ML062650059
| 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
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
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.
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-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-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 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 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