ML033370007
| ML033370007 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 12/02/2003 |
| From: | Crlenjak R Division of Reactor Safety I |
| To: | Mecredy R Rochester Gas & Electric Corp |
| References | |
| IR-03-008 | |
| Download: ML033370007 (26) | |
See also: IR 05000244/2003008
Text
December 2, 2003
Dr. Robert C. Mecredy
Vice President, Nuclear Operations
Rochester Gas and Electric Corporation
89 East Avenue
Rochester, New York 14649
SUBJECT:
R. E. GINNA - NRC INSPECTION REPORT NO. 05000244/2003008
Dear Dr. Mecredy:
On October 22, 2003, the U.S. Nuclear Regulatory Commission (NRC) completed the second
inspection of your application for renewal of the operating license for the R. E. Ginna Nuclear
Power Plant focusing on the manner by which you managed the effects of aging on systems,
components or structures previously determined to be within the scope of license renewal. The
results of the inspection, including a description of the inspection and its findings, were
progressively shared with members of your staff on July 25, August 8, and during a public exit
meeting on October 22, 2003.
The inspection was conducted in accordance with NRC Manual Chapter 2516, Policy and
Guidance for the License Renewal Inspection Program, using NRC Inspection Procedure 71002, License Renewal Inspections. The inspection was the second scheduled NRC team
inspection supporting your application for a renewed license for the R. E. Ginna facility. The
inspection consisted of a selected examination of procedures, representative records, and
interviews with personnel regarding the aging management of systems, structures and
components within the scope of license renewal in accordance with 10 CFR 54, in your license
renewal application.
The aging management portion of your license renewal activities was generally implemented or
planned as described in your license renewal application. The documentation supporting your
application was in an auditable and retrievable form. During the inspection the team identified
five items for which your staff must take further action to ensure that your aging management
programs are fully effective and consistent with regulatory guidance. Commitment and action
tracking system items have been generated for each of these issues. Except for these items,
the team determined that your aging management programs can acceptably identify and
manage the aging of the structures, systems, and components within the scope of license
renewal for the extended period of operation.
Dr. Robert C. Mecredy
2
In accordance with 10 CFR 2.790 of the NRCs Rules of Practice, a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) component of 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/
Richard V. Crlenjak, Deputy Director
Division of Reactor Safety
Docket No.
50-244
License No.
Enclosure:
Inspection Report 05000244/2003008
cc w/encl:
J. Laurito, President, Rochester Gas and Electric
P. Eddy, Electric Division, Department of Public Service, State of New York
C. Donaldson, Esquire, State of New York, Department of Law
N. Reynolds, Esquire, Winston & Strawn
P. Smith, Acting President, New York State Energy Research
and Development Authority
J. Spath, Program Director, New York State Energy Research
and Development Authority
D. Stenger, Ballard, Spahr, Andrews and Ingersoll, LLP
T. Wideman, Director, Wayne County Emergency Management Office
M. Meisenzahl, Administrator, Monroe County, Office of Emergency Preparedness
T. Judson, Central New York Citizens Awareness Network
Dr. Robert C. Mecredy
3
Distribution w/encl:
H. Miller, RA/J. Wiggins, DRA
J. Jolicoeur, RI EDO Coordinator
R. Laufer, NRR
K. Kolaczyk, SRI Ginna
M. Marshfield, DRP
J. Trapp, DRP
N. Perry, DRP
Region I Docket Room (with concurrences)
W. Lanning, DRS
R. Crlenjak, DRS
R. Lorson, DRS
M. Modes, DRS
DOCUMENT NAME: C:\\ORPCheckout\\FileNET\\ML033370007.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/DRS
RI/DRP
RI/DRS
NAME
MModes (via
email)
RLorson
JTrapp
RCrlenjak
DATE
11/26/03
11/26/03
12/01/03
12/02/03
OFFICIAL RECORD COPY
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No:
50-244
License No:
Report No:
Applicant:
Rochester Gas and Electric Corporation
Facility:
R. E. Ginna Nuclear Power Plant
Location:
1503 Lake Road
Ontario, New York 14519
Dates:
July 21 - 25, August 4-8, and September 17, 2003
Inspector:
Michael Modes, Team Leader, Region I
Fred Bower, Senior Reactor Engineer, Region I
Suresh Chaudhary, Senior Reactor Engineer, Region I
Alfred Lohmeier, Senior Reactor Engineer, Region I
Kamalamar Naidu, Senior Reactor Engineer, NRR
Approved by:
Raymond K. Lorson, Chief
Performance Evaluation Branch
Division of Reactor Safety
SUMMARY OF FINDINGS
IR 05000244/2003-009; 07/21/03-10/22/2003; R. E. Ginna Nuclear Power Plant; License
Renewal Application, Aging Management Programs.
This inspection of license renewal activities was performed by four regional specialist inspectors
with assistance from a reactor engineer from the Office of Nuclear Reactor Regulation. The
inspection conformed with NRC Manual Chapter 2516 and NRC Inspection Procedure 71002.
This inspection did not identify any findings as defined in NRC Manual Chapter 0612.
During the inspection the team identified five items for which your staff must take further action
to assure your aging management programs are complete and accurate: Commitment and
Action Tracking System (CATS) item 11329 to assure that license renewal documents are
revised as a consequence of the license renewal review process, CATS 11330 to modify
procedure EP-3P-0169 to clarify the requirements for evaluating bolting and hardware, CATS
11331 to compare M-92.2 to Regulatory Guide 1.127, CATS 11332 to formally notify the NRC
that fire system inspection and flushing periodicity is different than described in NRC NUREG 1801, and CATS 11333 to update the fire water system basis documents to incorporate
revisions and clarification identified by the NRC team inspection.
The inspection team concluded that the aging management programs referred to in Rochester
Gas and Electrics license renewal application were planned and/or conducted as described in
the license renewal application and that documentation supporting the application was in an
auditable and retrievable form.
The inspection team concluded there was reasonable assurance the aging management
processes, as described in the license renewal application, would adequately manage the
effects of aging.
REPORT DETAILS
01
LICENSE RENEWAL AGING MANAGEMENT ACTIVITIES
a.
Inspection Scope
This inspection was conducted to determine if the license renewal application (LRA)
submitted by Rochester Gas Electric Company (RGE), herein referred to as the
applicant, for the R. E. Ginna Station (Ginna), was in accordance with 10 CFR Part 54
for the aging management of systems, structures and components (SSC). The team
evaluated the applicants implementation of the aging management process by
reviewing the aging management programs identified in the application as applied to
selected risk significant plant systems and structures. The inspection objective was to
determine if the programs submitted for these selected systems and structures were
consistent with NRC guidance for license renewal. Applicable NRC guidance included
the statements of consideration that accompanied the license renewal rule (60FR22461,
published May 8, 1995); Regulatory Guide 1.188, Standard Format and Content for the
Application to Renew Nuclear Power Plant Operating Licenses, dated July 2001; and
the draft license renewal standard review plan, Standard Review Plan for the Review of
License Renewal Applications for Nuclear Power Plants, dated April 21, 2000, and
other staff guidance documents. The results of the review in this area are discussed
below.
1.
ASME Section XI, Subsections IWB, IWC, & IWD In-service Inspection Program
(A2.1.2)
The inspectors verified that the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code,Section XI, Subsections IWB, IWC, & IWD, In-service
Inspection (ISI) program was an existing program consistent with NUREG-1801 GALL
Report,Section XI.M1 (ASME Section XI ISI Program, Subsections IWB, IWC, and
IWD), and Section X1.M3 (Reactor Head Closure Studs). The inspectors noted that
aging effects were managed by the applicant through periodic visual, surface and
leakage tests of Class 1, 2, and 3 pressure retaining piping, components and
attachments identified for inspection in ASME Section XI within the scope of license
renewal.
In discussions with the applicants staff and examination of related documentation, the
inspectors reviewed the consistency of the program attributes in the areas of
preventative action, parameters monitored, detection of aging effects, monitoring and
trending, acceptance criteria, corrective actions, confirmation processes, administrative
controls, and operating experience review. The inspectors concluded that the applicant
conducted adequate evaluations as well as historical reviews to determine the aging
effects that can be managed by the ISI program. The applicant developed adequate
guidance to ensure aging effects would be appropriately managed. Thus, there is
reasonable assurance that the ASME XI program reflected in the application will be
maintained through the extended operating period.
2
2.
ASME Section XI, Subsections IWE & IWL In-service Inspection Program (A2.1.3)
The ISI Program is an existing program which is credited with managing the effects of
aging for:
(1) Carbon steel and miscellaneous polymeric materials and components that
provide a containment pressure boundary and leak tight barrier function, and are
tested/inspected per 10CFR50, Appendix J, and ASME,Section XI, Sub-section
IWE requirements.
(2) Containment Post-tensioning System, and
(3) Concrete and embedded steel (rebar) components of the containment that
are inspected per ASME,Section XI, Sub-section IWL.
The ISI program procedures, surveillance test procedures, and historic conditions for the
covered piping and supports were reviewed to determine the effectiveness of the
program. The review included technical adequacy of the procedure, conformance to the
applicable requirements, documentation of results, and corrective actions, if necessary.
The program was based on plant specific design bases, maintenance history, and
regulatory requirements, and included information available through NRC Generic
Letters, Bulletins, and Information Notices. Also, the program had previously been
reviewed by the NRC and found to be acceptable. The inspectors concluded that the
applicant conducted adequate evaluations as well as industry experience and historical
reviews to determine aging effects that can be managed by the ASME,Section XI, ISI
Program. The applicant provided adequate guidance to ensure aging effects would be
appropriately managed. Thus, there is reasonable assurance that covered systems and
components will be maintained through the period of extended operation.
3.
ASME Section XI, Subsection IWF In-service Inspection Program (ISI) (A2.1.4)
The ISI program, developed in accordance with 10 CFR 50.55a, is an existing program
credited to manage the effects of aging in the Class 1, 2, 3, and MC piping, and
components, and their associated supports. Ginna is in the first period of the fourth ten-
year interval of the ASME,Section XI, ISI program. The ISI program procedures,
surveillance test procedures, and historic conditions for the covered piping and supports
were reviewed to determine the program effectiveness, technical adequacy,
conformance to the regulatory requirements, documentation adequacy, and corrective
actions, as necessary.
The program was reviewed by the NRC, and determined to consistent with the licensing
basis of the plant. The inspectors concluded that the applicant conducted adequate
evaluations as well as industry experience and historical reviews to determine aging
effects that could be managed by the ASME,Section XI, ISI Program. The applicant
provided adequate guidance to ensure aging effects would be appropriately managed.
Thus, there is reasonable assurance that covered systems and components will be
maintained through the period of extended operation.
4.
Boric Acid Corrosion Inspection Program (A2.1.5)
3
The boric acid corrosion inspection (BACI) program is an existing program that has
been modified to manage the aging effects of boric acid wastage of non-RCS
components, including cable connectors and cable trays, as well as other susceptible
SSCs on which borated water may leak. The aging effects are managed by minimizing
borated water leakage through frequent monitoring of locations where potential leakage
could occur and by the timely repair of leaks.
Ginnas LRA, Section B2.1.6 - Boric Acid Corrosion, stated that the Ginna boric acid
corrosion control program would be consistent with NUREG-1801 Generic Aging
Lessons Learned (GALL). Ginna procedure IP-IIT-7, Boric Acid Corrosion Monitoring
Program, issued on March 13, 2003, was consistent with the GALL program elements.
The inspectors compared the LR boric acid control program plan to Sections XI.M10
and IP-IIT-7 of the GALL to determine the adequacy of the program. The inspectors
also walked down selected portions of the auxiliary building and the safety injection,
containment spray, spent fuel pool cooling and charging and volume control systems
with the boric acid control program coordinator to determine the effectiveness of the
program. Discrepancies identified during this walkdown were entered into the corrective
action program by the initiation of action requests (ARs) 2003-1766, 1767 and 1768.
The applicant has or is planning to develop additional guidance to implement the BACI
program to ensure aging effects of boric acid corrosion are appropriately managed.
Commitment and Action Tracking System (CATS) item 11329 was initiated to track to
completion the development of the guidance needed to implement the modified BACI
program. There is reasonable assurance the applicant will adequately manage the
effects of aging due to boric acid corrosion through the period of extended operation.
5.
Buried Piping and Tanks Program (A2.1.7)
The buried piping and tanks (BTNK) inspection program is a new aging management
program at Ginna that will use existing site procedures in conjunction with the one-time
inspection program and the periodic surveillance and preventive maintenance (PSPM)
program to manage a loss of material on the pressure-retaining capability of buried
piping and tanks due to generalized pitting, crevice corrosion, and microbiologically
influenced corrosion. Ginna also credited the BTNK Program with managing the aging
effects of external corrosion of the buried piping associated with the fire water and
service water systems and the fuel oil storage tanks for the emergency and technical
support center diesel generators. The applicant plans to perform visual inspections
under the one-time inspection program when buried piping and tanks are excavated for
maintenance or for any other reason. Ultrasonic thickness measurements of the
emergency diesel generator (EDG) fuel oil storage tanks will be periodically performed
in accordance with the PSPM program.
4
In lieu of periodic inspections, the BTNK inspection program credits inspections
performed in accordance with the one-time inspection program. The inspectors verified
that CATS 11320 was initiated to improve procedural guidance to ensure that these
inspections would be identified as part of the work control process.
During excavation in the yard in July 2003, Ginna performed an inspection of opportunity
on the B EDG fuel oil storage tank. The inspectors walked down the area of the
completed excavation, examined pictures taken during the inspection and reviewed the
inspection report. Additionally, the inspectors noted that Ginna is scheduled to perform
internal non-destructive inspections both EDG fuel oil storage tanks during an upcoming
refueling outage.
The applicant has or is planning to develop adequate guidance to implement the BTNK
program to adequately manage the aging effects on the pressure-retaining capability of
buried piping and tanks. Commitment and action tracking system item 11329 was
initiated to track to completion the development of the guidance needed to implement
the new BTNK program as described in the Ginna LR program plan.
The inspectors concluded that the applicant conducted adequate evaluations, as well as
industry experience and historical reviews, to determine aging effects that can be
managed by the BTNK program. Thus, there is reasonable assurance the integrity of
the buried tanks and piping will be maintained through the period of extended operation.
6.
Closed-Cycle Cooling Water (CCW) System Program (A2.1.8)
The CCW system program is an existing program credited with managing the effects of
aging in the cooling water system used to dissipate heat in various components through
out the plant. The aging effects are managed by maintaining control over the water
chemistry and the integrity of the tubes in the CCW heat exchangers.
The inspectors reviewed the CCW system program, supporting procedures, surveillance
test procedures, historic conditions, and water chemistry records. The program includes
preventive measures to minimize corrosion and surveillance testing and inspection to
monitor the effects of corrosion on the intended function of the component. The aging
effects are minimized or prevented by controlling the chemical species that cause the
underlying aging mechanisms. Surveillance testing and inspections were performed in
accordance with Electric Power Research Institute Topical Report TR-107396 to
evaluate the system and component performance. The inspectors noted that the
applicant has recently replaced the CCW heat exchangers tubes with tubes fabricated
from an admiralty brass material.
The inspectors concluded that the applicant conducted adequate evaluations, as well as
industry experience and historical reviews, to determine aging effects that can be
managed by the CCW program. The applicant provided adequate guidance to ensure
aging effects are appropriately managed. Thus, there is reasonable assurance the
integrity of the CCW will be maintained through the period of extended operation.
1.
Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental
Qualification Requirements Program (A2.1.9)
5
The electrical cables and connections not subject to 10 CFR 50.49 environmental
qualification (ECCNS-EQ) program is a new aging management program credited with
managing the effects of aging in cables that are exposed to adverse localized
environments caused by heat, radiation, or moisture. The aging effects are managed by
monitoring parameters such as the temperature of the cables in the balance of plant.
During the mid-1990s, the applicant implemented quarterly walkdowns of cables located
in accessible areas. During every refueling outage, cables in containment as well as
other inaccessible areas are inspected for outer jacket discoloration and cracking. In
cases where the temperatures exceed the established limits, the applicant evaluates the
cables to determine if the outer jacket exhibits deterioration such as, discoloration of
poly vinyl chloride (PVC) cables or cracking. Action requests are expected to be
generated to document and correct unacceptable conditions. The team reviewed
selected ARs and determined that the actions taken to correct the adverse findings were
adequate. Additionally, accessible electrical cables and connections installed in adverse
localized environments are scheduled to be visually inspected at least once every 10
years. The first inspection, for the purpose of license renewal, is planned to be
completed in 2009. This approach is in conformance with the GALL. Operating
experience shows that aging degradation is a slow process. A ten year inspection
frequency will provide two data points during a twenty year period to characterize the
degradation rate.
In the applicants response, dated July 11, 2003, to request for additional information
(RAI) 3.7-6(a), the applicant committed to perform thermographic inspections of 34-5kV
transformer yard components at least once per refueling cycle while the components are
energized. This inspection program is scheduled to begin before the end of the current
license period (i.e., September 2009). In the same letter, the applicant committed to
perform a visual inspection of the phase bus before year 2012. The applicant intends to
develop the ECCNS-EQ program with supporting procedures, surveillance test
procedures and historic conditions for detecting discolored, and cracked insulation
before the extended period of operation commences.
The inspectors concluded that the applicant conducted adequate evaluations, as well as
industry experience and historical reviews, to determine aging effects that can be
managed by the ECCNS-EQ program. The applicant plans to provide adequate
guidance to ensure aging effects are appropriately managed. Thus, there is reasonable
assurance the cables identified in the application will be adequately maintained through
the period of extended operation.
2.
Fire Protection Program (A2.1.10)
The fire protection (FP) program is an existing program which is credited with managing
the effects of aging for the fire seals, fire barriers, fire pumps, and the halon system.
The program manages the aging effects through periodic inspections of fire barriers and
periodic inspection and testing of fire pumps and the halon system. The LR
program basis document for the FP system identified inspections that must be
performed and specific changes that must be made to existing site procedures prior to
the end of the initial operating license for Ginna. The inspectors reviewed the LR FP
program plan and supporting documents to verify the effectiveness of the FP program.
6
The inspectors walked down selected portions of fire protection systems in the turbine,
intermediate and auxiliary buildings with the fire protection system engineer. The
inspectors noted that the LR FP program plan requires the operability of fire dampers to
be verified by drop-testing ten percent of the dampers on a rotating basis so that all
dampers are tested at least once in ten years. The inspectors also noted that
procedural guidance existed that required periodic inspection of selected fire dampers
for mechanical damage.
The inspectors concluded that the applicant conducted adequate evaluations, as well as
industry experience and historical reviews, to determine which aging effects and
systems and components can be managed by the FP program. The applicant has or is
planning to develop adequate guidance to implement the FP program to ensure aging
effects will be managed through periodic inspections of fire barriers and periodic
inspection and testing of fire pumps and the halon system. Commitment and action
tracking system item 11329 was initiated to track to completion the development of the
guidance needed to implement the new FP program.
3.
Fuel Oil Chemistry Program (A2.1.13)
The fuel oil chemistry (FOC) program is an existing program credited with managing
diesel oil used to operate the:
a) Emergency Diesel Generators;
b) Diesel Fire Pump (installed in the pump house);
c) Emergency Diesel Power to the Security System, and
d) Technical Support Diesel.
The aging effects are managed by minimizing the exposure of fuel oil to contaminants,
such as water and microbiological organisms, by periodic draining or cleaning of tanks
and by verifying the quality of new oil before its introduction into the storage tanks. The
applicant developed the fuel oil chemistry program to identify activities credited for
license renewal, and to describe how the program manages the identified aging effects.
identified in the aging management review process. The inspectors noted the applicant
has taken two exceptions to the GALL: a) not using biocides, and b) using a three
instead of an eight micron pore size filter recommended in ASTM Standard 2776. The
applicant currently tests the fuel oil in accordance with ASTM D 4176, Free Water and
Particulate Contamination in distillate Fuels (Clear and Bright Pass/Fail Procedures), as
required by Tech Specs. Because the purchased fuel oil contains a red dye, the
Clear and Bright test criterion is not considered meaningful and the applicant will not
perform this test during the period of extended operation. The principal goal of the FOC
program is to minimize corrosion on the internal surfaces of the diesel fuel storage tanks
and associated components. This is accomplished by following established procedures
that require periodically monitoring the viscosity of the fuel oil, water and sediment
content in the diesel fuel oil being stored.
The applicant purchases fuel as a commercial grade item to meet the requirements of
ANSI/ASTM D975-78, Standard Specification for Diesel Fuel Oils. On receipt, the
applicant performs an evaluation to upgrade the diesel fuel oil to a safety-related item
using procedure, CGIEE 90-001, revision 12, Commercial Grade Items Engineering
7
Evaluation. The CGIEE 90-001 procedure incorporates the requirements of ASTM
975-78. Technicians take samples from three levels of the storage tank and send them
to two independent laboratories to determine if the flash point, cloud point, water and
sediment content, carbon residues, weight percent, distillation temperatures and other
chemical content, are within the acceptable limits specified in ASTM D975-78. If the
samples meet the acceptance criteria, the fuel oil is accepted and stored offsite. When
fuel oil is requisitioned, a tanker collects sufficient oil from the off site storage and
delivers it to the EDG fuel tank(s). Chemistry procedure, CHA-DFOTP, Diesel Fuel Oil
Testing Program, summarizes the major elements of the GINNA Station fuel oil testing
program that are required to meet the requirements of Improved Technical Specification 5.5.12, Diesel Fuel Oil Testing. Every 92 days, the fuel being stored offsite is tested for
viscosity, water and sediment to verify if it meets the acceptance criteria mentioned in
Table 1 of ASTM D-975-78.
Until 1992, the applicant cleaned and inspected the main storage tanks and found no
discernable impurities. After consultation with the engine manufacturers and users
group, Ginna decided to clean and inspect the diesel storage tanks every 10 years. The
oil in the offsite 12,000 gallon fuel storage tank is sampled every 92 days in accordance
with ASTM D875. The oil in the two onsite 6000 gallon tanks is sampled every 60 days
to verify it meets ASTM D975-78 Table 1, for viscosity, water and sediment. In 1993,
the applicant drained the two fuel oil tanks (A&B) completely, cleaned, visually
inspected, pressure tested and refilled them.
The inspectors concluded the applicant conducted adequate evaluations, as well as
industry experience and historical reviews, to determine aging effects that can be
managed by the FOC program. The applicant provided adequate guidance to ensure
aging effects are appropriately managed. Thus, there is reasonable assurance the fuel
oil system can be maintained through the period of extended operation.
4.
Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling
System Program (A2.1.14)
The overhead heavy load and light load handling program is an existing program which
is credited with managing the effects of aging in the cranes and lifting devises for
refueling operations. The program demonstrates that aging effects are managed by
existing testing, surveillance, and maintenance program. The aging effects, such as
loss of material due to corrosion and wear, are evaluated by periodic examinations
under the current maintenance procedures, while over stressing is controlled by load
control procedures and vendor instructions.
8
The crane/lifting devices monitoring program and supporting procedures, testing and
surveillance procedures and historic conditions for maintenance and load control were
reviewed to determine the effectiveness of the program. The inspector noted R. E.
Ginna has included industry experience and NRC generic communications in its
program.
The inspector concluded that the applicant conducted adequate evaluations, as well as
industry experience and historical reviews, to determine aging effects that can be
managed by overhead heavy load and light load handling program. The applicant
provided adequate guidance to ensure aging effects are appropriately managed. Thus,
there is reasonable assurance that covered systems and components will be maintained
through the period of extended operation.
5.
Service Water System (SWS) Program (A2.1.16)
The inspectors verified the essential elements of the SWS program which is credited
with managing the aging effects in the SWS and other functionally related SSCs. Aging
management is accomplished utilizing a program developed in response to NRC Generic Letter 89-13. Aging management utilizes features of the program, such as
surveillance and control of bio-fouling, heat transfer capability, routine inspection and
maintenance related to corrosion, erosion silting, and bio-fouling degradation.
The inspectors discussed, with the applicants staff, the consistency of the SWS
attributes in the areas of preventative action, parameters monitored, detection of aging
effects, monitoring and trending, acceptance criteria, corrective actions, confirmation
processes, administrative controls, and operating experience review.
The inspectors concluded that the applicant conducted adequate evaluations, as well as
historical reviews, to determine aging effects that can be managed by the SWS
program. The applicant provided adequate guidance to ensure aging effects are
appropriately managed. Thus, there is reasonable assurance that the SWS program
reflected in the application will be maintained through the extended operating period.
6.
Periodic Surveillance and Preventive Maintenance (A2.1.17)
The periodic surveillance and preventive maintenance program is an existing reliability-
centered maintenance program. This program includes four key activities:
1.
Inspection and non-destructive testing of components. After completing the
inspections, engineering will review the inspection results, perform an
engineering evaluation, and determine if corrective actions are required, as well
as the need for additional inspections.
2.
Existing preventive maintenance inspections which are performed periodically
based on previous experience.
3.
Aging effect required management items. For these items, the applicant is
committed to preparing check lists for each material environment grouping which
may require management of multiple aging effects.
9
4.
Surveillance tests routinely performed on pumps, motors, diesel generators to
monitor performance.
The aging effects are managed by examining the wall thickness in pipes and flanges to
determine loss of material due to various types of corrosion, such as crevice corrosion,
galvanic corrosion, general corrosion, pitting corrosion, and microbiologically induced
corrosion. The interrogation of the components is implemented to predict rates of
degradation so that components can be replaced in a timely manner. For other
components, such as flexible hoses and connections, the program manages the effects
of changes in material properties and cracking. The program initiated a task to replace
the bladder for the spent fuel pool weir gate on a nine year frequency. When the
bladder is replaced, it will be inspected for degradation and change in material
properties. The inspectors observed that the applicant upgraded the existing
maintenance program, which was originally developed during the past years to routinely
replace aged components, and to address other age-related degradation.
The team concluded that the applicant conducted adequate evaluations, considered
industry experience, and historical reviews to determine aging effects that can be
managed by the existing periodic surveillance and preventive maintenance (PSaPM)
Program. Thus, there is reasonable assurance the existing PSaPM program will assure
that the aging effects are appropriately managed, and that the components will be
maintained through the period of extended operations.
7.
Reactor Vessel Internals Program (2.1.19)
The inspectors verified the reactor vessel internals program is a new program credited
with managing the effects of aging on the reactor vessel internals consistent with GALL,
Section XI.M16, PWR Vessel Internals, except that Ginna implements a VT-3
examination schedule different from that suggested in the GALL.
The inspectors verified that the aging effects are managed by identification of the most
limiting or susceptible components, development of techniques to permit detection and
characterization of degraded components, demonstration of inspection technique
effectiveness, and timeliness of the inspection. In discussions with the applicant and
documentation review, the inspectors reviewed elements of the aging management
program including irradiation embrittlement, neutron irradiation and thermal aging, void
swelling, chemistry programs of ASME Section XI, Subsection IWB, and industry
participation programs.
10
The inspectors concluded the applicant conducted adequate evaluations, as well as
historical reviews, to determine aging effects that can be managed by the reactor vessel
internals program. The applicant provided adequate guidance to ensure aging effects
are appropriately managed. Thus, there is reasonable assurance that the ASME XI
program reflected in the application will be maintained through the extended operating
period.
8.
Reactor Vessel Surveillance Program (A2.1.20)
The inspectors verified that the reactor vessel surveillance program is an existing plant
specific program that consists of ten elements credited with managing the effects of
aging on the reactor vessel and described in Appendix A of NUREG 1800, Standard
Review Plan for Review of License Renewal Applications for Nuclear Power Plants.
The program is generally consistent GALL Section XI.M.31, Reactor Vessel
Surveillance.
The inspectors verified the program monitoring methods are those in accordance with
10 CFR 50, Appendix H, which includes the testing of in-vessel capsules for fracture
toughness. The fracture toughness values are used to calculate an upper shelf energy,
fluence, uncertainty, life expressed as effective-full-power-years, development of
temperature-pressure limitations, and determination of low-temperature over-pressure
protection setpoints. In discussions with cognizant applicant personnel, the inspectors
noted that the applicant has utilized the attributes, history, and supporting evidence that
makes this program applicable to managing the aging effects.
The inspectors concluded that the applicant conducted adequate evaluations, as well as
historical reviews, to determine aging effects that can be managed by the reactor vessel
surveillance program. The applicant provided adequate guidance to ensure aging
effects are appropriately managed. Thus, there is reasonable assurance that the
ASME XI program, reflected in the application, will be maintained through the extended
operating period.
9.
Spent Fuel Pool Neutron Absorber Monitoring Program (A2.1.21)
The spent fuel pool neutron absorber monitoring program is an existing program which
is credited with managing the effects of aging in the borated stainless steel neutron
absorber material used at Ginna. The aging effects are managed by periodic visual
examination of test coupons; thickness measurements taken at representative locations
of creviced/galvanically coupled areas and exposed surfaces; and weighing the coupons
to the accuracy of 0.1 gram. These examinations and measurements are compared to
the reference photographs.
11
The neutron monitoring program and supporting procedures, surveillance test
procedures, and historic conditions for the borated stainless steel coupons were
reviewed to determine the effectiveness of the program. The inspectors noted that
Ginna has an appropriate plant specific program that addresses the ten elements
described in the Appendix A of the NUREG-1800. The intent of the program is
consistent with NUREG-1801.
The inspector concluded that the applicant conducted adequate evaluations, as well as
industry experience and historical reviews, to determine aging effects that can be
managed by spent fuel pool neutron monitoring program. The applicant provided
adequate guidance to ensure aging effects are appropriately managed. Thus, there is
reasonable assurance that covered systems and components will be maintained through
the period of extended operation.
10.
Structures Monitoring Program (A2.1.23)
The structures monitoring program is an existing program that has been modified to
include the structural bolting integrity program which is credited with managing the
effects of aging in the concrete and steel structures and appurtenances. The aging
effects are managed by systematically assessing the physical state of structures and
components by periodic surveillances, examinations, and tests to ensure that the
structures remain in an acceptable condition. The inspectors noted that Ginna has
developed the program to include all safety-related buildings, the containment structure
and structures within the containment, other buildings within the scope of license
renewal, and also some nonsafety-related component supports. It provides periodic
visual examination of concrete and steel structures and components, support steel
members and bolts, water control structures, and surveillance of containment pre-
stressing tendons.
The aging monitoring program describes aging effects, background and
operational/maintenance history, and actions that will assure continued integrity of
systems and components. The methodologies are technically valid and sufficiently
detailed to include known aging mechanisms and manifestations. They are generally
based on plant specific and industry experience.
The program covers the ten aging management program (AMP) attributes described in
the RLSB-1, Aging Management Review-Generic, which is included in Appendix A of
NUREG-1800. The program provides guidance for the attributes and the frequency of
inspection for various structures and components. The results are documented,
reviewed and evaluated for any corrective action, if needed. Additionally, the AMPs
have included information available through NRC Generic Letters, Bulletins, Information
Notices, and Vendor Notifications.
The scope of the monitoring program also includes non-structural items (i.e., joints and
elastomeric seals); architectural items - roofing, siding, and containment facade; and
miscellaneous items - flood barriers, dampers, and cathodic protection. The applicant
has performed surveillance and examinations of structures for serviceability and
structural degradation, and has implemented corrective actions where appropriate (e.g.,
containment pre-stressing tendons). The inspectors determined that the guidance
12
pertaining to the visual inspection and examination of structural bolting and fasteners
was not sufficiently detailed to readily disclose aging effects. The applicant initiated a
CATS item to review this issue.
The inspector concluded that the applicant conducted adequate evaluations, as well as
industry experience and historical reviews, to determine aging effects that can be
managed by the structures monitoring program. The applicants planned and completed
actions and guidance were considered adequate to ensure that aging effects would be
appropriately managed. Thus, there is reasonable assurance that covered systems and
components will be maintained through the period of extended operation.
11.
Thimble Tube Inspection Program (A2.1.25)
The program manages the integrity of the incore neutron monitoring thimble tubes,
which serve as a portion of the reactor coolant pressure boundary. The thimble tube
inspection program (TTIP) is an existing program that has been slightly modified to
include aging management attributes. The applicant has 36 thimble tubes, made of 316
type, stainless steel tubing, nominal OD of 0.300, "and nominal wall thickness of 0.049."
The aging effects are managed by measuring the wall thickness of the stainless steel
tubing every refueling outage, and replacing those which exhibit less than the minimum
required thickness. Multi-frequency eddy current examination of thimble tubes are
performed by qualified non-destructive examiners in accordance with approved
procedures, and the results are tabulated.
The team reviewed the TTIP and supporting procedures, surveillance test procedures
and historic conditions for minimum wall thickness, to determine the effectiveness of the
program. In response to NRC Information Notice No. 87-44, Thimble Tube Thinning in
Westinghouse Reactors, the applicant has performed thimble tube inspections during
every refueling outage (RFO). The focus of the program is to detect thimble tube wall
thinning due to wear caused by flow induced vibration and implement preventive
maintenance such as flushing, cleaning and replacement. Thimble tube wear is
detected at locations associated with geometric discontinuities or area changes along
the reactor coolant flow path. The program provides for evaluation of inspection results
and appropriate corrective actions. During the March 2002 RFO, thimble tube G6 was
retubed due to wear indication at the lower core plate, that was detected during the
September 2000 RFO. The wear indication was estimated at 69% through wall and was
capped and repositioned for planned replacement during the 2002 RFO. Thimble tube
G6 was replaced during the 2002 RFO with a 0.003" chrome plated, wear resistant,
thimble tube. This tube is coated approximately ten feet from the end for a distance of
approximately ten feet. The chrome coating location coincides with the lower core
support area which has shown to initiate thimble tube wear. The applicant will be
performing an examination on all thimble tubes during the next RFO to determine if
chrome plating will diminish the wear rate.
The inspectors concluded that the applicant conducted adequate evaluations, as well as
industry experience and historical reviews, to determine aging effects on thimble tube
can be managed by the program. The applicant provided adequate guidance to ensure
aging effects are appropriately managed. Thus, there is reasonable assurance the
thimble tubes will be maintained through the period of extended operation.
13
12.
Nickel-Alloy Nozzles and Penetrations Inspection Program (New)
The nickel-alloy nozzles and penetrations inspection program (RVH) is a new aging
management program which is credited with managing the aging effect manages crack
initiation and growth due to primary water stress corrosion cracking (PWSCC) of reactor
coolant system alloy 600/690 components, including reactor pressure vessel (RPV)
head penetrations and RPV bottom-mounted head penetrations. The program is also
credited with managing the aging effects of the replacement steam generator (SG) weld
overlay cladding (Alloy 82) on the tubesheets and the weld buttering (Alloy 152) on the
(SG) primary inlet and outlet nozzles. The RVH Program manages the aging effects
through: (a) PWSCC susceptibility assessment using industry models to identify
susceptible components, (b) monitoring and control of reactor coolant chemistry to
mitigate PWSCC, and (c) inspections of reactor vessel head penetrations and
nickel-alloy J-groove pressure boundary welds in accordance with RGEs commitments
to the NRC Order of February 11, 2003, and (d) routine inservice inspections conducted
in accordance with ASME,Section XI, Subsection IWB.
The inspectors reviewed documents supporting the RVH program including draft
procedure VT-116, Visual Examination of Reactor Vessel Head, to determine the
effectiveness of the program. Ginna implemented this procedure to inspect the RPV
bottom-mounted head penetrations during the fall 2003 refueling outage. The ISI
engineer was also interviewed. The inspectors noted that Ginna replaced the RPV head
during the fall 2003 refueling outage. The inspectors also noted that the RVH Program
incorporated the interim inspection requirements of the NRC Order of February 11,
2003. The applicant recognizes that the program may require updating to incorporate
revisions to the ASME Code, PSWCC susceptibility determinations and crack growth
rate information.
The inspectors concluded that the applicant conducted adequate evaluations, as well as
industry experience and historical reviews, to determine which aging effects and
systems and components can be managed by the RVH Program. The applicant has or
is planning to develop adequate guidance to implement the RVH Program to ensure the
aging effects of primary water stress corrosion cracking (PWSCC) are appropriately
managed. Commitment and Action Tracking System (CATS) item 11329 was initiated
to track to completion the development of the guidance needed to implement the
modified RVH program. Thus, there is reasonable assurance that the applicant has
demonstrated that the new RVH Program will adequately manage the effects of aging
due to PWSCC corrosion through the period of extended operation.
14
b.
Conclusion
The inspection team concluded that the aging management portion of Ginnas license
renewal activities were conducted as described in the license renewal application and
that documentation supporting the application is in an easily auditable and retrievable
form. During the inspection the team identified five items for which your staff must take
further action to assure your aging management programs are complete and accurate:
Commitment and Action Tracking System (CATS) item 11329 to assure that license
renewal documents are revised as a consequence of the license renewal review
process, CATS 11330 to modify procedure EP-3P-0169 to clarify the requirements for
evaluating bolting and hardware, CATS 11331 to compare M-92.2 to Regulatory Guide 1.127, CATS 11332 to formally notify the NRC that fire system inspection and flushing
periodicity is different than described in NRC NUREG 1801, and CATS 11333 to update
the fire water system basis documents to incorporate revisions and clarification
identified by the NRC team inspection.
The inspection team also concluded that applicant provided adequate information to
ensure that the aging effects would be appropriately managed. Thus, there is
reasonable assurance that covered systems and components will be maintained through
the period of extended operation.
02 MANAGEMENT MEETINGS
Exit Meeting Summary
The inspector presented the inspection results to Mr. George Wrobel, Project Manager
and other members of the applicants management at a public meeting on October 22,
2003. The applicant acknowledged the observations presented.
The applicant did not indicate that any of the information discussed was proprietary.
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Applicant Personnel:
Robert Mecredy, VP, Nuclear Operations
George Wrobel, Project Manager, License Renewal
David Wilson, License Renewal Engineer
George Herrick, License Renewal Engineer
Gerry Geiken, License Renewal Engineer
Yvonne Selbig, Human Resources, Ginna
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
None
LIST OF DOCUMENTS REVIEWED
General License Renewal Documents
Application for Renewed Operating License - R.E. Ginna - Chapter 2.0 Scoping and Screening
Methodology
NRC Inspection Manual - Inspection Procedure 71002, IMC 2516, License Renewal Inspection,
09/18/00
Regulatory Guide 1.188 - Standard Format and Content for Applications to Renew Nuclear
Power Plant Operating Licenses
Ginna Nuclear Power Plant, License Renewal Application Scoping and Screening Methodology
Mechanical Systems
10 CFR 54 Code of Federal Regulations - Energy - Requirements for Renewal of Operating
Licenses for Nuclear Power Plants
NEI 95-10 Rev. 3 Industry Guideline for Implementing the Requirements of 10 CFR Part 54 -
The License Rule, March 2001.
Ginna Probabilistic Assessment - Chapter 9
NEI 95-10 Industry Guideline for Implementing Licensee Renewal Rule
License Renewal Drawings
33013 -1262 Sheet 1 of 2 - LR and Sheet 2 of 2 - LR, Safety Injection System
33013 -1237 - LR, UFSAR Section 10.5.2, Auxiliary Feed Water
Miscellaneous
IP-QAP-1, Rev. 4; Structures, Systems and Component Safety Classification
System/Structure Scoping Report: LRSP-FIRE; Fire Protection (LR-21), Ginna Nuclear Power
Plant, Rev. 0, 4/30/2002
Procedure FPS-2, Rev. 0, Ginna Station Fire Barrier Penetration Seal Program
Procedure M-103, Rev. 12, Inspection and Maintenance of Fire Dampers, May 3, 2002
2
Procedure FPS-2.1, Rev. 5; Control and Verification of UFSAR and/or 10CFR50 Appendix R
Procedure FPS-1, Rev. 7; Fire Barrier Control Inspection
Procedure ME-302, Rev. 0; Installation Specification Ginna Station Fire Barrier Penetration Seal
Program
F-RAI 2.1-1
F-RAI 2.3.3.6-1
Completed Procedure FPS-2.1, Rev. 5; Control and Verification of UFSAR and/or 10CFR50
Appendix R Fire Barriers; September 2002
Completed Procedure FPS-2.1, Rev. 5; Control and Verification of UFSAR and/or 10CFR50
Appendix R Fire Barriers; October 2002
Completed Procedure FPS-2.1, Rev. 5; Control and Verification of UFSAR and/or 10CFR50
Appendix R Fire Barriers; November 2002
IP-CAP-1; Abnormal Condition Tracking Initiation or Notification (Action) Report
Ginna Station Fire Protection Program Report, Part IV Safe Shutdown Analysis
Ginna License Renewal Aging Management Program (LR-AMP) Basis Document, Nickel-Alloy
Nozzles and Penetrations Inspection Program, LR-RVH-PROGPLAN, Revision 0
Ginna License Renewal Aging Management Program (LR-AMP) Basis Document, Nickel-Alloy
Nozzles and Penetrations Inspection Program, LR-RVH-PROGPLAN, Revision 1 (Draft)
PT-7, ISI System Leakage Test, Reactor Coolant System
VT-110, Visual Examination of the Reactor Vessel and Removable Internal Structures
VT-109, Visual Examination for Leakage
VT-116, Visual Examination of the Reactor Vessel Head, Revision 6 (Draft)
VT-101, Visual Examination Acceptance Criteria, Revision 9 (Draft)
GMS-43-08-Tubesheet, Steam Generator Tubesheet Inspection
Work Order 20100793 - Component Summary I006200 for:
RCS ISI PT Inspection Report 02GP071
RCS ISI RT Inspection Report 02GRT191
RCS ISI PT Inspection Report 02GP069
RCS ISI RT Inspection Report 02GRT184
Ginna License Renewal Aging Management Program (LR-AMP) Basis Document, Boric Acid
Corrosion (BAC) Inspection Program, LR-BAC-PROGPLAN, Revision 0
RAI B2.1.6-1
RAI B2.1.6-2
RAI B2.1.23-2
IP-IIT-7, Boric Acid Corrosion Monitoring Program
IP-CAP-1.9, Boric Acid Leakage Initial Investigation Form
IP-HSC-3, Attachment 2, Part 5, Containment Component Leakage Monitoring Log
PT-39, Leakage Evaluation of Primary Coolant Sources Outside Containment
S-12.4, RCS Leakage Surveillance Record Instructions
S-12.2, Operator Action in the Event of Indication of Significant Increase in Leakage
VT-109, Visual Examination for Leakage
VT-101, Visual Examination Acceptance Criteria
CATS 11329, Update License Renewal Project Documents
3
Action Reports Reviewed:
AR2001-0327; Redundant Equipment With Conflicting Safety Classifications
AR2001-0383; Plant Components Used For Flood Prevention Measures Improperly Safety
Classified
AR2001-0388; Configuration Management Activities Associated With Chemical Nuclear Skid
Inadequate
AR2001-1732; Mis Classification Of Pressure Indicators
AR2001-1817; Solenoid Valve 14423S Misclassified As Safety Equipment
AR2000-0027; C Charging Pump Discharge Manifold Leak
AR2000-0041; SW Leak on Supply Line to Charging Pump Cooler B
AR2000-0061; Inability to Update Anti-virus Files
AR2000-0074; 115 KV Pipe Cable Pothead Leaking Oil
AR2000-0086; Rod Insertion Limit Alarm
AR2000-0134; MCB Alarm G-4 S/G A Hi Level Alert Alarmed Intermittently
AR2000-0200; AOV-110C Outlet Block Valve has Diaphragm Leak
AR2000-0250; Field Calibrator Source Stuck Open
AR2000-0270; New Parts from Stock Were Defective
AR2000-0272; Air Conditioning Unit Overflow in Central Records
AR2000-0277; Appendix R Emergency Lite RR-2 Failed
AR2000-0278; Appendix R Emergency Lite SB-20 Failed
AR2000-0291; Rust Found In LT-935 Displacer Column (Not on Transmitter)
AR2000-0295; Removed Bearing, Defective
AR2000-0306; Flux Mapping System Intermittent Power Supply Failure
AR2000-0325; Simulator Building - Sewer Pump Alarm
AR2000-0329; Battery Room A Structural Steel Fire Proofing Degradation
AR2000-0379; Simulator Problems Disrupted Evaluated Scenario Twice
AR2000-0412; Spare MQ-483 Failed In-storage Maintenance Worksheet
AR2000-0429; Valve Stem on V-691A, Isolation to PI-629B (B RHR Pump Discharge Press.)
AR2000-0468; RCP A and B Seal Delta Temperature Deviation
AR2000-0497; Resistance & Meggar Check Indicates Partial Open Circuit On Jacket Water
Heater
AR2000-0514; Bus 16 C Safety Injection Pump Main Control Board Switch Failed
AR2000-0522; RR-2 Lamp Failure Light is Lit
AR2000-0528; Simulator Aydin Display System Failures
AR2000-0546; Circulating Water Total Residual Chlorine Data Logger Malfunctioning
AR2000-0569; Boric Acid Heat Trace Circuit 21 Primary Has No Load Current
AR2000-0599; Potassium Chromate Leak, Exposure Hazard
AR2000-0629; SAS Failure - No Update
AR2000-0636; Degraded Flange Upstream of V-9545A
AR2000-0752; Increase in Plant Radio Gas Activity
AR2000-0765; Control Rods Moving Within Dead Band
AR2000-0768; Multiple Trouble Alarms On SSA and SSB
AR2000-0807; Excessive Wear Found in HHS Pipe
AR2000-0819; Security Radios Began Breaking Up During Security Contingency Drills
AR2000-0865; Low Meggar Readings On Jacket Water Heaters A D/G
AR2000-0877; Tube Wall Degradation In A D/G Jacket Water Heat Exchanger
AR2000-0900; Environmental Air Sampler Flow Meter Failed Calibration Check
AR2000-0904; Deteriorated Insulation On Electrical Cable
4
AR2000-0917; Radiation Monitor R-24 (AVT Mixed Bed B) Does Not Respond To Source
Check
AR2000-0937; Battery Electrolyte Level In Cell #48 Changed
AR2000-0952; Water Hammer In Area of V-5743B
AR2000-0962; Inline Dionex Computer Failed
AR2000-0964; Plant Vent Radiogas Increase During Performance of PT 2.5.4
AR2000-1033; R-15A Failure
AR2000-1042; Small Steam Leak Visible From The Insulation Around v-5731 In Cnmt
AR2000-1131; Spare Source Range Detector Unable To Be Calibrated
AR2000-1132; Leak Developed Upstream Of Steam Trap ZMS-02
AR2000-1133; AOV-392A Failed To Open With Less Than 257.5 PSID Per PT-2.6.4
AR2000-1140; A Atmospheric Relief Valve Nitrogen Supply Check Valve Failure
AR2000-1151; Minor Oil Spill In Turbine Oil Storage Room
AR2000-1159; Bus 11B Undervoltage Time Delay Relays Out Of Tolerance
AR2000-1184; Relay Holding Current Found Out Of Tolerance
AR2000-1251; Low UT Measurements At The Steam Extraction To 4B Heater Line
AR2000-1257; Debris Found In Turbine Lube Oil Guard Pipe
AR2000-1260; Cracks Identified In The 3516, 3517 Disc Seat Area
AR2000-1262; Wall Thickness For Comp #63 On Dwg M46B Is Below Min Allowable Thickness
AR2000-1263; Wall Thickness For Comp #08 On Dwg M46B Is Below Min Allowable Thickness
AR2000-1266; A Fuel Oil Transfer Pump Discharge Pressure Low
AR2000-1274; Work Activities Under WO 19903422 Results In Degraded HEMYC Wrap
AR2000-1275; Wall Thickness For Comp #2690-2680 On Dwg C381-358 Sht #3 Is Below Min.
Allowable Thickness
AR2000-1276; Cable (Unknown Circuit Schedule) Located In SIB2 (Front) Rack Has 4
Conductors With Degraded Wire Insulation
AR2000-1283; Insulation Broken On RTD Wires
AR2000-1325; Unable To Transmit Emergency Response Data (ERDS) During Quarterly Test
AR2000-1340; DC Switch DCPDPCB03A/10 Will Not Open. (Bus 13 Normal DC Power
AR2000-1341; High Carbon Monoxide Levels Identified In Off Loading Portal (Old Receiving
Bldg.) During Fork Lift Operation (Fork Lift Needs Service)
AR2000-1349; Gas Line Fitting Leaking
AR2000-1374; Containment Sump Level Indication Test Switch Broken
AR2000-1376; Thermocouple Number 25 Connector Broken In Bridge Cable Tray
AR2000-1414; 52/CRSF1B-P Breaker Failed To Close
AR2000-1415; Lead/Lag Unit Found Out of Tolerance
AR2000-1421; Main Steam Safety Valve Position Indication Failures (MSSV)
AR2000-1473; Moisture Separator Reheater Level Switch Wiring Found Melted, Mechanical
Failure
AR2000-1474; Broken Wires Found At Lugs On LAH-2100
AR2000-1483; Emergency Light 1BN-8 Failure
AR2000-1487; V-5088C Sprays Chlorine Into Contained Area When Opened
AR2000-1528; Charging Pump Leakoff Rate At .5 GPM For Charging Pump C
AR2000-1545; Main Steam Loop B Guide MSU-21 Has Bent I-Beam Flange
AR2000-1560; Minimum Charging Flow Acceptance Criteria
AR2000-1563; Wire In MCCS Contains PCBs And Wires Show Signs Of Degradation
AR2000-1575; R-31 Reading All ES
AR2000-1581; HEMYC Wrap Concerns On Cable Tray 111 (Appendix R)
AR2000-1629; R-9 Spiking Hi and Low
5
AR2000-1660; Coolant Leaking From Radiator On TSC Diesel Generator
AR2000-1664; AOV-4238, Condensate Recirc Valve, Repeatedly Alarms
AR2000-1666; RM-14A5 Shows Intermittent Spikes To 2E-6 UCI/CC
AR2000-1669; R-22 Fails PT-17.2
AR2000-1693; Heat Detector S-13 DX Failed To Alarm (North Detector)
AR2000-1694; Fire System S-13 (b Emer Diesel Gen) Approximately 25 Gals Of Water And 1
Cup Of Sand/Grit Drained From Sprinkler Header
AR2000-1702; Fire Suppression System S-13 Will Not Reset
AR2001-0011; Flowswitch Not Able To Be Calibrated
AR2001- 0019; Battery Charger BYCA Erratic Output
AR2001- 0022; Sirens Failed During Silent Test
AR2001- 0025; Siren Failed During Silent Test
AR2001- 0045; CRFC D Did Not Start When Switch Taken To Close (Cnmt Recirc Fan)
AR2001- 0046; A D/G Fuel Oil Transfer Pump Discharge Check Valve 5961 Failed Its Closure
Test
AR2001- 0051; Valve Stem Found Detached From Bonnet During Maintenance, 5961 (A D/G
Fuel Oil Transfer Check Valve)
AR2001- 0073; Plastic Bags Are Duct Taped Over Control Room Chlorine And Ammonia
Transmitters to Deflect Building Leakage
AR2001- 0094; Heat Detectors Failed During Fire System Testing
AR2001- 0194; Power Supplies Out of Tolerance On Channel 3 Nuclear Instruments
AR2001- 0205; Air Ejector Offgas Monitor RM-15A Channel 6 High Failure
AR2001- 0210; Bus 17 Undervoltage Control Cabinet Abnormal Indication
AR2001- 0306; New Storage Reel From Stock For Incore Detector Drive Unit Was Defective
AR2001- 0357; Aux Bldg Crane Failure
AR2001- 0393; Category Two Digital Pressure Indicator Found Out Of Tolerance
AR2001- 0400; Switch Malfunction
AR2001- 0406; Work Order 20101575 Identified Degraded Diesel Fire Pump Condition
AR2001- 0425; Valve 518 Pressurizer Mini Spray Had Packing Leak
AR2001- 0526; SFP Pump A Found Tripped
AR2001- 0612; Heavy Static Heard On Headset When Plugged Into Aux Benchboard Jack
AR2001- 1030; Relief Valve 9204R Flowing Water
AR2001- 1274; Leaking Capacitors On Siren 62
AR2001- 1344; Fuse Clip Loose
AR2001- 1478; Incore Drive A Display/Logic Problem
AR2001- 1570; Defective Neutron Meter ASP-1, NRD
AR2001- 1690; Possible Hole In AOV Diaphragm For V413
AR2001- 1761; Lube Oil Leak
AR2001- 1762; Lube Oil Leak
AR2001- 2029; PT-944 Found Out Of Tolerance
AR2001- 2183; Corrosion On B Battery Cell 29 And 11 Post To Link Connection
AR2001- 2249; R-9 Letdown Indications Read Low Locally
AR2002-0034; Low Voltage On Mux 2 Backup Power Supply
AR2002-0208; Bus 17 Undervoltage Relay 27D/B/17 Inoperable
AR2002-0294; Bast Pump B
AR2002-0590; J-10 And C-3 Fittings Found Damaged On Seal Table
AR2002-0698; Evaluate B RCP Boric Acid Leakage
AR2002-0798; Bus 11A Wattmeter RH
AR2002-0803; MCB 480V Bus Ammeters Not Working
6
AR2002-0809; B RCP Flange Stud Stretch Below Minimum
AR2002-0948; Aux Building Main HEPA Failed Testing
AR2002-1324; Small Leak On SW Piping In Aux Building
AR2002-1387; Bu2 14 Transformer Cooling Fan Reverse Rotation
AR2002-1398; Charging Pumps Exhibiting Increased Leakage
AR2002-1542; Service Water Leakage On Piping For RHR Pump Cooling Fan Cooler A
AR2002-1557; Bushing Cracked On A Charging Pump Plunger
AR2002-1623; R12, CV Gas Increases
AR2002-1627; A Hotwell Cation Conductivity Analyzer Failed Verification
AR2002-1647; RCS Lithium Outside Desired Range
AR2002-1767; Meteorology Tower 13A Wind Speed Not Recording
AR2002-1949; DMIMS Inoperability
AR2002-2035; B Inter Bldg Exh Fan Mounting Loose
AR2002-2055; Water On Top Of B Condensate Storage Tank Diaphragm
AR2002-2057; NIS Channel 2B Drawer Operation Selector Switch Not Working Properly
AR2002-2158; Service Water Leak: Supply To A RHR Fan Coolers
AR2002-2267; Rubber Mounts Are Out Of Their Retaining Rings
AR2002-2400; Calibrate UV Relay 27/17
AR2002-2420; Siren #61 Quit Alerting 60 Seconds Into Its 3 Minute Run.
AR2002-2447; RK-32, 33 Foot Wind Speed Recorder Has Broken String
AR2002-0223; V-300A Has Slight Packing Leak
AR2002-0229; Boric Acid Leak - V-893A Packing
AR2002-0393; Fire Water Booster Tank Filling Unexpectedly
AR2002-0408; Received Loss Of Communications When Testing Security Alarm Status Panels
AR2002-0640; Control Room Ammonia Analyzer Found Low Out Of Tolerance
AR2002-0990; High Rate On TSC Diesel Generator Battery Charger
AR2002-0794; Relay Room North Wall Block Needs Repair
AR2002-2076; Voids and Gaps in North Wall of Intermediate Building
AR2002-2322; Hole in Intermediate Building Wall Needs Repair
AR2003-1386; Classification Differences Between CMIS & FPPR
AR 2003 0310, Boric Acid Residue on C and D Containment Recirculation Fans
ARs 2003-1766, Boric Acid Residue\\Leak - PS102B
ARs 2003-1767, Boric Acid Residue\\Leak - PCH05
ARs 2003-1768, Boric Acid Residue\\Leak - PI-922A
7
LIST OF ACRONYMS
Aging Management Program
Action Request
American Society of Mechanical Engineers
BACI
Boric Acid Corrosion Inspection
BTNK
Buried Piping and Tanks
CATS
Commitment and Action Tracking System
Closed-Cycle Cooling Water
CGIEE
Commercial Grade Items Engineering Evaluation
Corrective Action Program
ECCNS-EQ
Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental
Qualification
FOC
Fuel Oil Chemistry
Fire Protection
Generic Aging Lessons Learned
In-service Inspection Program
kV
Kilo-volt
NRC
Nuclear Regulatory Commission
PSaPM
Periodic Surveillance and Preventive Maintenance
PSPM
Periodic Surveillance and Preventive Maintenance
Poly Vinyl Chloride
Primary Water Stress Corrosion Cracking
Request for Additional Information
Request for Additional Information
Refueling Outage
RGE
Rochester Gas Electric Company
RVH
Nickel-Alloy Nozzles and Penetrations Inspection
Standard Review Plan
Systems, Structures and Components
Service Water System
TTIP
Thimble Tube Inspection Program