Text

IR 05000286-09-002, on 01/01/2009 - 03/31/2009; Indian Point Nuclear Generating Unit 3; Radiological Access Control
	ML091320023
Person / Time
Site:	Indian Point
Issue date:	05/11/2009
From:	Mel Gray; Reactor Projects Branch 2
To:	Joseph E Pollock; Entergy Nuclear Operations
	gray mel
References
	IR-09-002
	Download: ML091320023 (37)
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UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION I

475 ALLENDALE ROAD

KING OF PRUSSIA, PA 19406-1415

May 11, 2009

Mr. Joseph E. Pollock

Site Vice President

Entergy Nuclear Operations, Inc.

Indian Point Energy Center

450 Broadway, GSB

Buchanan, NY 10511-0249

SUBJECT: INDIAN POINT NUCLEAR GENERATING UNIT No. 3 - NRC INTEGRATED

INSPECTION REPORT 05000286/2009002

Dear Mr. Pollock:

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

at Indian Point Nuclear Generating Unit No. 3. The enclosed integrated inspection report

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

members of your staff.

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

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

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

interviewed personnel.

This report documents one finding of very low safety significance (Green), which was also

determined to be a violation of NRC requirements. However, because of the very low safety

significance, and because the finding was entered into your corrective action program, the NRC

is treating the finding as a non-cited violation (NCV), consistent with Section VI.A.1 of the NRC

Enforcement Policy. If you contest this NCV, you should provide a written response within 30

days of the date of this inspection report, with the basis for your denial, to the Nuclear

Regulatory Commission, ATTN.: Document Control Desk, Washington D.C. 20555-0001; with

copies to the Regional Administrator, Region I; the Director, Office of Enforcement, United

States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Senior

Resident Inspector at Indian Point Nuclear Generating Unit 3. In addition, if you disagree with

the characterization of this finding, you should provide a response within 30 days of the date of

this inspection report, with the basis for your disagreement, to the Regional Administrator,

Region I, and the NRC Resident Inspectors at Indian Point Nuclear Generating Unit 3. The

information you provide will be considered in accordance with Inspection Manual Chapter 0305.

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) Part 2.390 of the

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

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

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

J. Pollock 2

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

Public Electronic Reading Room).

Sincerely,

/RA/

Mel Gray, Chief

Projects Branch 2

Division of Reactor Projects

Docket No. 50-286

License No. DPR-64

Enclosure: Inspection Report No. 05000286/2009002

w/ Attachment: Supplemental Information

cc w/encl:

Senior Vice President, Entergy Nuclear Operations

Vice President, Operations, Entergy Nuclear Operations

Vice President, Oversight, Entergy Nuclear Operations

Senior Manager, Nuclear Safety and Licensing, Entergy Nuclear Operations

Senior Vice President and COO, Entergy Nuclear Operations

Assistant General Counsel, Entergy Nuclear Operations

Manager, Licensing, Entergy Nuclear Operations

P. Tonko, President and CEO, New York State Energy Research and Development Authority

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

A. Donahue, Mayor, Village of Buchanan

J. G. Testa, Mayor, City of Peekskill

R. Albanese, Four County Coordinator

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

Chairman, Standing Committee on Energy, NYS Assembly

Chairman, Standing Committee on Environmental Conservation, NYS Assembly

Chairman, Committee on Corporations, Authorities, and Commissions

M. Slobodien, Director, Emergency Planning

P. Eddy, NYS Department of Public Service

Assemblywoman Sandra Galef, NYS Assembly

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

A. Spano, Westchester County Executive

R. Bondi, Putnam County Executive

C. Vanderhoef, Rockland County Executive

E. A. Diana, Orange County Executive

T. Judson, Central NY Citizens Awareness Network

M. Elie, Citizens Awareness Network

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

Public Citizen's Critical Mass Energy Project

M. Mariotte, Nuclear Information & Resources Service

J. Pollock 3

F. Zalcman, Pace Law School, Energy Project

L. Puglisi, Supervisor, Town of Cortlandt

Congressman John Hall

Congresswoman Nita Lowey

Senator Kirsten E. Gillibrand

Senator Charles Schumer

G. Shapiro, Senator Gillibrand's Staff

J. Riccio, Greenpeace

P. Musegaas, Riverkeeper, Inc.

M. Kaplowitz, Chairman of County Environment & Health Committee

A. Reynolds, Environmental Advocates

D. Katz, Executive Director, Citizens Awareness Network

K. Coplan, Pace Environmental Litigation Clinic

M. Jacobs, IPSEC

W. Little, Associate Attorney, NYSDEC

M. J. Greene, Clearwater, Inc.

R. Christman, Manager Training and Development

A. Peterson, New York State Energy Research, SLO Designee

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

J. Pollock 3

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

Public Electronic Reading Room).

Sincerely,

/RA/

Mel Gray, Chief

Projects Branch 2

Division of Reactor Projects

Docket No. 50-286

License No. DPR-64

Enclosure: Inspection Report No. 05000286/2009002

w/ Attachment: Supplemental Information

Distribution w/encl:

S. Collins, RA J. Boska, PM, NRR P. Cataldo, SRI - Indian Point-3

M. Dapas, DRA J. Hughey, NRR A. Koonce, RI - Indian Point-3

S. Campbell, RI OEDO M. Gray, DRP ROPReport resources

R. Nelson, NRR B. Bickett, DRP RGN-I Docket Room (w/concurrences)

M. Kowal, NRR D. Bearde, DRP D. Hochmuth, DRP

D. Lew, DRP J. Clifford, DRP

R. Nelson, NRR M. Kowal, NRR

J. Boska, PM, NRR J. Hughey, NRR

SUNSI Review Complete: bab (Reviewer=s Initials)

DOCUMENT NAME: G:\DRP\BRANCH2\A - INDIAN POINT 3\INSPECTION REPORTS\2009\IP3 2009-002\IP3.2009.002.R2.DOC

ML091320023

After declaring this document AAn 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 PCataldo/PC BBickett/bab MGray/mxg

DATE 05/ 07/09 05/07/09 05/11/09

OFFICIAL RECORD COPY

1

U.S. Nuclear Regulatory Commission

Region I

Docket No.: 50-286

License No.: DPR-64

Report No.: 05000286/2009002

Licensee: Entergy Nuclear Northeast (Entergy)

Facility: Indian Point Nuclear Generating Unit 3

Location: 450 Broadway, GSB

Buchanan, NY 10511-0249

Dates: January 1, 2009 through March 31, 2009

Inspectors: P. Cataldo, Senior Resident Inspector, Indian Point 3

A. Koonce, Resident Inspector, Indian Point 3

C. Hott, Resident Inspector, Indian Point 2

J. Commiskey, Health Physicist, Region 1

E. H. Gray, Senior Reactor Inspector

M. Patel, Reactor Inspector

Approved By: Mel Gray, Chief

Projects Branch 2

Division of Reactor Projects

Enclosure

2

TABLE OF CONTENTS

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

REPORT DETAILS..................................................................................................................... 5

1. REACTOR SAFETY .............................................................................................................. 5

1R01 Adverse Weather Protection ................................................................................... 5

1R04 Equipment Alignment ............................................................................................. 5

1R05 Fire Protection ........................................................................................................ 6

1R06 Flood Protection Measures ..................................................................................... 6

1R07 Heat Sink Performance ) ......................................................................................... 7

1R08 In-service Inspection ............................................................................................. 7

1R11 Licensed Operator Requalification Program .......................................................... 9

1R12 Maintenance Effectiveness .................................................................................... 9

1R13 Maintenance Risk Assessments/Emergent Work Control .................................... 10

1R15 Operability Evaluations ........................................................................................ 10

1R18 Plant Modifications .............................................................................................. 11

1R19 Post-Maintenance Testing ................................................................................... 11

1R20 Refueling and Outage Activities ........................................................................... 12

1R22 Surveillance Testing ............................................................................................ 13

2. RADIATION SAFETY ......................................................................................................... 13

2OS1 Access Control to Radiologically Significant Areas .............................................. 13

2OS2 ALARA Planning and Controls .............................................................................. 17

4. OTHER ACTIVITIES .......................................................................................................... 19

4OA1 Performance Indicator Verification ......................................................................... 19

4OA2 Identification and Resolution of Problems .............................................................. 19

4OA3 Event Followup ...................................................................................................... 20

4OA5 Other Activities ...................................................................................................... 21

4OA6 Meetings, including Exit ......................................................................................... 23

ATTACHMENT: SUPPLEMENTAL INFORMATION ................................................................ 24

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-8

Enclosure

3

SUMMARY OF FINDINGS

IR 05000286/2009-002; 01/01/2009 - 03/31/2009; Indian Point Nuclear Generating Unit 3;

Radiological Access Control.

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

One finding of very low significance (Green) was identified, which was also determined to be a

non-cited violation (NCV). The significance of most findings is indicated by their color (Green,

White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination

Process. The cross-cutting aspect for each finding was determined using IMC 0305,

Operating Reactor Assessment Program. Findings for which the significance determination

process (SDP) does not apply may be Green, or be assigned a severity level after NRC

management review. The NRCs program for overseeing safe operation of commercial nuclear

power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated

December 2006.

A. NRC-Identified and Self-Revealing Findings

Cornerstone: Occupational Radiation Safety

Green. The inspectors identified a Green non-cited violation of Technical

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

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

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

condition reports and adequately document the investigations for 21 instances of

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

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

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

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

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

01318.

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

Radiation Safety cornerstone attribute of programs and process, and adversely

affected the objective to ensure adequate protection of worker health and safety from

exposure to radiation. Moreover, the inspectors identified a programmatic deficiency

to maintain and implement programs to keep exposures as low as reasonably

achievable, because multiple examples were identified regarding the failure to satisfy

station radiation protection procedures. Using the Occupational Radiation Safety

Significance Determination Process, the inspectors determined that the finding was

of very low safety significance (Green) because it did not involve: (1) as low as is

reasonably achievable planning and controls, (2) an overexposure of an individual,

(3) a substantial potential for overexposure, or (4) an impaired ability to assess dose.

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

procedural adherence in the Work Practices component of the Human Performance

area. Specifically, Entergy employees did not follow procedures to generate

condition reports and document investigations when high dose-rate alarms were

received by workers. H.4(b) (Section 2OS1)

Enclosure

4

B. Licensee-Identified Violations

None.

Enclosure

5

REPORT DETAILS

Summary of Plant Status

Indian Point Nuclear Generating (Indian Point) Unit 3 began the inspection period at full reactor

power. On March 10, 2009, a planned downpower was initiated that culminated in the Unit

being taken off-line to begin refueling outage No. 15 (3R15). The Unit remained off-line to refuel

for the remainder of the inspection period.

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection (71111.01 - 2 samples)

.1 Impending Cold Weather Review

a. Inspection Scope

The inspectors performed a detailed review of Entergys procedures to address

impending cold weather conditions due to a forecasted arctic front on January 15, 2009.

The inspectors evaluated Entergys preparation and readiness for cold weather

conditions, evaluated applicable compensatory measures, conducted walk downs of

plant equipment, and verified that cold weather deficiencies from previous years have

been addressed. In addition, the inspectors reviewed the status of deficiencies identified

during the current seasonal preparations, and verified that adverse conditions were

being adequately addressed to ensure the impending cold weather conditions would not

have significant impact on plant operation and safety. The documents reviewed during

this inspection are listed in the Attachment. This review of cold weather preparations

represented one inspection sample.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment (71111.04Q - 3 samples)

a. Inspection Scope

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

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

where applicable, following return to service after maintenance. The inspectors

reviewed system procedures, the Updated Final Safety Analysis Report (UFSAR), and

system drawings to verify that the alignment of the applicable system or component

supported its required safety functions. The inspectors also reviewed applicable

condition reports or work orders to ensure that Entergy personnel had identified and

properly addressed equipment deficiencies that could potentially impair the capability of

the available train. The documents reviewed during this inspection are listed in the

Attachment.

Enclosure

6

The inspectors performed partial walkdowns of the following systems or components,

which represented three inspection samples:

Auxiliary feedwater system following return to service of the 31 auxiliary boiler

feedwater pump on February 20, 2009;

31 and 33 emergency diesel generators (EDG) while the 32 EDG was out-of-service

for 8-year and 16-year planned maintenance activities; and

31 and 33 safety injection (SI) pumps during planned maintenance on the 32 SI

pump on February 6, 2009.

b. Findings

No findings of significance were identified.

1R05 Fire Protection (71111.05Q - 6 samples)

a. Inspection Scope

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

operational status of applicable fire protection features. The inspectors verified,

consistent with the applicable administrative procedures, that: combustible material and

ignition sources were adequately controlled; passive fire barriers, manual fire-fighting

equipment, and suppression and detection equipment were appropriately maintained;

and compensatory measures for out-of-service, degraded, or inoperable fire protection

equipment were implemented in accordance with Entergys fire protection program. The

inspectors also evaluated the fire protection program against the requirements of

License Condition 2.K. Additionally, the inspectors reviewed the circumstances

surrounding a fire main component leak located at the header isolation valve associated

with the Outage Support Building (Fire Zones 391 and 392). The documents reviewed

during this inspection are listed in the Attachment.

This inspection represented six inspection samples and was conducted in the areas

covered by the following Pre-Fire Plans:

Pre-Fire Plan Nos. 391 and 392;

Pre-Fire Plan 306;

Pre-Fire Plan 306A;

Pre-Fire Plan 362;

Pre-Fire Plan 362A; and

Pre-Fire Plan 362B.

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures (71111.06 - 1 sample)

a. Inspection Scope

The inspectors reviewed the Unit 3 Individual Plant Examination, the UFSAR, and

IP-RPT-06-00071, "Indian Point Unit 3 Probabilistic Safety Assessment (PSA), Rev. 2,

Enclosure

7

concerning internal flooding events. The inspectors assessed flood mitigation attributes

within the turbine building that are utilized to minimize potential impacts of flooding on

the vital 480 Volt switchgear room that adjoins the turbine building. The inspectors also

reviewed a surveillance test conducted on February 3, 2009, associated with flood level

indicators in the turbine building, 3-PT-R22, "Turbine Building (Lower Level) Level

Sensors," Rev. 10. This inspection represented one sample for internal flood protection

measures.

b. Findings

No findings of significance were identified.

1R07 Heat Sink Performance (71111.07 - 1 sample)

a. Inspection Scope

The inspectors evaluated maintenance activities and reviewed inspection data

associated with periodic inspections of service water system piping. The inspectors

reviewed applicable design basis information and commitments associated with

Entergys Generic Letter 89-13 program to validate that maintenance activities were

adequate to ensure the system could perform its required safety function. The

inspectors reviewed radiographic results for selected piping segments to ensure pipe

corrosion and conditions adverse to quality were being identified and corrected. This

inspection represented one sample for heat sink performance.

b. Findings

No findings of significance were identified.

1R08 In-service Inspection (71111.08 - 1 sample)

a. Inspection Scope

Activities inspected during the Unit 3 refuel outage 15 (3R15) included observations of

ultrasonic testing (UT) calibration or in-progress component testing using manual and

computer based UT techniques. Manual UT observations included the main steam 6

diameter, loop 32, pipe welds 22 and 23 shown on drawing 2201, Rev 6, and review of

the UT data sheets for residual heat removal (RHR) pipe welds W13 and W16. The

sample of visual inspection (VT) included the areas of the containment inner boundary at

the containment liner and containment penetrations. The task work orders and test data

for several ultrasonic and visually identified indications were reviewed and confirmed to

be evaluated by Entergy personnel as part of the in-service inspection process.

The inspectors observed the video-visual examination results for a sample of the reactor

pressure vessel (RPV) upper head-to-control rod drive mechanism (CRDM) penetrations

per the EPRI guidelines. This inspection included the sequence of Entergy's evaluation

of the as-found conditions, conducted in accordance with procedure 3-PT-R203, Rev. 3,

which used a robot crawler to position a camera to view the circumference of each

CRDM for boric acid leakage. This inspection also included a comparison of the 2009

visual observations with those of the previous (2007) outage which included CRDMs 17,

24, 41, 53, 60, and 76.

Enclosure

8

The inspectors reviewed a sample of computer-based, eddy current (ET) and ultrasonic

testing (UT) records and results of the upper RPV-head-to-CRDM penetrations and weld

examinations as conducted from the underside of the RPV head.

In the area of boric acid corrosion control activities, the inspectors confirmed the extent

of boric acid walkdowns during plant operation and the plant shutdown process, and

verified that identified problem areas were documented in condition reports for

evaluation and resolution. In particular, the inspectors reviewed visual records of the as-

found and as-left conditions of a reactor vessel head mechanical penetration, Conoseal

3, which had experienced some leakage and was identified by Entergy personnel at

shutdown. The inspectors confirmed the Conoseal leakage to be from a mechanical

joint and not pressure boundary leakage that was repaired during this refueling outage.

Additionally, the inspectors evaluated the as-left condition on the RPV head in that area

and other potentially affected areas. The condition of the upper threads on vessel stud

29 and the status of eight other studs were visually inspected to confirm that no

significant degradation was present.

The inspectors noted that steam generator (SG) tube inspection results from the 2007

(3R14) outage provided the basis for not performing eddy current testing (ECT) of SG

tubes during the 3R15 outage. The inspectors reviewed the SG tube assessment

(Report IP-RPT-06-00186) for 3R14 and the documented review (Report IP-RPT-07-

00031) of the acceptability of SG operation for two cycles until 3R17. It was noted that

the operating conditions between 3R15 and 3R16 would be assessed to confirm that

those conditions were consistent with the IP-RPT-07-00031 report prior to the start of

RFO 3R16.

The inspectors reviewed computer-based ECT and UT records and examination results

of the four hot leg and four cold leg primary piping-to-reactor vessel nozzles consistent

with the dissimilar metal weld program under MRP-139. These welds were examined

under water from the inside of the reactor pressure vessel (RPV). The answers to the

applicable TI 2515/172 (temporary inspection) procedure are included in Section 4OA5

of this report. Additionally, the inspectors reviewed a sample of the computer-based

ECT and UT records and examination results of the bottom-mounted RPV penetrations

that were accessed from inside the RPV.

The inspectors reviewed the video record of the visual examinations of the three 6"

safety and one 4" pressure relief pressurizer upper cast head inner radius to nozzle

surfaces to verify the adequacy of the examination technique and to confirm the status of

the inner radius and related areas. The accessible areas around the 4" spray nozzle

were also viewed although the inner radius of the spray nozzle was not accessible. No

items of degradation were observed in any of the visually accessible areas.

The inspectors noted that the surge nozzle-to-pipe dissimilar metal, stainless steel weld,

located at the bottom of the pressurizer was ultrasonically examined after appropriate

preparation of the exterior surface by grinding flush. The inspectors examined the

grinding mockup, the as-ground condition, the engineering analysis including thickness

calculation, and the UT results.

b. Findings

No findings of significance were identified.

Enclosure

9

1R11 Licensed Operator Requalification Program (71111.11Q - 1 sample)

Quarterly Resident Inspector Evaluation

a. Inspection Scope

The inspectors observed licensed operator requalification training conducted on

February 25, 2009, in the Unit 3 plant-reference simulator. The inspectors assessed the

scope and breadth of the training, which focused on specific activities that were planned

for the Unit 3 refueling outage. In particular, the inspectors observed simulated activities

associated with the normal cooldown process that occurs following entry into the outage

as the plant transitions into lower modes of operation as defined by technical

specifications. The inspection also included the following: (1) discussions with Entergy

staff regarding deficiencies in operator performance and/or training being addressed in

the current requalification training cycle; and (2) assessment of the implementation of

abnormal operating procedures utilized by Unit 3 control room operators to respond to,

and mitigate the effects of, simulated loss of residual heat removal cooling.

The inspectors reviewed simulator fidelity to verify correlation with the actual plant

control room, and to verify that differences in fidelity that could potentially impact training

effectiveness were either identified or appropriately dispositioned. Licensed operator

training was evaluated for conformance with the requirements of 10 CFR 55, Operator

Licenses. Documents reviewed during this inspection are listed in the Attachment. This

review represented one inspection sample for licensed operator requalification training.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness (71111.12Q - 3 samples)

a. Inspection Scope

The inspectors reviewed performance-based problems that involved selected structures,

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

and to verify activities were conducted in accordance with site procedures and 10 CFR

50.65 (The Maintenance Rule). The reviews focused on:

Evaluation of Maintenance Rule scoping and performance criteria;

Verification that reliability issues were appropriately characterized;

Verification of proper system and/or component unavailability;

Verification that Maintenance Rule (a)(1) and (a)(2) classifications were

appropriate;

Verification that system performance parameters were appropriately trended; and

For SSCs classified as Maintenance Rule (a)(1), that goals and associated

corrective actions were adequate and appropriate for the circumstances.

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

Maintenance Rule basis documents. The documents reviewed during this inspection are

listed in the Attachment. The following Unit 3 systems and/or components were

reviewed and represented three inspection samples:

Enclosure

10

Intake structure;

Emergency diesel generators; and

RWST level indication system.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments/Emergent Work Control (71111.13 - 5 samples)

a. Inspection Scope

The inspectors reviewed maintenance activities to verify that the appropriate on-line and

shutdown risk assessments were performed prior to removing equipment for work as

required by 10 CFR 50.65 (a)(4). When planned work scope or schedules were altered

to address emergent or unplanned conditions, the inspectors verified that the plant risk

was promptly reassessed and managed. Additionally, the inspectors utilized IMC 0609,

Appendix G, during various refueling outage periods, to assist in the evaluation of

Entergy's shutdown risk assessments. The documents reviewed during this inspection

are listed in the Attachment. The following activities represented five inspection

samples:

Planned risk during containment fan cooler and N42 power range nuclear

instrumentation activities on January 26, 2009;

Planned risk during troubleshooting activities associated with 480-Volt safety bus

6A conducted on February 5, 2009;

Planned risk during quarterly calibrations of power range nuclear instrumentation

channels N41 and N42 on February 17, 2009;

Initial RCS drain down for reactor vessel head removal on March 13, 2009; and

Defense-in-depth contingency 3A during 138kV electrical system outage on

March 20, 2009.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations (71111.15 - 5 samples)

a. Inspection Scope

The inspectors reviewed operability evaluations to assess the acceptability of the

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

compliance with Technical Specifications. These reviews included verification that

operability determinations were performed in accordance with procedure ENN-OP-104,

Operability Determinations. The inspectors assessed the technical adequacy of the

evaluations to ensure consistency with the UFSAR and associated design and licensing

basis documents. The documents reviewed are listed in the Attachment. The following

operability evaluations were reviewed and represented four inspection samples:

CR-IP3-2009-00052: Noise from 32 service water pump bearing;

Enclosure

11

CR-IP3-2009-00138/00151: 33 control building exhaust fan deficiencies;

CR-IP3-2009-00135: Unit 3 auxiliary transformer tap changer deficiencies;

CR-IP3-2009-00408/00421: Safety injection room scaffolding deficiencies; and

CR-IP3-2008-01589: 33 emergency diesel generator past operability evaluation

(3R/4R cylinder lockout event from June/July 2008).

b. Findings

No findings of significance were identified.

1R18 Plant Modifications (71111.18 - 1 sample)

Temporary Emergency Diesel Generator (EDG) Cooling Water Modification

a. Inspection Scope

The inspectors reviewed design change documentation that supported Entergy's

installation of temporary end bells on the 31 and 33 EDG jacket water heat exchangers.

This modification diverted service water to a local storm drain to support maintenance

activities on the Service Water System. The inspectors verified that the design bases,

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

temporary modification. The inspectors verified that Entergy utilized established

procedures governing the use of temporary end bells while they were in service. In

addition, the inspectors interviewed plant staff, and reviewed issues that had been

entered into the corrective action program to determine whether Entergy had been

effective in identifying and resolving problems associated with temporary modifications.

The documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing (71111.19 - 7 samples)

a. Inspection Scope

The inspectors reviewed post-maintenance test procedures and associated testing

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

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

plant personnel. The inspectors verified that: test acceptance criteria were clear; tests

demonstrated operational readiness and were consistent with design basis

documentation; test instrumentation had current calibrations and appropriate range and

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

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

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

Post-maintenance testing was evaluated against the requirements of 10 CFR 50,

Appendix B, Criterion XI, Test Control. The following post-maintenance activities were

reviewed and represented seven inspection samples:

33 containment fan cooler air flow switch replacement on January 8, 2009;

Enclosure

12

Fuel storage building ventilation following charcoal and filter replacement on

January 26, 2009;

31 residual heat removal pump load sequence calibration on February 12, 2009;

Valve diagnostic test and calibration of MS-PCV-1134 on February 18, 2009;

32 EDG air receiver following liner installation on March 15, 2009;

32 EDG following 16-year and 8-year PMs on March 16, 2009; and

31 auxiliary boiler feedwater pump cutback controller repair.

b. Findings

No findings of significance were identified.

1R20 Refueling and Outage Activities (71111.20 - 1 partial sample)

Refueling Outage No. 15 (3R15)

a. Inspection Scope

The inspectors observed and/or evaluated the selected outage activities listed below to

verify that (1) shutdown risk was considered during schedule preparation and

implementation, and high risk significant evolutions such as mid-loop or reduced

inventory conditions; (2) defense-in-depth (DID) measures were utilized to mitigate

impacts on key safety functions (e.g., reactivity control, electrical power availability,

containment integrity, etc.) due to plant configuration control changes and ensure

compliance with technical specifications and the operating license throughout the outage

period; and (3) risk significant activities were conducted in accordance with procedures

and evaluated in a manner appropriate for the circumstances.

Fuel receipt and inspection activities; Special nuclear material (SNM) accountability

and transfer;

Plant shutdown, cooldown (in accordance with TS limits) entry into residual heat

removal operation; and refueling operations (e.g., reactor vessel head lift, core

offload, etc);

Changes in daily plant risk and implementation of DID measures;

Post-shutdown boric acid inspection inside the vapor containment to assess

effectiveness of unidentified leakage monitoring and compliance with TS;

Evaluated multiple reactor and refueling cavity draindown evolutions to verify

procedural compliance, and operability and functionality of the redundant and diverse

reactor coolant system level instrumentation;

A sample of lockout/tagouts and clearances, were reviewed to verify appropriate

controls of plant configuration changes were being implemented for the protection of

plant equipment and personnel;

Open outage constraints (work orders and condition reports) were reviewed to verify

appropriate disposition of issues, both technical and /or administratively, to ensure

compliance with procedural and/or TS requirements;

Vapor containment closure team DID measures (DID-C4) and contingency

implementation, team make-up, briefings, and inspection of staged tools;

Evaluated refueling cavity upender sheave failure and replacement activities;

Evaluated boration flowpath activities to ensure appropriate reactivity controls; and

Observed and/or evaluated several surveillance tests, which included:

Enclosure

13

o 3-PT-R145, "AMSAC System Functional Check," Rev. 14;3

o 3-PC-R62C, "Inadequate Core Cooling Monitor-86 Calibration," Rev. 12;

o 3-PC-R45, "Calibration Procedure For The Gamma-Metrics Excore Nuclear

Instrumentation System," Rev. 15;

o 3-PT-V51, "Overpressure Protection System Channel Operational Test," Rev.

2; and

o 3-PT-R003G, "31 EDG/2AT5A Interlock Test," Rev. 2.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing (71111.22 - 6 samples)

a. Inspection Scope

The inspectors witnessed performance of surveillance tests and/or reviewed test data of

selected risk-significant structures, systems, and components, to assess whether test

results satisfied Technical Specification, UFSAR, Technical Requirements Manual, and

Entergy procedure requirements. The inspectors verified that: test acceptance criteria

were sufficiently clear; tests demonstrated operational readiness and were consistent

with design basis documentation; test instrumentation had accurate calibrations and

appropriate range and accuracy for the application; tests were performed as written; and

applicable test prerequisites were satisfied. Following the tests, the inspectors verified

that the equipment was capable of performing the required safety functions. The

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

surveillance tests were reviewed and represented six inspection samples, which

includes RCS and IST surveillances:

3-PT-Q116C, 33 Safety Injection Pump Functional Test, Rev. 13, conducted on

January 28, 2009;

0-SOP-LEAKRATE-001, RCS Leakrate Surveillance, Evaluation and Leak

Identification, Rev. 1, conducted on March 9, 2009;

Bus 6A portion of 3-PT-R003B, Safety Injection System Test Breaker Sequencing/

Bus Stripping, Rev. 26, conducted on March 13, 2009;

3-PT-Q120C, "33 ABFP (Motor Driven) Surveillance And IST," Rev. 9, conducted on

January 23, 2009;

3-PT-M62C, "480V Undervoltage/Degraded Grid Protection System Bus 6A

Functional," Rev. 7, conducted on February 5, 2009; and

3-PT-R006A, "Main Steam Safety Valves Setting Test Using Set Pressure

Verification Device," Rev. 8, conducted on March 10, 2009.

b. Findings

No findings of significance were identified.

2. RADIATION SAFETY

Cornerstone: Occupational Radiation Safety (OS)

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

Enclosure

14

a. Inspection Scope

During March 23 - 27, 2009, the inspectors conducted the following activities to verify

that Entergy personnel were properly implementing physical, engineering, and

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

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

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

criteria contained in 10 CFR 20 and Entergys procedures required by the Technical

Specifications as criteria for determining compliance. During the inspection, the

inspectors interviewed the radiation protection manager, radiation protection

supervisors, and radiation workers. This inspection activity represents completion of

sixteen (16) samples relative to this inspection area.

The inspectors performed independent radiation dose rate measurements and reviewed

the following items:

Plant Walk Downs and Radiation Work Permit (RWP) Reviews

(1) Exposure-significant work areas were identified for review within radiation areas,

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

controls and surveys were reviewed for adequacy. This review included: Refuel

floor split pin and reactor head inspections, refuel floor lower internals removal

and installation, refuel floor and fuel support building fuel transport equipment

repairs requiring an underwater diver, reactor coolant pump (RCP) work including

RCP #31 impeller replacement, containment valve work including pressurizer

safety valves, various containment and auxiliary building activities.

(2) With the use of a survey instrument and assistance from a Health Physics

Technician, performed a walkdown of these areas to determine whether the

appropriate RWPs, procedure, and engineering controls were in place, and

whether surveys and postings were adequate.

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

of the plant including high radiation areas. Specified electronic personal

dosimeter alarm set points were reviewed with respect to current radiological

condition applicability, and workers were queried to verify their understanding of

plant procedures governing alarm response and knowledge of radiological

conditions in their work area.

(4) The inspectors noted there were no RWPs for airborne radioactivity areas with

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

(5) The inspectors noted there were no internal dose assessments that resulted in

actual internal exposures greater than 50 mrem CEDE. Internal assessments

were reviewed to determine adequacy and assurance that they were not in fact

equal to or greater than 50 mrem CEDE.

Problem Identification and Resolution

(6) The inspectors reviewed condition reports associated with access controls since

the last inspection in this area. Staff members were interviewed and documents

Enclosure

15

reviewed to determine whether follow-up activities were being conducted in an

effective and timely manner, commensurate with their safety and risk.

(7) For repetitive deficiencies or significant individual deficiencies in problem

identification and resolution, the inspectors determined if Entergy's assessment

activities were also identifying and addressing these deficiencies.

(8) The inspectors noted there were no events associated with performance

indicator occurrences that involved dose rates greater than 25 Rem/hour at 30

cm, dose rates greater than 500 Rem/hour at 1 meter, or unintended exposures

greater than 100 mrem TEDE (or greater than 5 Rem SDE or greater than 1.5

Rem LDE).

Job-in-Progress Reviews

(9) The inspectors observed aspects of various on-going activities to confirm that

radiological controls, such as required surveys, area postings, job coverage, and

job site preparations were conducted. The inspectors verified that personnel

dosimetry was properly worn and that workers were knowledgeable of work area

conditions. The inspectors attended pre-planning meetings for work described

earlier in the report.

(10) The inspectors reviewed the adequacy of underwater diving activities associated

with repairs to the fuel transport system, which included dosimetry requirements,

bioassay requirements and controls.

High Risk Significant, High Dose Rate High Radiation Areas (HRA) and Very HRA

Controls

(11) The inspectors reviewed the adequacy of inventory and key control for access to

LHRA and VHRA. The inspector verified that accessible LHRAs were properly

secured and posted during plant tours.

(12) The inspectors discussed with Radiation Protection supervision the adequacy of

high dose rate HRA and VHRA controls and procedures and verified that no

programmatic or procedural changes have occurred that reduce the

effectiveness and level of worker protection.

Radiation Worker Performance

(13) During observation of the work activities listed above, the inspectors evaluated

radiation worker performance with respect to the specific radiation protection

work requirements, and their knowledge of the radiological conditions in

applicable work areas.

(14) The inspectors reviewed condition reports related to radiation worker

performance to determine if an observable pattern, traceable to a similar cause

was evident.

Radiation Protection Technician Proficiency

Enclosure

16

(15) During observation of the work activities listed above, the inspectors evaluated

radiation protection technician work performance with respect to their knowledge

of the radiological conditions, the specific radiation protection work requirements

and radiation protection procedures.

(16) The inspectors reviewed condition reports related to radiation protection

technician proficiency to determine if an observable pattern traceable to a similar

cause was evident.

b. Findings

Introduction: The inspectors identified a Green non-cited violation of Technical

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

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

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

condition reports and adequately document the investigations for 21 instances of

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

2009 and March 2009.

Description: During the period January 2009 through March 2009, 21 instances of

electronic dosimeter dose rate alarms were recorded by the access control system.

During this period, Entergy personnel inconsistently utilized an informal process of

reviewing the alarms without a full investigation or approval process. Moreover, in three

of the 21 instances, the inspectors identified that no investigation or follow-up had

occurred. In some cases, the occurrences were over two months old, which the

inspectors noted would have made resultant investigations more challenging to perform.

In other cases, the alarms were not identified until the worker attempted to re-enter the

radiologically controlled area (RCA) and the access control system required manual

override to un-lock the occurrence to allow entry into the RCA. The inspectors noted

that the controlling Entergy procedure for this activity, EN-RP-203, Dose Assessment,

specifies that for a dose-rate alarm that is unanticipated or un-briefed, several actions

are required, one of which is to initiate a condition report, another is to document the

investigation using an attachment in the procedure. Contrary to EN-RP-203, for these

21 instances, no condition reports or attachments were generated with a detailed

investigation prior to the workers re-entering the radiologically controlled area. The

highest exposure received by these workers during their entry, as indicated by their

electronic dosimeter and logged by the access control system, was 33 mRem, while

most dosimeters indicated less than 1 mRem for the entry.

Analysis: The inspectors determined that the failure to generate a condition report, as

well as the failure to adequately investigate 21 unplanned or un-briefed electronic

dosimeter alarms prior to re-entry into the RCA, as required by station procedure was a

performance deficiency. This performance deficiency was within Entergy personnels

ability to foresee and correct, and should have been prevented. This issue was not

subject to traditional enforcement, in that it did not have actual safety consequence, it

was not an issue that had the potential to impact NRCs ability to perform its regulatory

function, and there were no willful aspects.

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

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

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

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

Enclosure

17

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

multiple examples were identified regarding the failure to satisfy station radiation

protection procedures. Specifically, in 21 cases, Entergy did not fully evaluate dose rate

alarms received by workers in radiologically controlled areas of the plant. Using the

Occupational Radiation Safety Significance Determination Process, the inspectors

determined that the finding was of very low safety significance (Green) because it did not

involve: (1) as low as is reasonably achievable planning and controls, (2) an

overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to

assess dose.

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

procedural adherence in the Work Practices component of the Human Performance

area. Specifically, Entergy employees did not follow procedures to generate condition

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

workers. (H.4 (b))

Enforcement. Technical Specification 5.4.1.a, Procedures, requires that Entergy

establish, implement, and maintain procedures specified in Regulatory Guide (RG) 1.33,

Revision 2, Appendix A., Section 7.e, radiation protection procedures for personnel

monitoring. Entergy procedure EN-RP-203, Revision 2, Section 5.11, requires that a

condition report be written for each unplanned or un-briefed electronic dosimeter dose-

rate alarm. Contrary to the above, the inspectors identified through a review of

electronic dosimeter log information from January 2009 through March 2009, 21

instances of unanticipated or un-briefed electronic dosimeter dose-rate alarms when the

procedure was not implemented and condition reports were not generated. Because

this finding was of very low safety significance and it was entered into the corrective

action program as CR-IP3-2009-001253 and CR-IP3-2009-001318, this violation is

being treated as an NCV, consistent with the NRC Enforcement Policy. (NCV 05000286/2009002-01, Failure to Follow Radiation Protection Procedures)

2OS2 ALARA Planning and Controls (71121.02 - 12 samples)

a. Inspection Scope

During March 23 - 27, 2009, the inspectors conducted the following activities to verify

that Entergy personnel were properly maintaining individual and collective radiation

exposures as low as is reasonably achievable (ALARA). Implementation of the ALARA

program was reviewed against the criteria contained in 10 CFR 20, applicable industry

standards, and Entergys procedures. This inspection activity represents completion of

twelve (12) samples relative to this inspection area.

Inspection Planning

(1) The inspectors reviewed pertinent information regarding cumulative exposure

history, current exposure trends, and on-going activities to assess current

performance and outage exposure challenges. The inspectors determined the

sites 3-year rolling collective average exposure.

(2) The inspectors reviewed Unit 3 outage work-related activities that occurred

during the inspection period, the associated ALARA plans, RWPs, ALARA

Committee Reviews, exposure estimates, actual exposures and post job reviews.

Work reviewed included: refuel floor split pin and reactor head inspections,

Enclosure

18

refuel floor lower internals removal and installation, refuel floor and fuel support

building fuel transport equipment repairs requiring an underwater diver, reactor

coolant pump (RCP) work, which included RCP #31 impeller replacement,

containment valve work including pressurizer safety valves, and various

containment and primary auxiliary building activities.

(3) The inspectors reviewed implementing procedures associated with maintaining

occupational exposures ALARA. This included a review of the processes used to

estimate and track work activity exposures.

Radiological Work Planning

(4) With respect to the work activities listed above, the inspectors reviewed dose

summary reports, related post-job ALARA reviews, related RWPs, exposure

estimates and actual exposures, and ALARA Committee meeting paperwork.

This review was also performed to verify that dose was appropriately managed

and evaluated by station management.

(5) ALARA work activity evaluations, exposure estimates, and exposure mitigating

requirements were reviewed for work packages previously mentioned, to verify

whether Entergy had established procedures, as well as engineering and work

controls, based on sound radiation protection principles.

(6) The inspectors compared the results achieved with the intended dose that was

established in the planning of the work. The inspectors determined the reasons

for inconsistencies between the intended and actual work activity doses and

station management awareness and involvement.

(7) The inspectors evaluated for adequacy, the interfaces between operations,

radiation protection, maintenance, maintenance planning and others for interface

problems or missing program elements.

Verification of Dose Estimates and Exposure Tracking Systems

(8) Methods for adjusting exposure estimates, or re-planning work, when

unexpected changes in scope or emergent work is encountered, were reviewed

by the inspectors for adequacy.

Job Site Inspections and ALARA Controls

(9) The inspectors reviewed work activities that present the highest radiological risk

to workers. The inspectors evaluated Entergy personnels use of engineering

controls to achieve dose reductions and to verify that procedures and controls

are consistent with ALARA reviews. Associated ALARA Plans and RWPs were

reviewed to determine if appropriate exposure and contamination controls were

being employed.

Radiation Worker Performance

(10) Through observations and interviews by the inspectors, workers and technicians

were found to be knowledgeable of the work area radiological conditions and low

dose waiting areas.

Enclosure

19

Declared Pregnant Workers

(11) The inspectors reviewed information associated with declared pregnant workers

during the assessment period and whether appropriate monitoring and controls

were being utilized to ensure compliance with 10CFR20.

Problem Identification and Resolution

(12) The inspectors reviewed elements of the Entergys corrective action program

related to implementing radiological controls to determine if problems are being

entered into the program for timely resolution.

b. Findings

No findings of significance were identified.

4. OTHER ACTIVITIES (OA)

4OA1 Performance Indicator Verification

Resident Inspector Baseline Inspection (71151 - 3 samples)

a. Inspection Scope

The inspectors reviewed performance indicator data for the cornerstones listed below

and used Nuclear Energy Institute 99-02, Regulatory Assessment Performance

Indicator Guideline, Revision 5, to verify individual performance indicator accuracy and

completeness. The documents reviewed during this inspection are listed in the

Attachment.

Initiating Events Cornerstone

Unplanned Scrams per 7000 Critical Hours;

Unplanned Power Changes per 7000 Critical Hours; and

Unplanned Scrams with Complications.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

.1 Routine Problem Identification and Resolution (PI&R) Program Review

a. Inspection Scope

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

and to identify repetitive equipment failures or specific human performance issues for

follow-up, the inspectors performed a daily screening of all items entered into Entergys

corrective action program. The review was accomplished by accessing Entergys

Enclosure

20

computerized database for condition reports, and attending condition report screening

meetings.

In accordance with the baseline inspection procedures, the inspectors selected

corrective action program items across the Initiating Events, Mitigating Systems, and

Barrier Integrity cornerstones for further follow-up and review. The inspectors assessed

Entergy personnels threshold for problem identification, the adequacy of the cause

analysis, extent of condition reviews, operability determinations, and the timeliness of

the associated corrective actions. The condition reports reviewed during this inspection

are listed in the applicable inspection sections.

b. Findings

No findings of significance were identified.

.2 Occupational Radiation Safety Cornerstone

a. Inspection Scope

The inspectors reviewed 23 corrective action condition reports associated with the

radiation protection program that were initiated between December 2008 and March

2009. The inspectors verified that problems identified by these condition reports were

properly characterized in the licensees event reporting system, and that applicable

cause and corrective actions were identified commensurate with the safety significance

of the radiological occurrences.

b. Findings

No findings of significance were identified.

.3 In-Service Inspection Activities (1R08)

a. Inspection Scope

The inspectors reviewed the extent of oversight of in-service inspection (ISI)

nondestructive examination (NDE) activities, including the topics of current ISI oversight

and surveillances. This review included a sample of issue reports, which are listed in

Attachment 1, to confirm that identified problems were being documented for evaluation

and proper resolution.

b. Findings

No findings of significance were identified.

4OA3 Event Follow-up

Loss of 480 Volt Emergency Safety Bus 6A During Surveillance Testing on January 2,

2009

Enclosure

21

a. Inspection Scope

The inspectors evaluated the response of control room personnel following the

unexpected loss of 480 Volt safeguards bus 6A that occurred during the performance of

a degraded grid/undervoltage relay surveillance test on January 2, 2009. The inspectors

reviewed plant computer data, evaluated plant parameter traces, and discussed the

event with plant personnel, to verify that plant equipment responded as expected, and to

ensure that operating procedures were appropriately implemented. The inspectors

verified that Entergys short term corrective actions were appropriate in response to the

event. This event was entered into Entergys corrective action program as CR IP3-2009-

00011.

b. Findings and Observations

No findings of significance were identified.

The inspectors noted, however, that corrective actions for the current event were tracked

in an on-going root cause evaluation for a similar event that occurred on October 9,

2008. In addition, the inspectors noted a failure analysis was planned for applicable

equipment and components, and extensive troubleshooting was planned for the current

3R15 outage period.

4OA5 Other Activities

.1 Continued Groundwater Sampling Effort to Monitor Tritium (Deviation Memorandum

Inspection)

a. Inspection Scope

During the week of March 23-27, 2009, the inspectors met with Entergy representatives

to review the results of recent groundwater samples, as well as those taken and

analyzed in 2008. The review was conducted against criteria contained in 10CFR20,

10CFR50, and applicable industry standards.

The review of the data included a comparison of Entergys data with split samples taken

by the NRC of monitoring wells MW-66 and MW-67, as well as the LaFarge sample

point. In all, 47 samples were analyzed and compared from January 2008 through

January 2009. Isotopic analyses were performed and compared at each of the sample

points for: Tritium, Strontium 90, Nickel 63, and gamma emitters such as Cobalt-60 and

Cesium-137. Results of the NRC samples can be found in ADAMS accession numbers:

ML081420676, ML082690244, ML082690202, ML082690237, ML082730830,

ML082730810, ML090400523, ML090400516, ML090400502, ML090923932,

ML090920949.

Entergy=s evaluation of recent groundwater results are documented in condition reports:

CR-IP2-2009-00883, CR-IP2-2009-01110, CR-IP2-2009-01111, CR-IP2-2009-01113,

and CR-IP2-2009-01114.

b. Findings

No findings of significance were identified.

Enclosure

22

The inspectors concluded that overall, there was agreement between Entergy

personnels results and those independently analyzed by the NRC, and that actions

taken by Entergy have been appropriate. The inspectors also noted that conservative

estimates indicate that the samples represent a very small fraction of the permissible

public dose limits and are negligible with respect to natural background radiation levels.

.2 Inspection Results for TI 2515/172, Reactor Coolant System (RCS) Dissimilar Metal Butt

Welds

a. Inspection Scope

The NRCs Temporary Instruction (TI) 2515/172, provides for confirmation that owners of

pressurized-water reactors (PWRs) have implemented the industry guidelines of the

Materials Reliability Program (MRP)-139 regarding nondestructive examination and

evaluation of certain dissimilar metal (DM) welds in reactor coolant systems containing

Alloy 600/82/182. The TI requires documentation of specific questions in an inspection

report, and those questions and responses applicable to Indian Point are included

below.

In summary, the Indian Point Units 2 and 3 have MRP-139 applicable Alloy 600/82/182

RCS welds in only the hot (HL) and cold leg (CL) pipe-to-reactor pressure vessel (RPV)

nozzle connections. These were examined from the inside diameter (ID) surface

volumetrically by ultrasonic testing (UT), and on the ID surface by eddy current testing

(ECT) at Unit 2 in the 2006 refueling outage, and on Unit 3 from the outside surface

visually during the 2007 refueling outage.

For Unit 3 during 3R15 in Spring 2009, eight alloy 82/182 welds were examined from the

nozzle inner diameter by ECT for the weld surface and UT for the weld volume with

ASME Section XI examination coverage confirmed. The safe end-to-pipe or cast elbow

stainless steel welds were also examined by ECT and UT methods. The inspector

evaluated the UT and ET techniques, which included the data analysis process and

qualifications of both the NDE procedures and the NDE examiners. No significant

indications were found on these welds. One very small indication in the weld cladding of

CL 34 was identified but found to be acceptable for continued service.

a. For MRP-139 baseline inspections of IP Unit 3 in 2009:

Qa1. Have the baseline inspections been performed or are they scheduled to be performed in

accordance with MRP-139 guidance?

A. Yes. The four HL and CL Unit 3 welds were scheduled for UT and ECT examinations

and performed during the Spring 2009 3R15 refueling outage.

Qa2. Is the licensee planning to take any deviations from the MRP-139 baseline inspection

requirements? If so, what deviations are planned and what is the general basis for the

deviation? If inspectors determine that a licensee is planning to deviate from any MRP-

139 baseline inspection requirements, NRR should be informed by email as soon as

possible.

A. No deviations are planned for Unit 3 as the 3R15 ECT and UT examinations complete

the MRP-required examination scope.

Enclosure

23

b. For each examination inspected at IP Unit 3 in 2009 was the activity:

Qb1. Performed in accordance with the examination guidelines in MRP-139 Section 5.1 for

unmitigated welds or mechanical stress improved welds and consistent with NRC staff

relief request authorization for weld overlaid welds?

A. For Unit 3, neither mechanical stress relief nor weld overlays were performed. The four

HL and CL Unit 3 weld UT and ECT examinations were performed from the nozzle

inside diameter at the DM weld location. Also, the outside surfaces of these welds were

visually examined in 2007.

Qb2. Performed by qualified personnel? (Briefly describe the personnel training/qualification

process used by the licensee for this activity.)

A. The UT was performed with a qualified procedure and by qualified individuals. The eddy

current examinations were done in accordance with procedure WDI-STD-146, Rev 9,

with review of the qualifications of the ECT individuals as part of the pre-job

preparations.

Qb3. Performed such that deficiencies were identified, dispositioned, and resolved?

A. One minor indication in the weld internal surface clad material was identified on the 34

CL. This UT-identified condition was reviewed and resolved by the Level III data

reviewer. The condition was not surface interfacing and was not an eddy current

indication.

b. Findings

No findings of significance were identified

.3 Quarterly Resident Inspector Observations of Security Personnel and Activities

a. Inspection Scope

During the inspection period, the inspectors conducted observations of security force

personnel and activities to ensure that these activities were consistent with Entergy

security procedures and applicable regulatory requirements. Although these

observations did not constitute additional inspection samples, the inspections were

considered an integral part of the normal, resident inspector plant status reviews during

implementation of the baseline inspection program.

b. Findings

No findings of significance were identified.

4OA6 Meetings, including Exit

Exit Meeting Summary

On April 15, 2009, the inspectors presented the inspection results to Mr. Joe Pollock and

other Energy staff members, who acknowledged the inspection results. While some

Enclosure

24

proprietary items were reviewed and returned during the inspection, no proprietary

information is presented in this report.

ATTACHMENT: SUPPLEMENTAL INFORMATION

Enclosure

A-1

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Entergy Personnel

J. Pollock, Site Vice President

A. Vitale, General Manager, Plant Operations

K. Davison, Assistant General Manager, Plant Operations

P. Conroy, Director, Nuclear Safety Assurance

D. Gagnon, Manager, Security

R. Walpole, Manager, Licensing

B. Beckman, Manager, Maintenance

J. Dinelli, Assistant Operations Manager, Unit 3

V. Myers, Supervisor, Mechanical Design Engineering

T. Orlando, Engineering Director

R. Burroni, Manager Programs, Components and Engineering

D. Loope, Manager, Radiation Protection

S. Verrochi, Manager System Engineering

F. Inzirillo, Manager, Quality Assurance

N. Azevedo, Supervisor, Code Programs

T. Morzello, Maintenance Supervisor

G. Dahl, Licensing Engineer

H. Anderson, Licensing Engineer

D. Smith, ALARA Specialist

G. Hocking, Supervisor, Radiation Protection Support

R. Blaine, Supervisor, Radiation Protection Operations

S. Sandike, Specialist, Effluent & Environmental Monitoring

P. Donahue, Specialist, Effluent & Environmental Monitoring

R. Mages, ALARA Specialist

N. Papayia, QA

B. Allen, Code Programs

Attachment

A-2

LIST OF ITEMS OPENED, CLOSED AND DISCUSSED

Open and Closed

05000286/2009002-01 NCV Failure to Follow Radiation Protection

Procedures (Section 2OS1)

LIST OF DOCUMENTS REVIEWED

Section 1R01: Adverse Weather Protection

Procedures

OAP-048, Seasonal Weather Preparation, Rev. 4

OAP-008, Severe Weather Preparations, Rev. 5

3-PT-W011, TSC Diesel Generator Support System Inspection, Rev. 18

Section 1R04: Equipment Alignment

Miscellaneous

CR-IP3-2009-00070

Other

Flow Diagram 9321-F-27503, "Safety Injection System, Sheet No. 2," Rev. 48

3-COL-FW-2, Auxiliary Feedwater System, Rev. 29

3-COL-EL-005, Diesel Generators, Rev. 33

3-COL-SI-001, "Safety Injection System," Rev. 38

Section 1R05: Fire Protection

Procedures

EN-DC-161, Control of Combustibles, Rev. 3

IP-SMM-DC-901, IPEC Fire Protection Program, Rev. 6

Pre-Fire Plans 391, 392, 306, 306A, 362, 362A, and 362B

3-COL-FP-2, Fire Protection System Ring Header, Rev. 10

Condition Reports (CR-IP3-2009-)

00499 00504 00511 00556 00584 00600 00604

Section 1R07: Heat Sink Performance

CR-IP3-2009-00535

IP CALC-09-00039

WO 00133315

RT Report # IP3-RT-09-008, Weld PAB-106, Line #408

Section 1R08: In-Service Inspection

ENN-NDE-9.04, Rev 2. Ultrasonic Examination of Ferritic Piping Welds (ASME Sect XI)

PDI-UT-1, Rev 20, Ultrasonic Examination of Ferritic Piping Welds

Attachment

A-3

3-PT-R114, Rev 10. RCS Boric Acid Leakage and Corrosion Inspection

3-PT-R131, Rev 11. RCS Integrity Leak Test

WDI-UT-004, Rev 12. IntraSpect UT Analysis Guidelines for RPV Upper Head CRDM welds

WDI-ET-013, Rev 13. IntraSpect ET Analysis Guidelines for RPV Upper Head CRDM welds

EN-DC-343, Rev 0. Buried Piping and Tanks Inspection and Monitoring Program

WDI-STD-146, Rev 9 ET RV Pipe Welds (ID)

WDI-STD-142, Rev 2 ET RV BMI Welds (ID)

WDI-STD-134, Rev 5 UT RV BMI Welds (ID)

WDI-STD-141, Rev 4 UT RV BMI Welds Analysis(ID)

3-REF-002-GEN, Section 3.7, Rev 2. RFO Procedure CETNA Conoseal Assembly

PDI-ISI-254-SE, Rev 2. Remote ISI Examination of Rx Nozzle to Pipe and Safe End

PDI-UT-10, ENN-NDE-9.10, Rev 2. UT procedure for pressurizer surge nozzle.

Drawings

A226192-18, IP U2 Reactor Coolant Pump Shaft Seal

9321-F-27453, Rev 30. Flow Diagram - Sampling System (valve 953)

9321-F-27383, Rev 27. Flow Diagram RC System (valve 514A)

322097-00, Rev 2, Replacement of Removed Liner Insulation (U2)

9321-F-1280-15, A200 168, Containment Liner Details (U2)

Pressurizer Drawing RCPCPRI, INT-1-2100, Rev 8.

6D30575, Rev 3. BMI NDE Calibration Sample Tube

Condition Reports (IP3-2009)

00898 01335 01242 01103 01097 01097

00779 00898 01016 00739 00805

Work Packages

WO-00154909-01, for NDE

WO-00172099-01, for upper RPV head to CRDMs VT

Other

IP3 Boric Acid Master List, dated 3/26/2009

Report WDI-PJF-1303956-FSR-002. Pressurizer Surge Nozzle Safe End Surface Preparation

and PDI UT Examination, dated March 23, 2009

ASME Section XI

ASME Section XI, Sub-Section IWE

Section 1R12: Maintenance Effectiveness

Procedures

3-PT-Q83, RWST Level Instrument Check and Calibration (LI-921)

3-PT-SA43, RWST Level Instrument Check and Calibration (Loop 920A/B)

3-ES-1.3, Transfer to Cold Leg Recirculation

Condition Reports (CR-IP3-)

2008-01027 2008-01080 2008-01088 2008-01139 2008-01264 2008-01327

2008-01340 2008-01490 2008-01520 2008-01554 2008-01566 2008-01577

2008-01594 2008-01599 2008-01601 2008-01602 2008-01670 2008-01723

2008-01828 2008-01849 2008-01870 2008-01873 2008-01878 2006-01671

2007-01179 2007-03884 2007-04038 2007-04352 2007-04409 2007-03155

2008-00726 2008-01844 2008-02211 2008-02875 2008-03019 2009-00348

Attachment

A-4

IP2-2009-00527 IP2-2009-01041

Maintenance Rule Monitoring Documents

EN-DC-143, System Health Reports, Rev. 8

EN-DC-159, System Monitoring Program, Rev. 3

EN-DC-167, Classification of Structures, Systems, and Components, Rev. 2

EN-DC-203, "Maintenance Rule Program," Rev. 1

EN-DC-204, "Maintenance Scope and Basis," Rev. 1

EN-DC-205, AMaintenance Rule Monitoring," Rev. 2

EN-DC-206, AMaintenance Rule (a)(1) Process,@ Rev. 1

Unit 3 EDG System health report for 4th Qtr 2008, Rev. 0

Unit 3 EDG Health Improvement Plan

SED-AD-22, Condition Monitoring of Maintenance Rule Structures, Rev. 4

Miscellaneous

Maintenance Rule Basis Document Residual Heat Removal System, dated 5/23/05

Work Orders

51667042 51802297 51797562 51694152 51688955 51679627

52024026 00176792

Section 1R13: Maintenance Risk Assessment and Emergent Work Control

Procedures

IP-SMM-WM-101, On-Line Risk Assessment, Rev. 3

Work Week Managers Operators Risk Report, Work Weeks 0905, 0906 and 0908

3R15 Refueling Outage Schedule Risk Assessment Report, Jan. 2009, Amended Feb. 2009

IP-SMM-OU-104, Attachment 9.1, "Shiftly Outage Shutdown Safety Assessment," Rev. 5

Section 1R15: Operability Evaluations

Procedures

EN-OP-104, Operability Determinations, Rev. 3

Indian Point Unit 3 Updated Final Safety Analysis Report, Rev. 2

3PT-Q124, Control Building Exhaust Fan Operational Test, Rev. 3

EN-MA-133, "Control of Scaffolding," Rev. 4

Calculations

IP-CALC-08-00208, Rev. 0

CN-CRA-08-11, Rev. 0

IP-CALC-04-00809, Rev. 2

IP3-ANAL-SI-02802, Rev. 0

IP3-CALC-ED-00207, Rev. 7

Condition Reports (CR-IP3-)

2008-01589

ML091320023

Text

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