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{{#Wiki_filter:UNITED STATES               NUCLEAR REGULATORY COMMISSION                                                         REGION I                                               475 ALLENDALE ROAD                               KING OF PRUSSIA, PA 19406-1415  
{{#Wiki_filter:UNITED STATES
 
                          NUCLEAR REGULATORY COMMISSION
  May 14, 2009  
                                            REGION I
 
                                        475 ALLENDALE ROAD
Mr. Joseph E. Pollock Site Vice President Entergy Nuclear Operations, Inc. Indian Point Energy Center 450 Broadway, GSB  
                                    KING OF PRUSSIA, PA 19406-1415
Buchanan, NY 10511-0249  
                                          May 14, 2009
SUBJECT: INDIAN POINT NUCLEAR GENERATING UNIT 2 - NRC INTEGRATED INSPECTION REPORT 05000247/2009002  
Mr. Joseph E. Pollock
Site Vice President
Dear Mr. Pollock:  
Entergy Nuclear Operations, Inc.
On March 31, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Indian Point Nuclear Generating Unit 2. The enclosed integrated inspection report documents the inspection results, which were discussed on April 15, 2009, with you and other  
Indian Point Energy Center
members of your staff.  
450 Broadway, GSB
The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations, and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.  
Buchanan, NY 10511-0249
SUBJECT:         INDIAN POINT NUCLEAR GENERATING UNIT 2 - NRC INTEGRATED
This report documents seven findings of very low safety significance (Green). Six of these findings were also determined to be violations of NRC requirements. However, because of their very low safety significance, and because the findings were entered into your corrective action program, the NRC is treating these findings as non-cited violations (NCVs) consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest any NCV in this report, you  
                INSPECTION REPORT 05000247/2009002
should provide a written response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington D.C. 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Senior Resident Inspector at Indian Point Nuclear Generating Unit 2. In addition, if you disagree with the characterization of any finding, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator and the NRC Resident Inspectors at Indian Point Nuclear Generating Unit 2. The information you provide will be considered in accordance with Inspection Manual Chapter 0305. 
Dear Mr. Pollock:
J. Pollock 2
On March 31, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection
In accordance with Title 10 of the Code of Federal Regulations Part 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room of from the Publicly Available Records (PARS) component of the NRC's document system (ADAMS). ADAMS is accessible from the NRC Web Site at http://www.nrc.gov/reading-rm/adams.html
at Indian Point Nuclear Generating Unit 2. The enclosed integrated inspection report
  (the Public Electronic Reading Room).  
documents the inspection results, which were discussed on April 15, 2009, with you and other
members of your staff.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations, and with the conditions of your
license. The inspectors reviewed selected procedures and records, observed activities, and
interviewed personnel.
This report documents seven findings of very low safety significance (Green). Six of these
findings were also determined to be violations of NRC requirements. However, because of their
very low safety significance, and because the findings were entered into your corrective action
program, the NRC is treating these findings as non-cited violations (NCVs) consistent with
Section VI.A.1 of the NRC Enforcement Policy. If you contest any NCV in this report, you
should provide a written response within 30 days of the date of this inspection report, with the
basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk,
Washington D.C. 20555-0001; with copies to the Regional Administrator, Region I; the Director,
Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC
20555-0001; and the NRC Senior Resident Inspector at Indian Point Nuclear Generating Unit 2.
In addition, if you disagree with the characterization of any finding, you should provide a
response within 30 days of the date of this inspection report, with the basis for your
disagreement, to the Regional Administrator and the NRC Resident Inspectors at Indian Point
Nuclear Generating Unit 2. The information you provide will be considered in accordance with
Inspection Manual Chapter 0305.


        Sincerely,
J. Pollock                                      2
      /RA/         Mel Gray, Chief       Projects Branch 2       Division of Reactor Projects  
In accordance with Title 10 of the Code of Federal Regulations Part 2.390 of the NRCs Rules
Docket No. 50-247 License No. DPR-26  
of Practice, a copy of this letter, its enclosure, and your response (if any) will be available
Enclosure: Inspection Report No. 05000247/2009002   w/ Attachment: Supplemental Information  
electronically for public inspection in the NRC Public Document Room of from the Publicly
cc w/encl:
Available Records (PARS) component of the NRCs document system (ADAMS).
Senior Vice President, Entergy Nuclear Operations Vice President, Operations, Entergy Nuclear Operations Vice President, Oversight, Entergy Nuclear Operations Senior Manager, Nuclear Safety and Licensing, Entergy Nuclear Operations Senior Vice President and COO, Entergy Nuclear Operations Assistant General Counsel, Entergy Nuclear Operations Manager, Licensing, Entergy Nuclear Operations  
ADAMS is accessible from the NRC Web Site at http://www.nrc.gov/reading-rm/adams.html
C. Donaldson, Esquire, Assistant Attorney General, New York Department of Law A. Donahue, Mayor, Village of Buchanan J. G. Testa, Mayor, City of Peekskill R. Albanese, Four County Coordinator S. Lousteau, Treasury Department, Entergy Services, Inc.  
(the Public Electronic Reading Room).
Chairman, Standing Committee on Energy, NYS Assembly Chairman, Standing Committee on Environmental Conservation, NYS Assembly Chairman, Committee on Corporations, Authorities, and Commissions M. Slobodien, Director, Emergency Planning P. Eddy, NYS Department of Public Service  
                                                        Sincerely,
Assemblywoman Sandra Galef, NYS Assembly T. Seckerson, County Clerk, Westchester County Board of Legislators A. Spano, Westchester County Executive R. Bondi, Putnam County Executive C. Vanderhoef, Rockland County Executive  
                                                        /RA/
E. A. Diana, Orange County Executive T. Judson, Central NY Citizens Awareness Network M. Elie, Citizens Awareness Network Public Citizen's Critical Mass Energy Project M. Mariotte, Nuclear Information & Resources Service  
                                                        Mel Gray, Chief
F. Zalcman, Pace Law School, Energy Project L. Puglisi, Supervisor, Town of Cortlandt
                                                        Projects Branch 2
J. Pollock 3
                                                        Division of Reactor Projects
Congressman John Hall Congresswoman Nita Lowey Senator Kirsten E. Gillibrand Senator Charles Schumer G. Shapiro, Senator Gillibrand 's Staff J. Riccio, Greenpeace
Docket No. 50-247
P.  Musegaas, Riverkeeper, Inc. M. Kaplowitz, Chairman of County Environment & Health Committee A. Reynolds, Environmental Advocates D. Katz, Executive Director, Citizens Awareness Network K. Coplan, Pace Environmental Litigation Clinic
License No. DPR-26
M. Jacobs, IPSEC W. Little, Associate Attorney, NYSDEC M. J. Greene, Clearwater, Inc. R. Christman, Manager Training and Development  J. Spath, New York State Energy Research, SLO Designee F. Murray, President & CEO, New York State Energy Research A. J. Kremer, New York Affordable Reliable Electricity Alliance (NY AREA)
Enclosure:     Inspection Report No. 05000247/2009002
                w/ Attachment: Supplemental Information
cc w/encl:
Senior Vice President, Entergy Nuclear Operations
Vice President, Operations, Entergy Nuclear Operations
Vice President, Oversight, Entergy Nuclear Operations
Senior Manager, Nuclear Safety and Licensing, Entergy Nuclear Operations
Senior Vice President and COO, Entergy Nuclear Operations
Assistant General Counsel, Entergy Nuclear Operations
Manager, Licensing, Entergy Nuclear Operations
C. Donaldson, Esquire, Assistant Attorney General, New York Department of Law
A. Donahue, Mayor, Village of Buchanan
J. G. Testa, Mayor, City of Peekskill
R. Albanese, Four County Coordinator
S. Lousteau, Treasury Department, Entergy Services, Inc.
Chairman, Standing Committee on Energy, NYS Assembly
Chairman, Standing Committee on Environmental Conservation, NYS Assembly
Chairman, Committee on Corporations, Authorities, and Commissions
M. Slobodien, Director, Emergency Planning
P. Eddy, NYS Department of Public Service
Assemblywoman Sandra Galef, NYS Assembly
T. Seckerson, County Clerk, Westchester County Board of Legislators
A. Spano, Westchester County Executive
R. Bondi, Putnam County Executive
C. Vanderhoef, Rockland County Executive
E. A. Diana, Orange County Executive
T. Judson, Central NY Citizens Awareness Network
M. Elie, Citizens Awareness Network
Public Citizen's Critical Mass Energy Project
M. Mariotte, Nuclear Information & Resources Service
F. Zalcman, Pace Law School, Energy Project
L. Puglisi, Supervisor, Town of Cortlandt


 
J. Pollock                                 3
J. Pollock 4
Congressman John Hall
In accordance with Title 10 of the Code of Federal Regulations Part 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room of from the Publicly Available Records (PARS) component of the NRC's document system (ADAMS). ADAMS is accessible from the NRC Web Site at http://www.nrc.gov/reading-rm/adams.html
Congresswoman Nita Lowey
  (the Public Electronic Reading Room).  
Senator Kirsten E. Gillibrand
        Sincerely,  
Senator Charles Schumer
              /RA/        Mel Gray, Chief        Projects Branch 2        Division of Reactor Projects
G. Shapiro, Senator Gillibrand 's Staff
J. Riccio, Greenpeace
P. Musegaas, Riverkeeper, Inc.
M. Kaplowitz, Chairman of County Environment & Health Committee
A. Reynolds, Environmental Advocates
D. Katz, Executive Director, Citizens Awareness Network
K. Coplan, Pace Environmental Litigation Clinic
M. Jacobs, IPSEC
W. Little, Associate Attorney, NYSDEC
M. J. Greene, Clearwater, Inc.
R. Christman, Manager Training and Development
J. Spath, New York State Energy Research, SLO Designee
F. Murray, President & CEO, New York State Energy Research
A. J. Kremer, New York Affordable Reliable Electricity Alliance (NY AREA)


Distribution w/encl: (via E-mail)  S. Collins, RA M. Dapas, DRA D. Lew, DRP
J. Pollock                                                          4
J. Clifford, DRP 
In accordance with Title 10 of the Code of Federal Regulations Part 2.390 of the NRCs Rules
M. Gray, DRP
of Practice, a copy of this letter, its enclosure, and your response (if any) will be available
B. Bickett, DRP
electronically for public inspection in the NRC Public Document Room of from the Publicly
A. Rosebrook, DRP S. McCarver, DRP J. Heinly, DRP 
Available Records (PARS) component of the NRCs document system (ADAMS).
G. Malone, DRP, SRI, IP2  C. Hott, DRP, RI, IP2 D. Hochmuth, DRP, OA S. Campbell, RI OEDO R. Nelson, NRR  
ADAMS is accessible from the NRC Web Site at http://www.nrc.gov/reading-rm/adams.html
M. Kowal, NRR  
(the Public Electronic Reading Room).
J. Boska, PM, NRR  
                                                                                  Sincerely,
J. Hughey, NRR  
                                                                                  /RA/
D. Bearde, DRP ROPreports@nrc.gov
                                                                                  Mel Gray, Chief
Region I Docket Room (w/concurrences)  
                                                                                  Projects Branch 2
  SUNSI Review Complete: ____BSB____ (Reviewer's Initial)
                                                                                  Division of Reactor Projects
  DOCUMENT NAME: G:\DRP\BRANCH2\A - INDIAN POINT 2\INSPECTION REPORTS\IP2 IR2009-002\IP2 2009002 REVFINAL.DOC  
Distribution w/encl: (via E-mail)                                                C. Hott, DRP, RI, IP2
S. Collins, RA                                                                    D. Hochmuth, DRP, OA
M. Dapas, DRA                                                                    S. Campbell, RI OEDO
D. Lew, DRP                                                                      R. Nelson, NRR
J. Clifford, DRP                                                                  M. Kowal, NRR
M. Gray, DRP                                                                      J. Boska, PM, NRR
B. Bickett, DRP                                                                  J. Hughey, NRR
A. Rosebrook, DRP                                                                D. Bearde, DRP
S. McCarver, DRP                                                                  ROPreports@nrc.gov
J. Heinly, DRP                                                                    Region I Docket Room (w/concurrences)
G. Malone, DRP, SRI, IP2
SUNSI Review Complete: ____BSB____ (Reviewers Initial)
DOCUMENT NAME: G:\DRP\BRANCH2\A - INDIAN POINT 2\INSPECTION REPORTS\IP2 IR2009-002\IP2
2009002 REVFINAL.DOC
After declaring this document An Official Agency Record it will be released to the Public
To Receive a copy of this document, indicate in the box: C = Copy without attachment/enclosure E = Copy with attachment/enclosure N = No copy
  Office                RI/DRP                              RI/DRP                                  RI/DRP
  Name                  GMalone/BSB for                      BBickett/                              MGray/
  Date                  05/14/09                            05/14/09                                05/14/09
                                                    OFFICAL AGENCY RECORD


After declaring this document "An Official Agency Record" it will be released to the Public
                                      1
To Receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure  "E" = Copy with attachment/enclosure  "N" = No copy Office RI/DRP  RI/DRP  RI/DRP  Name GMalone/BSB for BBickett/ MGray/ Date 05/14/09 05/14/09 05/14/09 OFFICAL AGENCY RECORD
                U.S. NUCLEAR REGULATORY COMMISSION
 
                                  REGION I
1 Enclosure 
Docket No.:  50-247
U.S. NUCLEAR REGULATORY COMMISSION  
License No.: DPR-26
REGION I   Docket No.:  50-247  
Report No.:  05000247/2009002
 
Licensee:   Entergy Nuclear Northeast (Entergy)
License No.: DPR-26  
Facility:   Indian Point Nuclear Generating Unit 2
  Report No.:  05000247/2009002  
Location:    450 Broadway, GSB
            Buchanan, NY 10511-0249
Licensee: Entergy Nuclear Northeast (Entergy)  
Dates:      January 1, 2009 through March 31, 2009
  Facility: Indian Point Nuclear Generating Unit 2  
Inspectors:  G. Malone, Senior Resident Inspector, Indian Point 2
            C. Hott, Resident Inspector, Indian Point 2
            J. Commisky, Health Physics Inspector, Region I
Approved By: Mel Gray, Chief
            Projects Branch 2
            Division of Reactor Projects
                                                                  Enclosure


  Location:  450 Broadway, GSB    Buchanan, NY 10511-0249
                                                            2
                                          TABLE OF CONTENTS
Dates:  January 1, 2009 through March 31, 2009
SUMMARY OF FINDINGS ............................................................................................................... 3
  Inspectors:  G. Malone, Senior Resident Inspector, Indian Point 2  
REPORT DETAILS........................................................................................................................... 8
  C. Hott, Resident Inspector, Indian Point 2    J. Commisky, Health Physics Inspector, Region I
1. REACTOR SAFETY .................................................................................................................... 8
  Approved By:  Mel Gray, Chief    Projects Branch 2    Division of Reactor Projects 
1R01   Adverse Weather Protection ............................................................................................... 8
2 Enclosure  TABLE OF CONTENTS  
1R04   Equipment Alignment ....................................................................................................... 10
  SUMMARY OF FINDINGS ...........................................................................................................
1R05   Fire Protection .................................................................................................................. 10
.... 3 REPORT DETAILS  
1R07   Heat Sink Performance .................................................................................................... 14
........................................................................................................................... 8
1R11   Licensed Operator Requalification Program ..................................................................... 15
  1. REACTOR SAFETY ................................................................................................................
1R12   Maintenance Effectiveness ............................................................................................... 15
.... 8 1R01 Adverse Weather Protection ............................................................................................... 8
1R13   Maintenance Risk Assessments and Emergent Work Control .......................................... 18
1R04 Equipment Alignment ....................................................................................................... 10
1R15   Operability Evaluations ..................................................................................................... 19
1R05 Fire Protection .................................................................................................................. 10 1R07 Heat Sink Performance .................................................................................................... 14
1R18   Plant Modifications ........................................................................................................... 20
1R11 Licensed Operator Requalification Program ..................................................................... 15
1R19   Post-Maintenance Testing ................................................................................................ 21
1R12 Maintenance Effectivene ss ...............................................................................................  
1R22   Surveillance Testing ......................................................................................................... 21
15 1R13 Maintenance Risk Assessments and Emergent Work Control .......................................... 18
1EP6   Drill Evaluation ................................................................................................................ 24
1R15 Operability Eval
2. RADIATION SAFETY ................................................................................................................ 24
uations ..................................................................................................... 1
2OS1 Access Control to Radiologically Significant Areas ........................................................... 24
9 1R18 Plant Modifications ........................................................................................................... 20 1R19 Post-Maintenance Testin g ................................................................................................  
2OS2 ALARA Planning and Controls.......................................................................................... 28
21 1R22 Surveillance Testing .........................................................................................................  
4. OTHER ACTIVITIES.................................................................................................................. 30
21 1EP6 Drill Evaluat
4OA1 Performance Indicator Verification ................................................................................... 30
ion  ................................................................................................................ 24 2. RADIATION SAFETY ................................................................................................................ 24 2OS1 Access Control to Radiologically Significant Areas ........................................................... 24
4OA2 Identification and Resolution of Problems ......................................................................... 31
2OS2 ALARA Planning and Controls .......................................................................................... 28
4OA3 Event Followup ................................................................................................................. 31
  4. OTHER ACTIVITIES .................................................................................................................. 30
4OA5 Other Activities ................................................................................................................. 32
4OA1 Performance Indicator Verification ................................................................................... 30
4OA6 Meetings........................................................................................................................... 33
4OA2 Identification and Resolution of Problems ......................................................................... 31
ATTACHMENT: SUPPLEMENTAL INFORMATION .................................................................... A-1
4OA3 Event Followup ................................................................................................................. 31 4OA5 Other Activities ................................................................................................................. 32 4OA6 Meetings ........................................................................................................................... 33
 
ATTACHMENT: SUPPLEMENTAL INFORMATION .................................................................... A-1
 
SUPPLEMENTAL INFORMATION ............................................................................................... A-1
SUPPLEMENTAL INFORMATION ............................................................................................... A-1
 
KEY POINTS OF CONTACT ........................................................................................................ A-1
KEY POINTS OF CONTACT ........................................................................................................ A-1 LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED ............................................................. A-1
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED ............................................................. A-1
  LIST OF DOCUMENTS REVIEWED ............................................................................................ A-2
LIST OF DOCUMENTS REVIEWED ............................................................................................ A-2
  LIST OF ACRONYMS .................................................................................................................. A-8
LIST OF ACRONYMS .................................................................................................................. A-8
3 Enclosure SUMMARY OF FINDINGS
                                                                                                                        Enclosure
  IR 05000247/2009-002; 01/01/2009 - 03/31/2009; Indian Point Nuclear Generating (Indian Point) Unit 2; Adverse Weather Protection; Fire Protection;  Maintenance Effectiveness; Maintenance Risk Assessments; Surveillance Testing; and Radiological Access Control.
This report covered a three-month period of inspection by resident and region based inspectors.  Seven findings of very low significance (Green) were identified, six of which were also determined to be non-cited violations (NCV).  The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609,
"Significance Determination Process."  The cross-cutting aspect for each finding was determined using IMC 0305, "Operating Reactor Assessment Program."  Findings for which the significance determination process (SDP) does not apply may be Green, or be assigned a severity level after NRC management review.  The NRC's program for overseeing safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006.
A. NRC-Identified and Self-Revealing Findings
  Cornerstone:  Initiating Events
  * Green. The inspectors identified a NCV of very low safety significance related to 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions," because Entergy did not promptly identify and correct an adverse condition related to an electrical fault.  Specifically,
personnel did not identify a safety-related cubicle had experienced an electrical fault prior to replacement of upstream fuses and restoration of power to the damaged cubicle.  Entergy entered the issue into the corrective action program as IP2-2009-00342 and IP2-2009-00483, trained all operations personnel on the requirements to replace fuses and re-energize electrical equipment, and plans to revise the operations procedure for
operating electrical equipment.
This issue was more than minor because the finding was associated with the external factors attribute of the Initiating Events cornerstone and impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge
critical safety systems during shutdown as well as power operations.  The inspectors determined that the issue increased the likelihood of a fire in the emergency diesel generator (EDG) building.  The condition was evaluated by a Senior Reactor Analyst utilizing Phase 2 of IMC 0609 Appendix F, "Fire Protection Significance Determination Process."  It was determined that in the event of a fire consuming the MCC, no transient
would be placed on the plant and no components required to safely shutdown the plant would be impacted.  As a result, in accordance with task 2.3.5 of Appendix F, the issue was screened to Green.
The inspectors determined that a cross-cutting aspect was associated with this finding in the area of human performance related to conservative decision making.  Specifically, Entergy's decision-making was non-conservative related to its decisions on the process
used to identify the source of the acrid odor; re-energize the damaged electrical equipment; and keep a damaged electrical component energized for 14 days prior to its removal from the MCC.  [H.1(b) per IMC 0305] (Section 1R05)
 
4 Enclosure 
* Green. The inspectors identified a NCV of very low safety significance related to TS 5.4.1, "Administrative Controls: Procedures," because Entergy did not maintain an adequate maintenance procedure for a safety-related electrical motor control center
(MCC).  Specifically, the eight-year maintenance procedure for the affected EDG ventilation MCC did not contain an adequate method to identify high resistance connections within the cubicle as was expected in the applicable preventative maintenance industry template.  Subsequently, a high resistance connection within the MCC developed into a phase-to-phase electrical fault on January 28, 2009.  Entergy
entered the issue into the corrective action program, scoped the affected MCC and 21 additional MCCs into the site's thermography program, and planned to revise the maintenance procedure.
This issue was more than minor because the finding was associated with the external
factors attribute of the Initiating Events cornerstone and impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety systems during shutdown as well as power operations. Specifically, the high resistance connection degraded into a phase-to-phase fault and increased the likelihood of a fire in the EDG building.  The condition was evaluated by a Senior Reactor Analyst utilizing Phase 2 of IMC 0609 Appendix F, "Fire Protection Significance Determination Process."  It was determined that in the event of a fire consuming the
MCC, no transient would be placed on the plant and no components required to safely shutdown the plant would be impacted.  As a result, in accordance with task 2.3.5 of Appendix F, the issue was screened to Green.
The inspectors determined that the finding had a cross-cutting aspect associated with
the area of problem identification and resolution related to the use of operating experience (OE).  Specifically, Entergy personnel did not implement industry recommended practices, or an alternate equivalent method, for identifying high resistance connections in electrical switchgear. [P.2(b) per IMC 0305] (Section 1R12)
Cornerstone:  Mitigating Systems
  * Green. The inspectors identified a finding of very low safety significance because Entergy personnel did not adequately implement procedure EN-LI-102, Corrective Action Process, and promptly identify a condition adverse to quality associated with open louvers in a fire protection pump room following pump testing on January 14, 2009.  The open louvers resulted in freezing conditions in fire protection piping located in the room
and cracked two six-inch header isolation valves on January 17, 2009.  Entergy entered the issue into the corrective action program and performed a site-wide extent-of-condition walkdown of louvers.
The finding was more than minor because it was associated with the protection against
external factors attribute of the Mitigating Systems cornerstone and it affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences.  This finding was evaluated using Phase 1 of IMC 0609 Appendix F, "Fire Protection Significance Determination Process."  The inspectors determined the issue was of very low safety significance (Green) because the cracked valves were easily isolated and did not pass sufficient water to render the fire header non-functional (low degradation rating). 


 
                                                    3
5 Enclosure  The inspectors determined that the finding had a cross-cutting aspect in the area of human performance related to work practices - human error prevention techniques. Specifically, Entergy personnel that routinely tour the 11 fire pump house did not question the abnormally cold room temperatures. [H.4(a) per IMC 0305] (Section 1R01)  
                                        SUMMARY OF FINDINGS
* Green. The inspectors identified a NCV of very low safety significance related to License Condition 2.K., fire protection program, because personnel did not promptly identify and correct a degraded three-hour rated fire door latch mechanism on the west entrance of  
IR 05000247/2009-002; 01/01/2009 - 03/31/2009; Indian Point Nuclear Generating (Indian
the 480-Volt switchgear room. Specifically, inspectors identified the fire door in a non-functional state on several instances over the course of a month.  Entergy personnel replaced the fire door latch mechanism on March 3, 2009. This issue was entered into the corrective action program as six condition reports spanning several weeks and included an extent of condition walkdown of site fire doors.
Point) Unit 2; Adverse Weather Protection; Fire Protection; Maintenance Effectiveness;
Maintenance Risk Assessments; Surveillance Testing; and Radiological Access Control.
This report covered a three-month period of inspection by resident and region based inspectors.
Seven findings of very low significance (Green) were identified, six of which were also
determined to be non-cited violations (NCV). The significance of most findings is indicated by
their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609,
Significance Determination Process. The cross-cutting aspect for each finding was
determined using IMC 0305, Operating Reactor Assessment Program. Findings for which the
significance determination process (SDP) does not apply may be Green, or be assigned a
severity level after NRC management review. The NRCs program for overseeing safe
operation of commercial nuclear power reactors is described in NUREG-1649, Reactor
Oversight Process, Revision 4, dated December 2006.
A.      NRC-Identified and Self-Revealing Findings
        Cornerstone: Initiating Events
    *   Green. The inspectors identified a NCV of very low safety significance related to 10 CFR
        50, Appendix B, Criterion XVI, Corrective Actions, because Entergy did not promptly
        identify and correct an adverse condition related to an electrical fault. Specifically,
        personnel did not identify a safety-related cubicle had experienced an electrical fault
        prior to replacement of upstream fuses and restoration of power to the damaged cubicle.
        Entergy entered the issue into the corrective action program as IP2-2009-00342 and
        IP2-2009-00483, trained all operations personnel on the requirements to replace fuses
        and re-energize electrical equipment, and plans to revise the operations procedure for
        operating electrical equipment.
        This issue was more than minor because the finding was associated with the external
        factors attribute of the Initiating Events cornerstone and impacted the cornerstone
        objective of limiting the likelihood of those events that upset plant stability and challenge
        critical safety systems during shutdown as well as power operations. The inspectors
        determined that the issue increased the likelihood of a fire in the emergency diesel
        generator (EDG) building. The condition was evaluated by a Senior Reactor Analyst
        utilizing Phase 2 of IMC 0609 Appendix F, Fire Protection Significance Determination
        Process. It was determined that in the event of a fire consuming the MCC, no transient
        would be placed on the plant and no components required to safely shutdown the plant
        would be impacted. As a result, in accordance with task 2.3.5 of Appendix F, the issue
        was screened to Green.
        The inspectors determined that a cross-cutting aspect was associated with this finding
        in the area of human performance related to conservative decision making. Specifically,
        Entergys decision-making was non-conservative related to its decisions on the process
        used to identify the source of the acrid odor; re-energize the damaged electrical
        equipment; and keep a damaged electrical component energized for 14 days prior to its
        removal from the MCC. [H.1(b) per IMC 0305] (Section 1R05)
                                                                                            Enclosure


The finding was more than minor because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. This fire door, when degraded, impacts the reliability of mitigating systems in the 480-Volt switchgear room that are relied upon during a postulated large fire in the turbine building, and vice versa. This finding was  
                                              4
evaluated using Phase 1 of IMC 0609 Appendix F, "Fire Protection Significance Determination Process."  Since the area in question had a fire watch posted during the time the door was degraded for an unrelated issue, an adequate level of protection was maintained to compensate for the degraded door. As such, according to task 1.3.1, the inspectors determined the finding was Green.
* Green. The inspectors identified a NCV of very low safety significance related to TS
  5.4.1, Administrative Controls: Procedures, because Entergy did not maintain an
  adequate maintenance procedure for a safety-related electrical motor control center
  (MCC). Specifically, the eight-year maintenance procedure for the affected EDG
  ventilation MCC did not contain an adequate method to identify high resistance
  connections within the cubicle as was expected in the applicable preventative
  maintenance industry template. Subsequently, a high resistance connection within the
  MCC developed into a phase-to-phase electrical fault on January 28, 2009. Entergy
  entered the issue into the corrective action program, scoped the affected MCC and 21
  additional MCCs into the sites thermography program, and planned to revise the
  maintenance procedure.
  This issue was more than minor because the finding was associated with the external
  factors attribute of the Initiating Events cornerstone and impacted the cornerstone
  objective of limiting the likelihood of those events that upset plant stability and challenge
  critical safety systems during shutdown as well as power operations. Specifically, the
  high resistance connection degraded into a phase-to-phase fault and increased the
  likelihood of a fire in the EDG building. The condition was evaluated by a Senior
  Reactor Analyst utilizing Phase 2 of IMC 0609 Appendix F, Fire Protection Significance
  Determination Process. It was determined that in the event of a fire consuming the
  MCC, no transient would be placed on the plant and no components required to safely
  shutdown the plant would be impacted. As a result, in accordance with task 2.3.5 of
  Appendix F, the issue was screened to Green.
  The inspectors determined that the finding had a cross-cutting aspect associated with
  the area of problem identification and resolution related to the use of operating
  experience (OE). Specifically, Entergy personnel did not implement industry
  recommended practices, or an alternate equivalent method, for identifying high
  resistance connections in electrical switchgear. [P.2(b) per IMC 0305] (Section 1R12)
  Cornerstone: Mitigating Systems
* Green. The inspectors identified a finding of very low safety significance because
  Entergy personnel did not adequately implement procedure EN-LI-102, Corrective Action
  Process, and promptly identify a condition adverse to quality associated with open
  louvers in a fire protection pump room following pump testing on January 14, 2009. The
  open louvers resulted in freezing conditions in fire protection piping located in the room
  and cracked two six-inch header isolation valves on January 17, 2009. Entergy entered
  the issue into the corrective action program and performed a site-wide extent-of-
  condition walkdown of louvers.
  The finding was more than minor because it was associated with the protection against
  external factors attribute of the Mitigating Systems cornerstone and it affected the
  cornerstone objective of ensuring the reliability of systems that respond to initiating
  events to prevent undesirable consequences. This finding was evaluated using Phase
  1 of IMC 0609 Appendix F, Fire Protection Significance Determination Process. The
  inspectors determined the issue was of very low safety significance (Green) because
  the cracked valves were easily isolated and did not pass sufficient water to render the
  fire header non-functional (low degradation rating).
                                                                                      Enclosure


The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy personnel did not thoroughly evaluate a degraded fire door latch on several occasions, such that the resolution of the problems addressed the causes. [P.1(c) per IMC 0305] (Section 1R05)  
                                              5
* Green. The inspectors identified a NCV of very low safety significance related to 10 CFR 50.65(a)(4), because Entergy personnel did not adequately assess the risk associated with the unavailability of the Refueling Water Storage Tank (RWST) level indication during planned maintenance on the level transmitters and instrumentation. Entergy entered the issue into the corrective action program (CR-IP2-2009-00342), updated the risk model to include the maintenance activity, assessed the risk, and appropriately  
  The inspectors determined that the finding had a cross-cutting aspect in the area of
coded the maintenance activity to ensure it would be risk assessed in the future.   
  human performance related to work practices - human error prevention techniques.
The inspectors determined that this finding was more than minor because it was a maintenance risk assessment issue in which personnel did not consider risk significant SSCs that were unavailable during maintenance. The RWST level indication is  
  Specifically, Entergy personnel that routinely tour the 11 fire pump house did not
specifically listed in Table 2 of the plant specific Phase 2 SDP risk-informed inspection notebook. The inspectors determined the significance of this issue in accordance with IMC 0609, Appendix K, "Maintenance Risk Assessment and Risk Management Significance Determination Process."  The inspectors determined that this finding was of very low safety significance because the Incremental Core Damage Probability Deficit was less than 1E-6.   
  question the abnormally cold room temperatures. [H.4(a) per IMC 0305] (Section 1R01)
* Green. The inspectors identified a NCV of very low safety significance related to License
The inspectors determined that the finding had a cross-cutting aspect in the area of human performance related to work control. Specifically, Entergy personnel did not
  Condition 2.K., fire protection program, because personnel did not promptly identify and
6 Enclosure appropriately plan work activities by incorporating risk insights for affected plant equipment.  [H.3(a) per IMC 0305] (Section 1R13)
  correct a degraded three-hour rated fire door latch mechanism on the west entrance of
* Green.  The inspectors identified a NCV of very low safety significance related to 10 CFR 50.55a, "Codes and standards," because Entergy's procedure, 2-PT-Q031A for an auxiliary component cooling water pump, did not contain appropriate acceptance criteria for positively determining that safety-related check valves performed their safety function when required in accordance with the American Society of Mechanical Engineers
  the 480-Volt switchgear room. Specifically, inspectors identified the fire door in a non-
(ASME) OM Code.  Specifically, the test used reverse rotation of a parallel pump to verify that the pump's discharge check valve was closed although previous site-specific experience demonstrated that the pump impeller would not rotate backwards when the check valve was stuck open.  Entergy entered this issue into their corrective action program as CR-2009-1312.
  functional state on several instances over the course of a month. Entergy personnel
The inspectors determined that the performance deficiency was greater than minor because it was associated with the procedure quality attribute of the Mitigating System cornerstone and it adversely affected the cornerstone's objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences.  Specifically, the test criterion used in procedure 2-PT-Q013A did not ensure that valve 755A reliably performed its safety function when tested as demonstrated by testing
  replaced the fire door latch mechanism on March 3, 2009. This issue was entered into
performed in January 2005.  The inspectors determined that the performance deficiency was of very low safety significance (Green) IMC 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings."  Specifically, the inspectors determined that this finding was of very low safety significance because the finding did not result in a loss of safety function and did not screen as potentially risk-significant due to external
  the corrective action program as six condition reports spanning several weeks and
events initiating events. 
  included an extent of condition walkdown of site fire doors.
The inspectors determined the finding had a cross-cutting aspect related to effective corrective actions in the corrective action program component of the problem identification and resolution area.  Specifically, Entergy personnel did not implement
  The finding was more than minor because it is associated with the protection against
effective corrective actions to resolve the testing inadequacy since 2005 and during subsequent quarterly testing. [P.1(d) per IMC 0305]  (Section 1R22)
  external factors attribute of the Mitigating Systems cornerstone and affected the
  cornerstone objective of ensuring the reliability of systems that respond to initiating
Cornerstone: Occupational Radiation Safety
  events to prevent undesirable consequences. This fire door, when degraded, impacts
* Green. The inspectors identified a NCV of very low safety significance related to Technical Specification 5.4.1.a, "Procedures," because Entergy personnel did not generate condition reports or investigation paperwork for multiple high dose-rate alarms as required by station procedures.  Specifically, personnel did not generate the required condition reports and adequately document the investigations for six instances of unplanned or un-briefed electronic dosimeter alarms that occurred between January 2009 and March 2009.  The performance deficiency resulted in workers receiving
  the reliability of mitigating systems in the 480-Volt switchgear room that are relied upon
unanticipated dose rate alarms with no formally-documented investigation prior to returning to work in a Radiologically Controlled Area.  Entergy entered the finding into the corrective action program as condition report CR-IP3-2009-01253 and 01318.
  during a postulated large fire in the turbine building, and vice versa. This finding was
The finding is more than minor because it is associated with the Occupational Radiation Safety cornerstone attribute of programs and process, and adversely affected the objective to ensure adequate protection of worker health and safety from exposure to
  evaluated using Phase 1 of IMC 0609 Appendix F, Fire Protection Significance
radiation.  Moreover, the inspectors identified a programmatic deficiency to maintain and implement programs to keep exposures as low as reasonably achievable, because 
  Determination Process. Since the area in question had a fire watch posted during the
7 Enclosure  multiple examples were identified regarding the failure to satisfy station radiation protection procedures.  Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance (Green) because it did not involve: (1) as low as is reasonably achievable planning and controls, (2) an overexposure of an individual, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. 
  time the door was degraded for an unrelated issue, an adequate level of protection was
  maintained to compensate for the degraded door. As such, according to task 1.3.1, the
  inspectors determined the finding was Green.
  The inspectors determined that the finding had a cross-cutting aspect in the area of
  problem identification and resolution because Entergy personnel did not thoroughly
  evaluate a degraded fire door latch on several occasions, such that the resolution of the
  problems addressed the causes. [P.1(c) per IMC 0305] (Section 1R05)
* Green. The inspectors identified a NCV of very low safety significance related to 10 CFR
  50.65(a)(4), because Entergy personnel did not adequately assess the risk associated
  with the unavailability of the Refueling Water Storage Tank (RWST) level indication
  during planned maintenance on the level transmitters and instrumentation. Entergy
  entered the issue into the corrective action program (CR-IP2-2009-00342), updated the
  risk model to include the maintenance activity, assessed the risk, and appropriately
  coded the maintenance activity to ensure it would be risk assessed in the future.
   The inspectors determined that this finding was more than minor because it was a
  maintenance risk assessment issue in which personnel did not consider risk significant
  SSCs that were unavailable during maintenance. The RWST level indication is
  specifically listed in Table 2 of the plant specific Phase 2 SDP risk-informed inspection
  notebook. The inspectors determined the significance of this issue in accordance with
  IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management
  Significance Determination Process. The inspectors determined that this finding was of
  very low safety significance because the Incremental Core Damage Probability Deficit
  was less than 1E-6.
   The inspectors determined that the finding had a cross-cutting aspect in the area of
  human performance related to work control. Specifically, Entergy personnel did not
                                                                                      Enclosure


The inspectors determined that the finding had a cross-cutting aspect related to procedural adherence in the work practices component of the human performance area. Specifically, Entergy personnel did not follow procedures to generate condition reports and document investigations when high dose-rate alarms were received by workers.
                                              6
[H.4(b) per IMC 0305] (Section 2OS1)  
  appropriately plan work activities by incorporating risk insights for affected plant
B. Licensee-Identified Violations
  equipment. [H.3(a) per IMC 0305] (Section 1R13)
   None.  
* Green. The inspectors identified a NCV of very low safety significance related to 10
8 Enclosure  REPORT DETAILS  
  CFR 50.55a, Codes and standards, because Entergys procedure, 2-PT-Q031A for an
  Summary of Plant Status
  auxiliary component cooling water pump, did not contain appropriate acceptance criteria
  Indian Point Nuclear Generating (Indian Point) Unit 2 began the inspection period at full reactor  
  for positively determining that safety-related check valves performed their safety function
power and remained at or near full power during the quarter.  
  when required in accordance with the American Society of Mechanical Engineers
1. REACTOR SAFETY  
  (ASME) OM Code. Specifically, the test used reverse rotation of a parallel pump to
  Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
  verify that the pumps discharge check valve was closed although previous site-specific
 
  experience demonstrated that the pump impeller would not rotate backwards when the
1R01 Adverse Weather Protection (71111.01 - 1 sample)  
  check valve was stuck open. Entergy entered this issue into their corrective action
  Impending Adverse Weather
  program as CR-2009-1312.
    a. Inspection Scope
  The inspectors determined that the performance deficiency was greater than minor
  The inspectors reviewed the overall preparations and protection of risk-significant systems for extremely cold weather conditions from January 14 - 19, 2009. The inspectors reviewed and assessed implementation of the site's adverse weather preparation procedures and compensatory measures for the affected conditions before the onset of and during the cold weather conditions. This included verification that  
  because it was associated with the procedure quality attribute of the Mitigating System
operator actions defined in their adverse weather procedure maintain readiness of essential systems that are vulnerable to freezing temperatures. The inspectors verified Entergy personnel implemented periodic equipment walkdowns or other measures to ensure the condition of plant equipment was operable.  
  cornerstone and it adversely affected the cornerstones objective to ensure the reliability
  of systems that respond to initiating events to prevent undesirable consequences.
The inspectors also reviewed Entergy's corrective action program to review previous issues associated with cold weather preparations and freezing conditions. Documents reviewed are listed in the attachment.  
  Specifically, the test criterion used in procedure 2-PT-Q013A did not ensure that valve
  b. Findings
  755A reliably performed its safety function when tested as demonstrated by testing
  Introduction. The inspectors identified a Green finding because Entergy personnel did not adequately implement procedure EN-LI-102, Corrective Action Process, and promptly identify a condition adverse to quality associated with stuck-open louvers in a fire protection pump room following pump testing on January 14, 2009.   
  performed in January 2005. The inspectors determined that the performance deficiency
  was of very low safety significance (Green) IMC 0609, Attachment 4, Phase 1 - Initial
  Screening and Characterization of Findings. Specifically, the inspectors determined
  that this finding was of very low safety significance because the finding did not result in
  a loss of safety function and did not screen as potentially risk-significant due to external
  events initiating events.
  The inspectors determined the finding had a cross-cutting aspect related to effective
  corrective actions in the corrective action program component of the problem
  identification and resolution area. Specifically, Entergy personnel did not implement
  effective corrective actions to resolve the testing inadequacy since 2005 and during
  subsequent quarterly testing. [P.1(d) per IMC 0305] (Section 1R22)
  Cornerstone: Occupational Radiation Safety
* Green. The inspectors identified a NCV of very low safety significance related to
  Technical Specification 5.4.1.a, Procedures, because Entergy personnel did not
  generate condition reports or investigation paperwork for multiple high dose-rate alarms
  as required by station procedures. Specifically, personnel did not generate the required
  condition reports and adequately document the investigations for six instances of
  unplanned or un-briefed electronic dosimeter alarms that occurred between January
  2009 and March 2009. The performance deficiency resulted in workers receiving
  unanticipated dose rate alarms with no formally-documented investigation prior to
  returning to work in a Radiologically Controlled Area. Entergy entered the finding into
  the corrective action program as condition report CR-IP3-2009-01253 and 01318.
  The finding is more than minor because it is associated with the Occupational Radiation
  Safety cornerstone attribute of programs and process, and adversely affected the
  objective to ensure adequate protection of worker health and safety from exposure to
  radiation. Moreover, the inspectors identified a programmatic deficiency to maintain and
  implement programs to keep exposures as low as reasonably achievable, because
                                                                                      Enclosure
 
                                            7
  multiple examples were identified regarding the failure to satisfy station radiation
  protection procedures. Using the Occupational Radiation Safety Significance
  Determination Process, the inspectors determined that the finding was of very low safety
  significance (Green) because it did not involve: (1) as low as is reasonably achievable
  planning and controls, (2) an overexposure of an individual, (3) a substantial potential for
  overexposure, or (4) an impaired ability to assess dose.
  The inspectors determined that the finding had a cross-cutting aspect related to
  procedural adherence in the work practices component of the human performance area.
  Specifically, Entergy personnel did not follow procedures to generate condition reports
  and document investigations when high dose-rate alarms were received by workers.
  [H.4(b) per IMC 0305] (Section 2OS1)
B. Licensee-Identified Violations
   None.
                                                                                      Enclosure
 
                                                  8
                                        REPORT DETAILS
Summary of Plant Status
Indian Point Nuclear Generating (Indian Point) Unit 2 began the inspection period at full reactor
power and remained at or near full power during the quarter.
1.     REACTOR SAFETY
      Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection (71111.01 - 1 sample)
      Impending Adverse Weather
  a.   Inspection Scope
        The inspectors reviewed the overall preparations and protection of risk-significant
        systems for extremely cold weather conditions from January 14 - 19, 2009. The
        inspectors reviewed and assessed implementation of the sites adverse weather
        preparation procedures and compensatory measures for the affected conditions before
        the onset of and during the cold weather conditions. This included verification that
        operator actions defined in their adverse weather procedure maintain readiness of
        essential systems that are vulnerable to freezing temperatures. The inspectors verified
        Entergy personnel implemented periodic equipment walkdowns or other measures to
        ensure the condition of plant equipment was operable.
        The inspectors also reviewed Entergys corrective action program to review previous
        issues associated with cold weather preparations and freezing conditions. Documents
        reviewed are listed in the attachment.
  b.     Findings
        Introduction. The inspectors identified a Green finding because Entergy personnel did
        not adequately implement procedure EN-LI-102, Corrective Action Process, and
        promptly identify a condition adverse to quality associated with stuck-open louvers in a
        fire protection pump room following pump testing on January 14, 2009.
        Description. On January 17, 2009, during a period of sustained cold weather which
        included sub-zero temperatures, control room personnel received a fire panel trouble
        alarm indicative of a low-pressure condition in the fire header and dispatched a plant
        operator to investigate. The operator identified spraying water from the body of a
        ruptured six-inch fire protection valve located in the 11 fire pump house. The operator
        isolated the broken valve from the fire header by shutting a manually-operated upstream
        valve which stopped the water spray. In addition, the operator observed that the pump
        house room was significantly colder than expected and subsequently identified the
        rooms ventilation louvers to the outside were mechanically bound in the open position.
        The operator disconnected the louver linkage and manually shut the louvers.
                                                                                        Enclosure
 
                                            9
On January 21, 2009, the inspectors identified a second six inch valve that was cracked
due to the previous cold weather (freezing) conditions in the fire pump house. Entergy
personnel entered this issue into the corrective action program and performed site
walkdowns to identify additional adverse conditions associated with the cold weather.
The inspectors determined that Entergy did not fully implement Entergy procedure EN-
LI-102, Corrective Action Process. Specifically, EN-LI-102 requires plant personnel to
identify adverse conditions, including cold-weather related conditions, and then enter
them into the CAP for resolution. Attachment 9.2 of the procedure provides examples of
adverse conditions expected to be reported; Section 1 of the Attachment contains
examples of operational conditions requiring entry into the CAP including "events or
conditions that could negatively impact reliability or availability." Additionally, plant
operators should have had heightened awareness to cold weather conditions because
Entergy procedure OAP-008, "Severe Weather Preparations," requires in step 4.3.7,
when freezing conditions are expected, that increased monitoring of plant areas to
monitor for adverse effects on plant equipment and verify that adequate protection is
provided. Operations personnel did not identify abnormal conditions in the 11 fire pump
room that led to the freezing and subsequent rupture of fire protection components.
The inspectors determined it was reasonable for Entergy personnel to identify this issue
because operators should have identified that the louvers failed to shut following a
routine operations test of 11 fire pump on January 14, 2009. In addition, operators
perform tours of the pump house every 12 hours and should have identified the room
was much colder than normal.
Analysis. The inspectors identified a performance deficiency because Entergy
personnel did not implement procedure guidance and identify stuck open louvers and a
subsequent second cracked fire header valve in the 11 fire pump house. The finding
was more than minor because it was associated with the protection against external
factors attribute of the Mitigating Systems cornerstone and it affected the cornerstone
objective of ensuring the reliability of systems that respond to initiating events to prevent
undesirable consequences. Specifically, the failure of the six-inch valves impacted the
reliability of the fire header until the ruptured valve was isolated.
This finding was evaluated using Phase 1 of Inspection Manual Chapter (IMC) 0609
Appendix F, Fire Protection Significance Determination Process. The inspectors
determined the issue was of very low safety significance (Green) because the cracked
fire valves were easily isolated and did not pass sufficient water to render the fire
header non-functional. Specifically, the inspectors assigned a low degradation rating to
the fire header because the fire pumps were able to maintain pressure in the fire header
until the ruptured valves were isolated.
The inspectors determined that the finding had a cross-cutting aspect in the area of
human performance related to work practices - human error prevention techniques.
Specifically, Entergy personnel routinely tour the 11 fire pump house did not question
the abnormally cold room temperatures. (H.4(a) per IMC 0305)
Enforcement: Enforcement action does not apply because the performance deficiency
did not involve a violation of a regulatory requirement. Because this finding does not
involve a violation of regulatory requirements and has very low safety significance, it is
identified as FIN 05000247/2009002-01, Failure to Identify Open Louvers in 11 Fire
Pump House.
                                                                                    Enclosure
 
                                                10
1R04 Equipment Alignment (71111.04Q - 3 samples)
      Partial System Walkdowns
a.  Inspection Scope
      The inspectors performed partial system walkdowns to verify the operability of redundant
      or diverse trains and components during periods of system train unavailability, or
      following periods of maintenance. The inspectors referenced the system procedures,
      the UFSAR, and system drawings to verify the alignment of the available train supported
      its required safety functions. The inspectors also reviewed applicable condition reports
      (CR) and work orders to ensure Entergy personnel identified and properly addressed
      equipment discrepancies that could potentially impair the capability of the available train,
      as required by Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix
      B, Criterion XVI, Corrective Action. The documents reviewed during these inspections
      are listed in the Attachment.
      The inspectors performed a partial walkdown on the following systems, which
      represented three inspection samples:
    *        21 and 22 component cooling water (CCW) system train when 23 CCW pump
              was tagged out for maintenance;
    *        City water system as a supply to auxiliary boiler feedwater (ABFW) when the
              condensate storage tank was declared inoperable due to leakage;
    *        21 and 23 ABFW trains when 22 ABFW pump was tagged out and temporary
              modifications were applied to 21 and 23 ABFW minimum flow lines.
b.  Findings
      No findings of significance were identified.
1R05 Fire Protection (71111.05Q - 5 samples)
a.  Inspection Scope
      The inspectors conducted tours of several fire areas to assess the material condition and
      operational status of fire protection features. The inspectors verified, consistent with the
      applicable administrative procedures, that: combustibles and ignition sources were
      adequately controlled; passive fire barriers, manual fire-fighting equipment, and
      suppression and detection equipment were appropriately maintained; and compensatory
      measures for out-of-service, degraded, or inoperable fire protection equipment were
      implemented in accordance with Entergys fire protection program. The inspectors
      evaluated the fire protection program for conformance with the requirements of License
      Condition 2.K. The documents reviewed during this inspection are listed in the
      Attachment. This inspection represented five inspection samples for fire protection
      tours, and was conducted in the following areas:
      *        FZ 65, Main Steam/Feed Regulating Valve Areas;
      *        FZ 23, 62A Auxiliary Feed Pump Room & Building;
      *        FZ 14, 480V Vital AC Switchgear Room;
      *        FZ 10, Emergency Diesel Generator Building; and
      *        FZ 360, Station Blackout Diesel Area.
                                                                                        Enclosure


Description.  On January 17, 2009, during a period of sustained cold weather which included sub-zero temperatures, control room personnel received a fire panel trouble alarm indicative of a low-pressure condition in the fire header and dispatched a plant operator to investigate.  The operator identified spraying water from the body of a ruptured six-inch fire protection valve located in the 11 fire pump house.  The operator isolated the broken valve from the fire header by shutting a manually-operated upstream
                                              11
valve which stopped the water spray.  In addition, the operator observed that the pump house room was significantly colder than expected and subsequently identified the room's ventilation louvers to the outside were mechanically bound in the open position.  The operator disconnected the louver linkage and manually shut the louvers. 
 
9 Enclosure  On January 21, 2009, the inspectors identified a second six inch valve that was cracked due to the previous cold weather (freezing) conditions in the fire pump house.  Entergy personnel entered this issue into the corrective action program and performed site walkdowns to identify additional adverse conditions associated with the cold weather. 
The inspectors determined that Entergy did not fully implement Entergy procedure EN-LI-102, Corrective Action Process.  Specifically, EN-LI-102 requires plant personnel to identify adverse conditions, including cold-weather related conditions, and then enter them into the CAP for resolution.  Attachment 9.2 of the procedure provides examples of adverse conditions expected to be reported; Section 1 of the Attachment contains
examples of operational conditions requiring entry into the CAP including "events or conditions that could negatively impact reliability or availability."  Additionally, plant operators should have had heightened awareness to cold weather conditions because Entergy procedure OAP-008, "Severe Weather Preparations," requires in step 4.3.7, when freezing conditions are expected, that increased monitoring of plant areas to monitor for adverse effects on plant equipment and verify that adequate protection is provided.  Operations personnel did not identify abnormal conditions in the 11 fire pump
room that led to the freezing and subsequent rupture of fire protection components. 
The inspectors determined it was reasonable for Entergy personnel to identify this issue
because operators should have identified that the louvers failed to shut following a
routine operations test of 11 fire pump on January 14, 2009.  In addition, operators perform tours of the pump house every 12 hours and should have identified the room was much colder than normal.   
Analysis.  The inspectors identified a performance deficiency because Entergy personnel did not implement procedure guidance and identify stuck open louvers and a subsequent second cracked fire header valve in the 11 fire pump house.  The finding was more than minor because it was associated with the protection against external
factors attribute of the Mitigating Systems cornerstone and it affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences.  Specifically, the failure of the six-inch valves impacted the reliability of the fire header until the ruptured valve was isolated. 
This finding was evaluated using Phase 1 of Inspection Manual Chapter (IMC) 0609 Appendix F, "Fire Protection Significance Determination Process."  The inspectors determined the issue was of very low safety significance (Green) because the cracked fire valves were easily isolated and did not pass sufficient water to render the fire header non-functional.  Specifically, the inspectors assigned a low degradation rating to the fire header because the fire pumps were able to maintain pressure in the fire header
until the ruptured valves were isolated. 
The inspectors determined that the finding had a cross-cutting aspect in the area of
human performance related to work practices - human error prevention techniques. 
Specifically, Entergy personnel routinely tour the 11 fire pump house did not question the abnormally cold room temperatures. (H.4(a) per IMC 0305)
Enforcement:  Enforcement action does not apply because the performance deficiency did not involve a violation of a regulatory requirement.  Because this finding does not involve a violation of regulatory requirements and has very low safety significance, it is identified as FIN 05000247/2009002-01, Failure to Identify Open Louvers in 11 Fire Pump House.
 
10 Enclosure  1R04 Equipment Alignment (71111.04Q - 3 samples)
  Partial System Walkdowns
    a. Inspection Scope
  The inspectors performed partial system walkdowns to verify the operability of redundant or diverse trains and components during periods of system train unavailability, or following periods of maintenance.  The inspectors referenced the system procedures, the UFSAR, and system drawings to verify the alignment of the available train supported its required safety functions.  The inspectors also reviewed applicable condition reports
(CR) and work orders to ensure Entergy personnel identified and properly addressed equipment discrepancies that could potentially impair the capability of the available train, as required by Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, "Corrective Action."  The documents reviewed during these inspections are listed in the Attachment. 
The inspectors performed a partial walkdown on the following systems, which
represented three inspection samples:
* 21 and 22 component cooling water (CCW) system train when 23 CCW pump was tagged out for maintenance;
* City water system as a supply to auxiliary boiler feedwater (ABFW) when the condensate storage tank was declared inoperable due to leakage;
* 21 and 23 ABFW trains when 22 ABFW pump was tagged out and temporary modifications were applied to 21 and 23 ABFW minimum flow lines.
  b. Findings
  b. Findings
  No findings of significance were identified.  
.1  Failure to Identify Damaged Components in EDG Ventilation Motor Control Center
    Introduction: The inspectors identified a NCV of very low safety significance (Green)
    related to 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, because Entergy
    personnel did not promptly identify and correct an adverse condition related to an
    electrical fault. Specifically, personnel did not identify a safety-related cubicle (bucket)
    had experienced a fault prior to replacement of upstream fuses and restoration of power
    to the cubicle.
    Description: On January 28, 2009, operations personnel detected an acrid odor coming
    from the emergency diesel generator (EDG) building. Operators entered the EDG
    building to investigate the source of the acrid odor and identified that a MCC was de-
    energized. Operations personnel did not identify external damage to the MCC; however,
    operators did not open MCC panels to inspect for internal damage. Operators checked
    the upstream 175 amp supply fuses, located in a different building, and identified that 2
    of 3 fuses had blown. Operators opened the downstream breakers on the MCC in the
    EDG building and then replaced the 175 amp supply fuses in the control building. Once
    operators replaced the blown fuses, they re-energized the EDG building MCC#1, and
    subsequently began to locally shut all of the cubicle switches. When operators
    attempted to shut the switch associated with cubicle 4N, the switch did not function as
    expected. Operators then opened the panel for cubicle 4N and identified charred
    electrical components.
    Entergy personnel generated a D level condition report (CR) for cubicle 4N on the
    basis that it supplies a non safety-related (NSR) EDG room heater. Entergy personnel
    closed the CR to a work request to troubleshoot and repair the NSR heater. However,
    the inspectors questioned the classification of the MCC and determined that the charred
    components were safety related (SR). Cubicle 4N contains a SR main line switch and
    SR 30 amp main line fuses. The 30 amp fuses are SR to isolate the NSR heaters from
    the MCC in the event of a room heater fault. The inspectors also questioned the
    appropriateness of leaving the damaged cubicle in the energized MCC. Following
    inspector questions, Entergy staff issued another CR and removed the damaged cubicle
    from the MCC on February 11. During removal of the charred cubicle, maintenance
    personnel were unable to disconnect the main line cables due to arc-welding at the
    termination and subsequently had to cut two of the three cables upstream of the
    termination and cubicle switch. These cables and the line side of the switch were
    energized from January 28 until February 11. After the damaged cubicle was removed,
    engineering personnel performed an inspection and determined that the fault originated
    from a high resistance connection on the C phase between the main fuse clip and the
    cubicle supply switch in the 4N cubicle.
    The inspectors determined that replacing the upstream 175 Amp fuses on and restoring
    power to the EDG ventilation MCC #1, which contained the charred 4N cubicle, without
    identifying the source of the acrid odor could have reinitiated the fault and increased the
    probability of a fire. In addition, operations personnel tried to locally close the damaged
    switch which could have also re-initiated the fault. Entergy staff also did not take action
    to remove or de-energize the charred cubicle after the condition was identified on
    January 28, 2009. The damaged cubicle was de-energized and removed from the MCC
    on February 11 in response to the inspectors questions.
                                                                                        Enclosure


1R05 Fire Protection (71111.05Q - 5 samples)
                                            12
  a. Inspection Scope
This issue was reasonable for the licensee to foresee and correct because acrid odor is
  The inspectors conducted tours of several fire areas to assess the material condition and operational status of fire protection features. The inspectors verified, consistent with the applicable administrative procedures, that: combustibles and ignition sources were adequately controlled; passive fire barriers, manual fire-fighting equipment, and suppression and detection equipment were appropriately maintained; and compensatory measures for out-of-service, degraded, or inoperable fire protection equipment were implemented in accordance with Entergy's fire protection program.  The inspectors
an indication of a fault. It was reasonable for Entergy personnel to open panel doors
evaluated the fire protection program for conformance with the requirements of License Condition 2.K. The documents reviewed during this inspection are listed in the Attachment. This inspection represented five inspection samples for fire protection tours, and was conducted in the following areas:
and perform visual inspections of the affected MCC prior to replacing upstream fuses
* FZ 65, Main Steam/Feed Regulating Valve Areas;
and restoring power to the fault. The inspectors determined that the National Electrical
* FZ 23, 62A Auxiliary Feed Pump Room & Building;
Code NFPA 70E, Standard for Electrical Safety in the Workplace, prohibits
* FZ 14, 480V Vital AC Switchgear Room;
reenergizing a circuit after a protective device has operated until it has been determined
* FZ 10, Emergency Diesel Generator Building; and  
that the automatic operation was a result of an overload and not a fault. The acrid odor
* FZ 360, Station Blackout Diesel Area.
in the EDG building was an indication of a fault vice an overload condition. In addition,
11 Enclosure  b. Findings
once Entergy personnel identified the cubicle was charred and experienced an electrical
  .1 Failure to Identify Damaged Components in EDG Ventilation Motor Control Center
fault, industry standards would have operators immediately secure power and/or
  Introduction: The inspectors identified a NCV of very low safety significance (Green) related to 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions," because Entergy
remove the damaged gear from the MCC.
personnel did not promptly identify and correct an adverse condition related to an electrical fault. Specifically, personnel did not identify a safety-related cubicle (bucket) had experienced a fault prior to replacement of upstream fuses and restoration of power to the cubicle. 
Entergy entered the issue into the corrective action program as IP2-2009-00342 and
IP2-2009-00483, trained all operations personnel on the requirements to replace fuses
Description: On January 28, 2009, operations personnel detected an acrid odor coming from the emergency diesel generator (EDG) building.  Operators entered the EDG building to investigate the source of the acrid odor and identified that a MCC was de-energized. Operations personnel did not identify external damage to the MCC; however, operators did not open MCC panels to inspect for internal damage.  Operators checked the upstream 175 amp supply fuses, located in a different building, and identified that 2 of 3 fuses had blown.  Operators opened the downstream breakers on the MCC in the  
and re-energize electrical equipment, and plans to review operations procedures for
EDG building and then replaced the 175 amp supply fuses in the control building.  Once operators replaced the blown fuses, they re-energized the EDG building MCC#1, and subsequently began to locally shut all of the cubicle switches.  When operators attempted to shut the switch associated with cubicle 4N, the switch did not function as expected.  Operators then opened the panel for cubicle 4N and identified charred
operating electrical equipment.
electrical components. 
Analysis: The inspectors determined that Entergys failure to promptly identify an
Entergy personnel generated a 'D' level condition report (CR) for cubicle 4N on the basis that it supplies a non safety-related (NSR) EDG room heater.  Entergy personnel closed the CR to a work request to troubleshoot and repair the NSR heater.  However,
adverse condition associated with damaged electrical components constituted a
the inspectors questioned the classification of the MCC and determined that the charred components were safety related (SR).  Cubicle 4N contains a SR main line switch and SR 30 amp main line fuses.  The 30 amp fuses are SR to isolate the NSR heaters from the MCC in the event of a room heater fault. The inspectors also questioned the appropriateness of leaving the damaged cubicle in the energized MCC. Following
performance deficiency. This issue was more than minor because the finding was
inspector questions, Entergy staff issued another CR and removed the damaged cubicle from the MCC on February 11.  During removal of the charred cubicle, maintenance personnel were unable to disconnect the main line cables due to arc-welding at the termination and subsequently had to cut two of the three cables upstream of the termination and cubicle switch.  These cables and the line side of the switch were
associated with the external factors attribute of the Initiating Events cornerstone and
energized from January 28 until February 11. After the damaged cubicle was removed, engineering personnel performed an inspection and determined that the fault originated from a high resistance connection on the 'C' phase between the main fuse clip and the cubicle supply switch in the 4N cubicle.  
impacted the cornerstone objective of limiting the likelihood of those events that upset
The inspectors determined that replacing the upstream 175 Amp fuses on and restoring power to the EDG ventilation MCC #1, which contained the charred 4N cubicle, without
plant stability and challenge critical safety systems during shutdown as well as power
identifying the source of the acrid odor could have reinitiated the fault and increased the probability of a fire. In addition, operations personnel tried to locally close the damaged switch which could have also re-initiated the fault. Entergy staff also did not take action to remove or de-energize the charred cubicle after the condition was identified on January 28, 2009.  The damaged cubicle was de-energized and removed from the MCC
operations. Specifically, operations personnel did not identify the source of the acrid
on February 11 in response to the inspectors' questions. 
odor, indicative of an electrical fault, in the EDG building; re-energized damaged
12 Enclosure  This issue was reasonable for the licensee to foresee and correct because acrid odor is an indication of a fault.  It was reasonable for Entergy personnel to open panel doors and perform visual inspections of the affected MCC prior to replacing upstream fuses and restoring power to the fault. The inspectors determined that the National Electrical Code NFPA 70E, "Standard for Electrical Safety in the Workplace," prohibits reenergizing a circuit after a protective device has operated until it has been determined
electrical equipment; and left damaged electrical components (cubicle 4N) energized for
that the automatic operation was a result of an overload and not a fault. The acrid odor in the EDG building was an indication of a fault vice an overload condition.  In addition, once Entergy personnel identified the cubicle was charred and experienced an electrical fault, industry standards would have operators immediately secure power and/or remove the damaged gear from the MCC. 
14 days prior to its removal from the MCC. The inspectors determined these issues
increased the likelihood of a fire in the EDG building. The condition was evaluated by a
Senior Reactor Analyst utilizing Phase 2 of IMC 0609 Appendix F, Fire Protection
Significance Determination Process. It was determined that in the event of a fire
consuming the MCC, no transient would be placed on the plant and no components
required to safely shutdown the plant would be impacted. As a result, in accordance
with task 2.3.5 of Appendix F, the issue was screened to Green.
The inspectors determined that a cross-cutting aspect was associated with this finding
in the area of human performance related to conservative decision making. Specifically,
Entergys decision-making was non-conservative as it related to the processes used to
identify the source of the acrid odor; re-energize the damaged electrical equipment; and
keep a damaged electrical component energized for 14 days prior to its removal from
the MCC. (H.1(b) per IMC 0305)
Enforcement: 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, requires that
measures shall be established to assure conditions adverse to quality, such as failures
and malfunctions are promptly identified and corrected. Contrary to the above, on
January 28, 2009, operations personnel did not identify that a safety-related bucket had
experienced a fault prior to replacing upstream fuses and restoring power to the bucket.
In addition, after replacing the upstream fuses, operations personnel tried to locally shut
the damaged cubicle switch and left damaged equipment energized until February 11,
2009. Entergy entered the issue into the corrective action program as IP2-2009-00342
and IP2-2009-00483, trained all operations personnel on the requirements to replace
fuses and re-energize electrical equipment, and plans to review operations procedures
                                                                                  Enclosure


Entergy entered the issue into the corrective action program as IP2-2009-00342 and IP2-2009-00483, trained all operations personnel on the requirements to replace fuses and re-energize electrical equipment, and plans to review operations procedures for operating electrical equipment.
                                            13
Analysis: The inspectors determined that Entergy's failure to promptly identify an adverse condition associated with damaged electrical components constituted a performance deficiency.  This issue was more than minor because the finding was associated with the external factors attribute of the Initiating Events cornerstone and impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety systems during shutdown as well as power
  for operating electrical equipment. Because the violation was of very low safety
operations. Specifically, operations personnel did not identify the source of the acrid odor, indicative of an electrical fault, in the EDG building; re-energized damaged electrical equipment; and left damaged electrical components (cubicle 4N) energized for 14 days prior to its removal from the MCC. The inspectors determined these issues increased the likelihood of a fire in the EDG building. The condition was evaluated by a  
  significance and it was entered into the licensees corrective action program, this
Senior Reactor Analyst utilizing Phase 2 of IMC 0609 Appendix F, "Fire Protection Significance Determination Process."  It was determined that in the event of a fire consuming the MCC, no transient would be placed on the plant and no components required to safely shutdown the plant would be impacted.  As a result, in accordance with task 2.3.5 of Appendix F, the issue was screened to Green.  
  violation is being treated as an NCV, consistent with the NRC Enforcement Policy: NCV
  05000247/2009002-02, Failure to Identify Damaged Components in EDG
  Ventilation Motor Control Center.
.2 Degraded Fire Door to the 480V Vital Bus Room
  Introduction: The inspectors identified a NCV of very low safety significance (Green)
  related to License Condition 2.K., fire protection program, because Entergy personnel
  did not promptly identify and correct a degraded three-hour rated fire door on the west
  entrance of the 480 Volt switchgear room.
  Description: License Condition 2.K., fire protection program, requires that Entergy
  implement and maintain in effect all provisions of the NRC-approved fire protection
  program, as approved in part by the NRC Safety Evaluation Report (SER) dated
  January 31, 1979. The January 31, 1979, SER requires administrative controls
  comparable to those described in NRC Branch Technical Position 9.5-1, Guidelines for
  Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976. Branch
  Technical Position (BTP) 9.5-1 requires that measures be established to assure that
  conditions adverse to fire protection, such as deficiencies, deviations, defective
  components, and non-conformities are promptly identified, reported, and corrected.
  On February 6, 2009, the inspectors performed a fire protection walkdown of the 480-
  Volt switchgear room. The inspectors noted the three-hour rated, swing-type fire door
  on the west side of the 480-Volt switchgear room was not latched closed. The
  inspectors observed the door being held open by the latch mechanism which had not
  repositioned to allow the door to shut. The inspectors observed the latch mechanism
  did not move freely preventing the door from shutting automatically. The inspectors
  shut the door and notified shift operations personnel who tightened latch screws on the
  door and wrote a condition report.
  On February 18, the inspectors identified the 480-Volt switchgear room door was not
  latched shut again. The inspectors determined the door could not be closed due to
  interference from the latch mechanism screw which had backed out. The inspectors
  notified operations of the fire door issue. Operations personnel re-inserted the latch
  mechanism screw and documented the issue in a condition report. The inspectors
  questioned whether it was appropriate to re-insert a screw that had backed out on its
  own in such a short period of time. Entergy personnel subsequently inspected the door
  on February 23 and identified the screws holding the latch mechanism to the door were
  stripped. Entergy personnel tapped new holes in the door latch mechanism and
  installed new screws.
  On March 3, inspectors identified the 480-Volt switchgear room fire door not latched
  shut again. The inspectors observed the door was being held open by the latch
  mechanism which had not repositioned to allow the door to shut. The inspectors noted
  the latch mechanism did not move freely preventing the door from shutting
  automatically. The inspectors notified operations personnel of the non-functioning fire
  door and Entergy subsequently had a locksmith inspect the latch. The locksmith
  installed a new latch mechanism on March 3 and determined the latch issues observed
  were age-related due to interaction of wear products from the latch interfering with the
  moving portions of the latch, as a result of latching and unlatching door operations.
                                                                                    Enclosure


The inspectors determined that a cross-cutting aspect was associated with this finding in the area of human performance related to conservative decision making. Specifically, Entergy's decision-making was non-conservative as it related to the processes used to identify the source of the acrid odor; re-energize the damaged electrical equipment; and
                                                14
keep a damaged electrical component energized for 14 days prior to its removal from the MCC. (H.1(b) per IMC 0305) 
      Entergy entered the issue into the corrective action program on March 3, performed an
Enforcement: 10 CFR 50 Appendix B, Criterion XVI, "Corrective Action," requires that measures shall be established to assure conditions adverse to quality, such as failures and malfunctions are promptly identified and corrected.  Contrary to the above, on January 28, 2009, operations personnel did not identify that a safety-related bucket had
      inspection of all fire doors onsite, and identified and corrected issues with other required
experienced a fault prior to replacing upstream fuses and restoring power to the bucket. In addition, after replacing the upstream fuses, operations personnel tried to locally shut the damaged cubicle switch and left damaged equipment energized until February 11, 2009. Entergy entered the issue into the corrective action program as IP2-2009-00342 and IP2-2009-00483, trained all operations personnel on the requirements to replace
      fire doors.
fuses and re-energize electrical equipment, and plans to review operations procedures 
      Analysis: The inspectors identified a performance deficiency because Entergy personnel
13 Enclosure  for operating electrical equipment.  Because the violation was of very low safety significance and it was entered into the licensee's corrective action program, this
      did not identify and correct the non-functional fire door. The finding was more than
violation is being treated as an NCV, consistent with the NRC Enforcement Policy:
      minor because it is associated with the protection against external factors attribute of
NCV 05000247/2009002-02, Failure to Identify Damaged Components in EDG Ventilation Motor Control Center.
      the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring
.2 Degraded Fire Door to the 480V Vital Bus Room
      the reliability of systems that respond to initiating events to prevent undesirable
  Introduction: The inspectors identified a NCV of very low safety significance (Green) related to License Condition 2.K., fire protection program, because Entergy personnel did not promptly identify and correct a degraded three-hour rated fire door on the west
      consequences. Specifically, in the event of a large fire in the 480-Volt switchgear room
entrance of the 480 Volt switchgear room.
      or the turbine building, the affected fire door is credited to prevent the spread of fire from
Description: License Condition 2.K., fire protection program, requires that Entergy implement and maintain in effect all provisions of the NRC-approved fire protection program, as approved in part by the NRC Safety Evaluation Report (SER) dated January 31, 1979. The January 31, 1979, SER requires administrative controls comparable to those described in NRC Branch Technical Position 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976."  Branch  
      one area to the other area. This fire door, when degraded, impacts the reliability of
Technical Position (BTP) 9.5-1 requires that measures be established to assure that conditions adverse to fire protection, such as deficiencies, deviations, defective components, and non-conformities are promptly identified, reported, and corrected.  
      mitigating systems in the 480-Volt switchgear room that are relied upon during a large
On February 6, 2009, the inspectors performed a fire protection walkdown of the 480-
      fire in the turbine building, and vice versa.
Volt switchgear room. The inspectors noted the three-hour rated, swing-type fire door on the west side of the 480-Volt switchgear room was not latched closed. The inspectors observed the door being held open by the latch mechanism which had not repositioned to allow the door to shut.  The inspectors observed the latch mechanism did not move freely preventing the door from shutting automatically.  The inspectors
      This finding was evaluated using Phase 1 of IMC 0609 Appendix F, Fire Protection
shut the door and notified shift operations personnel who tightened latch screws on the door and wrote a condition report.
      Significance Determination Process. Since the area in question had a fire watch
On February 18, the inspectors identified the 480-Volt switchgear room door was not
      posted during the time the door was degraded, an adequate level of protection was
latched shut again. The inspectors determined the door could not be closed due to interference from the latch mechanism screw which had backed out.  The inspectors notified operations of the fire door issue.  Operations personnel re-inserted the latch mechanism screw and documented the issue in a condition report.  The inspectors
      maintained to compensate for the degraded door and resulted in the finding being of
questioned whether it was appropriate to re-insert a screw that had backed out on its own in such a short period of time.  Entergy personnel subsequently inspected the door on February 23 and identified the screws holding the latch mechanism to the door were stripped.  Entergy personnel tapped new holes in the door latch mechanism and installed new screws.
      very low safety significance. As such according to task 1.3.1, the inspectors determined
      the finding was Green.
      The inspectors determined that the finding had a cross-cutting aspect in the area of
      problem identification and resolution because Entergy personnel did not thoroughly
      evaluate a degraded fire door latch on several occasions, such that the resolution of the
      problems addressed the causes. (P.1(c) per IMC 0305)
      Enforcement: License Condition 2.K., fire protection program, requires that Entergy
      implement and maintain in effect all provisions of the NRC-approved fire protection
      program, as approved in part by the NRC Safety Evaluation Report (SER) dated
      January 31, 1979. The January 31, 1979, SER requires administrative controls
      comparable to those described in NRC Branch Technical Position 9.5-1, Guidelines for
      Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976. Branch
      Technical Position 9.5-1 requires that measures be established to assure that conditions
      adverse to fire protection, such as deficiencies, deviations, defective components, and
      non-conformities are promptly identified, reported, and corrected.
      Contrary to the above, Entergy personnel did not promptly identify and then
      subsequently correct the non-functional 480-Volt switchgear fire door. This fire door
      was identified by inspectors in a non-functional state on February 6, February 18, and
      again on March 3, 2009. Entergy entered the issue into the corrective action program
      as IP2-2009-00526, IP2-2009-00680, IP2-2009-00709, IP2-2009-00834, IP2-2009-
      00842, and IP2-2009-00843. Because the violation was of very low safety significance
      and it was entered into the licensees corrective action program, this violation is being
      treated as an NCV, consistent with the NRC Enforcement Policy: NCV
      05000247/2009002-03, Failure to Identify and Promptly Correct Degraded 480-Volt
      Switchgear Room Fire Door.
1R07 Heat Sink Performance (71111.07A - 1 sample)
  a.  Inspection Scope
                                                                                        Enclosure


On March 3, inspectors identified the 480-Volt switchgear room fire door not latched shut again.  The inspectors observed the door was being held open by the latch mechanism which had not repositioned to allow the door to shut.  The inspectors noted the latch mechanism did not move freely preventing the door from shutting automatically.  The inspectors notified operations personnel of the non-functioning fire door and Entergy subsequently had a locksmith inspect the latch.  The locksmith
                                                15
installed a new latch mechanism on March 3 and determined the latch issues observed were age-related due to interaction of wear products from the latch interfering with the moving portions of the latch, as a result of latching and unlatching door operations. 
      The inspectors selected the 22 component water heat exchanger for review to
14 Enclosure  Entergy entered the issue into the corrective action program on March 3, performed an inspection of all fire doors onsite, and identified and corrected issues with other required fire doors. 
      determine the heat exchangers readiness and availability to perform its safety functions.
Analysis: The inspectors identified a performance deficiency because Entergy personnel did not identify and correct the non-functional fire door.  The finding was more than
      The inspectors reviewed the design basis for the component, reviewed Entergy
minor because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences.  Specifically, in the event of a large fire in the 480-Volt switchgear room or the turbine building, the affected fire door is credited to prevent the spread of fire from
      commitments to NRC Generic Letter 89-13, and reviewed engineering reports that
one area to the other area.  This fire door, when degraded, impacts the reliability of mitigating systems in the 480-Volt switchgear room that are relied upon during a large fire in the turbine building, and vice versa.
      documented results of previous internal inspections. The inspectors also observed the
This finding was evaluated using Phase 1 of IMC 0609 Appendix F, "Fire Protection Significance Determination Process."  Since the area in question had a fire watch posted during the time the door was degraded, an adequate level of protection was
      disassembly, inspection, and cleaning of the heat exchanger and reviewed engineering
maintained to compensate for the degraded door and resulted in the finding being of very low safety significance.  As such according to task 1.3.1, the inspectors determined the finding was Green. 
      results of the inspection to verify that appropriate corrective actions were initiated for
The inspectors determined that the finding had a cross-cutting aspect in the area of
      deficiencies that were discovered. The inspectors reviewed documents for and verified
problem identification and resolution because Entergy personnel did not thoroughly evaluate a degraded fire door latch on several occasions, such that the resolution of the problems addressed the causes. (P.1(c) per IMC 0305)
      that the amount of tubes plugged within the heat exchanger did not exceed the
Enforcement: License Condition 2.K., fire protection program, requires that Entergy implement and maintain in effect all provisions of the NRC-approved fire protection program, as approved in part by the NRC Safety Evaluation Report (SER) dated January 31, 1979. The January 31, 1979, SER requires administrative controls comparable to those described in NRC Branch Technical Position 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976." Branch
      maximum amount allowed. Documents reviewed are listed in the appendix.
Technical Position 9.5-1 requires that measures be established to assure that conditions adverse to fire protection, such as deficiencies, deviations, defective components, and non-conformities are promptly identified, reported, and corrected. 
b. Findings
Contrary to the above, Entergy personnel did not promptly identify and then
    No findings of significance were identified.
subsequently correct the non-functional 480-Volt switchgear fire door.  This fire door was identified by inspectors in a non-functional state on February 6, February 18, and again on March 3, 2009.  Entergy entered the issue into the corrective action program as IP2-2009-00526, IP2-2009-00680, IP2-2009-00709, IP2-2009-00834, IP2-2009-00842, and IP2-2009-00843.  Because the violation was of very low safety significance and it was entered into the licensee's corrective action program, this violation is being treated as an NCV, consistent with the NRC Enforcement Policy:
1R11 Licensed Operator Requalification Program
NCV 05000247/2009002-03, Failure to Identify and Promptly Correct Degraded 480-Volt Switchgear Room Fire Door. 
    Quarterly Review (71111.11Q - 1 sample)
1R07 Heat Sink Performance (71111.07A - 1 sample)
a. Inspection Scope
    On February 23, 2009, the inspectors observed licensed operator simulator training
  a. Inspection Scope
    associated with a sustained loss of all alternating current (AC) power scenario, to verify
 
    that operator performance was adequate, and that evaluators were identifying and
15 Enclosure 
    documenting crew performance problems. The inspectors evaluated the performance of
  The inspectors selected the 22 component water heat exchanger for review to determine the heat exchanger's readiness and availability to perform its safety functions. The inspectors reviewed the design basis for the component, reviewed Entergy commitments to NRC Generic Letter 89-13, and reviewed engineering reports that documented results of previous internal inspections. The inspectors also observed the  
    risk-significant operator actions, including the use of emergency operating procedures.
disassembly, inspection, and cleaning of the heat exchanger and reviewed engineering results of the inspection to verify that appropriate corrective actions were initiated for deficiencies that were discovered. The inspectors reviewed documents for and verified that the amount of tubes plugged within the heat exchanger did not exceed the maximum amount allowed. Documents reviewed are listed in the appendix.  
    The inspectors assessed the clarity and effectiveness of communications, the
  b. Findings
    implementation of appropriate actions in response to alarms, the performance of timely
  No findings of significance were identified.  
    control board operation and manipulation, and the oversight and direction provided by
1R11 Licensed Operator Requalification Program
    the control room supervisor. The inspectors also reviewed simulator fidelity with respect
  Quarterly Review (71111.11Q - 1 sample)  
    to the actual plant. The inspectors evaluated licensed operator training for conformance
  a. Inspection Scope
    with the requirements of 10 CFR Part 55, Operator Licenses. The documents
  On February 23, 2009, the inspectors observed licensed operator simulator training associated with a sustained loss of all alternating current (AC) power scenario, to verify that operator performance was adequate, and that evaluators were identifying and documenting crew performance problems. The inspectors evaluated the performance of risk-significant operator actions, including the use of emergency operating procedures. The inspectors assessed the clarity and effectiveness of communications, the  
    reviewed during this inspection are listed in the Attachment. This observation of
implementation of appropriate actions in response to alarms, the performance of timely control board operation and manipulation, and the oversight and direction provided by the control room supervisor. The inspectors also reviewed simulator fidelity with respect to the actual plant. The inspectors evaluated licensed operator training for conformance with the requirements of 10 CFR Part 55, "Operator Licenses."  The documents  
    operator simulator training represented one inspection sample.
reviewed during this inspection are listed in the Attachment. This observation of operator simulator training represented one inspection sample.  
b. Findings
  b. Findings
    No findings of significance were identified.
 
1R12 Maintenance Effectiveness (71111.12Q - 3 samples)
No findings of significance were identified.  
  a. Inspection Scope
1R12 Maintenance Effectiveness (71111.12Q - 3 samples)  
      The inspectors reviewed performance-based problems that involved structures,
  a. Inspection Scope
      systems, and components (SSCs) to assess the effectiveness of maintenance activities.
  The inspectors reviewed performance-based problems that involved structures,  
      When applicable, the reviews focused on:
systems, and components (SSCs) to assess the effectiveness of maintenance activities. When applicable, the reviews focused on:  
        *   Proper Maintenance Rule scoping in accordance with 10 CFR 50.65;
* Proper Maintenance Rule scoping in accordance with 10 CFR 50.65;  
        *   Characterization of reliability issues;
* Characterization of reliability issues;  
        *   Changing system and component unavailability;
* Changing system and component unavailability;
                                                                                          Enclosure
16 Enclosure
* 10 CFR 50.65(a)(1) and (a)(2) classifications;
* Identifying and addressing common cause failures;
* Trending of system flow and temperature values;
* Appropriateness of performance criteria for SSCs classified (a)(2); and
* Adequacy of goals and corrective actions for SSCs classified (a)(1).
The inspectors also reviewed system health reports, maintenance backlogs, and Maintenance Rule basis documents. The inspectors evaluated maintenance effectiveness and monitoring activities against the requirements of 10 CFR 50.65.  The documents reviewed during this inspection are listed in the Attachment. The following Maintenance Rule samples were reviewed and represented three inspection samples:
* RWST level indication system;
* EDG fuel injection system; and
* 480-Volt switchgear system.
  b. Findings
  Introduction: The inspectors identified a NCV of very low safety significance (Green) related to TS 5.4.1, "Administrative Controls: Procedures," because Entergy did not maintain an adequate maintenance procedure for a safety-related electrical motor control center (MCC).  Specifically, the eight-year maintenance procedure for the
affected EDG ventilation MCC did not contain an adequate method to identify high resistance connections within the cubicle.
Description: On January 28, 2009, operations personnel identified an acrid odor coming from the EDG building.  Subsequent personnel investigation revealed a charred cubicle
in a safety-related 480-Volt MCC.  Specifically, cubicle 4N, in the EDG ventilation MCC, experienced a phase-to-phase fault that caused the upstream 175 amp fuses to open and de-energize the MCC.  Entergy personnel subsequently generated a condition report (CR) that was closed to a work request to troubleshoot and repair the cubicle. 
Entergy personnel removed the damaged cubicle from the MCC on February 6 and determined the likely cause to be a high-resistance connection between the cubicle switch and 30 amp fuse clip on the 'C' phase resulting in long-term overheating.  This overheating condition degraded the insulation between two of the three phases over time and eventually resulted in a phase-to-phase fault on January 28, 2009. 


The inspectors reviewed the 8-year maintenance procedure 2-MCC-003-ELC, "Klockner-Moeller, Series 200, 480 Volt Motor Control Center Preventive Maintenance," which was performed on the affected EDG ventilation MCC on April 6, 2008. The inspectors noted that the procedure was revised the same day to allow performance of the maintenance without de-energizing the equipment. The revision resulted in portions of the cubicle cleaning and inspection procedure not being performed because they  
                                            16
could not be safely performed while the cubicle was energized. The inspectors determined that the procedure revision on April 6, 2008, was inappropriately treated as an editorial revision without a technical evaluation of the change performed. In addition, following interviews with Entergy personnel, it was determined that maintenance had not been performed on this MCC prior to April 6, 2008.
      *  10 CFR 50.65(a)(1) and (a)(2) classifications;
      *  Identifying and addressing common cause failures;
      *  Trending of system flow and temperature values;
      *  Appropriateness of performance criteria for SSCs classified (a)(2); and
      *  Adequacy of goals and corrective actions for SSCs classified (a)(1).
  The inspectors also reviewed system health reports, maintenance backlogs, and
  Maintenance Rule basis documents. The inspectors evaluated maintenance
  effectiveness and monitoring activities against the requirements of 10 CFR 50.65. The
  documents reviewed during this inspection are listed in the Attachment. The following
  Maintenance Rule samples were reviewed and represented three inspection samples:
  *  RWST level indication system;
  *  EDG fuel injection system; and
  *  480-Volt switchgear system.
b. Findings
  Introduction: The inspectors identified a NCV of very low safety significance (Green)
  related to TS 5.4.1, Administrative Controls: Procedures, because Entergy did not
  maintain an adequate maintenance procedure for a safety-related electrical motor
  control center (MCC). Specifically, the eight-year maintenance procedure for the
  affected EDG ventilation MCC did not contain an adequate method to identify high
  resistance connections within the cubicle.
  Description: On January 28, 2009, operations personnel identified an acrid odor coming
  from the EDG building. Subsequent personnel investigation revealed a charred cubicle
  in a safety-related 480-Volt MCC. Specifically, cubicle 4N, in the EDG ventilation MCC,
  experienced a phase-to-phase fault that caused the upstream 175 amp fuses to open
  and de-energize the MCC. Entergy personnel subsequently generated a condition
  report (CR) that was closed to a work request to troubleshoot and repair the cubicle.
  Entergy personnel removed the damaged cubicle from the MCC on February 6 and
  determined the likely cause to be a high-resistance connection between the cubicle
  switch and 30 amp fuse clip on the C phase resulting in long-term overheating. This
  overheating condition degraded the insulation between two of the three phases over
  time and eventually resulted in a phase-to-phase fault on January 28, 2009.
  The inspectors reviewed the 8-year maintenance procedure 2-MCC-003-ELC,
  Klockner-Moeller, Series 200, 480 Volt Motor Control Center Preventive Maintenance,
  which was performed on the affected EDG ventilation MCC on April 6, 2008. The
  inspectors noted that the procedure was revised the same day to allow performance of
  the maintenance without de-energizing the equipment. The revision resulted in portions
  of the cubicle cleaning and inspection procedure not being performed because they
  could not be safely performed while the cubicle was energized. The inspectors
  determined that the procedure revision on April 6, 2008, was inappropriately treated as
  an editorial revision without a technical evaluation of the change performed. In addition,
  following interviews with Entergy personnel, it was determined that maintenance had not
  been performed on this MCC prior to April 6, 2008.
                                                                                  Enclosure


 
                                          17
17 Enclosure  The inspectors reviewed industry guidance for performing switchgear maintenance and determined that Entergy did not include standard maintenance practices typically utilized by its staff that would have identified a high resistance connection in the cubicle. Specifically, continuity checks across contacts and switches were not performed, fuse clip tensions and tightness were not performed, and all terminations could not be checked due to the decision to perform the maintenance with portions of the cubicle  
The inspectors reviewed industry guidance for performing switchgear maintenance and
energized. In addition, the inspectors determined the EDG ventilation MCCs were not included in Entergy's thermography program, contrary to Entergy corporate preventive maintenance templates. The inspectors determined that not performing thermography on the EDG ventilation MCC constituted a missed opportunity to identify the high resistance condition.  
determined that Entergy did not include standard maintenance practices typically
utilized by its staff that would have identified a high resistance connection in the cubicle.
Specifically, continuity checks across contacts and switches were not performed, fuse
clip tensions and tightness were not performed, and all terminations could not be
checked due to the decision to perform the maintenance with portions of the cubicle
energized. In addition, the inspectors determined the EDG ventilation MCCs were not
included in Entergys thermography program, contrary to Entergy corporate preventive
maintenance templates. The inspectors determined that not performing thermography
on the EDG ventilation MCC constituted a missed opportunity to identify the high
resistance condition.
It is reasonable to consider the high resistance connection existed during the
maintenance performed on April 6, 2008, because high resistance connections do not
develop into phase-to-phase faults over a short period of time. This is an underlying
assumption for performing switchgear maintenance, which is intended to identify and
correct loose/high resistance connections, on an eight-year periodicity. In addition,
Entergys corporate template for switchgear maintenance recommends a six-year
periodicity and thermography every year. It is reasonable to expect Entergy to be aware
of the existing industry guidance as well as the Entergy corporate maintenance
templates.
Entergy entered the issue into the CAP as IP2-2009-00342 and IP2-2009-00483,
scoped the EDG ventilation MCC into the existing thermography program, performed an
extent-of-condition review that identified 21 additional panels that should be in the
thermography program, and plans to revise the maintenance procedure.
Analysis: The inspectors identified a performance deficiency because Entergy did not
maintain an adequate maintenance procedure for the safety-related EDG ventilation
MCC. This issue was more than minor because the finding was associated with the
external factors attribute of the Initiating Events cornerstone and impacted the initiating
events cornerstone objective of limiting the likelihood of those events that upset plant
stability and challenge critical safety systems during shutdown as well as power
operations. Specifically, the high resistance connection degraded into a phase-to-phase
fault and increased the likelihood of a fire in the EDG building. The condition was
evaluated by a Senior Reactor Analyst utilizing Phase 2 of IMC 0609 Appendix F, Fire
Protection Significance Determination Process. It was determined that in the event of a
fire consuming the MCC, no transient would be placed on the plant and no components
required to safely shutdown the plant would be impacted. As a result, in accordance
with task 2.3.5 of Appendix F, the issue was screened to Green.
The inspectors determined that the finding had a cross-cutting aspect associated with
the area of problem identification and resolution related to the use of operating
experience (OE). Specifically, Entergy personnel did not implement industry
recommended practices, or an alternate equivalent method, for identifying high
resistance connections in electrical switchgear. (P.2(b) per IMC 0305)
Enforcement. TS 5.4.1 Administrative Controls: Procedures, states, Written
procedures shall be established, implemented, and maintained covering the
requirements and recommendations of Appendix A of Regulatory Guide (RG) 1.33,
Revision 2. Appendix A of RG 1.33 requires procedures for maintenance activities that
                                                                                  Enclosure


It is reasonable to consider the high resistance connection existed during the maintenance performed on April 6, 2008, because high resistance connections do not develop into phase-to-phase faults over a short period of time.  This is an underlying assumption for performing switchgear maintenance, which is intended to identify and correct loose/high resistance connections, on an eight-year periodicity.  In addition, Entergy's corporate template for switchgear maintenance recommends a six-year
                                              18
periodicity and thermography every year.  It is reasonable to expect Entergy to be aware of the existing industry guidance as well as the Entergy corporate maintenance templates. 
      can affect the performance of safety related equipment. Contrary to the above, Entergy
Entergy entered the issue into the CAP as IP2-2009-00342 and IP2-2009-00483,
      did not maintain a maintenance procedure for a safety-related MCC cubicle.
scoped the EDG ventilation MCC into the existing thermography program, performed an extent-of-condition review that identified 21 additional panels that should be in the thermography program, and plans to revise the maintenance procedure.
      Specifically, the eight-year maintenance procedure, first performed on April 6, 2008, did
Analysis: The inspectors identified a performance deficiency because Entergy did not maintain an adequate maintenance procedure for the safety-related EDG ventilation MCC.  This issue was more than minor because the finding was associated with the external factors attribute of the Initiating Events cornerstone and impacted the initiating events cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety systems during shutdown as well as power
      not contain an adequate method to identify and correct high resistance connections in
operations.  Specifically, the high resistance connection degraded into a phase-to-phase fault and increased the likelihood of a fire in the EDG building.  The condition was evaluated by a Senior Reactor Analyst utilizing Phase 2 of IMC 0609 Appendix F, "Fire Protection Significance Determination Process."  It was determined that in the event of a fire consuming the MCC, no transient would be placed on the plant and no components
      the cubicle. Entergy entered the issue into the CAP as IP2-2009-00342 and IP2-2009-
required to safely shutdown the plant would be impacted.  As a result, in accordance with task 2.3.5 of Appendix F, the issue was screened to Green.
      00483. Because the violation was of very low safety significance and it was entered into
The inspectors determined that the finding had a cross-cutting aspect associated with the area of problem identification and resolution related to the use of operating experience (OE).  Specifically, Entergy personnel did not implement industry recommended practices, or an alternate equivalent method, for identifying high
      the licensees corrective action program, this violation is being treated as an NCV,
resistance connections in electrical switchgear. (P.2(b) per IMC 0305)
      consistent with the NRC Enforcement Policy: NCV 05000247/2009002-04, Inadequate
Enforcement.  TS 5.4.1 Administrative Controls: Procedures, states, "Written procedures shall be established, implemented, and maintained covering the requirements and recommendations of Appendix A of Regulatory Guide (RG) 1.33,
      Maintenance Procedure for EDG Ventilation Motor Control Center.
Revision 2." Appendix A of RG 1.33 requires procedures for maintenance activities that 
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 6 samples)
18 Enclosure  can affect the performance of safety related equipment. Contrary to the above, Entergy did not maintain a maintenance procedure for a safety-related MCC cubicle. Specifically, the eight-year maintenance procedure, first performed on April 6, 2008, did not contain an adequate method to identify and correct high resistance connections in the cubicle. Entergy entered the issue into the CAP as IP2-2009-00342 and IP2-2009-00483. Because the violation was of very low safety significance and it was entered into  
a. Inspection Scope
the licensee's corrective action program, this violation is being treated as an NCV,  
    The inspectors reviewed scheduled and emergent maintenance activities to verify the
consistent with the NRC Enforcement Policy: NCV 05000247/2009002-04, Inadequate Maintenance Procedure for EDG Ventilation Motor Control Center.  
    appropriate risk assessments were performed prior to removing equipment from service
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 6 samples)  
    for maintenance or repair. The inspectors verified that risk assessments were performed
  a. Inspection Scope
    as required by 10 CFR 50.65(a)(4) and were accurate and complete. When emergent
  The inspectors reviewed scheduled and emergent maintenance activities to verify the appropriate risk assessments were performed prior to removing equipment from service for maintenance or repair. The inspectors verified that risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate and complete. When emergent  
    work was performed, the inspectors verified the plant risk was promptly reassessed and
work was performed, the inspectors verified the plant risk was promptly reassessed and managed. Documents reviewed during this inspection are listed in the Attachment. The following activities represented six inspection samples:  
    managed. Documents reviewed during this inspection are listed in the Attachment. The
* Emergent maintenance on the 22 EDG lube oil pump during the 23 EDG maintenance outage;  
    following activities represented six inspection samples:
* Planned risk during 21 auxiliary boiler feedwater (ABFW) pump outage and reactor protection system testing;  
    *   Emergent maintenance on the 22 EDG lube oil pump during the 23 EDG
* Unplanned elevated risk condition due to delayed work on reactor protection system components during planned maintenance of 22 ABFW pump;  
          maintenance outage;
* Planned maintenance on a reactor water storage tank level indicator;  
    *   Planned risk during 21 auxiliary boiler feedwater (ABFW) pump outage and reactor
* Planned maintenance on the 22 ABFW pump while temporary modifications were applied to the 21 and 23 ABFW pumps; and  
          protection system testing;
* Planned risk during 23 EDG testing and maintenance.   
    *   Unplanned elevated risk condition due to delayed work on reactor protection system
  b. Findings
          components during planned maintenance of 22 ABFW pump;
  Introduction: The inspectors identified a NCV of very low safety significance (Green) related to 10 CFR 50.65(a)(4) because Entergy staff did not adequately assess the risk associated with the unavailability of the Refueling Water Storage Tank (RWST) level indication during planned maintenance on the level transmitters and instrumentation.
    *   Planned maintenance on a reactor water storage tank level indicator;
    *   Planned maintenance on the 22 ABFW pump while temporary modifications were
Description: On February 6, 2009, Entergy staff performed maintenance on the RWST level indication system. The inspectors identified that the online risk assessment did not consider planned maintenance on the RWST level indication, as required by 10 CFR 50.65(a)(4). The inspectors reviewed the work activity and noted the maintenance scheduling software used by Entergy did not have the RWST maintenance coded as a  
          applied to the 21 and 23 ABFW pumps; and
risk-significant activity. Entergy's maintenance planning process prompts the organization to evaluate the risk impact of all maintenance activities coded as risk-significant. Therefore, a risk assessment was not performed for the quarterly RWST level indication maintenance as required. In addition, the RWST level indication was not represented in Entergy's interactive risk model. Entergy staff subsequently updated the  
    *   Planned risk during 23 EDG testing and maintenance.
risk model to include the RWST level indication and subsequently assessed the online
  b. Findings
19 Enclosure risk for the maintenance which resulted in a measurable increase in the core damage frequency (CDF).  The increase in CDF was not large enough to require entrance into the higher risk category per Entergy procedures.  In addition, the increase in CDF (1.1E-6) combined with the limited duration of the maintenance (15 hours) resulted in a relatively small incremental core damage probability deficit (1.9E-9).
    Introduction: The inspectors identified a NCV of very low safety significance (Green)
    related to 10 CFR 50.65(a)(4) because Entergy staff did not adequately assess the risk
The inspectors determined this same maintenance activity is modeled in the Indian Point Unit 3 risk model.  Entergy entered the issue into the corrective action program (CR-IP2-2009-00342), updated the risk model to include the maintenance activity, assessed the risk, and appropriately coded the maintenance activity to ensure it would be risk assessed in the future. 
    associated with the unavailability of the Refueling Water Storage Tank (RWST) level
  Analysis:  The inspectors identified a performance deficiency in that Entergy staff did not assess the increase in plant risk resulting from planned maintenance activities on RWST level instrumentation as required by 10 CFR 50.65(a)(4).  The inspectors determined that this finding was more than minor because it was a risk assessment issue in which Entergy personnel did not consider risk significant SSCs that were unavailable during maintenance.  Specifically, RWST level indication is included in Table 2 of the plant
    indication during planned maintenance on the level transmitters and instrumentation.
specific Phase 2 SDP risk-informed inspection notebook.  The inspectors assessed the significance of this issue in accordance with IMC 0609, Appendix K, "Maintenance Risk Assessment and Risk Management Significance Determination Process."  The inspectors determined that this finding was of very low safety significance (Green) because the incremental core damage probability deficit was less than 1E-6.   
    Description: On February 6, 2009, Entergy staff performed maintenance on the RWST
    level indication system. The inspectors identified that the online risk assessment did not
    consider planned maintenance on the RWST level indication, as required by 10 CFR
    50.65(a)(4). The inspectors reviewed the work activity and noted the maintenance
    scheduling software used by Entergy did not have the RWST maintenance coded as a
    risk-significant activity. Entergys maintenance planning process prompts the
    organization to evaluate the risk impact of all maintenance activities coded as risk-
    significant. Therefore, a risk assessment was not performed for the quarterly RWST
    level indication maintenance as required. In addition, the RWST level indication was not
    represented in Entergys interactive risk model. Entergy staff subsequently updated the
    risk model to include the RWST level indication and subsequently assessed the online
                                                                                      Enclosure


The inspectors determined that the finding had a cross-cutting aspect in human performance related work control. Specifically, Entergy personnel did not appropriately plan work activities by incorporating risk insights for affected plant equipment. (H.3(a) per IMC 0305)  
                                              19
    risk for the maintenance which resulted in a measurable increase in the core damage
    frequency (CDF). The increase in CDF was not large enough to require entrance into
    the higher risk category per Entergy procedures. In addition, the increase in CDF (1.1E-
    6) combined with the limited duration of the maintenance (15 hours) resulted in a
    relatively small incremental core damage probability deficit (1.9E-9).
    The inspectors determined this same maintenance activity is modeled in the Indian Point
    Unit 3 risk model. Entergy entered the issue into the corrective action program (CR-IP2-
    2009-00342), updated the risk model to include the maintenance activity, assessed the
    risk, and appropriately coded the maintenance activity to ensure it would be risk
    assessed in the future.
    Analysis: The inspectors identified a performance deficiency in that Entergy staff did not
    assess the increase in plant risk resulting from planned maintenance activities on RWST
    level instrumentation as required by 10 CFR 50.65(a)(4). The inspectors determined
    that this finding was more than minor because it was a risk assessment issue in which
    Entergy personnel did not consider risk significant SSCs that were unavailable during
    maintenance. Specifically, RWST level indication is included in Table 2 of the plant
    specific Phase 2 SDP risk-informed inspection notebook. The inspectors assessed the
    significance of this issue in accordance with IMC 0609, Appendix K, Maintenance Risk
    Assessment and Risk Management Significance Determination Process. The
    inspectors determined that this finding was of very low safety significance (Green)
    because the incremental core damage probability deficit was less than 1E-6.
    The inspectors determined that the finding had a cross-cutting aspect in human
    performance related work control. Specifically, Entergy personnel did not appropriately
    plan work activities by incorporating risk insights for affected plant equipment. (H.3(a)
    per IMC 0305)
    Enforcement: 10 CFR 50.65 (a)(4) states, in part that licensees shall assess and
    manage the increase in risk that may result from the proposed maintenance activities
    before performing those activities. Contrary to the above, on February 6, 2009, Entergy
    performed maintenance on the RWST level indication system without assessing the
    increase in risk. Entergy entered the issue into the corrective action program (CR-IP2-
    2009-00342. Because this issue is of very low safety significance and is entered into
    Entergys corrective action program, this violation is being treated as an NCV consistent
    the NRC Enforcement Policy: NCV 05000247/2009002-05, Failure to Include RWST
    Level Maintenance In Online Risk Assessment.
1R15 Operability Evaluations (71111.15 - 7 samples)
a.  Inspection Scope
    The inspectors reviewed operability evaluations to assess the acceptability of the
    evaluations, the use and control of compensatory measures when applicable, and
    compliance with Technical Specifications. The inspectors reviews included verification
    that operability determinations were performed in accordance with procedure
    ENN-OP-104, Operability Determinations. The inspectors assessed the technical
    adequacy of the evaluations to ensure consistency with the Technical Specifications,
    UFSAR, and associated design basis documents. The documents reviewed are listed in
                                                                                      Enclosure


Enforcement:  10 CFR 50.65 (a)(4) states, in part that licensees shall assess and manage the increase in risk that may result from the proposed maintenance activities before performing those activities.  Contrary to the above, on February 6, 2009, Entergy performed maintenance on the RWST level indication system without assessing the
                                                  20
increase in risk.  Entergy entered the issue into the corrective action program (CR-IP2-2009-00342.  Because this issue is of very low safety significance and is entered into Entergy's corrective action program, this violation is being treated as an NCV consistent
    the Attachment. The following operability evaluations were reviewed and represented
the NRC Enforcement Policy: NCV 05000247/2009002-05, Failure to Include RWST Level Maintenance In Online Risk Assessment.
    seven inspection samples:
 
    *   Proximity of 480-Volt vital motor control center to an uninsulated steam line;
1R15 Operability Evaluations (71111.15 - 7 samples)
    *   Leakage from condensate storage tank (CST) return piping;
a. Inspection Scope
    *     Impacts of scaffolding built in the vicinity of the 21 and 22 component cooling water
  The inspectors reviewed operability evaluations to assess the acceptability of the evaluations, the use and control of compensatory measures when applicable, and
          heat exchangers;
compliance with Technical Specifications.  The inspectors' reviews included verification that operability determinations were performed in accordance with procedure ENN-OP-104, "Operability Determinations."  The inspectors assessed the technical adequacy of the evaluations to ensure consistency with the Technical Specifications, UFSAR, and associated design basis documents.  The documents reviewed are listed in 
    *     Impact on pressurizer surge line and reactor coolant system piping while performing
20 Enclosure  the Attachment. The following operability evaluations were reviewed and represented seven inspection samples:  
          reactor plant startups and shutdowns due to thermal transients;
* Proximity of 480-Volt vital motor control center to an uninsulated steam line;  
    *     Performance impact on the 21 and 22 auxiliary component cooling pumps (ACCPs)
* Leakage from condensate storage tank (CST) return piping;  
          with respect to a potential hydraulic lock-out condition of the 21 ACCP due to 22
* Impacts of scaffolding built in the vicinity of the 21 and 22 component cooling water heat exchangers;  
          ACCP larger impeller size;
* Impact on pressurizer surge line and reactor coolant system piping while performing reactor plant startups and shutdowns due to thermal transients;  
    *   Mechanical failure of a grease fitting on 21 service water pump; and
* Performance impact on the 21 and 22 auxiliary component cooling pumps (ACCPs) with respect to a potential hydraulic lock-out condition of the 21 ACCP due to 22 ACCP larger impeller size;  
    *   Low temperatures in condensate storage tank volume.
* Mechanical failure of a grease fitting on 21 service water pump; and  
b.   Findings
* Low temperatures in condensate storage tank volume.  
      No findings of significance were identified. With respect to the CST return piping, the
b. Findings
      inspectors determined Entergy operators maintained the CST aligned to supply water to
  No findings of significance were identified. With respect to the CST return piping, the inspectors determined Entergy operators maintained the CST aligned to supply water to the AFW pumps. The inspectors concluded the leakage did not prevent the CST from  
      the AFW pumps. The inspectors concluded the leakage did not prevent the CST from
fulfilling its safety function. Specifically, design features of the CST and the elevation of the return line relative to the leak location provided assurance that, in the event the CST return line leak increased significantly, the CST water volume would have been maintained above TS minimum required water level and able to supply the required water to the auxiliary feedwater system.  
      fulfilling its safety function. Specifically, design features of the CST and the elevation of
1R18 Plant Modifications (71111.18 - 2 samples)  
      the return line relative to the leak location provided assurance that, in the event the CST
.1 Temporary Modifications
      return line leak increased significantly, the CST water volume would have been
    a. Inspection Scope
      maintained above TS minimum required water level and able to supply the required
  The inspectors reviewed one temporary plant modification package for securing minimum flow lines on the motor driven auxiliary boiler feedwater pumps (ABFPs) and controlling the operation on the ABFPs through a temporary operating procedure during repairs of the CST return piping. The inspectors verified the design bases, licensing bases, and performance capability of the system was not degraded by the temporary  
      water to the auxiliary feedwater system.
modification. The inspectors' review included Entergy's engineering evaluation for determining the ABFPs could start with the pump's required minimum flow being achieved through the internal thrust balance lines while the minimum flow lines were isolated. In addition, the inspectors interviewed plant staff, and reviewed issues entered into the corrective action program to determine whether Entergy had been effective in  
1R18 Plant Modifications (71111.18 - 2 samples)
identifying and resolving problems associated with the temporary modification. The documents reviewed are listed in the Attachment. 
.1   Temporary Modifications
  b. Findings
a. Inspection Scope
  No findings of significance were identified.
    The inspectors reviewed one temporary plant modification package for securing
   
    minimum flow lines on the motor driven auxiliary boiler feedwater pumps (ABFPs) and
21 Enclosure  .2 Permanent Modifications
    controlling the operation on the ABFPs through a temporary operating procedure during
  a. Inspection Scope
    repairs of the CST return piping. The inspectors verified the design bases, licensing
  The inspectors reviewed modification documents associated with the installation of an additional nitrogen backup power supply for the 21- 24 steam generator atmospheric
    bases, and performance capability of the system was not degraded by the temporary
dump valves.  The inspector verified that the modification was reviewed adequately to verify the modification conformed to design criteria and did not interfere or invalidate previous design assumptions or functions.  The documents reviewed are listed in the Attachment.  
    modification. The inspectors review included Entergys engineering evaluation for
    determining the ABFPs could start with the pumps required minimum flow being
    achieved through the internal thrust balance lines while the minimum flow lines were
    isolated. In addition, the inspectors interviewed plant staff, and reviewed issues entered
    into the corrective action program to determine whether Entergy had been effective in
    identifying and resolving problems associated with the temporary modification. The
    documents reviewed are listed in the Attachment.
   b. Findings
   b. Findings
  No findings of significance were identified.  
    No findings of significance were identified.
1R19 Post-Maintenance Testing (71111.19 - 6 samples)
                                                                                          Enclosure
  a. Inspection Scope
  The inspectors reviewed post-maintenance test procedures and associated testing activities for selected risk-significant mitigating systems, and assessed whether the effect of maintenance on plant systems was adequately addressed by control room and engineering personnel.  The inspectors verified that: test acceptance criteria were clear,
the test demonstrated operational readiness and were consistent with design basis documentation; test instrumentation had current calibrations, and appropriate range and accuracy for the application; and the tests were performed as written, with applicable prerequisites satisfied.  Upon completion of the tests, the inspectors verified that equipment was returned to the proper alignment necessary to perform its safety function. 
Post-maintenance testing was evaluated for conformance with the requirements of 10 CFR 50, Appendix B, Criterion XI, "Test Control."  The documents reviewed are listed in the Attachment.  The following post-maintenance activities were reviewed and represented six inspection samples:
* Replacement of SG 23 pressure indicator PI-1355;
* 22 component cooling water heat exchanger following maintenance;
* 21 charging pump following recirculation valve maintenance;
* Condensate storage tank return line following pipe section replacement;
* Emergency diesel generator air compressor following quarterly maintenance; and
* 23 emergency diesel generator following quarterly engine maintenance.
  b. Findings
 
No findings of significance were identified.
1R22 Surveillance Testing (71111.22 - 6 samples)
  a. Inspection Scope
  The inspectors observed performance of portions of surveillance tests and/or reviewed test data for selected risk-significant SSCs to assess whether they satisfied Technical Specifications, UFSAR, Technical Requirements Manual, and Entergy procedure 
22 Enclosure  requirements.  The inspectors verified that: test acceptance criteria were identified, demonstrated operational readiness, and were consistent with design basis documentation; test instrumentation had accurate calibration, and appropriate range and accuracy for the application; and tests were performed as written, with applicable prerequisites satisfied.  Following the tests, the inspectors verified that the equipment was capable of performing the required safety functions.  The inspectors evaluated the
surveillance tests against the requirements in Technical Specifications.  The documents reviewed during this inspection are listed in the Attachment.  The following surveillance tests were reviewed and represented six inspection samples:
* 2-PT-Q031A, 21 Auxiliary Component Cooling Pump In-Service Test;
* 2-PT-Q054, Pressurizer Level Bistables;
* 2-PT-Q013 DS027, IST Valve Test of 888A (Safety Injection Pump Suction from Residual Heat Removal heat Exchanger);
* 2-PT-2M4, Safety Injection System Train "A" Actuation Logic and Master Relay Test;
* 2-PT-Q030C, 23 Component Cooling Water Pump; and
* 0-SOP-LEAKRATE-001, RCS Leakrate Surveillance, Evaluation, and Leak Identification.
  b.  Findings
  Introduction.  The inspectors identified a NCV of very low safety significance (Green) related to 10 CFR 50.55a, "Codes and standards," because Entergy's procedure 2-PT-
Q031A did not contain appropriate acceptance criteria for determining that safety-related check valves performed their safety function when required in accordance with the American Society of Mechanical Engineers (ASME) OM Code. 
Description.  Entergy procedure 2-PT-Q031A, "21 Auxiliary Component Cooling Pump (ACCP)", is an In-Service Test (IST) procedure that demonstrates the operability of the 21 ACCP, the pump bypass line check valve (755), the 21 ACCP discharge check valve (755B), and the 22 ACCP discharge check valve (755A) in accordance with Technical Specification (TS) 5.5.6, Inservice Testing Program.
The test established a single acceptance criterion to determine if the discharge check valve on the 22 ACCP train shuts when the parallel train's 21 ACCP is providing design
flow.  The acceptance criterion was that no reverse rotation is observed on the 22 ACCP.  Although NUREG-1482, "Guidelines for Inservice Testing at Nuclear Power Plants" identifies the methodology of using reverse pump rotation as an acceptable means of testing, Entergy's site-specific experience in 2005 demonstrated this particular method was not effective to maintain the ACCP discharge check valve safety function. 
Specifically, when 2-PT-Q031A was performed on January 19, 2005, the 21 ACCP failed the performance test because check valve 755A was determined to be in the open position.  However, the 22 ACCP did not rotate in the reverse direction.  Following disassembly of valve 755A, engineers determined the valve remained in the open position because of excessive clearances between the hinge pin and hinge pin
bushings.  Entergy personnel determined the check valve was likely in this condition following maintenance on the valve in late 2004.  CR-IP2-2005-0252 was written to document and evaluate the issue.  The issue was previously documented in LER 05000247/2005001-00 and NRC NCV 50-247/2005003-01.  At that time, Entergy personnel concluded the test criteria established in 2-PT-Q031A was acceptable but
that post-maintenance tests on the check valve should include amplifying comments 
23 Enclosure directing the performance of the IST following maintenance.  Entergy personnel concluded that the IST was adequate because the low pump head that caused the pump performance test to fail led to troubleshooting that identified that check valve 755A was stuck open.
The inspectors determined that the criterion for determining operability of 755A in test 2-
PT-Q013A was inadequate because the criterion in the procedure previously failed to identify that 755A remained in the open position in January 2005 and 2-PT-Q013A does not identify any other criteria, including using pump head, to determine operability of 755A.  Additionally, the inspectors determined the test criterion for check valve 755A and 755B were not consistent with the following ASME Code requirements:


  * The ASME OM Code 2001 Subsection ISTA-3160 states that "procedures shall contain the Owner-specified reference values and acceptance criteria";
                                                21
* The ASME OM Code 2001 Subsection ISTC-1400 (c) states "it is the Owner's responsibility to ensure that the application, method, and capability of each nonintrusive technique is qualified"; and
.2  Permanent Modifications
* The ASME OM Code 2001 Subsection ISTC-3530 states "obturator movement shall be determined by exercising the valve while observing an appropriate indicator."
  a.  Inspection Scope
Analysis. The inspectors determined that the performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating System cornerstone and adversely affected the cornerstone objective to ensure the  
    The inspectors reviewed modification documents associated with the installation of an
reliability of systems that respond to initiating events to prevent undesirable consequences.  Specifically, the test criterion used in procedure 2-PT-Q013A did not ensure that valve 755A reliably performed its safety function when tested as demonstrated by testing performed in January 2005. The inspectors determined that the performance deficiency was of very low safety significance (Green) using IMC 0609,
    additional nitrogen backup power supply for the 21- 24 steam generator atmospheric
Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings."  Specifically, the inspectors determined that this finding was of very low safety significance because the finding did not result in a loss of safety function and did not screen as potentially risk-significant due to external events initiating events.   
    dump valves. The inspector verified that the modification was reviewed adequately to
    verify the modification conformed to design criteria and did not interfere or invalidate
The inspectors determined the finding had a cross-cutting aspect related to effective corrective actions in the corrective action program component of the problem identification and resolution area.  Specifically, Entergy did not implement effective corrective actions to resolve the testing inadequacy since 2005 during subsequent quarterly testing.  Additionally, the issue was considered to be indicative of current
    previous design assumptions or functions. The documents reviewed are listed in the
performance because personnel when initially responding to inspector questions concluded the acceptance criteria were adequate. (P.1(d) per IMC 0305)
    Attachment.
Enforcement.  10 CFR 50.55a, "Codes and standards," states that pumps and valves which are classified as ASME code Class 1, Class 2, and Class 3 must meet the  
  b. Findings
inservice test requirements set forth in the ASME OM Code (2001 edition for Indian Point Unit 2).  Furthermore, inservice tests to verify operational readiness of pumps and valves, whose function is required for safety must comply with the requirements of the ASME OM Code.  The ASME OM Code 2001 Subsection ISTC-1400 (c) states "it is the Owner's responsibility to ensure that the application, method, and capability of each nonintrusive technique is qualified."  In addition, the ASME OM Code 2001 Subsection ISTC-3530 states "obturator movement shall be determined by exercising the valve 
    No findings of significance were identified.
24 Enclosure  while observing an appropriate indicator."  Contrary to the above, from February 2005 until February 2009, Entergy procedure 2-PT-Q031A, did not include appropriate acceptance criteria for demonstrating operability of valve 755A.  Specifically, the test did not utilize a qualified technique for testing the check-valve and did not verify check valve movement by observing an appropriate indicator.  Because ACCP performance tests since 2004 demonstrated satisfactory performance of the ACCPs at design flows, no
1R19 Post-Maintenance Testing (71111.19 - 6 samples)
actual impact to the operability of the ACCPs was evident. Because this violation was of very low safety significance and it was entered into Entergy's corrective action program (IP2-2009-1312), this violation is being treated as an NCV, consistent with the
   a. Inspection Scope
NRC Enforcement Policy.  NCV 2009002-06, Inadequate Test Acceptance Criteria for Auxiliary Component Cooling Check Valves. Cornerstone: Emergency Preparedness (EP)
    The inspectors reviewed post-maintenance test procedures and associated testing
 
    activities for selected risk-significant mitigating systems, and assessed whether the
1EP6 Drill Evaluation (71114.06 - 1 sample)
    effect of maintenance on plant systems was adequately addressed by control room and
  a. Inspection Scope
    engineering personnel. The inspectors verified that: test acceptance criteria were clear,
   The inspectors evaluated an emergency classification conducted on February 23, 2009, during a licensed-operator requalification simulator training evaluation.  The inspectors observed an operating crew in the simulator respond to various, simulated initiating events that ultimately resulted in the simulated implementation of the emergency plan.  In particular, the inspectors verified the adequacy and accuracy of the simulated
    the test demonstrated operational readiness and were consistent with design basis
emergency classification of a Site Area Emergency.  While other simulated classifications were made, the inspectors verified that the initial classification was appropriately credited as an opportunity toward NRC performance indicator data.  The inspectors observed the management evaluator and training critique following termination of the scenarios, and verified that significant performance deficiencies were
    documentation; test instrumentation had current calibrations, and appropriate range and
appropriately identified and addressed within the critique and the corrective action program. Also, the inspectors reviewed the summary performance report for the evaluation and verified that appropriate attributes of drill performance including deficiencies were captured.  This evaluation constituted one inspection sample.
    accuracy for the application; and the tests were performed as written, with applicable
    prerequisites satisfied. Upon completion of the tests, the inspectors verified that
    equipment was returned to the proper alignment necessary to perform its safety function.
    Post-maintenance testing was evaluated for conformance with the requirements of 10
    CFR 50, Appendix B, Criterion XI, Test Control. The documents reviewed are listed in
    the Attachment. The following post-maintenance activities were reviewed and
    represented six inspection samples:
    *   Replacement of SG 23 pressure indicator PI-1355;
    *  22 component cooling water heat exchanger following maintenance;
    *  21 charging pump following recirculation valve maintenance;
    *  Condensate storage tank return line following pipe section replacement;
    *  Emergency diesel generator air compressor following quarterly maintenance; and
    *  23 emergency diesel generator following quarterly engine maintenance.
   b. Findings
   b. Findings
   No findings of significance were identified.  
    No findings of significance were identified.
2. RADIATION SAFETY
1R22 Surveillance Testing (71111.22 - 6 samples)
  Cornerstone: Occupational Radiation Safety (OS)
  a. Inspection Scope
 
    The inspectors observed performance of portions of surveillance tests and/or reviewed
2OS1 Access Control to Radiologically Significant Areas (71121.01 - 16 samples)  
    test data for selected risk-significant SSCs to assess whether they satisfied Technical
  a. Inspection Scope
    Specifications, UFSAR, Technical Requirements Manual, and Entergy procedure
  From March 23 through March 27, 2009, the inspectors conducted the following activities to verify that Entergy was properly implementing physical, engineering, and administrative controls for access to high radiation areas, and other radiologically controlled areas, and that workers were adhering to these controls when working in these areas. Implementation of the access control program was reviewed against the
                                                                                      Enclosure
25 Enclosure  criteria contained in 10 CFR 20, site technical specifications, and Entergy's procedures required by the Technical Specifications as criteria for determining compliance.   This inspection activity represents completion of sixteen (16) samples relative to this inspection area. The inspector performed independent radiation dose rate measurements and reviewed the following items:  
 
                                            22
  requirements. The inspectors verified that: test acceptance criteria were identified,
  demonstrated operational readiness, and were consistent with design basis
  documentation; test instrumentation had accurate calibration, and appropriate range and
  accuracy for the application; and tests were performed as written, with applicable
  prerequisites satisfied. Following the tests, the inspectors verified that the equipment
  was capable of performing the required safety functions. The inspectors evaluated the
  surveillance tests against the requirements in Technical Specifications. The documents
  reviewed during this inspection are listed in the Attachment. The following surveillance
  tests were reviewed and represented six inspection samples:
  *    2-PT-Q031A, 21 Auxiliary Component Cooling Pump In-Service Test;
  *    2-PT-Q054, Pressurizer Level Bistables;
  *    2-PT-Q013 DS027, IST Valve Test of 888A (Safety Injection Pump Suction from
        Residual Heat Removal heat Exchanger);
  *    2-PT-2M4, Safety Injection System Train A Actuation Logic and Master Relay Test;
  *    2-PT-Q030C, 23 Component Cooling Water Pump; and
  *   0-SOP-LEAKRATE-001, RCS Leakrate Surveillance, Evaluation, and Leak
        Identification.
b. Findings
    Introduction. The inspectors identified a NCV of very low safety significance (Green)
    related to 10 CFR 50.55a, Codes and standards, because Entergys procedure 2-PT-
    Q031A did not contain appropriate acceptance criteria for determining that safety-
    related check valves performed their safety function when required in accordance with
    the American Society of Mechanical Engineers (ASME) OM Code.
    Description. Entergy procedure 2-PT-Q031A, 21 Auxiliary Component Cooling Pump
    (ACCP), is an In-Service Test (IST) procedure that demonstrates the operability of the
    21 ACCP, the pump bypass line check valve (755), the 21 ACCP discharge check valve
    (755B), and the 22 ACCP discharge check valve (755A) in accordance with Technical
    Specification (TS) 5.5.6, Inservice Testing Program.
    The test established a single acceptance criterion to determine if the discharge check
    valve on the 22 ACCP train shuts when the parallel trains 21 ACCP is providing design
    flow. The acceptance criterion was that no reverse rotation is observed on the 22
    ACCP. Although NUREG-1482, Guidelines for Inservice Testing at Nuclear Power
    Plants identifies the methodology of using reverse pump rotation as an acceptable
    means of testing, Entergys site-specific experience in 2005 demonstrated this particular
    method was not effective to maintain the ACCP discharge check valve safety function.
    Specifically, when 2-PT-Q031A was performed on January 19, 2005, the 21 ACCP
    failed the performance test because check valve 755A was determined to be in the
    open position. However, the 22 ACCP did not rotate in the reverse direction. Following
    disassembly of valve 755A, engineers determined the valve remained in the open
    position because of excessive clearances between the hinge pin and hinge pin
    bushings. Entergy personnel determined the check valve was likely in this condition
    following maintenance on the valve in late 2004. CR-IP2-2005-0252 was written to
    document and evaluate the issue. The issue was previously documented in LER
    05000247/2005001-00 and NRC NCV 50-247/2005003-01. At that time, Entergy
    personnel concluded the test criteria established in 2-PT-Q031A was acceptable but
    that post-maintenance tests on the check valve should include amplifying comments
                                                                                      Enclosure
 
                                          23
directing the performance of the IST following maintenance. Entergy personnel
concluded that the IST was adequate because the low pump head that caused the
pump performance test to fail led to troubleshooting that identified that check valve
755A was stuck open.
The inspectors determined that the criterion for determining operability of 755A in test 2-
PT-Q013A was inadequate because the criterion in the procedure previously failed to
identify that 755A remained in the open position in January 2005 and 2-PT-Q013A does
not identify any other criteria, including using pump head, to determine operability of
755A. Additionally, the inspectors determined the test criterion for check valve 755A
and 755B were not consistent with the following ASME Code requirements:
*    The ASME OM Code 2001 Subsection ISTA-3160 states that procedures shall
    contain the Owner-specified reference values and acceptance criteria;
*    The ASME OM Code 2001 Subsection ISTC-1400 (c) states it is the Owners
    responsibility to ensure that the application, method, and capability of each
    nonintrusive technique is qualified; and
*    The ASME OM Code 2001 Subsection ISTC-3530 states obturator movement
    shall be determined by exercising the valve while observing an appropriate
    indicator.
Analysis. The inspectors determined that the performance deficiency was more than
minor because it was associated with the procedure quality attribute of the Mitigating
System cornerstone and adversely affected the cornerstone objective to ensure the
reliability of systems that respond to initiating events to prevent undesirable
consequences. Specifically, the test criterion used in procedure 2-PT-Q013A did not
ensure that valve 755A reliably performed its safety function when tested as
demonstrated by testing performed in January 2005. The inspectors determined that
the performance deficiency was of very low safety significance (Green) using IMC 0609,
Attachment 4, Phase 1 - Initial Screening and Characterization of Findings.
Specifically, the inspectors determined that this finding was of very low safety
significance because the finding did not result in a loss of safety function and did not
screen as potentially risk-significant due to external events initiating events.
The inspectors determined the finding had a cross-cutting aspect related to effective
corrective actions in the corrective action program component of the problem
identification and resolution area. Specifically, Entergy did not implement effective
corrective actions to resolve the testing inadequacy since 2005 during subsequent
quarterly testing. Additionally, the issue was considered to be indicative of current
performance because personnel when initially responding to inspector questions
concluded the acceptance criteria were adequate. (P.1(d) per IMC 0305)
Enforcement. 10 CFR 50.55a, Codes and standards, states that pumps and valves
which are classified as ASME code Class 1, Class 2, and Class 3 must meet the
inservice test requirements set forth in the ASME OM Code (2001 edition for Indian
Point Unit 2). Furthermore, inservice tests to verify operational readiness of pumps and
valves, whose function is required for safety must comply with the requirements of the
ASME OM Code. The ASME OM Code 2001 Subsection ISTC-1400 (c) states it is the
Owners responsibility to ensure that the application, method, and capability of each
nonintrusive technique is qualified. In addition, the ASME OM Code 2001 Subsection
ISTC-3530 states obturator movement shall be determined by exercising the valve
                                                                                  Enclosure
 
                                                24
      while observing an appropriate indicator. Contrary to the above, from February 2005
      until February 2009, Entergy procedure 2-PT-Q031A, did not include appropriate
      acceptance criteria for demonstrating operability of valve 755A. Specifically, the test did
      not utilize a qualified technique for testing the check-valve and did not verify check valve
      movement by observing an appropriate indicator. Because ACCP performance tests
      since 2004 demonstrated satisfactory performance of the ACCPs at design flows, no
      actual impact to the operability of the ACCPs was evident. Because this violation was
      of very low safety significance and it was entered into Entergys corrective action
      program (IP2-2009-1312), this violation is being treated as an NCV, consistent with the
      NRC Enforcement Policy. NCV 2009002-06, Inadequate Test Acceptance Criteria
      for Auxiliary Component Cooling Check Valves.
    Cornerstone: Emergency Preparedness (EP)
1EP6 Drill Evaluation (71114.06 - 1 sample)
a.  Inspection Scope
    The inspectors evaluated an emergency classification conducted on February 23, 2009,
    during a licensed-operator requalification simulator training evaluation. The inspectors
    observed an operating crew in the simulator respond to various, simulated initiating
    events that ultimately resulted in the simulated implementation of the emergency plan.
    In particular, the inspectors verified the adequacy and accuracy of the simulated
    emergency classification of a Site Area Emergency. While other simulated
    classifications were made, the inspectors verified that the initial classification was
    appropriately credited as an opportunity toward NRC performance indicator data. The
    inspectors observed the management evaluator and training critique following
    termination of the scenarios, and verified that significant performance deficiencies were
    appropriately identified and addressed within the critique and the corrective action
    program. Also, the inspectors reviewed the summary performance report for the
    evaluation and verified that appropriate attributes of drill performance including
    deficiencies were captured. This evaluation constituted one inspection sample.
b.  Findings
    No findings of significance were identified.
2.   RADIATION SAFETY
    Cornerstone: Occupational Radiation Safety (OS)
2OS1 Access Control to Radiologically Significant Areas (71121.01 - 16 samples)
a. Inspection Scope
    From March 23 through March 27, 2009, the inspectors conducted the following
    activities to verify that Entergy was properly implementing physical, engineering, and
    administrative controls for access to high radiation areas, and other radiologically
    controlled areas, and that workers were adhering to these controls when working in
    these areas. Implementation of the access control program was reviewed against the
                                                                                        Enclosure
 
                                            25
criteria contained in 10 CFR 20, site technical specifications, and Entergys procedures
required by the Technical Specifications as criteria for determining compliance.
This inspection activity represents completion of sixteen (16) samples relative to this
inspection area. The inspector performed independent radiation dose rate
measurements and reviewed the following items:
Plant Walk Downs and Radiological Work Permit Reviews
Plant Walk Downs and Radiological Work Permit Reviews
  (1) Exposure significant work areas were identified by inspectors for review within radiation areas, high radiation areas, and airborne areas in the plant. Associated licensee controls and surveys were review for adequacy. Work reviewed  
(1)     Exposure significant work areas were identified by inspectors for review within
included: Refuel Floor Split Pin and Reactor Head Inspections, Refuel Floor Lower Internals Removal and Installation, Refuel Floor and Fuel Support Building Fuel Transport Equipment Repairs requiring an underwater diver, Reactor Coolant Pump work including RCP #31 Impeller replacement, Containment valve work including Pressurizer Safety Valves, Various Containment and Auxiliary Building activities.  
        radiation areas, high radiation areas, and airborne areas in the plant. Associated
        licensee controls and surveys were review for adequacy. Work reviewed
(2) With a survey instrument and assistance from a health physics technician, inspectors walked down the above mentioned areas to determine: whether the radiation work permits (RWPs), procedures and engineering controls were in place and whether surveys and postings were adequate.  
        included: Refuel Floor Split Pin and Reactor Head Inspections, Refuel Floor
        Lower Internals Removal and Installation, Refuel Floor and Fuel Support Building
(3) The inspectors reviewed RWPs that provide access to exposure significant areas of the plant including high radiation areas. Specified electronic personal dosimeter alarm set points were reviewed with respect to current radiological condition applicability and workers were queried to verify their understanding of plant procedures governing alarm response and knowledge of radiological  
        Fuel Transport Equipment Repairs requiring an underwater diver, Reactor
conditions in their work area.  
        Coolant Pump work including RCP #31 Impeller replacement, Containment valve
(4) There were no radiation work permits for airborne radioactivity areas with the potential for individual worker internal exposures of >50 mrem CEDE.  
        work including Pressurizer Safety Valves, Various Containment and Auxiliary
(5) There were no internal dose assessments that resulted in actual internal exposures greater than 50 mrem CEDE. Internal assessments were reviewed to determine adequacy and assurance that they were not in fact equal to or greater than 50 mrem CEDE.  
        Building activities.
Problem Identification and Resolution
(2)     With a survey instrument and assistance from a health physics technician,
  (6) Access controls related condition reports were reviewed since the last inspection in this area. Staff members were interviewed and documents reviewed to determine that follow-up activities are being conducted in an effective and timely manner, commensurate with their safety and risk.  
        inspectors walked down the above mentioned areas to determine: whether the
        radiation work permits (RWPs), procedures and engineering controls were in
(7) For repetitive deficiencies or significant individual deficiencies in problem identification and resolution, the inspectors determined if the licensee's assessment activities were also identifying and addressing these deficiencies.  
        place and whether surveys and postings were adequate.
(8) A review of events revealed no performance indicator occurrences that involved dose rates greater than 25 Rem/hour at 30 cm, dose rates greater than
(3)     The inspectors reviewed RWPs that provide access to exposure significant areas
26 Enclosure  500 Rem/hour at 1 meter, or unintended exposures greater than 100 mrem TEDE (or greater than 5 Rem SDE or greater than 1.5 Rem LDE)  
        of the plant including high radiation areas. Specified electronic personal
Job-in-Progress Reviews
        dosimeter alarm set points were reviewed with respect to current radiological
  (9) The inspectors observed aspects of various on-going activities to confirm that radiological controls, such as required surveys, area postings, job coverage, and job site preparations were conducted. The inspectors verified that personnel dosimetry was properly worn and that workers were knowledgeable of work area conditions. The inspectors attended pre-planning meetings for work described earlier in the report.  
        condition applicability and workers were queried to verify their understanding of
(10) Underwater diving activities associated with repairs to the fuel transport system were reviewed for adequacy. Dosimetry requirements, bioassay requirements, and controls were reviewed.  
        plant procedures governing alarm response and knowledge of radiological
High Risk Significant, High Dose Rate High Radiation Areas (HRA) and Very HRA Controls (11) Keys to locked and very HRA were reviewed for their controls and proper inventory. Accessible locked HRA were verified to be properly secured and posted during plant tours.  
        conditions in their work area.
(4)     There were no radiation work permits for airborne radioactivity areas with the
(12) The inspectors discussed with Radiation Protection supervision the adequacy of high dose rate HRA controls and procedures and verified that no programmatic or procedural changes have occurred that reduce the effectiveness and level of worker protection.  
        potential for individual worker internal exposures of >50 mrem CEDE.
(5)     There were no internal dose assessments that resulted in actual internal
        exposures greater than 50 mrem CEDE. Internal assessments were reviewed to
        determine adequacy and assurance that they were not in fact equal to or greater
        than 50 mrem CEDE.
Problem Identification and Resolution
(6)     Access controls related condition reports were reviewed since the last inspection
        in this area. Staff members were interviewed and documents reviewed to
        determine that follow-up activities are being conducted in an effective and timely
        manner, commensurate with their safety and risk.
(7)     For repetitive deficiencies or significant individual deficiencies in problem
        identification and resolution, the inspectors determined if the licensees
        assessment activities were also identifying and addressing these deficiencies.
(8)     A review of events revealed no performance indicator occurrences that involved
        dose rates greater than 25 Rem/hour at 30 cm, dose rates greater than
                                                                                    Enclosure
 
                                        26
      500 Rem/hour at 1 meter, or unintended exposures greater than 100 mrem
      TEDE (or greater than 5 Rem SDE or greater than 1.5 Rem LDE)
Job-in-Progress Reviews
(9)   The inspectors observed aspects of various on-going activities to confirm that
      radiological controls, such as required surveys, area postings, job coverage, and
      job site preparations were conducted. The inspectors verified that personnel
      dosimetry was properly worn and that workers were knowledgeable of work area
      conditions. The inspectors attended pre-planning meetings for work described
      earlier in the report.
(10)   Underwater diving activities associated with repairs to the fuel transport system
      were reviewed for adequacy. Dosimetry requirements, bioassay requirements,
      and controls were reviewed.
High Risk Significant, High Dose Rate High Radiation Areas (HRA) and Very HRA
Controls
(11)   Keys to locked and very HRA were reviewed for their controls and proper
      inventory. Accessible locked HRA were verified to be properly secured and
      posted during plant tours.
(12)   The inspectors discussed with Radiation Protection supervision the adequacy of
      high dose rate HRA controls and procedures and verified that no programmatic
      or procedural changes have occurred that reduce the effectiveness and level of
      worker protection.
Radiation Worker Performance
Radiation Worker Performance
  (13) During observation of the work activities listed above, radiation worker performance was evaluated with respect to the specific radiation protection work requirements and their knowledge of the radiological conditions in their work  
(13)   During observation of the work activities listed above, radiation worker
areas. (14) The inspectors reviewed condition reports, related to radiation worker performance to determine if an observable pattern traceable to a similar cause was evident.  
      performance was evaluated with respect to the specific radiation protection work
Radiation Protection Technician Proficiency
      requirements and their knowledge of the radiological conditions in their work
  (15) During observation of the work activities listed above, radiation protection technician work performance was evaluated with respect to their knowledge of the radiological conditions, the specific radiation protection work requirements and radiation protection procedures.  
      areas.
(16) The inspectors reviewed condition reports, related to radiation worker performance to determine if an observable pattern traceable to a similar cause was evident.  
(14)   The inspectors reviewed condition reports, related to radiation worker
      performance to determine if an observable pattern traceable to a similar cause
 
      was evident.
27 Enclosure    b. Findings
Radiation Protection Technician Proficiency
  Introduction. The inspectors identified a NCV of very low safety significance (Green) related to Technical Specification 5.4.1.a, "Procedures," because Entergy personnel did not generate condition reports or investigation paperwork for multiple high dose-rate alarms as required by station procedures. Specifically, personnel did not generate the  
(15)   During observation of the work activities listed above, radiation protection
required condition reports and adequately document the investigations for six instances of unplanned or un-briefed electronic dosimeter alarms received by individuals in the Unit 2 radiologically controlled area (RCA) that occurred between January 2009 and March 2009.  
      technician work performance was evaluated with respect to their knowledge of
      the radiological conditions, the specific radiation protection work requirements
Description. During the period January 2009 through March 2009, six instances of electronic dosimeter dose rate alarms were recorded by the access control system for Unit 2 personnel in the RCA (Unit 3 had 15 instances). During this period, Entergy personnel inconsistently utilized an informal process of reviewing the alarms without a full investigation or approval process. Moreover, in one of the six instances at Unit 2, the inspectors identified that no investigation or follow-up had occurred. In some cases, the occurrences were over two months old, which the inspectors noted would have  
      and radiation protection procedures.
made resultant investigations more challenging to perform. In other cases, the alarms were not identified until the worker attempted to re-enter the RCA and the access control system required manual override to "un-lock" the occurrence to allow entry into the RCA. The inspectors noted that the controlling Entergy procedure for this activity, EN-RP-203, "Dose Assessment," specifies that for a dose-rate alarm that is unanticipated or un-
(16)   The inspectors reviewed condition reports, related to radiation worker
briefed, several actions are required, one of which is to initiate a condition report, another is to document the investigation using an attachment in the procedure. Contrary to EN-RP-203, for these 21 instances, no condition reports or attachments were generated with a detailed investigation prior to the workers re-entering the radiologically controlled area. The highest exposure received by these workers during their entry, as  
      performance to determine if an observable pattern traceable to a similar cause
indicated by their electronic dosimeter and logged by the access control system, was 33 mRem, while most dosimeters indicated less than 1 mRem for the entry.  
      was evident.
Analysis. The inspectors determined that the failure to generate a condition report, as well as the failure to adequately investigate six unplanned or un-briefed electronic  
                                                                                  Enclosure
dosimeter alarms prior to re-entry into the Unit 2 RCA, as required by station procedure was a performance deficiency. This performance deficiency was within Entergy personnel's ability to foresee and correct, and should have been prevented. This issue was not subject to traditional enforcement, in that it did not have actual safety consequence, it was not an issue that had the potential to impact NRC's ability to  
 
perform its regulatory function, and there were no willful aspects.  
                                            27
The finding is more than minor because it is associated with the Occupational Radiation Safety cornerstone attribute of programs and process, and adversely affected its objective to ensure adequate protection of worker health and safety from exposure to radiation. Moreover, the inspectors identified a programmatic deficiency to maintain and implement programs to keep exposures as low as reasonably achievable, because  
b. Findings
multiple examples were identified regarding the failure to satisfy station radiation protection procedures. Specifically, in six cases, Entergy did not fully evaluate dose rate alarms received by workers in radiologically controlled areas of the plant. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance (Green) because it did not  
  Introduction. The inspectors identified a NCV of very low safety significance (Green)
involve: (1) as low as is reasonably achievable planning and controls, (2) an
  related to Technical Specification 5.4.1.a, Procedures, because Entergy personnel did
28 Enclosure  overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose.  
  not generate condition reports or investigation paperwork for multiple high dose-rate
The inspectors determined that the finding had a cross-cutting aspect related to procedural adherence in the Work Practices component of the Human Performance area. Specifically, Entergy employees did not follow procedures to generate condition  
  alarms as required by station procedures. Specifically, personnel did not generate the
reports and document investigations when high-dose rate alarms were received by workers. (H.4 (b) per IMC 0305)  
  required condition reports and adequately document the investigations for six instances
Enforcement. Technical Specification 5.4.1.a, "Procedures," requires that Entergy establish, implement, and maintain procedures specified in Regulatory Guide (RG) 1.33,  
  of unplanned or un-briefed electronic dosimeter alarms received by individuals in the
Revision 2, Appendix A., Section 7.e, radiation protection procedures for personnel monitoring. Entergy procedure EN-RP-203, Revision 2, Section 5.11, requires that a condition report be written for each unplanned or un-briefed electronic dosimeter dose-rate alarm. Contrary to the above, the inspectors identified through a review of electronic dosimeter log information from January 2009 through March 2009, six instances of unanticipated or un-briefed electronic dosimeter dose-rate alarms when the procedure was not implemented and condition reports were not generated. Because  
  Unit 2 radiologically controlled area (RCA) that occurred between January 2009 and
this finding was of very low safety significance and it was entered into the corrective action program as CR-IP3-2009-001253 and CR-IP3-2009-001318, this violation is being treated as an NCV, consistent with the NRC Enforcement Policy.  
  March 2009.
NCV 05000247/2009002-07, Failure to Follow Radiation Protection Procedures.
  Description. During the period January 2009 through March 2009, six instances of
  2OS2 ALARA Planning and Controls (71121.02 - 12 samples)  
  electronic dosimeter dose rate alarms were recorded by the access control system for
  a. Inspection Scope
  Unit 2 personnel in the RCA (Unit 3 had 15 instances). During this period, Entergy
  From March 23 through March 27, 2009, the inspectors conducted the following  
  personnel inconsistently utilized an informal process of reviewing the alarms without a
activities to verify that Entergy was properly maintaining individual and collective radiation exposures as low as is reasonably achievable (ALARA). Implementation of the ALARA program was reviewed by inspectors against the criteria contained in 10 CFR 20, applicable industry standards, and Entergy's procedures.  
  full investigation or approval process. Moreover, in one of the six instances at Unit 2,
  the inspectors identified that no investigation or follow-up had occurred. In some cases,
This inspection activity represents completion of twelve (12) samples relative to this inspection area.  
  the occurrences were over two months old, which the inspectors noted would have
Inspection Planning
  made resultant investigations more challenging to perform. In other cases, the alarms
  (1) The inspectors reviewed pertinent information regarding cumulative exposure history, current exposure trends, and on-going activities to assess current performance and outage exposure challenges. The inspectors determined the site's 3-year rolling collective average exposure.  
  were not identified until the worker attempted to re-enter the RCA and the access control
(2) The inspectors reviewed unit 3 outage work related activities occurring during the inspection period, the associated ALARA plans, RWPs, ALARA Committee  
  system required manual override to un-lock the occurrence to allow entry into the RCA.
Reviews, exposure estimates, actual exposures and post job reviews. Work reviewed included: Refuel Floor Split Pin and Reactor Head Inspections, Refuel Floor Lower Internals Removal and Installation, Refuel Floor and Fuel Support Building Fuel Transport Equipment Repairs requiring an underwater diver, Reactor Coolant Pump work including RCP #31 Impeller replacement,
  The inspectors noted that the controlling Entergy procedure for this activity, EN-RP-203,
29 Enclosure Containment valve work including Pressurizer Safety Valves, Various Containment and Auxiliary Building activities.
  Dose Assessment, specifies that for a dose-rate alarm that is unanticipated or un-
(3) The inspectors reviewed implementing procedures associated with maintaining occupational exposures ALARA.  This included a review of the processes used to estimate and track work activity exposures.
  briefed, several actions are required, one of which is to initiate a condition report,
Radiological Work Planning
  another is to document the investigation using an attachment in the procedure. Contrary
  (4) With respect to the work activities listed above, the inspectors reviewed dose summary reports, related post-job ALARA reviews, related RWPS, exposure
  to EN-RP-203, for these 21 instances, no condition reports or attachments were
estimates and actual exposures, and ALARA Committee meeting paperwork. Through this review, the inspector determined that dose was appropriately managed and evaluated by Station Management.
  generated with a detailed investigation prior to the workers re-entering the radiologically
(5) ALARA work activity evaluations, exposure estimates, and exposure mitigating requirements were reviewed for work packages previously mentioned.  The inspectors determined that Entergy established procedures, engineering and
  controlled area. The highest exposure received by these workers during their entry, as
work controls, based on sound radiation protection principles.
  indicated by their electronic dosimeter and logged by the access control system, was 33
(6) The inspectors compared the results achieved with the intended dose that was established in the planning of the work.  The inspectors determined the reasons for any inconsistencies between the intended and actual work activity doses and
  mRem, while most dosimeters indicated less than 1 mRem for the entry.
station management awareness and involvement.
  Analysis. The inspectors determined that the failure to generate a condition report, as
(7) The inspectors evaluated for adequacy, the interfaces between operations, radiation protection, maintenance, maintenance planning and others for interface problems or missing program elements.
  well as the failure to adequately investigate six unplanned or un-briefed electronic
Verification of Dose Estimates and Exposure Tracking Systems
  dosimeter alarms prior to re-entry into the Unit 2 RCA, as required by station procedure
  (8) Methods for adjusting exposure estimates, or re-planning work, when unexpected changes in scope or emergent work is encountered, was reviewed
  was a performance deficiency. This performance deficiency was within Entergy
by the inspectors for adequacy.
  personnels ability to foresee and correct, and should have been prevented. This issue
Job Site Inspections and ALARA Controls
  was not subject to traditional enforcement, in that it did not have actual safety
  (9) The inspectors reviewed work activities that present the highest radiological risk to workers.  The inspectors evaluated Entergy's use of engineering controls to achieve dose reductions and to verify that procedures and controls are consistent with ALARA reviews.  Associated ALARA Plans and RWPs were reviewed to determine if appropriate exposure and contamination controls were being employed.
  consequence, it was not an issue that had the potential to impact NRCs ability to
Radiation Worker Performance
  perform its regulatory function, and there were no willful aspects.
  (10) Through observations and interviews, workers and technicians were found to be knowledgeable of the work area radiological conditions and low dose waiting areas. 
  The finding is more than minor because it is associated with the Occupational Radiation
 
  Safety cornerstone attribute of programs and process, and adversely affected its
30 Enclosure  Declared Pregnant Workers
  objective to ensure adequate protection of worker health and safety from exposure to
  (11) The inspectors reviewed information associated with declared pregnant workers during the assessment period and whether appropriate monitoring and controls
  radiation. Moreover, the inspectors identified a programmatic deficiency to maintain and
were being utilized to ensure compliance with 10CFR Part 20.
  implement programs to keep exposures as low as reasonably achievable, because
Problem Identification and Resolution
  multiple examples were identified regarding the failure to satisfy station radiation
  (12) The inspectors reviewed elements of the Entergy's corrective action program related to implementing radiological controls to determine if problems are being entered into the program for timely resolution.
  protection procedures. Specifically, in six cases, Entergy did not fully evaluate dose rate
  b. Findings
  alarms received by workers in radiologically controlled areas of the plant. Using the
  No findings of significance were identified.
  Occupational Radiation Safety Significance Determination Process, the inspectors
4. OTHER ACTIVITIES [OA]
  determined that the finding was of very low safety significance (Green) because it did not
  4OA1 Performance Indicator Verification (71151 - 3 samples)
  involve: (1) as low as is reasonably achievable planning and controls, (2) an
  a. Inspection Scope
                                                                                      Enclosure
  The inspectors reviewed performance indicator data for the cornerstones listed below and used Nuclear Energy Institute 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 5, to verify individual performance indicator accuracy and completeness.  The documents reviewed during this inspection are listed in the
 
Attachment.
                                                28
Initiating Events Cornerstone
    overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to
  * Unplanned Scrams per 7000 Critical Hours (January 2008 to December 2008)
    assess dose.
* Unplanned Transients per 7000 Critical Hours (January 2008 to December 2008)
    The inspectors determined that the finding had a cross-cutting aspect related to
The inspectors reviewed data and plant records from January 2008 to December 2008.  The records included PI data summary reports, licensee event reports, operator
    procedural adherence in the Work Practices component of the Human Performance
narrative logs, Entergy's corrective action program, and Maintenance Rule records.  The inspectors verified the accuracy of the number of critical hours reported, and interviewed the system engineers and operators responsible for data collection and evaluation.
    area. Specifically, Entergy employees did not follow procedures to generate condition
Barrier Integrity Cornerstone
    reports and document investigations when high-dose rate alarms were received by
  * RCS Activity (January 2008 to December 2008)
    workers. (H.4 (b) per IMC 0305)
The inspectors reviewed data and plant records from January 2008 to December 2008.  The records included performance indicator data summary reports, licensee event reports, operator narrative logs, Entergy's corrective action program, and Maintenance Rule records.  The inspectors verified the accuracy of the number of critical hours reported, and interviewed the system engineers and operators responsible for data collection and evaluation.
    Enforcement. Technical Specification 5.4.1.a, Procedures, requires that Entergy
    establish, implement, and maintain procedures specified in Regulatory Guide (RG) 1.33,
    Revision 2, Appendix A., Section 7.e, radiation protection procedures for personnel
    monitoring. Entergy procedure EN-RP-203, Revision 2, Section 5.11, requires that a
    condition report be written for each unplanned or un-briefed electronic dosimeter dose-
    rate alarm. Contrary to the above, the inspectors identified through a review of
    electronic dosimeter log information from January 2009 through March 2009, six
    instances of unanticipated or un-briefed electronic dosimeter dose-rate alarms when the
    procedure was not implemented and condition reports were not generated. Because
    this finding was of very low safety significance and it was entered into the corrective
    action program as CR-IP3-2009-001253 and CR-IP3-2009-001318, this violation is
    being treated as an NCV, consistent with the NRC Enforcement Policy. NCV
    05000247/2009002-07, Failure to Follow Radiation Protection Procedures.
2OS2 ALARA Planning and Controls (71121.02 - 12 samples)
a. Inspection Scope
    From March 23 through March 27, 2009, the inspectors conducted the following
    activities to verify that Entergy was properly maintaining individual and collective
    radiation exposures as low as is reasonably achievable (ALARA). Implementation of the
    ALARA program was reviewed by inspectors against the criteria contained in 10 CFR
    20, applicable industry standards, and Entergys procedures.
    This inspection activity represents completion of twelve (12) samples relative to this
    inspection area.
    Inspection Planning
    (1)     The inspectors reviewed pertinent information regarding cumulative exposure
              history, current exposure trends, and on-going activities to assess current
              performance and outage exposure challenges. The inspectors determined the
              sites 3-year rolling collective average exposure.
    (2)     The inspectors reviewed unit 3 outage work related activities occurring during the
              inspection period, the associated ALARA plans, RWPs, ALARA Committee
              Reviews, exposure estimates, actual exposures and post job reviews. Work
              reviewed included: Refuel Floor Split Pin and Reactor Head Inspections, Refuel
              Floor Lower Internals Removal and Installation, Refuel Floor and Fuel Support
              Building Fuel Transport Equipment Repairs requiring an underwater diver,
              Reactor Coolant Pump work including RCP #31 Impeller replacement,
                                                                                        Enclosure


    
                                          29
31 Enclosure    b. Findings
        Containment valve work including Pressurizer Safety Valves, Various
  No findings of significance were identified.         4OA2 Identification and Resolution of Problems (71152)
        Containment and Auxiliary Building activities.
.1 Routine Problem Identification & Resolution Program Review
(3)    The inspectors reviewed implementing procedures associated with maintaining
      a. Inspection Scope
        occupational exposures ALARA. This included a review of the processes used to
  As required by Inspection Procedure 71152, "Identification and Resolution of Problems,"
        estimate and track work activity exposures.
and to identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of all items entered into Entergy's corrective action program. The review was accomplished by accessing Entergy's computerized database for condition reports, and attending condition report screening meetings.  
Radiological Work Planning
In accordance with the baseline inspection modules, the inspectors selected corrective  
(4)    With respect to the work activities listed above, the inspectors reviewed dose
action program items across the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones for further follow-up and review. The inspectors assessed Entergy's threshold for problem identification, adequacy of the causal analysis, extent of condition reviews, and operability determinations, and timeliness of the associated corrective actions. The condition reports reviewed during this inspection are listed in the  
        summary reports, related post-job ALARA reviews, related RWPS, exposure
Attachment.  
        estimates and actual exposures, and ALARA Committee meeting paperwork.
  b. Findings
        Through this review, the inspector determined that dose was appropriately
  No findings of significance were identified
        managed and evaluated by Station Management.
4OA3 Event Followup
(5)    ALARA work activity evaluations, exposure estimates, and exposure mitigating
  .1 Condensate Return Line Leak on February 15, 2009
        requirements were reviewed for work packages previously mentioned. The
  a. Inspection Scope
        inspectors determined that Entergy established procedures, engineering and
  On February 15, 2009, an operator observed indications of wetness in a pipe sleeve in the floor of the auxiliary feed pump building. The operator notified the control room. Chemistry samples of the water were drawn and analyzed. On February 16, Entergy  
        work controls, based on sound radiation protection principles.
determined the chemistry results indicated the water was from the condensate storage tank (CST) return line. The inspectors reviewed the technical specifications (TS) to determine whether operators entered the applicable TS action statements for the CST and completed required actions to administratively determine the back-up on-site city water tank was available, if needed, to provide water to the auxiliary feedwater pumps. The inspectors reviewed Entergy's operability evaluation of the CST to determine whether it was technically supported. In addition, the inspectors reviewed the impact of  
(6)    The inspectors compared the results achieved with the intended dose that was
the leak on the auxiliary feed water system which utilizes the CST as a primary source of water and circulates water back to the CST through the CST return piping. The inspectors also reviewed chemistry and radiological samples taken of the water to assess the environmental impact of the leak and determine if the release was below NRC regulatory limits for liquid effluents.  
        established in the planning of the work. The inspectors determined the reasons
 
        for any inconsistencies between the intended and actual work activity doses and
32 Enclosure b. Findings and Observations
        station management awareness and involvement.
  No findings of significance were identified.  
(7)    The inspectors evaluated for adequacy, the interfaces between operations,
Entergy excavated a portion of the CST piping in the area of the identified leakage and determined that the CST return pipe was leaking due to a hole the pipe where a small  
        radiation protection, maintenance, maintenance planning and others for interface
area of a protective coating was missing. Entergy also identified two additional areas of piping with metal loss that did not exceed ASME Code minimum required wall thickness. However, the areas were repaired while the opportunity existed. Entergy removed the portion of pipe with the localized defects and sent the specimen to a laboratory for analysis to identify the causes. The inspectors determined that the actions Entergy  
        problems or missing program elements.
implemented to evaluate and repair the leaking CST pipe to restore operability to the CST were adequate
Verification of Dose Estimates and Exposure Tracking Systems
and in accordance with their operating license. Additionally, the inspectors determined that the evaluations and actions Entergy performed to evaluate and maintain operability of the auxiliary feed pumps were adequate. Entergy analyzed the water leaking up through the sleeve and determined it was CST water based on hydrazine and tritium levels. The amount of tritium detected in the water was consistent with that found in the CST, for example, analyses of samples of water from the leak  
(8)    Methods for adjusting exposure estimates, or re-planning work, when
returned 2000 - 2300 picocuries per liter (pCi/l). The release was determined to be below the NRC regulatory limits for liquid effluents. For added perspective, while not drinking water, the Environmental Protection Agency environmental limit for drinking water requires tritium levels less than 20,000 pCi/l.  
        unexpected changes in scope or emergent work is encountered, was reviewed
        by the inspectors for adequacy.
Entergy initiated a root cause analysis to determine causes of the leak that is scheduled to be completed in May 2009. At the end of the inspection period, the inspectors were monitoring the performance of Entergy in implementing its corrective action program to address the issue and develop a root cause evaluation and further corrective actions.  
Job Site Inspections and ALARA Controls
(9)    The inspectors reviewed work activities that present the highest radiological risk
        to workers. The inspectors evaluated Entergys use of engineering controls to
        achieve dose reductions and to verify that procedures and controls are consistent
        with ALARA reviews. Associated ALARA Plans and RWPs were reviewed to
        determine if appropriate exposure and contamination controls were being
        employed.
Radiation Worker Performance
(10)   Through observations and interviews, workers and technicians were found to be
        knowledgeable of the work area radiological conditions and low dose waiting
        areas.
                                                                                Enclosure
 
                                              30
    Declared Pregnant Workers
    (11)    The inspectors reviewed information associated with declared pregnant workers
              during the assessment period and whether appropriate monitoring and controls
              were being utilized to ensure compliance with 10CFR Part 20.
    Problem Identification and Resolution
    (12)    The inspectors reviewed elements of the Entergys corrective action program
              related to implementing radiological controls to determine if problems are being
              entered into the program for timely resolution.
b.  Findings
    No findings of significance were identified.
4.  OTHER ACTIVITIES [OA]
4OA1 Performance Indicator Verification (71151 - 3 samples)
a.  Inspection Scope
    The inspectors reviewed performance indicator data for the cornerstones listed below
    and used Nuclear Energy Institute 99-02, Regulatory Assessment Performance
    Indicator Guideline, Revision 5, to verify individual performance indicator accuracy and
    completeness. The documents reviewed during this inspection are listed in the
    Attachment.
    Initiating Events Cornerstone
    *    Unplanned Scrams per 7000 Critical Hours (January 2008 to December 2008)
    *    Unplanned Transients per 7000 Critical Hours (January 2008 to December 2008)
    The inspectors reviewed data and plant records from January 2008 to December 2008.
    The records included PI data summary reports, licensee event reports, operator
    narrative logs, Entergys corrective action program, and Maintenance Rule records. The
    inspectors verified the accuracy of the number of critical hours reported, and interviewed
    the system engineers and operators responsible for data collection and evaluation.
    Barrier Integrity Cornerstone
    *   RCS Activity (January 2008 to December 2008)
    The inspectors reviewed data and plant records from January 2008 to December 2008.
    The records included performance indicator data summary reports, licensee event
    reports, operator narrative logs, Entergys corrective action program, and Maintenance
    Rule records. The inspectors verified the accuracy of the number of critical hours
    reported, and interviewed the system engineers and operators responsible for data
    collection and evaluation.
                                                                                      Enclosure
 
                                                31
  b. Findings
      No findings of significance were identified.
4OA2 Identification and Resolution of Problems (71152)
.1   Routine Problem Identification & Resolution Program Review
  a. Inspection Scope
      As required by Inspection Procedure 71152, Identification and Resolution of Problems,
      and to identify repetitive equipment failures or specific human performance issues for
      follow-up, the inspectors performed a daily screening of all items entered into Entergys
      corrective action program. The review was accomplished by accessing Entergys
      computerized database for condition reports, and attending condition report screening
      meetings.
      In accordance with the baseline inspection modules, the inspectors selected corrective
      action program items across the Initiating Events, Mitigating Systems, and Barrier
      Integrity cornerstones for further follow-up and review. The inspectors assessed
      Entergys threshold for problem identification, adequacy of the causal analysis, extent of
      condition reviews, and operability determinations, and timeliness of the associated
      corrective actions. The condition reports reviewed during this inspection are listed in the
      Attachment.
  b. Findings
      No findings of significance were identified
4OA3 Event Followup
.1   Condensate Return Line Leak on February 15, 2009
a.   Inspection Scope
    On February 15, 2009, an operator observed indications of wetness in a pipe sleeve in
    the floor of the auxiliary feed pump building. The operator notified the control room.
    Chemistry samples of the water were drawn and analyzed. On February 16, Entergy
    determined the chemistry results indicated the water was from the condensate storage
    tank (CST) return line. The inspectors reviewed the technical specifications (TS) to
    determine whether operators entered the applicable TS action statements for the CST
    and completed required actions to administratively determine the back-up on-site city
    water tank was available, if needed, to provide water to the auxiliary feedwater pumps.
    The inspectors reviewed Entergys operability evaluation of the CST to determine
    whether it was technically supported. In addition, the inspectors reviewed the impact of
    the leak on the auxiliary feed water system which utilizes the CST as a primary source of
    water and circulates water back to the CST through the CST return piping. The
    inspectors also reviewed chemistry and radiological samples taken of the water to assess
    the environmental impact of the leak and determine if the release was below NRC
    regulatory limits for liquid effluents.
                                                                                        Enclosure
 
                                              32
  b. Findings and Observations
    No findings of significance were identified.
      Entergy excavated a portion of the CST piping in the area of the identified leakage and
      determined that the CST return pipe was leaking due to a hole the pipe where a small
      area of a protective coating was missing. Entergy also identified two additional areas of
      piping with metal loss that did not exceed ASME Code minimum required wall thickness.
      However, the areas were repaired while the opportunity existed. Entergy removed the
      portion of pipe with the localized defects and sent the specimen to a laboratory for
      analysis to identify the causes. The inspectors determined that the actions Entergy
      implemented to evaluate and repair the leaking CST pipe to restore operability to the
      CST were adequate and in accordance with their operating license. Additionally, the
      inspectors determined that the evaluations and actions Entergy performed to evaluate
      and maintain operability of the auxiliary feed pumps were adequate. Entergy analyzed
      the water leaking up through the sleeve and determined it was CST water based on
      hydrazine and tritium levels. The amount of tritium detected in the water was consistent
      with that found in the CST, for example, analyses of samples of water from the leak
      returned 2000 - 2300 picocuries per liter (pCi/l). The release was determined to be
      below the NRC regulatory limits for liquid effluents. For added perspective, while not
      drinking water, the Environmental Protection Agency environmental limit for drinking
      water requires tritium levels less than 20,000 pCi/l.
      Entergy initiated a root cause analysis to determine causes of the leak that is scheduled
      to be completed in May 2009. At the end of the inspection period, the inspectors were
      monitoring the performance of Entergy in implementing its corrective action program to
      address the issue and develop a root cause evaluation and further corrective actions.
4OA5 Other Activities
4OA5 Other Activities
  .1 Continued Groundwater Sampling Effort to Monitor Tritium (Deviation Memorandum Inspection)
.1   Continued Groundwater Sampling Effort to Monitor Tritium (Deviation Memorandum
    Inspection)
a.  Inspection Scope
    During the week of March 23-27, 2009, the inspectors met with Entergy representatives
    to review the results of recent groundwater samples, as well as those taken and
    analyzed in 2008. The review was conducted against criteria contained in 10CFR20,
    10CFR50, and applicable industry standards.
    The review of the data included a comparison of Entergys data with split samples taken
    by the NRC of monitoring wells MW-66 and MW-67, as well as the LaFarge sample
    point. In all, 47 samples were analyzed and compared from January 2008 through
    January 2009. Isotopic analyses were performed and compared at each of the sample
    points for: Tritium, Strontium 90, Nickel 63, and gamma emitters such as Cobalt-60 and
    Cesium-137. Results of the NRC samples can be found in ADAMS accession numbers:
    ML081420676, ML082690244, ML082690202, ML082690237, ML082730830,
    ML082730810, ML090400523, ML090400516, ML090400502, ML090923932,
    ML090920949.
                                                                                      Enclosure
 
                                              33
    Entergy=s evaluation of recent groundwater results are documented in condition reports:
    CR-IP2-2009-00883, CR-IP2-2009-01110, CR-IP2-2009-01111, CR-IP2-2009-01113,
    and CR-IP2-2009-01114.
  b. Findings
    No findings of significance were identified.
    The inspectors concluded that overall, there was agreement between Entergy
    personnels results and those independently analyzed by the NRC, and that actions
    taken by Entergy have been appropriate. The inspectors also noted that conservative
    estimates indicate that the samples represent a very small fraction of the permissible
    public dose limits and are negligible with respect to natural background radiation levels.
.2  Quarterly Resident Inspector Observations of Security Personnel and Activities
   a. Inspection Scope
   a. Inspection Scope
  During the week of March 23-27, 2009, the inspectors met with Entergy representatives
    During the inspection period, the inspectors conducted observations of security force
to review the results of recent groundwater samples, as well as those taken and analyzed in 2008.  The review was conducted against criteria contained in 10CFR20,
    personnel and activities to ensure that these activities were consistent with Entergy
10CFR50, and applicable industry standards.
    security procedures and applicable regulatory requirements. Although these
The review of the data included a comparison of Entergy's data with split samples taken by the NRC of monitoring wells MW-66 and MW-67, as well as the LaFarge sample point.  In all, 47 samples were analyzed and compared from January 2008 through January 2009.  Isotopic analyses were performed and compared at each of the sample points for:  Tritium, Strontium 90, Nickel 63, and gamma emitters such as Cobalt-60 and
    observations did not constitute additional inspection samples, the inspections were
Cesium-137.  Results of the NRC samples can be found in ADAMS accession numbers:  ML081420676, ML082690244, ML082690202, ML082690237, ML082730830, ML082730810, ML090400523, ML090400516, ML090400502, ML090923932, ML090920949.
    considered an integral part of the normal, resident inspector plant status reviews during
 
    implementation of the baseline inspection program.
33 Enclosure  Entergy=s evaluation of recent groundwater results are documented in condition reports:  CR-IP2-2009-00883, CR-IP2-2009-01110, CR-IP2-2009-01111, CR-IP2-2009-01113, and CR-IP2-2009-01114.
  b. Findings
  No findings of significance were identified.
The inspectors concluded that overall, there was agreement between Entergy personnel's results and those independently analyzed by the NRC, and that actions
taken by Entergy have been appropriate.  The inspectors also noted that conservative estimates indicate that the samples represent a very small fraction of the permissible public dose limits and are negligible with respect to natural background radiation levels.
.2 Quarterly Resident Inspector Observations of Security Personnel and Activities
    a. Inspection Scope
  During the inspection period, the inspectors conducted observations of security force personnel and activities to ensure that these activities were consistent with Entergy  
security procedures and applicable regulatory requirements. Although these observations did not constitute additional inspection samples, the inspections were considered an integral part of the normal, resident inspector plant status reviews during implementation of the baseline inspection program.  
   b. Findings
   b. Findings
  No findings of significance were identified.  
    No findings of significance were identified.
4OA6 Meetings
4OA6 Meetings
  Exit Meeting Summary  
    Exit Meeting Summary
  On April 15, 2009, the inspectors presented the inspection results to Joe Pollock and other Entergy staff members, who acknowledged the inspection results presented. Entergy did not identify any material as proprietary.  
    On April 15, 2009, the inspectors presented the inspection results to Joe Pollock and
    other Entergy staff members, who acknowledged the inspection results presented.
ATTACHMENT: SUPPLEMENTAL INFORMATION 
    Entergy did not identify any material as proprietary.
A-1  Attachment
ATTACHMENT: SUPPLEMENTAL INFORMATION
  SUPPLEMENTAL INFORMATION  
                                                                                      Enclosure
KEY POINTS OF CONTACT
Entergy Personnel
  J. Pollock,  Site Vice President A. Vitale,  General Manager, Plant Operations P. Conroy,  Director of Nuclear Safety Assurance
A. Williams,  Site Operations Manager B. Sullivan,  Emergency Planning Manager S. Verrochi,  System Engineering Manager R. Walpole,  Licensing Manager D. Loope, Manager, Radiation Protection
  LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
  05000247/2009002-01                    FIN Failure to Identify Open Louvers in 11 Fire Pump House (Section 1R01)
05000247/2009002-02                  NCV Failure to Identify Damaged Components in EDG Ventilation Motor Control Center #2 (Section 1R05)
05000247/2009002-03                  NCV Failure to identify and Promptly Correct Degraded 480 Volt Switchgear Room Fire Door (Section 1R05)
05000247/2009002-04                  NCV Inadequate Maintenance Procedure for EDG Ventilation Motor Control Center #2 (Section 1R12)
05000247/2009002-05                  NCV Failure to Include RWST Level Maintenance In Online Risk Assessment (Section 1R13)
05000247/2009002-06                  NCV Inadequate Test Acceptance Criteria for Auxiliary Component Cooling Check Valves (Section 1R22)
05000247/2009002-07  NCV  Failure to Follow Radiation Protection Procedures (Section 2OS1)
 
A-2  Attachment 
LIST OF DOCUMENTS REVIEWED
Section 1R01:  Adverse Weather Protection
  Procedures
OAP-048, Rev. 4, Seasonal Weather Preparation OAP-008, Rev. 5, Severe Weather Preparations 2-AOP-SSD-1, Rev. 13, Control Room Inaccessibility Safe Shutdown Control OAP-017, Rev. 5, Plant Surveillance and Operator Rounds EN-OP-115, Rev. 5, Conduct of Operations


  Condition Reports
                                            A-1
  IP2-2009-00197 IP2-2009-00207 IP2-2009-00208 IP2-2009-00211 IP2-2009-00212 IP2-2009-00214 IP2-2009-00215 IP2-2009-00226
                            SUPPLEMENTAL INFORMATION
Orders 00152922 00153082 00153083 00179583
                                KEY POINTS OF CONTACT
Entergy Personnel
J. Pollock, Site Vice President
A. Vitale,  General Manager, Plant Operations
P. Conroy,  Director of Nuclear Safety Assurance
A. Williams, Site Operations Manager
B. Sullivan, Emergency Planning Manager
S. Verrochi, System Engineering Manager
R. Walpole, Licensing Manager
D. Loope,    Manager, Radiation Protection
                  LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
05000247/2009002-01              FIN            Failure to Identify Open Louvers in 11 Fire
                                                Pump House (Section 1R01)
05000247/2009002-02              NCV            Failure to Identify Damaged Components in
                                                EDG Ventilation Motor Control Center #2
                                                (Section 1R05)
05000247/2009002-03              NCV            Failure to identify and Promptly Correct
                                                Degraded 480 Volt Switchgear Room Fire
                                                Door (Section 1R05)
05000247/2009002-04              NCV            Inadequate Maintenance Procedure for
                                                EDG Ventilation Motor Control Center #2
                                                (Section 1R12)
05000247/2009002-05              NCV            Failure to Include RWST Level
                                                Maintenance In Online Risk Assessment
                                                (Section 1R13)
05000247/2009002-06              NCV            Inadequate Test Acceptance Criteria for
                                                Auxiliary Component Cooling Check Valves
                                                (Section 1R22)
05000247/2009002-07              NCV            Failure to Follow Radiation Protection
                                                Procedures (Section 2OS1)
                                                                                  Attachment


                                              A-2
Section 1R04: Equipment Alignment
                              LIST OF DOCUMENTS REVIEWED
 
Section 1R01: Adverse Weather Protection
Procedures
OAP-048, Rev. 4, Seasonal Weather Preparation
OAP-008, Rev. 5, Severe Weather Preparations
2-AOP-SSD-1, Rev. 13, Control Room Inaccessibility Safe Shutdown Control
OAP-017, Rev. 5, Plant Surveillance and Operator Rounds
EN-OP-115, Rev. 5, Conduct of Operations
Condition Reports
IP2-2009-00197        IP2-2009-00207        IP2-2009-00208        IP2-2009-00211
IP2-2009-00212        IP2-2009-00214        IP2-2009-00215        IP2-2009-00226
Orders
00152922      00153082      00153083      00179583
Section 1R04: Equipment Alignment
Procedures
2-PT-M103, Rev. 2, Auxiliary Feedwater System Monthly Alignment Verification
2-COL-4.1.1, Rev. 22, Component Cooling System
Section 1R05: Fire Protection
Procedures
SAO-703, Rev. 25, Fire Protection Impairment Criteria and Surveillance
EN-DC-161, Rev. 2, Control of Combustibles
OAP-037, Rev. 2, Operations Electrical Equipment Operating Guidelines
IP-SMM-IS-103, Rev. 0, IPEC Site Management Manual Electrical Safety
2-PT-SA020, Rev. 0, Swing Fire Doors
Condition Reports
IP2-2009-00904        IP2-2009-00526        IP2-2009-00680        IP2-2009-00709
IP2-2009-00834        IP2-2009-00342        IP2-2009-00483        IP2-2004-05336
IP2-2007-03561        IP2-2007-04645        IP2-2008-05447
Orders
51645822      51676572
Miscellaneous
Indian Point Nuclear Generating Station, Unit 2, Fire Protection Program Plan, Rev. 9
Indian Point Pre-Fire Plans Unit 2 - Nuclear
IP2-RPT-03-00015, IP2 Fire Hazards Analysis, Rev. 3
1R07: Heat Sink Performance
Procedures
Procedures
2-PT-M103, Rev. 2, Auxiliary Feedwater System Monthly Alignment Verification
SEP-SW-001, NRC Generic Letter 89-13 Service Water Program
2-COL-4.1.1, Rev. 22, Component Cooling System
PT-2Y10B, 22 CCW HX Test
Section 1R05:  Fire Protection
2-HTX-004-CCW, Component Cooling Water Heat Exchanger Maintenance
  Procedures
                                                                                    Attachment
SAO-703, Rev. 25, Fire Protection Impairment Criteria and Surveillance EN-DC-161, Rev. 2, Control of Combustibles OAP-037, Rev. 2, Operations Electrical Equipment Operating Guidelines IP-SMM-IS-103, Rev. 0, IPEC Site Management Manual Electrical Safety 2-PT-SA020, Rev. 0, Swing Fire Doors


Condition Reports
                                                A-3
IP2-2009-00904 IP2-2009-00526 IP2-2009-00680 IP2-2009-00709  IP2-2009-00834 IP2-2009-00342 IP2-2009-00483 IP2-2004-05336 IP2-2007-03561 IP2-2007-04645 IP2-2008-05447
Work Orders
51675733
Condition Reports
IP2-2005-0673        IP2-2005-0768          IP2-2005-1268      IP2-2006-7126
IP2-2006-3974
Miscellaneous
EPRI NP-7552, Heat Exchanger Performance Monitoring Guidelines
Preliminary Report of Eddy Current Testing dated 2/10/09
21 CCW Hx Inspection Reports dated 2/23/2005 and 1/8/2007
22 CCW Hx Inspection Reports dated 2/23/2005 and 12/12/2006
Section 1R11: Licensed Operator Requalification Program
Procedures
OAP-033, Conduct of Operations Simulator Training, Evaluations, and Debriefs, Rev. 4
OAP-032, Operations Training Program, Rev. 9
2-E-0, Rev. 0, Reactor Trip or Safety Injection
2-ECA-0.0, Rev. 3, Loss of All AC Power
2-AOP-480V-1, Rev. 5, Loss of Normal Power to any 480V Bus
Miscellaneous
LRQ-SES-21, Rev. 0, IPEC Evalauted Scenario for Loss of All AC Power
Section 1R12: Maintenance Effectiveness
Procedures
2-MCC-003-ELC, Rev 0, Klockner-Moeller, Series 200, 480 Volt Motor Control Center
    Preventive Maintenance
2-PC-Q2, Rev. 19, Refueling Water Storage Tank Level
0-MS-412, Rev. 0, Inspection and Cleaning of Bus Bars, Contacts, Ground Connections, Wiring
    and Insulators
IP-SMM-IS-103, Rev. 0, IPEC Site Management Manual Electrical Safety
0-GNR-404-ELC, Rev. 1, Emergency Diesel Generator 2-Year Inspection
2-GNR-015-ELC, Rev. 2, Emergency Diesel Generator Preventive Maintenance 2-Year
2-PT-M021B, Rev. 17, Emergency Diesel Generator 22 Load Test
Condition Reports
IP2-2009-00527        IP2-2009-00532          IP2-2009-01041    IP2-2003-00948
IP2-2009-00342       IP2-2009-00483         IP2-2004-03106    IP2-2007-01893
IP2-2008-05382        IP2-2009-00486          IP2-2009-00041    IP2-2009-00178
IP2-2006-04101        IP2-2009-00093          IP2-2007-03476    IP2-2007-04921
IP2-2008-00454        IP2-2008-00907          IP2-2008-03976
Orders
51557262      51676147      06-16146        51696697    51322921    51268313
00181009      00167536      04-26645        57696714    51649505    51654261
00118733      07-03476      07-04921        08-00454    08-00907    09-00532
Drawing
309030-02, Loop diagram RWST level indication
3WS-463-610-14-20101-3, Schematic for EDG HVAC Heater
                                                                                  Attachment


Orders 51645822 51676572
                                              A-4
Miscellaneous
IP2-S-000231-04, Schematic for EDG Building Ventilation Distribution
Indian Point Nuclear Generating Station, Unit 2, Fire Protection Program Plan, Rev. 9 Indian Point Pre-Fire Plans Unit 2 - Nuclear
B248513-12, 480V MCC 26C and CCR Ventilation Distribution
IP2-RPT-03-00015, "IP2 Fire Hazards Analysis," Rev. 3
B228434-02, Class A Boundary for Electrical Systems
1R07: Heat Sink Performance
Miscellaneous
  Procedures
Maintenance Rule Basis Document Residual Heat Removal System, dated 5/23/05
SEP-SW-001, NRC Generic Letter 89-13 Service Water Program PT-2Y10B, 22 CCW HX Test 2-HTX-004-CCW, Component Cooling Water Heat Exchanger Maintenance 
Maintenance Rule Basis Document HVAC Emergency Diesel Building, dated 5/23/05
A-3  Attachment  Work Orders
IP-SMM-AD-102, Att 10.2, dated 4/6/08, for revision to procedure 2-MCC-003-ELC
51675733  Condition Reports
Vendor Manual, Klockner-Moeller Series 200 Motor Control Center
IP2-2005-0673 IP2-2005-0768 IP2-2005-1268 IP2-2006-7126  IP2-2006-3974
Vendor Manual, Qmark MUH Series Modular Unit Heaters
 
Vendor Manual, ALCO Fuel Injection Nozzle and Holder
Miscellaneous
Maintenance Rule Expert Panel Meeting Minutes dated 2/14/05
EPRI NP-7552, Heat Exchanger Performance Monitoring Guidelines Preliminary Report of Eddy Current Testing dated 2/10/09 21 CCW Hx Inspection Reports dated 2/23/2005 and 1/8/2007
Tagout 2-480V-Panel-MCC26C dated 4/3/08
22 CCW Hx Inspection Reports dated 2/23/2005 and 12/12/2006
DRN-08-01336 dated 4/6/08 for procedure 2-MCC-003-ELC
Section 1R11: Licensed Operator Requalification Program
PMCR ER-06-33534, to establish maintenance activity for EDG HVAC MCC
 
Section 1R13: Maintenance Risk Assessments and Emergent Work Control
Procedures
Procedures
OAP-033, "Conduct of Operations Simulator Training, Evaluations, and Debriefs," Rev. 4 OAP-032, "Operations Training Program," Rev. 9
IP-SMM-WM-101, On-Line Risk Assessment
2-E-0, Rev. 0, Reactor Trip or Safety Injection 2-ECA-0.0, Rev. 3, Loss of All AC Power 2-AOP-480V-1, Rev. 5, Loss of Normal Power to any 480V Bus
2-PC-Q109, Recalibration of Nis and OT/OP delta T parameters
Miscellaneous
PT-Q17A, Verify ASSS supply to 21 AFP
LRQ-SES-21, Rev. 0, IPEC Evalauted Scenario for Loss of All AC Power
2-PT-Q027A, 21 Auxiliary Feed Pump
Section 1R12:  Maintenance Effectiveness
2-PC-Q2, Rev. 19, Refueling Water Storage Tank Level
  Procedures
2-ES-1.3, Rev. 2, Transfer to Cold Leg Recirculation
2-MCC-003-ELC, Rev 0, Klockner-Moeller, Series 200, 480 Volt Motor Control Center Preventive Maintenance 2-PC-Q2, Rev. 19, Refueling Water Storage Tank Level 0-MS-412, Rev. 0, Inspection and Cleaning of Bus Bars, Contacts, Ground Connections, Wiring and Insulators IP-SMM-IS-103, Rev. 0, IPEC Site Management Manual Electrical Safety 0-GNR-404-ELC, Rev. 1, Emergency Diesel Generator 2-Year Inspection 2-GNR-015-ELC, Rev. 2, Emergency Diesel Generator Preventive Maintenance 2-Year 2-PT-M021B, Rev. 17, Emergency Diesel Generator 22 Load Test
Condition Reports
Condition Reports
IP2-2009-00527 IP2-2009-00532 IP2-2009-01041 IP2-2003-00948 IP2-2009-00342 IP2-2009-00483 IP2-2004-03106 IP2-2007-01893  IP2-2008-05382 IP2-2009-00486 IP2-2009-00041 IP2-2009-00178 IP2-2006-04101 IP2-2009-00093 IP2-2007-03476 IP2-2007-04921 IP2-2008-00454 IP2-2008-00907 IP2-2008-03976
IP2-2009-00018       IP2-2009-00027       IP2-2009-00139       IP2-2009-00143
IP2-2009-00148       IP2-2009-00389
Orders 51557262 51676147 06-16146 51696697 51322921 51268313 00181009 00167536 04-26645 57696714 51649505 51654261 00118733 07-03476 07-04921 08-00454 08-00907 09-00532
Work Orders
00165604     51654961       51692571     51692351       51696697
Drawing 309030-02, Loop diagram RWST level indication 3WS-463-610-14-20101-3, Schematic for EDG HVAC Heater 
A-4  Attachment  IP2-S-000231-04, Schematic for EDG Building Ventilation Distribution B248513-12, 480V MCC 26C and CCR Ventilation Distribution B228434-02, Class "A" Boundary for Electrical Systems
Miscellaneous
Maintenance Rule Basis Document Residual Heat Removal System, dated 5/23/05
Maintenance Rule Basis Document HVAC Emergency Diesel Building, dated 5/23/05 IP-SMM-AD-102, Att 10.2, dated 4/6/08, for revision to procedure 2-MCC-003-ELC Vendor Manual, Klockner-Moeller Series 200 Motor Control Center Vendor Manual, Qmark MUH Series Modular Unit Heaters Vendor Manual, ALCO Fuel Injection Nozzle and Holder 
Maintenance Rule Expert Panel Meeting Minutes dated 2/14/05 Tagout 2-480V-Panel-MCC26C dated 4/3/08 DRN-08-01336 dated 4/6/08 for procedure 2-MCC-003-ELC PMCR ER-06-33534, to establish maintenance activity for EDG HVAC MCC
Section 1R13:  Maintenance Risk Assessments and Emergent Work Control
  Procedures
IP-SMM-WM-101, On-Line Risk Assessment 2-PC-Q109, Recalibration of Nis and OT/OP delta T parameters PT-Q17A, Verify ASSS supply to 21 AFP 2-PT-Q027A, 21 Auxiliary Feed Pump
2-PC-Q2, Rev. 19, Refueling Water Storage Tank Level 2-ES-1.3, Rev. 2, Transfer to Cold Leg Recirculation
Condition Reports
IP2-2009-00018 IP2-2009-00027 IP2-2009-00139 IP2-2009-00143  
IP2-2009-00148 IP2-2009-00389  
Work Orders
00165604 51654961 51692571 51692351 51696697  
Miscellaneous
Miscellaneous
Equipment Out-Of-Service (EOOS) risk assessment reports  
Equipment Out-Of-Service (EOOS) risk assessment reports
Section 1R15: Operability Evaluations
Section 1R15: Operability Evaluations
Procedures
  Procedures
2-PT-Q031A, 21 Auxiliary Component Cooling Pump
2-PT-Q031A, 21 Auxiliary Component Cooling Pump 2-PT-Q031B, 22 Auxiliary Component Cooling Pump EN-MA-133, Control of Scaffolding 2-AOP-IB-1, Loss of Power to an Instrument Bus 2-PT-M021B, Rev. 17, Emergency Diesel Generator 22 Load Test 2-SOP-AFW-002, Rev. 1, Auxiliary Feedwater System Operation Support Procedure  
2-PT-Q031B, 22 Auxiliary Component Cooling Pump
EN-MA-133, Control of Scaffolding
2-AOP-IB-1, Loss of Power to an Instrument Bus
2-PT-M021B, Rev. 17, Emergency Diesel Generator 22 Load Test
2-SOP-AFW-002, Rev. 1, Auxiliary Feedwater System Operation Support Procedure
Drawings
A249955-21, 480V AC MCC 29 & 29A
Calculation
IP3-CALC-FW-01482, Rev. 0, Feedwater Stratification and Auxiliary Feedwater
                                                                                Attachment


Drawings A249955-21, 480V AC MCC 29 & 29A
                                              A-5
Calculation
Condition Reports
IP3-CALC-FW-01482, Rev. 0, Feedwater Stratification and Auxiliary Feedwater
IP2-2009-0500           IP2-2009-0505       IP2-2008-3749       IP2-2009-0547
 
IP2-2009-0567           IP2-2009-0509       IP2-2005-0252       IP2-2009-0552
A-5 Attachment  Condition Reports
IP2-2009-0655           IP2-2008-2705       IP2-2009-0041       IP2-2009-0093
IP2-2009-0500 IP2-2009-0505 IP2-2008-3749 IP2-2009-0547 IP2-2009-0567 IP2-2009-0509 IP2-2005-0252 IP2-2009-0552 IP2-2009-0655 IP2-2008-2705 IP2-2009-0041 IP2-2009-0093  
Work Orders
NP-99-07694
Miscellaneous
WCAP-12312, Rev. 2, Safety Evaluation for an Ultimate Heat Sink Temperature Increase to 95F at Indian Point Unit 2 Heat exchanger data sheet for containment recirculation pump number 22 motor cooler WCAP-7829, Fan Cooler Motor Unit Test Environmental Qualification Report for Containment Recirculation Pump Motors IP2-CCW-DBD, Component Cooling Water design bases document IP2-DBD-207, Design Basis Document for 118V AC Electrical System AMSE OM-2001 Edition Unit 2 active scaffold list
VM 1073-1.2, Vendor manual for auxiliary component cooling pumps VM 1100, vendor manual for 118V AC solid state static inverters Work order NP-89-43777, replacement of 22 ACCP impeller IP2-AFW-DBD, Rev. 1, AFW Design Basis Document
Section 1R18:  Plant Modifications
  Procedures
2-SOP-18-1, Main and Reheat Steam System TP-SQ-11.016, Post Work Test Program (historical)
 
Condition Reports
IP2-2009-0983 IP2-2009-0137 IP2-2008-5636 IP2-2009-0077 IP2-2009-0069 IP2-2009-0062 IP2-2008-5621 IP2-2009-0781
Work Orders
Work Orders
IP2-03-11725  IP2-02-32013  51305160 
NP-99-07694
Drawings B235623-6, Atmospheric Steam Dump Panel
Miscellaneous
9321-F-70313, Auxiliary Boiler Feed Pump Room Instrument Piping 
WCAP-12312, Rev. 2, Safety Evaluation for an Ultimate Heat Sink Temperature Increase to 95F
Miscellaneous
  at Indian Point Unit 2
IP2 Maintenance Rule Basis for Main Steam System IP2-MS-DBD, Design Basis Document for the Main Steam System IPT-RPT-05-00071, Appendix R Safe Shutdown Analysis SEE-03-5, Indian Point Unit 2 RHR Cooldown Analysis for the 5% Power Uprate
Heat exchanger data sheet for containment recirculation pump number 22 motor cooler
IP2 Inservice Testing Program Basis Data Sheets for PCV-1136 & 1137 (23/24 SG ADVs) ER 06-2-012, Install Secondary Backup Nitrogen Cylinders at both S/G ADV Local Control Panels in the ABFP Building
WCAP-7829, Fan Cooler Motor Unit Test
Environmental Qualification Report for Containment Recirculation Pump Motors
IP2-CCW-DBD, Component Cooling Water design bases document
   
IP2-DBD-207, Design Basis Document for 118V AC Electrical System
A-6  Attachment  Section 1R19: Post-Maintenance Testing
AMSE OM-2001 Edition
 
Unit 2 active scaffold list
VM 1073-1.2, Vendor manual for auxiliary component cooling pumps
VM 1100, vendor manual for 118V AC solid state static inverters
Work order NP-89-43777, replacement of 22 ACCP impeller
IP2-AFW-DBD, Rev. 1, AFW Design Basis Document
Section 1R18: Plant Modifications
Procedures
Procedures
OAP-24, "Operations Testing," Rev. 3 2-PT-M021C, Rev. 16, Emergency Diesel Generator 23 Load Test 0-GNR-403-ELC, Emergency Diesel Generator Quarterly Inspection
2-SOP-18-1, Main and Reheat Steam System
2-PT-Q033B, 21 Charging Pump 2-SOP-4.1.2, Rev. 34, Component Cooling System Operation
TP-SQ-11.016, Post Work Test Program (historical)
Orders 51797559 51797558 52027651 00183296 00157710 51675732
Condition Reports
IP2-2009-0983          IP2-2009-0137      IP2-2008-5636        IP2-2009-0077
IP2-2009-0069          IP2-2009-0062      IP2-2008-5621        IP2-2009-0781
Work Orders
IP2-03-11725            IP2-02-32013        51305160
Drawings
B235623-6, Atmospheric Steam Dump Panel
9321-F-70313, Auxiliary Boiler Feed Pump Room Instrument Piping
Miscellaneous
IP2 Maintenance Rule Basis for Main Steam System
IP2-MS-DBD, Design Basis Document for the Main Steam System
IPT-RPT-05-00071, Appendix R Safe Shutdown Analysis
SEE-03-5, Indian Point Unit 2 RHR Cooldown Analysis for the 5% Power Uprate
IP2 Inservice Testing Program Basis Data Sheets for PCV-1136 & 1137 (23/24 SG ADVs)
ER 06-2-012, Install Secondary Backup Nitrogen Cylinders at both S/G ADV Local Control
        Panels in the ABFP Building
                                                                                Attachment


                                              A-6
Section 1R22: Surveillance Testing
Section 1R19: Post-Maintenance Testing
 
Procedures
Procedures
2-PT-2M4, Safety Injection System Train "A" Actuation Logic and Master Relay Test 2-PT-Q013, Inservice Valve Tests 2-PT-Q013-DS027, Valve 888A IST Data Sheet
OAP-24, Operations Testing, Rev. 3
0-SOP-LEAKRATE-001, Rev. 1, RCS Leakrate Surveillance, Evaluation and Leak Identification 2-PT-Q030C, Rev. 18, 23 Component Cooling Water Pump
2-PT-M021C, Rev. 16, Emergency Diesel Generator 23 Load Test
Drawings 11497, Valve 888A
0-GNR-403-ELC, Emergency Diesel Generator Quarterly Inspection
 
2-PT-Q033B, 21 Charging Pump
Condition Reports
2-SOP-4.1.2, Rev. 34, Component Cooling System Operation
IP2-2007-1754 IP2-2008-1443 IP2-2008-2002 IP2-2007-3329
Orders
Orders 51694305  Miscellaneous
51797559      51797558      52027651      00183296      00157710      51675732
IP2-ESF DBD, Design Basis Document for Engineered Safeguards Features System IP2 Inservice Testing Program Data Sheet - Valve 888A
Section 1R22: Surveillance Testing
PGI-00066-01, 888 A & B Diff Pr Calc
Section 1EP6: Drill Evaluation
 
Procedures
Procedures
IP-EP-120, Rev. 3, Emergency Classification
2-PT-2M4, Safety Injection System Train A Actuation Logic and Master Relay Test
Miscellaneous
2-PT-Q013, Inservice Valve Tests
IP-EP-115, Rev. 24, form EP-1 radiological emergency data forms dated 2/23/09
2-PT-Q013-DS027, Valve 888A IST Data Sheet
0-SOP-LEAKRATE-001, Rev. 1, RCS Leakrate Surveillance, Evaluation and Leak Identification
Section 2OS1:  Access Control to Radiologically Significant Areas and
2-PT-Q030C, Rev. 18, 23 Component Cooling Water Pump
Section 2OS2: ALARA Planning and Controls
Drawings
 
11497, Valve 888A
Condition Reports
IP2-2007-1754        IP2-2008-1443        IP2-2008-2002        IP2-2007-3329
Orders
51694305
Miscellaneous
IP2-ESF DBD, Design Basis Document for Engineered Safeguards Features System
IP2 Inservice Testing Program Data Sheet - Valve 888A
PGI-00066-01, 888 A & B Diff Pr Calc
Section 1EP6: Drill Evaluation
Procedures
Procedures
EN-RP-100, Rev. 03, Radworker Expectations EN-RP-101, Rev. 04, Access Control for Radiologically Controlled Areas EN-RP-102, Rev. 02, Radiological Control
IP-EP-120, Rev. 3, Emergency Classification
EN-RP-105, Rev. 04, Radiation Work Permits EN-RP-108, Rev. 07, Radiation Protection Posting EN-RP-110, Rev. 05, ALARA Program 
A-7  Attachment  EN-RP-121, Rev. 04, Radioactive Material Control EN-RP-131, Rev. 06, Air Sampling EN-RP-141, Rev. 04, Job Coverage EN-RP-151, Rev. 02, Radiological Diving EN-RP-202, Rev. 06, Personnel Monitoring EN-RP-203, Rev. 02, Dose Assessment
EN-RP-204, Rev. 02, Special Monitoring Requirements EN-RP-205, Rev. 02, Prenatal Monitoring EN-RP-208, Rev. 02, Whole Body Counting and In-Vitro Bioassay
Condition Reports
CR-IP3-2009-00752, CR-IP3-2009-00785, CR-IP3-2009-00857, CR-IP3-2009-00885 CR-IP3-2009-00886, CR-IP3-2009-00937, CR-IP3-2009-00998, CR-IP3-2009-01006 CR-IP3-2009-01107, CR-IP3-2009-01154, CR-IP3-2009-01169, CR-IP3-2009-01171 CR-IP3-2009-01183, CR-IP3-2009-01253, CR-IP3-2009-01293, CR-IP3-2009-01295 CR-IP3-2009-01296, CR-IP3-2009-01318, CR-IP2-2009-00883, CR-IP2-2009-01110,  CR-IP2-2009-01111, CR-IP2-2009-01113, CR-IP2-2009-01114
Miscellaneous
Miscellaneous
Radiation Protection Attention Logs (Electronic Dosimeter Alarms) TEDE ALARA Evaluations ALARA Committee Reviews RP-STD-XX, Rev. X, "Unreported Dosimeter Alarms and Anomolies" (Draft)
IP-EP-115, Rev. 24, form EP-1 radiological emergency data forms dated 2/23/09
IPEC Snapshot Self-Assessment Report (IP3-LO-2007-0010) July 2007 - June 2008. RWP's: 2009-002, 2009-003, 2009-2021, 2009-3001, , 2009-3002, 2009-3056, 2009-3501, 2009-3504, 2009-3515, 2009-3529
Section 2OS1: Access Control to Radiologically Significant Areas and
Section 4OA1:  Performance Indicator Verification
Section 2OS2: ALARA Planning and Controls
 
Procedures
EN-EP-201, "Performance Indicators," Rev. 6 EN-LI-114, "Performance Indicator Process," Rev. 3 NEI 99-02, "Regulatory Assessment Performance Indicator Guideline," Rev. 5 0-CY-2765, Rev. 3, Coolant Activity Limits
EN-RP-100, Rev. 03, Radworker Expectations
EN-RP-101, Rev. 04, Access Control for Radiologically Controlled Areas
EN-RP-102, Rev. 02, Radiological Control
EN-RP-105, Rev. 04, Radiation Work Permits
EN-RP-108, Rev. 07, Radiation Protection Posting
EN-RP-110, Rev. 05, ALARA Program
                                                                                  Attachment


                                              A-7
Section 4OA2: Identification and Resolution of Problems
EN-RP-121, Rev. 04, Radioactive Material Control
 
EN-RP-131, Rev. 06, Air Sampling
EN-RP-141, Rev. 04, Job Coverage
EN-RP-151, Rev. 02, Radiological Diving
EN-RP-202, Rev. 06, Personnel Monitoring
EN-RP-203, Rev. 02, Dose Assessment
EN-RP-204, Rev. 02, Special Monitoring Requirements
EN-RP-205, Rev. 02, Prenatal Monitoring
EN-RP-208, Rev. 02, Whole Body Counting and In-Vitro Bioassay
Condition Reports
CR-IP3-2009-00752, CR-IP3-2009-00785, CR-IP3-2009-00857, CR-IP3-2009-00885
CR-IP3-2009-00886, CR-IP3-2009-00937, CR-IP3-2009-00998, CR-IP3-2009-01006
CR-IP3-2009-01107, CR-IP3-2009-01154, CR-IP3-2009-01169, CR-IP3-2009-01171
CR-IP3-2009-01183, CR-IP3-2009-01253, CR-IP3-2009-01293, CR-IP3-2009-01295
CR-IP3-2009-01296, CR-IP3-2009-01318, CR-IP2-2009-00883, CR-IP2-2009-01110,
CR-IP2-2009-01111, CR-IP2-2009-01113, CR-IP2-2009-01114
Miscellaneous
Radiation Protection Attention Logs (Electronic Dosimeter Alarms)
TEDE ALARA Evaluations
ALARA Committee Reviews
RP-STD-XX, Rev. X, Unreported Dosimeter Alarms and Anomolies (Draft)
IPEC Snapshot Self-Assessment Report (IP3-LO-2007-0010) July 2007 - June 2008.
RWPs: 2009-002, 2009-003, 2009-2021, 2009-3001, , 2009-3002, 2009-3056, 2009-3501,
2009-3504, 2009-3515, 2009-3529
Section 4OA1: Performance Indicator Verification
EN-EP-201, "Performance Indicators," Rev. 6
EN-LI-114, Performance Indicator Process, Rev. 3
NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Rev. 5
0-CY-2765, Rev. 3, Coolant Activity Limits
Section 4OA2: Identification and Resolution of Problems
Procedures
Procedures
EN-LI-102, Rev. 13, Corrective Action Process  
EN-LI-102, Rev. 13, Corrective Action Process
Condition Reports
IP2-2009-00342        IP2-2009-00483        IP2-2004-03106        IP2-2007-01893
IP2-2008-05382        IP2-2009-00486        IP2-2009-00027        IP2-2009-00139
IP2-2009-00143        IP2-2009-00148
                                                                                Attachment


Condition Reports
                  A-8
IP2-2009-00342 IP2-2009-00483 IP2-2004-03106 IP2-2007-01893  IP2-2008-05382 IP2-2009-00486  IP2-2009-00027 IP2-2009-00139  IP2-2009-00143 IP2-2009-00148
        LIST OF ACRONYMS
   
ALARA  as low as is reasonably achievable
A-8 Attachment  LIST OF ACRONYMS  
ABFW   auxiliary boiler feedwater
ABFP   auxiliary boiler feedwater pump
ALARA  as low as is reasonably achievable ABFW   auxiliary boiler feedwater ABFP   auxiliary boiler feedwater pump ACCP   auxiliary component cooling pump  
ACCP   auxiliary component cooling pump
ADAMS  Agency-wide Document and Management System ASME   American Society of Mechanical Engineers CAP   corrective action program CCW   component cooling water CDF   core damage frequency  
ADAMS  Agency-wide Document and Management System
CFR   Code of Federal Regulations CST   condensate storage tank EDO   Executive Director of Operations EDG   emergency diesel generator ENTERGY   Entergy Nuclear Northeast EP   Emergency Preparedness HRA   high radiation area  
ASME   American Society of Mechanical Engineers
IMC   Inspection Manual Chapter IPEC   Indian Point Energy Center IST   in-service test MCC   motor control center NCV   non-cited violation  
CAP     corrective action program
NDE   non-destructive examination NRC   Nuclear Regulatory Commission NRR   Nuclear Reactor Regulation NSR   non safety-related PARS   Publicly Available Records System  
CCW     component cooling water
PI   performance indicator RCA   radiologically controlled area RCS   reactor coolant system RWP   radiation work permit RWST   refueling water storage tank  
CDF     core damage frequency
SDP   significance determination process SER   safety evaluation report SG   steam generator SR   safety related SSC   structures, systems, and components  
CFR     Code of Federal Regulations
TS   Technical Specification UFSAR  Updated Final Safety Evaluation Report URI   unresolved item WO   work order
CST     condensate storage tank
EDO     Executive Director of Operations
EDG     emergency diesel generator
ENTERGY Entergy Nuclear Northeast
EP     Emergency Preparedness
HRA     high radiation area
IMC     Inspection Manual Chapter
IPEC   Indian Point Energy Center
IST     in-service test
MCC     motor control center
NCV     non-cited violation
NDE     non-destructive examination
NRC     Nuclear Regulatory Commission
NRR     Nuclear Reactor Regulation
NSR     non safety-related
PARS   Publicly Available Records System
PI     performance indicator
RCA     radiologically controlled area
RCS     reactor coolant system
RWP     radiation work permit
RWST   refueling water storage tank
SDP     significance determination process
SER     safety evaluation report
SG     steam generator
SR     safety related
SSC     structures, systems, and components
TS     Technical Specification
UFSAR  Updated Final Safety Evaluation Report
URI     unresolved item
WO     work order
                                                Attachment
}}
}}

Latest revision as of 05:51, 14 November 2019

IR 05000247-09-002, on 01/01/09 to 03/31/09, Indian Point Nuclear Generating Unit 2; Adverse Weather Protection; Fire Protection; Maintenance Effectiveness; Maintenance Risk Assessments; Surveillance Testing; and Radiological Access Control
ML091340445
Person / Time
Site: Indian Point Entergy icon.png
Issue date: 05/14/2009
From: Mel Gray
Reactor Projects Branch 2
To: Pollack J
Entergy Nuclear Operations
Gray M, RI/DRP/BR2/610-337-5209
References
FOIA/PA-2011-0021 IR-09-002
Download: ML091340445 (45)
v • d • e


See also: IR 05000247/2009002

Text

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION I

475 ALLENDALE ROAD

KING OF PRUSSIA, PA 19406-1415

May 14, 2009

Mr. Joseph E. Pollock

Site Vice President

Entergy Nuclear Operations, Inc.

Indian Point Energy Center

450 Broadway, GSB

Buchanan, NY 10511-0249

SUBJECT: INDIAN POINT NUCLEAR GENERATING UNIT 2 - NRC INTEGRATED

INSPECTION REPORT 05000247/2009002

Dear Mr. Pollock:

On March 31, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection

at Indian Point Nuclear Generating Unit 2. The enclosed integrated inspection report

documents the inspection results, which were discussed on April 15, 2009, with you and other

members of your staff.

The inspection examined activities conducted under your license as they relate to safety and

compliance with the Commissions rules and regulations, and with the conditions of your

license. The inspectors reviewed selected procedures and records, observed activities, and

interviewed personnel.

This report documents seven findings of very low safety significance (Green). Six of these

findings were also determined to be violations of NRC requirements. However, because of their

very low safety significance, and because the findings were entered into your corrective action

program, the NRC is treating these findings as non-cited violations (NCVs) consistent with

Section VI.A.1 of the NRC Enforcement Policy. If you contest any NCV in this report, you

should provide a written response within 30 days of the date of this inspection report, with the

basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk,

Washington D.C. 20555-0001; with copies to the Regional Administrator, Region I; the Director,

Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC

20555-0001; and the NRC Senior Resident Inspector at Indian Point Nuclear Generating Unit 2.

In addition, if you disagree with the characterization of any finding, you should provide a

response within 30 days of the date of this inspection report, with the basis for your

disagreement, to the Regional Administrator and the NRC Resident Inspectors at Indian Point

Nuclear Generating Unit 2. The information you provide will be considered in accordance with

Inspection Manual Chapter 0305.

J. Pollock 2

In accordance with Title 10 of the Code of Federal Regulations Part 2.390 of the NRCs Rules

of Practice, a copy of this letter, its enclosure, and your response (if any) will be available

electronically for public inspection in the NRC Public Document Room of from the Publicly

Available Records (PARS) component of the NRCs document system (ADAMS).

ADAMS is accessible from the NRC Web Site at http://www.nrc.gov/reading-rm/adams.html

(the Public Electronic Reading Room).

Sincerely,

/RA/

Mel Gray, Chief

Projects Branch 2

Division of Reactor Projects

Docket No. 50-247

License No. DPR-26

Enclosure: Inspection Report No. 05000247/2009002

w/ Attachment: Supplemental Information

cc w/encl:

Senior Vice President, Entergy Nuclear Operations

Vice President, Operations, Entergy Nuclear Operations

Vice President, Oversight, Entergy Nuclear Operations

Senior Manager, Nuclear Safety and Licensing, Entergy Nuclear Operations

Senior Vice President and COO, Entergy Nuclear Operations

Assistant General Counsel, Entergy Nuclear Operations

Manager, Licensing, Entergy Nuclear Operations

C. Donaldson, Esquire, Assistant Attorney General, New York Department of Law

A. Donahue, Mayor, Village of Buchanan

J. G. Testa, Mayor, City of Peekskill

R. Albanese, Four County Coordinator

S. Lousteau, Treasury Department, Entergy Services, Inc.

Chairman, Standing Committee on Energy, NYS Assembly

Chairman, Standing Committee on Environmental Conservation, NYS Assembly

Chairman, Committee on Corporations, Authorities, and Commissions

M. Slobodien, Director, Emergency Planning

P. Eddy, NYS Department of Public Service

Assemblywoman Sandra Galef, NYS Assembly

T. Seckerson, County Clerk, Westchester County Board of Legislators

A. Spano, Westchester County Executive

R. Bondi, Putnam County Executive

C. Vanderhoef, Rockland County Executive

E. A. Diana, Orange County Executive

T. Judson, Central NY Citizens Awareness Network

M. Elie, Citizens Awareness Network

Public Citizen's Critical Mass Energy Project

M. Mariotte, Nuclear Information & Resources Service

F. Zalcman, Pace Law School, Energy Project

L. Puglisi, Supervisor, Town of Cortlandt

J. Pollock 3

Congressman John Hall

Congresswoman Nita Lowey

Senator Kirsten E. Gillibrand

Senator Charles Schumer

G. Shapiro, Senator Gillibrand 's Staff

J. Riccio, Greenpeace

P. Musegaas, Riverkeeper, Inc.

M. Kaplowitz, Chairman of County Environment & Health Committee

A. Reynolds, Environmental Advocates

D. Katz, Executive Director, Citizens Awareness Network

K. Coplan, Pace Environmental Litigation Clinic

M. Jacobs, IPSEC

W. Little, Associate Attorney, NYSDEC

M. J. Greene, Clearwater, Inc.

R. Christman, Manager Training and Development

J. Spath, New York State Energy Research, SLO Designee

F. Murray, President & CEO, New York State Energy Research

A. J. Kremer, New York Affordable Reliable Electricity Alliance (NY AREA)

J. Pollock 4

In accordance with Title 10 of the Code of Federal Regulations Part 2.390 of the NRCs Rules

of Practice, a copy of this letter, its enclosure, and your response (if any) will be available

electronically for public inspection in the NRC Public Document Room of from the Publicly

Available Records (PARS) component of the NRCs document system (ADAMS).

ADAMS is accessible from the NRC Web Site at http://www.nrc.gov/reading-rm/adams.html

(the Public Electronic Reading Room).

Sincerely,

/RA/

Mel Gray, Chief

Projects Branch 2

Division of Reactor Projects

Distribution w/encl: (via E-mail) C. Hott, DRP, RI, IP2

S. Collins, RA D. Hochmuth, DRP, OA

M. Dapas, DRA S. Campbell, RI OEDO

D. Lew, DRP R. Nelson, NRR

J. Clifford, DRP M. Kowal, NRR

M. Gray, DRP J. Boska, PM, NRR

B. Bickett, DRP J. Hughey, NRR

A. Rosebrook, DRP D. Bearde, DRP

S. McCarver, DRP ROPreports@nrc.gov

J. Heinly, DRP Region I Docket Room (w/concurrences)

G. Malone, DRP, SRI, IP2

SUNSI Review Complete: ____BSB____ (Reviewers Initial)

DOCUMENT NAME: G:\DRP\BRANCH2\A - INDIAN POINT 2\INSPECTION REPORTS\IP2 IR2009-002\IP2

2009002 REVFINAL.DOC

After declaring this document An Official Agency Record it will be released to the Public

To Receive a copy of this document, indicate in the box: C = Copy without attachment/enclosure E = Copy with attachment/enclosure N = No copy

Office RI/DRP RI/DRP RI/DRP

Name GMalone/BSB for BBickett/ MGray/

Date 05/14/09 05/14/09 05/14/09

OFFICAL AGENCY RECORD

1

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket No.: 50-247

License No.: DPR-26

Report No.: 05000247/2009002

Licensee: Entergy Nuclear Northeast (Entergy)

Facility: Indian Point Nuclear Generating Unit 2

Location: 450 Broadway, GSB

Buchanan, NY 10511-0249

Dates: January 1, 2009 through March 31, 2009

Inspectors: G. Malone, Senior Resident Inspector, Indian Point 2

C. Hott, Resident Inspector, Indian Point 2

J. Commisky, Health Physics Inspector, Region I

Approved By: Mel Gray, Chief

Projects Branch 2

Division of Reactor Projects

Enclosure

2

TABLE OF CONTENTS

SUMMARY OF FINDINGS ............................................................................................................... 3

REPORT DETAILS........................................................................................................................... 8

1. REACTOR SAFETY .................................................................................................................... 8

1R01 Adverse Weather Protection ............................................................................................... 8

1R04 Equipment Alignment ....................................................................................................... 10

1R05 Fire Protection .................................................................................................................. 10

1R07 Heat Sink Performance .................................................................................................... 14

1R11 Licensed Operator Requalification Program ..................................................................... 15

1R12 Maintenance Effectiveness ............................................................................................... 15

1R13 Maintenance Risk Assessments and Emergent Work Control .......................................... 18

1R15 Operability Evaluations ..................................................................................................... 19

1R18 Plant Modifications ........................................................................................................... 20

1R19 Post-Maintenance Testing ................................................................................................ 21

1R22 Surveillance Testing ......................................................................................................... 21

1EP6 Drill Evaluation ................................................................................................................ 24

2. RADIATION SAFETY ................................................................................................................ 24

2OS1 Access Control to Radiologically Significant Areas ........................................................... 24

2OS2 ALARA Planning and Controls.......................................................................................... 28

4. OTHER ACTIVITIES.................................................................................................................. 30

4OA1 Performance Indicator Verification ................................................................................... 30

4OA2 Identification and Resolution of Problems ......................................................................... 31

4OA3 Event Followup ................................................................................................................. 31

4OA5 Other Activities ................................................................................................................. 32

4OA6 Meetings........................................................................................................................... 33

ATTACHMENT: SUPPLEMENTAL INFORMATION .................................................................... A-1

SUPPLEMENTAL INFORMATION ............................................................................................... A-1

KEY POINTS OF CONTACT ........................................................................................................ A-1

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED ............................................................. A-1

LIST OF DOCUMENTS REVIEWED ............................................................................................ A-2

LIST OF ACRONYMS .................................................................................................................. A-8

Enclosure

3

SUMMARY OF FINDINGS

IR 05000247/2009-002; 01/01/2009 - 03/31/2009; Indian Point Nuclear Generating (Indian

Point) Unit 2; Adverse Weather Protection; Fire Protection; Maintenance Effectiveness;

Maintenance Risk Assessments; Surveillance Testing; and Radiological Access Control.

This report covered a three-month period of inspection by resident and region based inspectors.

Seven findings of very low significance (Green) were identified, six of which were also

determined to be non-cited violations (NCV). The significance of most findings is indicated by

their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609,

Significance Determination Process. The cross-cutting aspect for each finding was

determined using IMC 0305, Operating Reactor Assessment Program. Findings for which the

significance determination process (SDP) does not apply may be Green, or be assigned a

severity level after NRC management review. The NRCs program for overseeing safe

operation of commercial nuclear power reactors is described in NUREG-1649, Reactor

Oversight Process, Revision 4, dated December 2006.

A. NRC-Identified and Self-Revealing Findings

Cornerstone: Initiating Events

  • Green. The inspectors identified a NCV of very low safety significance related to 10 CFR

50, Appendix B, Criterion XVI, Corrective Actions, because Entergy did not promptly

identify and correct an adverse condition related to an electrical fault. Specifically,

personnel did not identify a safety-related cubicle had experienced an electrical fault

prior to replacement of upstream fuses and restoration of power to the damaged cubicle.

Entergy entered the issue into the corrective action program as IP2-2009-00342 and

IP2-2009-00483, trained all operations personnel on the requirements to replace fuses

and re-energize electrical equipment, and plans to revise the operations procedure for

operating electrical equipment.

This issue was more than minor because the finding was associated with the external

factors attribute of the Initiating Events cornerstone and impacted the cornerstone

objective of limiting the likelihood of those events that upset plant stability and challenge

critical safety systems during shutdown as well as power operations. The inspectors

determined that the issue increased the likelihood of a fire in the emergency diesel

generator (EDG) building. The condition was evaluated by a Senior Reactor Analyst

utilizing Phase 2 of IMC 0609 Appendix F, Fire Protection Significance Determination

Process. It was determined that in the event of a fire consuming the MCC, no transient

would be placed on the plant and no components required to safely shutdown the plant

would be impacted. As a result, in accordance with task 2.3.5 of Appendix F, the issue

was screened to Green.

The inspectors determined that a cross-cutting aspect was associated with this finding

in the area of human performance related to conservative decision making. Specifically,

Entergys decision-making was non-conservative related to its decisions on the process

used to identify the source of the acrid odor; re-energize the damaged electrical

equipment; and keep a damaged electrical component energized for 14 days prior to its

removal from the MCC. H.1(b) per IMC 0305] (Section 1R05)

Enclosure

4

  • Green. The inspectors identified a NCV of very low safety significance related to TS 5.4.1, Administrative Controls: Procedures, because Entergy did not maintain an

adequate maintenance procedure for a safety-related electrical motor control center

(MCC). Specifically, the eight-year maintenance procedure for the affected EDG

ventilation MCC did not contain an adequate method to identify high resistance

connections within the cubicle as was expected in the applicable preventative

maintenance industry template. Subsequently, a high resistance connection within the

MCC developed into a phase-to-phase electrical fault on January 28, 2009. Entergy

entered the issue into the corrective action program, scoped the affected MCC and 21

additional MCCs into the sites thermography program, and planned to revise the

maintenance procedure.

This issue was more than minor because the finding was associated with the external

factors attribute of the Initiating Events cornerstone and impacted the cornerstone

objective of limiting the likelihood of those events that upset plant stability and challenge

critical safety systems during shutdown as well as power operations. Specifically, the

high resistance connection degraded into a phase-to-phase fault and increased the

likelihood of a fire in the EDG building. The condition was evaluated by a Senior

Reactor Analyst utilizing Phase 2 of IMC 0609 Appendix F, Fire Protection Significance

Determination Process. It was determined that in the event of a fire consuming the

MCC, no transient would be placed on the plant and no components required to safely

shutdown the plant would be impacted. As a result, in accordance with task 2.3.5 of

Appendix F, the issue was screened to Green.

The inspectors determined that the finding had a cross-cutting aspect associated with

the area of problem identification and resolution related to the use of operating

experience (OE). Specifically, Entergy personnel did not implement industry

recommended practices, or an alternate equivalent method, for identifying high

resistance connections in electrical switchgear. P.2(b) per IMC 0305] (Section 1R12)

Cornerstone: Mitigating Systems

  • Green. The inspectors identified a finding of very low safety significance because

Entergy personnel did not adequately implement procedure EN-LI-102, Corrective Action

Process, and promptly identify a condition adverse to quality associated with open

louvers in a fire protection pump room following pump testing on January 14, 2009. The

open louvers resulted in freezing conditions in fire protection piping located in the room

and cracked two six-inch header isolation valves on January 17, 2009. Entergy entered

the issue into the corrective action program and performed a site-wide extent-of-

condition walkdown of louvers.

The finding was more than minor because it was associated with the protection against

external factors attribute of the Mitigating Systems cornerstone and it affected the

cornerstone objective of ensuring the reliability of systems that respond to initiating

events to prevent undesirable consequences. This finding was evaluated using Phase

1 of IMC 0609 Appendix F, Fire Protection Significance Determination Process. The

inspectors determined the issue was of very low safety significance (Green) because

the cracked valves were easily isolated and did not pass sufficient water to render the

fire header non-functional (low degradation rating).

Enclosure

5

The inspectors determined that the finding had a cross-cutting aspect in the area of

human performance related to work practices - human error prevention techniques.

Specifically, Entergy personnel that routinely tour the 11 fire pump house did not

question the abnormally cold room temperatures. H.4(a) per IMC 0305] (Section 1R01)

  • Green. The inspectors identified a NCV of very low safety significance related to License

Condition 2.K., fire protection program, because personnel did not promptly identify and

correct a degraded three-hour rated fire door latch mechanism on the west entrance of

the 480-Volt switchgear room. Specifically, inspectors identified the fire door in a non-

functional state on several instances over the course of a month. Entergy personnel

replaced the fire door latch mechanism on March 3, 2009. This issue was entered into

the corrective action program as six condition reports spanning several weeks and

included an extent of condition walkdown of site fire doors.

The finding was more than minor because it is associated with the protection against

external factors attribute of the Mitigating Systems cornerstone and affected the

cornerstone objective of ensuring the reliability of systems that respond to initiating

events to prevent undesirable consequences. This fire door, when degraded, impacts

the reliability of mitigating systems in the 480-Volt switchgear room that are relied upon

during a postulated large fire in the turbine building, and vice versa. This finding was

evaluated using Phase 1 of IMC 0609 Appendix F, Fire Protection Significance

Determination Process. Since the area in question had a fire watch posted during the

time the door was degraded for an unrelated issue, an adequate level of protection was

maintained to compensate for the degraded door. As such, according to task 1.3.1, the

inspectors determined the finding was Green.

The inspectors determined that the finding had a cross-cutting aspect in the area of

problem identification and resolution because Entergy personnel did not thoroughly

evaluate a degraded fire door latch on several occasions, such that the resolution of the

problems addressed the causes. P.1(c) per IMC 0305] (Section 1R05)

  • Green. The inspectors identified a NCV of very low safety significance related to 10 CFR

50.65(a)(4), because Entergy personnel did not adequately assess the risk associated

with the unavailability of the Refueling Water Storage Tank (RWST) level indication

during planned maintenance on the level transmitters and instrumentation. Entergy

entered the issue into the corrective action program (CR-IP2-2009-00342), updated the

risk model to include the maintenance activity, assessed the risk, and appropriately

coded the maintenance activity to ensure it would be risk assessed in the future.

The inspectors determined that this finding was more than minor because it was a

maintenance risk assessment issue in which personnel did not consider risk significant

SSCs that were unavailable during maintenance. The RWST level indication is

specifically listed in Table 2 of the plant specific Phase 2 SDP risk-informed inspection

notebook. The inspectors determined the significance of this issue in accordance with

IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management

Significance Determination Process. The inspectors determined that this finding was of

very low safety significance because the Incremental Core Damage Probability Deficit

was less than 1E-6.

The inspectors determined that the finding had a cross-cutting aspect in the area of

human performance related to work control. Specifically, Entergy personnel did not

Enclosure

6

appropriately plan work activities by incorporating risk insights for affected plant

equipment. H.3(a) per IMC 0305] (Section 1R13)

  • Green. The inspectors identified a NCV of very low safety significance related to 10

CFR 50.55a, Codes and standards, because Entergys procedure, 2-PT-Q031A for an

auxiliary component cooling water pump, did not contain appropriate acceptance criteria

for positively determining that safety-related check valves performed their safety function

when required in accordance with the American Society of Mechanical Engineers

(ASME) OM Code. Specifically, the test used reverse rotation of a parallel pump to

verify that the pumps discharge check valve was closed although previous site-specific

experience demonstrated that the pump impeller would not rotate backwards when the

check valve was stuck open. Entergy entered this issue into their corrective action

program as CR-2009-1312.

The inspectors determined that the performance deficiency was greater than minor

because it was associated with the procedure quality attribute of the Mitigating System

cornerstone and it adversely affected the cornerstones objective to ensure the reliability

of systems that respond to initiating events to prevent undesirable consequences.

Specifically, the test criterion used in procedure 2-PT-Q013A did not ensure that valve

755A reliably performed its safety function when tested as demonstrated by testing

performed in January 2005. The inspectors determined that the performance deficiency

was of very low safety significance (Green) IMC 0609, Attachment 4, Phase 1 - Initial

Screening and Characterization of Findings. Specifically, the inspectors determined

that this finding was of very low safety significance because the finding did not result in

a loss of safety function and did not screen as potentially risk-significant due to external

events initiating events.

The inspectors determined the finding had a cross-cutting aspect related to effective

corrective actions in the corrective action program component of the problem

identification and resolution area. Specifically, Entergy personnel did not implement

effective corrective actions to resolve the testing inadequacy since 2005 and during

subsequent quarterly testing. P.1(d) per IMC 0305] (Section 1R22)

Cornerstone: Occupational Radiation Safety

  • Green. The inspectors identified a NCV of very low safety significance related to

Technical Specification 5.4.1.a, Procedures, because Entergy personnel did not

generate condition reports or investigation paperwork for multiple high dose-rate alarms

as required by station procedures. Specifically, personnel did not generate the required

condition reports and adequately document the investigations for six instances of

unplanned or un-briefed electronic dosimeter alarms that occurred between January

2009 and March 2009. The performance deficiency resulted in workers receiving

unanticipated dose rate alarms with no formally-documented investigation prior to

returning to work in a Radiologically Controlled Area. Entergy entered the finding into

the corrective action program as condition report CR-IP3-2009-01253 and 01318.

The finding is more than minor because it is associated with the Occupational Radiation

Safety cornerstone attribute of programs and process, and adversely affected the

objective to ensure adequate protection of worker health and safety from exposure to

radiation. Moreover, the inspectors identified a programmatic deficiency to maintain and

implement programs to keep exposures as low as reasonably achievable, because

Enclosure

7

multiple examples were identified regarding the failure to satisfy station radiation

protection procedures. Using the Occupational Radiation Safety Significance

Determination Process, the inspectors determined that the finding was of very low safety

significance (Green) because it did not involve: (1) as low as is reasonably achievable

planning and controls, (2) an overexposure of an individual, (3) a substantial potential for

overexposure, or (4) an impaired ability to assess dose.

The inspectors determined that the finding had a cross-cutting aspect related to

procedural adherence in the work practices component of the human performance area.

Specifically, Entergy personnel did not follow procedures to generate condition reports

and document investigations when high dose-rate alarms were received by workers.

H.4(b) per IMC 0305] (Section 2OS1)

B. Licensee-Identified Violations

None.

Enclosure

8

REPORT DETAILS

Summary of Plant Status

Indian Point Nuclear Generating (Indian Point) Unit 2 began the inspection period at full reactor

power and remained at or near full power during the quarter.

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection (71111.01 - 1 sample)

Impending Adverse Weather

a. Inspection Scope

The inspectors reviewed the overall preparations and protection of risk-significant

systems for extremely cold weather conditions from January 14 - 19, 2009. The

inspectors reviewed and assessed implementation of the sites adverse weather

preparation procedures and compensatory measures for the affected conditions before

the onset of and during the cold weather conditions. This included verification that

operator actions defined in their adverse weather procedure maintain readiness of

essential systems that are vulnerable to freezing temperatures. The inspectors verified

Entergy personnel implemented periodic equipment walkdowns or other measures to

ensure the condition of plant equipment was operable.

The inspectors also reviewed Entergys corrective action program to review previous

issues associated with cold weather preparations and freezing conditions. Documents

reviewed are listed in the attachment.

b. Findings

Introduction. The inspectors identified a Green finding because Entergy personnel did

not adequately implement procedure EN-LI-102, Corrective Action Process, and

promptly identify a condition adverse to quality associated with stuck-open louvers in a

fire protection pump room following pump testing on January 14, 2009.

Description. On January 17, 2009, during a period of sustained cold weather which

included sub-zero temperatures, control room personnel received a fire panel trouble

alarm indicative of a low-pressure condition in the fire header and dispatched a plant

operator to investigate. The operator identified spraying water from the body of a

ruptured six-inch fire protection valve located in the 11 fire pump house. The operator

isolated the broken valve from the fire header by shutting a manually-operated upstream

valve which stopped the water spray. In addition, the operator observed that the pump

house room was significantly colder than expected and subsequently identified the

rooms ventilation louvers to the outside were mechanically bound in the open position.

The operator disconnected the louver linkage and manually shut the louvers.

Enclosure

9

On January 21, 2009, the inspectors identified a second six inch valve that was cracked

due to the previous cold weather (freezing) conditions in the fire pump house. Entergy

personnel entered this issue into the corrective action program and performed site

walkdowns to identify additional adverse conditions associated with the cold weather.

The inspectors determined that Entergy did not fully implement Entergy procedure EN-

LI-102, Corrective Action Process. Specifically, EN-LI-102 requires plant personnel to

identify adverse conditions, including cold-weather related conditions, and then enter

them into the CAP for resolution. Attachment 9.2 of the procedure provides examples of

adverse conditions expected to be reported; Section 1 of the Attachment contains

examples of operational conditions requiring entry into the CAP including "events or

conditions that could negatively impact reliability or availability." Additionally, plant

operators should have had heightened awareness to cold weather conditions because

Entergy procedure OAP-008, "Severe Weather Preparations," requires in step 4.3.7,

when freezing conditions are expected, that increased monitoring of plant areas to

monitor for adverse effects on plant equipment and verify that adequate protection is

provided. Operations personnel did not identify abnormal conditions in the 11 fire pump

room that led to the freezing and subsequent rupture of fire protection components.

The inspectors determined it was reasonable for Entergy personnel to identify this issue

because operators should have identified that the louvers failed to shut following a

routine operations test of 11 fire pump on January 14, 2009. In addition, operators

perform tours of the pump house every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and should have identified the room

was much colder than normal.

Analysis. The inspectors identified a performance deficiency because Entergy

personnel did not implement procedure guidance and identify stuck open louvers and a

subsequent second cracked fire header valve in the 11 fire pump house. The finding

was more than minor because it was associated with the protection against external

factors attribute of the Mitigating Systems cornerstone and it affected the cornerstone

objective of ensuring the reliability of systems that respond to initiating events to prevent

undesirable consequences. Specifically, the failure of the six-inch valves impacted the

reliability of the fire header until the ruptured valve was isolated.

This finding was evaluated using Phase 1 of Inspection Manual Chapter (IMC) 0609

Appendix F, Fire Protection Significance Determination Process. The inspectors

determined the issue was of very low safety significance (Green) because the cracked

fire valves were easily isolated and did not pass sufficient water to render the fire

header non-functional. Specifically, the inspectors assigned a low degradation rating to

the fire header because the fire pumps were able to maintain pressure in the fire header

until the ruptured valves were isolated.

The inspectors determined that the finding had a cross-cutting aspect in the area of

human performance related to work practices - human error prevention techniques.

Specifically, Entergy personnel routinely tour the 11 fire pump house did not question

the abnormally cold room temperatures. (H.4(a) per IMC 0305)

Enforcement: Enforcement action does not apply because the performance deficiency

did not involve a violation of a regulatory requirement. Because this finding does not

involve a violation of regulatory requirements and has very low safety significance, it is

identified as FIN 05000247/2009002-01, Failure to Identify Open Louvers in 11 Fire

Pump House.

Enclosure

10

1R04 Equipment Alignment (71111.04Q - 3 samples)

Partial System Walkdowns

a. Inspection Scope

The inspectors performed partial system walkdowns to verify the operability of redundant

or diverse trains and components during periods of system train unavailability, or

following periods of maintenance. The inspectors referenced the system procedures,

the UFSAR, and system drawings to verify the alignment of the available train supported

its required safety functions. The inspectors also reviewed applicable condition reports

(CR) and work orders to ensure Entergy personnel identified and properly addressed

equipment discrepancies that could potentially impair the capability of the available train,

as required by Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix

B, Criterion XVI, Corrective Action. The documents reviewed during these inspections

are listed in the Attachment.

The inspectors performed a partial walkdown on the following systems, which

represented three inspection samples:

  • 21 and 22 component cooling water (CCW) system train when 23 CCW pump

was tagged out for maintenance;

  • City water system as a supply to auxiliary boiler feedwater (ABFW) when the

condensate storage tank was declared inoperable due to leakage;

  • 21 and 23 ABFW trains when 22 ABFW pump was tagged out and temporary

modifications were applied to 21 and 23 ABFW minimum flow lines.

b. Findings

No findings of significance were identified.

1R05 Fire Protection (71111.05Q - 5 samples)

a. Inspection Scope

The inspectors conducted tours of several fire areas to assess the material condition and

operational status of fire protection features. The inspectors verified, consistent with the

applicable administrative procedures, that: combustibles and ignition sources were

adequately controlled; passive fire barriers, manual fire-fighting equipment, and

suppression and detection equipment were appropriately maintained; and compensatory

measures for out-of-service, degraded, or inoperable fire protection equipment were

implemented in accordance with Entergys fire protection program. The inspectors

evaluated the fire protection program for conformance with the requirements of License

Condition 2.K. The documents reviewed during this inspection are listed in the

Attachment. This inspection represented five inspection samples for fire protection

tours, and was conducted in the following areas:

  • FZ 65, Main Steam/Feed Regulating Valve Areas;
  • FZ 23, 62A Auxiliary Feed Pump Room & Building;
  • FZ 14, 480V Vital AC Switchgear Room;
  • FZ 360, Station Blackout Diesel Area.

Enclosure

11

b. Findings

.1 Failure to Identify Damaged Components in EDG Ventilation Motor Control Center

Introduction: The inspectors identified a NCV of very low safety significance (Green)

related to 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, because Entergy

personnel did not promptly identify and correct an adverse condition related to an

electrical fault. Specifically, personnel did not identify a safety-related cubicle (bucket)

had experienced a fault prior to replacement of upstream fuses and restoration of power

to the cubicle.

Description: On January 28, 2009, operations personnel detected an acrid odor coming

from the emergency diesel generator (EDG) building. Operators entered the EDG

building to investigate the source of the acrid odor and identified that a MCC was de-

energized. Operations personnel did not identify external damage to the MCC; however,

operators did not open MCC panels to inspect for internal damage. Operators checked

the upstream 175 amp supply fuses, located in a different building, and identified that 2

of 3 fuses had blown. Operators opened the downstream breakers on the MCC in the

EDG building and then replaced the 175 amp supply fuses in the control building. Once

operators replaced the blown fuses, they re-energized the EDG building MCC#1, and

subsequently began to locally shut all of the cubicle switches. When operators

attempted to shut the switch associated with cubicle 4N, the switch did not function as

expected. Operators then opened the panel for cubicle 4N and identified charred

electrical components.

Entergy personnel generated a D level condition report (CR) for cubicle 4N on the

basis that it supplies a non safety-related (NSR) EDG room heater. Entergy personnel

closed the CR to a work request to troubleshoot and repair the NSR heater. However,

the inspectors questioned the classification of the MCC and determined that the charred

components were safety related (SR). Cubicle 4N contains a SR main line switch and

SR 30 amp main line fuses. The 30 amp fuses are SR to isolate the NSR heaters from

the MCC in the event of a room heater fault. The inspectors also questioned the

appropriateness of leaving the damaged cubicle in the energized MCC. Following

inspector questions, Entergy staff issued another CR and removed the damaged cubicle

from the MCC on February 11. During removal of the charred cubicle, maintenance

personnel were unable to disconnect the main line cables due to arc-welding at the

termination and subsequently had to cut two of the three cables upstream of the

termination and cubicle switch. These cables and the line side of the switch were

energized from January 28 until February 11. After the damaged cubicle was removed,

engineering personnel performed an inspection and determined that the fault originated

from a high resistance connection on the C phase between the main fuse clip and the

cubicle supply switch in the 4N cubicle.

The inspectors determined that replacing the upstream 175 Amp fuses on and restoring

power to the EDG ventilation MCC #1, which contained the charred 4N cubicle, without

identifying the source of the acrid odor could have reinitiated the fault and increased the

probability of a fire. In addition, operations personnel tried to locally close the damaged

switch which could have also re-initiated the fault. Entergy staff also did not take action

to remove or de-energize the charred cubicle after the condition was identified on

January 28, 2009. The damaged cubicle was de-energized and removed from the MCC

on February 11 in response to the inspectors questions.

Enclosure

12

This issue was reasonable for the licensee to foresee and correct because acrid odor is

an indication of a fault. It was reasonable for Entergy personnel to open panel doors

and perform visual inspections of the affected MCC prior to replacing upstream fuses

and restoring power to the fault. The inspectors determined that the National Electrical

Code NFPA 70E, Standard for Electrical Safety in the Workplace, prohibits

reenergizing a circuit after a protective device has operated until it has been determined

that the automatic operation was a result of an overload and not a fault. The acrid odor

in the EDG building was an indication of a fault vice an overload condition. In addition,

once Entergy personnel identified the cubicle was charred and experienced an electrical

fault, industry standards would have operators immediately secure power and/or

remove the damaged gear from the MCC.

Entergy entered the issue into the corrective action program as IP2-2009-00342 and

IP2-2009-00483, trained all operations personnel on the requirements to replace fuses

and re-energize electrical equipment, and plans to review operations procedures for

operating electrical equipment.

Analysis: The inspectors determined that Entergys failure to promptly identify an

adverse condition associated with damaged electrical components constituted a

performance deficiency. This issue was more than minor because the finding was

associated with the external factors attribute of the Initiating Events cornerstone and

impacted the cornerstone objective of limiting the likelihood of those events that upset

plant stability and challenge critical safety systems during shutdown as well as power

operations. Specifically, operations personnel did not identify the source of the acrid

odor, indicative of an electrical fault, in the EDG building; re-energized damaged

electrical equipment; and left damaged electrical components (cubicle 4N) energized for

14 days prior to its removal from the MCC. The inspectors determined these issues

increased the likelihood of a fire in the EDG building. The condition was evaluated by a

Senior Reactor Analyst utilizing Phase 2 of IMC 0609 Appendix F, Fire Protection

Significance Determination Process. It was determined that in the event of a fire

consuming the MCC, no transient would be placed on the plant and no components

required to safely shutdown the plant would be impacted. As a result, in accordance

with task 2.3.5 of Appendix F, the issue was screened to Green.

The inspectors determined that a cross-cutting aspect was associated with this finding

in the area of human performance related to conservative decision making. Specifically,

Entergys decision-making was non-conservative as it related to the processes used to

identify the source of the acrid odor; re-energize the damaged electrical equipment; and

keep a damaged electrical component energized for 14 days prior to its removal from

the MCC. (H.1(b) per IMC 0305)

Enforcement: 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, requires that

measures shall be established to assure conditions adverse to quality, such as failures

and malfunctions are promptly identified and corrected. Contrary to the above, on

January 28, 2009, operations personnel did not identify that a safety-related bucket had

experienced a fault prior to replacing upstream fuses and restoring power to the bucket.

In addition, after replacing the upstream fuses, operations personnel tried to locally shut

the damaged cubicle switch and left damaged equipment energized until February 11,

2009. Entergy entered the issue into the corrective action program as IP2-2009-00342

and IP2-2009-00483, trained all operations personnel on the requirements to replace

fuses and re-energize electrical equipment, and plans to review operations procedures

Enclosure

13

for operating electrical equipment. Because the violation was of very low safety

significance and it was entered into the licensees corrective action program, this

violation is being treated as an NCV, consistent with the NRC Enforcement Policy: NCV 05000247/2009002-02, Failure to Identify Damaged Components in EDG

Ventilation Motor Control Center.

.2 Degraded Fire Door to the 480V Vital Bus Room

Introduction: The inspectors identified a NCV of very low safety significance (Green)

related to License Condition 2.K., fire protection program, because Entergy personnel

did not promptly identify and correct a degraded three-hour rated fire door on the west

entrance of the 480 Volt switchgear room.

Description: License Condition 2.K., fire protection program, requires that Entergy

implement and maintain in effect all provisions of the NRC-approved fire protection

program, as approved in part by the NRC Safety Evaluation Report (SER) dated

January 31, 1979. The January 31, 1979, SER requires administrative controls

comparable to those described in NRC Branch Technical Position 9.5-1, Guidelines for

Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976. Branch

Technical Position (BTP) 9.5-1 requires that measures be established to assure that

conditions adverse to fire protection, such as deficiencies, deviations, defective

components, and non-conformities are promptly identified, reported, and corrected.

On February 6, 2009, the inspectors performed a fire protection walkdown of the 480-

Volt switchgear room. The inspectors noted the three-hour rated, swing-type fire door

on the west side of the 480-Volt switchgear room was not latched closed. The

inspectors observed the door being held open by the latch mechanism which had not

repositioned to allow the door to shut. The inspectors observed the latch mechanism

did not move freely preventing the door from shutting automatically. The inspectors

shut the door and notified shift operations personnel who tightened latch screws on the

door and wrote a condition report.

On February 18, the inspectors identified the 480-Volt switchgear room door was not

latched shut again. The inspectors determined the door could not be closed due to

interference from the latch mechanism screw which had backed out. The inspectors

notified operations of the fire door issue. Operations personnel re-inserted the latch

mechanism screw and documented the issue in a condition report. The inspectors

questioned whether it was appropriate to re-insert a screw that had backed out on its

own in such a short period of time. Entergy personnel subsequently inspected the door

on February 23 and identified the screws holding the latch mechanism to the door were

stripped. Entergy personnel tapped new holes in the door latch mechanism and

installed new screws.

On March 3, inspectors identified the 480-Volt switchgear room fire door not latched

shut again. The inspectors observed the door was being held open by the latch

mechanism which had not repositioned to allow the door to shut. The inspectors noted

the latch mechanism did not move freely preventing the door from shutting

automatically. The inspectors notified operations personnel of the non-functioning fire

door and Entergy subsequently had a locksmith inspect the latch. The locksmith

installed a new latch mechanism on March 3 and determined the latch issues observed

were age-related due to interaction of wear products from the latch interfering with the

moving portions of the latch, as a result of latching and unlatching door operations.

Enclosure

14

Entergy entered the issue into the corrective action program on March 3, performed an

inspection of all fire doors onsite, and identified and corrected issues with other required

fire doors.

Analysis: The inspectors identified a performance deficiency because Entergy personnel

did not identify and correct the non-functional fire door. The finding was more than

minor because it is associated with the protection against external factors attribute of

the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring

the reliability of systems that respond to initiating events to prevent undesirable

consequences. Specifically, in the event of a large fire in the 480-Volt switchgear room

or the turbine building, the affected fire door is credited to prevent the spread of fire from

one area to the other area. This fire door, when degraded, impacts the reliability of

mitigating systems in the 480-Volt switchgear room that are relied upon during a large

fire in the turbine building, and vice versa.

This finding was evaluated using Phase 1 of IMC 0609 Appendix F, Fire Protection

Significance Determination Process. Since the area in question had a fire watch

posted during the time the door was degraded, an adequate level of protection was

maintained to compensate for the degraded door and resulted in the finding being of

very low safety significance. As such according to task 1.3.1, the inspectors determined

the finding was Green.

The inspectors determined that the finding had a cross-cutting aspect in the area of

problem identification and resolution because Entergy personnel did not thoroughly

evaluate a degraded fire door latch on several occasions, such that the resolution of the

problems addressed the causes. (P.1(c) per IMC 0305)

Enforcement: License Condition 2.K., fire protection program, requires that Entergy

implement and maintain in effect all provisions of the NRC-approved fire protection

program, as approved in part by the NRC Safety Evaluation Report (SER) dated

January 31, 1979. The January 31, 1979, SER requires administrative controls

comparable to those described in NRC Branch Technical Position 9.5-1, Guidelines for

Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976. Branch

Technical Position 9.5-1 requires that measures be established to assure that conditions

adverse to fire protection, such as deficiencies, deviations, defective components, and

non-conformities are promptly identified, reported, and corrected.

Contrary to the above, Entergy personnel did not promptly identify and then

subsequently correct the non-functional 480-Volt switchgear fire door. This fire door

was identified by inspectors in a non-functional state on February 6, February 18, and

again on March 3, 2009. Entergy entered the issue into the corrective action program

as IP2-2009-00526, IP2-2009-00680, IP2-2009-00709, IP2-2009-00834, IP2-2009-

00842, and IP2-2009-00843. Because the violation was of very low safety significance

and it was entered into the licensees corrective action program, this violation is being

treated as an NCV, consistent with the NRC Enforcement Policy: NCV 05000247/2009002-03, Failure to Identify and Promptly Correct Degraded 480-Volt

Switchgear Room Fire Door.

1R07 Heat Sink Performance (71111.07A - 1 sample)

a. Inspection Scope

Enclosure

15

The inspectors selected the 22 component water heat exchanger for review to

determine the heat exchangers readiness and availability to perform its safety functions.

The inspectors reviewed the design basis for the component, reviewed Entergy

commitments to NRC Generic Letter 89-13, and reviewed engineering reports that

documented results of previous internal inspections. The inspectors also observed the

disassembly, inspection, and cleaning of the heat exchanger and reviewed engineering

results of the inspection to verify that appropriate corrective actions were initiated for

deficiencies that were discovered. The inspectors reviewed documents for and verified

that the amount of tubes plugged within the heat exchanger did not exceed the

maximum amount allowed. Documents reviewed are listed in the appendix.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program

Quarterly Review (71111.11Q - 1 sample)

a. Inspection Scope

On February 23, 2009, the inspectors observed licensed operator simulator training

associated with a sustained loss of all alternating current (AC) power scenario, to verify

that operator performance was adequate, and that evaluators were identifying and

documenting crew performance problems. The inspectors evaluated the performance of

risk-significant operator actions, including the use of emergency operating procedures.

The inspectors assessed the clarity and effectiveness of communications, the

implementation of appropriate actions in response to alarms, the performance of timely

control board operation and manipulation, and the oversight and direction provided by

the control room supervisor. The inspectors also reviewed simulator fidelity with respect

to the actual plant. The inspectors evaluated licensed operator training for conformance

with the requirements of 10 CFR Part 55, Operator Licenses. The documents

reviewed during this inspection are listed in the Attachment. This observation of

operator simulator training represented one inspection sample.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness (71111.12Q - 3 samples)

a. Inspection Scope

The inspectors reviewed performance-based problems that involved structures,

systems, and components (SSCs) to assess the effectiveness of maintenance activities.

When applicable, the reviews focused on:

  • Proper Maintenance Rule scoping in accordance with 10 CFR 50.65;
  • Characterization of reliability issues;
  • Changing system and component unavailability;

Enclosure

16

  • Identifying and addressing common cause failures;
  • Trending of system flow and temperature values;
  • Appropriateness of performance criteria for SSCs classified (a)(2); and
  • Adequacy of goals and corrective actions for SSCs classified (a)(1).

The inspectors also reviewed system health reports, maintenance backlogs, and

Maintenance Rule basis documents. The inspectors evaluated maintenance

effectiveness and monitoring activities against the requirements of 10 CFR 50.65. The

documents reviewed during this inspection are listed in the Attachment. The following

Maintenance Rule samples were reviewed and represented three inspection samples:

  • RWST level indication system;
  • EDG fuel injection system; and
  • 480-Volt switchgear system.

b. Findings

Introduction: The inspectors identified a NCV of very low safety significance (Green)

related to TS 5.4.1, Administrative Controls: Procedures, because Entergy did not

maintain an adequate maintenance procedure for a safety-related electrical motor

control center (MCC). Specifically, the eight-year maintenance procedure for the

affected EDG ventilation MCC did not contain an adequate method to identify high

resistance connections within the cubicle.

Description: On January 28, 2009, operations personnel identified an acrid odor coming

from the EDG building. Subsequent personnel investigation revealed a charred cubicle

in a safety-related 480-Volt MCC. Specifically, cubicle 4N, in the EDG ventilation MCC,

experienced a phase-to-phase fault that caused the upstream 175 amp fuses to open

and de-energize the MCC. Entergy personnel subsequently generated a condition

report (CR) that was closed to a work request to troubleshoot and repair the cubicle.

Entergy personnel removed the damaged cubicle from the MCC on February 6 and

determined the likely cause to be a high-resistance connection between the cubicle

switch and 30 amp fuse clip on the C phase resulting in long-term overheating. This

overheating condition degraded the insulation between two of the three phases over

time and eventually resulted in a phase-to-phase fault on January 28, 2009.

The inspectors reviewed the 8-year maintenance procedure 2-MCC-003-ELC,

Klockner-Moeller, Series 200, 480 Volt Motor Control Center Preventive Maintenance,

which was performed on the affected EDG ventilation MCC on April 6, 2008. The

inspectors noted that the procedure was revised the same day to allow performance of

the maintenance without de-energizing the equipment. The revision resulted in portions

of the cubicle cleaning and inspection procedure not being performed because they

could not be safely performed while the cubicle was energized. The inspectors

determined that the procedure revision on April 6, 2008, was inappropriately treated as

an editorial revision without a technical evaluation of the change performed. In addition,

following interviews with Entergy personnel, it was determined that maintenance had not

been performed on this MCC prior to April 6, 2008.

Enclosure

17

The inspectors reviewed industry guidance for performing switchgear maintenance and

determined that Entergy did not include standard maintenance practices typically

utilized by its staff that would have identified a high resistance connection in the cubicle.

Specifically, continuity checks across contacts and switches were not performed, fuse

clip tensions and tightness were not performed, and all terminations could not be

checked due to the decision to perform the maintenance with portions of the cubicle

energized. In addition, the inspectors determined the EDG ventilation MCCs were not

included in Entergys thermography program, contrary to Entergy corporate preventive

maintenance templates. The inspectors determined that not performing thermography

on the EDG ventilation MCC constituted a missed opportunity to identify the high

resistance condition.

It is reasonable to consider the high resistance connection existed during the

maintenance performed on April 6, 2008, because high resistance connections do not

develop into phase-to-phase faults over a short period of time. This is an underlying

assumption for performing switchgear maintenance, which is intended to identify and

correct loose/high resistance connections, on an eight-year periodicity. In addition,

Entergys corporate template for switchgear maintenance recommends a six-year

periodicity and thermography every year. It is reasonable to expect Entergy to be aware

of the existing industry guidance as well as the Entergy corporate maintenance

templates.

Entergy entered the issue into the CAP as IP2-2009-00342 and IP2-2009-00483,

scoped the EDG ventilation MCC into the existing thermography program, performed an

extent-of-condition review that identified 21 additional panels that should be in the

thermography program, and plans to revise the maintenance procedure.

Analysis: The inspectors identified a performance deficiency because Entergy did not

maintain an adequate maintenance procedure for the safety-related EDG ventilation

MCC. This issue was more than minor because the finding was associated with the

external factors attribute of the Initiating Events cornerstone and impacted the initiating

events cornerstone objective of limiting the likelihood of those events that upset plant

stability and challenge critical safety systems during shutdown as well as power

operations. Specifically, the high resistance connection degraded into a phase-to-phase

fault and increased the likelihood of a fire in the EDG building. The condition was

evaluated by a Senior Reactor Analyst utilizing Phase 2 of IMC 0609 Appendix F, Fire

Protection Significance Determination Process. It was determined that in the event of a

fire consuming the MCC, no transient would be placed on the plant and no components

required to safely shutdown the plant would be impacted. As a result, in accordance

with task 2.3.5 of Appendix F, the issue was screened to Green.

The inspectors determined that the finding had a cross-cutting aspect associated with

the area of problem identification and resolution related to the use of operating

experience (OE). Specifically, Entergy personnel did not implement industry

recommended practices, or an alternate equivalent method, for identifying high

resistance connections in electrical switchgear. (P.2(b) per IMC 0305)

Enforcement. TS 5.4.1 Administrative Controls: Procedures, states, Written

procedures shall be established, implemented, and maintained covering the

requirements and recommendations of Appendix A of Regulatory Guide (RG) 1.33,

Revision 2. Appendix A of RG 1.33 requires procedures for maintenance activities that

Enclosure

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can affect the performance of safety related equipment. Contrary to the above, Entergy

did not maintain a maintenance procedure for a safety-related MCC cubicle.

Specifically, the eight-year maintenance procedure, first performed on April 6, 2008, did

not contain an adequate method to identify and correct high resistance connections in

the cubicle. Entergy entered the issue into the CAP as IP2-2009-00342 and IP2-2009-

00483. Because the violation was of very low safety significance and it was entered into

the licensees corrective action program, this violation is being treated as an NCV,

consistent with the NRC Enforcement Policy: NCV 05000247/2009002-04, Inadequate

Maintenance Procedure for EDG Ventilation Motor Control Center.

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 6 samples)

a. Inspection Scope

The inspectors reviewed scheduled and emergent maintenance activities to verify the

appropriate risk assessments were performed prior to removing equipment from service

for maintenance or repair. The inspectors verified that risk assessments were performed

as required by 10 CFR 50.65(a)(4) and were accurate and complete. When emergent

work was performed, the inspectors verified the plant risk was promptly reassessed and

managed. Documents reviewed during this inspection are listed in the Attachment. The

following activities represented six inspection samples:

maintenance outage;

  • Planned risk during 21 auxiliary boiler feedwater (ABFW) pump outage and reactor

protection system testing;

components during planned maintenance of 22 ABFW pump;

  • Planned maintenance on a reactor water storage tank level indicator;

applied to the 21 and 23 ABFW pumps; and

  • Planned risk during 23 EDG testing and maintenance.

b. Findings

Introduction: The inspectors identified a NCV of very low safety significance (Green)

related to 10 CFR 50.65(a)(4) because Entergy staff did not adequately assess the risk

associated with the unavailability of the Refueling Water Storage Tank (RWST) level

indication during planned maintenance on the level transmitters and instrumentation.

Description: On February 6, 2009, Entergy staff performed maintenance on the RWST

level indication system. The inspectors identified that the online risk assessment did not

consider planned maintenance on the RWST level indication, as required by 10 CFR

50.65(a)(4). The inspectors reviewed the work activity and noted the maintenance

scheduling software used by Entergy did not have the RWST maintenance coded as a

risk-significant activity. Entergys maintenance planning process prompts the

organization to evaluate the risk impact of all maintenance activities coded as risk-

significant. Therefore, a risk assessment was not performed for the quarterly RWST

level indication maintenance as required. In addition, the RWST level indication was not

represented in Entergys interactive risk model. Entergy staff subsequently updated the

risk model to include the RWST level indication and subsequently assessed the online

Enclosure

19

risk for the maintenance which resulted in a measurable increase in the core damage

frequency (CDF). The increase in CDF was not large enough to require entrance into

the higher risk category per Entergy procedures. In addition, the increase in CDF (1.1E-

6) combined with the limited duration of the maintenance (15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />) resulted in a

relatively small incremental core damage probability deficit (1.9E-9).

The inspectors determined this same maintenance activity is modeled in the Indian Point

Unit 3 risk model. Entergy entered the issue into the corrective action program (CR-IP2-

2009-00342), updated the risk model to include the maintenance activity, assessed the

risk, and appropriately coded the maintenance activity to ensure it would be risk

assessed in the future.

Analysis: The inspectors identified a performance deficiency in that Entergy staff did not

assess the increase in plant risk resulting from planned maintenance activities on RWST

level instrumentation as required by 10 CFR 50.65(a)(4). The inspectors determined

that this finding was more than minor because it was a risk assessment issue in which

Entergy personnel did not consider risk significant SSCs that were unavailable during

maintenance. Specifically, RWST level indication is included in Table 2 of the plant

specific Phase 2 SDP risk-informed inspection notebook. The inspectors assessed the

significance of this issue in accordance with IMC 0609, Appendix K, Maintenance Risk

Assessment and Risk Management Significance Determination Process. The

inspectors determined that this finding was of very low safety significance (Green)

because the incremental core damage probability deficit was less than 1E-6.

The inspectors determined that the finding had a cross-cutting aspect in human

performance related work control. Specifically, Entergy personnel did not appropriately

plan work activities by incorporating risk insights for affected plant equipment. (H.3(a)

per IMC 0305)

Enforcement: 10 CFR 50.65 (a)(4) states, in part that licensees shall assess and

manage the increase in risk that may result from the proposed maintenance activities

before performing those activities. Contrary to the above, on February 6, 2009, Entergy

performed maintenance on the RWST level indication system without assessing the

increase in risk. Entergy entered the issue into the corrective action program (CR-IP2-

2009-00342. Because this issue is of very low safety significance and is entered into

Entergys corrective action program, this violation is being treated as an NCV consistent

the NRC Enforcement Policy: NCV 05000247/2009002-05, Failure to Include RWST

Level Maintenance In Online Risk Assessment.

1R15 Operability Evaluations (71111.15 - 7 samples)

a. Inspection Scope

The inspectors reviewed operability evaluations to assess the acceptability of the

evaluations, the use and control of compensatory measures when applicable, and

compliance with Technical Specifications. The inspectors reviews included verification

that operability determinations were performed in accordance with procedure

ENN-OP-104, Operability Determinations. The inspectors assessed the technical

adequacy of the evaluations to ensure consistency with the Technical Specifications,

UFSAR, and associated design basis documents. The documents reviewed are listed in

Enclosure

20

the Attachment. The following operability evaluations were reviewed and represented

seven inspection samples:

  • Proximity of 480-Volt vital motor control center to an uninsulated steam line;
  • Leakage from condensate storage tank (CST) return piping;
  • Impacts of scaffolding built in the vicinity of the 21 and 22 component cooling water

heat exchangers;

reactor plant startups and shutdowns due to thermal transients;

  • Performance impact on the 21 and 22 auxiliary component cooling pumps (ACCPs)

with respect to a potential hydraulic lock-out condition of the 21 ACCP due to 22

ACCP larger impeller size;

  • Mechanical failure of a grease fitting on 21 service water pump; and
  • Low temperatures in condensate storage tank volume.

b. Findings

No findings of significance were identified. With respect to the CST return piping, the

inspectors determined Entergy operators maintained the CST aligned to supply water to

the AFW pumps. The inspectors concluded the leakage did not prevent the CST from

fulfilling its safety function. Specifically, design features of the CST and the elevation of

the return line relative to the leak location provided assurance that, in the event the CST

return line leak increased significantly, the CST water volume would have been

maintained above TS minimum required water level and able to supply the required

water to the auxiliary feedwater system.

1R18 Plant Modifications (71111.18 - 2 samples)

.1 Temporary Modifications

a. Inspection Scope

The inspectors reviewed one temporary plant modification package for securing

minimum flow lines on the motor driven auxiliary boiler feedwater pumps (ABFPs) and

controlling the operation on the ABFPs through a temporary operating procedure during

repairs of the CST return piping. The inspectors verified the design bases, licensing

bases, and performance capability of the system was not degraded by the temporary

modification. The inspectors review included Entergys engineering evaluation for

determining the ABFPs could start with the pumps required minimum flow being

achieved through the internal thrust balance lines while the minimum flow lines were

isolated. In addition, the inspectors interviewed plant staff, and reviewed issues entered

into the corrective action program to determine whether Entergy had been effective in

identifying and resolving problems associated with the temporary modification. The

documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

Enclosure

21

.2 Permanent Modifications

a. Inspection Scope

The inspectors reviewed modification documents associated with the installation of an

additional nitrogen backup power supply for the 21- 24 steam generator atmospheric

dump valves. The inspector verified that the modification was reviewed adequately to

verify the modification conformed to design criteria and did not interfere or invalidate

previous design assumptions or functions. The documents reviewed are listed in the

Attachment.

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing (71111.19 - 6 samples)

a. Inspection Scope

The inspectors reviewed post-maintenance test procedures and associated testing

activities for selected risk-significant mitigating systems, and assessed whether the

effect of maintenance on plant systems was adequately addressed by control room and

engineering personnel. The inspectors verified that: test acceptance criteria were clear,

the test demonstrated operational readiness and were consistent with design basis

documentation; test instrumentation had current calibrations, and appropriate range and

accuracy for the application; and the tests were performed as written, with applicable

prerequisites satisfied. Upon completion of the tests, the inspectors verified that

equipment was returned to the proper alignment necessary to perform its safety function.

Post-maintenance testing was evaluated for conformance with the requirements of 10

CFR 50, Appendix B, Criterion XI, Test Control. The documents reviewed are listed in

the Attachment. The following post-maintenance activities were reviewed and

represented six inspection samples:

  • Replacement of SG 23 pressure indicator PI-1355;
  • 22 component cooling water heat exchanger following maintenance;
  • 21 charging pump following recirculation valve maintenance;
  • Condensate storage tank return line following pipe section replacement;

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing (71111.22 - 6 samples)

a. Inspection Scope

The inspectors observed performance of portions of surveillance tests and/or reviewed

test data for selected risk-significant SSCs to assess whether they satisfied Technical

Specifications, UFSAR, Technical Requirements Manual, and Entergy procedure

Enclosure

22

requirements. The inspectors verified that: test acceptance criteria were identified,

demonstrated operational readiness, and were consistent with design basis

documentation; test instrumentation had accurate calibration, and appropriate range and

accuracy for the application; and tests were performed as written, with applicable

prerequisites satisfied. Following the tests, the inspectors verified that the equipment

was capable of performing the required safety functions. The inspectors evaluated the

surveillance tests against the requirements in Technical Specifications. The documents

reviewed during this inspection are listed in the Attachment. The following surveillance

tests were reviewed and represented six inspection samples:

  • 2-PT-Q031A, 21 Auxiliary Component Cooling Pump In-Service Test;
  • 2-PT-Q054, Pressurizer Level Bistables;
  • 2-PT-Q013 DS027, IST Valve Test of 888A (Safety Injection Pump Suction from

Residual Heat Removal heat Exchanger);

  • 2-PT-2M4, Safety Injection System Train A Actuation Logic and Master Relay Test;
  • 2-PT-Q030C, 23 Component Cooling Water Pump; and
  • 0-SOP-LEAKRATE-001, RCS Leakrate Surveillance, Evaluation, and Leak

Identification.

b. Findings

Introduction. The inspectors identified a NCV of very low safety significance (Green)

related to 10 CFR 50.55a, Codes and standards, because Entergys procedure 2-PT-

Q031A did not contain appropriate acceptance criteria for determining that safety-

related check valves performed their safety function when required in accordance with

the American Society of Mechanical Engineers (ASME) OM Code.

Description. Entergy procedure 2-PT-Q031A, 21 Auxiliary Component Cooling Pump

(ACCP), is an In-Service Test (IST) procedure that demonstrates the operability of the

21 ACCP, the pump bypass line check valve (755), the 21 ACCP discharge check valve

(755B), and the 22 ACCP discharge check valve (755A) in accordance with Technical

Specification (TS) 5.5.6, Inservice Testing Program.

The test established a single acceptance criterion to determine if the discharge check

valve on the 22 ACCP train shuts when the parallel trains 21 ACCP is providing design

flow. The acceptance criterion was that no reverse rotation is observed on the 22

ACCP. Although NUREG-1482, Guidelines for Inservice Testing at Nuclear Power

Plants identifies the methodology of using reverse pump rotation as an acceptable

means of testing, Entergys site-specific experience in 2005 demonstrated this particular

method was not effective to maintain the ACCP discharge check valve safety function.

Specifically, when 2-PT-Q031A was performed on January 19, 2005, the 21 ACCP

failed the performance test because check valve 755A was determined to be in the

open position. However, the 22 ACCP did not rotate in the reverse direction. Following

disassembly of valve 755A, engineers determined the valve remained in the open

position because of excessive clearances between the hinge pin and hinge pin

bushings. Entergy personnel determined the check valve was likely in this condition

following maintenance on the valve in late 2004. CR-IP2-2005-0252 was written to

document and evaluate the issue. The issue was previously documented in LER 05000247/2005001-00 and NRC NCV 50-247/2005003-01. At that time, Entergy

personnel concluded the test criteria established in 2-PT-Q031A was acceptable but

that post-maintenance tests on the check valve should include amplifying comments

Enclosure

23

directing the performance of the IST following maintenance. Entergy personnel

concluded that the IST was adequate because the low pump head that caused the

pump performance test to fail led to troubleshooting that identified that check valve

755A was stuck open.

The inspectors determined that the criterion for determining operability of 755A in test 2-

PT-Q013A was inadequate because the criterion in the procedure previously failed to

identify that 755A remained in the open position in January 2005 and 2-PT-Q013A does

not identify any other criteria, including using pump head, to determine operability of

755A. Additionally, the inspectors determined the test criterion for check valve 755A

and 755B were not consistent with the following ASME Code requirements:

  • The ASME OM Code 2001 Subsection ISTA-3160 states that procedures shall

contain the Owner-specified reference values and acceptance criteria;

  • The ASME OM Code 2001 Subsection ISTC-1400 (c) states it is the Owners

responsibility to ensure that the application, method, and capability of each

nonintrusive technique is qualified; and

  • The ASME OM Code 2001 Subsection ISTC-3530 states obturator movement

shall be determined by exercising the valve while observing an appropriate

indicator.

Analysis. The inspectors determined that the performance deficiency was more than

minor because it was associated with the procedure quality attribute of the Mitigating

System cornerstone and adversely affected the cornerstone objective to ensure the

reliability of systems that respond to initiating events to prevent undesirable

consequences. Specifically, the test criterion used in procedure 2-PT-Q013A did not

ensure that valve 755A reliably performed its safety function when tested as

demonstrated by testing performed in January 2005. The inspectors determined that

the performance deficiency was of very low safety significance (Green) using IMC 0609,

Attachment 4, Phase 1 - Initial Screening and Characterization of Findings.

Specifically, the inspectors determined that this finding was of very low safety

significance because the finding did not result in a loss of safety function and did not

screen as potentially risk-significant due to external events initiating events.

The inspectors determined the finding had a cross-cutting aspect related to effective

corrective actions in the corrective action program component of the problem

identification and resolution area. Specifically, Entergy did not implement effective

corrective actions to resolve the testing inadequacy since 2005 during subsequent

quarterly testing. Additionally, the issue was considered to be indicative of current

performance because personnel when initially responding to inspector questions

concluded the acceptance criteria were adequate. (P.1(d) per IMC 0305)

Enforcement. 10 CFR 50.55a, Codes and standards, states that pumps and valves

which are classified as ASME code Class 1, Class 2, and Class 3 must meet the

inservice test requirements set forth in the ASME OM Code (2001 edition for Indian

Point Unit 2). Furthermore, inservice tests to verify operational readiness of pumps and

valves, whose function is required for safety must comply with the requirements of the

ASME OM Code. The ASME OM Code 2001 Subsection ISTC-1400 (c) states it is the

Owners responsibility to ensure that the application, method, and capability of each

nonintrusive technique is qualified. In addition, the ASME OM Code 2001 Subsection

ISTC-3530 states obturator movement shall be determined by exercising the valve

Enclosure

24

while observing an appropriate indicator. Contrary to the above, from February 2005

until February 2009, Entergy procedure 2-PT-Q031A, did not include appropriate

acceptance criteria for demonstrating operability of valve 755A. Specifically, the test did

not utilize a qualified technique for testing the check-valve and did not verify check valve

movement by observing an appropriate indicator. Because ACCP performance tests

since 2004 demonstrated satisfactory performance of the ACCPs at design flows, no

actual impact to the operability of the ACCPs was evident. Because this violation was

of very low safety significance and it was entered into Entergys corrective action

program (IP2-2009-1312), this violation is being treated as an NCV, consistent with the

NRC Enforcement Policy. NCV 2009002-06, Inadequate Test Acceptance Criteria

for Auxiliary Component Cooling Check Valves.

Cornerstone: Emergency Preparedness (EP)

1EP6 Drill Evaluation (71114.06 - 1 sample)

a. Inspection Scope

The inspectors evaluated an emergency classification conducted on February 23, 2009,

during a licensed-operator requalification simulator training evaluation. The inspectors

observed an operating crew in the simulator respond to various, simulated initiating

events that ultimately resulted in the simulated implementation of the emergency plan.

In particular, the inspectors verified the adequacy and accuracy of the simulated

emergency classification of a Site Area Emergency. While other simulated

classifications were made, the inspectors verified that the initial classification was

appropriately credited as an opportunity toward NRC performance indicator data. The

inspectors observed the management evaluator and training critique following

termination of the scenarios, and verified that significant performance deficiencies were

appropriately identified and addressed within the critique and the corrective action

program. Also, the inspectors reviewed the summary performance report for the

evaluation and verified that appropriate attributes of drill performance including

deficiencies were captured. This evaluation constituted one inspection sample.

b. Findings

No findings of significance were identified.

2. RADIATION SAFETY

Cornerstone: Occupational Radiation Safety (OS)

2OS1 Access Control to Radiologically Significant Areas (71121.01 - 16 samples)

a. Inspection Scope

From March 23 through March 27, 2009, the inspectors conducted the following

activities to verify that Entergy was properly implementing physical, engineering, and

administrative controls for access to high radiation areas, and other radiologically

controlled areas, and that workers were adhering to these controls when working in

these areas. Implementation of the access control program was reviewed against the

Enclosure

25

criteria contained in 10 CFR 20, site technical specifications, and Entergys procedures

required by the Technical Specifications as criteria for determining compliance.

This inspection activity represents completion of sixteen (16) samples relative to this

inspection area. The inspector performed independent radiation dose rate

measurements and reviewed the following items:

Plant Walk Downs and Radiological Work Permit Reviews

(1) Exposure significant work areas were identified by inspectors for review within

radiation areas, high radiation areas, and airborne areas in the plant. Associated

licensee controls and surveys were review for adequacy. Work reviewed

included: Refuel Floor Split Pin and Reactor Head Inspections, Refuel Floor

Lower Internals Removal and Installation, Refuel Floor and Fuel Support Building

Fuel Transport Equipment Repairs requiring an underwater diver, Reactor

Coolant Pump work including RCP #31 Impeller replacement, Containment valve

work including Pressurizer Safety Valves, Various Containment and Auxiliary

Building activities.

(2) With a survey instrument and assistance from a health physics technician,

inspectors walked down the above mentioned areas to determine: whether the

radiation work permits (RWPs), procedures and engineering controls were in

place and whether surveys and postings were adequate.

(3) The inspectors reviewed RWPs that provide access to exposure significant areas

of the plant including high radiation areas. Specified electronic personal

dosimeter alarm set points were reviewed with respect to current radiological

condition applicability and workers were queried to verify their understanding of

plant procedures governing alarm response and knowledge of radiological

conditions in their work area.

(4) There were no radiation work permits for airborne radioactivity areas with the

potential for individual worker internal exposures of >50 mrem CEDE.

(5) There were no internal dose assessments that resulted in actual internal

exposures greater than 50 mrem CEDE. Internal assessments were reviewed to

determine adequacy and assurance that they were not in fact equal to or greater

than 50 mrem CEDE.

Problem Identification and Resolution

(6) Access controls related condition reports were reviewed since the last inspection

in this area. Staff members were interviewed and documents reviewed to

determine that follow-up activities are being conducted in an effective and timely

manner, commensurate with their safety and risk.

(7) For repetitive deficiencies or significant individual deficiencies in problem

identification and resolution, the inspectors determined if the licensees

assessment activities were also identifying and addressing these deficiencies.

(8) A review of events revealed no performance indicator occurrences that involved

dose rates greater than 25 Rem/hour at 30 cm, dose rates greater than

Enclosure

26

500 Rem/hour at 1 meter, or unintended exposures greater than 100 mrem

TEDE (or greater than 5 Rem SDE or greater than 1.5 Rem LDE)

Job-in-Progress Reviews

(9) The inspectors observed aspects of various on-going activities to confirm that

radiological controls, such as required surveys, area postings, job coverage, and

job site preparations were conducted. The inspectors verified that personnel

dosimetry was properly worn and that workers were knowledgeable of work area

conditions. The inspectors attended pre-planning meetings for work described

earlier in the report.

(10) Underwater diving activities associated with repairs to the fuel transport system

were reviewed for adequacy. Dosimetry requirements, bioassay requirements,

and controls were reviewed.

High Risk Significant, High Dose Rate High Radiation Areas (HRA) and Very HRA

Controls

(11) Keys to locked and very HRA were reviewed for their controls and proper

inventory. Accessible locked HRA were verified to be properly secured and

posted during plant tours.

(12) The inspectors discussed with Radiation Protection supervision the adequacy of

high dose rate HRA controls and procedures and verified that no programmatic

or procedural changes have occurred that reduce the effectiveness and level of

worker protection.

Radiation Worker Performance

(13) During observation of the work activities listed above, radiation worker

performance was evaluated with respect to the specific radiation protection work

requirements and their knowledge of the radiological conditions in their work

areas.

(14) The inspectors reviewed condition reports, related to radiation worker

performance to determine if an observable pattern traceable to a similar cause

was evident.

Radiation Protection Technician Proficiency

(15) During observation of the work activities listed above, radiation protection

technician work performance was evaluated with respect to their knowledge of

the radiological conditions, the specific radiation protection work requirements

and radiation protection procedures.

(16) The inspectors reviewed condition reports, related to radiation worker

performance to determine if an observable pattern traceable to a similar cause

was evident.

Enclosure

27

b. Findings

Introduction. The inspectors identified a NCV of very low safety significance (Green)

related to Technical Specification 5.4.1.a, Procedures, because Entergy personnel did

not generate condition reports or investigation paperwork for multiple high dose-rate

alarms as required by station procedures. Specifically, personnel did not generate the

required condition reports and adequately document the investigations for six instances

of unplanned or un-briefed electronic dosimeter alarms received by individuals in the

Unit 2 radiologically controlled area (RCA) that occurred between January 2009 and

March 2009.

Description. During the period January 2009 through March 2009, six instances of

electronic dosimeter dose rate alarms were recorded by the access control system for

Unit 2 personnel in the RCA (Unit 3 had 15 instances). During this period, Entergy

personnel inconsistently utilized an informal process of reviewing the alarms without a

full investigation or approval process. Moreover, in one of the six instances at Unit 2,

the inspectors identified that no investigation or follow-up had occurred. In some cases,

the occurrences were over two months old, which the inspectors noted would have

made resultant investigations more challenging to perform. In other cases, the alarms

were not identified until the worker attempted to re-enter the RCA and the access control

system required manual override to un-lock the occurrence to allow entry into the RCA.

The inspectors noted that the controlling Entergy procedure for this activity, EN-RP-203,

Dose Assessment, specifies that for a dose-rate alarm that is unanticipated or un-

briefed, several actions are required, one of which is to initiate a condition report,

another is to document the investigation using an attachment in the procedure. Contrary

to EN-RP-203, for these 21 instances, no condition reports or attachments were

generated with a detailed investigation prior to the workers re-entering the radiologically

controlled area. The highest exposure received by these workers during their entry, as

indicated by their electronic dosimeter and logged by the access control system, was 33

mRem, while most dosimeters indicated less than 1 mRem for the entry.

Analysis. The inspectors determined that the failure to generate a condition report, as

well as the failure to adequately investigate six unplanned or un-briefed electronic

dosimeter alarms prior to re-entry into the Unit 2 RCA, as required by station procedure

was a performance deficiency. This performance deficiency was within Entergy

personnels ability to foresee and correct, and should have been prevented. This issue

was not subject to traditional enforcement, in that it did not have actual safety

consequence, it was not an issue that had the potential to impact NRCs ability to

perform its regulatory function, and there were no willful aspects.

The finding is more than minor because it is associated with the Occupational Radiation

Safety cornerstone attribute of programs and process, and adversely affected its

objective to ensure adequate protection of worker health and safety from exposure to

radiation. Moreover, the inspectors identified a programmatic deficiency to maintain and

implement programs to keep exposures as low as reasonably achievable, because

multiple examples were identified regarding the failure to satisfy station radiation

protection procedures. Specifically, in six cases, Entergy did not fully evaluate dose rate

alarms received by workers in radiologically controlled areas of the plant. Using the

Occupational Radiation Safety Significance Determination Process, the inspectors

determined that the finding was of very low safety significance (Green) because it did not

involve: (1) as low as is reasonably achievable planning and controls, (2) an

Enclosure

28

overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to

assess dose.

The inspectors determined that the finding had a cross-cutting aspect related to

procedural adherence in the Work Practices component of the Human Performance

area. Specifically, Entergy employees did not follow procedures to generate condition

reports and document investigations when high-dose rate alarms were received by

workers. (H.4 (b) per IMC 0305)

Enforcement. Technical Specification 5.4.1.a, Procedures, requires that Entergy

establish, implement, and maintain procedures specified in Regulatory Guide (RG) 1.33,

Revision 2, Appendix A., Section 7.e, radiation protection procedures for personnel

monitoring. Entergy procedure EN-RP-203, Revision 2, Section 5.11, requires that a

condition report be written for each unplanned or un-briefed electronic dosimeter dose-

rate alarm. Contrary to the above, the inspectors identified through a review of

electronic dosimeter log information from January 2009 through March 2009, six

instances of unanticipated or un-briefed electronic dosimeter dose-rate alarms when the

procedure was not implemented and condition reports were not generated. Because

this finding was of very low safety significance and it was entered into the corrective

action program as CR-IP3-2009-001253 and CR-IP3-2009-001318, this violation is

being treated as an NCV, consistent with the NRC Enforcement Policy. NCV 05000247/2009002-07, Failure to Follow Radiation Protection Procedures.

2OS2 ALARA Planning and Controls (71121.02 - 12 samples)

a. Inspection Scope

From March 23 through March 27, 2009, the inspectors conducted the following

activities to verify that Entergy was properly maintaining individual and collective

radiation exposures as low as is reasonably achievable (ALARA). Implementation of the

ALARA program was reviewed by inspectors against the criteria contained in 10 CFR

20, applicable industry standards, and Entergys procedures.

This inspection activity represents completion of twelve (12) samples relative to this

inspection area.

Inspection Planning

(1) The inspectors reviewed pertinent information regarding cumulative exposure

history, current exposure trends, and on-going activities to assess current

performance and outage exposure challenges. The inspectors determined the

sites 3-year rolling collective average exposure.

(2) The inspectors reviewed unit 3 outage work related activities occurring during the

inspection period, the associated ALARA plans, RWPs, ALARA Committee

Reviews, exposure estimates, actual exposures and post job reviews. Work

reviewed included: Refuel Floor Split Pin and Reactor Head Inspections, Refuel

Floor Lower Internals Removal and Installation, Refuel Floor and Fuel Support

Building Fuel Transport Equipment Repairs requiring an underwater diver,

Reactor Coolant Pump work including RCP #31 Impeller replacement,

Enclosure

29

Containment valve work including Pressurizer Safety Valves, Various

Containment and Auxiliary Building activities.

(3) The inspectors reviewed implementing procedures associated with maintaining

occupational exposures ALARA. This included a review of the processes used to

estimate and track work activity exposures.

Radiological Work Planning

(4) With respect to the work activities listed above, the inspectors reviewed dose

summary reports, related post-job ALARA reviews, related RWPS, exposure

estimates and actual exposures, and ALARA Committee meeting paperwork.

Through this review, the inspector determined that dose was appropriately

managed and evaluated by Station Management.

(5) ALARA work activity evaluations, exposure estimates, and exposure mitigating

requirements were reviewed for work packages previously mentioned. The

inspectors determined that Entergy established procedures, engineering and

work controls, based on sound radiation protection principles.

(6) The inspectors compared the results achieved with the intended dose that was

established in the planning of the work. The inspectors determined the reasons

for any inconsistencies between the intended and actual work activity doses and

station management awareness and involvement.

(7) The inspectors evaluated for adequacy, the interfaces between operations,

radiation protection, maintenance, maintenance planning and others for interface

problems or missing program elements.

Verification of Dose Estimates and Exposure Tracking Systems

(8) Methods for adjusting exposure estimates, or re-planning work, when

unexpected changes in scope or emergent work is encountered, was reviewed

by the inspectors for adequacy.

Job Site Inspections and ALARA Controls

(9) The inspectors reviewed work activities that present the highest radiological risk

to workers. The inspectors evaluated Entergys use of engineering controls to

achieve dose reductions and to verify that procedures and controls are consistent

with ALARA reviews. Associated ALARA Plans and RWPs were reviewed to

determine if appropriate exposure and contamination controls were being

employed.

Radiation Worker Performance

(10) Through observations and interviews, workers and technicians were found to be

knowledgeable of the work area radiological conditions and low dose waiting

areas.

Enclosure

30

Declared Pregnant Workers

(11) The inspectors reviewed information associated with declared pregnant workers

during the assessment period and whether appropriate monitoring and controls

were being utilized to ensure compliance with 10CFR Part 20.

Problem Identification and Resolution

(12) The inspectors reviewed elements of the Entergys corrective action program

related to implementing radiological controls to determine if problems are being

entered into the program for timely resolution.

b. Findings

No findings of significance were identified.

4. OTHER ACTIVITIES [OA]

4OA1 Performance Indicator Verification (71151 - 3 samples)

a. Inspection Scope

The inspectors reviewed performance indicator data for the cornerstones listed below

and used Nuclear Energy Institute 99-02, Regulatory Assessment Performance

Indicator Guideline, Revision 5, to verify individual performance indicator accuracy and

completeness. The documents reviewed during this inspection are listed in the

Attachment.

Initiating Events Cornerstone

  • Unplanned Scrams per 7000 Critical Hours (January 2008 to December 2008)
  • Unplanned Transients per 7000 Critical Hours (January 2008 to December 2008)

The inspectors reviewed data and plant records from January 2008 to December 2008.

The records included PI data summary reports, licensee event reports, operator

narrative logs, Entergys corrective action program, and Maintenance Rule records. The

inspectors verified the accuracy of the number of critical hours reported, and interviewed

the system engineers and operators responsible for data collection and evaluation.

Barrier Integrity Cornerstone

  • RCS Activity (January 2008 to December 2008)

The inspectors reviewed data and plant records from January 2008 to December 2008.

The records included performance indicator data summary reports, licensee event

reports, operator narrative logs, Entergys corrective action program, and Maintenance

Rule records. The inspectors verified the accuracy of the number of critical hours

reported, and interviewed the system engineers and operators responsible for data

collection and evaluation.

Enclosure

31

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems (71152)

.1 Routine Problem Identification & Resolution Program Review

a. Inspection Scope

As required by Inspection Procedure 71152, Identification and Resolution of Problems,

and to identify repetitive equipment failures or specific human performance issues for

follow-up, the inspectors performed a daily screening of all items entered into Entergys

corrective action program. The review was accomplished by accessing Entergys

computerized database for condition reports, and attending condition report screening

meetings.

In accordance with the baseline inspection modules, the inspectors selected corrective

action program items across the Initiating Events, Mitigating Systems, and Barrier

Integrity cornerstones for further follow-up and review. The inspectors assessed

Entergys threshold for problem identification, adequacy of the causal analysis, extent of

condition reviews, and operability determinations, and timeliness of the associated

corrective actions. The condition reports reviewed during this inspection are listed in the

Attachment.

b. Findings

No findings of significance were identified

4OA3 Event Followup

.1 Condensate Return Line Leak on February 15, 2009

a. Inspection Scope

On February 15, 2009, an operator observed indications of wetness in a pipe sleeve in

the floor of the auxiliary feed pump building. The operator notified the control room.

Chemistry samples of the water were drawn and analyzed. On February 16, Entergy

determined the chemistry results indicated the water was from the condensate storage

tank (CST) return line. The inspectors reviewed the technical specifications (TS) to

determine whether operators entered the applicable TS action statements for the CST

and completed required actions to administratively determine the back-up on-site city

water tank was available, if needed, to provide water to the auxiliary feedwater pumps.

The inspectors reviewed Entergys operability evaluation of the CST to determine

whether it was technically supported. In addition, the inspectors reviewed the impact of

the leak on the auxiliary feed water system which utilizes the CST as a primary source of

water and circulates water back to the CST through the CST return piping. The

inspectors also reviewed chemistry and radiological samples taken of the water to assess

the environmental impact of the leak and determine if the release was below NRC

regulatory limits for liquid effluents.

Enclosure

32

b. Findings and Observations

No findings of significance were identified.

Entergy excavated a portion of the CST piping in the area of the identified leakage and

determined that the CST return pipe was leaking due to a hole the pipe where a small

area of a protective coating was missing. Entergy also identified two additional areas of

piping with metal loss that did not exceed ASME Code minimum required wall thickness.

However, the areas were repaired while the opportunity existed. Entergy removed the

portion of pipe with the localized defects and sent the specimen to a laboratory for

analysis to identify the causes. The inspectors determined that the actions Entergy

implemented to evaluate and repair the leaking CST pipe to restore operability to the

CST were adequate and in accordance with their operating license. Additionally, the

inspectors determined that the evaluations and actions Entergy performed to evaluate

and maintain operability of the auxiliary feed pumps were adequate. Entergy analyzed

the water leaking up through the sleeve and determined it was CST water based on

hydrazine and tritium levels. The amount of tritium detected in the water was consistent

with that found in the CST, for example, analyses of samples of water from the leak

returned 2000 - 2300 picocuries per liter (pCi/l). The release was determined to be

below the NRC regulatory limits for liquid effluents. For added perspective, while not

drinking water, the Environmental Protection Agency environmental limit for drinking

water requires tritium levels less than 20,000 pCi/l.

Entergy initiated a root cause analysis to determine causes of the leak that is scheduled

to be completed in May 2009. At the end of the inspection period, the inspectors were

monitoring the performance of Entergy in implementing its corrective action program to

address the issue and develop a root cause evaluation and further corrective actions.

4OA5 Other Activities

.1 Continued Groundwater Sampling Effort to Monitor Tritium (Deviation Memorandum

Inspection)

a. Inspection Scope

During the week of March 23-27, 2009, the inspectors met with Entergy representatives

to review the results of recent groundwater samples, as well as those taken and

analyzed in 2008. The review was conducted against criteria contained in 10CFR20,

10CFR50, and applicable industry standards.

The review of the data included a comparison of Entergys data with split samples taken

by the NRC of monitoring wells MW-66 and MW-67, as well as the LaFarge sample

point. In all, 47 samples were analyzed and compared from January 2008 through

January 2009. Isotopic analyses were performed and compared at each of the sample

points for: Tritium, Strontium 90, Nickel 63, and gamma emitters such as Cobalt-60 and

Cesium-137. Results of the NRC samples can be found in ADAMS accession numbers:

ML081420676, ML082690244, ML082690202, ML082690237, ML082730830,

ML082730810, ML090400523, ML090400516, ML090400502, ML090923932,

ML090920949.

Enclosure

33

Entergy=s evaluation of recent groundwater results are documented in condition reports:

CR-IP2-2009-00883, CR-IP2-2009-01110, CR-IP2-2009-01111, CR-IP2-2009-01113,

and CR-IP2-2009-01114.

b. Findings

No findings of significance were identified.

The inspectors concluded that overall, there was agreement between Entergy

personnels results and those independently analyzed by the NRC, and that actions

taken by Entergy have been appropriate. The inspectors also noted that conservative

estimates indicate that the samples represent a very small fraction of the permissible

public dose limits and are negligible with respect to natural background radiation levels.

.2 Quarterly Resident Inspector Observations of Security Personnel and Activities

a. Inspection Scope

During the inspection period, the inspectors conducted observations of security force

personnel and activities to ensure that these activities were consistent with Entergy

security procedures and applicable regulatory requirements. Although these

observations did not constitute additional inspection samples, the inspections were

considered an integral part of the normal, resident inspector plant status reviews during

implementation of the baseline inspection program.

b. Findings

No findings of significance were identified.

4OA6 Meetings

Exit Meeting Summary

On April 15, 2009, the inspectors presented the inspection results to Joe Pollock and

other Entergy staff members, who acknowledged the inspection results presented.

Entergy did not identify any material as proprietary.

ATTACHMENT: SUPPLEMENTAL INFORMATION

Enclosure

A-1

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Entergy Personnel

J. Pollock, Site Vice President

A. Vitale, General Manager, Plant Operations

P. Conroy, Director of Nuclear Safety Assurance

A. Williams, Site Operations Manager

B. Sullivan, Emergency Planning Manager

S. Verrochi, System Engineering Manager

R. Walpole, Licensing Manager

D. Loope, Manager, Radiation Protection

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000247/2009002-01 FIN Failure to Identify Open Louvers in 11 Fire

Pump House (Section 1R01)05000247/2009002-02 NCV Failure to Identify Damaged Components in

EDG Ventilation Motor Control Center #2

(Section 1R05)05000247/2009002-03 NCV Failure to identify and Promptly Correct

Degraded 480 Volt Switchgear Room Fire

Door (Section 1R05)05000247/2009002-04 NCV Inadequate Maintenance Procedure for

EDG Ventilation Motor Control Center #2

(Section 1R12)05000247/2009002-05 NCV Failure to Include RWST Level

Maintenance In Online Risk Assessment

(Section 1R13)05000247/2009002-06 NCV Inadequate Test Acceptance Criteria for

Auxiliary Component Cooling Check Valves

(Section 1R22)05000247/2009002-07 NCV Failure to Follow Radiation Protection

Procedures (Section 2OS1)

Attachment

A-2

LIST OF DOCUMENTS REVIEWED

Section 1R01: Adverse Weather Protection

Procedures

OAP-048, Rev. 4, Seasonal Weather Preparation

OAP-008, Rev. 5, Severe Weather Preparations

2-AOP-SSD-1, Rev. 13, Control Room Inaccessibility Safe Shutdown Control

OAP-017, Rev. 5, Plant Surveillance and Operator Rounds

EN-OP-115, Rev. 5, Conduct of Operations

Condition Reports

IP2-2009-00197 IP2-2009-00207 IP2-2009-00208 IP2-2009-00211

IP2-2009-00212 IP2-2009-00214 IP2-2009-00215 IP2-2009-00226

Orders

00152922 00153082 00153083 00179583

Section 1R04: Equipment Alignment

Procedures

2-PT-M103, Rev. 2, Auxiliary Feedwater System Monthly Alignment Verification

2-COL-4.1.1, Rev. 22, Component Cooling System

Section 1R05: Fire Protection

Procedures

SAO-703, Rev. 25, Fire Protection Impairment Criteria and Surveillance

EN-DC-161, Rev. 2, Control of Combustibles

OAP-037, Rev. 2, Operations Electrical Equipment Operating Guidelines

IP-SMM-IS-103, Rev. 0, IPEC Site Management Manual Electrical Safety

2-PT-SA020, Rev. 0, Swing Fire Doors

Condition Reports

IP2-2009-00904 IP2-2009-00526 IP2-2009-00680 IP2-2009-00709

IP2-2009-00834 IP2-2009-00342 IP2-2009-00483 IP2-2004-05336

IP2-2007-03561 IP2-2007-04645 IP2-2008-05447

Orders

51645822 51676572

Miscellaneous

Indian Point Nuclear Generating Station, Unit 2, Fire Protection Program Plan, Rev. 9

Indian Point Pre-Fire Plans Unit 2 - Nuclear

IP2-RPT-03-00015, IP2 Fire Hazards Analysis, Rev. 3

1R07: Heat Sink Performance

Procedures

SEP-SW-001, NRC Generic Letter 89-13 Service Water Program

PT-2Y10B, 22 CCW HX Test

2-HTX-004-CCW, Component Cooling Water Heat Exchanger Maintenance

Attachment

A-3

Work Orders

51675733

Condition Reports

IP2-2005-0673 IP2-2005-0768 IP2-2005-1268 IP2-2006-7126

IP2-2006-3974

Miscellaneous

EPRI NP-7552, Heat Exchanger Performance Monitoring Guidelines

Preliminary Report of Eddy Current Testing dated 2/10/09

21 CCW Hx Inspection Reports dated 2/23/2005 and 1/8/2007

22 CCW Hx Inspection Reports dated 2/23/2005 and 12/12/2006

Section 1R11: Licensed Operator Requalification Program

Procedures

OAP-033, Conduct of Operations Simulator Training, Evaluations, and Debriefs, Rev. 4

OAP-032, Operations Training Program, Rev. 9

2-E-0, Rev. 0, Reactor Trip or Safety Injection

2-ECA-0.0, Rev. 3, Loss of All AC Power

2-AOP-480V-1, Rev. 5, Loss of Normal Power to any 480V Bus

Miscellaneous

LRQ-SES-21, Rev. 0, IPEC Evalauted Scenario for Loss of All AC Power

Section 1R12: Maintenance Effectiveness

Procedures

2-MCC-003-ELC, Rev 0, Klockner-Moeller, Series 200, 480 Volt Motor Control Center

Preventive Maintenance

2-PC-Q2, Rev. 19, Refueling Water Storage Tank Level

0-MS-412, Rev. 0, Inspection and Cleaning of Bus Bars, Contacts, Ground Connections, Wiring

and Insulators

IP-SMM-IS-103, Rev. 0, IPEC Site Management Manual Electrical Safety

0-GNR-404-ELC, Rev. 1, Emergency Diesel Generator 2-Year Inspection

2-GNR-015-ELC, Rev. 2, Emergency Diesel Generator Preventive Maintenance 2-Year

2-PT-M021B, Rev. 17, Emergency Diesel Generator 22 Load Test

Condition Reports

IP2-2009-00527 IP2-2009-00532 IP2-2009-01041 IP2-2003-00948

IP2-2009-00342 IP2-2009-00483 IP2-2004-03106 IP2-2007-01893

IP2-2008-05382 IP2-2009-00486 IP2-2009-00041 IP2-2009-00178

IP2-2006-04101 IP2-2009-00093 IP2-2007-03476 IP2-2007-04921

IP2-2008-00454 IP2-2008-00907 IP2-2008-03976

Orders

51557262 51676147 06-16146 51696697 51322921 51268313

00181009 00167536 04-26645 57696714 51649505 51654261

00118733 07-03476 07-04921 08-00454 08-00907 09-00532

Drawing

309030-02, Loop diagram RWST level indication

3WS-463-610-14-20101-3, Schematic for EDG HVAC Heater

Attachment

A-4

IP2-S-000231-04, Schematic for EDG Building Ventilation Distribution

B248513-12, 480V MCC 26C and CCR Ventilation Distribution

B228434-02, Class A Boundary for Electrical Systems

Miscellaneous

Maintenance Rule Basis Document Residual Heat Removal System, dated 5/23/05

Maintenance Rule Basis Document HVAC Emergency Diesel Building, dated 5/23/05

IP-SMM-AD-102, Att 10.2, dated 4/6/08, for revision to procedure 2-MCC-003-ELC

Vendor Manual, Klockner-Moeller Series 200 Motor Control Center

Vendor Manual, Qmark MUH Series Modular Unit Heaters

Vendor Manual, ALCO Fuel Injection Nozzle and Holder

Maintenance Rule Expert Panel Meeting Minutes dated 2/14/05

Tagout 2-480V-Panel-MCC26C dated 4/3/08

DRN-08-01336 dated 4/6/08 for procedure 2-MCC-003-ELC

PMCR ER-06-33534, to establish maintenance activity for EDG HVAC MCC

Section 1R13: Maintenance Risk Assessments and Emergent Work Control

Procedures

IP-SMM-WM-101, On-Line Risk Assessment

2-PC-Q109, Recalibration of Nis and OT/OP delta T parameters

PT-Q17A, Verify ASSS supply to 21 AFP

2-PT-Q027A, 21 Auxiliary Feed Pump

2-PC-Q2, Rev. 19, Refueling Water Storage Tank Level

2-ES-1.3, Rev. 2, Transfer to Cold Leg Recirculation

Condition Reports

IP2-2009-00018 IP2-2009-00027 IP2-2009-00139 IP2-2009-00143

IP2-2009-00148 IP2-2009-00389

Work Orders

00165604 51654961 51692571 51692351 51696697

Miscellaneous

Equipment Out-Of-Service (EOOS) risk assessment reports

Section 1R15: Operability Evaluations

Procedures

2-PT-Q031A, 21 Auxiliary Component Cooling Pump

2-PT-Q031B, 22 Auxiliary Component Cooling Pump

EN-MA-133, Control of Scaffolding

2-AOP-IB-1, Loss of Power to an Instrument Bus

2-PT-M021B, Rev. 17, Emergency Diesel Generator 22 Load Test

2-SOP-AFW-002, Rev. 1, Auxiliary Feedwater System Operation Support Procedure

Drawings

A249955-21, 480V AC MCC 29 & 29A

Calculation

IP3-CALC-FW-01482, Rev. 0, Feedwater Stratification and Auxiliary Feedwater

Attachment

A-5

Condition Reports

IP2-2009-0500 IP2-2009-0505 IP2-2008-3749 IP2-2009-0547

IP2-2009-0567 IP2-2009-0509 IP2-2005-0252 IP2-2009-0552

IP2-2009-0655 IP2-2008-2705 IP2-2009-0041 IP2-2009-0093

Work Orders

NP-99-07694

Miscellaneous

WCAP-12312, Rev. 2, Safety Evaluation for an Ultimate Heat Sink Temperature Increase to 95F

at Indian Point Unit 2

Heat exchanger data sheet for containment recirculation pump number 22 motor cooler

WCAP-7829, Fan Cooler Motor Unit Test

Environmental Qualification Report for Containment Recirculation Pump Motors

IP2-CCW-DBD, Component Cooling Water design bases document

IP2-DBD-207, Design Basis Document for 118V AC Electrical System

AMSE OM-2001 Edition

Unit 2 active scaffold list

VM 1073-1.2, Vendor manual for auxiliary component cooling pumps

VM 1100, vendor manual for 118V AC solid state static inverters

Work order NP-89-43777, replacement of 22 ACCP impeller

IP2-AFW-DBD, Rev. 1, AFW Design Basis Document

Section 1R18: Plant Modifications

Procedures

2-SOP-18-1, Main and Reheat Steam System

TP-SQ-11.016, Post Work Test Program (historical)

Condition Reports

IP2-2009-0983 IP2-2009-0137 IP2-2008-5636 IP2-2009-0077

IP2-2009-0069 IP2-2009-0062 IP2-2008-5621 IP2-2009-0781

Work Orders

IP2-03-11725 IP2-02-32013 51305160

Drawings

B235623-6, Atmospheric Steam Dump Panel

9321-F-70313, Auxiliary Boiler Feed Pump Room Instrument Piping

Miscellaneous

IP2 Maintenance Rule Basis for Main Steam System

IP2-MS-DBD, Design Basis Document for the Main Steam System

IPT-RPT-05-00071, Appendix R Safe Shutdown Analysis

SEE-03-5, Indian Point Unit 2 RHR Cooldown Analysis for the 5% Power Uprate

IP2 Inservice Testing Program Basis Data Sheets for PCV-1136 & 1137 (23/24 SG ADVs)

ER 06-2-012, Install Secondary Backup Nitrogen Cylinders at both S/G ADV Local Control

Panels in the ABFP Building

Attachment

A-6

Section 1R19: Post-Maintenance Testing

Procedures

OAP-24, Operations Testing, Rev. 3

2-PT-M021C, Rev. 16, Emergency Diesel Generator 23 Load Test

0-GNR-403-ELC, Emergency Diesel Generator Quarterly Inspection

2-PT-Q033B, 21 Charging Pump

2-SOP-4.1.2, Rev. 34, Component Cooling System Operation

Orders

51797559 51797558 52027651 00183296 00157710 51675732

Section 1R22: Surveillance Testing

Procedures

2-PT-2M4, Safety Injection System Train A Actuation Logic and Master Relay Test

2-PT-Q013, Inservice Valve Tests

2-PT-Q013-DS027, Valve 888A IST Data Sheet

0-SOP-LEAKRATE-001, Rev. 1, RCS Leakrate Surveillance, Evaluation and Leak Identification

2-PT-Q030C, Rev. 18, 23 Component Cooling Water Pump

Drawings

11497, Valve 888A

Condition Reports

IP2-2007-1754 IP2-2008-1443 IP2-2008-2002 IP2-2007-3329

Orders

51694305

Miscellaneous

IP2-ESF DBD, Design Basis Document for Engineered Safeguards Features System

IP2 Inservice Testing Program Data Sheet - Valve 888A

PGI-00066-01, 888 A & B Diff Pr Calc

Section 1EP6: Drill Evaluation

Procedures

IP-EP-120, Rev. 3, Emergency Classification

Miscellaneous

IP-EP-115, Rev. 24, form EP-1 radiological emergency data forms dated 2/23/09

Section 2OS1: Access Control to Radiologically Significant Areas and

Section 2OS2: ALARA Planning and Controls

Procedures

EN-RP-100, Rev. 03, Radworker Expectations

EN-RP-101, Rev. 04, Access Control for Radiologically Controlled Areas

EN-RP-102, Rev. 02, Radiological Control

EN-RP-105, Rev. 04, Radiation Work Permits

EN-RP-108, Rev. 07, Radiation Protection Posting

EN-RP-110, Rev. 05, ALARA Program

Attachment

A-7

EN-RP-121, Rev. 04, Radioactive Material Control

EN-RP-131, Rev. 06, Air Sampling

EN-RP-141, Rev. 04, Job Coverage

EN-RP-151, Rev. 02, Radiological Diving

EN-RP-202, Rev. 06, Personnel Monitoring

EN-RP-203, Rev. 02, Dose Assessment

EN-RP-204, Rev. 02, Special Monitoring Requirements

EN-RP-205, Rev. 02, Prenatal Monitoring

EN-RP-208, Rev. 02, Whole Body Counting and In-Vitro Bioassay

Condition Reports

CR-IP3-2009-00752, CR-IP3-2009-00785, CR-IP3-2009-00857, CR-IP3-2009-00885

CR-IP3-2009-00886, CR-IP3-2009-00937, CR-IP3-2009-00998, CR-IP3-2009-01006

CR-IP3-2009-01107, CR-IP3-2009-01154, CR-IP3-2009-01169, CR-IP3-2009-01171

CR-IP3-2009-01183, CR-IP3-2009-01253, CR-IP3-2009-01293, CR-IP3-2009-01295

CR-IP3-2009-01296, CR-IP3-2009-01318, CR-IP2-2009-00883, CR-IP2-2009-01110,

CR-IP2-2009-01111, CR-IP2-2009-01113, CR-IP2-2009-01114

Miscellaneous

Radiation Protection Attention Logs (Electronic Dosimeter Alarms)

TEDE ALARA Evaluations

ALARA Committee Reviews

RP-STD-XX, Rev. X, Unreported Dosimeter Alarms and Anomolies (Draft)

IPEC Snapshot Self-Assessment Report (IP3-LO-2007-0010) July 2007 - June 2008.

RWPs: 2009-002, 2009-003, 2009-2021, 2009-3001, , 2009-3002, 2009-3056, 2009-3501,

2009-3504, 2009-3515, 2009-3529

Section 4OA1: Performance Indicator Verification

EN-EP-201, "Performance Indicators," Rev. 6

EN-LI-114, Performance Indicator Process, Rev. 3

NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Rev. 5

0-CY-2765, Rev. 3, Coolant Activity Limits

Section 4OA2: Identification and Resolution of Problems

Procedures

EN-LI-102, Rev. 13, Corrective Action Process

Condition Reports

IP2-2009-00342 IP2-2009-00483 IP2-2004-03106 IP2-2007-01893

IP2-2008-05382 IP2-2009-00486 IP2-2009-00027 IP2-2009-00139

IP2-2009-00143 IP2-2009-00148

Attachment

A-8

LIST OF ACRONYMS

ALARA as low as is reasonably achievable

ABFW auxiliary boiler feedwater

ABFP auxiliary boiler feedwater pump

ACCP auxiliary component cooling pump

ADAMS Agency-wide Document and Management System

ASME American Society of Mechanical Engineers

CAP corrective action program

CCW component cooling water

CDF core damage frequency

CFR Code of Federal Regulations

CST condensate storage tank

EDO Executive Director of Operations

EDG emergency diesel generator

ENTERGY Entergy Nuclear Northeast

EP Emergency Preparedness

HRA high radiation area

IMC Inspection Manual Chapter

IPEC Indian Point Energy Center

IST in-service test

MCC motor control center

NCV non-cited violation

NDE non-destructive examination

NRC Nuclear Regulatory Commission

NRR Nuclear Reactor Regulation

NSR non safety-related

PARS Publicly Available Records System

PI performance indicator

RCA radiologically controlled area

RCS reactor coolant system

RWP radiation work permit

RWST refueling water storage tank

SDP significance determination process

SER safety evaluation report

SG steam generator

SR safety related

SSC structures, systems, and components

TS Technical Specification

UFSAR Updated Final Safety Evaluation Report

URI unresolved item

WO work order

Attachment

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