ML060390411

IR 05000247-05-005; 10/01/2005 - 12/31/2005; Indian Point Nuclear Generating Unit 2; Maintenance Rule; Maintenance Risk Assessment and Emergent Work; Emergency Planning; Problem Identification and Resolution

ML060390411
Person / Time
Site:	Indian Point
Issue date:	02/07/2006
From:	Brian Mcdermott Reactor Projects Branch 2
To:	Dacimo F Entergy Nuclear Operations
McDermott, B J, RGN-I/DRP/610-337-5233
References
FOIA/PA-2016-0148 IR-05-005
Download: ML060390411 (49)

See also: IR 05000247/2005005

Text

February 7, 2006

Mr. Fred R. Dacimo

Site Vice President

Entergy Nuclear Operations, Inc.

Indian Point Energy Center

295 Broadway, Suite 1

P.O. Box 249

Buchanan, NY 10511-0249

SUBJECT: INDIAN POINT NUCLEAR GENERATING UNIT 2 - NRC INTEGRATED

INSPECTION REPORT NO. 05000247/2005005

Dear Mr. Dacimo:

On December 31, 2005, the U.S. Nuclear Regulatory Commission (NRC) completed an

inspection at the Indian Point Nuclear Generating Unit 2 (IP2). The enclosed integrated

inspection report documents the inspection findings, which were discussed on

January 11, 2006, with Mr. Paul Rubin 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. Within these areas, the inspection consisted of a selected examination of procedures

and representative records, observations of activities, and interviews with personnel.

Based on the results of the inspection, six findings were identified. Four of these findings were

determined to be violations of NRC requirements, including one finding that was determined to

be a Severity Level IV violation. However, because of the very low safety significance, and

because they were entered into your corrective action program, the NRC is treating these four

findings as non-cited violations (NCVs) consistent with Section VI.A of the NRC Enforcement

Policy. If you contest the NCVs in this report, you should provide a 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; and the NRC Senior

Resident Inspector at Indian Point 2.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its

enclosure will be available electronically for public inspection in the NRC Public Document

Room or from the Publicly Available Records (PARS) component of the NRCs document

Mr. Fred R. Dacimo 2

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/

Brian J. McDermott, Chief

Projects Branch 2

Division of Reactor Projects

Docket No. 50-247

License No. DPR-26

Enclosure: Inspection Report No. 05000247/2005005

w/Attachment: Supplemental Information

G. J. Taylor, Chief Executive Officer, Entergy Operations

M. R. Kansler, President, Entergy Nuclear Operations Inc. (ENO)

J. T. Herron, Senior Vice President and Chief Operations Officer (ENO)

C. Schwarz, Vice President, Operations Support (ENO)

P. Rubin, General Manager Operations (ENO)

O. Limpias, Vice President, Engineering (ENO)

J. McCann, Director, Licensing (ENO)

C. D. Faison, Manager, Licensing (ENO)

M. J. Colomb, Director of Oversight (ENO)

J. Comiotes, Director, Nuclear Safety Assurance (ENO)

P. Conroy, Manager, Licensing (ENO)

T. C. McCullough, Assistant General Counsel, Entergy Nuclear Operations, Inc.

P. R. Smith, President, New York State Energy, Research and Development Authority

P. Eddy, Electric Division, New York State Department of Public Service

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

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

B. Brandenburg, Assistant General Counsel

Assemblywoman Sandra Galef, NYS Assembly

County Clerk, Westchester County Legislature

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

Mr. Fred R. Dacimo 3

M. Elie, Citizens Awareness Network

D. Lochbaum, Nuclear Safety Engineer, Union of Concerned Scientists

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

Congresswoman Sue W. Kelly

Congresswoman Nita Lowey

Senator Hillary Rodham Clinton

Senator Charles Schumer

J. Riccio, Greenpeace

A. Matthiessen, Executive Director, Riverkeeper, Inc.

M. Kaplowitz, Chairman of County Environment & Health Committee

A. Reynolds, Environmental Advocates

M. Jacobs, Director, Longview School

D. Katz, Executive Director, Citizens Awareness Network

P. Leventhal, The Nuclear Control Institute

K. Coplan, Pace Environmental Litigation Clinic

W. DiProfio, PWR SRC Consultant

D. C. Poole, PWR SRC Consultant

W. Russell, PWR SRC Consultant

W. Little, Associate Attorney, NYSDEC

R. Christman, Manager Training and Development

Mr. Fred R. Dacimo 4

Distribution w/encl: (via E-mail)

S. Collins, RA

M. Dapas, DRA

S. Lee, RI OEDO

R. Laufer, NRR

J. Boska, PM, NRR

P. Tam, PM (backup)

B. McDermott, DRP

D. Jackson, DRP

C. Long, DRP

M. Cox, DRP, Senior Resident Inspector - Indian Point 2

G. Bowman, DRP, Resident Inspector - Indian Point 2

R. Martin, DRP, Resident OA

Region I Docket Room (w/concurrences)

ROPreports@nrc.gov

DOCUMENT NAME:E:\Filenet\ML060390411.wpd

SISP REVIEW COMPLETE ________DEJ________ (Reviewers initials)

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 MCox/DEJ for DJackson/DEJ BMcDermott/BJM

DATE 02/07 /06 02/07/06 02/07/06

OFFICIAL RECORD COPY

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket No. 50-247

License No. DPR-26

Report No. 05000247/2005005

Licensee: Entergy Nuclear Northeast

Facility: Indian Point Nuclear Generating Unit 2

Location: 295 Broadway, Suite 3

Buchanan, NY 10511-0308

Dates: October 1, 2005 - December 31, 2005

Inspectors: M. Cox, Senior Resident Inspector, IP2

T. Hipschman, Senior Resident Inspector, IP3

G. Bowman, Resident Inspector, IP2

C. Long, Acting Resident Inspector, IP2

S. Barr, Senior Operations Engineer, Region I

T. Fish, Senior Operations Engineer, Region I

R. Fuhrmeister, Senior Project Engineer, Region I

D. Jackson, Senior Project Engineer, Region I

R. Kahler, Senior Emergency Preparedness Specialist, NSIR

J. Noggle, Senior Health Physicist, Region I

D. Silk, Senior Emergency Preparedness Inspector, Region I

D. Johnson, Reactor Inspector, Region I

T. Sicola, Reactor Inspector, Region I

Approved by: Brian J. McDermott, Chief

Projects Branch 2

Division of Reactor Projects

i Enclosure

TABLE OF CONTENTS

SUMMARY OF FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii

Summary of Plant Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

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

1R01 Adverse Weather . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

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

1R05 Fire Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

1R06 Flood Protection Measures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

1R07 Heat Sink Performance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

1R11 Operator Requalification Inspection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

1R12 Maintenance Effectiveness . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6

1R13 Maintenance Risk Assessments and Emergent Work Control . . . . . . . . . . . . . . 9

1R14 Personnel Performance During Non-routine Plant Evolutions and Events . . . 13

1R15 Operability Evaluations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14

1R16 Operator Workarounds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14

1R19 Post-Maintenance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15

1R22 Surveillance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

1R23 Temporary Plant Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

1EP2 Alert and Notification System Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17

1EP3 Emergency Response Organization (ERO) Augmentation Testing . . . . . . . . . 17

1EP4 Emergency Action Level (EAL) and Emergency Plan Changes . . . . . . . . . . . 18

1EP5 Correction of Emergency Preparedness Weaknesses and Deficiencies . . . . . 20

1EP6 Drill Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20

4. OTHER ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21

4OA1 Performance Indicator Verification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21

4OA2 Identification and Resolution of Problems . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22

4OA4 Cross-Cutting Aspects of Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27

4OA6 Meetings, including Exit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27

ATTACHMENT: SUPPLEMENTAL INFORMATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27

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

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . A-2

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

LIST OF ACRONYMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-11

ii Enclosure

SUMMARY OF FINDINGS

IR 05000247/2005005; 10/01/2005 - 12/31/2005; Indian Point Nuclear Generating Unit 2;

Maintenance Rule; Maintenance Risk Assessment and Emergent Work; Emergency Planning;

Problem Identification and Resolution.

The report covers a 3-month period of inspection by resident inspectors, 8 regional inspectors,

and one inspector from the NRCs Office of Nuclear Security and Incident Response. Six

findings were identified, four of which were non-cited violations (NCVs), including one that was

determined to be a Severity Level IV violation. The significance of most findings is indicated by

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

Significance Determination Process, (SDP). Findings for which the SDP does not apply may

be Green or be assigned a severity level after NRC management reviews. The NRCs program

for overseeing the safe operation of commercial nuclear power reactors is described in

NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.

A. NRC Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

• Green

This finding greater than minor because it is associated with the Mitigating

Systems cornerstone attribute of Equipment Performance, and affected the

cornerstone objective of ensuring the availability, reliability, and capability of

systems that respond to initiating events to prevent undesirable consequences.

Specifically, the reliability of the control rod drive mechanism fans, which are

to cool the control rod drive mechanisms during normal operation and

are used in emergency operating procedures to prevent void formation in the

reactor head region during natural circulation cool down, was adversely affected.

This finding is of very low safety significance because while equipment reliability

was degraded, there was no actual loss of function

. (Section 1R12)

• Green. The NRC identified a Green NCV of Technical Specification 5.4.1

. Specifically,

iii Enclosure

. Entergy entered this issue into the

corrective action program and took action to revise their work control procedure

to modify their definition of emergency work.

This finding greater than minor, because if left uncorrected it would become a

more significant safety concern. Failure to complete required prior to

work on safety-related equipment could impact the operability of risk-significant

components.

This finding is of very low

safety significance, because the safety-related work performed without an

approved evaluation did not result in the actual loss of safety function of a

system and did not impact fire, flooding, seismic, or severe weather initiating

events.

(Section 1R13)

Cornerstone: Barrier Integrity

• Green. The NRC identified a Green NCV of 10 CFR 50, Appendix B, Criterion V,

Instructions, Procedures, and Drawings,

to FCV-447, the safety-related

feedwater flow control valve to the 24 steam generator. Specifically, while

implementing a modification to grind material from the valve actuator cap screw

heads, maintenance personnel removed more material than allowed by the

modification package.

Entergy entered

this issue into the corrective action program and completed an operability

assessment to show that remained operable.

This finding is greater than minor because it is associated with the Barrier

Integrity cornerstone attribute of Barrier Performance, and affected the

cornerstone objective of ensuring the availability and reliability of components

used for containment isolation. Improper implementation of this modification

could have resulted in the inability of this valve to perform its safety function.

This finding is of very low safety significance because while the modification was

incorrectly implemented, subsequent analysis showed that the valve remained

operable.

(Section 1R13)

iv Enclosure

Cornerstone: Emergency Preparedness

This finding is greater than minor because it is associated with the Emergency

Preparedness cornerstone attribute of Facilities and Equipment, and affected the

cornerstone objective of ensuring that the licensee is capable of implementing

adequate measures to protect the health and safety of the public in the event of

a radiological emergency. The deficiency is not greater than Green because it

did not result in the Risk-Significant Planning Standard Function being lost or

degraded. Section 4.4 of Manual Chapter 0609, Appendix B, provides examples

for use in assessing emergency preparedness related findings. One example of

a Green finding states, The EAL classification process would not declare any

Alert or Notification of Unusual Event that should be declared. Since the

declaration of an UE based on low service water bay level could have been

missed or delayed, this finding was considered consistent with the example

provided and was therefore determined to be of very low safety significance

(Green). Because this issue is of very low safety significance and has been

entered into Entergys corrective action program, it is being treated as an NCV.

(Section 1EP4)

• Green. The inspectors identified a Green finding for a failure to implement timely

corrective actions for multiple frame relay system problems dating back to 2003.

Specifically, for issues related to the reliability of the frame relay system,

adequate actions to prevent recurrence were not implemented in a timely

manner. Entergys corrective actions in response to the August 2005 frame

relay failures resulted in a more thorough assessment of this issue and

reasonable actions to prevent recurrence.

This finding was determined to be more than minor because it is associated with

the Emergency Preparedness cornerstone attribute of Facilities and Equipment.

It affected the cornerstone objective of ensuring that the licensee is capable of

implementing adequate measures to protect the health and safety of the public in

the event of a radiological emergency. This finding is not suitable for

Significance Determination Process evaluation but has been reviewed by NRC

management and is determined to be a finding of very low safety significance.

This issue is not greater than Green, because of the short periods that the frame

relay system was unavailable and, because the alert and notification system

design included a secondary method (i.e., back-up radio system) to actuate the

sirens. (Section 4AO2)

v Enclosure

• Severity Level IV. A Severity Level IV violation of 10 CFR 50.72(b)(3)(xiii) was

identified for not formally reporting a siren system problem that occurred on

August 5, 2005. The inspectors noted that the duration of the siren system

problem was short, the NRC was informally notified, the process for back-up

route alerting was available, and the capability to actuate the sirens via a manual

siren initiation method was not lost. Subsequent to this event, Entergy

implemented corrective actions to formalize the manual siren system actuation

method. Notwithstanding these circumstances, a formal notification to the NRC

was required, because the normal processes for actuation of the sirens were not

available and Entergy did not have formal procedures for, and had limited

experience with, method.

This deficiency was evaluated using the traditional enforcement process since

the failure to make a required report could adversely impact the NRCs ability to

carry out its regulatory mission.

(Section 4OA2)

B. Licensee Identified Violations.

None.

vi Enclosure

Report Details

Summary of Plant Status

Indian Point 2 (IP2) began the inspection period at full power and operated at or near full power

until December 22. On December 22, power was reduced to approximately 3 percent and the

main turbine was taken off-line to repack FCV-447, the feedwater regulating valve to the 24

steam generator (S/G). Work was completed and the plant was returned to full power on

December 23. The plant remained at full power for the remainder of the inspection period.

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R01 Adverse Weather

a. Inspection Scope (71111.01 - 1 sample of system-related weather preparations)

The inspectors reviewed Entergys administrative controls and implementation of a

maintenance program to ensure adequate protection of safety-related water sources

from freezing conditions. These systems were selected because their safety-related

functions could be affected by adverse weather. Specifically, the inspectors reviewed

Entergys strategy for coping with cold weather effects on the condensate storage tank

(CST), the primary water storage tank (PWST), and the refueling water storage tank

(RWST). The inspectors also reviewed work orders (WOs) and condition reports (CRs)

associated with these external tanks which had the potential to impact cold weather

performance. In addition, the inspectors walked down the accessible areas of piping

and instrumentation to evaluate the insulation and heat tracing material condition. The

specific information reviewed is listed in the attachment. Cumulatively, this inspection of

selected tanks and support systems constituted one inspection sample.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment

.1 Partial System Walkdown

a. Inspection Scope (71111.04Q - 3 samples)

The inspectors performed three 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

and drawings in order to verify that the alignment of the available train was proper to

support its required safety functions. The inspectors also reviewed applicable CRs and

WOs to assure that Entergy had identified and properly addressed equipment

discrepancies that could potentially impair the capability of the available train.

Enclosure

2

Referenced documents are listed in the attachment at the end of this report. The

following system walkdowns were counted as inspection samples:

• Gas Turbine (GT) 3 with GT-1 Out of Service for Preventative Maintenance

• 22 and 23 Auxiliary Boiler Feed Pumps (ABFP) with the 21 ABFP Out of Service

for Maintenance

• Emergency Diesel Generators 21, 22, and 23 Following Periodic Testing

b. Findings

No findings of significance were identified.

.2 Full Equipment Alignment

a. Inspection Scope (71111.04S - 1 sample)

The inspectors performed an extensive walkdown of the component cooling water

(CCW) system. The inspectors walked down the system using 2-PT-Q90, Component

Cooling Water System Quarterly Alignment Verification, Revision 0, and the system

flow diagrams. The inspectors verified that all accessible system components were in

the proper position and verified that any discrepancies were properly documented.

Additionally, the inspectors evaluated the physical condition of the equipment during the

walkdown and reviewed open CRs and WOs to evaluate if any had the potential to

impact system operability. This system walkdown was considered one inspection

sample.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

a. Inspection Scope (71111.05Q - 6 samples)

The inspectors toured areas that were identified as important to plant safety and risk

significance. The inspectors consulted the Indian Point 2 Individual Plant Examination

for External Events (IPEEE), Section 4.0, Internal Fires Analysis, and the top risk-

significant fire zones in Table 4.6-2, Summary of Core Damage Frequency

Contributions from Fire Zones. The objective of this inspection was to determine if

Entergy had adequately controlled combustibles and ignition sources within the plant,

effectively maintained fire detection and suppression capability, and had adequately

established compensatory measures for degraded fire protection equipment. The

inspectors evaluated conditions related to: (1) control of transient combustibles and

ignition sources; (2) the material condition, operational status, and operational lineup of

fire protection systems, equipment, and features; (3) the fire barriers used to prevent fire

damage or fire propagation; (4) compensatory measures for out-of-service, degraded, or

inoperable fire protection equipment were implemented in accordance with Entergys fire

Enclosure

3

plan. Reference material used by the inspectors to determine the acceptability of the

observed conditions in the fire zones are referenced in the attachment at the end of this

report. The following areas were counted as inspection samples:

• Zone 11

• Zones 5, 6, 7

• Zone 14

• Zones 9, 12A, 13A

• Zones 19, 20, 45A

• Zone 1

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures

a. Inspection Scope (71111.06 - 1 internal flooding sample)

The inspectors reviewed Entergys internal flood analysis, flood mitigation procedures,

and design features to verify that they were consistent with IP2's design

requirements. The inspectors walked down the emergency diesel generator (EDG)

building and evaluated the condition and adequacy of mitigation equipment to assess

whether flood protection design features were adequate. This walkdown constituted one

inspection sample.

The inspectors reviewed Entergys flood mitigation procedures. In addition, the

inspectors reviewed the corrective action program (CAP) to determine if there was any

history of flood problems in this area. The specific information reviewed is referenced in

the attachment at the end of this report.

b. Findings

Introduction: The inspectors identified an Unresolved Item (URI) associated with the

potential vulnerability of the normal and emergency 480 VAC vital alternating current

(AC) power sources to flooding in the EDG building. Approximately 30 to 50 oil

absorbing pads of varying sizes were found on the floor underneath all three EDGs.

During a flooding event, these pads could be swept into the five building drainage

sumps, preventing water from being drained from the building.

Discussion: On November 29, 2005, the inspectors reviewed the internal flood

protection measures for the EDG building. All three 480 VAC EDGs are located in this

building in a common area separated by installed fire barrier walls. The EDGs are

approximately five feet above the concrete floor of the building and access to the EDGs

is afforded by metal grate flooring. The major sources of potential flooding for the space

are fire protection piping and the essential service water (ESW) system piping in the

building. The building is designed such that water drains toward five shallow sumps that

Enclosure

4

are connected to a common 12-inch-diameter drain line that discharges to the site

drainage system. Each sump has two 3-inch-diameter openings that have backwater

ball check valves to prevent back-leakage into the EDG building. Inspectors observed

between 30 and 50 oil absorbent pads on the concrete floor underneath the EDGs.

These pads were not fixed to the floor by any means, and would be free to migrate to

the building sumps along with water during a flooding event. The pads were of sufficient

size to effectively block the 3-inch holes in each of the building sumps as water level

rises in the building. The IPEEE credits the building drains being sufficiently sized to

prevent significant accumulation of water due to a break of fire protection piping in the

room. This assumes that the function of these drains is not impeded by foreign material

blockage. In addition, both the IPEEE and the IP2 Probabilistic Safety Assessment

(PSA) state that a break of an ESW line is bounded by the fact that the EDGs are

cooled by ESW and would be the only equipment negatively impacted by the flooding,

and that this occurrence is analyzed by the total loss of service water event. There are

inconsistencies between the IPEEE, dated 1995, and the PSA, which was completed in

the 1998 time frame. The PSA does not account for the fire protection header as being

a potential source of flooding for the EDG building, whereas the IPEEE does. Both

analyses credit open ventilation louvers along the building north wall at grade level to

drain water if the buildings installed drain capacity is insufficient. However, during the

winter months these louvers are maintained shut. In addition, the IPEEE mentions an

EDG building flood alarm in the control room and specific isolation procedures in the

event of flooding. Neither the alarm, nor the specific isolation procedures, currently

exist. Finally, the inspectors identified that 480 VAC normal feeder breaker control

power exists in each EDG control cabinet. Flood water that reaches the bottom of the

EDG control cabinets due to insufficient building drain capacity, and can not be relieved

through closed building doors and closed ventilation louvers, could potentially render all

three EDGs unavailable and trip the normal feeder breakers to all 480 VAC vital AC

buses. In response to the , Entergy removed the oil absorbent

pads from the EDG building and entered the issue into the corrective action

program (CR-IP2-05-4868). This issue will be treated as a URI pending additional

licensee evaluation and inspector review of the potential impact of flooding in the EDG

building on the normal and emergency vital AC power sources:

URI 05000247/2005005-01, Emergency Diesel Generator Building Flood Mitigation

Capability.

1R07 Heat Sink Performance

a. Inspection Scope (71111.07 - 1 sample)

The inspectors performed a review of the instrument air closed cooling water (IACCW)

heat exchangers to verify that Entergy was monitoring performance on a continuing

basis and to ensure that any potential deficiencies which could mask degraded

performance were identified. The inspectors reviewed the design basis documents and

Final Safety Analysis Report (FSAR) to validate that testing acceptance criteria were

appropriate. The inspectors also reviewed the latest inspection reports for both the 21

and 22 IACCW heat exchangers, evaluated the results of eddy current testing, and

Enclosure

5

ensured that the appropriate tube plugging criteria were used. In addition, the

inspectors verified that Entergy was maintaining their commitments from Generic Letter 89-13 concerning heat exchanger inspection and testing. The inspection of the IACCW

heat exchangers constituted one inspection sample.

b. Findings

No findings of significance were identified.

1R11 Operator Requalification Inspection

.1 Resident Inspector Quarterly Review

a. Inspection Scope (71111.11Q - 2 samples)

On November 30, 2005, the inspectors observed an Emergency Plan drill

implementation by licensed operators in the simulator. The inspectors reviewed the

simulator scenario performed as a part of the overall drill to determine if the scenario

contained: (1) clear event descriptions with realistic initial conditions, (2) clear start and

end points, (3) clear descriptions of visible plant symptoms for the crew to recognize,

and (4) clear expectations of operator actions in response to abnormal conditions. The

scenario involved a simulated reactor coolant system leak, small break loss of coolant

accident, large break loss of coolant accident, and a loss of emergency coolant

recirculation capability.

During the simulator exercise, the inspectors evaluated the teams performance for: (1)

clarity and formality of communications, (2) correct use and implementation of

emergency operating procedures (EOPs) and abnormal operating procedures (AOPs),

(3) operators ability to properly interpret and verify alarms, (4) operators ability to

classify events in a timely fashion, and (4) operators ability to take timely actions in a

safe direction based on transient conditions. In addition, the inspectors evaluated the

control room supervisors ability to exercise effective oversight and control of the crews

actions during the exercise.

On December 1, 2005, the inspectors observed in-plant training of 2-AOP-SSD-1,

Control Room Inaccessibility Safe Shutdown Control. The training involved rotating

three groups of licensed operators between the primary auxiliary building, auxiliary

feedwater pump building, and turbine hall to walk-through the complex procedure. The

instructors were knowledgeable and asked probing questions of the students throughout

the training. Actions in the procedures were simulated as actual performance was not

possible due to plant operation.

The simulator scenario observation was counted as one inspection sample, and the

observation of the walk through training was counted as a second inspection sample.

b. Findings

Enclosure

6

No findings of significance were identified.

.2 Annual Review of Operating Test and Comprehensive Written Exam Results

a. Inspection Scope (71111.11B - 1 sample)

On December 19, 2005, the inspector conducted an in-office review of licensee annual

operating test results and comprehensive written exam results for 2005, constituting one

inspection sample. The inspection assessed whether pass rates were consistent with

the guidance of NRC Manual Chapter 0609, Appendix I, Operator Requalification

Human Performance Significance Determination Process (SDP). The inspector verified

that:

• Crew failure rate was less than 20%. (Crew failure rate was 0%.)

• Individual failure rate on the dynamic simulator test was less than or equal to

20%. (Individual failure rate was 0%.)

• Individual failure rate on the walk-through test was less than or equal to 20%.

(Individual failure rate was 2%.)

• Individual failure rate on the comprehensive written exam was less than or equal

to 20%. (Individual failure rate was 11%.)

• Overall pass rate among individuals for all portions of the exam was greater than

or equal to 75%. (Overall pass rate was 87%.)

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

.1 Maintenance Rule Implementation - Quarterly

a. Inspection Scope (71111.12Q - 3 samples)

The inspectors also reviewed WOs, and associated post-maintenance test activities to

assess whether: (1) the effect of maintenance work in the plant had been adequately

Enclosure

7

addressed by control room personnel, (2) work planning was adequate for the

maintenance performed, (3) the acceptance criteria were clear and adequately

demonstrated operational readiness consistent with design and licensing documents,

and (4) the equipment was effectively returned to service. Referenced documents are

listed in the Supplemental Information attachment at the end of this report. The below-

listed systems maintenance activities were observed and/or evaluated. Each system

review constituted one inspection sample.

• Control Rod Drive Mechanism Fans

• 480 VAC Circuit Breakers

• CCW System

b. Findings

Introduction: The inspectors identified a Green finding in that Entergy failed to maintain

adequate design control of the control rod drive mechanism (CRDM) fans. This directly

resulted in loss of lubrication and failure of the 23 CRDM fan.

Description: On June 2, 2005, the 23 CRDM fan failed during operation. These fans

are used to remove the heat generated by the CRDMs when the plant is at power.

Restrictions are placed on plant operation and additional actions by operators are

required if more than one fan is out of service. The CRDM fans are also used in the

EOPs to facilitate natural circulation cooldown of the plant by preventing void formation

in the reactor head region. Alternate mitigation strategies are available to the operators

in the event the fans are unavailable, however, this complicates and slows down the

cooldown process.

Entergys failure analysis for the 23 CRDM fan determined that the failure was due to a

lack of grease in a motor bearing. It was further determined that the fan motor bearings

were not built in accordance with design change package MSAP-00-00524-FFX. This

design change, which installed a more robust bearing design, was implemented in 2000

to improve CRDM fan reliability and recover the system from Maintenance Rule (a)(1)

status. The original motor design used light duty ball bearings, while the upgraded

design used heavier duty bearings with a thrust bearing at one end. The thrust bearing

required the installation of shields to ensure it could run for a 24 month cycle without

additional grease. The shields in the 23 CRDM fan had not been properly installed,

which ultimately resulted in bearing seizure due to lack of lubrication.

The inspectors also noted that following a failure of the 24 CRDM fan in April 2005 from

an unrelated cause, Entergy identified that the spare motor was not properly configured.

The companys investigation determined that there was an error in the purchase order

for fans installed in the November 2004 plant outage, and Entergy could not be certain

of the configuration of any of the installed motors.

. Entergy entered this issue into the corrective action program (CR-

IP2-05-2210) and has taken actions to ensure that the deficiency will be corrected

during the upcoming refueling outage.

Enclosure

8

Analysis: The inspectors determined that the failure to maintain design control of the

CRDM fans was a performance deficiency since it was the direct result of errors in

Entergys procurement process. This issue directly resulted in the failure of the 23

CRDM fan. It is reasonable that Entergy could have recognized and prevented this

problem. Traditional enforcement does not apply since there were no actual safety

consequences or potential for impacting the NRCs regulatory function, and the finding

was not the result of any willful violation of NRC requirements or Entergys procedures.

This finding is greater than minor since it is associated with the Design Control attribute

of the Mitigating Systems cornerstone objective to ensure availability, reliability, and

capability of systems that respond to initiating events to prevent undesirable

consequences. This finding was evaluated using Phase 1 of Inspection Manual Chapter

(IMC) 0609, Appendix A, Significance Determination of Reactor Inspection Findings for

At-Power Situations. The finding is of very low safety significance since the

performance deficiency does not represent an actual loss of function and did not

screen as risk-significant due to seismic, flooding, or severe weather initiating events.

was out of service, the other three fans were available to perform the

necessary system functions.

Enforcement: No violation of regulatory requirements occurred since the design control

issues involved the non-safety-related CRDM fans which are outside the scope of 10

CFR 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel

Reprocessing Plants. This finding is identified as FIN 05000247/2005005-02, Failure

to Maintain Design Control of Control Rod Drive Mechanism Fans.

.2 Maintenance Rule Implementation - Biennial

a. Inspection Scope (71111.12B - 4 samples)

The inspector conducted a review of the periodic evaluation of implementation of the

Maintenance Rule as required by 10 CFR 50.65(a)(3) for IP2. The evaluation covered a

period from April 2003 to April 2005. The purpose of this review was to ensure that IP2

established appropriate goals, and effectively assessed system performance and

preventive maintenance activities. The inspector verified that the evaluation was

completed within the required time period and that industry operating experience was

utilized, where applicable. Additionally, the inspector verified that Indian Point

appropriately balanced equipment reliability and availability and made adjustments when

appropriate.

The inspector selected a sample of four risk-significant systems to verify that: (1) the

structures, systems, and components were properly characterized; (2) goals and

performance criteria were appropriate; (3) corrective action plans were adequate; and

(4) performance was being effectively monitored in accordance with station procedure

ENN-DC-121, Maintenance Rule. The following systems were selected for detailed

review and constituted four inspection samples:

Enclosure

9

• Auxiliary Feedwater System

• Chemical and Volume Control System

• Control Rod Drives

• Gas Turbines

These systems were either in (a)(1) status, had been in (a)(1) status at some time

during the assessment period, or had experienced degraded performance. The

inspector reviewed corrective action documents for malfunctions and failures of these

systems to determine if: (1) system failures had been correctly categorized as functional

failures, and (2) system performance was adequately monitored to determine if

classifying a system as (a)(1) was appropriate.

The inspector interviewed the maintenance rule coordinator and system engineers,

reviewed documentation for applicable systems, and reviewed a sample of condition

reports. The documents that were reviewed are listed in the attachment to this report.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope (71111.13 - 4 samples)

The inspectors observed selected portions of routinely scheduled and emergent

maintenance work activities to assess Entergys risk management in accordance with 10

CFR 50.65(a)(4). The inspectors verified that Entergy took the necessary steps to plan

and control emergent work activities, to minimize the probability of initiating events, and

to maintain the functional capability of mitigating systems. The inspectors observed

and/or discussed risk management actions with maintenance and operations personnel.

The following emergent work activities were observed, and constituted four inspection

samples:

• WO IP2-05-25862, Replacement of the 22 S/G High Steam Flow SI Initiation

Bistable

• CR IP2-05-4759, PCV-455C Low Nitrogen Pressure Alarms

• WO IP2-01-23320, Lighting Bus 22/23 Tie Breaker Maintenance

• FCV-447 Packing Leakage Corrective Actions

Introduction: The inspectors identified a Green non-cited violation (NCV) in that

modification documents and procedures were not followed while implementing a

temporary modification to the 24 S/G feedwater regulating valve, FCV-447. This was

determined to be a violation of 10 CFR 50, Appendix B, Criterion V, Instructions,

Procedures and Drawings.

Enclosure

10

Description: On September 27, 2005, a modification was performed on FCV-447 in

which the cap screws holding the valve actuator onto the valve body were ground down

at an angle to allow clearance between the cap screws and the packing gland follower.

This valve is a safety-related component required for isolation of one of the four

feedwater lines following a feed or steam line break inside the containment to minimize

peak containment pressure. The additional clearance was required to allow further

packing adjustments to prevent feedwater/steam leakage through the valves packing.

The temporary alteration package, TA-05-2-107, specified that the maximum thickness

removed was to be three-eighths of an inch, which would leave a minimum of three-

eighths of an inch of the cap screw head remaining. The supporting calculation to

assure structural integrity following this modification also assumed the cap screw

thickness to be three-eighths of an inch following the grinding.

The inspectors reviewed Entergys analysis to show that the structural integrity of the

cap screws would be maintained, and completed a walkdown of the valve to ensure the

modification had been properly implemented. The inspectors identified that more

material had been ground from the cap screws than allowed by the analysis. The

inspectors also found that Entergys structural integrity evaluation failed to consider

seismic stresses. Based on the inspectors observations, Entergy declared the valve

inoperable and entered a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement as required by Technical

Specifications. Entergy entered this issue into the corrective action program (CR-IP2-

05-4615) and completed an additional evaluation which showed that the actuator cap

screws on FCV-447 would remain operable under design conditions in the as-left

condition. Following this evaluation, Entergy exited the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement.

Analysis: The inspectors determined that this is a performance deficiency since Entergy

exceeded the grinding depth as specified in the alteration package and failed to identify

this condition. It is reasonable that Entergy could have recognized and prevented this

problem. Traditional enforcement does not apply since there were no actual safety

consequences or potential for impacting the NRCs regulatory function and the finding

was not the result of any willful violation of NRC requirements or Entergys procedures.

This finding is greater than minor, because it was associated with the Barrier Integrity

cornerstone attribute of Human Performance, and affected the cornerstone objective of

ensuring the containment would remain functional to protect the public from radionuclide

releases caused by accidents or events. This finding was evaluated using Phase 1 of

IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for

At-Power Situations, and is of very low safety significance since the work performed on

FCV-447 did not result in an actual loss of its safety function.

This finding is also related to the cross-cutting area of Human Performance in that

maintenance personnel failed to implement the modification as specified in the

temporary alteration package. This error was not identified by the maintenance workers

or engineering personnel upon completion of the modification. (Section 4OA4).

Enforcement: 10 CFR 50, Appendix B, Criterion V states, in part, that activities affecting

quality shall be prescribed by procedures and shall be accomplished in accordance with

these procedures. Contrary to this, Entergy failed to implement the modification of FCV-

Enclosure

11

447 as prescribed by the modification requirements. Because this failure to follow

modification requirements for the valve is of very low safety significance and has been

entered into Entergys corrective actions program (CR-IP2-2005-4615) this violation is

being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy:

NCV 05000247/2005005-03, Failure to Follow Procedural Requirements During

Modification of a Safety-Related Valve.

2. Introduction: The inspectors identified a Green NCV of Technical Specification (TS)

5.4.1, Administrative Controls - Procedures, because Indian Points work management

procedure inappropriately allowed actions to be implemented to perform a modification

on a safety-related component before the modification package was issued.

Description: The inspectors reviewed procedure IP-SMM-WM-100, Work Management

Process. This procedure, which is specific to Indian Point, defines emergency work, in

part, as: Actions required to prevent a real or potential plant transient or forced

shutdown. TS 5.4.1 requires Indian Point to establish, implement, and maintain those

procedures discussed in Appendix A of Regulatory Guide (RG) 1.33, Quality Assurance

Program Requirements (Operation). RG 1.33 states that maintenance which can affect

the performance of safety-related equipment should be properly preplanned and

performed in accordance with written procedures, documented instructions, or drawings

appropriate to the circumstances.

Entergys work management procedure provided steps to bypass the established work

control process in order to prevent a forced shutdown of the plant based on the

procedural definition of emergency work. 10 CFR 50.54(x) allows licensees to take

actions that depart from a Technical Specification requirement in an emergency, when

that action is immediately needed to protect public health and safety, and there is no

immediately apparent action consistent with the Technical Specifications which can

provide adequate or equivalent protection. The inspectors determined that Entergys

definition of emergency work, specifically work to prevent a forced shutdown,

.

Therefore, there was no valid basis to bypass the established work control process, as

required by TS 5.4.1, to prevent forced shutdowns or plant transients. In addition, the

inspectors noted that EN-WM-100, Work Request Generation, Screening and

Classification, the Entergy fleet-wide governing document for the site specific work

control procedure, did not classify work to prevent a forced shutdown as emergency

work. The inspectors determined that Indian Points work management procedure

inappropriately allowed maintenance which could result in a shutdown or transient to be

declared emergency work thus allowing the required work controls process to be

bypassed.

On September 27, 2005, and again on November 8, 2005, Entergy modified capscrews

on a safety related valve (FCV-447), using the work management procedures allowance

for emergency work. Entergys decision to use this process was based on the

assumption that prompt action was required to prevent excessive damage to the

packing, which would make future packing adjustments unsuccessful and leave valve

Enclosure

12

re-packing or leak repair as the only viable maintenance alternatives. The declaration of

emergency work allowed the maintenance to be performed prior to completion of work

procedures, a modification package, and the associated engineering analysis.

In response to the inspectors concerns,

Additionally, procedure IP-SMM-WM-100 was

revised to prevent the declaration of emergency work for issues which could result in a

plant shutdown or transient. Entergy also completed an engineering analysis to show

that FCV-447 would still be able to perform its safety function.

Analysis: The inspectors determined that Indian Points development of a procedure

actions contrary to the plants Technical Specifications is a performance

deficiency and directly resulted in actions to commence modifications on a safety-

related component before evaluation was completed. It is reasonable that

Entergy could have recognized and prevented this problem. Traditional enforcement

does not apply because there were no actual safety consequences or potential for

impacting the NRCs regulatory function, and the finding was not the result of any willful

violation of NRC requirements or Entergy procedures. This finding is greater than

minor, because if left uncorrected, the issue could become a more significant safety

concern. This finding was evaluated using Phase 1 of IMC 0609, Appendix A,

Significance Determination of Reactor Inspection Findings for At-Power Situations.

The inspectors determined that the violation of TS 5.4.1 is of very low safety significance

since the work performed on FCV-447 did not result in the actual loss of safety function

of a system and did not impact fire, flooding, seismic, or severe weather initiating

events.

This finding is associated with the Human Performance cross-cutting area in that the

decision to implement the modification to FCV-447 without an adequate evaluation was

based on inappropriate procedural guidance (see Section 4OA4).

Enforcement: TS 5.4.1 requires that written procedures be established, implemented,

and maintained covering the activities recommended in RG 1.33. RG 1.33 states that

maintenance which can affect the performance of safety-related equipment should be

properly preplanned and performed in accordance with written procedures, documented

instructions, or drawings appropriate to the circumstances. Contrary to the above,

Entergy developed a site work control procedure which allowed a modification to be

performed on an in-service safety-related component prior to the modification

documents being completed and formalized. Because this violation is of very low safety

significance

, it is being treated as an NCV consistent with Section VI.A.1 of the NRC

Enforcement Policy: NCV 05000247/2005005-04, Inadequate Procedure for Control

of Work on Safety-Related Components.

Enclosure

13

November 17, 2005

• On December 22, 2005, the inspectors monitored Entergys actions to reduce

reactor power to approximately 3 percent and take the turbine off-line to repack

FCV-447. The inspectors reviewed Entergys procedures for plant shutdown and

observed the evolution in the control room. The inspectors also observed

important activities associate with the power ascension following completion of

the maintenance. A list of documents reviewed is included in the attachment to

this report.

b. Findings

No findings of significance were identified.

Enclosure

14

1R15 Operability Evaluations

a. Inspection Scope (71111.15 - 5 samples)

The inspectors selected operability evaluations that Entergy had generated that

warranted review on the basis of potential risk significance. The selected samples are

addressed in the CRs listed below. The inspectors assessed the accuracy of the

evaluations, the use and control of compensatory measures, if needed, and compliance

with the TSs. The inspectors review included a verification that the operability

evaluations were made as specified by procedure ENN-OP-104, Operability

Determinations. The technical adequacy of the evaluations was reviewed and

compared to the TSs, Technical Requirements Manual (TRM), FSAR, and associated

design basis documents. The operability evaluations that were inspected constituted

five inspection samples.

• CR IP2-05-4246, Operations with the Low Feed Flow Bypass Isolation Valves

Open

• CR IP2-05-4414, Failure of FCV-1176 (EDG Service Water Outlet Valve) to

Close Following Surveillance Testing

• CR IP2-05-4642, EDG Fuel Oil Inventory

• CR-IP2-05-4792, Identification of a Gas Void Between the Outlet of the 21

Residual Heat Removal Heat Exchanger and Valve HCV-638

• CR IP2-05-4841, Cable Separation Issues Identified During Walkdowns Outside

Containment

b. Findings

No findings of significance were identified.

1R16 Operator Workarounds

a. Inspection Scope (71111.16 - 2 samples)

The inspectors reviewed the operator workaround associated with failure of a power

supply for Control Room alarm panel AS-1. The inspectors reviewed the individual

alarms lost, impact on plant operation, and the compensatory measures established by

Entergy. The inspectors evaluated these measures to ensure they were appropriately

scoped within Entergys operator burdens program and that the required actions could

feasibly be performed by the operations staff. This operator workaround review

constitutes one inspection sample.

The inspectors also focused on the operator workaround associated with one

pressurizer spray valve (PCV-455A) being isolated. The inspectors verified that the

Operational Decision Making Process was followed for this issue, and that appropriate

compensatory actions were taken. The inspectors used OAP-45, Operator Burden

Program, and EN-OP-111, Operational Decision Making Issue Process, to evaluate

plant deficiencies and their effects on plant operation.

Enclosure

15

The sample related to operation with PCV-455A isolated was originally documented in

Inspection Report 05000247/2005-04. However, due to an oversight, it was not counted

as a completed sample. It is being documented again in this report to correct the error.

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope (71111.19 - 8 samples)

The inspector reviewed post-maintenance test (PMT) procedures and associated testing

activities to assess whether: (1) the effect of testing in the plant was adequately

addressed by control room personnel; (2) testing was adequate for the maintenance

WO performed; (3) the acceptance criteria was clear and adequately demonstrated

operational readiness consistent with design and licensing documents; (4) test

instrumentation had current calibrations, range, and accuracy for the application; and

(5) test equipment was removed following testing.

The selected testing activities involved components that were risk-significant as

identified in the IP2 Individual Plant Examination (IPE). The regulatory references for

the inspection included TSs and 10 CFR 50, Appendix B, Criterion XIV, Inspection,

Test, and Operating Status. The following testing activities were evaluated, and

constituted eight inspection samples:

• WO IP2-05-2522, Containment Sump Stop Valve 885B Following Mechanical

Interlock Replacement

• WO IP2-05-26316, 21 CCP Following Internal Valve Replacements Due to Flow

Oscillation

• WO IP2-05-27328, PWT Following On-Line Leak Repair of PCV-1135, the 22

S/G Atmospheric Dump Valve

• WO IP2-05-28030, 23 Station Battery Following Replacement of Cell 3 Due to

Low Cell Voltage

• WO IP2-05-20640, 21 CCW Pump Following a Planned Component

Maintenance Outage Period

• WO IP2-05-21932, 24 Service Water Pump (SWP) Following a Planned

Component Maintenance Outage Period

• WO IP2-04-12427, 25 SWP Following a Planned Component Maintenance

Outage Period

• WO IP2-05-26680, PWT for 3M Oil and Filter Change on 23 EDG Starting Air

Compressor

b. Findings

No findings of significance were identified.

Enclosure

16

1R22 Surveillance Testing

a. Inspection Scope (71111.22 - 5 samples)

The inspectors reviewed surveillance test procedures and observed testing activities to

assess whether: (1) the test preconditioned the component tested; (2) the effect of the

testing was adequately addressed in the control room; (3) the acceptance criteria

demonstrated operational readiness consistent with design calculations and licensing

documents; (4) the test equipment range and accuracy were adequate and the

equipment was properly calibrated; (5) the test was performed per the procedure; (6)

test equipment was removed following testing; and (7) the test discrepancies were

appropriately evaluated. The surveillance tests observed were based on risk-significant

components as identified in the IP2 IPE. The regulatory requirements that provided the

acceptance criteria for this review were 10 CFR 50, Appendix B, Criterion V,

Instructions, Procedures, and Drawings; Criterion XIV, Inspection, Test, and

Operating Status; Criterion XI, Test Control; and TS 6.8.1.a. The following test

activities were reviewed and constituted five inspection samples:

• 2PT-Q29B, 22 Safety Injection Pump Quarterly Test, Revision 15

• 2PI-3Y2A, 23 Auxiliary Boiler Feed Pump Suction and Discharge Inservice

Inspection Pressure Test, Revision 4

• 2PT-Q27A, 21 Auxiliary Boiler Feed Pump Quarterly Test, Revision 12

• 2PT-SA067, Main Turbine Stop and Control Valves, Revision 1

• 2PT-M21A, Emergency Diesel Generator 21 Load Test, Revision 11

b. Findings

No findings of significance were identified.

1R23 Temporary Plant Modifications

a. Inspection Scope (71111.23 - 2 samples)

The inspectors reviewed two temporary modifications, that constituted two inspection

samples, to ensure that the effects on plant operation were well understood and to

ensure that no unintended, adverse consequences would result from the modification.

The inspectors evaluated the modification documentation for accuracy and

completeness, the basis for the modification, and any associated procedures or

changes to procedures to control the temporary modification operation. The following

temporary modifications were reviewed:

• WO IP2-05-27331, On-line Leak Repair of PCV-1135, 22 S/G Atmospheric

Dump Valve

• TA-05-2-039, Removal and Installation of SWN-840 Actuator on 22 CCW Heat

Exchanger Valve SWN-35-1

Enclosure

17

b. Findings

No findings of significance were identified.

Cornerstone: Emergency Preparedness

1EP2 Alert and Notification System Testing

a. Inspection Scope (71114.02 - 1 Sample)

An onsite review of Entergys alert and notification system (ANS) was conducted to

ensure prompt notification of the public for taking protective actions. During the

inspection at Indian Point, the inspectors reviewed the test and maintenance

documentation for the siren system. Distribution records were sampled pertaining to the

tone alert radio portion of the ANS. CRs generated as a result of siren testing were

reviewed for causes, trends, and corrective actions. The inspectors interviewed

personnel responsible for the alert and notification program. The inspection was

conducted in accordance with NRC Inspection Procedure 71114, Attachment 02.

Planning standard 10 CFR 50.47(b)(5) and the related requirements of 10 CFR 50,

Appendix E were used as reference criteria.

b. Findings

No findings of significance were identified.

1EP3 Emergency Response Organization (ERO) Augmentation Testing

a. Inspection Scope (71114.03 - 1 Sample)

A review of Indian Points ERO augmentation staffing requirements and the process for

notifying the ERO was conducted to ensure the readiness of key staff for responding to

an event and to ensure timely facility activation. The inspectors reviewed procedures

and CRs associated with the ERO notification system and process. The inspectors

interviewed personnel responsible for the ERO augmentation process. The inspection

was conducted in accordance with NRC Inspection Procedure 71114, Attachment 03.

Planning standard 10 CFR 50.47(b)(2) and related requirements of 10 CFR 50,

Appendix E were used as reference criteria.

b. Findings

No findings of significance were identified.

Enclosure

18

1EP4 Emergency Action Level (EAL) and Emergency Plan Changes

.1 EAL Review

a. Inspection Scope

The inspectors reviewed changes to Entergys EALs to ensure that the changes did not

decrease the effectiveness of the Emergency Plan. The inspectors reviewed Entergy

procedures to determine if an EAL scheme had been changed in a manner that

decreased its effectiveness such that the EALs may not produce the appropriate

emergency classification. The inspectors verified that the EAL scheme continued to

meet the planning standard.

b. Findings

Introduction: The inspectors identified a Green NCV associated with emergency

planning standard 10 CFR 50.47(b)(4). The inspectors determined that a performance

deficiency existed in that inadequate indications were available for operators to

determine if a threshold for an unusual event (UE), based on service water bay level,

had been met. This issue did not result in the loss or degradation of a risk significant

planning standard based on the inspectors assessment of the criteria in NRC Manual

Chapter 0609, Appendix B, Emergency Preparedness Significance Determination

Process.

Description: A combination of low tides and debris on the intake structure trash bars

resulted in a low service water bay level at the Indian Point Unit 3 (IP3) intake structure

between November 23 to November 25, 2005. Operators were alerted to this condition

due to the occasional trips of the non-safety related screen wash pumps. EAL 8.4.3

requires the declaration of a UE if service water bay level drops to 4 feet 5 inches below

mean sea level. In response to the low water conditions, the operators improvised a

means to measure the service water bay level and determined that the UE criteria had

not been met. The inspectors discussed the availability of instrumentation for

assessment of the UE entry criteria with IP operations and emergency planning staff,

reviewed relevant plant procedures, and performed a walkdown of the intake structure.

The inspectors determined that Entergy had no established means of indication or

instrumentation for operators to assess the service water bay level and evaluate the

associated entry criteria of EAL 8.4.3. Upon further review,

Analysis: The performance deficiency is that no established means of indication or

procedures were readily available for operators to determine if the service water bay

level met the threshold declaration of an UE described in EAL 8.4.3. The failure to

provide adequate indication for assessment of EAL entry criteria could impact the timely

declaration of an emergency and is contrary to 10 CFR 50.54(q) and 50.47(b)(4). This

finding is greater than minor because it was associated with the Emergency

Enclosure

19

Preparedness (EP) cornerstone attribute of Facilities and Equipment, and affected the

cornerstone objective of ensuring that a licensee is capable of implementing adequate

measures to protect the health and safety of the public in the event of a radiological

emergency. This finding was evaluated using Inspection Manual Chapter 0609,

Appendix B, Emergency Preparedness Significance Determination Process, Sheet 1,

Failure to Comply. This finding is associated with a failure to meet or implement a

regulatory requirement. The deficiency is not greater than Green because it did not

result in the Risk-Significant Planning Standard Function being lost or degraded.

Section 4.4 of Manual Chapter 0609, Appendix B, provides examples for use in

assessing emergency preparedness related findings. One example of a Green finding

states, The EAL classification process would not declare any Alert or Notification of

Unusual Event that should be declared. Since the declaration of an UE based on low

service water bay level could have been missed or delayed, this finding was considered

consistent with the example provided and was therefore determined to be of very low

safety significance (Green).

Enforcement: 10 CFR 50.54(q) requires that the facility licensee follow and maintain in

effect emergency plans which meet the standards in 10 CFR 50.47(b). 10 CFR

50.47(b)(4) requires, in part, that emergency response plans include a standard

emergency classification and action level scheme, the bases of which include facility

system and effluent parameters. The emergency classification and action level scheme

is required to be used by the nuclear facility licensee, and State and local response

plans rely on information provided by facility licensees for determinations of minimum

initial offsite response measures. Contrary to the above, prior to November 2005,

Entergy did not have adequate means of indication or procedures to support an EAL

classification based on service water bay intake level. Entergy entered this issue into its

CAP as CR-IP3-2005-5380 and installed temporary level indication pending the

development of permanent corrective actions. Because this issue is of very low safety

significance and has been entered into Entergy's CAP, it is being treated as an NCV

consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000247/2005-

05-05, Inadequate Equipment to Assess Threshold for Emergency Action Level

8.4.3.

.2 Emergency Plan Change Review

a. Inspection Scope

Prior to this inspection, the NRC had received and acknowledged the changes made to

the Indian Point Emergency Plan and implementing procedures. These changes were

made in accordance with 10 CFR 50.54(q), which Entergy had determined did not result

in a decrease in effectiveness to the Plan and concluded that the changes continued to

meet the requirements of 10 CFR 50.47(b) and Appendix E of 10 CFR 50. During this

inspection, the inspectors conducted a sampling review of the changes which could

potentially result in a decrease in effectiveness. This review does not constitute an

approval of the changes and, as such, the changes are subject to future NRC

inspection. The inspection was conducted in accordance with NRC Inspection

Enclosure

20

Procedure 71114, Attachment 4. The requirements in 10 CFR 50.54(q) were used as

reference criteria.

b. Findings

No findings of significance were identified.

1EP5 Correction of Emergency Preparedness Weaknesses and Deficiencies

a. Inspection Scope (71114.05 - 1 Sample)

The inspectors reviewed CRs initiated by Indian Point from drills, self-assessments, and

audits, and the associated corrective actions to determine the significance of the issues

and to determine if repeat problems were occurring. A list of the CRs reviewed are

contained in the attachment to this report. Also, the 2004 and 2005 audit reports were

reviewed to assess Indian Points ability to identify issues, assess repetitive issues, and

evaluate the effectiveness of corrective actions through their independent audit process.

This inspection was conducted according to NRC Inspection Procedure 71114,

Attachment 05. Planning standard 10 CFR 50.47(b)(14) and the related requirements of

10 CFR 50, Appendix E were used as reference criteria.

b. Findings

No findings of significance were identified.

1EP6 Drill Evaluation

a. Inspection Scope

The inspectors observed an EP drill conducted on November 30. The inspectors used

NRC Inspection Procedure 71114.06, Drill Evaluation, as guidance and criteria for

evaluation of the drill. The drill consisted of an Emergency Notification Siren test and

Emergency Response Organization Drill. The inspectors observed the drill and

conducted reviews from the Indian Point Emergency Operations Facility (EOF). The

inspectors focused the reviews on the identification of weaknesses and deficiencies in

the classification and notification timeliness and quality and accountability of essential

personnel during the drill. The inspectors were briefed on Entergys critique results and

compared the NRC-identified weaknesses and deficiencies to those identified by

Entergy to ensure that problem areas were properly identified. Inspection of this EP Drill

constitutes one inspection sample.

b. Findings

No findings of significance were identified.

Enclosure

21

4. OTHER ACTIVITIES

4OA1 Performance Indicator Verification

.1 Occupational Exposure Control Effectiveness

a. Inspection Scope (71151 - 1 sample)

The inspector reviewed implementation of Entergys Occupational Exposure Control

Effectiveness Performance Indicator (PI) Program. Specifically, the inspector reviewed

CRs, and radiological controlled area dosimeter exit logs for the past four calendar

quarters. These records were reviewed for occurrences involving locked high radiation

areas, very high radiation areas, and unplanned exposures against the criteria specified

in Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Performance Indicator

Guideline, Revision 2, to verify that all occurrences that met the NEI criteria were

identified and reported as performance indicators. This inspection activity represents

the completion of one sample relative to this inspection area, completing the annual

inspection requirement.

b. Findings

No findings of significance were identified.

.2 RETS/ODCM Radiological Effluent Occurrences

a. Inspection Scope (71122.01 - 1 Sample)

The inspector reviewed a listing of relevant effluent release reports for the past four

calendar quarters, for issues related to the public radiation safety PI, which measures

radiological effluent release occurrences per site that exceed 1.5 mrem/quarter whole

body or 5.0 mrem/quarter organ dose for liquid effluents, 5 mrad/quarter gamma air

dose, 10 mrad/quarter beta air dose, and 7.5 mrads/quarter for organ dose for gaseous

effluents. This inspection activity represents the completion of one sample relative to

this inspection area, completing the annual inspection requirement.

The inspector reviewed the following documents to ensure Entergy met all requirements

of the performance indicator:

• monthly projected dose assessment results due to radioactive liquid and

gaseous effluent releases;

• quarterly projected dose assessment results due to radioactive liquid and

gaseous effluent releases; and

• dose assessment procedures.

b. Findings

No findings of significance were identified.

Enclosure

22

.3 Emergency Preparedness

a. Inspection Scope (71151 - 3 Samples)

The inspectors reviewed Entergys procedure for developing the data for the EP PIs

which are: (1) Drill and Exercise Performance (DEP), (2) ERO Drill Participation, and

(3) ANS Reliability. The inspectors also reviewed Entergys drill and exercise reports,

training records, and ANS testing data to verify the accuracy of the reported data. Data

generated since the June 2004 EP PI verification was reviewed during this inspection.

Therefore, data submitted from the second quarter of 2004 through the end of the third

quarter of 2005 were reviewed. The review was conducted in accordance with NRC

Inspection Procedure 71151. The acceptance criteria used for the review were 10 CFR

50.9 and NEI 99-02, Revision 3, Regulation Assessment Performance Indicator

Guideline.

d. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

.1 Daily Review

a. Inspection Scope (71152)

As required by Inspection Procedure 71152, Identification and Resolution of Problems,

and in order to help identify repetitive failures or specific human performance issues for

follow-up, the inspectors screened all items entered into Entergys corrective action

program. This review was accomplished by reviewing copies each condition report

(CR).

b. Findings

No findings of significance were identified.

.2 Semi-annual Trend Review

a. Inspection Scope (71152 - 1 sample)

The inspectors performed a semi-annual review to identify trends that might indicate the

existence of a more significant safety issue. The inspectors included in this review

repetitive or closely related issues that may have been documented by Entergy outside

of the normal CAP, such as trend reports, PIs, major equipment problem lists,

maintenance rework lists, departmental challenges, system health reports, maintenance

rule assessments, and maintenance and CAP backlogs.

Enclosure

23

The inspectors reviewed Entergys CAP database during 2005 in order to assess the

total number and significance of CRs written in various subject areas such as equipment

or processes, and to discern any notable trends in these areas. The CRs entered into

the CAP in all quarters included those written as a result of NRC findings. This semi-

annual review represented one inspection sample.

b. Findings

No findings of significance were identified.

.3 Identification and Resolution of Problems - Emergency Preparedness

a. Inspection Scope (71152 - 2 samples)

The inspectors reviewed Entergys corrective actions for recent problems associated

with components used to actuate the siren system. These problems included

malfunctions of the frame relay telephone network that connects the county actuation

points with the siren system (primary actuation method) and problems associated with

the radio system (back-up actuation method). The inspectors reviewed CR evaluations

and associated root or apparent cause reports, and interviewed licensee and contractor

personnel responsible for maintenance of the siren system and the corrective action

program. The inspection was conducted per NRC Inspection Procedure 71152. The

applicable emergency preparedness planning standards, 10 CFR 50.47(b) and the

requirements of 10 CFR 50 Appendix E were used as reference criteria.

Introduction: The inspectors identified a Green finding for a failure to implement timely

corrective actions for multiple frame relay system problems.

Description: While reviewing documentation pertaining to an August 5, 2005, frame

relay problem, the inspectors noted nine condition reports referenced in Entergys higher

tier apparent cause (CR-IP2-2005-3345) for various frame relay problems dating back to

September 23, 2003. Following the inspection, Entergy identified an additional 13 CRs

pertaining to frame relay issues, the oldest going back to March 21, 2003. The

inspectors found that the evaluation and corrective actions described in CR IP2-2005-

3345 following the August 5, 2005, frame relay system failure to be appropriate.

Entergys corrective action program, as described in procedure, EN-LI-102, Corrective

Action Process, Revision 3, groups CRs into significance categories. Category C and

D issues can be closed by fixing the immediate problem or by confirming that the

condition has been corrected. Category B CRs require an apparent cause evaluation be

conducted to address the apparent causes for the failures. The inspectors noted that

the nine frame relays system CRs, referenced in CR-IP2-2005-3345, had been

characterized as Category C or D. The inspectors also noted that Entergy had not

performed any type of apparent cause evaluation to identify the underlying causes

and to prevent recurrence of these repeat unplanned frame relay system outages.

Enclosure

24

Entergy disagreed with the characterization of this finding and stated that the frame

relay system availability had improved during the period and also that there was no

identified connection between the frame relay problems dating back to 2003 and the

corrective actions identified in the recent apparent cause evaluation (CR IP2-2005-

3345). The inspectors considered this information and concluded that Entergy should

have acted in a more timely manner to identify and correct the underlying causal factors

that led to the earlier frame relay system outages. The failure to implement a timely and

thorough evaluation of these failures adversely impacted the reliability of the frame relay

system.

Analysis: The performance deficiency involved the failure to implement timely corrective

actions to prevent repeat unplanned failures of the frame relay system. This finding was

determined to be more than minor because the finding is associated with the EP

cornerstone attribute of Facilities and Equipment (alert and notification system

availability). It affected the cornerstone objective of ensuring that Entergy is capable of

implementing adequate measures to protect the health and safety of the public in the

event of a radiological emergency. This finding is not suitable for Significance

Determination Process evaluation but has been reviewed by NRC management and is

determined to be a finding of very low safety significance. This issue is not greater than

Green because of the short periods that the frame relay system was unavailable and

because the ANS design included a secondary method (i.e., back-up radio system).

This finding was associated with the Problem Identification and Resolution cross-cutting

area because it was related to Entergys failure to implement timely corrective actions

for reliability issues with the frame relay system.

Enforcement: No violation of regulatory requirements occurred. This finding of very low

significance was entered into Entergys corrective action process (CR-IP2-2005-4475).

FIN 05000247/2005-05-06; Inadequate Corrective Actions for Frame Relay System

Problems.

Introduction: An NCV of 10 CFR 50.72(b)(3)(xiii) was identified for not formally reporting

a siren system problem that occurred on August 5, 2005.

Description: At about 0830 on August 5, 2005, Entergy identified a frame relay system

problem that prevented use of the primary siren system actuation method from the

Putnam County warning point. Entergy contacted the vendor to correct this condition.

At about 1200, Entergy noted that the frame relay system, used to provide the primary

siren system actuation method for all four counties located in the Indian Point

emergency planning zone, was out of service. Entergy again contacted the vendor to

effect repairs and learned that the entire frame relay system had been inoperable since

approximately 0901. Entergy checked the back-up radio activation system for each of

the four counties at about 1200 and identified that the radio activation system for

Westchester County was non-functional. The back-up radio system for the remaining

three counties had remained functional. The frame relay system for all four counties

was restored at about 1435 and the back-up radio system for Westchester County was

restored by 1820. The inspectors determined that the primary and back-up systems

Enclosure

25

relied upon to actuate the sirens in Westchester County had been non-functional from

about 0900 to 1435 (approximately five and one-half hours).

Entergy reported that the counties were informed regarding the above actuation system

problems and the NRC was also informally notified regarding the above actuation

system problems. The inspectors questioned why Entergy did not formally report the

problem associated with actuation of the Westchester County sirens per 10 CFR 50.72.

Entergy indicated that formal reporting of this problem was not required since manual

actuation from the Indian Point emergency operations facility was available to actuate

the sirens upon a request by the County. The inspectors reviewed the procedures,

protocols, and practices that would have been relied upon to implement this alternate

manual siren actuation method at the time of the loss of the normal and back-up siren

actuation methods for Westchester County and noted that they were not described in

any formal procedure or documentation. The inspectors also questioned how Entergy

had practiced or demonstrated the manual notification method and were informed that

the manual actuation method was attempted during a practice exercise in 2004.

Entergy documented (CR IP2-2005-3245) several problems associated with the manual

activation method including the lack of formal guidance and protocol for Indian Point and

County emergency response staff. Subsequent to this event, Entergy implemented

appropriate corrective actions to address the identified manual actuation system

problems.

The inspectors reviewed the process for manual actuation of the sirens and considered

that successful actuation would have involved a series of tasks including identification of

an actuation problem and effective interaction with county representatives to obtain

direction and permission to actuate the sirens. The inspectors reviewed NUREG-1022

which provided guidance for reportabilty under 10 CFR 50.72, and noted that Entergy

could refrain from reporting emergency notification system problems based upon, the

existence of procedures or practices to compensate for the lost emergency sirens. The

inspectors determined that Entergy should have reported the Westchester County siren

system actuation problem per 10 CFR 50.72 based on the lack of formal procedures for

using the manual actuation method and also based on the limited experience and

practices where Entergy had demonstrated use of this method. Entergy disagreed with

this conclusion and indicated that the manual actuation method should be considered a

practice as described in NUREG-1022. The inspectors, in coordination with specialists

from the Office of Nuclear Reactor Regulation and from the Office of Nuclear Security

and Incident Response, reviewed Entergys position and concluded that Entergy should

have formally reported the siren system problem for the Westchester County sirens per

10 CFR 50.72.

Analysis: The performance deficiency involved the failure to formally notify the NRC

regarding a siren system actuation problem as required by 10 CFR 50.72. This

deficiency was evaluated using the traditional enforcement process since the failure to

make a required report could adversely impact the NRCs ability to carry out its

regulatory mission.

Enclosure

26

While reviewing this finding, the inspectors considered the short duration of the siren

system problem, the fact that the NRC was informally notified, that back-up route

alerting was available, and also that the capability to actuate the sirens via the manual

siren initiation method was not lost. The inspectors also noted that subsequent to this

event Entergy implemented corrective actions to formalize the manual siren system

actuation method as described in CR IP2-2005-3245. The inspectors considered the

above and evaluated the severity of this violation using the criteria contained in

Supplement I - Reactor Operations and Section VI.A.1 of the NRCs Enforcement Policy

and determined that this finding met the criteria for disposition as a non-cited violation.

Enforcement: 10 CFR 50.72(b)(3)(xiii) requires that problems associated with operation

of the off-site notification system be reported to the NRC. Contrary to the above, on

August 5, 2005, Entergy did not formally report a problem that affected the primary and

back-up actuation systems for the sirens located in Westchester County. This is a

violation of 10 CFR 50.72(b)(3)(xiii). Because this finding met the criteria contained in

Section VI.A.1 of the NRCs Enforcement Policy, it is being dispositioned as a non-cited

violation. NCV 05000247/2005-05-07; Failure to Make a 10 CFR 50.72(b)(3)(xiii)

Notification.

.4 PI&R Annual Sample - Fire Brigade

a. Inspection Scope (71152 - 1 sample)

The inspector reviewed CRs pertaining to the fire brigade which were generated during

calender year 2005. The inspector also reviewed procedures controlling fire brigade

activities, reviewed fire brigade drill reports and attendance sheets, and discussed fire

brigade performance with the Site Fire Protection Engineer to determine whether

Entergy was identifying areas for improvement and entering them into the corrective

action program.

b. Findings and Observations

No findings were identified. During calender year 2005, Indian Point generated 65 CRs

related to fire brigade activities. Issues included equipment, manning, drill response

times, and access through locked doors on-site, among others. Fire brigade response

times were noted to be outside the administrative requirements on several drills, from

several areas both within and outside the protected area. Fire protection personnel are

monitoring response times for each brigade member, and recording both the response

time and the area of the plant from which the individual responded. This effort was

intended to continue through the end of the year when results were to be evaluated for

additional enhancements. Actions taken to date include having operators outside the

protected area taking logs remain in direct contact with the control room, providing a

company vehicle to operators leaving the protected area on assigned duties, and having

the brigade member pre-stage personnel protective equipment near the security access

point. Assigned fire brigade members are prohibited from attending training outside the

protected area.

Enclosure

27

Several instances were identified where fire brigade members were unable to enter

areas of the facility to respond to alarms due to locked doors. This delayed response to

smoke detector alarms on more than one occasion. When problems were identified,

actions were taken to provide appropriate keys or keycards to the brigade. In the event

of an actual fire, the brigade could use the available forcible entry tools to gain access.

On those occasions where deficiencies were identified with regard to fire brigade

equipment, actions were implemented to restore the inventory and condition of

equipment to that required.

4OA4 Cross-Cutting Aspects of Findings

Section 1R13 describes a finding related to improper implementation of a temporary

alteration on a safety-related component that had a cross-cutting aspect in the area of

human performance. Specifically, maintenance personnel exceeded the maximum

allowable material removal allowed by the modification package while grinding the cap

screw heads on FCV-447.

Section 1R13 describes a finding related to a work control procedure which

inappropriately allowed a modification to be implemented on a safety-related component

prior to the completion of an adequate evaluation. This finding had a cross-cutting

aspect in the area of human performance because the decision to implement the

modification was based on inappropriate procedural guidance.

Section 4AO2 describes a finding related to Entergys failure to take timely corrective

actions for a series of problems with the frame relay portion of the alert and notification

system. This finding had a cross-cutting aspect in the area of problem identification and

resolution because Entergy properly evaluate a problem with the ANS system, and as a

result did not take timely and effective corrective actions.

4OA6 Meetings, including Exit

Exit Meeting Summary

On January 11, 2006, the inspectors presented the inspection results to Mr. P. Rubin

and other Entergy staff members, who acknowledged the inspection results presented.

The inspectors asked Entergy what materials examined during the inspection should be

considered proprietary. No proprietary information is presented in this report.

ATTACHMENT: SUPPLEMENTAL INFORMATION

Enclosure

A-1

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

F. Dacimo Site Vice President

P. Rubin Plant Manager

J. Comiotes Director, Nuclear Safety Assurance

A. Vitale Site Operations Manager

T. Barry Security Manager

P. Conroy Manager, Licensing

R. DeCensi Technical Support Manager

F. Inzirillo Emergency Planning Manager

S. Petrosi Manager, Design Engineering

J. Ventosa Engineering Manager

C. Wend Radiation Protection Manager

T. Orlando Systems Engineering Manager

D. Mayer Unit 1 Project Manager

G. Hinrichs Project manager

J. Tuohy Manager, Cable Separation Program Improvements

J. Baker Shift Manager

P. Studley Shift Manager

H. Santis Project Construction Manager

G. Schwartz ISFSI Project Manager

G. Dean Assistant Operations Manager - Training

R. Drake Supervisor, Mechanical Design Engineering

V. Andreozzi Systems Engineering Electrical Systems Supervisor

D. Gately Assistant Radiation Protection Superintendent

M. Sicard I&C Superintendent

L. Lee Systems Engineering Primary Systems Supervisor

J. Raffaele Supervisor, Electrical Design Engineering

E. Anderson Lead Engineer, Cable Separation Program Improvements

T. Beasley System Engineer

C. Bergeren In-Service Testing Engineer

J, Bretti System Engineer

T. Chan System Engineer

G. Dahl Technical Specialist, Licensing

R. Daley System Engineer

T. Foley System Engineer

D. Friedle System Engineer

M. Johnson System Engineer

T. Jones Nuclear Safety/Licensing Specialist, Licensing

T. Lowe System Engineer

W. Mahlmeister Technical Lead, Cable Separation Program Improvements

B. Meek System Engineer

P. Peloquin Project Engineer

Attachment

A-2

B. Rokes Licensing Engineer

J. Skonieczny Project Engineer

D. Smith Scheduling and Work Order Coordinator

A. Stewart Licensing

R. Sutton System Engineer

S. Wilkie System Engineer

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000247/2005005-01 URI Emergency Diesel Generator Building Flooding

(Section 1R06)

Opened and Closed

05000247/2005005-02 FIN Failure to Maintain Design Control of Control Rod

Drive Mechanism Fans (Section 1R12)05000247/2005005-03 NCV Failure to Follow Procedural Requirements During

Modification of a Safety-Related Valve (Section

1R13)05000247/2005005-04 NCV Inadequate Procedure for Control of Work on

Safety-Related Components (Section 1R13)05000247/2005005-05 NCV Inadequate Equipment to Assess Threshold for

Emergency Action Level 8.4.3. (Section 1EP4)05000247/2005005-06 FIN Inadequate Corrective Actions for Frame Relay

System Problems05000247/2005005-07 NCV Failure to Make a 10 CFR 50.72(b)(3)(xiii)

Notification

LIST OF DOCUMENTS REVIEWED

Section 1R01: Adverse Weather Protection

Procedures

OAP-008, Severe Weather Preparations, Revision 0

2-SOP-20.2, Condensate System Operation, Revision 40

2-SOP-11.5, Space Heating and Winterization, Revision 31

OAP-048, Seasonal Weather Preparations, Revision 2

Attachment

A-3

SOP-30.1, Electric Heat Trace System, Revision 24

Condition Reports

IP2-03-02655 IP2-04-06749 IP2-05-04130

IP2-04-06700 IP2-05-00378 IP2-05-04667

Work Orders

IP2-03-20868 IP2-03-21069 IP2-04-24084

IP2-03-21054 IP2-03-21591 IP2-04-24091

Section 1R04: Equipment Alignment

Procedures

2-AOP-CCW-1, Loss of Component Cooling Water, Revision 1

2-COL-27.3.1, Diesel Generators, Revision 25

2-SOP-4.1.2, Component Cooling System Operation, Revision 29

2-COL-31.3, Gas Turbine 3, Revision 6

Condition Reports

IP2-04-00213 IP2-04-01706 IP2-05-01179

IP2-04-01328 IP2-05-00541

Drawings

9321-F-2028-36, Jacket Water to Diesel Generators, Revision 1

9321-H-2029-49, Starting Air to Diesel Generators, Revision 11

9321-F-2030-39, Fuel Oil to Diesel Generators, Revision 39

302775-04, Fuel Oil System Flow Diagram

304122-05, Fuel Forwarding System Flow Diagram

UFSAR Figure 9.3-1, Auxiliary Coolant System - Flow Diagram Sheet 1, Revision 17B

UFSAR Figure 9.3-1, Auxiliary Coolant System - Flow Diagram Sheet 2, Revision 17B

UFSAR Figure 9.3-1, Auxiliary Coolant System - Flow Diagram Sheet 3, Revision 1

Miscellaneous

CCWS DBD, Design Bases Document for the Component Cooling Water System, Revision 0

Section 1R05: Fire Protection

Procedures

PFP-207, General Floor Plan - Primary Auxiliary Building, Revision 0

Attachment

A-4

PFP-209, Component Cooling Pump Room - Primary Auxiliary Building, Revision 0

PFP-255D, Boiler Feed Pumps - Turbine Building, Revision 0

PFP-252, Cable Spreading Room - Control Building, Revision 0

PFP-211, General Floor Plan - Primary Auxiliary Building, Revision 0

Section 1R06: Flood Protection Measures

Condition Reports

IP2-03-04868 IP2-05-00543 IP2-05-04686

Miscellaneous

IPEEE (1995) Section 5, Internal Flooding

IP2 PSA Section 4.7, Internal Flooding Analysis

Section 1R07: Heat Sink Performance

Calculations

PGI-00089-00, 22 Instrument Air Closed Cooling Heat Exchanger Performance, Revision 0

PGI-00090-00, 21 Instrument Air Closed Cooling Heat Exchanger Performance, Revision 0

Condition Reports

IP2-05-00674

Heat Exchanger Test Reports

Report 21-87, Eddy Current Inspection Results for 21 Instrument Air Closed Cooling Heat

Exchanger, June 8, 2004

Report 21-106, Eddy Current Inspection Results for 22 Instrument Air Closed Cooling Heat

Exchanger, February 14, 2005

SE-330, Attachment III, Heat Exchanger Inspection Report for 22 Instrument Air Closed

Cooling Water Heat Exchanger, February 14, 2005

Miscellaneous

Indian Point Response to Generic Letter 89-13, Service Water System Problems Affecting

Safety-Related Equipment, February 2, 1990

Indian Point 2 Maintenance Rule Basis Document - Instrument Air Closed Cooling Water

Work Orders

IP2-02-60313

IP2-03-11687

Attachment

A-5

Section 1R11: Operator Requalification

Miscellaneous

Lesson Plan IP2LPLORAOP003, Safe Shutdown/ Control Room Inaccessibility Review and

Drill

Section 1R12: Maintenance Effectiveness

Maintenance Rule (A)(1) Evaluations

Maintenance Rule Action Plan, Gas Turbine Reliability, Revision 6

Maintenance Rule Action Plan, Control Room HVAC System

Maintenance Rule Action Plan, Component Cooling Water System

Maintenance Rule Action Plan, Vapor Containment System, July 2005

Maintenance Rule Action Plan, Rod Control System

Maintenance Rule Action Plan, CVCS System, June 2005

Maintenance Rule Action Plan, IP2 and IP3 Structural Monitoring Program

Maintenance Rule Basis Documents

Maintenance Rule SSC Basis Document:: Auxiliary Feedwater System, Revision 4

Maintenance Rule SSC Basis Document:: Chemical and Volume Control System, Revision 2

Maintenance Rule SSC Basis Document:: Nuclear Instrumentation System

Maintenance Rule SSC Basis Document:: Control Rod Drive System

Maintenance Rule SSC Basis Document:: 345 kV Electrical System

Maintenance Rule SSC Basis Document:: Emergency Lighting System

Maintenance Rule SSC Basis Document:: Heating, Ventilation, and Air Conditioning, Revision 2

Maintenance Rule SSC Basis Document:: Control Rod Drive Fans, Revision 2

Maintenance Rule SSC Basis Document:: Component Cooling Water System, Revision

Administrative Documents

ENN-DC-121, Maintenance Rule, Revision 2

ENN-DC-171, Maintenance Rule Monitoring, Revision 2

ENN-DC-172, Maintenance Rule (a)(3) Periodic Assessment, Revision 0

ENN-MS-S-008, ENN Engineering Standard - Action Plans, Revision 0

IP3-LO-2005-00208, Maintenance Rule Periodic Assessment, June 2005

IP-SMM-WM-100, Work Control Process, Revision 4

Condition Reports

IP2-03-06179 IP2-04-03669 IP2-05-00794

IP2-03-07485 IP2-04-04366 IP2-05-01101

IP2-04-02334 IP2-05-00211 IP2-05-01350

IP2-04-02668 IP2-05-00212 IP2-05-01402

IP2-04-02668 IP2-05-00568 IP2-05-01662

Attachment

A-6

IP2-05-01691 IP2-05-02210 IP2-05-03670

IP2-05-01691 IP2-05-02717 IP2-05-04580

IP2-05-01884 IP2-05-02729 IP2-05-04704

IP2-05-01908 IP2-05-03018 IP2-05-04744

IP2-05-01944 IP2-05-03206

Miscellaneous

ES-0.2 Deviation, Natural Circulation Cooldown, Revision 17

Indian Point Energy Center Maintenance Rule Program Quarterly Report second Quarter 2005

Shulz Electric Motor Failure Report, Q-9993

Maintenance Rule Expert Panel Meeting #2005-03 Minutes, May 9, 2005

Maintenance Rule Expert Panel Meeting #2005-04 Minutes, May 23, 2005

IPEC Maintenance Rule Program Quarterly Report, First Quarter 2005

IPEC Maintenance Rule Program Quarterly Report, Second Quarter 2005

IPEC Maintenance Rule Program Quarterly Report, Fourth Quarter 2005

Top Ten Technical Issue - IPEC NI System Reliability, Revision 5, May 2005

Top Ten Equipment Reliability Action Plan - Control Rod Drive System

Performance Criteria/Goal Evaluation - 345 kV System

System Health Reports

Auxiliary Feedwater System Health Report, Third Quarter 2005

CVCS System Health Report, Third Quarter 2005

Control Rod Drive Annual System Health Report, 2004

345 kV Annual System Health Report, 2004

Control Room HVAC System Health Report, Third Quarter 2005

Component Cooling Water System, First Quarter 2004

Component Cooling Water System, Second Quarter 2004

Component Cooling Water System, Fourth Quarter 2004

Component Cooling Water System, Second Quarter 2005

Component Cooling Water System, Third Quarter 2005

HVAC Annual Report, June 2004 to June 2005

Section 1R13: Maintenance Risk Assessment and Emergent Work Control

Work Orders

IP2-05-25862

Calculations

1487203-C-003, Assessment of Actuator Bolting with Shaved Cap Screw Heads

IP3-Calc-FW-02760, Allowable Margin for Cap Screw Grinding

FPX-00281-00, Nitrogen Supply Capacity Requirement of an Accumulator for Power-Operated

Relief Valves PCV-455C and PCV456, Revision 0

Attachment

A-7

Condition Reports:

IP2-05-03847 IP2-05-04544 IP2-05-04671

IP2-05-04137 IP2-05-04615 IP2-05-04926

IP2-05-04354 IP2-05-04666 IP2-05-04979

IP2-05-04408

Miscellaneous

Operators Daily Risk Report 10/28/05

IEEE Standard C37.106-2003, Guide for Abnormal Frequency Protection for Power Generating

Plants

Westinghouse I.L. 14429-A, Instructions of A/200 Series Mechanical Interlocks

Regulatory Guide 1.33, Quality Assurance Program Requirements, Revision 2

TA-05-2-107, Grind the socket head cap screws to facilitate packing adjustment

Procedures

2-ARP-SKF, Bearing Monitor, Revision 23

EN-WM-100, Work Request Generation, Screening and Classification, Revision 0

0-VLV-416-AOV, Copes-Vulcan Steam Generator Feedwater Regulating Valve Maintenance,

Revision 1

Drawings

A225103-08, Logic Diagram Steam Generator Trip Signals

D252556-4, Flow & Pressure Channel 1 (SG #22)

Section 1R14: Non-Routine Events

Procedures

2-TOP-006, Permanent Tave Increase to 565F, Revision 0

OAP-030, Infrequently Performed Tests and Evolutions, Revision 0

2-POP-1.3, Plant Startup, Mode 2 to Mode 1, Revision 69

2-POP-3.1, Plant Shutdown, Mode 1 to Mode 3, Revision 47

2-SOP-26.4, Turbine Generator Startup, Synchronizing, Voltage Control and Shutdown,

Revision 48

2-ARP-SCF, Condensate and Boiler Feed, Revision 35

2-ARP-SAF, Reactor Coolant System, Revision 34

Miscellaneous

IP-EP-AD13 Attachment 2, EAL Technical Bases, Revision 2

Operator Logs November 30, 2005

Operator Logs December 22-23, 2005

Attachment

A-8

Section 1R15: Operability Evaluations

Calculations

IP-CALC-05-00951, Evaluation of an Increase in SG Feedwater Regulating Valve Stroke Time

Revision 0

CN-CRA-02-38, Indian Point 2 Steamline Break Inside Containment with Feedwater Valve

Failure, Revision 0

CN-CRA-03-20, Indian Point 2 Steamline Break Inside Containment Analysis for Stretch Power

Uprate, Revision 0`

Condition Reports

IP2-04-01075 IP2-05-04246 IP2-05-04642

IP2-04-06818 IP2-05-04414 IP2-05-04792

IP2-04-06837

Drawings

A235296, Flow Diagram - Safety Injection System, Revision 65

B206909-5, Inservice Inspection Isometric of Safety Injection Line 361 - Inside Containment

(Sheet 1 of 2) (RHR Return), November 21, 1975

D252680, EDGs Jacket Water and Lube Oil Coolers Cooling Water System, Revision 4

Miscellaneous

10 CFR 50.59 Evaluation EVL-IP2-05-26384, Operation of Feedwater Bypass BFD-90 Series

and 417L Series Valves at Stretch Uprate Power, Revision 0

UFSAR Section 14.2.5, Rupture of a Steam Pipe, Revision 17

IP2-RPT-05-00112, Outside VC Cable Tray Walkdown Summary

Procedures

2-PT-M108, RHR/SI System Venting, Revision 2

2-PT-Q013, Inservice Valve Tests, Revision 36

Work Orders

IP2-05-02442

Section 1R19: Post Maintenance Testing

Condition Reports

IP2-03-05294 IP2-05-04125 IP2-05-04138

IP2-04-06150 IP2-05-04137 IP2-05-04795

Attachment

A-9

Work Orders

IP2-04-12427 IP2-05-02522 IP2-05-25316

IP2-04-35956 IP2-05-20640 IP2-05-26680

IP2-05-02517 IP2-05-21932 IP2-05-27328

IP2-05-02521

Drawings

IP2-S-000221-01, RHR Pump Suction from Containment Pump Vaive-MOV 885A

9321-F-3006-92, Single Line Diagram 480V MCC 26A and 26B

Miscellaneous

IEEE Std 43-2000, Recommended Practice for Testing Insulation Resistance of Rotating

Machinery

Procedures

2-PT-Q013-DS249, Valve PCV1135 IST Data Sheet, Revision 24

2-PT-Q026D, 24 Service Water Pump, Revision 8

2-PT-Q030A, 21 Component Cooling Water Pump, Revision 12

BAT-C-001-A, Replacement of Battery Cells, Revision 8

PT-A35C, 23 Station Battery Intercell Resistance Checks, Revision 1

PT-Q17E, Alternate Safe Shutdown Supply Verification to 24 SWP, Revision 8

PMP-006-CVCS, Replacement of Fluid Cylinder Valves - Union QX-300 Charging Pump,

Revision 6

PT-Q33A, 21 Charging Pump, Revision 9

PT-Q68A, 21 Charging Pump Check Valves, Revision 3

2-PT-Q026E, 25 Service Water Pump, Revision 9

2-PT-Q013, Inservice Valve Tests, Revision 36

Section 1R22: Surveillance Testing

Procedures

2-PT-SA067, Main Turbine Stop and Control Valves, Revision 1

2-PT-M021A, Emergency Diesel Generator 21 Load Test, Revision 14

2-SOP-27.3.1.1, 21 Emergency Diesel Generator Manual Operation, Revision 13

PI-3Y2A, Inservice Inspection Pressure Tests - 23 AFP Suction & Discharge, Revision 1

Condition Reports

IP2-05-04269

IP2-05-04504

Attachment

A-10

Drawings

UFSAR Figure 10.2-7, Flow Diagram Boiler Feedwater

Section 1R23: Temporary Alterations

Calculations

IP-CALC-05-1032, Evaluation of Leak Repair Injection for PCV-1135, 22 S/G Atmospheric

Relief, Revision 0

Condition Reports

IP2-05-04912

Engineering Requests

IP2-05-27331-RS, Leak Repair on PCV-1135 Lubricator Nipple,

Procedures

0-LKR-401-GEN, Temporary On-Line Leak Repairs, Revision 0

ENN-ME-S-001, Attachment 7.3, Leak Repair Evaluation, Revision 0

Miscellaneous

10 CFR 50.59 Screening for PCV-1135 Leak Repair

Work Orders

IP2-05-02498

Section 1EP4: Emergency Action Level (EAL) and Emergency Plan Changes

Procedures

IPEC-EP-AD-13, IPEC Emergency Plan Administrative Procedures, Revision 2

3-ARP-012, Panel SJF - Cooling Water and Air, Revision 44

3-AOP-SW-1, Service Water Malfunction, Revision 1

3-ARP-049, Panel Local - Intake Structure, Revision 1

Drawings

9321-F-20015, Screenwash Pump

9321-F-10113-8, Intake Structure Top Slab Plan

9321-F-10143-7, Intake Structure Miscellaneous Steel Details

Attachment

A-11

Condition Reports

IP3-05-05375 IP3-05-05388 IP3-05-05401

IP3-05-05380 IP3-05-05389

Calcalations

IP3-CALC-SWS-03622, Service Water Header Pressure

Section 4OA1: Performance Indicator Verification

Condition Reports

IP2-05-05457

Section 4OA2: Problem Identification and Resolution

Procedures

IPEC Operations Night Orders, December 6, 2005

IP-SMM-TQ-122, Fire Protection Training Program, Revision 1

IP-SMM-DC-901, IPEC Fire Protection Program Plan, Revision 2

OAP-001, Conduct of Operations, Revision 8

OASL-15.21, Shift Manning Requirements, Revision 5

OASL-15.22, Fire Brigade Requirements, Revision 7

Condition Reports

IP2-2005-00429 IP2-2005-03999 IP3-2005-02763

IP2-2005-00530 IP2-2005-04033 IP3-2005-02776

IP2-2005-00584 IP2-2005-04257 IP3-2005-02912

IP2-2005-00808 IP2-2005-04282 IP3-2005-03081

IP2-2005-01763 IP2-2005-04546 IP3-2005-03882

IP2-2005-02555 IP2-2005-04606 IP3-2005-04138

IP2-2005-02700 IP2-2005-04760 IP3-2005-05047

IP2-2005-03308 IP3-2005-00471 IP3-2005-05060

IP2-2005-03319 IP3-2005-00675 IP3-2005-05231

IP2-2005-03354 IP3-2005-00744 IP3-2005-05620

IP2-2005-03448

Miscellaneous

Standing Order 05-02

Temporary Procedure Change 03-0229

LIST OF ACRONYMS

ABFP Auxiliary Boiler Feedwater Pump

ADAMS Agencywide Document Management System

AOP Abnormal Operating Procedure

CAP Corrective Action Program

Attachment

A-12

CCP Coolant Charging Pump

CCW Component Cooling Water

CFR Code of Federal Regulations

CR Condition Report

CRDM Control Rod Drive Mechanism

CST Condensate Storage Tank

EDG Emergency Diesel Generator

EOP Emergency Operating Procedure

EP Emergency Preparedness

ESW Emergency Service Water

FSAR Final Safety Analysis Report

GT Gas Turbine

IACCW Instrument Air Closed Cooling Water

IMC Inspection Manual Chapter

IP2 Indian Point 2

IP3 Indian Point 3

IPE Individual Plant Examination

IPEEE Individual Plant Examination of External Events

MR Maintenance Rule

NCV Non-Cited Violation

NEI Nuclear Energy Institute

NRC Nuclear Regulatory Commission

PARS Publically Available Records System

PI Performance Indicator

PSA Probabilistic Safety Assessement

PWST Primary Water Storage Tank

PWT Post Work Test

RG Regulatory Guide

RWP Radiation Work Permit

RWST Refueling Water Storage Tank

SDP Significance Determination Process

SFP Spent Fuel Pool

S/G Steam Generator

SI Safety Injection

SWP Service Water Pump

Tave Average Coolant Temperature

TRM Technical Requirements Manual

TS Technical Specification

URI Unresolved Item

WO Work Order

Attachment