ML062650059

2006/09/21-IR 05000219-06-007; Amergen Energy Company; 03/13-17/2006 03/27-31/2006; Oyster Creek Generating Station; Inspection of the Scoping of Non-Safety Systems and the Proposed Aging Management Procedures for Application for Renewed Li

ML062650059
Person / Time
Site:	Oyster Creek
Issue date:	09/21/2006
From:	Diane Jackson Engineering Region 1 Branch 1
To:	Crane C AmerGen Energy Co
Jackson D E
References
%dam200612 IR-06-007
Download: ML062650059 (46)

See also: IR 05000219/2006007

Text

September 21, 2006

Mr. Christopher M. Crane

President and CEO

AmerGen Energy Company, LLC

200 Exelon Way, KSA 3-E

Kennett Square, PA 19348

SUBJECT: OYSTER CREEK GENERATING STATION - NRC LICENSE RENEWAL

INSPECTION REPORT 05000219/2006007

Dear Mr. Crane:

On March 31, 2006, the NRC completed the onsite portion of the inspection of your application

for license renewal of your Oyster Creek Generating Station. The inspection continued in our

Region I office until early September 2006. The enclosed report documents the results of the

inspection, which were discussed on September 13, 2006, with members of your staff in an exit

meeting open for public observation at the Lacey Township Town Hall.

The purpose of this inspection was to examine the plant activities and documents that

supported the application for a renewed license of Oyster Creek Generating Station. The

inspection reviewed the screening and scoping of non-safety related systems, structures, and

components, as required in 10 CFR 54.4(a)(2), and determined whether the proposed aging

management programs are capable of reasonably managing the effects of aging. These NRC

inspection activities constitute one of several inputs into the NRC review process for license

renewal applications.

The inspection team concluded screening and scoping of non-safety related systems,

structures, and components, was implemented as required in 10 CFR 54.4(a)(2), and the aging

management portion of the license renewal activities were conducted as described in the

License Renewal Application. The inspection results support a conclusion that the proposed

activities will reasonably manage the effects of aging in the systems, structures, and

components identified in your application. The inspection concluded the documentation

supporting the application was in an auditable and retrievable form.

In accordance with 10 CFR 2.390 of the NRC's "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 NRCs document system

(ADAMS). ADAMS is accessible from the NRC Website at

http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Donald E. Jackson, Acting Chief

Engineering Branch 1

Division of Reactor Safety

Docket No. 50-219

License No. DPR-16

C. Crane 2

Enclosure: Inspection Report 05000219/2006007

cc w/encl:

Chief Operating Officer, AmerGen

Site Vice President, Oyster Creek Nuclear Generating Station, AmerGen

Plant Manager, Oyster Creek Generating Station, AmerGen

Regulatory Assurance Manager, Oyster Creek, AmerGen

Senior Vice President - Nuclear Services, AmerGen

Vice President - Mid-Atlantic Operations, AmerGen

Vice President - Operations Support, AmerGen

Vice President - Licensing and Regulatory Affairs, AmerGen

Director Licensing, AmerGen

Manager Licensing - Oyster Creek, AmerGen

Vice President, General Counsel and Secretary, AmerGen

T. ONeill, Associate General Counsel, Exelon Generation Company

J. Fewell, Assistant General Counsel, Exelon Nuclear

Correspondence Control Desk, AmerGen

J. Matthews, Esquire, Morgan, Lewis & Bockius LLP

Mayor of Lacey Township

K. Tosch, Chief, Bureau of Nuclear Engineering, NJ Dept of Environmental Protection

R. Shadis, New England Coalition Staff

N. Cohen, Coordinator - Unplug Salem Campaign

W. Costanzo, Technical Advisor - Jersey Shore Nuclear Watch

E. Gbur, Chairwoman - Jersey Shore Nuclear Watch

E. Zobian, Coordinator - Jersey Shore Anti Nuclear Alliance

P. Baldauf, Assistant Director, Radiation Protection and Release Prevention, State of

New Jersey

R. Webster, Rutgers Environmental Law Clinic

C. Crane 3

Distribution w/encl: (VIA E-MAIL)

S. Collins, RA

M. Dapas, DRA

R. Bellamy, DRP

M. Ferdas, DRP, Senior Resident Inspector

R. Treadway, DRP, Resident Inspector

J. DeVries, DRP, Resident OA

B. Sosa, RI OEDO

D. Roberts, NRR

E. Miller PM, NRR

T. Valentine, Backup PM (Interim), NRR

D. Ashley, NRR

ROPreports@nrc.gov

Region I Docket Room (with concurrences)

A. Blough, DRS

M. Gamberoni, DRS

D. Jackson, DRS

M. Modes, DRS

M. Young, OGC

SUNSI Review Complete: DEJ (Reviewers Initials)

DOCUMENT NAME: E:\Filenet\ML062650059.wpd

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/DRS RI/DRP RI/DRS

NAME MModes * (MM) RBellamy * (RB) DJackson (DJ)

DATE 09/19/06 09/21/06 09/19/06

OFFICE

NAME

DATE

OFFICIAL RECORD COPY

U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket No: 50-219

License No: DPR-16

Report No: 05000219/20006007

Licensee: AmerGen Energy Company, LLC

Facility: Oyster Creek Generating Station

Location: Forked River, New Jersey

Dates: March 13 - 17, 2006 and March 27 - 31, 2006

Inspectors: M. Modes, Team Leader, Division of Reactor Safety (DRS)

P. Kaufman, Sr. Reactor Inspector, DRS

G. Meyer, Sr. Reactor Inspector, DRS

S. Chaudhary, Health Physicist, Division of Nuclear Materials Safety

(DNMS)

T. OHara, Reactor Inspector, DRS

J. Lilliendahl, Reactor Inspector, DRS

D. Johnson, Reactor Inspector, DRS

D. Werkheiser, Resident Inspector, Division of Reactor Projects (DRP)

Approved By: Donald E. Jackson, Acting Chief

Engineering Branch 1

Division of Reactor Safety

Enclosure

SUMMARY OF FINDINGS

IR 05000219/2006007; 03/13/2006 - 03/17/2006, 03/27/2006 - 03/31/2006, Oyster Creek

Generating Station; Inspection of the Scoping of Non-Safety Systems and the Proposed Aging

Management Procedures for the Oyster Creek Generating Station Application for Renewed

License.

This inspection of license renewal activities was performed by eight regional office engineering

inspectors. The inspection was conducted in accordance with NRC Manual Chapter 2516 and

NRC Inspection Procedure 71002. This inspection did not identify any findings as defined in

NRC Manual Chapter 0612. The inspection team concluded screening and scoping of non-

safety related systems, structures, and components, were implemented as required in 10 CFR 54.4(a)(2), and the aging management portions of the license renewal activities were conducted

as described in the License Renewal Application. The inspection results support a conclusion

that the proposed activities will reasonably manage the effects of aging in the systems,

structures, and components identified in your application. The inspection concluded the

documentation supporting the application was in an auditable and retrievable form.

ii Enclosure

TABLE OF CONTENTS

Page

SUMMARY OF FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii

4. OTHER ACTIVITIES (OA) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

4OA2 Other - License Renewal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

a. Inspection Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

a.1. Scoping of Non Safety-Related Systems, Structures, and

Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

a.2. Programs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

One-Time Inspection Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

Bolting Integrity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

Buried Piping Inspection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

Flow-Accelerated Corrosion Program . . . . . . . . . . . . . . . . . . . . . . . . . . 4

Water Chemistry Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

Closed-Cycle Cooling Water Systems Program . . . . . . . . . . . . . . . . . . . 5

10 CFR Part 50, Appendix J Program . . . . . . . . . . . . . . . . . . . . . . . . . . 5

Fuel Oil Chemistry Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6

Boiling Water Reactor Feedwater Nozzle Program . . . . . . . . . . . . . . . . 6

Boiling Water Reactor Stress Corrosion Cracking Program . . . . . . . . . . 7

Periodic Inspection Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8

Wooden Utility Pole Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

Periodic Testing of Containment Spray Nozzles . . . . . . . . . . . . . . . . . 10

Inaccessible Medium-Voltage Cables Not Subject to 10 CFR 50.49

Environmental Qualification Requirements . . . . . . . . . . . . . . . . 10

Electrical Cables and Connections Not Subject to 10 CFR 50.49

Environmental Qualification Requirements . . . . . . . . . . . . . . . . 11

Electrical Cables and Connections Not Subject to 10 CFR 50.49

Environmental Qualification Requirements Used in Instrument

Circuits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

Fire Protection Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

Fire Water System Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13

Periodic Inspection of Ventilation Systems Program . . . . . . . . . . . . . . 14

Periodic Inspection - Forked River Combustion Turbine . . . . . . . . . . . 14

ASME,Section XI, Subsection IWE Program . . . . . . . . . . . . . . . . . . . . 16

Protective Coating Monitoring and Maintenance Program . . . . . . . . . . 16

Above-Ground Outdoor Tank Monitoring Program . . . . . . . . . . . . . . . . 17

ASME Section XI, Subsection IWF . . . . . . . . . . . . . . . . . . . . . . . . . . . 17

Masonry Wall Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18

Structures Monitoring Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19

Inspection of Water Control Structures . . . . . . . . . . . . . . . . . . . . . . . . 20

Metal Fatigue of Reactor Coolant Pressure Boundary . . . . . . . . . . . . . 21

Isolation Condenser System Review . . . . . . . . . . . . . . . . . . . . . . . . . . 21

b. Observation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23

c. Overall Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24

40A6 Meetings, Including Exit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24

iii Enclosure

TABLE OF CONTENTS (Contd)

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

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

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

LIST OF ACRONYMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-15

iv Enclosure

Report Details

4. OTHER ACTIVITIES (OA)

4OA2 Other - License Renewal

a. Inspection Scope

This inspection was conducted by NRC Region I and headquarters based inspectors in

order to evaluate the thoroughness and accuracy of the screening and scoping of non-

safety related systems, structures, and components, as required in 10 CFR 54.4(a)(2)

and to evaluate whether aging management programs will be capable of managing the

identified aging effect in a appropriate manner.

The inspection team selected a number of systems for review, using the NRC accepted

guidance, in order to determine if the methodology applied by the applicant appropriately

captured the non-safety systems affecting the safety functions of a system, component,

or structure within the scope of license renewal.

The inspection team selected a sample of aging management programs to verify the

adequacy of the applicants documentation and implementation activities. The selected

aging management programs were reviewed to determine whether the proposed aging

management implementing process would adequately manage the effects of aging on

the system.

The inspectors reviewed supporting documentation and interviewed applicant personnel

to confirm the accuracy of the license renewal application conclusions. For a sample of

plant systems and structures, inspectors performed visual examinations of accessible

portions of the systems to observe aging effects.

a.1. Scoping of Non Safety-Related Systems, Structures, and Components

To assess the thoroughness and accuracy of the methods used to bring systems,

structures, and components within scope of the application and to screen non-safety

related systems, structures, and components, as required in 10 CFR 54.4(a)(2), the

inspectors reviewed the applicants program guidance procedures and summaries of

results for Oyster Creek. The inspectors determined the applicants procedures to be

consistent with the NRC accepted guidance in Sections 3, 4, and 5 of Appendix F to

NEI 95-10, Revision 5 (3: non-safety related systems, structures, and components

within scope of the current licensing basis, 4: non-safety related systems, structures,

and components directly connected to safety-related systems, structures, and

components, and 5: non-safety related systems, structures, and components not directly

connected to safety-related systems, structures, and components). Also, the inspectors

determined that the applicant appropriately utilized the guidance in their process for

determining which systems were within scope.

Enclosure

2

The applicant based the scoping and screening results on a technical review and

walkdown of all applicable plant areas by qualified plant personnel. The inspectors

reviewed the set of license renewal drawings, which were color-coded based on the

results. The inspectors interviewed personnel and independently inspected numerous

areas within the plant to confirm that appropriate systems, structures, and components

had been included within the license renewal scope, that systems, structures, and

components excluded from the license renewal scope had an acceptable basis, and that

the boundary for determining scope within the systems, including anchors, was

appropriate. For systems, structures, and components selected from the results, the

inspectors confirmed that the in-plant configuration was accurate and acceptably

categorized, and for systems, structures, and components selected within the plant, the

inspectors confirmed that the categorization result in program documents was

appropriate. The in-plant areas and systems reviewed included the following:

• Reactor Building;

• Turbine Building;

• Intake Structure;

• Ventilation Stack;

• Diesel Generator Building;

• Diesel Fuel Oil Building;

• Fire Protection System;

• Isolation Condenser System;

• Hardened Vent System;

• Nitrogen Supply System;

• Instrument Air System; and

• Service Water System.

The inspectors determined the personnel involved in the process were knowledgeable

and appropriately trained, and that the applicant had implemented an acceptable

method of scoping and screening of non-safety related systems, structures, and

components.

a.2. Programs

One-Time Inspection Program

The One-Time Inspection Program is a new aging management program intended to

verify the effectiveness of other aging management programs, including Water

Chemistry, Closed Cycle Cooling Water Systems, and Fuel Oil Chemistry Programs, by

reviewing various aging effects for impact. Where corrosion resistant materials and/or

non-corrosive environments exist, the One-Time Inspection Program is intended to

verify that an aging management program is not needed during extended operations by

confirming that aging effects are not occurring or are occurring in a manner that does

not affect the safety function of systems, structures, and components within the scope

of the application. Non-destructive evaluation will be performed by qualified personnel

using procedures and processes consistent with the American Society of Mechanical

Engineers Boiler and Pressure Vessel Code (ASME) and 10 CFR 50, Appendix B. The

Enclosure

3

One-Time Inspection Program will be implemented prior to the period of extended

operation.

The inspectors reviewed the program description, implementation plan, and inspection

sample basis, and discussed the planned activities with the responsible staff.

For the One-Time Inspection Program, the inspectors concluded the applicant

performed adequate evaluations and reviews of industry experience and plant history to

determine an acceptable approach to identifying, assessing and managing any aging

effects detected. The applicant developed adequate guidance for implementation of

the One-Time Inspection Program.

Bolting Integrity

The Bolting Integrity Program is an existing program credited with managing the loss of

material, cracking, and loss of prestress aging effects in safety-related bolting at Oyster

Creek. The aging effects are managed by visual inspection for leakage during system

pressure tests, normal plant operation, and periodic system maintenance, and repaired

in accordance with maintenance procedures and the ASME Code.

The inspectors reviewed the program basis document, implementing procedures,

documented reviews, and a bolting-related apparent cause evaluation, and interviewed

the responsible plant personnel regarding these documents. In addition, the inspectors

walked down portions of the Standby Liquid Control, Isolation Condenser, Control Rod

Drive, and Reactor Building Closed Cooling Water Systems to confirm that the program

had maintained acceptable bolting conditions.

For the Bolting Integrity Program, the inspectors concluded that the applicant had

performed adequate evaluations as well as industry experience and historical reviews to

determine the aging effects are managed by the Bolting Integrity Program. The

applicant provided adequate guidance to ensure the aging effects are appropriately

managed.

Buried Piping Inspection

The Buried Piping Inspection Program is an existing program credited with managing

the loss of material aging effects on the external surfaces of piping in a soil

environment, including the service water, emergency service water, and condensate

transfer systems. The aging effects are managed by preventive measures, i.e.,

coatings, wrapping, and condition monitoring measures, including visual inspections and

periodic system pressure testing. As described in Appendix B, Part 1.26 of the

application, the applicant plans to enhance the program by augmenting the visual

inspections prior to extended operations and performing periodic visual inspections, and

to include additional piping, such as the fire protection system.

Enclosure

4

The inspectors reviewed the program basis document, system drawings, implementing

procedures, and documented reviews, and interviewed the responsible plant personnel

regarding these documents. Also, the inspectors walked down the service water and

emergency service water systems in the vicinity of buried piping.

For the Buried Piping Inspection Program, the inspectors concluded that the applicant

had performed adequate evaluations as well as industry experience and historical

reviews to determine the aging effects managed by the Buried Piping Inspection

Program. The applicant provided adequate guidance to ensure the aging effects are

appropriately managed.

Flow-Accelerated Corrosion Program

The Flow-Accelerated Corrosion Program is an existing program credited with managing

the corrosion aging effects in all carbon steel piping and components containing high-

energy fluids at Oyster Creek Generating Station. The aging effects are managed by

using ultrasonic and radiographic testing to detect wall thinning and by predicting wear

rates to support the proactive replacement of system piping. In addition, the program

provides for the performance of follow-up inspections to confirm predictions and to

determine the need for repairs or replacements as necessary.

The inspectors reviewed the piping ultrasonic testing wall thickness results from

previous inspections and reviewed the CHECWORKS computer analysis of the future

wall thickness forecasts. The inspector also reviewed recent changes to the

CHECWORKS model to ensure previously identified deficiencies have been corrected.

The inspectors noted that recent replacements, initiated as a result of this program,

were implemented preventively due to identified flow-accelerated corrosion. The

replacement piping material was more resistant to corrosion than the original piping

material.

For the Flow-Accelerated Corrosion Program, the inspectors concluded the applicant

conducted adequate evaluations as well as industry experience and historical reviews

and, as a consequence, the effects of aging will be reasonably managed by the

proposed program.

Water Chemistry Program

The Water Chemistry Program is an existing program credited with managing the

effects of aging on piping, piping components, piping elements, and systems, such as

the condensate and feedwater, and condensate storage tank in Oyster Creek

Generating Station. The aging effects are managed by monitoring and control of

reactor water chemistry to minimize contaminant concentration and mitigate loss of

material due to general, crevice and pitting corrosion and cracking caused by stress

corrosion cracking.

Enclosure

5

Water chemistry control is administered in accordance with the Boiling Water Reactor

Vessel and Internals Project guideline BWRVIP-29 and Electric Power Research

Institute guideline EPRI TR-103515. The inspectors reviewed the chemistry procedures

and sampling results to confirm that the guidance contained in BWRVIP-29 and EPRI

TR-103515 was being implemented.

For the Water Chemistry Program, the inspectors concluded the applicant had

conducted adequate evaluations as well as industry experience and historical reviews to

determine aging effects managed by an aging management program. The applicant

provided adequate guidance to ensure aging effects are appropriately managed.

Closed-Cycle Cooling Water Systems Program

The Closed-Cycle Cooling Water Systems Program is an existing program credited with

managing loss of material, cracking, and buildup-of-deposit aging effects in components

exposed to closed-cycle cooling water environments at the Oyster Creek Generating

Station. Systems within the scope of the closed-cycle cooling water program include the

turbine building closed cooling, reactor building closed cooling, and emergency diesel

generator closed cooling water systems. The aging effects are managed by monitoring

and control of cooling water chemistry, performing surveillance tests, and through

periodic inspection of system components in a manner consistent with EPRI TR-107396

guidelines.

The inspectors observed cleaning of the turbine closed-cooling water heat exchanger

and performed a walkdown of the system with plant personnel. In addition, the

inspectors reviewed closed-cycle cooling water chemistry procedures and reviewed past

chemistry sample results to confirm that the requirements of EPRI TR-107396 are being

met.

For the Closed-Cycle Cooling Water Systems Program, the inspectors concluded the

applicant had conducted adequate evaluations as well as industry experience and

historical reviews to determine aging effects managed by an aging management

program. The applicant provided adequate guidance to ensure aging effects are

appropriately managed.

10 CFR Part 50, Appendix J Program

The 10 CFR Part 50, Appendix J Program is an existing program credited with

managing the aging degradation of pressure retaining boundaries of piping and

components of the various systems penetrating the containment at the Oyster Creek

Generating Station. In addition, the program also detects age related degradation in

material properties of gaskets, o-rings, and packing materials for the primary

containment pressure boundary access points. The aging effects are managed by

performing containment leak rate tests to assure that leakage through primary

containment and systems and components penetrating primary containment does not

exceed allowable leakage limits specified in the Technical Specifications.

Enclosure

6

The inspectors reviewed Oyster Creeks procedures for leak rate testing. In addition,

the inspectors reviewed corrective actions for components that did not meet leak rate

test acceptance criteria. The inspectors noted that corrective actions taken to repair

these components were acceptable.

For the 10 CFR Part 50, Appendix J Program, the inspectors concluded the applicant

had conducted adequate evaluations as well as industry experience and historical

reviews to determine aging effects managed by an aging management program. The

applicant provided adequate guidance to ensure aging effects are appropriately

managed.

Fuel Oil Chemistry Program

The Fuel Oil Chemistry Program is an existing program that will be modified for the

purpose of managing the affects of pitting and corrosion in the diesel fuel oil tank at the

Oyster Creek Generating Station. The aging effects are managed by the addition of

biocides and corrosion inhibitors to minimize biological activity and mitigate corrosion,

periodic cleaning, and applying coating to the internal surfaces of the tank.

The inspectors reviewed the schedule for implementation of the enhancements. The

inspectors reviewed recent sample results and tank thickness measurements to verify

that results were within the acceptable range.

For the Fuel Oil Chemistry Program, the inspectors concluded the applicant had

conducted adequate evaluations as well as industry experience and historical reviews to

determine aging effects managed by an aging management program. The applicant

provided adequate guidance to ensure aging effects are appropriately managed.

Boiling Water Reactor Feedwater Nozzle Program

The Boiling Water Reactor Feedwater Nozzle Program is an existing program that will

be modified to implement the recommendations of the Boiling Water Owners Group

Licensing Topical Report: General Electric NE-523-A71-0594. These enhancements will

be implemented prior to entering the period of extended operation per Oyster Creek

Assignment Report AR# 00330592, A.1.05 Commitment (BWR Feedwater Nozzle). The

program is credited with managing the aging effects of cracking in the feedwater

nozzles. The program is administered by the station in-service inspection plan ER-OC-

330-1001, ISI Program Plan Fourth Ten-Year Inspection Interval, and implemented by

station procedure ER-AA-330-002, In-service Inspection of Section XI Welds and

Components. The station in-service inspection program incorporates the requirements

of the ASME Code. The aging effects are managed by periodic ultrasonic testing

inspections of critical regions of the feedwater nozzles. The ultrasonic test inspections

are performed at intervals not exceeding ten years and was embraced in an NRC safety

evaluation.

Enclosure

7

Inspections performed in 1977 identified cracks in the Oyster Creek nozzles. These

cracks were repaired. The inspectors reviewed plant modification #166-76-4,

Feedwater Nozzle Cladding Removal and Sparger Replacement and reviewed

selected ultrasonic testing examination reports of the feedwater nozzles. To minimize

thermal cycling and fatigue induced cracking, the thermal sleeves were modified to a

piston design. Subsequent inspections found no indications in the feedwater nozzles.

The inspectors reviewed Focused Area Self-Assessment Report Oyster Creek Inservice

Inspection Program, completed in June 2004. The inspectors determined the feedwater

nozzle program at Oyster Creek effectively monitored the feedwater nozzles for

cracking.

For the Boiling Water Reactor Feedwater Nozzle Program, the inspectors concluded the

applicant had conducted adequate evaluations as well as industry experience and

historical reviews to determine aging effects managed by an aging management

program. The applicant provided adequate guidance to ensure aging effects are

appropriately managed.

Boiling Water Reactor Stress Corrosion Cracking Program

The Boiling Water Reactor Stress Corrosion Cracking Program is an existing aging

management program credited with managing crack initiation and growth due to

intergranular stress corrosion cracking in stainless steel and nickel alloy reactor coolant

pressure boundary piping, welds, components and piping four inches and larger nominal

pipe size exposed to reactor coolant above 200EF. The aging effects are managed by

preventive measures which include monitoring and controlling water impurities by

improved water chemistry control activities and by providing replacement stainless steel

components in a solution annealed condition with a maximum carbon content of 0.035%

wt. and a minimum ferrite level of 7.5%. Inspection and flaw evaluations are conducted

in accordance with Oyster Creek in-service inspection program plan ER-OC-330-1001,

ISI Program Plan Fourth Ten-Year Inspection Interval and Oyster Creeks augmented

inspection program for IGSCC ER-OC-330-1002, IGSCC Inspection Plan Fourth Ten-

Year Inspection Interval, which incorporates the technical basis and guidance described

in NUREG-0313, NRC Generic Letter 88-01, and staff-reviewed Boiling Water Reactor

Vessel Internal Inspection Program BWRVIP-75.

The inspectors noted that where pre-emptive piping replacement was accomplished the

replacement piping material used was more resistant to intergranular stress corrosion

cracking than the original piping material. The applicant replaced the following system

piping material with intergranular stress corrosion cracking resistant material:

1) all isolation condenser large bore piping outside the drywell from the drywell

penetrations to the isolation condensers during refueling outage 1R13 in 1991;

2) all piping within the four isolation condenser drywell penetrations and the two

reactor water cleanup system drywell penetrations which contained welds that

were not inspectible;

3) the head cooling spray nozzle assembly, the 4 inch tee and flange of the

reactor vent line.

Enclosure

8

To further mitigate the initiation and propagation of intergranular stress corrosion

cracking the applicant implemented hydrogen water chemistry during cycle 12 in 1990

and noble metals chemical additions during 1R19 refueling outage in 2002.

Additionally, all accessible welds susceptible to intergranular stress corrosion cracking

in reactor coolant boundary piping systems inside the drywell (except the reactor water

cleanup system) were stress improved.

The Boiling Water Reactor stress corrosion cracking aging management program uses

ultrasonic testing to detect intergranular stress corrosion cracking flaws in the reactor

coolant boundary piping prior to loss of intended functions of the components. Of the

380 welds included in the scope of Generic Letter 88-01, Oyster Creek identified, during

the period the program was implemented, there were 11 welds with indications of

intergranular stress corrosion cracking. Nine welds have been repaired with full

structural overlays (four in the core spray system, four in the reactor recirculation

system, and one in the shutdown cooling system). Two reactor recirculation system

welds, which were both stress improved before initial inspections had indications of

intergranular stress corrosion cracking, remained in service without repair. Both of

these welds in the reactor recirculation system have been re-examined in 2002 and

2004 using the Improved Performance Demonstration Initiative ultrasonic test

examination technique and the welds did not exhibit any indication of intergranular

stress corrosion cracking. No new indications of intergranular stress corrosion cracking

have been detected by inspections during the past six refueling outages. As a result of

the implemented preventive measures to mitigate intergranular stress corrosion cracking

Oyster Creek has no indications of intergranular stress corrosion cracking at this time.

Therefore the inspectors determined that the Boiling Water Reactor Stress Corrosion

Program at Oyster Creek has been effective in monitoring and mitigating intergranular

stress corrosion cracking in the reactor coolant boundary piping systems.

For the Boiling Water Reactor Stress Corrosion Cracking Program, the inspectors

concluded the applicant had conducted adequate evaluations as well as industry

experience and historical reviews to determine aging effects managed by an aging

management program. The applicant provided adequate guidance to ensure aging

effects are appropriately managed.

Periodic Inspection Program

The Periodic Inspection Program is a new program under development at Oyster Creek

that consists of periodic inspections of selected systems in the scope of license renewal

that require periodic monitoring of aging effects, and are not covered by other existing

periodic monitoring programs to verify the integrity of the systems and confirm the

absence of identified aging effects. The Periodic Inspection Program manages the

aging effect of change in material properties, loss of material and reduction of heat

transfer for systems, components, and environments. The aging effects are managed

by periodic condition monitoring examinations performed at susceptible locations in the

systems, intended to assure that existing environmental conditions are not causing

material degradation that could result in a loss of system intended functions. The initial

periodic inspections of this new aging management program will be implemented near

Enclosure

9

the end of the current operating term but prior to the period of extended operation.

Subsequent periodic inspections will be performed on a frequency not to exceed once

every ten years.

The Periodic Inspection Program provides inspection criteria, requires evaluation of the

inspection results, and provides recommendations for additional inspections, as

necessary. Inspections will be performed in accordance with station procedures that are

based on applicable codes and standards. Inspection methods may include visual

examinations VT-1 or VT-3 of disassembled components or volumetric non-destructive

examination techniques. Some of the implementing procedures for the Periodic

Inspection Program were reviewed by the inspectors, including existing nondestructive

examination procedures ER-OC-330-1001, ISI Program Plan Fourth Ten-Year

Inspection Interval, ER-AA-35-014, VT-1 Visual Examinations, ER-AA-335-016, VT-3

Visual Examination of Component Supports and Attachments, and ER-AA-335-032,

Ultrasonic Through Wall Sizing in Pipe Welds. A periodic inspection table, which was

in draft at the time of this inspection, is a listing of selected systems and components to

be periodically inspected to verify the integrity of the system and confirm the absence of

identified aging effects was also reviewed. Based on review of the implementing

documents and procedures, the inspectors determined that the Periodic Inspection

Program, when implemented at Oyster Creek, will provide assurance that systems and

components are routinely inspected for age related degradation of change in material

properties, loss of material and reduction of heat transfer for systems, components, and

environments, and will adequately manage the identified aging effects.

For the Periodic Inspection Program, the inspectors concluded the applicant had

conducted adequate evaluations as well as industry experience and historical reviews to

determine aging effects managed by an aging management program. The applicant

provided adequate guidance to ensure aging effects are appropriately managed.

Wooden Utility Pole Program

The Wooden Utility Pole Program is a new program credited with managing the aging

effects of loss of material and change in material properties in all wooden utility poles

which support an intended function for the offsite power systems at the Oyster Creek

Generating Station. The aging effects are managed by inspection of wooden poles

every ten years by a qualified inspector.

The team reviewed program bases documents and industry guidance. The inspectors

also conducted interviews and performed walkdowns with plant personnel. During the

walkdown, one pole (JC 514A L) was noted to be degraded. The applicant was able to

show that the condition had been previously analyzed and that plans are in place to

adequately reinforce the pole.

For the Wooden Utility Pole Program, the inspectors concluded the applicant had

conducted adequate evaluations as well as industry experience and historical reviews to

determine aging effects managed by the Wooden Utility Pole Program. The applicant

provided adequate draft guidance to ensure aging effects will be appropriately

managed.

Enclosure

10

Periodic Testing of Containment Spray Nozzles

The Containment Spray Nozzle Program is an existing program credited with

demonstrating that the drywell and torus spray nozzles are not blocked by debris or

corrosion products. Carbon steel piping upstream of the drywell and torus spray nozzles

is subject to possible general corrosion that could result in plugging nozzles with rust.

Periodic air tests verify that the drywell and torus spray nozzles are free from plugging

and are therefore available to provide the steam quenching functions of the nozzles.

The team conducted interviews and reviewed program bases documents and previous

test results. The team noted that the existing program has been effective at identifying

and correcting degraded conditions.

For the Periodic Testing of Containment Spray Nozzles Program, the inspectors

concluded the applicant had conducted adequate evaluations as well as industry

experience and historical reviews to determine aging effects managed by the

Containment Spray Nozzle Program. The applicant provided adequate guidance to

ensure aging effects are appropriately managed.

Inaccessible Medium-Voltage Cables Not Subject to 10 CFR 50.49 Environmental

Qualification Requirements

The Inaccessible Medium-Voltage Cables Not Subject to 10 CFR 50.49 Environmental

Qualification Requirements Program is a new program developed for the purpose of

aging management credited with managing the moisture related aging effects in

medium-voltage cable systems at the Oyster Creek Generating Station. The aging

effects are managed by cable testing and periodic inspection of manholes.

The team reviewed program bases documents and industry guidance. The inspectors

also conducted interviews and performed walkdowns with plant personnel. The

manhole inspection frequency was initially established at the NUREG-1801,

Revision 1,"Generic Aging Lessons Learned (GALL) Report, recommended two-year

frequency based on Oyster Creeks operating experience that does not indicate a trend

or recurrence of cable submergence in manholes. However, NRC inspectors identified,

approximately 2 inches of water in the manhole selected by the NRC team for

inspection. Consequently, the applicant entered this issue into the corrective action

system (CA #IR 469998, #IR 471363) and documented the need to re-evaluate the

adequacy of the manhole inspection frequency.

Due to several medium-voltage cable failures in Oyster Creeks operating experience, a

medium-voltage cable testing program is currently in place. Because of the limited

success of the previous DC step voltage testing method, Oyster Creek has begun

implementing a new method of cable testing provided by DTE Energy for most of the

medium-voltage cables. NUREG 1801 (XI.E3) specifies the test method should be

state-of-the-art at the time the test is performed. Although the new DTE Energy testing

method is not yet recognized as an industry standard, it is a form of partial discharge

testing (partial discharge testing is one of the recognized standards specifically listed in

Enclosure

11

the NUREG-1801), and the applicant expects formal acceptance of the new testing

method as an industry standard prior to extended operation.

The applicant has agreed to maintain the current testing frequency limit of six years in

LRCR 289 for the first six years, after which the frequency may be re-evaluated and

extended up to ten years. This change will provide sufficient time for successful

operating experience prior to expanding to the NUREG 1801 recommended ten year

frequency.

For the Inaccessible Medium-Voltage Cables Not Subject to 10 CFR 50.49

Environmental Qualification Requirements Program, the inspectors concluded the

applicant had conducted adequate evaluations as well as industry experience and

historical reviews to determine aging effects managed by the Inaccessible Medium-

Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements

Program. The applicant provided adequate guidance to ensure aging effects are

appropriately managed.

Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental

Qualification Requirements

The Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental

Qualification Requirements Program is a new program credited with managing the heat,

radiation, and moisture aging effects in non-environmentally qualified electrical cables

and connections in Oyster Creek Generating Station. Connections include splices,

terminal blocks, connectors, and fuse blocks. The aging effects are managed by

periodic inspections.

The team reviewed program bases documents, a draft implementing procedure and

industry guidance. The inspectors also conducted interviews and performed walkdowns

with plant personnel. Inspections will be done of all accessible cables and connections

in adverse localized environments. This aging management program focuses on a

representative sample of accessible cables and connections with sampling structured to

include key areas of concern. Plant locations containing cables within scope that do not

include adverse general or localized conditions may be excluded from inspections based

on engineering evaluations.

Because there were several examples of polyvinyl chloride cable insulation bleeding in

Oyster Creeks operating experience, the applicant agreed to specifically include

polyvinyl chloride cable insulation bleeding as an aging effect to be addressed in this

program. Although polyvinyl chloride cable insulation bleeding has not led to any

equipment degradation at Oyster Creek, there have been instances cited in NRCs

Information Notices 91-20 and 94-78 where polyvinyl chloride insulation bleeding under

unfavorable configurations caused hardened plasticizer to degrade equipment. As a

consequence of the NRCs review, the applicant entered this issue into their corrective

action system (AR 00472707) in order to evaluate the current extent-of-condition of

polyvinyl chloride cable insulation bleeding and determine if their original screening of

this aging affect should be revised.

Enclosure

12

For the Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental

Qualification Requirements Program, the inspectors concluded the applicant had

conducted adequate evaluations as well as industry experience and historical reviews to

determine aging effects managed by the Electrical Cables and Connections Not Subject

to 10 CFR 50.49 Environmental Qualification Requirements Program. The applicant

provided adequate draft guidance to ensure aging effects are appropriately managed.

Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental

Qualification Requirements Used in Instrument Circuits

The Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental

Qualification Requirements Used in Instrument Circuits Program is an existing program

modified for the purpose of aging management that manages aging of the cables of the

Intermediate Range Monitoring, Local Power Range Monitoring/Average Power Range

Monitoring, Reactor Building High Radiation Monitoring, and Air Ejector Off-Gas

Radiation Monitoring systems that are sensitive instrumentation circuits with low-level

signals and are located in areas where the cables and connections could be exposed to

adverse localized environments caused by heat, radiation, or moisture. The aging

effects are managed by calibration, current/voltage, and time domain reflectometry

testing. The current program will be enhanced to include a review of the calibration and

cable testing results for cable aging degradation.

The team reviewed program bases documents, draft implementing procedure and

industry guidance. The inspectors also conducted interviews and performed walkdowns

with plant personnel.

For the Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental

Qualification Requirements Used in Instrument Circuits Program, the inspectors

concluded the applicant had conducted adequate evaluations as well as industry

experience and historical reviews to determine aging effects managed by the Electrical

Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification

Requirements Used in Instrument Circuits Program. The applicant provided adequate

draft guidance to ensure aging effects are appropriately managed.

Fire Protection Program

The Fire Protection Program is an existing program modified for the purpose of aging

management credited with managing the fire barrier function aging effects in fire

protection systems and a diesel-driven fire pump inspection program. The aging effects

are managed by periodic inspection of fire barrier penetration seals, fire barrier walls,

ceilings, floors, and all fire rated doors. The program is credited with managing loss of

material aging effects in fuel oil lines of the diesel driven fire pump through periodic

testing of the pump. This aging management program will also manage the aging

effects of in-scope carbon dioxide and halon suppression systems, once enhancements

are made to periodically inspect these systems.

Enclosure

13

The inspectors reviewed the Fire Protection Program as well as supporting documents

to verify the effectiveness of the Fire Protection Program. The inspectors also

conducted interviews and performed walkdowns of various fire protection systems with

plant personnel to observe the effectiveness of the existing Fire Protection Program.

Enhancements to the existing program include guidance to identify fire barrier

degradation, surface integrity and clearance on fire doors inspected every two years, fire

pump diesel fuel supply system external surface corrosion examinations, and external

corrosion and damage inspections for halon and low-pressure carbon dioxide fire

suppression systems. The inspectors noted an acceptable exception in the application

of the NUREG-1801 guidance for 6-month periodicity on visual inspection and functional

testing of halon and carbon dioxide fire suppressions. Oyster Creek Generating Station

performs in-depth operational tests and inspections on an 18-month periodicity. The

applicant does perform a weekly tank/charge check and a monthly valve position

alignment check and will include visual inspections of external surfaces as an

enhancement prior to the period of extended operation.

For the Fire Protection Program, the inspectors concluded the applicant had conducted

adequate evaluations as well as industry experience and historical reviews to determine

aging effects managed by the fire protection program. The applicant has provided

adequate guidance to ensure aging effects are appropriately managed.

Fire Water System Program

The Fire Water System Program is an existing program modified for the purpose of

aging management credited with managing the loss of material, microbiological

influenced corrosion, and biofouling aging effects in fire water systems at Oyster Creek

Generating Station. The aging effects are managed by periodic maintenance, testing,

and inspection of system piping and components in accordance with codes and

standards. The inspectors reviewed program bases documents, completed testing and

maintenance procedures, corrective action reports, design documents, and industry

guidance. The inspectors also conducted interviews and performed walkdowns of the

fire water system with plant personnel. The fire water system is maintained in a

pressurized state which provides the applicant with constant system integrity status.

The piping internals are routinely inspected at various locations throughout the system

for loss of material and biofouling. The following enhancements have been noted:

Sprinkler head inspections in accordance with NFPA 25 Standard for the

Inspection, Testing, and Maintenance of Water-Based Fire Protection Systems

(1998 Edition).

Samples will be submitted to a testing laboratory prior to being in service 50

years.

Inclusion of inspection of selected portions of the fire protection system piping

located aboveground and exposed to water, by non-intrusive volumetric

examinations.

Performance of water sampling for the presence of microbiological corrosion

every 5 years.

Enclosure

14

Inclusion of visual inspection of the water storage tank heater pressure boundary

components during the periodic tank internal inspection.

For the Fire Water System Program, the inspectors concluded the applicant had

conducted adequate evaluations, as well as industry experience and historical reviews

to determine aging effects managed by the fire water system program. The applicant

provided an acceptable plan to implement adequate guidance and terms to ensure

aging effects are appropriately managed.

Periodic Inspection of Ventilation Systems Program

The Periodic Inspection of Ventilation Systems Program is an existing program at

Oyster Creek Generating Station modified for the purpose of aging management. The

program is credited with managing loss of material, changes in material properties, and

degradation of heat transfer in ventilation systems in the scope of license renewal

(flexible connections, fan and filter housings, and access door seals). Instrument piping

and valves, restricting orifices and flow elements, thermowells, and Standby Gas

Treatment System ducting exposed to soil or sand will be added to the scope as

enhancements to the program. The aging effects are managed by periodic inspections

that will be condition monitoring examinations performed at susceptible locations in the

systems, intended to assure that existing environmental conditions are not causing

material degradation that could result in a loss of system intended functions.

The inspectors reviewed program bases documents, completed testing and

maintenance procedures, corrective action reports, design documents, and industry

guidance. The inspectors also conducted interviews and performed walkdowns of

accessible portions of Standby Gas Treatment and Reactor Building Ventilation Systems

with plant personnel.

Complete visual inspections and performance tests of all ventilation systems in scope

are performed during system preventive maintenance activities on a frequency not to

exceed five years. This includes system leakage and filter efficiency tests for Standby

Gas Treatment, Reactor Building and Control Room Ventilation systems. An additional

noted enhancement includes adding specific guidance to inspect for loss of material and

material property changes.

For the Periodic Inspection of Ventilation Systems Program, the inspectors concluded

the applicant had conducted adequate evaluations as well as industry experience and

historical reviews to determine aging effects managed by an aging management

program. The applicant provided adequate guidance to ensure aging effects are

appropriately managed.

Periodic Inspection - Forked River Combustion Turbine

The Periodic Inspection - Forked River Combustion Turbine program is a new program

credited with addressing the two Forked River Combustion Turbine power plant

components in the scope of license renewal that require periodic monitoring of aging

Enclosure

15

effects, and are not covered by other aging management programs. In the applicants

response to the NRCs requests for additional information, RAI 2.5.1.19-1, dated

October 12, 2005 and November 11, 2005, the applicant expanded the single aging

management for the Forked River Combustion Turbine to twelve aging management

programs. This periodic inspection program is one of twelve programs that monitor the

aging effects of the Forked River Combustion Turbine.

The Periodic Inspection - Forked River Combustion Turbine aging management program

manages the aging effect of change in material properties, loss of material and reduction

of heat transfer for systems, components, and environments. The aging effects are

managed by periodic inspections that will be condition monitoring examinations

performed at susceptible locations in the systems, intended to assure that existing

environmental conditions are not causing material degradation that could result in a loss

of system intended functions. These inspections will be performed on a periodicity not to

exceed once every 10 years and will coincide with major combustion turbine

maintenance inspections.

The two Forked River Combustion Turbines are owned, operated, and maintained by

FirstEnergy, under contract to supply station blackout services to Oyster Creek

Generating Station. The inspectors reviewed program bases documents, maintenance

rule performance data, walkdown reports, and action logs. Applicable portions of the

Interconnect and Station Blackout Agreements were reviewed. The inspectors also

conducted interviews and performed walkdowns with Oyster Creek Generating Station

and FirstEnergy personnel of the Forked River Combustion Turbine facility and portions

of its switchyard. The inspectors observed maintenance activities conducted by General

Electric for FirstEnergy on Forked River Combustion Turbine #2 during a minor outage.

Though the Forked River Combustion Turbines are operated and maintained by

FirstEnergy, Oyster Creek Generating Station assigns a system engineer to monitor their

performance via monthly data sets and logs received from the onsite FirstEnergy

engineers. Significant events and maintenance on the Forked River Combustion Turbine

are logged and evaluated by the Oyster Creek system engineer for further action.

At the time of this inspection, the implementing procedures for this program were not

developed. Hence, the aging management program elements have not been negotiated

with FirstEnergy to be added into the Station Blackout Agreement. The Office of Nuclear

Reactor Regulation accepted AmerGens response to 2.5.1.15-1 and 2.5.1.19-1 and

requests for additional information. Based on discussions with applicant personnel and

reviews of supporting documents, the inspectors concluded that the applicant has plans

to develop adequate guidance and terms for implementation of the Periodic Inspection -

Forked River Combustion Turbine Program. AmerGen will negotiate those terms into the

station blackout agreement.

For the Periodic Inspection - Forked River Combustion Turbine Program, the inspectors

concluded the applicant had conducted adequate evaluations as well as industry

experience and historical reviews to determine aging effects managed by an aging

management program. The applicant provided an acceptable plan to implement

adequate guidance and terms to ensure aging effects are appropriately managed.

Enclosure

16

ASME,Section XI, Subsection IWE Program

The ASME,Section XI, Subsection IWE Program is an existing program modified for

the purpose of aging management credited with managing the aging effects in drywell

containment systems in Oyster Creek Generating Station. ASME Section XI,

Subsection IWE provides for inspection of primary containment components and the

containment vacuum breakers system piping and components. It covers steel

containment shells and their integral attachments; containment hatches and airlocks,

seals and gaskets, containment vacuum breakers system piping and components, and

pressure retaining bolting. The aging effects are managed by periodic visual

inspections, and periodic ultrasonic testing wall thickness measurements. Additionally,

the applicant will conduct monitoring of leakage from the drywell sand bed region drains

going forward, as an additional method to detect conditions favorable for corrosion to

occur. Only the visual and ultrasonic examinations are given credit for managing the

affects of aging.

The inspectors reviewed all of the licensees ultrasonic thickness testing inspection

results for the condition of the drywell from 1983 through 2002, evaluations and

calculations of corrosion rates and projections of wall thickness for several locations on

the drywell. Also, the inspectors reviewed video records of the sand bed region

condition and the removal of the sand from the sand bed region. The inspectors

reviewed the structural analysis performed to confirm the structural integrity of the

drywell after the amount of corrosion had been determined. The inspectors reviewed

the most recently completed visual inspection results of the drywell sand bed exterior

coating and the UT measurements from higher elevations of the drywell.

For the ASME,Section XI, Subsection IWE Program, the inspectors concluded the

applicant had conducted adequate evaluations as well as industry experience and

historical reviews to determine aging effects managed by the ASME,Section XI,

Subsection IWE Program. The applicant provided adequate guidance to ensure aging

effects are appropriately managed, pending resolution of Safety Evaluation Report

Open Items OI 4.7.2-1.1 through OI 4.7.1-1.4, and OI 4.7.2-3.

Protective Coating Monitoring and Maintenance Program

The Protective Coating Monitoring and Maintenance Program is an existing program

credited with managing the aging effects on the internal and external surfaces of the

torus and the condition of the drywell in the sand bed region in systems in Oyster Creek

Generating Station. The aging effects are managed by visual inspections of the

protective coatings on each component, and examination, evaluation and repair of all

coating defects observed.

The inspectors reviewed the past inspection results in each area to understand what

conditions are being documented, the method of evaluation of recorded indications, the

repair methods used to fix any damaged or degraded coating. The inspectors also

looked at the licensees cause determination for the underlying corrosion phenomena

and actions being taken to monitor the condition.

Enclosure

17

The team concluded that as long as the coating integrity was maintained by this

program, the presence of water, as indicated by collection from the former sandbed

area drains, would not affect the rate of corrosion of the drywell at the former sandbed

area.

For the Protective Coating Monitoring and Maintenance Program, the inspectors

concluded the applicant had conducted adequate evaluations as well as industry

experience and historical reviews to determine aging effects managed by the Protective

Coating Monitoring and Maintenance Program. The applicant provided adequate

guidance to ensure aging effects are appropriately managed.

Above-Ground Outdoor Tank Monitoring Program

The Above-Ground Outdoor Tank Monitoring Program is a new program credited with

managing the aging effects on above ground steel tanks in systems at the Oyster

Creek Generating Station. The aging effects will be managed by periodic visual

inspections, some nondestructive evaluation inspections based upon maintenance

history and industry experience.

The inspectors reviewed the Oyster Creek Generation Station template used to guide

and control this inspection effort, conducted field walkdowns of four of the tanks

covered by the program and reviewed the industry operating experience which the

licensee has used to prepare this inspection program.

For the Above-Ground Outdoor Tank Monitoring Program, the inspectors concluded the

applicant had conducted adequate evaluations as well as industry experience and

historical reviews to determine aging effects managed by the Above Ground Outdoor

Tank Monitoring Program. The applicant provided adequate guidance to ensure aging

effects are appropriately managed.

ASME Section XI, Subsection IWF

The ASME Section XI, Subsection IWF Program is an existing program credited with

managing the aging effects in the ASME Section XI, Subsection IWF. Subsection IWF

provides for periodic visual examination of ASME Section XI Class 1, 2, 3 and MC

components and piping support members for loss of mechanical function and loss of

material. Bolting is also included with these components, inspecting for loss of material

and for loss of preload by inspecting for missing, detached, or loosened bolts.

The aging effects are managed by periodic visual examinations for corrosion and loss

of material in structural members, loss of preload in bolting; missing, detached, or

loosened members or bolts; and any degradation of protective coatings. The program

has been enhanced by including additional MC components in the approved ASME

Section XI, Inservice Inspection program.

Enclosure

18

The inspectors reviewed the program description, program basis documents, the

currently approved ASME Section XI, Subsection IWF program, and the results of

previous inspections and examinations. The documents reviewed and discussions with

cognizant individuals indicated the operating experience of the In-service Inspection

program at Oyster Creek, which includes ASME Section XI, Subsection IWF aging

management activities, has not shown any adverse trend. Periodic self-assessments of

the program have been performed to identify the areas that need improvement to

maintain the quality and integrity of the program. The proposed aging management

program based on the ASME Section XI, Subsection IWF, is generally consistent with

the elements of XI.S3 of NUREG-1801 with some exceptions; e.g., NUREG1801

specifies ASME Section XI, 2001edition, including the 2002 and 2003 addenda,

whereas, the station program is based on ASME Section XI, 1995 edition with 1996

addenda, an acceptable alternate edition of the code. The enhancements include

additional MC supports and inspection of underwater supports.

For the ASME Section XI, Subsection IWF Program, the inspectors concluded the

applicant had conducted adequate evaluations as well as industry experience and

historical reviews to determine aging effects managed by an aging management

program. The applicant provided adequate guidance to ensure aging effects are

appropriately managed.

Masonry Wall Program

The Masonry Wall Program is credited with managing the aging effects in masonry

walls at the Oyster Creek Generating Station as part of the Structural Monitoring

Program. The aging effects are managed by a program of inspection of masonry walls

for cracking on a frequency of four years to assure that the established evaluation basis

for each masonry wall remains valid during the period of extended operation.

The inspectors reviewed the program description, program basis documents, the

currently approved station procedures, the results of prior inspections, discussions with

cognizant personnel, and a walkthrough visual examination of accessible masonry walls

to assess the effectiveness of the current program. The scope of the program includes

all masonry walls that perform intended functions in accordance with 10 CFR 54.4, and

were covered by I. E.Bulletin 80-11.

The inspections are implemented though station procedures. Maintenance history has

revealed minor degradation of masonry block walls; but none that could impact their

intended function. In response to I.E.Bulletin 80-11, Masonry Wall Design," and

Information Notice 87-67, Lessons Learned from Regional Inspections of Licensee

Actions in Response to I.E.Bulletin 80-11," various actions have been taken. Actions

have included program enhancements, follow-up inspections to substantiate masonry

wall analyses and classifications, and the development of procedures for tracking and

recording changes to the walls. These actions have addressed all concerns raised by

I.E.Bulletin 80-11 and Information Notice 87-67, namely unanalyzed conditions,

improper assumptions, improper classification, and lack of procedural controls. A

review of operating experience indicates that the program is effective for managing

aging effects of masonry walls.

Enclosure

19

For the Masonry Wall Program, the inspectors concluded the applicant had conducted

adequate evaluations as well as industry experience and historical reviews to determine

aging effects managed by an aging program. The applicant provided adequate

guidance to ensure aging effects are appropriately managed.

Structures Monitoring Program

The Structures Monitoring Program is an existing program that has been modified, and

will be further modified, for the purpose of aging management of structures and

structural components, including structural bolting within the scope of license renewal at

Oyster Creek Station. The program was developed based on Regulatory Guide 1.160,

Revision 2, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, and

NUMARC 93-01 Revision 2, Industry Guidelines for Monitoring the Effectiveness of

Maintenance at Nuclear Power Plants, to satisfy the requirement of 10 CFR 50.65,

Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power

Plants.

The scope of the program also includes condition monitoring of masonry walls and

water-control structures as described in the Masonry Wall Program and in the RG 1.127, and Inspection of Water-Control Structures Associated With Nuclear Power

Plants aging management program. The enhanced program includes structures that

are not monitored under the current term but require monitoring during the period of

extended operation. Aging effects are managed by periodic visual inspections by

qualified personnel to monitor structures and components for applicable aging effects.

Specifically, concrete structures are inspected for loss of material, cracking, and a

change in material properties. Steel components are inspected for loss of material due

to corrosion. Masonry walls are inspected for cracking, and elastomers will be

monitored for a change in material properties. Earthen structures associated with

water-control structures and the Fire Pond Dam will be inspected for loss of material

and loss of form. Component supports will be inspected for loss of material, reduction

or loss of isolation function, and reduction in anchor capacity due to local concrete

degradation.

Exposed surfaces of bolting are monitored for loss of material, due to corrosion, loose

nuts, missing bolts, or other indications of loss of preload. The scope of the program

will be enhanced to include structures that are not monitored under the current term but

require monitoring during the period of extended operation.

The inspectors reviewed the program description, program basis documents, the

currently approved station procedures, the results of prior inspections, discussions with

cognizant personnel, and a walkthrough visual examination of accessible structural

items, including reinforced concrete and structural steel members, components and

systems to assess the effectiveness of the current program. The scope of the program

also includes all masonry walls that perform intended functions in accordance with

10 CFR 54.4, and were covered by I. E.Bulletin 80-11. The inspections included a

review of station procedures, maintenance history, inspection findings and followup of

inspection findings, and current inspection schedules. Inspection frequency is every

Enclosure

20

four years; except for submerged portions of water-control structures, which will be

inspected when the structures are dewatered, or on a frequency not to exceed ten

years. The program contains provisions for more frequent inspections to ensure that

observed conditions that have the potential for impacting an intended function are

evaluated or corrected in accordance with the corrective action process. The

Structures Monitoring Program is consistent with the ten elements of aging

management program XI.S6, "Structures Monitoring Program," specified in NUREG-

1801.

For the Structures Monitoring Program, the inspectors concluded the applicant had

conducted adequate evaluations as well as industry experience and historical reviews to

determine aging effects managed by an aging management program. The applicant

provided adequate guidance to ensure aging effects are appropriately managed.

Inspection of Water Control Structures

The Inspection of Water Control Structures Program is an existing program modified for

the purpose of the aging management program credited with managing the aging

effects in Water Control Structure systems at the Oyster Creek Generating Station.

The aging effects are managed by periodic inspections of the water control structures

for structural and hydraulic degradation, and potential loss of function of intended

service. The Water Control Structure Program is a subpart of the main Structures

Monitoring Program. It is based on the guidance provided in RG 1.127 and ACI 349.3R

and will provide for periodic inspection of the Intake Structure and Canal, the Fire Pond

Dam, and the Dilution structure. The program will be used to manage loss of material,

cracking, and change in material properties for concrete components, loss of material

and change in material properties for wooden components, and loss of material, and

loss of form of the dam, and the canal slopes. Inspection frequency is every four years;

except for submerged portions of the structures, which will be inspected when the

structures are de-watered, or on a frequency not to exceed ten years. The program will

be enhanced to ensure that water-control structures aging effects are adequately

managed during the period of extended operation.

The inspectors reviewed the program description, program basis documents, the

currently approved station procedures, the results of prior inspections, discussions with

cognizant personnel, and a walkthrough visual examination of accessible water control

structures, including components and systems to assess the effectiveness of the

current program. As the Water Control Structures Monitoring Program is a subpart of

the larger Structures Monitoring Program, this review was performed in conjunction with

the comprehensive review of the main Structures Monitoring Program. Inspection of

Water-Control Structures Associated with Nuclear Power Plants program is consistent

with the ten elements of aging management program.

For the Inspection of Water Control Structures Program, the inspectors concluded the

applicant had conducted adequate evaluations as well as industry experience and

historical reviews to determine aging effects managed by an aging management

program. The applicant provided adequate guidance to ensure aging effects are

appropriately managed.

Enclosure

21

Metal Fatigue of Reactor Coolant Pressure Boundary

The Metal Fatigue of Reactor Coolant Pressure Boundary Aging Management Program

monitors select components in the reactor coolant pressure boundary by tracking and

evaluating contributing plant events. The Metal Fatigue of Reactor Coolant Pressure

Boundary program monitors operating transients and, by way of a computer program,

calculates up-to-date fatigue usage factors.

The design basis metal fatigue analyses for the reactor coolant pressure boundary are

considered time limited aging analysis for the purposes of license renewal. The Metal

Fatigue of Reactor Coolant Pressure Boundary Program provides an analytical basis for

confirming that the number of cycles, established by the analysis of record, will not be

exceeded before the end of the period of extended operation. In order to determine

cumulative usage factors more accurately, the program will implement FatiguePro

fatigue monitoring software. FatiguePro calculates cumulative fatigue using both

cycle-based and stress-based monitoring. This provides an analytical basis for

confirming that the number of cycles established by the analysis of record will not be

exceeded before the end of the period of extended operation.

For the Metal Fatigue of Reactor Coolant Pressure Boundary Aging Management

Program, the inspectors reviewed the program including the basis calculations, ongoing

monitoring, corrective actions, limiting components, and current cumulative usage

factors for the limiting components. The applicant provided adequate guidance to

ensure aging effects are appropriately managed.

Isolation Condenser System Review

The Oyster Creek license renewal application listed a number of plant systems within

the scope of license renewal. From this list the inspectors selected the isolation

condenser system for a focused review to determine whether the applicants aging

management programs were adequate to effectively manage aging effects related to

this component. The following aging management programs are credited for managing

aging effects of the isolation condenser system: ASME Section XI In-service

Inspection, Subsection IWB, IWC, and IWD; Bolting Integrity; BWR Stress Corrosion

Cracking; One Time Inspection; Structures Monitoring Program; and, Water Chemistry.

The inspectors focused on the loss of material aging effect to determine how it would

be managed by the identified programs applied specifically to the Isolation Condenser

System.

Although the Oyster Creek 10 CFR 50, Appendix K, design basis event analysis, no

longer takes credit for the Isolation Condenser it is very important for post-accident heat

removal and mitigation of event consequences. It ranks very high on the probabilistic

risk worth for this reason. Because of its risk importance the inspectors reviewed the

aging management programs given credit for managing the affects of aging in the

system.

Enclosure

22

The Isolation Condenser System contains safety-related components relied upon to

remain functional during and following design basis events. For example the primary

coolant boundary must be maintained through the condenser. Additionally the failure of

nonsafety-related structures and components in the Isolation Condenser System could

potentially prevent the satisfactory accomplishment of a safety-related function. The

isolation condenser also performs functions that support fire protection and station

blackout.

AmerGen is committed, in their application documents, to maintaining the water

environment of the secondary side because the integrity of the heat exchanger tubes

can be affected from both the inside and outside. Additionally, the heat exchanger

shell, and therefore, the secondary water environment, is part of the One-Time Aging

Management program because it is required to maintain structural integrity during a

design basis earthquake to support the heat exchanger tubing and the attached reactor

coolant/steam line piping.

The applicant proposed using the ASME Section XI In-service Inspection, Subsections

IWB, IWC, and IWD aging management program with the water chemistry aging

management program to manage loss of material of the Isolation Condensers. The

inspectors reviewed the Oyster Creek in-service inspection program procedure ER-OC-

330-1001, ISI Program Plan Fourth Ten-Year Inspection Interval to verify that it was

modified to include inspections of the isolation condenser tube side components, eddy

current testing of the tubes, and inspection (VT or UT) of the tube sheet and channel

head to ensure that degradation is not occurring and the components intended function

will be maintained. The inspectors reviewed selected NDE reports of isolation

condenser system piping and components, where degradation would result, to verify

compliance with ASME Section XI Code.

The inspectors reviewed the UT wall thickness data sheet 96-023-03 from 1R16

refueling outage which documented shell thickness measurements of the B Isolation

Condenser. The UT results indicate that the shell thickness was over 0.350 inches in

most cases with one reading at 0.312 inches. The vendor drawing 1691-655-20

indicates the shell thickness is 0.375 inches with 0.100 inches corrosion allowance with

a minimum of 0.275 inches. Therefore, the B Isolation Condenser meets the original

design specifications. The inspectors noted that coating inspections performed by the

applicant of the inside shell surface of the B Isolation Condenser during 1R16 in 1996

blistering of the coating was observed in most of the submerged sections. The coating

on the inside shell of the isolation condensers is not credited in the aging management

programs.

Based on discussions with applicant personnel and reviews of the One-time Aging

Management Program basis documents, the inspectors determined the applicant has

elected to perform a one-time aging management program inspection of the shell of

isolation condenser prior to entering extended plant operations.

Enclosure

23

b. Observation

The inspectors identified an observation related to the monitoring of liquid leakage from

the former drywell sand bed region related to the current operating license period. This

observation was determined not to be safety significant and has been entered into the

applicants ongoing corrective action system.

A current commitment for monitoring the sand bed drains is in a staff Safety Evaluation

Report transmitted by letter November 1, 1995. This Safety Evaluation Report

requested a commitment to perform inspections 3 months after the discovery of any

water leakage. Subsequent correspondence from General Public Utilities Nuclear

Corporation the licensee, at the time, clarified the commitment after discussions with

the staff. The commitment made and accepted by the staff in a February 15, 1996,

letter was to perform an evaluation of the impact of any leakage during power

operations and conduct additional inspections of the drywell approximately 3 months

after discovery of the water leakage if the evaluation determines that it is warranted.

This commitment was not meant to apply to minor leakage from normal refueling

activities.

During the inspection, the NRC team requested a walkdown of the torus room.

AmerGen staff walked down the torus room prior to the NRC team making entry. Water

collection jugs, fed by tubing from the former drywell sand bed drains, were emptied

prior to the NRCs walkdown, without taking samples of the water in the jugs or

recording water levels. The fact that water was present in the jugs meant that leakage

had been occurring. The applicant informed the NRC team that the bottles had been

improperly emptied without measurement or analysis. Upon further investigation, the

applicant could not find documentation that showed prior surveillance of the water

drains had been completed. AmerGen staff also could not find documented evidence

that strippable coating of the refueling channel had been applied. This strippable

coating is used as a measure to limit or prevent water leakage during refueling

operations.

The applicant stated that, although there was no formal leakage monitoring in place,

there has been no previous reported evidence of leakage from the former sand bed

drains. Issue Report #348545 was submitted into the corrective action process when

the missed commitment and the improper emptying of the bottles were discovered.

This corrective action will capture the commitment in the applicants computerized

scheduling process so that the required actions will be automatically prompted.

Because there was no previously reported leakage, the applicant did not investigate the

source of leakage, take corrective actions, evaluate the impact of leakage, or perform

additional drywell inspections.

The applicant further stated that a number of actions had been taken to alleviate the

previous water leakage problem since discovery of the consequent drywell shell

corrosion in the early 1990's. Some of the significant actions consisted of inspections

of the reactor cavity wall, remote visual inspection of the trough area below the reactor

cavity bellows seal area, and subsequent repair of the trough area and clearing of its

drain. Clearing of the trough drain and repair of the trough routed any leakage away

from the drywell shell. In addition, AmerGen believes that the strippable coating was

Enclosure

24

applied to the reactor cavity walls before the reactor cavity is filled with water as part of

refueling activities to minimize the likelihood of leakage into the trough area.

The license renewal application does not take credit for the use of the strippable

coating, or the monitoring of the water leakage in managing the aging affects on the

liner. As long as the coating of the exterior surface of the former sand bed area is

maintained, any amount of water can be present and have no affect on the corrosion

rate. The thickness of the cylindrical portion of the liner is managed using ultrasonic

testing and this program will capture any changes in corrosion rate due to water in the

liner gap. AmerGen has taken corrective actions to ensure, in the future, the drains are

monitored, and the strippable coating is applied.

c. Overall Findings

The inspection verified that there is an adequate approach to monitor and control the

effects of aging so that the intended function(s) of systems, structures, and

components, for which an aging management review is required, will be maintained

consistent with the current licensing basis during the period of extended operation. The

inspection verified documentation, procedures, guidance, and personnel, appropriately

supported the license renewal application.

40A6 Meetings, Including Exit

The inspectors presented the inspection results to Mr. T. Rausch, Oyster Creek

Generating Station Vice President, and other members of the licensees staff in a

meeting that was open for public observation on September 13, 2006. The licensee

had no objections to the NRC observations. No proprietary information was provided to

the inspectors during this inspection. The State of New Jersey, Department of

Environmental Protection attended the exit meeting, and made a statement at the

meeting concerning the observation associated with the monitoring of liquid leakage

from the former drywell sandbed drains. In addition, they stated that a letter concerning

this issue was sent to the NRC Region I Regional Administrator dated September 13,

2006. A copy of this letter is available in the NRC ADAMS document management

system under ML062630218.

Enclosure

A-1

ATTACHMENT

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

J. Camire System Manager

L. Corsi Mechanical Engineer, LR Project

M. Gallagher Vice-President, License Renewal

J. Hufnagel Licensing Lead, LR Project

K. Muggleston Mechanical Engineer, LR Project

A. Ouaou Civil Engineer, LR Project

F. Polaski License Renewal Manager

T. Quintenz Site Lead Engineer, LR Project

D. Warfel Technical Lead, LR Project

R. Francis App J. Program Manager

K. Muggleston Licensing

S. Getz, License Renewal

L. Corsi License Renewal

R. Gayley FAC Program Manager

J. Watley CCCW System Engineer

C. Roth TBCCW System Engineer

R. Artz Chemist

M. Miller License Renewal

T. Trettel Fire Protection System Engineer

J. Yuen System Engineer - Ventilation

C. Micklo License Renewal

J. Esch FirstEnergy Engineer

R. Bonelli FirstEnergy Engineer

R. Skelskey System Engineer - FRCT

M. Filippone System Manager

E. Johnson System Manager

R. Pruthi System Manager

S. Schwartz System Manager

D. Spamer Senior Engineer, Electrical

Attachment

A-2

LIST OF DOCUMENTS REVIEWED

Drawings

Complete Set of License Renewal Drawings:

LR-BR-2002, Rev. 0

LR-BR-2003, Rev. 0

LR-BR-2004, Rev. 0

LR-BR-2005, Rev. 0

LR-BR-2006, Rev. 0

LR-BR-2007, Rev. 0

LR-BR-2008, Rev. 0

LR-BR-2009, Rev. 0

LR-BR-2010, Rev. 0

LR-BR-2011, Rev. 0

LR-BR-2012, Rev. 0

LR-BR-2013, Rev. 0

LR-BR-2014, Rev. 0

LR-BR-2015, Rev. 0

LR-BR-M0012, Rev. 0

LR-FP-SE-5419, Rev. 0

LR-GE-107C5339, Rev. 0

LR-GE-148F262, Rev. 0

LR-GE-148F437, Rev. 0

LR-GE-148F444, Rev. 0

LR-GE-148F711 Rev. 0

LR-GE-148F712, Rev. 0

LR-GE-148F723, Rev. 0

LR-GE-148F740, Rev. 0

LR-GE-197E871, Rev. 0

LR-GE-234R166, Rev. 0

LR-GE-237E487, Rev. 0

LR-GE-237E756, Rev. 0

LR-GE-237E798, Rev. 0

LR-GE-713E802, Rev. 0

LR-GE-865D741, Rev. 0

LR-GE-885D781 Rev. 0

LR-GU-3E-243-21-1000, Rev. 0

LR-GU-3E-551-21-1000, Rev. 0

LR-GU-3E-551-21-1001, Rev. 0

LR-GU-3E-666-21-1000, Rev. 0

LR-GU-3E-822-21-1000, Rev. 0

LR-GU-3E-861-21-1000, Rev. 0

LR-GU-3E-861-21-1001, Rev. 0

LR-GU-3E-861-21-1002, Rev. 0

LR-GU-3E-862-21-1000, Rev. 0

Attachment

A-3

LR-GU-3E-871-21-1000, Rev. 0

LR-JC-147434, Rev. 0

LR-JC-19479, Rev. 0

LR-JC-19616, Rev. 0

LR-JC-19629, Rev. 0

LR-OC-010520, Rev. 0

LR-SN-13432.19, Rev. 0

Other Drawings

Drawing 4059-2, Sheet 2 or 3, Reactor Bldg. First Floor At Elev. 23' 6", Sections & Details -

SH.2

Drawing 3E-SK-5-85, 1986 Drywell Data UT Location Plan

Drawing BE-SK-S-89, Revision 0, 10/16/89; Ultrasonic Testing Drywell Level 30'2" - 67'5"

M0123, Post Accident Sampling Isometric, Rev. 2

M0124, Post Accident Sampling Isometric, Rev. 2

M0278, Diesel Fuel Oil Storage Tank Isometric, Rev. 0

GU 3E-000-A3-002, Sheet 7, Rev.1, Isometric Composite Various Systems IGSCC Weld

History

Foster Wheeler Drawing1691-655-20, Outline & Section of Emergency Condenser, Rev. F

Drawing 4059-2, Sheet 2 or 3, Reactor Bldg. First Floor At Elev. 23' 6", Sections & Details -

SH.2

Drawing 3E-SK-5-85, 1986 Drywell Data UT Location Plan

Drawing BE-SK-S-89, Revision 0, 10/16/89; Ultrasonic Testing Drywell Level 30'2" - 67'5"

Procedures

MA-AA-723-500, Inspection of Non-EQ Cables and Connections for Managing Adverse

Localized Environments, Draft Rev 2A.

621.3.005, High Radiation Monitor Calibration, Draft Rev 48A.

621.3.002, Air Ejector Off Gas Radiation Monitor Check Source Functional Test, Draft Rev

26A.

2400-SMI-3623.09, Calibration and Operation of the LPRM Diagnostic System, Rev 11.

2400-SMI-3623.08, IRM Detector Current-Voltage (I/V) Testing, Rev 6.

2400-SMI-3623.03, IRM, SRM, LPRM, Characterization Trending and Diagnostics, Rev 7.

2400-SME-3780.05, Power Factor Testing of 5kV Cables, Rev 2.

2400-SME-3780.06, Dielectric Testing for 2.3kV and 5kV Cables and Equipment, Rev 8.

Exelon Technical Specification for Distribution System Wood Pole Inspection and Remediation,

Dated 1/1/05.

ECR OC 05-00275-00: Revise C-1302-187-E310-037, Revision 2

ER-AA-330-007

ER-AA-335-018

ER-AA-330, Revision 3: Conduct Of Inservice Inspection Activities

ER-AA-330-007, Revision 3: Visual Examination Of Section XI Class MC Surfaces And Class

CC Liners

Attachment

A-4

ER-AA-330-018, Revision 2: General, VT-1, VT-1C, VT-3 And VT-3C, Visual Examination Of

ASME Class MC And CC Containment Surfaces And Components

2400-GMM-3900.52, Revision 3: Inspection And Torquing Of Bolted Connections

SM-AA-300, Revision 0; Procurement Engineering Support Activities

WO R2064827-06, Disassemble Reactor Vessel For Refuel Outage, Prepare Areas & Apply

Cavity Coating, 11/1/06

WO R2068582-03, Perform Reactor Vessel Reassembly, Remove Cavity Coating And Decon

Cavity, 11/1/06

Procedure No. 666.5.007, Revision 16; Primary Containment Integrated Leak Rate Test

PP-03, Criteria for Scoping Systems and Structures Relied upon to Demonstrate Compliance

with 10 CFR 54.4 (a)(2), Rev. 3

PP-04, Systems and Structures Relied upon to Demonstrate Compliance with 10 CFR 50.63 -

Station Blackout, Rev. 4

PP-05, Systems and Structures Relied upon to Demonstrate Compliance with 10 CFR 50.62 -

ATWS, Rev. 1

PP-13, Abnormal Operating Transients, Rev. 2

PP-15, Standard Materials, Environment, and Aging Effects, Rev. 5

Inspection Sample Basis, Aug. 16, 2005

License Renewal Project Level Instruction 5 (PLI-5), Aging Management Reviews, Rev. 5

2400-GMM-3900.52, Inspection and Torquing of Bolted Connections, Rev. 3

2400-SMM-3900.04, System Pressure Test Procedure (ASME XI), Rev. 8

ER-AA-330-008, Protective Coatings, Rev. 3

ER-AA-2030, Attachment 4, System Walkdown Standards, Rev. 3

SA-AA-117, Excavation, Trenching, and Shoring, Rev. 3

SA-AA-117, Excavation, Trenching, and Shoring, Rev. 4b

SP-1302-12-261, Specification for Pipe Integrity Inspection Program, Rev. 7

SP-9000-06-004, Specification for Application and Repair of Service Level III Coatings, Rev. 0

101.2, Oyster Creek Fire Protection Program, Rev. 54

CC-AA-211, Fire Protection Program, Rev. 1

645.6.003, Fire Hose Station, Hose House and Fire Hydrant Inspection, Rev. 17

645.6.007, Fire Protection System Flush, Rev. 15

645.6.011, Deluge and Sprinkler System Inspection, Rev. 10

645.6.013, Fire Suppression System Halon Functional Test, Rev. 19

645.6.026, Fire Damper Inspection, Rev. 11

645.6.017, Fire Barrier Penetration Surveillance, Rev. 10

ER-OC-330-1001, ISI Program Plan Fourth Ten-Year Inspection Interval (Draft)

OC-2, IGSCC Inspection Plan Fourth Ten-Year Inspection Interval, Rev. 1

ER-OC-330-1002, IGSCC Inspection Plan Fourth Ten-Year Inspection Interval (Draft)

ER-AA-330-002, In-service Inspection of Section XI Welds and Components, Rev. 5a

ER-AA-330-009, ASME Section XI Repair/Replacement Program, Rev. 4a

ER-AA-380, Rev. 3, Primary Containment Leakage Rate Testing Program

ER-OC-380, Rev. 0, Oyster Creek Containment Leakage Rate Testing Program

MA-AA-723-500, 50-Year Sample Testing of Fire Water System Sprinkler Heads, Rev. 0

(Draft)

New Oyster Creek PM Task defined in AR 00330592.20, Wall Thickness Measurements of Fire

Water Systems, March 2006 (Draft)

Attachment

A-5

R0801533-Annual, Recurring work task for Fire Pond Screens & Rake Clean and Lubricate

(System 176)

Oyster Creek Generating Station Procedure No. 665.5.020, Rev. 19, Integrated Local Leak

Rate Test Summary

Oyster Creek Generating Station Procedure No. 327.1, Rev. 31, Fuel Oil Receipt and Fuel

Handling Procedure

Oyster Creek Generating Station Procedure No. 828.7, Rev. 22, Secondary Systems Analysis:

Plant Oil

ER-AA-430, Rev. 1, Conduct of Flow Accelerated Flow Accelerated Corrosion Activities

ER-AA-430-1001, Rev. 1, Guidelines for Flow Accelerated Corrosion Activities

CY-OC-120-110, Rev. 0, Chemistry Limits and Frequencies

CY-AA-120-400, Rev. 8, Closed Cooling Water Chemistry

CY-AB-120-1000, Rev. 2, BWR Chemistry Optimization

CY-AB-120-1100, Rev. 3, Reactor Water Hydrogen Water Chemistry, Noble Chem and Zinc

Injection

CY-AB-120-320, Rev. 2, Control Rod Drive Water Chemistry

CY-AB-120-310, Rev. 2, Suppression Pool/Torus Chemistry

CY-AB-120-300, Rev. 5, Spent Fuel Pool

CY-AB-120-200, Rev. 4, Storage Tanks Chemistry

CY-AB-120-130, Rev. 4, BWR Shutdown Chemistry

CY-AB-120-120, Rev. 4, BWR Startup Chemistry

CY-AB-120-100, Rev. 8, Condensate and Feedwater Chemistry

CY-AB-120-100, Rev. 7, Reactor Water Chemistry

CY-OC-120-1107, Rev. 0, Fuel Oil Sample and Analysis Schedule

ECR OC 05-00275-00: Revise C-1302-187-E310-037, Revision 2

ER-AA-330-007

ER-AA-335-018

ER-AA-330, Revision 3: Conduct Of Inservice Inspection Activities

ER-AA-330-007, Revision 3: Visual Examination Of Section XI Class MC Surfaces And Class

CC Liners

ER-AA-330-018, Revision 2: General, VT-1, VT-1C, VT-3 And VT-3C, Visual Examination Of

ASME Class MC And CC Containment Surfaces And Components

2400-GMM-3900.52, Revision 3: Inspection And Torquing Of Bolted Connections

SM-AA-300, Revision 0; Procurement Engineering Support Activities

WO R2064827-06, Disassemble Reactor Vessel For Refuel Outage, Prepare Areas & Apply

Cavity Coating, 11/1/06

WO R2068582-03, Perform Reactor Vessel Reassembly, Remove Cavity Coating And Decon

Cavity, 11/1/06

Procedure No. 666.5.007, Revision 16; Primary Containment Integrated Leak Rate Test

PM0001AC (RTWO R0800287) Control Room HVAC Sys B Inspection, Draft Revision

PM01279M (RTWO R0802279) Lubricate SGTS Fan EF-1-8, Draft Revision

ST 654.3.004 Control Room HVAC System A Flow and Differential Pressure Test, Draft

Revision

ST 651.4.001 Standby Gas Treatment System Test, Draft Revision

PM Draft for AR00330592.40-Measure wall thickness for SGTS buried exhaust ducts/trains A &

B, dated March 13, 2006

ST 678.4.004 Station Blackout Combustion Turbine - Test, Rev. 7

Attachment

A-6

Aging Management Review Technical Basis Documents

OC-AMR-2.3.3.17, Hardened Vent System, Rev. 0

Documents

OC-AMR-2.3.1.3, Rev. 0, Isolation Condenser System

OC-AMR-2.3.3.15 Vol 1, AMR Technical Basis Document - Fire Protection System, March

2006

OC-AMR-2.3.3.15 Vol 2, AMR Technical Basis Document - Fire Protection System, March

2006

OCLR Tracking Item AMP-213 response, Fire Protection System B.1.19 Operating Experience

review to support frequency of visual and functional testing of halon and CO2 Fire

Suppression Systems, dated January 24, 2006

OCLR Tracking Database Open Item 1571, OC-AMR-2.3.3.15 Fire Protection Table

Incomplete, dated February 01, 2006

Interconnection Agreement (Partial) for the Oyster Creek Nuclear Generating Station between

AmerGen Energy Company LLC and Jersey Central Power & Light Company d/b/a

GPU Energy - Schedule A: Interconnection Facilities, dated October 15, 1999

Station Blackout Agreement Between GPU Energy and AmerGen Energy Company, L.L.C,

dated April 14, 2000

Closed Cycle Cooling Water Chemistry Assessment: Oyster Creek Nuclear Generating Station

Final Report

Report of NRC Information Requests Concerning Oyster Creek License Renewal Application,

Topic: Closed Cycle Cooling Water

BWRVIP-130: BWR Vessel and Internals Project - BWR Water Chemistry Guidelines, 2004

Revision

NFPA-25, Standard for the Inspection, Testing and Maintenance of Water-Based Fire

Protection Systems, 1998 Edition

AM-2003-07, Corporate Engineering Oversight Self-Assessment Report, Oyster Creek Flow

Accelerated Corrosion Program, February 2003

EPRI NSAC-202L-R2, Recommendation for an Effective Flow-Accelerated Corrosion

Program, April 1999

EPIR TR-109623, Rev. 2003, Erosion, Corrosion, and Flow Accelerated Corrosion

SP-1302-12-237, Rev. 11, Nuclear Safety Related Pipe Wall Thinning Inspections for Oyster

Creek Nuclear Generating Station Erosion/Corrosion Program

TDR #943, Rev. 3, Oyster Creek Flow Accelerated Corrosion Inspection History

Focus Area Self-Assessment Report, Reactor Water Chemistry Control, April 2004

NEI 03-08 Guidelines for the Management of Materials Issue, May 2003

EPRI BWRVIP-62: BWR Vessel and Internal Project Technical Basis for Inspection Relief for

TDR-1048, Technical Data Report-SGTS Duct Failure in the Tunnel, Rev. 0

EPRI 1007933, Aging Assessment Field Guide, Dec 2003

BWR Internals Components with Hydrogen Injection (TR-108705) - Final Report, December

1998

System 743, SBO Combustion Turbine and Support System OC-7 Functional Failure

Definition, dated April 24, 2002

Attachment

A-7

ASME Code Case -597

Oyster Creek UFSAR, Section 6.2: Containment Systems

Video Tape #1; Bay 7, 0-62 Wall, 9/26/92; Bay 3, 701-739 Heavy Scale

Video Tape #2; Bay 11 & 17,Coating Video Exam, 8/9/96

Video Tape #3; Bay 3, 2/26 & 2/27/92

Video Tape #4; Core Bore Drill Drywell Liner

Video Tape #5; Bay 9, 2/20/92, 10:51 AM, No Drywell Wall

Video Tape #6; Bay 9, 2/18 & 2/19/92

Video Tape #7; Bay 9 & Bay 7, 2/21 & 2/24/92; Bay 7, 600-695, 705 Wall Condition; Bay 3,

2/25/92 w. Arauera 721 Rota Router F804 TUD Guide Tube; Vacuuming from guide

tube Reban 45-56

Video Tape #8; Bay 9, 2/13/92, No drywell wall

Video Tape #9; Bay 4, 2/28 & 3/2/92

Video Tape #10; 2/11/92 Inspection Bay 11, 2/13/92, Bay 11 Rota Router in Drain 11, Bay 7

0-62 Wall, 9/26/92

Video Tape #11; Drywell Liner Sample Areas Plug #s 1, 2, 3, 4, 5, 6,& 7, Bay 7

System Manager Walkdowns of Service Water System, completed Sept. 24, 2005 & Dec. 1,

2005

Operating Experience Review - Hardened Vent - System 822, Rev. 0

Topical Report 116, Oyster Creek Underground Piping Program Description and Status, Rev. 1

Topical Report 140, Emergency Service Water & Service Water System Piping Plan, Rev. 2

Technical Data Report 829, Pipe Integrity Inspection Program, Rev. 4

System & Structure Scoping Form - Torus Water Storage and Transfer System, Rev. 1

System & Structure Scoping Form - Source Range Monitoring, Rev. 0

Assessment of Structural Support at SR/NSR Interfaces

IR & CR

CR

AR 00330592.24.01 to .22

AR 00441639

AR 002114568

IR 469998

IR 471363

Program Basis Documents

PBD-AMP-B.1.5, Rev. 0, BWR Feedwater Nozzle

PBD-AMP-B.1.7, Rev. 2, BWR Stress Corrosion Cracking

PBD-AMP-B.1.11, Rev. 0, Flow Accelerated Corrosion

PBD-AMP-B.1.14, Rev. 0, Closed Cycle Cooling Water Systems

PBD-AMP-B.1.19, Program Basis Document - Fire Protection System, March 2006

PBD-AMP-B.1.19, Program Proof Document - Fire Protection System, March 2006

PBD-AMP-B.1.20, Program Basis Document - Fire Water System, March 2006

PBD-AMP-B.1.20, Program Proof Document - Fire Water System, March 2006

Attachment

A-8

PBD-AMP-B.1.29, Rev. 0, 10 CFR Part 50, Appendix J

PBD-AMP-B.1.02, Rev. 0, Water Chemistry

PBD-AMP-B.2.5, Rev. 0, Periodic Inspection Program

PBD-AMP-B.2.5a, Rev. 0, Periodic Inspection Program-FRCT

PBD-AMP-B.22, Rev. 0, Fuel Oil Chemistry

PBD-AMP-B.2.04, Program Basis Document - Periodic Inspection of Ventilation Systems,

Rev. 0

Program Basis Document, PBD-AMP-B.1.27, Revision 0, ASME Section XI, Subsection IWE

Program Basis Document, PBD-AMP-B.1.33, Revision 0, Protective Coating Monitoring And

Maintenance Program

Program Basis Document, PBD-AMP-B.1.21, Aboveground Steel Tanks

PBD-AMP-B.1.12, Bolting Integrity, Rev. 0

PBD-AMP-B.1.13, Open Cycle Cooling Water System, Rev. 0

PBD-AMP-B.1.24, One-Time Inspection, Rev. 0

PBD-AMP-B.1.26, Buried Piping Inspection, Rev. 0

System Health Reports

In-service Inspection Program 1st, 2nd, 3rd & 4th Quarter 2005 Reports

Oyster Creek Appendix J Program, 4th Qtr 2005 System Health Report

Flow Accelerated Corrosion Program Controlling Document: ER-AA-430, 4th Qtr. 2003 System

Health Report

FRCT Walkdown Report, dated October 17-19, 2005

System 743, CT-1 & CT-2 two year Maintenance Rule Performance data, dated March 2006

Calculations

13432.46-Z-012, Pipe Supports Design - Instrument Air and Nitrogen, Rev. 0

C-1302-187-8610-030 Statistical Analyses of Drywell Thickness Data thru September 1996"

C-1302-187-5300-028, Rev. 0, Statistical Analyses of Drywell Thickness Data thru September

1994"

C-1302-87-5300-021, Rev. 0, Statistical Analyses of Drywell Thickness thru May 1992"

030681, Rev. 0, CHECWORK Flow Accelerated Corrosion Model, August 4, 2003

010663-02, Rev. 1, Flow Accelerated Corrosion Susceptible Non-Modeled Analysis, August 4,

2003

Oyster Creek Nuclear Power Plant, Unit no. 1; Primary Containment Design Report

TDR 277, Revision 0; Oyster Creek Pressure Suppression Chamber Materials Coating

Evaluation, 7/10/85

TDR 851, Revision 0; Assessment Of Oyster Creek Drywell Shell, 12/27/88

TDR 854, Revision 1; Drywell Sand Bed Region Corrosion Assessment, 4/22/87

TDR 948, Revision 1; Statistical Analysis Of Drywell Thickness Data, 2/1/89

TDR 922, Revision 1; 8/5/88, Drywell Upper Elevation - Wall Thinning

TDR 1080, Revision 0; Oyster Creek - Torus Internal Coating

TDR 948, Revision 1, 2/1/89; Statistical Analysis Of Drywell Thickness Data

TDR 854, Revision 1, 4/2/87; Drywell Sand Bed Region Corrosion Assessment

Attachment

A-9

Calculation C-1302-187-5300-008, Revision 0, Statistical Analysis Of Drywell Thickness Data

Thru 2/8/90

Calculation C-1302-187-5300-005, Revision 0,2/1/89; Statistical Analysis Of Drywell Thickness

Data Thru 12/31/88

Calculation C-1302-187-5300-011, Revision 1,6/13/90; Statistical Analysis Of Drywell

Thickness Data Thru 4/24/90

Calculation C-1302-187-5300-017, Revision 0, Statistical Analysis Of Drywell Thickness Data

Thru May 1991

Calculation C-1302-187-5300-021, Revision 0, Statistical Analysis Of Drywell Thickness Data

Thru May 1992

Calculation C-1302-187-5300-015, Revision 0; Statistical Analysis Of Drywell Thickness Data

Thru March 1991

Calculation C-1302-187-5300-016, Revision 0; OCDW Projected Thickness Using Data Thru

3/3/91

Calculation C-1302-187-5300-022, Revision 0; OCDW Projected Thickness Using Data Thru

5/31/92

Calculation C-1302-187-5300-025, Revision 0; Statistical Analysis Of Drywell Thickness Data

Thru December 1992

Calculation C-1302-187-5300-025, Revision 1; Statistical Analysis Of Drywell Thickness Data

Thru December 1992

Calculation C-1302-187-5300-028, Revision 0; Statistical Analysis Of Drywell Thickness Data

Thru September 1994

Calculation C-1302-187-310-037, Revision 2; Statistical Analysis Of Drywell Vessel Thickness

Data Thru September 2000

Calculation C-1302-187-E310-037, Revision 1; Statistical Analysis Of Drywell Vessel Thickness

Data Thru September 2000

Calculation C-1302-241-E610-081, Revision 2; Suction Strainer Debris Generation and

Transport, 10/5/98

Calculation C-1302-187-5320-024, Revision 0; OC Drywell Ext. UT Evaluation in Sandbed,

4/16/93

Document 990-2174, Letter Report On Additional Sandbed Analyses, GE Materials Monitoring

& Structural Analysis Services, 12/11/92

TDR 1108, Revision 0; Oyster Creek Drywell Vessel Corrosion Mitigation

CC-AA-309-1001, Revision 0, Oyster Creek Torus Corrosion Allowable Pit Depth; Calculation

C-1302-187-E310-038

Engineering Evaluation 82-74-9, Oyster Creek Torus Shell Thickness, 7/15/76

Engineering Evaluation 82-74-4, Oyster Creek Torus Shell Thickness, 3/4/76

Calculation C-1302-187-E310-038, Revision 0; Oyster Creek Torus Corrosion Allowable Pit

Depth

Calculation C-1302-187-5360-006, Revision 0; O. C. Drywell - Projected Thickness Thru June

1992, 1/27/89

Calculation C-1302-187-8610-003, Revision 0; Statistical Analysis of Drywell Thickness Data

thru September 1996

ECR OC 05-00275-000, Revise C-1302-187-E310-037 Revision 1 To Revision 2

ER 84-006-00, Oyster Creek, Torus Corrosion Pitting and Missing Structural Welds

Attachment

A-10

Specifications

Isolation Condenser B Inspection and Re-coating, Rev. 0, 2/29/1996

Specification 100579-000, 8/3/77; Specification For Coating The Exterior Of The Torus Oyster

Creek Nuclear Generating Station, Toms River, New Jersey

Specification SP-1302-06-009, Revision 3, 7/25/91; Specification For Application And Repair

Of Service Level I Coatings On Ferrous Metal Surfaces; Oyster Creek Nuclear

Generating Station, Toms River, New Jersey

EPRI TR-109937, Guideline on Nuclear Safety-Related Coatings, April 1998

Specification IS-328227-005, Revision 12: Functional Requirements For Drywell Containment

Vessel Thickness Examination

Specification IS-402950-001; Functional Requirements For Augmented Drywell Inspection

Specification SP-1302-08-002, Revision 1; Inspection Of The Torus Coating, 2/17/83

Specification SP-1302-52-094, Revision 1; Drywell Shell Coating Touch-Up

Specification SP-1302-52-120; Inspection And Localized Repair Of The Torus And Vent

System Coating

Specification SP-1302-32-035, Revision 0; Inspection And Minor Repair Of Coating On

Concrete & Drywell Shell Surfaces In The Sandbed Region

Specification 9000-06-003, Revision 4; Application And Repair Of Service Level II And Balance

Of Plant Coatings

Specification #125-75-10, Torus Shell Welding Repair, 6/13/77

Specification OCIS 328001-001, Installation Specification For Torus Coating, Oyster Creek

Nuclear Generating Station Pressure Suppression Chamber, 6/28/83

Specification SP-1302-52-120, Revision 2, 10/03/02; Specification For Inspection And

Localized Repair Of The Torus And Vent System Coating

Safety Evaluations

Safety Evaluation SE-000243-002, Revision 14

Safety Evaluation 000243-002, Revision 0: Drywell Shell Plate Thickness Reduction At The

Base Sand Cushion Entrenchment Region

Safety Evaluation 402950-005, Revision 3: Removal Of Sand From Drywell Sand Bed

Safety Evaluation 315403-019, Revision 1: Drywell Design Pressure Reduction - Tech Spec

Change

Safety Evaluation 000243-002, Revision 14, 8/2/95: Drywell Steel Shell Plate Thickness

Reduction

Safety Evaluation 000187-004, Revision 0: Inspection/Repair Of Torus/Vent System Coating

Safety Evaluation 000187-001, Revision 1, 1/14/91: Evaluation Of Blistered Torus Coating

SP-1302-06-013, Post-Fire Safe Shutdown Program Requirements at Oyster Creek Nuclear

Generating Station, Rev. 1

Safety Evaluation, SE-000822-023, Safety Evaluation - Repair of SGTS Duct at the Stack,

Rev. 0

Attachment

A-11

NRC Documents

NRC Regulatory Guide 1.147

NRC Ltr. C321-95-2235/5000-95; Oyster Creek Nuclear Generating Station (OCNGS) Docket

No. 50-219 Facility Operating License No. DPR-19, Drywell Corrosion Monitoring

Program

NRC Information Notice 89-79: Degraded Coatings And Corrosion Of Steel Containment

Vessels

NRC Information Notice 89-79, Supplement 1: Degraded Coatings And Corrosion Of Steel

Containment Vessels

NRC Information Notice 97-10: Liner Plate Corrosion In Concrete Containments

NRC Generic Letter 98-04; Potential for Degradation of the Emergency Core Cooling System

and the Containment Spray System After a Loss-Of-Cooling Accident Because of

Construction and Protective Coating Deficiencies and Foreign Material in Containment,

7/14/98

NRC Ltr. Dated 1/5/87; Docket No. 50-219; December 10, 1986, Meeting With GPU Nuclear

Corporation (GPUN) To Discuss Corrosion Of The Outer Surface Of The Drywell Shell

NRC Inspection Report No. 50-219/87-27

NRC Ltr. Dated 10/16/90, Docket No. 50-219; Drywell Corrosion Program - Oyster Creek

Nuclear Generating Station

NRC Inspection Report No. 50-219/90-21

NRC Information Notice 86-99, Supplement 1: Degradation Of Steel Containments

NRC Ltr. Dated 11/1/95, Docket No. 50-219; Drywell Corrosion Program - Oyster Creek

Nuclear Generating Station

NUREG-1522 Assessment of Inservice Conditions of Safety-Related Nuclear Plant Structures,

6/95

NRC Ltr. Dated 12/29/86, Docket No. 50-219; Interim Operation For Cycle 12 Following

Corrosion Of The Outer Surface Of The Drywell Shell (TAC 64016)

NRC Ltr. Dated 9/22/87, Docket No. 50-219; Licensee Actions Taken And Action Plans For

Mitigating The Corrosive Attack On The Drywell Sheel Of The Oyster Creek Nuclear

Generating Station (TAC 65448)

NRC Ltr. Dated 9/13/93, Docket No. 50-219; Issuance Of License Amendment No. 165 -

Change In Containment Drywell Design Pressure of 62 psig To New Design Pressure

of 44 psig

NUREG-6706 Capacity of Steel and Concrete Containment Vessels With Corrosion Damage,

2/01

NUREG-0661 Safety Evaluation Report Mark I Containment Long-Term Program, 7/80

Licensee Letters

AmerGen Ltr. 2130-05-20037, Oyster Creek Generating Station Refueling Outage 20 (1R20)

Inservice Inspection (ISI) Summary Report

GPUN Ltr. 5000-86-1116, 12/18/86: Oyster Creek Nuclear Generating Station Docket No.

50-219, Licensing No. DPR-16, Oyster Creek Drywell Containment

GPUN Ltr. 5/12/87: Oyster Creek Nuclear Generating Station Docket No. Generic Letter 87-05

GPUN Ltr. 5200-87-0061, 5/29/87: Oyster Creek Nuclear Generating Station Docket No.; NRC

Meeting 6/11/87

Attachment

A-12

GPUN Ltr. 5000-89-1717, 2/9/89: Oyster Creek Nuclear Generating Station Docket No,

License No. DPR-16; Drywell Containment

GPUN Ltr. 5000-90-1995, 12/5/90: Oyster Creek Nuclear Generating Station Docket No.

50-219, License No. DPR-16, Oyster Creek Drywell Containment

GPUN Ltr. 5000-95-2235, 9/15/95: Oyster Creek Nuclear Generating Station Docket No.

50-219, License No. DPR-16, Drywell Corrosion Monitoring Program

GPUN Ltr. 1940-99-20661, 12/17/99: Oyster Creek Nuclear Generating Station Docket No.

50-219, ASME XI Relief Requests

Jersey Central Power & Light Company Ltr. EA-76-686, 7/16/76: Oyster Creek Nuclear

Generating Station Docket No. 50-219, Oyster Creek Torus Shell Thickness Evaluation

GPUN Ltr. C321-93-2153, 5/25/93; Oyster Creek Nuclear Generating Statio Docket No. 50-219

Reactor Containment Building Integrated Leak Rate Test

GPU Nuclear SDBD-OC-243(MPR), Revision 11; Design Basis Document For Containment

System Oyster Creek Nuclear Generating Station

Applicant Response Letter for NRC RAI 2.5.1.19-1, dated October 12, 2005

Applicant Supplemental Response Letter for NRC RAI 2.5.1.19-1, dated November 11, 2005

Vendor Documents

Condition Assessment of Cable Circuits at Exelon AmerGen Oyster Creek Nuclear Power

Plant, Dated 6/1/05

Condition Assessment of Cable Circuits at Exelon AmerGen Oyster Creek Nuclear Power

Plant, Dated 8/4/04

VM -OC-5001, Care and Operation of Isolation Condensers, Rev. 2

Oyster Creek Nuclear Power Plant Unit No. 1, Primary Containment Design Report, prepared

by Ralph M. Parsons Company for GE

Final Inspection Report, Torus Coating Inspection and Repair and ECCS Suction Strainer

Replacement; S.G. Pinney & Associates

MPR-1322, Revision 0, Results of Painting Process Qualification Tests For the Drywell Exterior

in the Sand Bed Area at Oyster Creek

GE Evaluation No. 87-178-003, Revision 1; Corrosion Evaluation Of The Oyster Creek Drywell,

3/6/87

Final Report Summary Of Torus Decontamination Underwater Inspection And Coating Repair,

12R Outage, MCF Job 9401; by S.G. Pinney & Associates, Inc., 4/21/89

Final Report, Exelon/AmerGen, Oyster Creek Nuclear Generating Station, 1R19 Refueling

Outage, Torus Desludging, Torus Coating And Corrosion Inspection, Torus Coating

Repair; Underwater Construction Corporation, 12/2/2002

Final Engineering Report, No. FER-7047, Revision 0; GPU Nuclear, Oyster Creek Nuclear

Generating Station, 1R13 Coating And Corrosion Inspection Report, Torus Immersion

And Vent Header; Underwater Engineering Services, Inc., 7/29/91

Final Engineering Report, Torus Coating Inspection And Repair and ECCS Suction Strainer

Replacement, GPU Nuclear, Oyster Creek Nuclear Generating Station, by S.G. Pinney

& Associates, Inc., 12/22/98

MPR Calculation 83-179-001, Revision 0; Oyster Creek Torus Shell Thickness Margin For

Fatigue Loading

Attachment

A-13

S. G. Pinney Report 990-2587, 12/16/96; Design Basis Accident (DBA) and Irradiation Testing

of Coating Repair Materials for Use in Boiling Water Reactor Suppression Chamber

Immersion Areas

GE Technical Report TR-7377-1, Justification For Use Of Section III, Subsection NE, Guidance

In Evaluating The Oyster Creek Drywell, November 1990

DRF #00664, Index No. 9-1, Revision 0; An ASME Section III Evaluation Of Oyster Creek

Drywell Part I Stress Analysis, November 1990

DRF #00664, Index No. 9-2, Revision 0; An ASME Section III Evaluation Of Oyster Creek

Drywell Part II Stability Analysis, November 1990

DRF #00664, Index No. 9-3, Revision 0; An ASME Section III Evaluation Of Oyster Creek

Drywell For Without Sand Case, Part I Stress Analysis, February 1991

DRF #00664, Index N0. 9-4, Revision 2; An ASME Section III Evaluation Of Oyster Creek

Drywell For Without Sand Case; Part 2 Stability Analysis, November 1992

SGPAI Procedure QCP-10-2-OCNGS-7101, Revision 2, 11/30/92; Underwater Coating Repair

Work Orders

C2012115

C2003517

A/R # A2101209

C2009903, DTE Testing of Medium Voltage Cables, Dated 6/20/05

C2008036, DTE Testing of Medium Voltage Cables, Dated 6/30/05

R2046003, Radiation Monitor Functional Test, Dated 12/21/05

R2071072, High Radiation Monitor Calibration and Test, Dated 10/6/05

R2073609, NI Cable Test Data Review, Dated 9/1/05

R0807890, IRM/SRM Characterization Trending and Diagnostics, Dated 12/2/04

R0808284, Containment Spray Nozzle Verification, Dated 10/21/02

00543366, Containment Spray Nozzle Test, Dated 11/9/00

00034392, Engineering Evaluation of Containment Spray Nozzle #6, Dated 1/4/93

Corrective Action Program

O2004-2340 O2005-1445 O2003-0318 O2000-0309

O2005-2249 O2005-0265 O2004-0200 O2001-0634

O2003-0799 O2000-1772 O2005-1152 O2000-1531

O2003-2454 O2004-3586 O2004-3745 O2000-1429

O2003-2493 O2004-3437 O2004-3550 O2000-0401

O2004-0990 O2000-1429 O2003-1865 O2000-0634

O2004-2304 O2000-1355 O2003-2076 O2005-1445

O2004-1542 O2002-1564 O2005-2249 O2005-0265

O2002-1842 O2004-3442 O2003-0799 O2000-1772

O2000-0309 O2005-1350 O2003-2454 O2004-3586

O2001-0634 O2000-1578 O2003-2493 O2004-3437

O2000-1531 O2000-1607 O2004-0990 O2000-1429

O2000-1429 O2003-1903 O2004-2304 O2000-0609

O2000-0401 O2003-0488 O2004-1542 O2004-2153

O2000-0634 O2002-1280 O2002-1842 O2005-1772

Attachment

A-14

O2003-1000

O2004-0313

O2004-1644

O2003-1308

O2004-1314

O2002-1937

O2000-1788

A2097892

A2107513

A2073455

00472707*

00472141*

00471363*

00469998*

00471867*

00472346

00470325

00472090

00472346

00470325

00472090

00461639

00348545*

• As a result of this inspection

WO R0806127-01

WO R2027889-01

WO R2071967-06

CAP O2003-2586, #2 Diesel-Driven Fire Pump Cooling Water Line pin-hole leaks

CAP O2002-0916, Water leakage from base of Fire Hydrant #9

CAP O2005-2288, Debris in SGTS #1 Filter Train

Aging Management Programs

PBD-AMP-B.1.27, Revision 0, ASME Section XI, Subsection IWE

PBD-AMP-B.1.33, Revision 0, Protective Coating Monitoring And Maintenance Program

PBD-AMP-B.1.21, Aboveground Steel Tanks

PBD-AMP-B.1.34, Electrical Cables and Connections not Subject to 10 CFR 50.49

Environmental Qualification Requirements, Rev 0

PBD-AMP-B.2.01, Periodic Testing of Containment Spray Nozzles, Rev 0

PBD-AMP-B.1.35, Electrical Cables and Connections not Subject to 10 CFR 50.49

Environmental Qualification Requirements used in Instrumentation Circuits, Rev 0

Attachment

A-15

PBD-AMP-B.1.36, Inaccessible Medium-Voltage Cables not Subject to 10 CFR 50.49

Environmental Qualification Requirements, Rev 0

PBD-AMP-B.2.06, Wooden Utility Pole Program, Rev 0

License Renewal Change Requests

LRCR 291

LRCR 290

LRCR 289

LIST OF ACRONYMS

ADAMS Agency-wide Documents Access and Management System

ASME American Society Mechanical Engineers

PARS Publicly Available Records

GALL Generic Aging Lessons Learned Report

Attachment