-856 Nitrogen Pressure Regulator Closed Open [52]138 FT-1200 AFW to Steam Generator 21 Operating Operating
[8][10][12]
Flow Transmitter 139 FT-1203 AFW to Steam Generator 24 Operating Operating
[8][11][121 Flow Transmitter 140 FlI-1200 AFW to Steam Generator 21 Operating Operating
[10]Flow Indication AFW to Steam Generator 24 141 F1-1203 Flow Indication Operating Operating
[11]142 FCV-405A Auxiliary Feed Flow Control Closed Open Manual operation with handwheel
[8][10]Valve SG 21 143 FCV-405D Auxiliary Feed Flow Control Closed Open Manual operation with handwheel
[8][11]Valve SG 24 144 LT-417D Steam Generator 21 WR Level Operating Operating
[8][14][16]
Transmitter 145 LT-447D Steam Generator 24 WR Level Operating Operating
[8][15][16]
Transmitter Steam Generator 21 WR Level 146 LI-417D Indicator Operating Operating
[14]147 LI-447D Steam Generator 24 WR Level Operating Operating
[15]Indicator Page A-12 Oiily Indian Point Unit 2 ESEP Report ESEL K Equipment .Operating State Item Notes/Comments.
References Number ID .Description NormaI State Desired State K -.148 LQM-417D SG 21 WR Level Operating Operating
[26][56]149 LQM-447D SG 24 WR Level Operating Operating
[26][571 150 PT1 22 AFW Pump Local Control Operating Operating
[14][67]Panel Steam Generator 22 Steam 151 P1-1354 Pressure Indicator Operating Operating
[71[58]152 PI-1355 Steam Generator 23 Steam Operating Operating
[7](581 Pressure Indicator 153 PI-1353 Steam Generator 21 Steam Operating Operating
[7][591 Pressure Indicator 154 PI-1356 Steam Generator 24 Steam Operating Operating
[7][59]Pressure Indicator 155 ACCUM. 21 Accumulator Tank 21 Intact Intact [24][59]156 ACCUM. 22 Accumulator Tank 22 Intact Intact [24][59]157 ACCUM. 23 Accumulator Tank 23 Intact Intact [24][59]158 ACCUM. 24 Accumulator Tank 24 Intact Intact [24][59]159 LCV-459 Letdown Isolation Valve Open Close [53]RHR Purification Line Control 160 CH-HCV-133 VleOpen Close -[53]Valve 161 MOV-882 RHR Pump Suction Isolation Close Open [59][62]Valve Page A-13 Indian Point Unit 2 ESEP Report ESEL Equipment., Operating State. : Item Notes/Comments References Number ID X Description Normal State Desired State: 162 EXC6 Terminal Box Intact Intact [55]163 LT. 1312 Reactor Vessel Level Operating Operating
[54]Transmitter WR 164 LI-1312 Reactor Vessel Level Indicator Operating Operating
[47]WR 165 LIS-1311 Hydraulic Isolators Intact Intact [54][551][59][61]166 LIS-1312 Hydraulic Isolators Intact Intact [54][55][591[611 Reactor Level Hydraulic Operating Operating
[60][611 167 EXG7 Isolator Rack 168 PNL SN Supervisory Panel SN Operating Operating
[54]169 Rack B1 CCR Rack B1 Operating Operating
[26][65]170 MS-577 TDAFWP Turbine Trip Standby Operating
[8][9][59]
[63][661 171 22AFPT TDAFWP Turbine Standby Operating Mounted on TDAFWP skid [8][9][631 172 -TDAFWP Lube Oil Coolers Standby Operating Mounted on TDAFWP skid [8][9][63]
173 PWST Primary Water Storage Tank Intact Intact [59][64]174 300KFPT Fire Water Storage Tank Intact Intact [58]175 TE-1313 Reactor Vessel Upper On On [55][68]Compensation Temperature Page A-14 Indian Point Unit 2 ESEP Report ESEL Equipment
>..* Operating State it.em....
Notes/Conmments.
References Number ID Description
.Normal State Desired State 176 TE-1314 Reactor Vessel Upper On On [55][68]Compensation Temperature 177 TE-1317 Reactor Vessel Conduit On On [55][68]Compensation Temperature 178 TE-1318 Reactor Vessel Conduit On On [55][68]Compensation Temperature 179 TE-1319 Reactor Vessel Lower Tap On On [55][68]180 SP3 Terminal Box Intact Intact [69][70]181 EWV8 Terminal Box Intact Intact [71][72]Page A-1S Or nori m..Indian Point Unit 2 ESEP Report ATTACHMENT B -ESEP HCLPF VALUES AND FAILURE MODES TABULATION Page B-1 Indian Point Unit 2 ESEP Report Ite :Equipment ID Equipment:Description ScreeningL Failur Comments" No. eel mode;1 MS-45B Steam Generator 22 Safety > RLGM Screened Relief Valve 2 MS-45C Steam Generator 23 Safety > RLGM Screened Relief Valve 3 PCV-1135 Steam Generator 22 > RLGM Screened Atmospheric Dump Valve 4 PCV-1136Steam Generator 23 > RLGM Screened Atmospheric Dump Valve 5 O022AFP Turbine Driven Auxiliary
> RLGM Screened Note 1 Feed Pump 22 6 HCV-1118 22 AFW Pump Speed > RLGM Screened Control Valve 7 PCV-1139 22 AFW Pump Steam Supply > RLGM Screened 8 PCV-1213 22 AFW Pump Bearing > RLGM Screened Cooling 9 PCV-1310A 22 AFW Pump Steam > RLGM Screened Isolation 10 PCV-1310B 22 AFW Pump Steam > RLGM Screened Isolation AFW to Steam Generator 22 11 FT-1201 Flo t ransmiter>
RLGM Screened Flow Transmitter 12 FT-1202 AWtStaGerto23
> RLGM Screened Flow Transmitter 13 FI-1201 AFW to Steam Generator 22 > RLGM Screened Flow Indication AFW to Steam Generator 23 14 F1-1202 FoIniaon>
RLGM Screened Flow Indication 15 INST RK 5 Instrument Rack 5 > RLGM Screened Note 1 Condenser
& Feedwater 16 PNL SC > RLGM Screened Note 1 Supervisory Panel 17 FCV-405B Auxiliary Feed Flow Control > RLGM Screened Valve SG 22 18 FCV-405C Auxiliary Feed Flow Control > RLGM Screened Valve SG 23 Steam Generator 22 WR 19 LT-427D TBD TBD Note 3 Level Transmitter 20 LT-437D Steam Generator 23 WR TBD TBD Note 3 Level Transmitter 21 LI-427D Steam Generator 22 WR > RLGM Screened Level Indicator Page B-2 0rL Indian Point Unit 2 ESEP Report HcLPF(g)" Item E."l,.. .*..Failure
..EqNupmentlD
.... Equipment Description Screening.
Mod Comments..
No. ... ..** M odle .: , Level 22 LI-437D Steam Generator 23 WR > RLGM Screened Level Indicator 23 INST RK 21 Instrument Rack 21 TBD TBD Note 3 Steam Generator 22 Steam 24 PC-429 > RLGM Screened Pressure Controller 25 PC-439 Steam Generator 23 Steam > RLGM Screened Pressure Controller 26 PT-429A Steam Generator 22 Steam > RLGM Screened Pressure Transmitter 27 PT-439A Steam Generator 23 Steam > RLGM Screened Pressure Transmitter 28 PI-429A Steam Generator 22 Steam > RLGM Screened Pressure Indicator Steam Generator 23 Steam 29 PI-439A > RLGM Screened Pressure Indicator 30 INST RK 9 Instrument Rack 9 > RLGM Screened Note 1 31 PNL FB Flight Panel FB > RLGM Screened Note 2 32 CST Condensate Storage Tank 0.42 Tank Shell Buckling 33 LT-1128 CST Level Transmitter
> RLGM Screened 34 LI-1128 CST Level indicator
> RLGM Screened 35 MOV-894A Boronated Water Injection TBD TBD Note 3 Valve 36 MOV-894B Boronated Water Injection TBD TBD Note 3 Valve 37 MOV-894C Boronated Water Injection TBD TBD Note 3 valve Boronated Water Injection 38 MOV-894D TBD TBD Note 3 Valve 480V Motor Control Center 39 MCC-26A > RLGM Screened Note 2 26A 480V Motor Control Center 40 MCC-26B > RLGM Screened Note 2 26B 41 CH-MOV-222 Seal Water Return Isolation
> RLGM Screened Valve CCW Return from RCP 42 FCV-625 > RLGM Screened Thermal Barrier Page B-3 a U~t;~ "IJ Indian Point Unit 2 ESEP Report HCLPF (g)- Fu.Item Failure." * " ./: "" .* " ... ..." Ie *.Equipment ID, Equipment Description Screening:.ailur,, .'Comments No... .. .....eeMode: 43 0021RWST 21 Refueling Water Storage 0.18 Buckling of Tank Tank Shell Modifications required.44 O021RHRHX RHR Heat Exchanger
#21 > RLGM Screened Notes 2 and 3 45 O022RHRHX RHR Heat Exchanger
#22 > RLGM Screened Notes 2 and 3 46 HCV-3100 Remote Reactor Head Vent TBD TBD Note 3 MOV 47 HCV-3101 Remote Reactor Head Vent TBD TBD Note 3 47 HCV-3101 MBOTDVot MOV 48 PNLSA PanelSA > RLGM Screened Note 2 49 EHT PNL 24 Electric Heat Trace Panel 24 > RLGM Screened Note 2 480V Motor Control Center 50 MCC-27 > RLGM Screened Note 2 27 51 PT-948A Containment Pressure > RLGM Screened Transmitter Containment Pressure 52 PI-948A > RLGM Screened Indicator 53 TE-1203-1 Fan Coil Unit 21 TBD TBD Note 3 Temperature Element 54 TE-1203-2 Fan Coil Unit 22 TBD TBD Note 3 Temperature Element 55 TE-1203-3 Fan Coil Unit 23 TBD TBD Note 3 Temperature Element 56 TE-1203-4 Fan Coil Unit 24 TBD TBD Note 3 Temperature Element 57 TE-1203-5 Fan Coil Unit 25 TBD TBD Note 3 Temperature Element Temperature Controller 58 TIC-1203 Containment Average > RLGM Screened Temperature 59 TT-1203 Containment T-Ave > RLGM Screened Temperature Transmitter 60 TI-1203A Containment Temperature
> RLGM Screened Indicator 61 INST RK 24 Instrument Rack 24 > RLGM Screened Note 2 62 21FOST EDG Fuel Oil Storage Tank > RLGM Screened 6 21 1 R S Page B-4 Indian Point Unit 2 ESEP Report' " ... HCLPF (g) i .." :. ":. ...* .... ....ID "CLP. .... .' Failure Equipment ID Equipment Description Screening Mode Comments.Level 63 22FOST EDG Fuel Oil Storage Tank > RLGM Screened 22 64 23FOST EDG Fuel Oil Storage Tank > RLGM Screened 23 65 BATT21 Battery Bank 0.31 Blockwall Note 2 66 BATT22 Battery Bank 0.31 Blockwall Note 2 67 BATT23 Battery Bank 0.33 Blockwall Note 1 68 BATT24 Battery Bank 0.31 Blockwall Note 1 69 BATTCHG21 Battery Charger 21 0.3 Anchorage 70 BATTCHG22 Battery Charger 22 0.3 Anchorage 71 BATTCHG23 Battery Charger 23 0.3 Anchorage 72 BATTCHG24 Battery Charger 24 0.3 Anchorage 73 BUS 2A 480VAC Bus 2A 0.3 Anchorage 74 BUS 3A 480VAC Bus 3A 0.3 Anchorage 75 BUS 5A 480VAC Bus 5A 0.3 Anchorage 76 BUS 6A 480VAC Bus 6A 0.3 Anchorage 77 EPA9 125 VDC Power Panel 24 0.38 Blockwall 78 EPB3 125 VDC Power Panel 23 0.38 Blockwall 79 EPE1 118 VAC Instrument Bus 21A > RLGM Screened Note 2 80 EPE2 118 VAC Instrument Bus 22A > RLGM Screened Note 2 81 EPE3 118 VAC Instrument Bus 23A > RLGM Screened Note 2 125 VDC Distribution Panel 82 EPF1 > RLGM Screened Note 2 21AA 125 VDC Distribution Panel 83 EPF3 > RLGM Screened Note 2 23AA Page B-5 Indian Point Unit 2 ESEP Report HLPF (g)/Item H 4  .Failure.No. EquipmentlD Equipment Description Screening Mode C ommensts Level 84 MCC-24A Motor Control Center 24A 0.49 Blockwall Note 1 85 MCC-29A Motor Control Center 29A 0.3 Blockwall Note 1 86 PC1 125 VDC Power Panel 21 0.38 Blockwall 87 PC2 125 VDC Power Panel 22 0.38 Blockwall 125 VDC Distribution Panel > RLGM Screened Note 2 88 P3> RLGM Screened Note 2 21 89 PC4 12 D itiuinPnl
> RLGM Screened Note 2 22 90 PE-6 118 VAC Instrument Bus 24 > RLGM Screened Note 2 91 PE-7 118 VAC Instrument Bus 23 > RLGM Screened Note 2 92 PE-8 118 VAC Instrument Bus 21 > RLGM Screened Note 2 93 PE-9 118 VAC Instrument Bus 22 > RLGM Screened Note 2 94 NC-31D Neutron Source Range TBD TBD Note 3 Detector 95 NC-32D Neutron Source Range TBD TBD Note 3 Detector 96 NC-41D Neutron Source Range TBD TBD Note 3 Detector 97 NC-42D Neutron Source Range TBD TBD Note 3 Detector 98 NI-31 Source Range Indication
> RLGM Screened 99 NI-32 Source Range Indication
> RLGM Screened 100 NI-35 Source Range Indication
> RLGM Screened 101 NI-36 Source Range Indication
> RLGM Screened 102 PNL FC Flight Panel FC > RLGM Screened Note 1 103 NI-41 Source Range Indication
> RLGM Screened 104 NI-42 Source Range Indication
> RLGM Screened Page B-6 I- fl Indian Point Unit 2 ESEP Report Item Equiment D Equ. ...ipmet HCLPF (g) / Failure ItemFalr Equipment ID Equipment Description Screening Mode Comments........M Level 105 PNL FD Flight Panel FD > RLGM Screened Note 1 106 LT-459 Pressurizer Level TBD TBD Note 3 Transmitter 107 LI-459 Pressurizer Level Indicator
> RLGM Screened Reactor Vessel Level 108 LT-1311 > RLGM Screened Transmitter Narrow RCP WR Pressure 109 PT-402 Transmitter and RVLIS Train TBD TBD Note 3 A 110 TE-411A Temperature Element for TBD TBD Note 3 RVLIS-Train A RVLIS Cabinet RVLIS Cabinet > RLGM Screened Note 1 EPH8 112 PNL-D8 Core Exit Thermocouple
> RLGM Screened Note 2 Monitoring (D-8)Containment Pressure 113 PM-948A > RLGM Screened Repeater RACK B9 114 MERLIN ID: IP2- Reactor Protection CH IV > RLGM Screened Note 1 CB-53-CCR Rack Instrument Logic Rack B9 B9 115 LI-1311 Reactor Vessel Level > RLGM Screened Indicator Narrow PNL A#3 MERLIN ID: IP2-116 C CR AAS Assessment Panel #3 > RLGM Screened Note 1 CB-53-CCR AAS PNL3 117 PM-429A Steam Generator 22 Steam > RLGM Screened Pressure Isolator RACKA3 MERLIN ID: 1P2- Reactor Protection CH I 118 > RLGM Screened Note 1 CB-53-CCR Rack Instrument Logic Rack A3 A3 119 PM-439A Steam Generator 23 Steam > RLGM Screened Pressure RACKA2 MERLIN ID: 1P2- Reactor Protection CH I 120 > RLGM Screened Note 1 CB-53-CCR Rack Instrument Logic Rack A2 A2 121 LQM-427D Steam Generator 22 WR > RLGM Screened Level Page B-7 Indian Point Unit 2 ESEP Report HCLPF (g) /Item Failure. Comments Equipment ID Equipment Description Screening ModeComments Level RACK BS MERLIN ID: IP2-122 C2R Rack Instrument Rack B5 > RLGM Screened Note 1 CB-53-CCR Rack B5 Steam Generator 23 WR 123 LQM-437D L evel > RLGM Screened Level RACK B1O MERLIN ID: IP2- Reactor Protection CH IV 124 > RLGM Screened Note 1 CB-53-CCR Rack Instrument Logic Rack BIO B1O 125 LM-459A Pressurizer Level Module > RLGM Screened RACK A4 MERLIN ID: Reactor Protection CH I lP2-CB-53-CCR Instrument Logic Rack A4 Rack A4 PNL SB-1 MERLIN ID: IP2-127 MER PNL Supervisory Panel SB-1 > RLGM Screened Note 1 CB-53-CCR PNL SB-1 PNLSB-2 18 MERLIN ID: 128 IP2-CB-53-CCR Supervisory Panel SB-2 > RLGM Screened Note 1 PNL SB-2 129 MS-45-A Steam Generator 21 Safety >RLGM Screened Relief Valve 130 MS-45-D Steam Generator 24 Safety >RLGM Screened Relief Valve 131 PCV-1134 Steam Generator 21 >RLGM Screened Atmospheric Dump Valve Steam Generator 24 132 PCV-1137 Atmospheric Dm >RLGM Screened Atmospheric Dump Valve 133 PNL #1 Atmospheric Steam Dump >RLGM Screened Note 2 Panel 134 PNL #2 Atmospheric Steam Dump >RLGM Screened Note 2 Panel 135 N2 Tanks Backup Nitrogen Cylinders
>RLGM Screened Note 2 136 N-854 Nitrogen Pressure Regulator
>RLGM Screened 137 N-856 Nitrogen Pressure Regulator
>RLGM Screened 138 FT-1200 AFW to Steam Generator 21 >RLGM Screened Flow Transmitter Page B-8 Indian Point Unit 2 ESEP Report Item FiHCLPFl(g)/
'lure No I Equipment ID Equipment Description Screening
'M Comments No.-ipi~ ':Mode .... .*Level 139 FT-1203 AFW to Steam Generator 24 >RLGM Screened Flow Transmitter 140 FI-.1200 AFW to Steam Generator 21 >RLGM Screened Flow Indication 141 FI-1203 AFW to Steam Generator 24 >RLGM Screened Flow Indication 142 FCV-4O5A Auxiliary Feed Flow Control >RLGM Screened Valve SG 21 143 FCV-405D Auxiliary Feed Flow Control >RLGM Screened Valve SG 24 144 LT-417D Steam Generator 21 WR TBD TBD Note 3 Level Transmitter 145 LT-447D Steam Generator 24 WR TBD TBD Note 3 Level Transmitter Steam Generator 21 WR 146 LI-417D Levendiator
>RLGM Screened Level Indicator 147 LI-447D Steam Generator 24 WR >RLGM Screened Level Indicator 148 LQM-417D SG 21 WR Level >RLGM Screened 149 LQM-447D SG 24 WR Level >RLGM Screened 150 PT1 22 AFW Pump Local Control >RLGM Screened Note 2 Panel 151 P1-1354 Steam Generator 22 Steam >RLGM Screened Pressure Indicator Steam Generator 23 Steam 152 P1-1355 PrsueIdctr>RLGM Screened Pressure Indicator 154 PI-1356 Steam Generator 21 Steam >RLGM Screened Pressure Indicator 154 PI-1356 Sta eeao 4Sem >RLGM Screened Pressure Indicator 155 ACCUM. 21 Accumulator Tank 21 TBD TBD Note 3 156 ACCUM. 22 Accumulator Tank 22 TBD TBD Note 3 157 ACCUM. 23 Accumulator Tank 23 TBD TBD Note 3 158 ACCUM. 24 Accumulator Tank 24 TBD TBD Note 3 159 LCV-459 Letdown Isolation Valve TBD TBD Note 3 Page B-9 Indian Point Unit 2 ESEP Report HCLPF (g)Item .t Failure No. Equipment ID Equipment Description.
Screening
... Moe Comments.Level 5 Mode ,<160 CH-HCV-133 RHR Purification Line TBD TBD Note 3 Control Valve 161 MOV-882 RHR Pump Suction Isolation TBD TBD Note 3 Valve 162 EXC6 Terminal Box TBD TBD Note 3 163 LT-1312 Reactor Vessel Level TBD TBD Note 3 Transmitter WR 164 LI-1312 Reactor Vessel Level >RLGM Screened Indicator WR 165 LIS-1311 Hydraulic Isolators TBD TBD Note 3 166 LIS-1312 Hydraulic Isolators TBD TBD Note 3 167 EXG7 Reactor Level Hydraulic TBD TBD Note 3 isolator Rack 168 PNL SN Supervisory Panel SN >RLGM Screened Note 1 169 Rack B1 CCR Rack B1 >RLGM Screened Note 1 170 MS-577 TDAFWP Turbine Trip >RLGM Screened 171 22AFPT TDAFWP Turbine >RLGM Screened 172 -TDAFWP Lube Oil Coolers >RLGM Screened 173 PWST Primary Water Storage Tank 0.39 Anchorage 174 300KFPT Fire Water Storage Tank 0.17 Anchorage Modifications required.175 TE-1313 Reactor Vessel Upper TBD TBD Note 3 Compensation Temperature 176 TE-1314 Reactor Vessel Upper TBD TBD Note 3 Compensation Temperature 177 TE-1317 ReactorVesselConduit TBD TBD Note 3 Compensation Temperature Reactor Vessel Conduit 178 TE-1318 TBD TBD Note 3 Compensation Temperature 179 TE-1319 Reactor Vessel Lower Tap TBD TBD Note 3 180 SP3 Terminal Box >RLGM Screened Note 2 Page B-10 F ~. ~fo ~t~on Indian Point Unit 2 ESEP Report.tem. HCLPF(g) / Failure .omment, Equipment ID: , Equipment Description Screening Comments No. .. ...Mode"" Level 181 EWV8 Terminal Box TBD TBD Note 3 Notes: 1.2.3.Anchorage screened out based on available margin during walkdown by SRT.Anchorage screened out during walkdown validation by SRT.Inaccessible.
Per EPRI NP-6041-SLR1, Sec. 2, Seismic Capability Walkdown, Step 5 -This component was not walked down.Page B-11 ENCLOSURE 2 TO NL-14-152 EXPEDITED SEISMIC EVALUATION PROCESS REPORT FOR INDIAN POINT UNIT 3 ENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 DOCKET NOS. 50-247 AND 50-286 ATAHMENT 9.1F ENGINEERING REPORT COVER SHEET & INSTRUCTIONS SHEET 1 OF 2 Engineering Report No. IP-RPT- 14-00038 Rev 0 Page 1 of 67 Entergy ENTERGY NUCLEAR Engineering Report Cover Sheet Engineering Report Title: Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 3 Engineering Report Type: New [] Revision []Cancelled El Superseded E]Superseded by: Applicable Site(s)IP I l IP2 El ANQI El ANO2 U IP3 0 ECH El JAF El GGNS El PNPS El RBS El vy El WF3 El wPo 11 PLP El EC No. 54071 Report Origin: El Entergy ED Vendor Vendor Document No.: 51-9230673-001 Quality-Related:
El Yes 0 No Prepared by: Areva Responsible Engineer (Print Name/Sign)
Design Verified: Reviewed by: Approved by: N/A.Design Verifier (if required) (Pr'nt Name/Sign)
Frank Madero/,,, ,aZ &#xfd;a Date: 12/18/14 Date: Date: __-/_-//Date: Ri-"v" er (Print ame/S ),n) / 9 Richard Drake/ /__ Z 'W/ sup: >/R 1anagerxlfnL Na&ii/Sign)
EN-DC-147 REV 6 20004-021 (01/30/2014)
AREVA AREVA Inc.Engineering Information Record Document No.: 51 -9230673 -001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 3 Page 1 of 66 AREVADocument No.: 61-923O81n-OIX E 8qedle ~tmfcEveiueUan Piocer6 -(ES P).Repoittfbrlndi Pplint Uhft Safety Related? E E N Vacstbli document estullsh dcaigu ar tealinial reqalremenh EJYBB ENO.Does flu ducumient
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Document No.: 51-9230673-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit. 3 Signature Block (continued)
Project Manager Approval of Customer References (N/A if not applicable)
Name Title (printed or typed) (printed or typed) Signature Datb Jennifer Butler Project Manager 6 12/13/14 Page 3 A AREVA 20004-021 (01/30/2014)
Document No.: 51-9230673-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 3 Record of Revision Revision PageslSections/
No. Paragraphs Changed Brief Description
/ Change Authorization 000 All Initial release 001 Section 2.0 Section 2.0* References were added due to additional components based on new Appendix A -Section 2.0, revision of supporting document.3.0, 3.1, 3.1.3, 4.2, 5.1, 6.0, Appendix A 6.1, 6.2, 6.3.3, 6.4, 6.5, 7.1,
* Section 2.0, 3.0, 3.1, 3.1.3, 4.2, 5.1, 6.0, 6.1, 6.2, 6.3.3, 6.4, 6.5, 7.2, 8.1, 8.2, 8.3, 8.4, 9.0, 7.1, 7.2, 8.1, 8.2, 8.3, and 8.4 were modified to incorporate Entergy Attachment A, and Comments [84] on Revision 0 of the document and updated with Attachment B additional components based on new revision of supporting document.* Attachment A -modified to incorporate Entergy Comments [84]and to update with additional components based on new revision of supporting document.* Attachment B -modified to incorporate Entergy Comments [84]and to update with additional components based on new revision of supporting document.4 t 4 t i i*1 t+ t+ t Page 4 A AREVA Document No.: 51-9230673-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 3 Table of Contents Page S IG N A T U R E B LO C K ................................................................................................................................
2 R E C O R D O F R E V IS IO N ..........................................................................................................................
4 1.0 DOCUMENTATION
......................................................................................................................
6 2 .0 R E F E R E N C E S ..............................................................................................................................
6 APPENDIX A: EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT FOR INDIAN POINT UNIT 3 ........................................................................................
A-1 Page 5 A AR EVA Document No.: 51-9230673-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 3 1.0 DOCUMENTATION This document contains the Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 3. This document is located in Appendix A and is presented in the customer requested format.
 
==2.0 REFERENCES==
 
References identified with an (*) are maintained within Indian Point Unit 3 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, 2012.2. 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 NL-1 3-042 "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. ML13079A348.
: 4. Entergy Letter to U.S. NRC, letter number NL-14-031, "Indian Point Energy Center's Second Six-Month Status Report for the Implementation of Order EA-1 2-049 Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," February 27, 2014, NRC ADAMS Accession No. ML14070A365.
: 5. Entergy Letter to U.S. NRC, letter number NL-14-1 10, "Indian Point Energy Center's Third Six-Month Status Report for the Implementation of Order EA-12-049 Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," August 27, 2014, NRC ADAMS Accession No. ML14251A227.
: 6. *Entergy Engineering Evaluation, EC No. 45874, Revision 1, "FLEX-Beyond Design Basis External Event Phases 1, 11, and III Strategy Development Evaluation." 7. *Entergy Drawing 9321-F-20173, Revision 72, "Flow Diagram Main Steam." 8. *Entergy System Design Description 21.2, Revision 8, "System Description 21.2, Auxiliary Feedwater System." 9. *Entergy Drawing 9321-F-70313, Revision 17, "Auxiliary Boiler Feed Pump Room Instrument Piping Sheet No.1 Instrumentation." 10. *Entergy Drawing 9321-F-27233, Revision 40, "Flow Diagram Nitrogen to Nuclear Equipment." 11. *Entergy Plant Equipment Database for 1P3.12. *Entergy Drawing 9321-F-20193, Revision 62, "Flow Diagram Boiler Feedwater." 13. *Entergy Drawing 9321-LD-72123, Sheet 9, Revision 2, "Aux. F.W. Flow to Steam Generator#31 Loop F-1200 Diagram." Page 6
%1~*J ~.J'U A AR EVA Document No.: 51-9230673-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 3 14. *Entergy Drawing 9321-LD-72123, Sheet 10 Revision 2, "Aux. F.W. Flow to Steam Generator#32 Loop F-1201 Diagram." 15. *Entergy Drawing 9321 -LD-72123, Sheet 11, Revision 2, "Aux. F.W. Flow to Steam Generator#33 Loop F-1202 Diagram." 16. *Entergy Drawing 9321-LD-72123, Sheet 12, Revision 2, "Aux. F.W. Flow to Steam Generator#34 Loop F-1203 Diagram." 17. *Entergy Drawing 9321-F-31673, Revision 28, "Wiring Diagram 480V Switchgear Miscellaneous." 18. *Entergy Drawing 9321-F-70033, Revision 17, "Transmitter Racks Piping Arrangement
-Sheet No. 2 Instrumentation for Indian Point Energy Center Unit No. 3." 19. *Entergy System Design Description 21.1, Revision 4, "System Description 21.1, Steam Generator Water Level Control." 20. *Entergy Drawing 9321-F-70253, Revision 10, "Primary Plant Instrument Piping & Supports -Sheet No. 1 Instrumentation." 21. *Entergy Drawing 9321-H-39903 Sheet 70, Revision 5, "Rack D-9 Layout." 22. *Entergy Drawing 9321-F-32273, Revision 41, "Wiring Diagram Supervisory Control Panel SC." 23. *Entergy Drawing 9321-F-70513, Revision 17, "Transmitter Racks Piping Arrangement-Sheet No. 4 Instrumentation." 24. *Entergy Drawing 9321-F-10023, Revision 22, "Plot Plan." 25. *Entergy Drawing 9321-F-20183 Sheet 1, Revision 63, "Flow Diagram Condensate
& Boiler Feed Pump Suction." 26. *Entergy Drawing 9321-F-27353, Revision 42, "Flow Diagram Safety Injection System Sheet No. 1." 27. *Entergy System Design Description 10.1, Revision 10, "System Description 10.1, Safety Injection System." 28. *Entergy System Design Description 3.0, Revision 8, "System Description 3.0, Chemical and Volume Control System." 29. *Entergy System Design Description 4.2, Revision 7, "System Description 4.2, Residual Heat Removal System." 30. *Entergy System Design Description 1.4, Revision 7, "System Description 1.4, Pressurizer
&Pressurizer Relief Tank." 31. *Entergy Drawing 9321-F-33853, Revision 19, "Electrical Distribution
& Transmission System." 32. *Entergy Drawing 9321-F-30063 Sheet 1, Revision 81, "Single Line Diagram 480V Motor Control Center No.'s 36A, 36B & 36C." 33. *Entergy Drawing 9321-F-27363, Revision 52, Flow Diagram Chemical & Volume Control System Sheet No. 1." 34. *Entergy Drawing 9321-F-27513 Sheet 1, Revision 31, "Flow Diagram Auxiliary Coolant System In PAB & FSB Sheet No. 1." Page 7 A AR EVA Document No.: 51-9230673-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 3 35. *Entergy Drawing 9321-F-27503, Revision 55, "Flow Diagram Safety Injection System Sheet No. 2." 36. *Entergy Drawing 9321-F-27203, Revision 29, "Flow Diagram Auxiliary Coolant System Inside Containment." 37. *Entergy Drawing 9321-F-27473, Revision 43, "Flow Diagram Reactor Coolant System Sheet No. 2." 38. *Entergy Drawing 9321-F-36383, Revision 4, "Miscellaneous Wiring Details RCS-SOV-652, RCS-SOV-653, RCS-SOV-654, & RCS-SOV-655." 39. *Entergy Drawing 9321-F-30053, Revision 72, "Single Line Diagram 480V Motor Control Centers 37, 38, 39, & 311." 40. *Entergy Drawing 9321-F-30083, Revision 60, "Single Line Diagram D.C. System." 41. *Entergy Drawing IP3V-0454-0041, Revision 1, Structural Detail for Seismic Category 1 Instrument Rack." 42. *Entergy Calculation IP-CALC-07-00154, Revision 0, "Containment Atmospheric Temperature." 43. *Entergy Drawing 9321-F-33433, Revision 7, "Containment Parameters System Wiring Diagram." 44. *Entergy Drawing 9321-H-39913 Sheet 8, Revision 8, "External Connection Diagram R.P.S.Rack No. 8 (A-7)." 45. *Entergy Drawing 9321-H-36723, Revision 0, "Cover Plates on Flight Pnl. "FCF" & Supervisory Pnl, "SCF" -Fabrication Mounting Details." 46. *Entergy Drawing 9321-F-20303, Revision 30, "Flow Diagram Fuel Oil to Diesel Generators." 47. *Entergy Drawing 9321-F-39893, Revision 43, "Single Line Diagram 118VAC Instrument Buses 31, 31A, 32, 32A, 33, 33A, 34, & 34A." 48. *Entergy Document IP3-RPT-UNSPEC-02182, Indian Point Three Nuclear Power Plant Individual Plant Examination of External Events," September 1997.49. *Entergy Drawing 9321-F-32723, Revision 28, Wiring Diagram Flight Control Pnl. FCF & FCR." 50. *Entergy Drawing 9321-LD-72453 Sheet 21A, Revision 2, "Overpressurization System Channel 1 Loop P/T-413 Diagram." 51. *Entergy Drawing 9321-LD-72453 Sheet 23A, Revision 2, "Overpressurization System Channel 4 Loop P/T-443 Diagram." 52. *Entergy Drawing 9321-LD-72453 Sheet 21, Revision 3, "Overpressurization System Channel 1 Loop P/T-413 Diagram." 53. *Entergy Drawing 9321-LD-72453 Sheet 23, Revision 3, "Overpressurization System Channel 4 Loop P/T-443 Diagram." 54. *Entergy Drawing 9321-F-27383, Revision 28, "Flow Diagram Reactor Coolant System Sheet No. 1." 55. *Entergy Drawing 9321-LL-36853, Sheet 1, Revision 3, "Schematic Block Diagram Reactor Vessel Level Instrument System Train "A"." Page 8 A AR EVA Document No.: 51-9230673-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 3 56. *Entergy Drawing 9321-F-33203, Sheet 1, Revision 24, "Conduit & Tray Connection Schematic Containment Building." 57. *Entergy Drawing 9321-F-33313, Sheet 2, Revision 6, "Conduit & Tray Connection Schematic Fan House." 58. *Entergy Drawing 9321-F-30793, Revision 50, "Conduit Layout Containment Building Piping Penetration Area -Fan House." 59. *Entergy Drawing 932 1-F-72043, Revision 7, "Containment Building Reactor Vessel Level Instrumentation System Flow Diagram." 60. *Entergy Drawing 9321-F-70283, Revision 25, Containment Building Instrument Arrangement Sheet No. 2 Instrumentation." 61. *Entergy Drawing 9321-F-39933, Revision 19 "Conduit Layout TSI Room, CFM Multiplexer Room, Control Building EL. 53'-0" & Roof El. 72'-7." 62. *Entergy Drawing 9321-F-95273 Sheet 1, Revision 6, "Control Room RVLIS Rack -Train "A" Interconnection Wiring Diagram." 63. *Entergy Procedure 3-ECA-0.0, Revision 9, "Loss of All AC Power." 64. *Entergy Drawing 9321-F-32383, Revision 31, "Wiring Diagram Supervisory Control Panel SB2." 65. *Entergy System Design Description 1.1, Revision 5, "System Description 1.1, Reactor Coolant System." 66. *Entergy Drawing IP3V-0245-0001, Revision 0, "40'-0" OD x 40'-0" High Fire Protection Water Storage Tanks "FP-Tk-1" & "FP-Tk-2" Pipe Support Details." 67. *Entergy System Design Description 18.0, Revision 7, "System Description 18.0, Main and Reheat Steam." 68. 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 2013.69. Entergy Letter NL-14-043, John A. Ventosa to NRC, "Entergy Seismic Hazard and Screening 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 dated March 31, 2014." NRC ADAMS Accession No.ML14099A111.
: 70. *"Indian Point Energy Center Unit 3 Updated Final Safety Analysis Report," Revision 5, Docket No. 50-286, 2013.71. EPRI-NP-6041-SL, "Methodology for Assessment of Nuclear Power Plant Seismic Margin," Revision 1, August 1991.72. EPRI TR-103959, "Methodology for Developing Seismic Fragilities," July 1994.73. NRC NUREG-1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," June 1991.74. SQUG, "Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment, Seismic Qualification Utility Group," Revision 3A, December 2001.Page 9 A AR EVA Document No.: 51-9230673-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 3 75. 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, 2014.76. 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, 2014.77. 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.
: 78. 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, 2013.79. *Entergy Document EC54071, "ESEP Reports," the following AREVA documents are captured in the plant document management system: a. AREVA Document 51-9212951-006, "ESEP Expedited Seismic Equipment List (ESEL) -Indian Point Unit 3." b. AREVA Calculation 32-9227208-000, "Indian Point Unit 3 ESEP HCLPF Calculation
-Station Service Transformers 2, 3, 5, and 6." c. AREVA Calculation 32-9227381-000, "Indian Point Unit 3 ESEP HCLPF Calculation
-Battery Bank 33 (BATT 33)." d. AREVA Calculation 32-9227576-000, "Indian Point Unit 3 ESEP HCLPF Calculation
-Battery Chargers 31, 32, & 34." e. AREVA Calculation 32-9230353-000, "Indian Point Unit 3 ESEP HCLPF Calculation
-Refueling Water Storage Tank, RWST-31." f. AREVA Calculation 32-9230692-000, "Indian Point Unit 3 ESEP HCLPF Calculation
-RCS Overpressure Racks H-1 and H-3." g. AREVA Document 32-9232897-000, "Indian Point Unit 3 ESEP Calculation
-Fire Water Storage Tanks FP-T-1 and FP-T-2." The following references are AREVA references which were used as input for Appendix A.80. AREVA Calculation 32-9224585-002, "Indian Point Unit 3 ESEP Binning and Screening." 81. AREVA Document 51-9230419-001, "Input to Entergy ESEP Report Sections 2 and 3 for Indian Point 3." 82. AREVA Document 51-9227403-000, "Input to Entergy ESEP Report Sections 4 and 5 for Indian Point Unit 3." 83. AREVA Document 51-9230505-000, "Input to Entergy ESEP Report Sections 6, 7, and 8 for Indian Point Unit 3." 84. AREVA Document 38-9232223-000, "Indian Point Unit 3 ESEP Report Comment Resolution Form." Page 10 A Ct i A AR EVA Document No.: 51-9230673-001 Expedited Seismic Evaluation Process (ESEP) Report for Indian Point Unit 3 APPENDIX A: EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT FOR INDIAN POINT UNIT 3 Note: Customer requested formatting begins on the following page.Page A-1 EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT FOR INDIAN POINT UNIT 3 (IP3)Page I Indian Point Unit 3 ESEP Report Table of Contents Page LIST O F TA B LES ....... .....................................................................................................................................
4 LIST O F FIG U R ES ..........................................................................................................................................
5 1.0 PURPOSE AND OBJECTIVE
...............................................................................................................
6 2.0 BRIEF
 
==SUMMARY==
OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES
..................................
6 3.0 EQUIPM ENT SELECTION PROCESS AND ESEL .............................................................................
7 3.1 Equipment Selection Process and ESEL ...........................................................................
8 3.1.1 ESEL Development
..........................................................................................
9 3.1.2 Power Operated Valves ...................................................................................
9 3 .1.3 Pu ll B o xes ...........................................................................................................
10 3.1.4 Termination Cabinets ......................................................................................
10 3.1.5 Critical Instrumentation Indicators
................................................................
10 3.1.6 Phase 2 and 3 Piping Connections
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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) ...............................................................................
14 5,1 Description of RLGM Selected .....................................................................................
14 5.2 Method to Estimate In-Structure Response Spectra (ISRS) ..........................................
16 6.0 SEISMIC MARGIN EVALUATION APPROACH .............................................................................
16 6.1 Summary of Methodologies Used .................................................................................
17 6.2 HCLPF Screening Process ..............................................................................................
17 6.3 Seismic W alkdown Approach .......................................................................................
18 6.3.1 W alkdown Approach ......................................................................................
18 6.3.2 Application of Previous W alkdown Information
............................................
19 6.3.3 Significant W alkdown Findings ......................................................................
19 6.4 HCLPF Calculation Process ............................................................................................
19 6.5 Functional Evaluations of Relays .................................................................................
20 6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes) ......................................
20 7.0 INACCESSIBLE ITEMS .....................................................................................................................
20 7.1 Identification of ESEL Item Inaccessible for W alkdowns ...............................................
20 7.2 Planned W alkdown / Evaluation Schedule / Close Out ...............................................
22 8.0 ESEP CONCLUSIONS AND RESULTS ..........................................................................................
22 8.1 Supporting Information
....................................................................................................
22 Page 2 ly infornn,", I* %Jrl, Indian Point Unit 3 ESEP Report Table of Contents (continued)
Page 8.2 Identification of Planned Modifications
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24 8.3 Modification Implementation Schedule .......................................................................
24 8.4 Summary of Regulatory Commitments
.........................................................................
24 9 .0 R EFER EN C ES ..................................................................................................................................
2 5 ATTACHMENT A -INDIAN POINT UNIT 3 ESEL ....................................................................................
A-1 ATTACHMENT B -ESEP HCLPF VALUES AND FAILURE MODES TABULATION
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B-1 Page 3 Indian Point Unit 3 ESEP Report List of Tables Page TABLE 4-1: GIVIRS FOR INDIAN POINT UNIT 3 ......................................................................................
11 TABLE 4-2: SSE FOR INDIAN POINT UNIT 3 ..........................................................................................
13 TABLE 5-1: RLGM FOR INDIAN POINT UNIT 3 .....................................................................................
14 Page 4 Indian Point Unit 3 ESEP Report List of Figures Page FIGURE 4-1: GM RS FOR INDIAN POINT UNIT 3 ....................................................................................
12 FIGURE 4-2: GMRS TO SSE COMPARISON FOR INDIAN POINT UNIT 3 .................................................
13 FIGURE 5-1: RLGM FOR INDIAN POINT UNIT 3 ...................................................................................
16 Page 5 Indian Point Unit 3 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 subsequent tsunami, the Nuclear Regulatory Commission (NRC)established a Near-Term Task Force (NTTF) to conduct a systematic review of NRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena.
Subsequently, the NRC issued a 50.54(f) letter on March 12, 2012 [1], requesting information to assure that these recommendations are addressed by all U.S. nuclear power plants. The 50.54(f) letter 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 Indian Point Unit 3. 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 Indian Point Unit 3 FLEX strategies for Reactor Core Cooling and Heat Removal, Reactor Inventory Control/Long Term Subcriticality, and Containment Function are summarized below. This summary is derived from the Indian Point Energy Center Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order EA-12-049
[3], and is consistent with the second and third six-month status reports [4][5] and supplemented by supporting FLEX engineering calculations
[6].Core Cooling and Heat Removal The Phase 1 FLEX strategy at Indian Point Unit 3 for this function is to use Atmospheric Dump Valves (ADVs) and Main Steam Safety Valves (MSSVs) to remove heat, with the steam generator being fed by the turbine-driven Auxiliary Feedwater (AFW) pump. Suction for the AFW pump is from the Condensate Storage Tank (CST). Backup nitrogen cylinders are available to support cycling the ADVs.During Phase 2 of the FLEX strategy, portable diesel-driven pumps will be staged to provide makeup to the CST or to the steam generator feedwater pump suction. The diesel-driven steam generator FLEX feed pump will be staged to provide feedwater to steam generators in the event that the turbine-driven AFW pump becomes unavailable.
The Primary Water Storage Tank (PWST) or Fire Water Storage Tanks (FWSTs) will be used as makeup sources to the CST. Diesel fuel for FLEX equipment can be provided from existing onsite Emergency Diesel Generator (EDG) Fuel Oil Storage Tanks.Page 6 I nl Indian Point Unit 3 ESEP Report The key parameters to be monitored are: steam generator level, steam generator pressure, CST level, Reactor Coolant System (RCS) pressure, and RCS temperature.
RCS Inventory Control For At Power modes In Phase 1, plant cooldown and depressurization will occur. Inventory control is achieved via the accumulators.
During Phase 2, to avoid adverse effects on the RCS natural circulation flow, the cold-leg accumulator isolation valves are electrically closed during the cooldown to prevent nitrogen injection into the RCS.A FLEX pump will be used to provide RCS makeup with borated water supplied by the Refueling Water Storage Tank (RWST). To allow borated water injection into the RCS, the reactor head vent can be opened, if necessary, to provide a letdown path.If the Extended Loss of AC Power (ELAP) event occurs during cold weather months when freezing of the RWST could possibly occur, a FLEX diesel generator can be used to repower the Electric Heat Trace (EHT) system.For Shutdown modes In Phase 1, if the refueling canal is full, RCS makeup will be supplied by gravity feed from the RWST.During Phase 2, a FLEX pump will be used to provide RCS makeup from the RWST in the same manner as for the At Power modes.Additional key parameters to be monitored are pressurizer level and reactor vessel level and nuclear instrumentation.
Containment Function Containment function is not expected to be challenged during Phase 1 or Phase 2 for an ELAP event occurring when the plant is in Mode 1-4. Therefore, no FLEX strategy beyond monitoring containment pressure and temperature was developed to support containment function.For Modes 5 and 6, containment pressure could be challenged unless a vent path is established.
Methods to establish a vent path will be used and include: opening penetration UU, which is used during outages as an additional air supply, deflating the sealing ring of the equipment hatch (if installed), or another vent path identified and evaluated.
Supporting Systems Necessary electrical components are outlined in the Indian Point Unit 3 QIP and primarily entail station batteries, Direct Current (DC) buses, distribution panels, inverters, battery chargers, and instrument buses.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 Indian Point Unit 3 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], [38], [39], [40], [41], [42], [43], [44], [45], [46], [47], [48], [49], [50],[51], [52], [53], [54], [55], [56], [57], [58], [59], [60], [61], [62], [63], [64], [65], [66], and [67].Page 7 Indian Point Unit 3 ESEP Report 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 Indian Point Unit 3 OIP in Response to the March 12, 2012, Commission Order EA-12-049
[3] and is consistent with the second and third six-month status report issued to the NRC [4][5]. The OIP provides the Indian Point Unit 3 FLEX mitigation strategy and serves as the basis for equipment selected for the ESEP.The scope of "installed plant equipment" includes equipment relied upon for the FLEX strategies to sustain the critical functions of core cooling and containment integrity consistent with the Indian Point Unit 3 0IP. 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 subcriticality, and containment integrity functions.
Portable and pre-staged FLEX equipment (not permanently installed) are excluded from the ESEL per EPRI 3002000704.
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 in Table 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 Indian Point Unit 3 0IP.2. The scope of components is limited to installed plant equipment, and FLEX connections necessary to implement the Indian Point Unit 3 OIP as described in Section 2 of 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
[2] guidance are:* Structures (e.g. containment, reactor building, control building, auxiliary building, etc.).* Piping, cabling, conduit, HVAC, and their supports.* Manual valves and rupture disks.* Power-operated valves not required to change state as part of the FLEX mitigation strategies.
* Nuclear steam supply system components (e.g. RPV 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.Page 8 Indian Point Unit 3 ESEP Report 3.1.1 ESEL Development The ESEL was developed by reviewing the Indian Point OIP [3], second and third six-month status reports [4][5] to determine the major equipment involved in the FLEX strategies.
Further reviews of plant drawings (e.g., Piping and Instrumentation Diagrams (P&IDs) and Electrical One Line Diagrams)were performed to identify the boundaries of the flow paths to be used in the FLEX strategies and to identify specific components in the flow paths needed to support implementation of the FLEX strategies.
Boundaries were established at an electrical or mechanical isolation device (e.g., isolation amplifier, valve, etc.) in branch circuits / branch lines off the defined strategy electrical or fluid flow path. P&IDs were the primary reference documents used to identify mechanical components and instrumentation.
The flow paths used for FLEX strategies were selected and specific components were identified using detailed equipment and instrument drawings, piping isometrics, electrical schematics and one-line drawings, system descriptions, design basis documents, as necessary.
Cabinets and equipment controls containing relays, contactors, switches, potentiometers, circuit breakers and other electrical and instrumentation that could be affected by high-frequency earthquake motions and that impact the operation of equipment in the ESEL are required to be on the ESEL. These cabinets and components were identified in the ESEL.For each parameter monitored during the FLEX implementation, a single indication was selected for inclusion in the ESEL. 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 not included in 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 as part of the FLEX mitigation strategies 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 Indian Point Unit 3 ESEL for functional failure modes associated with power operated valves:* Power operated valves that remain energized during the ELAP events (such as DC powered valves), were included on the ESEL." Power operated valves not required to change state as part of the FLEX mitigation strategies were not included on the ESEL. The seismic event also causes the ELAP event; therefore, the valves are incapable of spurious operation as they would be de-energized." Power operated valves not required to change state as part of the FLEX mitigation strategies during Phase 1, and are re-energized and operated during subsequent Phase 2 and 3 strategies, were not evaluated for spurious valve operation as the seismic event that caused the ELAP has passed before the valves are re-powered.
Page 9 Indian Point Unit 3 ESEP Report 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
[2].3.1.4 Termination Cabinets Termination cabinets, including cabinets necessary for FLEX Phase 2 and Phase 3 connections, provide consolidated locations for permanently connecting multiple cables. The termination cabinets and the internal connections provide a completely passive function; however, the cabinets are included in the ESEL to ensure industry knowledge on panel/anchorage failure vulnerabilities is addressed.
 
====3.1.5 Critical====
Instrumentation Indicators Critical indicators and recorders are typically physically located on panels/cabinets and are included as separate components; however, seismic evaluation of the instrument indication may be included in the panel/cabinet seismic evaluation (rule-of-the-box).
 
====3.1.6 Phase====
2 and 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 Indian Point Unit 3 0IP as described in Section 2." Item 3 in Section 3.1 also notes that "The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e., either "Primary" or "Back-up/Alternate")." Item 6 in Section 3.1 above goes on to explain that "Piping, cabling, conduit, HVAC, and their supports" are excluded from the ESEL scope in accordance with EPRI 3002000704
[2].Therefore, piping and pipe supports associated with FLEX Phase 2 and Phase 3 connections are excluded from the scope of the ESEP evaluation.
However, any active valves in FLEX Phase 2 and Phase 3 connection flow path are included in the ESEL.3.2 Justification for Use of Equipment That is Not the Primary Means for FLEX Implementation No equipment that was not part of the primary success path was selected for the Indian Point Unit 3 ESEL.4.0 GROUND MOTION RESPONSE SPECTRUM (GMRS)4.1 Plot of GMRS Submitted by the Licensee In accordance with the guidance provided in Section 2.4.2 of the SPID [681 for rock sites, the Safe Shutdown Earthquake (SSE) control point elevation is defined at the top of hard-rock and is applicable at grade in the free field as well as the various foundations elevations
[69]. Table 4-1 shows the GMRS acceleration for a range of spectral frequencies
[69]. The GMRS at the control point is shown in Figure 4-1.Page 10 0 i" n To ror Indian Point Unit 3 ESEP Report Table 4-1: GMRS for Indian Point Unit 3 Frequency GMRS (Hz) (g)100 4.12E-01 90 4.46E-01 80 5.04E-01 70 5.94E-01 60 7.04E-01 50 8.06E-01 45 8.42E-01 40 8.66E-01 35 8.77E-01 30 8.75E-01 25 8.58E-01 20 8.28E-01 15 7.67E-01 12.5 7.17E-01 10 6.48E-01 9 6.04E-01 8 5.55E-01 7 5.02E-01 6 4.46E-01 5 3.85E-01 4 3.14E-01 3 2.36E-01 2.5 1.94E-01 2 1.59E-01 1.5 1.17E-01 1.25 9.42E-02 1 7.04E-02 0.9 6.40E-02 0.8 5.71E-02 0.7 4.99E-02 0.6 4.25E-02 0.5 3.48E-02 Page 11 II ~l ~Indian Point Unit 3 ESEP Report Table 4-1: GMRS for Indian Point Unit 3 (continued)
Frequency GMRS (Hz) (g)0.4 2.78E-02 0.3 2.09E-02 0.2 1.39E-02 0.167 1.16E-02 0.125 8.69E-03 0.1 6.95E-03 GMRS at Control Point for Indian Point Unit 3, 5% Damping 1.00 0.90 0.80 0.70 0.60' 0.50 0.40 0.30 0.20 0.10 0.00 0.1 10 100 Frequency (Hz)Figure 4-1: GMRS for Indian Point Unit 3 4.2 Comparison to SSE The SSE corresponds to a horizontal acceleration of 0.15g [70]. The SSE is defined in the Updated Final Safety Analysis Report [70] in terms of a Peak Ground Acceleration (PGA) and a design response spectrum.
These spectra have been digitized and tabulated
[69]. Table 4-2 shows the spectral acceleration values at selected frequencies for the 5% damped horizontal SSE.Page 12 rc;o (L &#xfd;i!.Indian Point Unit 3 ESEP Report Table 4-2: SSE for Indian Point Unit 3 Frequency Spectral Acceleration (Hz) (g)100 0.15 25 0.15 10 0.168 5 0.228 2.5 0.234 1 0.127 0.5 0.075 GMRS to SSE Comparison for Indian Point Unit 3, 5% Damping 0.90 0.70 0.60 OMS 0.40 0.30 0.20 0.10 0.00 0.1 10 Frequency (Hz}Figure 4-2: GMRS to SSE Comparison for Indian Point Unit 3 The SSE envelops the GMRS for lower frequencies up to nearly 3 Hz. The GMRS exceeds the SSE beyond that point. As the GMRS exceeds the SSE in the 1 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-frequency GMRS exceedances at Low Seismic Hazard Sites and b) Narrow Band Exceedances in the 1 to 1OHz range, provide criteria for accepting specific GMRS exceedances.
However, the GMRS exceedances occur in the frequency range of interest and cannot be characterized as narrow-band exceedances.
Therefore, these special screening considerations do not apply for Indian Point Unit 3 and High Confidence of a Low Probability of Failure (HCLPF) evaluations are to be performed.
Page 13 rn at, a.Indian Point Unit 3 ESEP Report 5.0 REVIEW LEVEL GROUND MOTION (RLGM)5.1 Description of RLGIvM Selected The RLGM is selected based on Approach 1 in Section 4 of EPRI 3002000704
[2]. The RLGM is developed based on the SSE. The maximum GMRS/SSE ratio between I and 10 Hz range occurs at 10 Hz where the ratio is 0.648/0.168
= 3.86. As the maximum ratio of the GMRS to the SSE over the 1 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 IP3 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.Table 5-1: RLGM for Indian Point Unit 3 Frequency RLGM at 5% Damping (Hz) (g)100.00 0.30 33.00 0.30 15.78 0.31 13.65 0.31 12.20 0.32 10.09 0.34 8.43 0.36 7.23 0.38 6.55 0.40 5.89 0.43 5.34 0.44 4.94 0.46 4.59 0.47 4.17 0.48 3.83 0.49 3.67 0.49 3.41 0.50 3.15 0.49 2.89 0.49 2.65 0.48 2.55 0.47 2.24 0.45 2.02 0.42 Page 14 F c~ ~: K ()rrr ~ '~ic'Indian Point Unit 3 ESEP Report Table 5-1: RLGM for Indian Point Unit 3 (continued)
Frequency RLGM at 5% Damping (Hz) (g)1.78 0.40 1.64 0.38 1.57 0.36 1.46 0.34 1.34 0.32 1.23 0.30 1.16 0.28 1.09 0.27 1.01 0.26 0.96 0.24 0.91 0.23 0.86 0.22 0.81 0.21 0.76 0.21 0.74 0.20 0.71 0.19 0.68 0.19 0.64 0.18 0.60 0.17 0.56 0.16 0.53 0.15 0.10 0.14 Page 15 l(ffl Indian Point Unit 3 ESEP Report Review Level Ground Motion (2xSSE) Response Spectra -Horizontal Direction 0.600I I I I. -500o0un~pn 0.50 I I.... ..... ... ....... ... ....... ....... i ..... ........ , ., ....... .. .... ... ..0.40 ....... .............
... .............
.-I 0.30I, I 0.20 4 i I I 0.10 I i i 0.00: i ' : iii 0.10 1.00 10.00 100.00 Frequency (Hz)Figure 5-1: RLGM for Indian Point Unit 3 5.2 Method to Estimate In-Structure Response Spectra (ISRS)The RLGM ISRS for Indian Point Unit 3 are generated by scaling the SSE ISRS [70]. 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 1 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) [71].Page 16
'I ~K~) ~Indian Point Unit 3 ESEP Report 2. Probabilistic approach using the fragility analysis methodology of EPRI TR-103959, Methodology for Developing Seismic Fragilities
[72].6.1 Summary of Methodologies Used Indian Point Unit 3 was classified as a 0.3g full scope plant in NUREG-1407
[73] and performed a SPRA as part of Individual Plant Examination for External Events (IPEEE) program. The SPRA is documented in [48]. Indian Point Unit 3 IPEEE program followed the NUREG-1407 methodology for seismic evaluation with plant seismic walkdowns using the EPRI NP-6041-SL
[71] and Generic Implementation Procedure
[74]. Walkdown efforts were coordinated for evaluations pertaining to the IPEEE and Unresolved Safety Issue (USI) A-46. Section 3.3 and Appendix B of [69] established that in accordance with the criteria established in SPID [68] Section 3.3, the IPEEE and reassessment of IHS are adequate to support screening of the updated seismic hazard for Indian Point Unit 3. Hence, the risk insights obtained from the IPEEE are used to assess risk for ESEP where applicable.
For ESEP, the evaluation consisted of screening walkdowns and HCLPF calculations.
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 2xSSE with a PGA of 0.3g PGA as shown on Figure 5-1.6.2 HCLPF Screening Process For ESEP, the components are screened considering RLGM (2xSSE) with a 0.3g PGA. The screening tables in EPRI NP-6041-SL
[71] are based on ground peak spectral accelerations of 0.8g and 1.2g. These both exceed the RLGM peak spectral acceleration.
The ESEL components were prescreened based on Table 2-4 of EPRI NP-6041-SL.
Additional pre-screening, specifically for anchorage, considered walkdown results and documentation from NTTF 2.3 and SEWS from IPEEE and USI A-46. Equipment anchorage was screened out in cases where previous evaluations showed large available margin against SSE. The remaining components (i.e., components that do not screen out), were identified as requiring HCLPF calculations.
ESEL components were walked down and based on the equipment and anchorage conditions, prescreening decisions were confirmed and a final list of required HCLPF calculations was generated.
Equipment for which the screening caveats were met and for which the anchorage capacity exceeded the RLGM seismic demand are screened out from ESEP seismic capacity determination because the HCLPF capacity exceeds the RLGM.The Indian Point Unit 3 ESEL contains 194 items. Of these, 30 are valves. In accordance with Table 2-4 of EPRI NP-6041-SL, 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. 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.Page 17 Indian Point Unit 3 ESEP Report 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
[71] for the Seismic Margin Assessment process. Pages 2-26 through 2-30 of EPRI NP-6041-SL describe the seismic walkdown criteria, including the following key criteria."The SRT [Seismic Review Team] should "walk by" 100% of all components which are reasonably accessible and in non-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 found so as to enable the inspection to be performed while the plant is in operation.
At least for the one component of each type which is selected, anchorage should be thoroughly inspected.
The walkdown procedure should be performed in an ad hoc manner. For each class of components the SRT should look closely at the first items and compare the field configurations with the construction drawings and/or specifications.
If a one-to-one correspondence is found, then subsequent items do not have to be inspected in as great a detail. Ultimately the walkdown becomes a "walk by" of the component class as the SRT becomes confident that the construction pattern is typical. This procedure for inspection should be repeated for each component class; although, during the actual walkdown the SRT may be inspecting several classes of components in parallel.
If serious exceptions to the drawings or questionable construction practices are found then the system or component class must be inspected in closer detail until the systematic deficiency is defined.The 100% "walk by" is to look for outliers, lack of similarity, anchorage which is different from that shown on drawings or prescribed in criteria for that component, potential SI [Seismic Interaction]
problems, situations that are at odds with the team members' past experience, and any other areas of serious seismic concern. If any such concerns surface, then the limited sample size of one component of each type for thorough inspection will have to be increased.
The increase in sample size which should be inspected will depend upon the number of outliers and different anchorages, etc., which are observed.
It is up to the SRT to ultimately select the sample size since they are the ones who are responsible for the seismic adequacy of all elements which they screen from the margin review. Appendix D gives guidance for sampling selection." Page 18 I Indian Point Unit 3 ESEP Report 6.3.2 Application of Previous Walkdown Information Several ESEL items were previously walked down during the Indian Point Unit 3 seismic IPEEE program, for seismic IPEEE outlier resolutions in accordance with 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 the previous walkdown, 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
[71], no significant outliers or anchorage concerns were identified during the Indian Point Unit 3 seismic walkdowns.
Based on walkdown results, HCLPF capacity evaluations were recommended for the following thirteen (13) components:
* RWST-31, Refueling Water Storage Tank* BATT CHGR 31, Battery Charger 31* BATT CHGR 32, Battery Charger 32* BATT CHGR 34, Battery Charger 34* BATT 33, Battery Bank 33* BUS2A, Bus 2A 480V" BUS3A, Bus3A480V* BUS5A, Bus 5A480V* BUS6A, Bus6A480V* Rack H1, CCR Aux. Panel Analog Rack H1* Rack H3, CCR Aux. Panel Analog Rack H3* FP-T-1, 31 Fire Water Storage Tank* FP-T-2, 32 Fire Water Storage Tank 6.4 HCLPF Calculation Process ESEL items identified for ESEP at Indian Point Unit 3 were evaluated using the criteria in EPRI NP-6041-SI [71] 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, 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 Page 19 Indian Point Unit 3 ESEP Report* 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 six (6) HCLPF calculations were performed to address the thirteen (13) components.
* Calculation "Battery Chargers 31, 32, & 34" addressing three (3) components BATT CHGR 31, BATT CHGR 32 and BATT CHGR 34* Calculation "Battery Bank 33" addressing a single component BATT 33* Calculation "Refueling Water Storage Tank" addressing a single component RWST-31* Calculation "Station Service Transformers 2, 3, 5, and 6" addressing transformers adjacent to our (4) components BUS2A, BUS3A, BUS5A and BUS6A* Calculation "RCS Overpressure Racks H-1 and H-3" addressing two (2) instrument racks HI and H3* Calculation "Fire Water Storage Tanks FP-T-1 and FP-T-2" addressing two (2) components FP-T-1 and FP-T-2 6.5 Functional Evaluations of Relays No seal in/lockout type relays were identified on Indian Point Unit 3 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
[71] 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, the evaluated ESEL components were determined to have adequate capacity for the design basis loads and HCLPF greater than RLGM.7.0 INACCESSIBLE ITEMS 7.1 Identification of ESEL Item Inaccessible for Walkdowns Forty-one (41) components on the ESEL were inaccessible and not walked down since they are located in the Primary Containment Building in a locked high radiation area at the time of the walkdowns and there were no alternate means of evaluating these items:* ACAHRS1, RHR HTEXCH # 31* ACAHRS2, RHR HTEXCH # 32 Page 20 Indian Point Unit 3 ESEP Report" ACCUM 31, Accumulator Tank 31" ACCUM 32, Accumulator Tank 32* ACCUM 33, Accumulator Tank 33" ACCUM 34, Accumulator Tank 34" CH-HCV-133, RHR LP BYPASS Valve" CH-LCV-459, Letdown Isolation Valve* FE1, Preamplifier For NE-31* LT-417D, Steam Generator 31 Level Transmitter
* LT-447D, Steam Generator 34 Level Transmitter" LT-427D, Steam Generator 32 Level Transmitter" LT-437D, Steam Generator 33 Level Transmitter" LT-459, Pressure Level Transmitter
* PT-402, Loop 31 Hot Leg Pressure Transmitter
* PT-413, Loop 31 Hot Leg Pressure Transmitter
* PT-443, Loop 34 Hot Leg Pressure Transmitter
* Rack 19, Instrument Rack* RACK 21, Steam Generators Level Transmitter Rack* RCS-SOV-652, Reactor Head Vent* RCS-SOV-653, Reactor Head Vent* SI-MOV-894A, NO. 31 Accumulator ISOLATION VALVE" SI-MOV-894B, NO. 32 Accumulator Isolation Valve" SI-MOV-894C, NO. 33 Accumulator Isolation Valve" SI-MOV-894D, NO. 34 Accumulator Isolation Valve* TE423A, Temperature Element" TE-1313, Upper Tap Compensation Temperature Element* TE-1314, Upper Tap Compensation Temperature Element* TE-1317, RVWL Conduit Compensation Temperature Element* TE-1318, RVWL Conduit Compensation
* TE-1319, RVWL Lower Tap Capillary Temperature Element* TE-1416-1, Fan 31 Temperature Element* TE-1416-2, Fan 32 Temperature Element* TE-1416-3, Fan 33 Temperature Element Page 21 vi ~ ~Indian Point Unit 3 ESEP Report* TE-1416-4, Fan 34 Temperature Element* TE-1416-5, Fan 35 Temperature Element* TE-413A, RCS Loop 31 Hot Leg Wide Range Temperature Element* TE-423A3, RCS Loop 32 Hot Leg Wide Range Temperature Element" Y32, Terminal Box* Y36,TerminalBox" Y39,TerminalBox Also, the two (2) Hydraulic Isolators, one (1) valve and one (1) Terminal Box listed below were not walked down due to the plant condition (inaccessible due to contamination/high radiation) at the time of the walkdowns.
Subject components were evaluated based on the review of the recent photos of the components and the general area.* LIS-1311, Hydraulic Isolator* LIS-1312, Hydraulic Isolator* MOV-882, RHR Pump Suction Isolation Valve* Y29, Terminal Box 7.2 Planned Walkdown / Evaluation Schedule / Close Out The walkdowns of the inaccessible items identified in Section 7.1 are scheduled to be performed no later than the second planned refueling outage after December 31, 2014.8.0 ESEP CONCLUSIONS AND RESULTS 8.1 Supporting Information Indian Point Unit 3 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 Indian Point Unit 3 response to the NRC's 50.54(f) letter. On March 12, 2014, NEI submitted to the NRC results of a study [76] 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 [69]. 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 [75] concluded 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 Page 22 Indian Point Unit 3 ESEP Report 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 Indian Point Unit 3 was included in the fleet risk evaluation submitted in the March 12, 2014 NEI letter [76]; therefore, the conclusions in the NRC's May 9 letter also apply to Indian Point Unit 3.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 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
* 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 [68]. As identified in the Indian Point Unit 3 Seismic Hazard and GMRS submittal
[69], Indian Point Unit 3 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 Page 23 Indian Point Unit 3 ESEP Report characterization.
Indian Point Unit 3 will complete that evaluation in accordance with the schedule identified in NEI's letter dated April 9, 2013 [77] and endorsed by the NRC in their May 7, 2013 letter[78].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. Subject modifications are planned to provide additional seismic margin such that the HCLPF will exceed the RLGM.* FP-T-1, 31 Fire Water Storage Tank* FP-T-2, 32 Fire Water Storage Tank 8.3 Modification Implementation Schedule Plant modifications described in Section 8.2 will be performed in accordance with the schedule identified in NEI letter dated April 9, 2013 [77], 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, 2014.8.4 Summary of Regulatory Commitments The following actions will be performed as a result of the ESEP.Equipment
......Action # Equipment ID Descript i o in. .Action Description Completion Date Perform seismic walkdowns, No later than the generate HCLPF calculations and end of the second 1 N/A N/A design and implement any planned refueling necessary modifications for plane rfuel inaccessible items listed in outage after Section 7.1 December 31, 2014.2 FP-T 31 Fire Water Modify tank and anchorage such As described in Storage Tank that HCLPF>RLGM Section 8.3 3 32 Fire Water Modify tank and anchorage such As described in Storage Tank that HCLPF>RLGM Section 8.3 Submit a letter to NRC summarizing the HCLPF results of Within 60 days Items 1 through 3 confirming following 4 N/A N/A implementation of the plant completion of ESEP modifications associated with activities, including items 1 through 3.Page 24 Indian Point Unit 3 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, 2012.2. 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 NL-13-042 "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. ML13079A348.
: 4. Entergy Letter to U.S. NRC, letter number NL-14-031, "Indian Point Energy Center's Second Six-Month Status Report for the Implementation of Order EA-12-049 Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," February 27, 2014, NRC ADAMS Accession No. ML14070A365.
: 5. Entergy Letter to U.S. NRC, letter number NL-14-110, "Indian Point Energy Center's Third Six-Month Status Report for the Implementation of Order EA-12-049 Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," August 27, 2014, NRC ADAMS Accession No. ML14251A227.
: 6. Entergy Engineering Evaluation, EC No. 45874, Revision 1, "FLEX-Beyond Design Basis External Event Phases I, II, and III Strategy Development Evaluation." 7. Entergy Drawing 9321-F-20173, Revision 72, "Flow Diagram Main Steam." 8. Entergy System Design Description 21.2, Revision 8, "System Description 21.2, Auxiliary Feedwater System." 9. Entergy Drawing 9321-F-70313, Revision 17, "Auxiliary Boiler Feed Pump Room Instrument Piping Sheet No.1 Instrumentation." 10. Entergy Drawing 9321-F-27233, Revision 40, "Flow Diagram Nitrogen to Nuclear Equipment." 11. Entergy Plant Equipment Database for Indian Point Unit 3.12. Entergy Drawing 9321-F-20193, Revision 62, "Flow Diagram Boiler Feedwater." 13. Entergy Drawing 9321-LD-72123, Sheet 9, Revision 2, "Aux. F.W. Flow to Steam Generator
#31 Loop F-1200 Diagram." 14. Entergy Drawing 9321-LD-72123, Sheet 10 Revision 2, "Aux. F.W. Flow to Steam Generator
#32 Loop F-1201 Diagram." 15. Entergy Drawing 9321-LD-72123, Sheet 11, Revision 2, "Aux. F.W. Flow to Steam Generator
#33 Loop F-1202 Diagram." 16. Entergy Drawing 9321-LD-72123, Sheet 12, Revision 2, "Aux. F.W. Flow to Steam Generator
#34 Loop F-1203 Diagram." 17. Entergy Drawing 9321-F-31673, Revision 28, "Wiring Diagram 480V Switchgear Miscellaneous." Page 25 Yl o r n:,,,,-, i Indian Point Unit 3 ESEP Report 18. Entergy Drawing 9321-F-70033, Revision 17, "Transmitter Racks Piping Arrangement
-Sheet No. 2 Instrumentation for Indian Point Energy Center Unit No. 3." 19. Entergy System Design Description 21.1, Revision 4, "System Description 21.1, Steam Generator Water Level Control." 20. Entergy Drawing 9321-F-70253, Revision 10, "Primary Plant Instrument Piping & Supports -Sheet No. 1 Instrumentation." 21. Entergy Drawing 9321-H-39903 Sheet 70, Revision 5, "Rack D-9 Layout." 22. Entergy Drawing 9321-F-32273, Revision 41, "Wiring Diagram Supervisory Control Panel SC." 23. Entergy Drawing 9321-F-70513, Revision 17, "Transmitter Racks Piping Arrangement
-Sheet No. 4 Instrumentation." 24. Entergy Drawing 9321-F-10023, Revision 22, "Plot Plan." 25. Entergy Drawing 9321-F-20183 Sheet 1, Revision 63, "Flow Diagram Condensate
& Boiler Feed Pump Suction." 26. Entergy Drawing 9321-F-27353, Revision 42, "Flow Diagram Safety Injection System Sheet No.1." 27. Entergy System Design Description 10.1, Revision 10, "System Description 10.1, Safety Injection System." 28. Entergy System Design Description 3.0, Revision 8, "System Description 3.0, Chemical and Volume Control System." 29. Entergy System Design Description 4.2, Revision 7, "System Description 4.2, Residual Heat Removal System." 30. Entergy System Design Description 1.4, Revision 7, "System Description 1.4, Pressurizer
&Pressurizer Relief Tank." 31. Entergy Drawing 9321-F-33853, Revision 19, "Electrical Distribution
& Transmission System." 32. Entergy Drawing 9321-F-30063 Sheet 1, Revision 81, "Single Line Diagram 480V Motor Control Center No.'s 36A, 36B & 36C." 33. Entergy Drawing 9321-F-27363, Revision 52, Flow Diagram Chemical & Volume Control System Sheet No. 1." 34. Entergy Drawing 9321-F-27513 Sheet 1, Revision 31, "Flow Diagram Auxiliary Coolant System In PAB & FSB Sheet No. 1." 35. Entergy Drawing 9321-F-27503, Revision 55, "Flow Diagram Safety Injection System Sheet No.2.1 36. Entergy Drawing 9321-F-27203, Revision 29, "Flow Diagram Auxiliary Coolant System Inside Containment." 37. Entergy Drawing 9321-F-27473, Revision 43, "Flow Diagram Reactor Coolant System Sheet No.2." Page 26 Indian Point Unit 3 ESEP Report 38. Entergy Drawing 9321-F-36383, Revision 4, "Miscellaneous Wiring Details RCS-SOV-652, RCS-SOV-653, RCS-SOV-654, & RCS-SOV-655." 39. Entergy Drawing 9321-F-30053, Revision 72, "Single Line Diagram 480V Motor Control Centers 37, 38, 39, & 311." 40. Entergy Drawing 9321-F-30083, Revision 60, "Single Line Diagram D.C. System." 41. Entergy Drawing IP3V-0454-0041, Revision 1, Structural Detail for Seismic Category 1 Instrument Rack." 42. Entergy Calculation IP-CALC-07-00154, Revision 0, "Containment Atmospheric Temperature." 43. Entergy Drawing 9321-F-33433, Revision 7, "Containment Parameters System Wiring Diagram." 44. Entergy Drawing 9321-H-39913 Sheet 8, Revision 8, "External Connection Diagram R.P.S. Rack No. 8 (A-7)." 45. Entergy Drawing 9321-H-36723, Revision 0, "Cover Plates on Flight Pnl. "FCF" & Supervisory Pnl."SCF" -Fabrication Mounting Details." 46. Entergy Drawing 9321-F-20303, Revision 30, "Flow Diagram Fuel Oil to Diesel Generators." 47. Entergy Drawing 9321-F-39893, Revision 43, "Single Line Diagram 118VAC Instrument Buses 31, 31A, 32, 32A, 33, 33A, 34, & 34A." 48. Entergy Document IP3-RPT-UNSPEC-02182, Indian Point Three Nuclear Power Plant Individual Plant Examination of External Events," September 1997.49. Entergy Drawing 9321-F-32723, Revision 28, Wiring Diagram Flight Control Pnl. FCF & FCR." 50. Entergy Drawing 9321-LD-72453 Sheet 21A, Revision 2, "Overpressurization System Channel 1 Loop P/T-413 Diagram." 51. Entergy Drawing 9321-LD-72453 Sheet 23A, Revision 2, "Overpressurization System Channel 4 Loop P/T-443 Diagram." 52. Entergy Drawing 9321-LD-72453 Sheet 21, Revision 3, "Overpressurization System Channel 1 Loop P/T-413 Diagram." 53. Entergy Drawing 9321-LD-72453 Sheet 23, Revision 3, "Overpressurization System Channel 4 Loop P/T-443 Diagram." 54. Entergy Drawing 9321-F-27383, Revision 28, "Flow Diagram Reactor Coolant System Sheet No.1." 55. Entergy Drawing 9321-LL-36853, Sheet 1, Revision 3, "Schematic Block Diagram Reactor Vessel Level Instrument System Train "A"." 56. Entergy Drawing 9321-F-33203, Sheet 1, Revision 24, "Conduit & Tray Connection Schematic Containment Building." 57. Entergy Drawing 9321-F-33313, Sheet 2, Revision 6, "Conduit & Tray Connection Schematic Fan House." 58. Entergy Drawing 9321-F-30793, Revision 50, "Conduit Layout Containment Building Piping Penetration Area -Fan House." Page 27
-I Indian Point Unit 3 ESEP Report 59. Entergy Drawing 9321-F-72043, Revision 7, "Containment Building Reactor Vessel Level Instrumentation System Flow Diagram." 60. Entergy Drawing 9321-F-70283, Revision 25, Containment Building Instrument Arrangement Sheet No. 2 Instrumentation." 61. Entergy Drawing 9321-F-39933, Revision 19 "Conduit Layout TSI Room, CFM Multiplexer Room, Control Building EL. 53'-0" & Roof El. 72'-7." 62. Entergy Drawing 9321-F-95273 Sheet 1, Revision 6, "Control Room RVLIS Rack -Train "A" Interconnection Wiring Diagram." 63. Entergy Procedure 3-ECA-0.0, Revision 9, "Loss of All AC Power." 64. Entergy Drawing 9321-F-32383, Revision 31, "Wiring Diagram Supervisory Control Panel SB2." 65. Entergy System Design Description 1.1, Revision 5, "System Description 1.1, Reactor Coolant System." 66. Entergy Drawing IP3V-0245-0001, Revision 0, "40'-0" OD x 40'-0" High Fire Protection Water Storage Tanks "FP-Tk-1" & "FP-Tk-2" Pipe Support Details." 67. Entergy System Design Description 18.0, Revision 7, "System Description 18.0, Main and Reheat Steam." 68. 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 2013.69. Entergy Letter NL-14-043, John A. Ventosa to NRC, "Entergy Seismic Hazard and Screening 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 dated March 31, 2014." NRC ADAMS Accession No.ML14099A111.
: 70. "Indian Point Energy Center Unit 3 Updated Final Safety Analysis Report," Revision 5, Docket No. 50-286, 2013.71. EPRI-NP-6041-SL, "Methodology for Assessment of Nuclear Power Plant Seismic Margin," Revision 1, August 1991.72. EPRI TR-103959, "Methodology for Developing Seismic Fragilities," July 1994.73. NRC NUREG-1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," June 1991.74. SQUG, "Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment, Seismic Qualification Utility Group," Revision 3A, December 2001.75. 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.Page 28 Indian Point Unit 3 ESEP Report 76. 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, 2014.77. 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.
: 78. 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, 2013.79. Entergy Document EC54071, "ESEP Reports," the following AREVA documents are captured in the plant document management system: a. AREVA Document 51-9212951-006, "ESEP Expedited Seismic Equipment List (ESEL) -Indian Point Unit 3." b. AREVA Calculation 32-9227208-000, "Indian Point Unit 3 ESEP HCLPF Calculation
-Station Service Transformers 2, 3, 5, and 6." c. AREVA Calculation 32-9227381-000, "Indian Point Unit 3 ESEP HCLPF Calculation
-Battery Bank 33 (BATT 33)." d. AREVA Calculation 32-9227576-000, "Indian Point Unit 3 ESEP HCLPF Calculation
-Battery Chargers 31, 32, & 34." e. AREVA Calculation 32-9230353-000, "Indian Point Unit 3 ESEP HCLPF Calculation
-Refueling Water Storage Tank, RWST-31." f. AREVA Calculation 32-9230692-000, "Indian Point Unit 3 ESEP HCLPF Calculation
-RCS Overpressure Racks H-1 and H-3." g. AREVA Document 32-9232897-000, "Indian Point Unit 3 ESEP Calculation
-Fire Water Storage Tanks FP-T-1 and FP-T-2." Page 29 0 r il"IGi Indian Point Unit 3 ESEP Report ATTACHMENT A -INDIAN POINT UNIT 3 ESEL Page A-1 0 f mto~Ol Indian Point Unit 3 ESEP Report Equipment
""Operiting State, ....~ESEL Itemi~.Num&#xfd;be.ID .. Description; Norm6l State Desired State >Notes/Comments
*. .References 1 MS452 Steam Generator 32 Safety Closed Open [7]Relief Valve 2 MS-45-3 Steam Generator 33 Safety Closed Open [7]Relief Valve ATM Steam Relief Valve 32 3 PCV-1135 Steam Generator Closed Cycled [7]4 PCV-1136 ATM Steam Relief Valve 33 Closed Cycled [7]Steam Generator 5 PNL#1 ATM Steam Dump Panel #1 On On [9][67]6 PNL #2 ATM Steam Dump Panel #2 On On [9][67]7 32 ABFP Turbine Driven Auxiliary Standby Operating
[7][12](8]
Feedwater Pump No. 32 8 BFD-PCV-1213 Pressure Control Valve for Closed Open [12][8)Bearing Cooler 9 HCV-1118 32 AFW Pump Turbine Closed Open [12][8]Governor 10 MS-PCV.1139 Main Steam to AFW Turbine Closed Open [71[8]PCV 32 Auxiliary Boiler Feed Fails open on loss of instrument air or 11 MS-PCV-1310A Pump Steam Supply First Closed Open manually open per procedure
[7][63][8]
Isolation 32 Auxiliary Boiler Feed Fails open on loss of instrument air or 12 MS-PCV-1310B Pump Steam Supply Second Closed Open manually open per procedure
[7][63][8]
Isolation 13 PNL PT2 Auxiliary Boiler Feed Pump Off On [17][8]Control Station Page A-2 Indian Point Unit 3 ESEP Report ESLIe Equipment Operating State .................
ID Description Normal State Desired State ...-......_,,... : .Notes/CommentsI "References 14 FT-1201 AFW to SG 32 Flow Off On [12][141 Transmitter 15 FT-1202 AFW to SG 33 Flow Off On [12][15]Transmitter 16 FI-1201 AFW to SG 32 Flow Indicator On On [12][14]17 FI-1202 AFW to SG 33 Flow Indicator On On [12](15]18 RACK 26 Pressure Transmitter Rack On On [18][13][14]
#26 [15)[16][21]19 RACK D-9 CCR Rack 9" (NIS MISC On On [13][14][15]
Instrument)
[161 PNLSC [22]20 Supervisory Condenser
& Feedwater On On [13][14][15]
Panel Supervisory Panel [16]21 BFD-FCV-405B No. 32 AFW Pump Manual Closed Open Per Local Equipment Procedure
[12]18]Flow Control to 32 SG Manually Open Valves 22 BFD-FCV-405C No. 32 AFW Pump Manual Closed Open Per Local Equipment Procedure
[12][8]Flow Control to 33 SG Manually Open Valves 23 LT-427D SG 32 Level Transmitter On On [12][23][191 1 [11][201 24 LT-437D SG 33 Level Transmitter On On [12][23][19]
[11][20]25 LI-427D SG 32 Level Indicator On On [19]26 LI-437D SG 33 Level Indicator On On [19]Page A-3
~or !nfc~matVn On~y Indian Point Unit 3 ESEP Report Equipment Operating State* ESEL Item Number ..ID. Description Normal State Desired State* ." ... ... *Notes/Comnments*
.References*
Steam Generators Level*27 RACK 21 Trnmte akOn On [23][201 Transmitter Rack 28 PT-429C SG 32 Steam Pressure On On [7][18]Transmitter 29 PT-439C SG 33 Steam Pressure On On [7181 Transmitter 30 P1354 SG 32 Steam Pressure On On Indication in AFW Pump Local Control Indicator Station 31 P1355 SG 33 Steam Pressure On On Indication in AFW Pump Local Control (7]Indicator Station 32 RACK 9 Instrument Rack 9 Operating Operating
[18]33 CST Condensate Storage Tank Available Available
[24][251 34 LI 1102-S CST Level Indicator Operating Operating
[25]35 LCV-1158-2 CST Low Level Control Valve Open Closed [25]CST to Condensers Level 36 LCV-1158-1 Control Valve Open Closed [25]No. 31 Accumulator Isolation 37 SI-MOV-894A No. Open Closed [26][27]Valve No. 32 Accumulator Isolation 38 SI-MOV-894B No. Open Closed [26][27]Valve No. 33 Accumulator Isolation 39 SI-MOV-894C No. Open Closed [26][27]Valve No. 34 Accumulator Isolation 40 SI-MOV-8940 No. Open Closed [26][27]Valve Page A-4 For nlni"T, o ~Indian Point Unit 3 ESEP Report ESEL Item .."Equipment Operating State... ." ..Number N be ,,ID ...... .Description
: NormalState Desired State , N s m .*e s..* ,,
* * .. .... .... .... " ... "..... ... ..... Notes/Comments>..*
: : References 41 36AMCC Primary Auxiliary Building On On [31][321 Motor Control Center 36A 42 36BMCC Primary Auxiliary Building On On [31][32]Motor Control Center 36B 43RCP Seal Water Return Per Loss of All AC Power Procedure
[3311631 Isolation Valve Manually Close Valves Thermal Barrier Isolation Per Loss of All AC Power Procedure
[34][63]4 AC-FCV-6 TherValve Open Closed Manually Close Valves 45 RWST-31 Refueling Water Storage Available Available
-[35]Tank 46 ACAHRS1 RHR HTEXCH # 31 Intact Intact Gravity feed path from RWST through [36][29]heat exchanger 47 ACAHRS2 RHR HTEXCH # 32 Intact Intact Gravity feed path from RWST through [36](29]heat exchanger 48 RCS-SOV-652 Reactor Head Vent Closed Open 125VDC Distribution Panel 32A [37][38][65]
49 RCS-SOV-653 Reactor Head Vent Closed Open 125VDC Distribution Panel 32A [37][38][65]
50 PNL SBF-1 Supervisory Panel SBF1 On On -[38]51 EHT Panel 34 Electric Heat Trace Panel 34 On On Powered by MCC 37 [39]Primary Auxiliary Building 480VAC Bus 6A feeds to Primary 52 37MCC Motor Control Center 37 On On Auxiliary Building 480V MCC 37 which [31][39]powers Battery Charger 32 53 PNLK49 125 VDC Distribution Panel On On [31][40]32A 54 PT-1421 CTMT Pressure Transmitter On On (35][41]Page A-5 I'r c i &#xfd;71afon Only Indian Point Unit 3 ESEP Report Eie " Operating.State
: Number I.AD Description Desired State .. som t ..*e',.. '._:.. * """_""' RensNotes/CommentsR r 55 PR-1421 CTMT Pressure Recorder On On [43]56 TE-1416-1 Fan 31 Temperature Element On On [44][421 57 TE-1416-2 Fan 32 Temperature Element On On [44](42]58 TE-1416-3 Fan 33 Temperature Element On On [44][42]59 TE-1416-4 Fan 34 Temperature Element On On [44][421 60 TE-1416-5 Fan 35 Temperature Element On On [44][421 61 RACK 24A Instrument Rack On On [41]EDG-31-FO-62 STN Fuel Oil Storage Tank 31 Available Available
[46][48]STN K 63 EDG-32-FO-Fuel Oil Storage Tank 32 Available Available
[46][48]STNK 64 EDG-33-FO-Fuel Oil Storage Tank 33 Available Available
[46][481 STN K 65 311B 118V AC Instrument Bus 31 On On [31][47)Channel II 66 32113 118V AC Instrument Bus 32 On On [31][47]Channel I* 67 341B 118V AC Instrument Bus 34 On On [31][47]Channel III I n I n_[[47 68 3318 118V AC Instrument Bus 33 On On [31][47]Channel IV Page A-6
:or Information
()Wd Indian Point Unit 3 ESEP Report"..........
... "'Equipment Operating State :..ESEL Item i.I.... ..i1 ..",.* .. (,: .Nube ..ID Description.
Normal State. Desired State Notes/Comments Reeene* ...... .** .. .. * ... .... ..." !':..i,..Notes/iCom ments .... :.Referen e~s*69 31AIB 118V AC Instrument Bus 31A On On [31][471 Channel II 70 32AIB 118V AC Instrument Bus 32A On On [311[471 Channel I 71 34AIB 118V AC Instrument Bus 34A On On [31][471 Channel III 72 33AIB 118V AC Instrument Bus 33A On On -31][47]Channel IV 73 34 INVERTER Static Inverter 34 On On [31][40][48]
74 33 INVERTER Static Inverter 33 On On [31][40][48]
75 31 INVERTER Static Inverter 31 On On [31][40][481 76 32 INVERTER Static Inverter 32 On On [31][40][48]
77 31DP 125VDC Distribution Panel 31 On On [31][40][48]
78 32DP 125VDC Distribution Panel 32 On On [31][40]79 33DP 125VDC Distribution Panel 33 On On [31][40][48]
80 34DP 125VDC Distribution Panel 34 On On [31][40][481 81 PNL K48 125 VDC Distribution Panel On On [31][40][48]
31A 82 31PP 125VDC Power Panel 31 On On [31][40][48]
Page A-7 iation Only Indian Point Unit 3 ESEP Report S" Equipment.
Operating State..ESEL Item ._" .* ' ,. ...ID Description..
Normal State Desired State S.. .. ,..._ Notes/Comments References 83 32PP 125VDC Power Panel 32 On On [31][40][48]
84 33PP 125VDC Power Panel 33 On On [31][40][48]
85 34PP 125VDC Power Panel 34 On On [31][40][48]
86 BATT CHGR 31 Battery Charger 31 On On [31][40]87 BATn CHGR 32 Battery Charger 32 On On [31][40]88 BATT CHGR 33 Battery Charger 33 On On [31][40]89 BATT CHGR 34 Battery Charger 34 On On [31][40]90 BATn 31 Battery Bank 31 On On [31][40]91 BATI 32 Battery Bank 32 On On [31][40]92 BATT 34 Battery Bank 34 On On [31][40]93 BAnT 33 Battery Bank 33 On On [31][40]94 BUS2A 480V(SWGR
: 31) Bus 2A On On [31]95 BUS3A 480V(SWGR
: 32) Bus 3A On On [31]96 BUSSA 480V(SWGR31)
Bus 5A On On [31]Page A-8 For 'iniration ODn y Indian Point Unit 3 ESEP Report'SE "te Equipment Operating State Number ID ..Description.
Normal State Desired State: Notes/Comments.
.. References 97 BUS6A 480V(SWGR
: 32) Bus 6A On On [31]98 36CMCC Primary Auxiliary Building On On [31]Motor Control Center 36C 99 39MCC Control Building Motor On On [31]Control Center 39 100 32MCC Turbine-Generator Building On On [31]Motor Control Center 32 101 NI 31B Source Range Count Rate On On [48]Meter 102 NI 31D Source Range Count Rate On On [48]Meter 103 FE1 Preamplifier For NE-31 On On [48]104 PNL FCF Flight Control Panel FC On On [48][49]105 PI-413K Loop 31 Hot Leg Pressure On On [501 Indicator 106 P-443K Loop 34 Hot Leg Pressure On On [51]Indicator 107 PT-413 Loop 31 Hot Leg Pressure On On [52][54]Transmitter 108 PT443 Loop 34 Hot Leg Pressure On On [53][54]Transmitter 109 TE-1313 Upper Tap Compensation On On [59]Temperature Element 110 TE-1314 Upper Tap Compensation On On [59]Temperature Element Page A-9 iron,-Otion Onl~y Indian Point Unit 3 ESEP Report.t Equip'ment Operating State.Number ...ID Description Normal State Desired State*.. .Notes/Comments
.. ..References.
111 TE-1317 RVWL Conduit Compensation On On [59]Temperature Element 112 TE-1318 RVWL Conduit Compensation On On [59]113 TE-1319 RVWL Lower Tap Capillary On On [59]Temperature Element 114 TE-413A RCS Loop 31 Hot Leg Wide On On [60]Range Temperature Element 115 TE-423A3 RCS Loop 32 Hot Leg Wide n On [48][601 Range Temperature Element 116 CAB JR9 RVLIS Cabinet On On [48][61][62]
117 L1-1311 RVWL Narrow Range On On [48]Indicator 118 LI-1312 RVWL Wide Range Indicator On On [48]119 LT-1311 Reactor Vessel Level On On [48][59]Transmitter Narrow Range Reactor Vessel Level 120 LT-1312 Ranttr Narrow Rane On On [48][59]Transmitter Narrow Range 121 TI-1416 Containment Atmospheric On On [44]Temperature Indicator 122 PM-413K Loop 31 Hot Leg Pressure On On [50]Containment Parameters 123 31AIB-2 (J01) Recording Cabinet (Channel On On [43]124 PM-443K Loop 34 Hot Leg Pressure On On [51]Page A-10 For Information Only Indian Point Unit 3 ESEP Report Equipment Operating State .ESEL Item .....__...__.._..____.__.__"_._____.,___'_
Number ID Description Normal State Desired State.. Notes/Comments..
Ntes/ ommReferences.." Not:s..omm,.ts References.
125 PC-413 Loop 31 Hot Leg Pressure On On [521 126 PC-443 Loop 34 Hot Leg Pressure On On [53]127 PQ-413 Loop 31 Hot Leg Pressure On On [52]128 PQ-443 Loop 34 Hot Leg Pressure On On [53]129 PNL SBF-2 Supervisory Control Panel SB- On On [641 2 130 RACK Hi CCR Auxiliary Panel Analog On On [52]Rack H1 131 RACK H3 CCR Auxiliary Panel Analog On On [53]Rack H3 132 Panel SFF Supervisory Panel SF On On [50][51]RVLIS RACK 133 TRAIN A RVLIS Rack Train A On On [48][59]134 MS-45-1 Steam Generator 31 Safety Closed Open [7]Relief Valve 135 MS-45-4 Steam Generator 34 Safety Closed Open [71 Relief Valve 136 PCV-1134 ATM Steam Relief Valve 31 Closed Cycled Manual Operation
[7]Steam Generator 137 PCV-1137 ATM Steam Relief Valve 34 Closed Cycled Manual Operation
[7]Steam Generator 138 N2 TANKS Backup Nitrogen Cylinders Intact Intact -[67][9]Page A-11 For Information Only Indian Point Unit 3 ESEP Report.......> 4.. .. ... , ,. ,' Vo, : : ,! .'. '" 4" : .. .ESE Item ..Equipment Operating State :. , .. ... -:." ESEL Item. .*. ..* ., ...& .Number "" .ID Description Normal State Desired State ... , .... >... ..... : "Notes/Comments.
... ..References 139 IA-PCV-1278 Nitrogen Pressure Regulator Closed Open [67][9][10]
[11]140 IA-PCV-1277 Nitrogen Pressure Regulator Closed Open [67][9][101
___________[11]
141 MS-577 32 AFW Pump Overspeed Intact Intact Part of the TDAFWP Skid [8][11]Trip & Governor Valve 142 FT-1200 AFW to SG 31 Flow Off On [12][13]Transmitter 143 FT-1203 AFW to SG 34 Flow Off On [12161 Transmitter 144 FI-1200 AFW to SG 31 Flow Indicator On On [12][13]145 FI-1203 AFW to SG 34 Flow Indicator On On [12][161 146 BFD-FCV-405A No. 32 AFW Pump Manual Closed Open Manually Open Valves (121 Flow Control to 31 SG 147 BFD-FCV-405D No. 32 AFW Pump Manual Closed Open Manually Open Valves [12]Flow Control to 34 SG 148 LT-417D SG 31 Level Transmitter On On [12][23][19]
[11][20]149 LT-447D SG 34 Level Transmitter On On [12][23][191
[11][20]150 LI-417D SG 31 Level Indicator On On [19]151 LI-447D SG 34 Level Indicator On On [19]152 LQ-417D Steam Generator Level On On [11]Page A-12 I.rnforrnation Only Indian Point Unit 3 ESEP Report'Eupment .Operating State ... ... .. -.. ...ESEL Item. .....Number ,.. .D** ... .... .ID Description Normal State Desired State ,,,,~Notes/iComments 5 References.
153 LQ-427D Steam Generator Level On On [11]154 LQ-437D Steam Generator Level On On [11]155 LQ-447D Steam Generator Level On On f11]156 RACK BlO Instrument Rack On On [11]157 RACK B5 Instrument Rack On On [11l SG 31 Steam Pressure 158 P1-1353 Indicato r On On Local Indication
[7]indicator 159 P1-1356 SG 34 Steam Pressure On On Local Indication
[7]Indicator 160 ACCUM. 31 Accumulator Tank 31 Intact Intact [26][27][11]
161 ACCUM. 32 Accumulator Tank 32 Intact Intact [26][27][111 162 ACCUM. 33 Accumulator Tank 33 Intact Intact [26][27][11]
163 ACCUM. 34 Accumulator Tank 34 Intact Intact [26][27][11]
164 CH-LCV-459 Letdown Isolation Valve Open Closed Fail Closed on Loss of instrument air [28][111 165 CH-HCV-133 RHR LP Bypass Valve Open Closed -[28][11]166 MOV882 RHR Pump Suction Isolation Closed Open Powered from MCC 36B [29]Valve Page A-13 For information Onvy Indian Point Unit 3 ESEP Report Equipment Operating State ..I:ESEL.Item
* J  Number ID Description Normal State Desired State Notes/Comments References 167 LT-459 Pressurizer Level Transmitter On On [30]168 LI-459A Pressurizer Level Indicator On On [30]169 LM-459A Pressurizer Level On On [11]170 Rack A4 CCR Rack A4 On On [11]171 Rack 19 Instrument Rack On On [11]172 Y39 Terminal Box Intact Intact [55]173 Y32 Terminal Box Intact Intact [55]174 Y36 Terminal Box Intact Intact [55][56]175 LIS-1311 Hydraulic Isolators Intact Intact [55][11]176 LIS-1312 Hydraulic Isolators Intact Intact [55][11]177 Y29 Terminal Box Intact Intact [55][57][58]
178 PNLSN Supervisory Panel SN (JB9) On On [55]179 PT-402 Pressure Transmitter On On [55]180 TE 423A Temperature Element On On [55]Page A-14 Fo)r fnao 0W Indian Point Unit 3 ESEP Report ESEL: Item , Equipment
.. ..Operating State .ESEL Item ..... .. .'" .. '' , ." , .." ": " ": " Numer ... ...ID Description Normal State Desired State NotesCommets.." Reeecs* .... .. .. ., :..;,N "t...,...m ents: ...
 
==References:==
:
181 RACK 22 (C9) Instrument Rack On On [55][11]182 RACK 6 (A9) Instrument Rack On On [55][11]183 TDAFWP Turbine Standby Operating Part of the TDAFWP skid [7][12][8]
184 TDAFWP Lube Oil Coolers Standby Operating Part of the TDAFWP skid [7][12][8]
185 TC-1416-1 Temperature Converter On On [42]186 TC-1416-2 Temperature Converter On On [42]187 TC-1416-3 Temperature Converter On On [42]188 TC-1416-4 Temperature Converter On On [42]189 TC-1416-5 Temperature Converter On On [42]190 Rack A-7 CCR Rack A-7 On On [42]191 FP-T-1 31 Fire Water Storage Tank Intact Intact [661 192 FP-T-2 32 Fire Water Storage Tank Intact Intact [66]193 PW-S-TK 31 Primary Water Storage Intact Intact [11]Tank 194 Rack B4 CCR Rack B-4 On On [11]Page A-15 Indian Point Unit 3 ESEP Report ATTACHMENT B -ESEP HCLPF VALUES AND FAILURE MODES TABULATION Page B-1 Indian Point Unit 3 ESEP Report.HCLPF (g)Item Failure Equipment ID Equipment Description Screening Comments No. ~~~Mode ~ Cmet Level 'e 1 MS-45-2 Steam Generator 32 >RLGM Screened Safety Relief Valve 2 MS-45-3 Steam Generator 33 >RLGM Screened Safety Relief Valve 3 PCV-1135 ATM Steam Relief Valve >RLGM Screened 32 Steam Generator 4 PCV-1136 ATM Steam Relief Valve >RLGM Screened 33 Steam Generator 5 PNL#1 ATM Steam Dump Panel >RLGM Screened Note 2#1 6 PNL #2 ATM Steam Dump Panel >RLGM Screened Note 2#2 7 32 ABFP Turbine Driven Auxiliary
>RLGM Screened Note 1 Feedwater Pump No. 32 Pressure Control Valve for 8 BFD-PCV-1213
>RLGM Screened Bearing Cooler 9 HCV-1118 32 AFW Pump Turbine >RLGM Screened Governor 10 MS-PCV-1139 Main Steam to AFW >RLGM Screened Turbine PCV 32 Auxiliary Boiler Feed 11 MS-PCV-1310A Pump Steam Supply First >RLGM Screened Isolation 32 Auxiliary Boiler Feed 12 MS-PCV-1310B Pump Steam Supply >RLGM Screened Second Isolation 13 PNL PT2 Auxiliary Boiler Feed >RLGM Screened Note 1 Pump Control Station 14 FT-1201 AFW to SG 32 Flow >RLGM Screened Transmitter 15 FT-1202 AFW to SG 33 Flow >RLGM Screened Transmitter 16 FI-1201 AFW to SG 32 Flow >RLGM Screened Indicator 17 FI-1202 AFW to SG 33 Flow >RLGM Screened Indicator 18 RACK 26 Pressure Transmitter Rack >RLGM Screened Note 2#26 19 RACK D-9 CCR Rack "D9" (NIS MISC >RLGM Screened Note 2 Instrument)
PNL SC Condenser
& Feedwater 20 Supervisory Supervisory>RLGM Screened Note 2 PanelSPag Page B-2 Indian Point Unit 3 ESEP Report Item Equipment
' HCLPF (g) / Failure No. Equipment ID Equipment Description Screening Mode Comments Level 21 BFD-FCV-405B No. 32 AFW Pump Manual >RLGM Screened Flow Control to 32 SG 22 BFD-FCV-405C No. 32 AFW Pump Manual >RLGM Screened Flow Control to 33 SG 23 LT-427D SG 32 Level Transmitter TBD TBD Note 3 24 LT-437D SG 33 Level Transmitter TBD TBD Note 3 25 LI-427D SG 32 Level Indicator
>RLGM Screened 26 LI-437D SG 33 Level Indicator
>RLGM Screened 27 RACK 21 Steam Generators Level TBD TBD Note 3 Transmitter Rack SG 32 Steam Pressure 28 PT-429C Transmiter
>RLGM Screened Transmitter 29 PT-439C SG 33 Steam Pressure >RLGM Screened Transmitter 30 P-1354SG 32 Steam Pressure >RLGM Screened Indicator SG 33 Steam Pressure 31 P1-1355 Indicat >RLGM Screened Indicator 32 RACK 9 Instrument Rack 9 >RLGM Screened Note 2 33 CST Condensate Storage Tank >RLGM Screened Note 1 34 LI 1102-S CST Level Indicator
>RLGM Screened 35 LCV-1158-2 CST Low Level Control >RLGM Screened Valve 36 LCV-1158-1 CST to Condensers Level >RLGM Screened Control Valve 37 SI-MOV-894A No. 31 Accumulator TBD TBD Note 3 Isolation Valve 38No. 32 Accumulator TBD TBD Note 3 Isolation Valve 39 SI-MOV-894C No. 33 Accumulator TBD TBD Note 3 Isolation Valve 40 SI-MOV-894D No. 34 Accumulator TBD TBD Note 3 Isolation Valve 41 36AMCC Primary Auxiliary Building >RLGM Screened Note 1 Motor Control Center 36A Page B-3 frnofaiio ,n'y Indian Point Unit 3 ESEP Report Item.HCLPF (g) / Failure Equipment ID Equipment Descriptio...n Screening Mode comments, Level 42 36BMCC Primary Auxiliary Building >RLGM Screened Note 1 Motor Control Center 36B 43 CH-MOV-222 RCP Seal Water Return >RLGM Screened Isolation Valve Thermal Barrier Isolation 44 AC-FCV-625 Val >RLGM Screened Valve 45 RWST-31 Refueling Water Storage 0.41 Tank Tank Sloshing 46 ACAHRS1 RHR HTEXCH # 31 TBD TBD Note 3 47 ACAHRS2 RHR HTEXCH # 32 TBD TBD Note 3 48 RCS-SOV-652 Reactor Head Vent TBD TBD Note 3 49 RCS-SOV-653 Reactor Head Vent TBD TBD Note 3 50 PNL SBF-1 Supervisory Panel SBF1 >RLGM Screened Note 2 51 EHTPanel34 Electric Heat Trace Panel >RLGM Screened 34 52 37MCC Primary Auxiliary Building >RLGM Screened Note 1 Motor Control Center 37 53 PNL125 VDC Distribution
>RLGM Screened Note 2 Panel 32A 54 PT-1421 >RLGM Screened Transmitter 55 PR-1421 CTMT Pressure Recorder >RLGM Screened 56 TE-1416-1 Fan 31 Temperature TBD TBD Note 3 Element 57 TE-1416-2 Fan 32 Temperature TBD TBD Note 3 Element 58 TE-1416-4 Fan 33 Temperature TBD TBD Note 3 Element 59 TE-1416-5 Fan 34 Temperature TBD TBD Note 3 Element 60 TE-1416-5 Fan 35 Temperature TBD TBD Note 3 Element 61 RACK 24A Instrument Rack >RLGM Screened Note 2 62 EDG-31-FO-Fuel Oil Storage Tank 31 >RLGM Screened Note 2 Page B-4 FUr lrVorl- ~iinQ Indian Point Unit 3 ESEP Report":. HCLPF (g) a Item creig Failure Ie... Equipment ID Equipment Description MSceenig COmmentsode N o .%,ii ... ,,.... .. .....*... ... ... .: M o d e
.. ./ <3/4 Level<63 EDG-32-FO-Fuel Oil Storage Tank 32 >RLGM Screened Note 2 STNK 64 EDG-33-FO-Fuel Oil Storage Tank 33 >RLGM Screened Note 2 STNK 65 311B 118V AC Instrument Bus >RLGM Screened Note 2 31 Channel II 66 321B 118V AC Instrument Bus >RLGM Screened Note 2 32 Channel 1 67 341B 118V AC Instrument Bus >RLGM Screened Note 2 34 Channel III 68 331B 118V AC Instrument Bus >RLGM Screened Note 2 33 Channel IV 69 31AIB 118V AC Instrument Bus >RLGM Screened Note 2 31A Channel II 70 32AIB118V AC Instrument Bus >RLGM Screened Note 2 32A Channel 1 71 34AIB 118V AC Instrument Bus >RLGM Screened Note 2 34A Channel III 72 33AIB 118V AC Instrument Bus >RLGM Screened Note 2 33A Channel IV 73 34 INVERTER Static Inverter 34 >RLGM Screened Note 1 74 33 INVERTER Static Inverter 33 >RLGM Screened Note 2 75 31 INVERTER Static Inverter 31 >RLGM Screened Note 2 76 32 INVERTER Static Inverter 32 >RLGM Screened Note 2 77 31DP 125VDC Distribution Panel >RLGM Screened Note 2 31 78 32DP 12VDC Distribution Panel >RLGM Screened Note 2 32 79 33DP 125VDC Distribution Panel >RLGM Screened Note 2 33 80 34DP 125VDC Distribution Panel >RLGM Screened Note 2 34 81 PNLK48 125 VDC Distribution
>RLGM Screened Note 2 Panel 31A 82 31PP 125VDC Power Panel 31 >RLGM Screened Note 2 83 32PP 125VDC Power Panel 32 >RLGM Screened Note 2 Page B-5 ia*~iori Cnlv Indian Point Unit 3 ESEP Report Item Equipment HCLPF (g) / Failure Equipment ID. Equipment Description Screening Mod Comments____ ~LeveUI 84 33PP 125VDC Power Panel 33 >RLGM Screened Note 2 85 34PP 125VDC Power Panel 34 >RLGM Screened Note 2 86 BATT CHGR 31 Battery Charger 31 0.36 Anchorage 87 BATT CHGR 32 Battery Charger 32 0.36 Anchorage 88 BATT CHGR 33 Battery Charger 33 >RLGM Screened Note 1 89 BATT CHGR 34 Battery Charger 34 0.36 Anchorage 90 BATT 31 Battery Bank 31 >RLGM Screened Note 1 91 BATT 32 Battery Bank 32 >RLGM Screened Note 1 92 BATT 34 Battery Bank 34 >RLGM Screened Note 2 93 BATT 33 Battery Bank 33 0.41 Anchorage 94 BUS2A 480V(SWGR
: 31) Bus 2A >RLGM Screened Note 2 95 BUS3A 480V(SWGR
: 32) Bus 3A >RLGM Screened Note 2 96 BUS5A 480V(SWGR31)
Bus 5A >RLGM Screened Note 2 97 BUS6A 480V(SWGR
: 32) Bus 6A >RLGM Screened Note 2 98 36CMCC Primary Auxiliary Building >RLGM Screened Note 2 Motor Control Center 36C 99 39MCC Control Building Motor >RLGM Screened Note 2 Control Center 39 Turbine-Generator 100 32MCC Building Motor Control >RLGM Screened Note 1 Center 32 101 NI 31B Source Range Count Rate >RLGM Screened Meter 102 NI 31D Source Range Count Rate >RLGM Screened Meter 103 FE1 Preamplifier For NE-31 TBD TBD Note 3 Page B-6 Fo r m i L 0-,rah,, Indian Point Unit 3 ESEP Report.."' .... HCLPF (g) / ai r ' .C "m:en'.I e .. ..* , ." " ," ....... .* " F a ilu re :. :: *..." ...Item Equipment ID. Equipment Description
...Scr.eening...
Codments No. Equipmen ""Level 104 PNL FCF Flight Control Panel FC >RLGM Screened Note 2 105 PI-413K Loop 31 Hot Leg Pressure >RLGM Screened Indicator 106 PI-443K Loop 34 Hot Leg Pressure >RLGM Screened Indicator 107 PT-413 Loop 31 Hot Leg Pressure TBD TBD Note 3 Transmitter 108 PT-443 Loop 34 Hot Leg Pressure TBD TBD Note 3 Transmitter 109 TE-1313 Upper Tap Compensation TBD TBD Note 3 Temperature Element 110 TE-1314 Upper Tap Compensation TBD TBD Note 3 Temperature Element RVWL Conduit 111 TE-1317 Compensation TBD TBD Note 3 Temperature Element 112 TE-1318 RVWL Conduit TBD TBD Note 3 Compensation 113 TE-1319 RVWL Lower Tap Capillary TBD TBD Note 3 Temperature Element RCS Loop 31 Hot Leg Wide 114 TE-413A Range Temperature TBD TBD Note 3 Element RCS Loop 32 Hot Leg Wide 115 TE-423A3 Range Temperature TBD TBD Note 3 Element 116 CAB JR9 RVLIS Cabinet >RLGM Screened Note 2 117 LI-1311 RVWL Narrow Range >RLGM Screened Indicator 118 LI-1312 RVWL Wide Range >RLGM Screened Indicator 119 LT-1311 Reactor Vessel Level >RLGM Screened Transmitter Narrow Range 120 LT-1312 Reactor Vessel Level >RLGM Screened Transmitter Narrow Range 121 TI-1416 Containment Atmospheric
>RLGM Screened Temperature Indicator 122 PM-413K Loop 31 Hot Leg Pressure >RLGM Screened Page B-7
~1L L >rn? ~Orn1V Indian Point Unit 3 ESEP Report''Item Equipmt HCLPF (g) / Failure Comments Equipment ID Equipment Description Screening:
C t sMode Level Containment Parameters 123 31AIB-2 (J01) Recording Cabinet >RLGM Screened Note 2 (Channel I)124 PM-443K Loop 34 Hot Leg Pressure >RLGM Screened 125 PC-413 Loop 31 Hot Leg Pressure >RLGM Screened 126 PC-443 Loop 34 Hot Leg Pressure >RLGM Screened 127 PQ-413 Loop 31 Hot Leg Pressure >RLGM Screened 128 PQ-443 Loop 34 Hot Leg Pressure >RLGM Screened 129 PNL SBF-2 Supervisory Control Panel >RLGM Screened Note 1 SB-2 130 Rack H1 CCR Auxiliary Panel Analog 0.47 Functional Rack H1 131 Rack H3 CCR Auxiliary Panel Analog 0.47 Functional Rack H3 132 PANEL SFF Supervisory Panel SF >RLGM Screened Note 2 133 RVLIS RACK RVLIS Rack Train A >RLGM Screened Note 2 TRAIN A 134 MS-45-1 Steam Generator 31 >RLGM Screened Safety Relief Valve 135 MS-45-4 Steam Generator 34 >RLGM Screened Safety Relief Valve 136 PCV-1134 ATM Steam Relief Valve >RLGM Screened 31 Steam Generator 137 PCV-1137 ATM Steam Relief Valve >RLGM Screened 34 Steam Generator 138 N2 TANKS Backup Nitrogen Cylinders
>RLGM Screened Note 2 139 IA-PCV-1278 Nitrogen Pressure >RLGM Screened Regulator 140 IA-PCV-1277 Nitrogen Pressure >RLGM Screened Regulator 141 MS-577 32 AFW Pump Overspeed
>RLGM Screened Trip & Governor Valve 142 FT-1200 AFW to SG 31 Flow >RLGM Screened Transmitter Page B-8 iation C.-, I I I '.-)nly Indian Point Unit 3 ESEP Report Item E up etHCLPF (g) Failure .Equipment ID Equipment Description Screening,, Comments:N,:,... ::Level 143 FT-1203 AFW to SG 34 Flow >RLGM Screened Transmitter 144 FI-1200 AFW to SG 31 Flow >RLGM Screened Indicator 145 FI-1203 AFW to SG 34 Flow >RLGM Screened Indicator 146 BFD-FCV-405A No. 32 AFW Pump Manual >RLGM Screened Flow Control to 31 SG 147 BFD-FCV-405D No. 32 AFW Pump Manual >RLGM Screened Flow Control to 34 SG 148 LT-417D SG 31 Level Transmitter TBD TBD Note 3 149 LT-447D SG 34 Level Transmitter TBD TBD Note 3 150 LI-417D SG 31 Level Indicator
>RLGM Screened 151 LI-447D SG 34 Level Indicator
>RLGM Screened 152 LQ-417D Steam Generator Level >RLGM Screened 153 LQ-427D Steam Generator Level >RLGM Screened 154 LQ-437D Steam Generator Level >RLGM Screened 155 LQ-447D Steam Generator Level >RLGM Screened 156 RACK B1O Instrument Rack >RLGM Screened Note 1 157 RACK BS Instrument Rack >RLGM Screened Note 1 SG 31 Steam Pressure 158 P1-1353 Indicat >RLGM Screened Indicator 159 P1-1356 SG 34 Steam Pressure >RLGM Screened Indicator 160 ACCUM. 31 Accumulator Tank 31 TBD TBD Note 3 161 ACCUM. 32 Accumulator Tank 32 TBD TBD Note 3 162 ACCUM. 33 Accumulator Tank 33 TBD TBD Note 3 163 ACCUM. 34 Accumulator Tank 34 TBD TBD Note 3 Page B-9
'I Indian Point Unit 3 ESEP Report*item Equipme......nt.
.*. HCLPF Failure(NIte Equipment.D.
Equipment Description.
Screening Failue Comments Level Mode.164 CH-LCV-459 Letdown Isolation Valve TBD TBD Note 3 165 CH-HCV-133 RHR LP Bypass Valve TBD TBD Note 3 166 MOV-882 RHR Pump Suction >RLGM Screened Isolation Valve 167 LT-459 Pressurizer Level TBD TBD Note 3 Transmitter 168 LI-459A Pressurizer Level Indicator
>RLGM Screened 169 LM-459A Pressurizer Level >RLGM Screened 170 Rack A4 CCR Rack A4 >RLGM Screened Note 1 171 Rack 19 Instrument Rack TBD TBD Note 3 172 Y39 Terminal Box TBD TBD Note 3 173 Y32 Terminal Box TBD TBD Note 3 174 Y36 Terminal Box TBD TBD Note 3 175 LIS-1311 Hydraulic Isolators
>RLGM Screened Note 2 176 LIS-1312 Hydraulic Isolators
>RLGM Screened Note 2 177 Y29 Terminal Box >RLGM Screened Note 2 178 PNL SN Supervisory Panel SN (JB9) >RLGM Screened Note 1 179 PT-402 Pressure Transmitter TBD TBD Note 3 180 TE 423A Temperature Element TBD TBD Note 3 181 RACK 22 (C9) Instrument Rack >RLGM Screened Note 1 182 RACK 6 (A9) Instrument Rack >RLGM Screened Note 1 183 -TDAFWP Turbine >RLGM Screened 184 -TDAFWP Lube Oil Coolers >RLGM Screened Page B-10 Fo1 Information 0nFy Indian Point Unit 3 ESEP Report HCLPF (g)Item Failur Equipment ID Equipment Description Screeningm Comments*NO. Mode N o ... .. .. * ....... ~ ~Leve l-,., .. .:.. .... ... ..,.,. .185 TC-1416-1 Temperature Converter
>RLGM Screened 186 TC-1416-2 Temperature Converter
>RLGM Screened 187 TC-1416-3 Temperature Converter
>RLGM Screened 188 TC-1416-4 Temperature Converter
>RLGM Screened 189 TC-1416-5 Temperature Converter
>RLGM Screened 190 Rack A-7 CCR Rack A-7 >RLGM Screened Note 1 191 FP-T-1 31 Fire Water Storage 0.24 Anchorage Modifications required.Tank 192 FPT2 32 Fire Water Storage 0.24 Anchorage Modifications required.Tank 193 PW-S-TK 31 Primary Water Storage >RLGM Screened Note 1 Tank 194 Rack B4 CCR Rack B-4 >RLGM Screened Note 1 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.Page B-11 Remote User Iglande http:1/w071 Osi 3.enne.entergy.com/pcrs/asp/crdetc 12/22/14 08:24 AM CR Detail Page I of 1 PCRS Condition Summary CR-IP3-2014-3313 PCRS WebLink This report contains only summary information.
Please consult PCRS application for full detail.Discovered datettime:
Originated By: Operability:
Reportability:
Affected Systems: 12/22/2014 12:49:06 AM Dignam,John M ( Operations Shift Staff IP3)EQUIP NON-FUNCTIONAL CR Status: Open Responsible Dept: Classification:
Significance:
Affected Equipment:
Condition
 
==
Description:==
 
The Met tower 10 meter wind direction appears to be stuck at about 175 degrees. See Unit 2 CR IP2-2014-06587 Immediate Action
 
== Description:==
 
Verified wind direction is available through alternate means per IP-EP-510 Suggested Action
 
== Description:==
 
Corrective Actions: No Corrective action has been issued for this CR.Close Window http://w071Os13.enne.entergy.con/pcrs/asp/crdetail-14_3.asp?crid=
11517831 12/22/2014 Tagout Tag Hang List Clearance:
3-PLANT LABEL Tagout: LO 0004 Indian Point Tag Hang Sheet 12/22/2014 07:44:29 rag Type Equipment Ver. Place. Placement Configuration Place. 1st Verif Place. 2nd Verif---.. --------------------------------.
Req. Seq. ----------------------
Date/Time Date/Time Serial No.
* Equipment Description
* Notes* Equipment Location Plant Label 2-LO:' -VALVE -LO-988 ..... ...............
562
* 22 MBFP HIGH PRESSURE STOP VALVE TEST SOLENOID TH -15 Plant Label 2-10 -VALVE -LO-987 563
* 22 MBFP LOW PRESSURE STOP VALVE TEST SOLENOID ...........
"* .._____ TH ~15-- ---Plant Label 2-LO -VALVE -LO-974 .... ...................
...... ... ...564
* 21 MBFP LOW PRESSURE TEST STOP SOLENOID ..TH -15 ....--Plant Label 565 2-LO -VALVE -LO-973 21 MBFP HIGH PRESSURE STOP VLV TEST SOLENOID TH -15 ,- ---*
ATTACHMENT TO NL-14-152 LIST OF REGULATORY COMMITMENTS ENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 DOCKET NOS. 50-247 AND 50-286 NL-14-152 Docket Nos. 50-247 and 50-286 Attachment Page 1 of 2 List of Reaulatorv Commitments The following table identifies those actions committed to by Entergy 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 COMMITMENT ONE- COMPLETION DATE TIE CONTINUING TIME COMPLIANCE (If Required)ACTION Entergy will perform seismic walkdowns at No later than the end IP2 for inaccessible items listed in Section IV] of the first planned aP2 7. refueling outage after December 31, 2014.No later than 90 days Entergy will generate HCLPF calculations following the end of for IP2 inaccessible items listed in Section [V] the first planned IP2 7.1 refueling outage after December 31, 2014.Entergy will implement any necessary IP2 December 31, 2016 or modifications for inaccessible items listed in no later than the end Section 7.1 based on the schedule IV] IP2 refueling outage commitment to complete this activity in NL- after December 31, 13-069 dated April 29, 2013 2014 p mbeNL-13-069 December 3L1, 20169o Entergy will modify the IP2 RWST and Fire December 31,t2016 or Water Storage Tank and anchorages so of the second planned that HCLPF>RLGM based on the schedule [V1P2 refueling outage commitment to complete this activity in NL- after December 31, 13-069 dated April 29, 2013 2014 p mberL-1 , 2014 per NL-13-069 Entergy will submit a letter to NRC summarizing the IP2 HCLPF results and confirming implementation of the plant Within 60 days modifications associated with the IP2 IV] following completion commitments to complete modifications for of IP2 ESEP activities inaccessible items and modifications of the RWST and Fire Water Storage Tank.Entergy will perform seismic walkdowns at No later than the end IP3 for inaccessible items listed in Section [of the first planned P3 7.1 refueling outage after December 31, 2014.No later than 90 days Entergy will generate HCLPF calculations following the end of for IP3 inaccessible items listed in Section [,] the first planned 1P3 7.1 refueling outage after December 31, 2014.
NL-14-152 Docket Nos. 50-247 and 50-286 Attachment Page 2 of 2 Entergy will implement any necessary IP3 December 31, 2016 or modifications for inaccessible items listed in no later than the end Section 7.1 based on the schedule [o'] IP3 refueling outage commitment to complete this activity in NL- after December 31, 13-069 dated April 29, 2013 2014 p mbeNL-13-069 December 31,2016 or Entergy will modify the IP3 Fire Water no later than the end Storage Tanks 31 and 32 and anchorages of the end so that HCLPF>RLGM on the schedule []of the second planned commitment to complete this activity in NL- aP3 refueling outage 13-069 dated April 29, 2013 after December 31, 2014 per NL-13-069 Entergy will submit a letter to NRC summarizing the IP3 HCLPF results and confirming implementation of the plant Within 60 days modifications associated with the IP3 [1] following completion commitments to complete modifications for of IP3 ESEP activities inaccessible items and modifications of the Fire Water Storage Tanks 31 and 32.}}

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