ML12090A767

Entergy Pre-Filed Hearing Exhibit ENT000427, Letter from M. Gray, NRC, to J. Pollock, Entergy, Indian Point Nuclear Generating Unit 2;NRC Integrated Inspection Report 05000247/2009002

ML12090A767
Person / Time
Site:	Indian Point
Issue date:	05/14/2009
From:	Mel Gray Reactor Projects Branch 2
To:	Joseph E Pollock Entergy Nuclear Operations, Atomic Safety and Licensing Board Panel
SECY RAS
Shared Package
ML12090A727	List: ML12090A728 ML12090A729 ML12090A730 ML12090A731 ML12090A732 ML12090A733 ML12090A734 ML12090A735 ML12090A736 ML12090A738 ML12090A739 ML12090A740 ML12090A742 ML12090A744 ML12090A745 ML12090A746 ML12090A747 ML12090A748 ML12090A750 ML12090A751 ML12090A752 ML12090A754 ML12090A755 ML12090A756 ML12090A757 ML12090A758 ML12090A760 ML12090A762 ML12090A763 ML12090A764 ML12090A766 ML12090A767 ML12090A768 ML12090A771
References
RAS 22156, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01 IR-09-002
Download: ML12090A767 (45)

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.

ENT000427

Submitted: March 30, 2012

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)

S. Collins, RA

M. Dapas, DRA

D. Lew, DRP

J. Clifford, DRP

M. Gray, DRP

B. Bickett, DRP

A. Rosebrook, DRP

S. McCarver, DRP

J. Heinly, DRP

G. Malone, DRP, SRI, IP2

C. Hott, DRP, RI, IP2

D. Hochmuth, DRP, OA

S. Campbell, RI OEDO

R. Nelson, NRR

M. Kowal, NRR

J. Boska, PM, NRR

J. Hughey, NRR

D. Bearde, DRP

ROPreports@nrc.gov

Region I Docket Room (w/concurrences)

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

Enclosure

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

2

Enclosure

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

3

Enclosure

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)

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 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).

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)

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

6

Enclosure

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

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.

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.

8

Enclosure

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.

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 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.

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

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.

11

Enclosure

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.

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

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

13

Enclosure

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.

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.

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

15

Enclosure

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;

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

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.

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

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

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

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:

Emergent maintenance on the 22 EDG lube oil pump during the 23 EDG

maintenance outage;

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

protection system testing;

Unplanned elevated risk condition due to delayed work on reactor protection system

components during planned maintenance of 22 ABFW pump;

Planned maintenance on a reactor water storage tank level indicator;

Planned maintenance on the 22 ABFW pump while temporary modifications were

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

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 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

20

Enclosure

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;

Impact on pressurizer surge line and reactor coolant system piping while performing

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.

21

Enclosure

.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;

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

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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 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

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;

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

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

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

25

Enclosure

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

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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)

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.

27

Enclosure

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

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Enclosure

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,

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Enclosure

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.

30

Enclosure

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.

31

Enclosure

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.

32

Enclosure

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.

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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

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

A-1

Attachment

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)

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

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

A-3

Attachment

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

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

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

A-5

Attachment

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

A-6

Attachment

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

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

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

A-8

Attachment

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