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IR 05000263-06-006(DRS); Nuclear Management Company; on 01/23/2006 - 02/22/2006; Monticello Nuclear Generating Plant; License Renewal Inspection Program, Scoping, Screening, and Aging Management Programs
	ML060940194
Person / Time
Site:	Monticello
Issue date:	03/30/2006
From:	Stone A M; NRC/RGN-III/DRS/EB2
To:	Conway J; Nuclear Management Co
References
	FOIA/PA-2010-0209 IR-06-006
	Download: ML060940194 (66)
	v • d • e

March 30, 2006

Mr. J. ConwaySite Vice President Monticello Nuclear Generating Plant Nuclear Management Company, LLC 2807 West County Road 75 Monticello, MN 55362-9637

SUBJECT: MONTICELLO NUCLEAR GENERATING PLANT NRC LICENSE RENEWALSCOPING, SCREENING, AND AGING MANAGEMENT INSPECTION REPORT 05000263/2006006(DRS)

Dear Mr. Conway:

On February 22, 2006, the NRC completed an inspection regarding your application for licenserenewal for the Monticello Nuclear Generating Plant. The enclosed report documents the inspection results, which were discussed on February 22, 2006, with members of your staff in an exit meeting open for public observation at the Monticello Community Center in Monticello, Minnesota.The purpose of this inspection was an examination of activities that support the application for arenewed license for Monticello. The inspection addressed the processes of scoping and screening plant equipment to select equipment subject to an aging management review and development and implementation of aging management programs to support a period of extended operation. As part of the inspection, the NRC examined procedures andrepresentative records, interviewed personnel, and visually examined accessible portions of various systems, structures or components to verify license renewal boundaries and to observeany effects of equipment aging.The inspection concluded that the scoping, screening, and aging management license renewalactivities were generally conducted as described in the license renewal application, as supplemented through your responses to requests for additional information from the NRC. The inspection also concluded that documentation supporting the application is generally in anauditable and retrievable form. Existing aging management programs were determined to be functioning adequately and, when all the programs are implemented as described in your license renewal application, there is reasonable assurance that the intended functions of vitalplant systems, structures, and components will be maintained through the period of extendedoperation.

Conway-2-In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter,its enclosure, and any response you provide will be available electronically for public inspectionin the NRC Public Document Room or from the Publicly Available Records System (PARS)

component of NRC's 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/Ann Marie Stone, ChiefEngineering Branch 2 Division of Reactor SafetyDocket Nos. 50-263 License Nos. DPR-22

Enclosure:

Inspection Report 05000263/2006006(DRS)

w/Attachment:

Supplemental Informationcc w/encl:M. Sellman, Chief Executive Officer and Chief Nuclear Officer Manager, Regulatory Affairs J. Rogoff, Vice President, Counsel, and Secretary Nuclear Asset Manager, Xcel Energy, Inc.

Commissioner, Minnesota Department of Health R. Nelson, President Minnesota Environmental Control Citizens Association (MECCA)

Commissioner, Minnesota Pollution Control Agency D. Gruber, Auditor/Treasurer, Wright County Government Center Commissioner, Minnesota Department of Commerce Manager - Environmental Protection Division Minnesota Attorney General's Office Conway-2-In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter,its enclosure, and any response you provide will be available electronically for public inspectionin the NRC Public Document Room or from the Publicly Available Records System (PARS)

component of NRC's 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/Ann Marie Stone, ChiefEngineering Branch 2 Division of Reactor SafetyDocket Nos. 50-263 License Nos. DPR-22

Enclosure:

Inspection Report 05000263/2006006(DRS)

w/Attachment:

Supplemental Informationcc w/encl:M. Sellman, Chief Executive Officer and Chief Nuclear Officer Manager, Regulatory Affairs J. Rogoff, Vice President, Counsel, and Secretary Nuclear Asset Manager, Xcel Energy, Inc.

Commissioner, Minnesota Department of Health R. Nelson, President Minnesota Environmental Control Citizens Association (MECCA)

Commissioner, Minnesota Pollution Control Agency D. Gruber, Auditor/Treasurer, Wright County Government Center Commissioner, Minnesota Department of Commerce Manager - Environmental Protection Division Minnesota Attorney General's OfficeDOCUMENT NAME: E:\Filenet\ML060940194.wpdTo receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copyOFFICERIIIRIIIRIIINAMEPLougheedBBurgessAMStoneDATE03/28/0603/30/0603/30/06OFFICIAL RECORD COPY C. Conway-3-ADAMS Distribution

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GEG KGO CST CAA1 C. Pederson, DRS DRPIII DRSIII PLB1 JRK1 DXM2 ROPreports@nrc.gov EnclosureU.S. NUCLEAR REGULATORY COMMISSIONREGION IIIDocket Nos:50-263 License Nos:DPR-22 Report No:05000263/2006006 Applicant:Nuclear Management Company, LLC Facility:Monticello Nuclear Generating PlantLocation:Monticello Nuclear Generating PlantNuclear Management Company, LLC 2807 West County Road 75 Monticello, MN 55362-9637Dates:January 23 through February 22, 2006Inspectors:P. Lougheed, Senior Engineering Inspector (lead)Z. Falevits, Senior Engineering Inspector M. Holmberg, Senior Engineering Inspector J. Neurauter, Senior Engineering Inspector D. Merzke, License Renewal Project Manager, NRR C. Acosta-Acevado, Engineering InspectorApproved by:Ann Marie Stone, ChiefEngineering Branch 2 Division of Reactor Safety Enclosure 1

SUMMARY OF FINDINGS

IR 05000263/2006006(DRS); Nuclear Management Company; on 01/23/2006 - 02/22/2006;Monticello Nuclear Generating Plant; License Renewal Inspection Program, Scoping,

Screening, and Aging Management ProgramsThis inspection of the applicant's license renewal aging management review was performed byfive regional office inspectors and one staff member from the NRC's office of Nuclear ReactorRegulation. The team used NRC Manual Chapter 2516 and NRC Inspection Procedure 71002as guidance for performing this inspection. No "findings" as defined in NRC Manual Chapter0612 were identified.The team concluded that, in general, the applicant performed their license renewal scoping,screening, and aging management review in accordance with the Monticello license renewal application. No impediments to the applicant receiving an extended operating license wereidentified.

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REPORT DETAILS

A.Inspection ScopeThis inspection was conducted by NRC Region III inspectors and the license renewalproject manager from the Office of Nuclear Reactor Regulation (NRR). The inspectionwas performed in accordance with NRC Manual Chapter 2516 and NRC InspectionProcedure 71002, "License Renewal Inspection," dated February 18, 2005.This inspection looked at both the applicant's scoping and screening methodology andaging management programs, as described in the license renewal application (LRA),

submitted to the NRC on March 24, 2005.The attachments to this report list the applicant personnel contacted, the documentsreviewed, and the acronyms used.B.Visual Observation of Plant EquipmentDuring this inspection, the inspectors performed walkdown inspections of portions ofmany of the plant systems, structures, and components (SSCs), including some SSCswhich were outside the scope of license renewal (LR). The walkdowns were intended to determine the acceptability of the scoping boundaries, to observe the current conditionof the SSCs, and to assess the likelihood that a proposed aging management program would successfully manage any aging effects. Specific comments on the walkdown results are presented in the sections below. Portions of the following systems werewalked down:*Condensate and Feedwater Systems;*Diesel Generator Starting Air System;

Electrical Power Distribution System;

Emergency Power Distribution System;

Fire Protection System;

Fuel Pool Cooling System;

Heating and Ventilation System;

High Pressure Coolant Injection System;

Instrument and Service Air Systems;

Liquid and Solid Radwaste Systems;

Main Condenser;

Non-Essential Diesel;

Reactor Building Closed Cooling Water System;

Reactor Water Cleanup System; and

Service and Seal Water Systems.

3The following structures were walked down:*Auxiliary Building;*Block Walls;

Buried Cable Manholes;

Electrical Penetrations;

Emergency Diesel Generator Rooms;

Exterior Equipment Foundations;

High Pressure Coolant Injection Pump Room;

Intake Structure Interior Area;

Overhead Cranes;

Non-Segregated Bus Duct;

Pipe Supports;

Plant Power, Control and Instrumentation Cable Installations;

Reactor Core Isolation Cooling Pump Room;

Residual Heat Removal Pump Room;

Torus Room Area;

Turbine Building;

Substation;

4.16 kV Switchgear Rooms; and

985 Pump Room.The inspectors reviewed videotaped examinations of the following reactor vesselinternals conducted by the applicant during the 2003 and 2005 refueling outages:*Core Shroud Welds;*Core Spray Headers and Spargers;

Core Spray Piping "T-Box" Repair Hardware;

Core Spray Piping Welds;

Jet Pump Components;

Shroud Support Welds;

Steam Dryer Support Lug Attachment Welds;

Steam Dryer Welds; and

Surveillance Sample Holder Attachment Welds.C.Review of Scoping and Screening MethodologyIn order to assess the applicant's scoping and screening methodology, the inspectionconcentrated on those non-safety-related systems whose failure could prevent safety-related SSCs from accomplishing a safety function, in accordance with 10 CFR 54.4(a)(2). The inspection specifically focused on the SSCs, or portions of SSCs, which the applicant had determined to be outside the scope of license renewal.

To verify that non-safety-related SSCs were correctly captured within or omitted from the scope of license renewal, the inspectors reviewed LR documents, interviewed personnel, and walked down the selected SSCs.

41.Condensate and FeedwaterThe condensate and feedwater (CFW) system is primarily a non-safety-relatedsystem which supplies condensate from the main condenser to the reactorvessel at an elevated temperature and pressure. The CFW system includes thecondensate demineralizer, the reactor feedwater pump seal, and the zinc injection passivation subsystems. Some components in the CFW system arewithin the scope of LR because they perform a safety-related function, in accordance with 10 CFR 54.4(a)(1). Some components are within scope because their failure could affect the capability of safety-related components toperform their safety function, in accordance with 10 CFR 54.4(a)(2). In addition, some components are within scope because they support station blackout, in accordance with 10 CFR 54.4(a)(3).The inspectors reviewed the LR boundary drawings, the application, and theupdated safety analysis report (USAR), and interviewed personnel responsible for the program. The inspectors also performed system walkdowns of accessible portions of the condensate and feedwater systems. The inspectorsconcluded that the applicant had performed scoping and screening for the condensate and feedwater systems in accordance with the methodologydescribed in the LRA and the rule.2.Diesel Generator Starting AirThe purpose of the diesel starting air system is to provide the motive force toinitially put the diesel engine in motion and begin the diesel cycle. The dieselengine is started by compressed air stored in the starting air system receivertanks. In-scope components are located between the air compressor discharge check valves and the diesel engine air start motors. The air compressors and driers are not in-scope of LR because they are not required to start the diesel engine.The inspectors reviewed the LR boundary drawings, the application, the scopingand screening reports and the USAR, and interviewed personnel responsible for the program. The inspectors also performed system walkdowns of accessible portions of the diesel starting air system. The inspectors concluded that theapplicant had performed scopi ng and screening for the diesel starting air systemin accordance with the methodology described in the LRA and the rule.3.Drywell Atmospheric CoolingThe drywell atmospheric cooling system is a non-safety-related system which theapplicant considered outside the scope of license renewal. The system providesnormal cooling to the drywell to maintain the bulk average drywell ambient temperature less than 135F and localized temperatures below 150F duringnormal plant operation. This assures a sustained life for insulation and gasket materials or sealants inside the drywell. The system does not have any safetyfunction and is not required for any regulated event.

5The inspectors reviewed the USAR, the scoping and screening output report,and the operations manual for the drywell atmospheric system. The inspectorsalso reviewed drawings of the system and held discussions with applicantpersonnel. The inspectors agreed with the applicant's assessment that this non-safety-related system did not meet the requirements to be included within thescope of license renewal.4.Electric Power Distribution SystemThe electric power distribution system provides alternating-current power to bothsafety-related and non-safety-related SSCs in the plant. Components were placed into scope either because they individually had a safety-related function or were required to perform during a regulated event. Additionally, some classes of components were placed in scope on a commodities basis. The majority of the electrical components screened out as active and did not require an aging management program.The inspectors reviewed the related electrical LR boundary drawing, theapplication, and the applicable USAR sections and interviewed personnel responsible for the electrical power distribution system and the LR program. Theinspectors also performed system walkdowns of accessible and partiallyinaccessible portions of the electrical power distribution system.The inspectors identified discrepancies between the scoping and screeningreport and electrical boundary drawing number LR-36298. Specifically, the LR classification of breakers fed from several 480 Volt load centers were not consistent with the LR scoping and screening document. The applicant also identified several other load center cubicles which were not previously shown as being in-scope, but which should have been. The applicant determined that theadditional components brought into scope of LR were all active components and therefore screened out and did not required aging management. Therefore, there were no new aging management concerns.The inspectors concluded that the applicant had performed scoping andscreening for the electrical power distribution system in accordance with themethodology described in the LRA and the rule.4.Fuel Pool Cooling and CleanupThe purpose of the fuel pool cooling and cleanup system is to remove decayheat generated and to maintain pool water purity and clarity. Components in the fuel pool cooling and cl eanup system are non-safety-related and their failurecould affect the capability of safety-related SSCs to perform their safety function;therefore, they are in-scope in accordance with 10 CFR 54.4(a)(2).The inspectors reviewed the LR boundary drawings, the application, and theUSAR, and interviewed personnel responsible for the program. The inspectors also performed system walkdowns of accessible portions of the fuel pool coolingand cleanup system. The inspectors concluded that the applicant had performed 6scoping and screening for the fuel pool cooling and cl eanup system inaccordance with the methodology described in the LRA and the rule.5.Heating and VentilationThe purpose of the heating and ventilation (HTV) system is to provideappropriate ambient environmental conditions for plant safety-related equipment, specifically for the high pressure coolant injection and core spray systems. Another purpose is to provide for controlled flow direction and release of radioactive gases during non-accident conditions. In addition, components of

the system are included in the secondary containment isolation function. Thereare some portions of the HTV system which perform a safety-related function, in accordance with 10 CFR 54.4(a)(1). The HTV system also is credited for use inmitigating the regulated events of fire protection and environmental qualification(EQ).The inspectors reviewed the LR boundary drawings, the application, and theUSAR, and interviewed personnel responsible for the program. The inspectors also performed system walkdowns of accessible portions of the heating andventilation system. As a result of the system walkdowns, the inspectors identified additional piping and valves which should have been included within the scope of LR in accordance with 10 CFR 54.4(a)(2). One portion of piping along with two valves and a steam trap inside the emergency diesel generator room, as well as a section of non-safety-related piping attached to safety-related piping outside the torus were brought into scope. The inspectors concluded that the applicant had performed scoping and screening for the HTV system inaccordance with the methodology described in the LRA and the rule.6.Instrument AirThe instrument air system is a non-safety-relat ed system which provides theplant with a continuous supply of oil-free compressed air. The instrument air portion of the system supplies dried compressed air for most of the pneumaticinstruments and controls in the plant. The instrument air system is not requiredpost-accident since equipment requiring compressed air for operation during or immediately subsequent to an accident receives air from local accumulators or other pneumatic sources. However, the instrument air system interfaces withsafety related systems resulting in some structures and components within thissystem being in-scope for license renewal. Components in the instrument airsystem whose failure could affect the capability of safety related components toperform their safety function, are in-scope for license renewal. In addition, some structures and components are in-scope due to environmental qualification. The remainder of the system was excluded because it was considered not to performa safety-related function, not to potentially impact the function of another safety system, and not provide a function related to one of the regulated events.The inspectors reviewed the LR boundary drawings, the application, the scopingand screening reports and the USAR, and interviewed personnel responsible for the program. The inspectors also performed system walkdowns of accessible 7portions of the instrument and service air system. The inspectors identified thatLR boundary drawings LR-36049-10 at locations D-6 and B-6 showed the lines continuing to drawing LR-36049-12 at location B-6. However, the continuation line on drawing LR-36049-12 at location B-6 LR did not show the lines as being in-scope. The inspectors determined through walkdowns and discussion with the applicant that the lines should be in-scope until a physical boundary wasreached. Based on this clarification, the inspectors concluded that the applicanthad performed scoping and screening for the instrument and service air systemin accordance with the methodology described in the LRA and the rule.7.Liquid and Solid RadwasteThe liquid and solid radwaste systems are non-safety-rela ted systems designedto collect, process, and dispose of radioactive and potentially radioactive wastes in a controlled and safe manner without limiting plant power output or availability. The liquid radwaste system is designed to accommodate the radioactive inputresulting from the design basis maximum fuel leakage condition. The solid radwaste system is also designed to package, store, monitor, and provideshielded storage facilities to allow for radioactive decay and for temporarystorage prior to shipment for off-site disposal. Some components in these systems were in-scope in accordance with 10 CFR 54.4(a)(1) and othercomponents were brought into scope due to supporting the environmental qualification regulated event in accordance with 10 CFR 54.4(a)(3). Additionally, since the failure of some non-safety-related components could affect the capability of safety-related components to perform their safety function, thosecomponents were brought into scope in accordance with 10 CFR 54.4(a)(2).The inspectors reviewed the LR boundary drawings, the application, the scopingand screening output report, and the USAR. The inspectors also performed system walkdowns of accessible portions of the liquid and solid radwastesystems, with an emphasis on the boundaries where the applicant determinedthe systems would no longer be in-scope for license renewal. Additionally, theinspectors interviewed personnel responsible for the systems and familiar withthe LR process. The inspectors determined that, in some cases, the LR boundary drawings did not correctly show the break point where radwaste SSCs transitioned from being in or out of scope. The applicant noted each of these items and wrote license renewal action items to clarify the boundary location.Additionally, the inspectors noted that the application did not correctly describethe basis for concluding that portions of the liquid radwaste system inside thereactor building were out of scope for license renewal. Specifically, the application stated that all radwaste system components existing in either theturbine or reactor buildings, and constituting a liquid pressure boundary, were in-scope. However, the inspectors noted that the portions of the liquid radwastesystem inside a room - designated as the 985 pump room - were considered outof scope in the scoping and screening output report and on the LR boundary drawings. The inspectors determined that the 985 pump room was physically inside the reactor building. Furthermore the room opened to and could only be accessed from the reactor building. Therefore, the inspectors questioned the 8dichotomy between the statement in the application and the boundariesdescribed in the other documents.The inspectors performed a physical walkdown of the 985 pump room and itsvicinity. The inspectors acknowledged that the 985 pump room was reasonably physically separate from the rest of the reactor building, that all systems andcomponents inside the 985 pump room were non-safety-related and that failure of the components within the room would not adversely impact any safety-related components. Therefore, the inspectors agreed that components inside the 985pump room did not need to be in-scope for LR under the provisions of 10 CFR 54(a)(2). The applicant stated that a change would be made to the application wording to explain that components in the 985 pump room were out of scope for license renewal. The applicant planned to include the revision in the annual update to the application. With this revision, the inspectors concluded that the applicant had performed scoping and screening for the liquid and solid radwaste systems in accordance with the methodology described in the LRA andthe rule.8.Main CondenserThe purposes of the main condenser are to provide a heat sink for the steamcycle, to remove non-condensable gases, and to serve as a central collectionpoint for system drains. The system is non-safety-related, but is credited forpost-accident plate out and holdup of radioactive iodine in the loss of coolant accident and control rod drop accident analyses in the USAR. Therefore, the applicant placed portions of the system in scope for license renewal.The inspectors reviewed the LR boundary drawings, the application, and theUSAR, and interviewed personnel responsible for the program. The inspectors also perform ed system walkdowns of accessible portions of the main condensersystem. The inspectors concluded that the applicant had performed scoping andscreening for the main condenser system in accordance with the methodologydescribed in the LRA and the rule.9.Non-Essential DieselThe non-essential diesel is a non-safety-related component which the applicantconsidered outside the scope of license renewal. The diesel provides a source of power for the safety parameter display system, should normal power be lost. Additionally, plant procedures allow the operators to use the diesel during a station blackout. However, the non-essential diesel is not credited in the stationblackout analysis and does not perform any safety-related function.The inspectors reviewed the USAR, the scoping and screening output report, theoperations manual, and abnormal procedures for the non-essential diesel.

Additionally, the station blackout analysis design basis document was reviewed and a walkdown of the area was performed. The inspectors agreed with the applicant's assessment that this non-safety-related system did not meet therequirements to be included within the scope of license renewal.

910.Reactor Water CleanupThe reactor water cleanup system is a filtering and ion exchange system thatmaintains water purity in the reactor and recirculation lines during all modes of plant operation. This non-safety-related system provides for continuouspurification of a portion of the reactor recirculation system flow with a minimum ofheat loss and water loss from the cycle. Portions of the system perform a safetyfunction and are within the scope of LR in accordance with 10 CFR 54(a)(1).

Other components were identified as being required for the regulated events ofanticipated transient without scram and environment qualification and so were in-scope under 10 CFR 54(a)(3). Finally the failure of some non-safety-related components could affect the function of safety-related components, putting some components in-scope under 10 CFR 54(a)(2).The inspectors reviewed the LR boundary drawings, the application, and thescoping and screening output report. The inspectors also perform ed syst emwalkdowns of accessible portions of the reactor water cleanup system, with anemphasis on the boundaries where the applicant determined t he system wouldno longer be in-scope for license renewal. Additionally, the inspectors interviewed personnel responsible for the systems and familiar with the LRprocess. The inspectors determined that portions of the reactor water cleanup system physically located within the 985 pump room were shown in scope. Uponfurther review, the applicant confirmed that those portions were non-safety-related and did not affect the classification of the 985 pump room and components contained therein as being out-of-scope for license renewal. Similar to the description for the radwaste systems, the application did not describe thecomponents within the 985 pump room as being out-of-scope. The application revision described above would correct this issue as well. Therefore, the inspectors concluded that the applicant had performed scoping and screening for the reactor water cleanup system in accordance with the methodology describedin the LRA and the rule.11.Reactor Vessel InternalsThe reactor pressure vessel internals consists of all the structures andcomponents within the reactor vessel that provide support for the core, control rod system support, instrumentation support, steam quality enhancement, andthat direct coolant flow. The portions of the reactor pressure vessel internals containing components subject LR include the core shroud and core plate, top guide, core spray lines and spargers, jet pump assemblies, fuel support and control rod drive housing, and guide tubes. The nuclear fuel is not addressed because the fuel is periodically replaced thereby making it short-lived. Other core internal components which were not considered within scope of 10 CFR 54.4(a)(1)(2) or

(3) included the steam separator assembly, the feedwater spargers, and the surveillance sample holders.The inspectors reviewed the LRA and the USAR, and boiling water reactor(BWR) vessel internals project (VIP) document BWRVIP-06 "BWR Vessel Internals Project Safety Assessment of BWR Internals." Based upon these 10reviews, the inspectors did not identify any discrepancies in the applicantsscoping and screening for the reactor vessel internals. Therefore, the inspectors concluded that the applicant had identified the appropriate core internalscomponents subject to aging management in accordance with the methodology described in the LRA and the rule.12.Service and Seal WaterThe purpose of the service and seal water (SSW) system is to supply screenedand strained raw cooling water from the Mississippi river to various non-essential plant heat loads and services during all modes of operation. The seal water portion of the SSW system provides filtered well water to the shaft seals for various pumps including the service water pumps, residual heat removal service water (RSW) pumps, and the circulating water pumps. The service water portion is used to keep the RSW subsystem filled and pressurized during normal plantoperation, and serves as a backup supply for RSW motor thrust bearing oil coolers. The SSW system also supplies water to the sodium hypochloritesubsystem. The SSW system is normally in service during plant operation andshutdown. The SSW system is not required during or immediately subsequentto a design basis accident and is not safety related.The SSW system lines have the potential for spatial interactions with safetyrelated equipment. Also, portions of service water piping were upgraded to Class I seismic requirements for internal flooding concerns. The SSW system isconnected to the fire water system and supplies water to the fire system jockeypump. SSW system valves provide a pressure boundary to prevent backflowfrom the fire system; therefore these portions of the SSW system were placed in scope for the regulated event of fire protection. The remainder of t he syst emwas excluded because it was considered not to perform a safety-related function, not to potentially impact the function of another safety system, and not provide afunction related to one of the regulated events.The inspectors reviewed the LR boundary drawings, the application, the scopingand screening reports, and the USAR, and interviewed personnel responsible for the program. The inspectors also performed system walkdowns of accessible portions of the SSW system. During review of the LR boundary drawing LR-36665, the inspectors noticed that a continuation of in-scope non-safety-relatedSSW piping at location C-5 occurred going to location A-5 on the same drawing.

However, the continuation at location A-5 was not identified as within scope of license renewal. Following discussion and a plant walkdown, the inspectors determined that the continuation line needed to be shown as in-scope for LR upto the first physical barrier. With this minor issue corrected, the inspectors concluded that the applicant had performed scoping and screening for the SSW system in accordance with the methodology described in the LRA and the rule.D.Review of Aging Management ProgramsThe inspection assessed the adequacy of current implementation of existing agingmanagement programs (AMPs) credited in the applicant's LR program. This included 11verification that current AMPs would ensure that aging effects would be managed sothat there was reasonable assurance that an SSCs intended function would bemaintained throughout the period of extended operation. For those programs indicatedby the applicant as being consistent with NUREG 1801, "Generic Aging LessonsLearned (GALL) Report," the inspectors confirmed that the applicant's program included the GALL attributes. For those programs which the applicant indicated were new or being enhanced, the inspectors confirmed that commitments existed and were sufficient to support future implementation. For those programs where the applicant indicated thatthey intended to take exception to the GALL, the inspectors reviewed the exceptionsagainst the GALL recommendations and evaluated the acceptability of the applicant'sproposal.The inspection also consisted of walkdowns of selected in-scope SSCs to assess howplant equipment was being maintained under the current operating license and to visually observe examples of non-safety-related equipment determined to be in scope due to their proximity to safety-related equipment and their potential for failure due to aging effects.1.10 CFR Part 50, Appendix J (B2.1.01)The 10 CFR Part 50, Appendix J program is an existing program that isconsistent with the recommendations of NUREG-1801,Section XI.S4, "10 CFRPart 50, Appendix J." However, the applicant also identified some exceptions to the GALL program. The 10 CFR Part 50, Appendix J program specifies pneumatic pressure tests and visual examinations to verify the structural and leak tight integrity of the primary containment.An overall (Type A) pressure test assesses the capacity of the containment toretain design basis accident pressure. This test also measures total leakage through the containment pressure-retaining boundary. Local (Type B and C)tests measure leakage through individual penetration isolation barriers. These barriers are maintained as required to keep overall and local leakage under Technical Specification and plant administrative limits.The inspectors reviewed the LR program basis documentation, agingmanagement review documents, existing plant procedures, and recently completed inspection results. The inspectors reviewed the applicant's plant-specific operating experience through a corrective action program search for degraded penetrations which exceeded the administrative leakage limits and verified that the applicant performed adequate historic reviews of plant-specificexperience to determine aging effects.The inspectors concluded that the 10 CFR Part 50, Appendix J programeffectively manages aging effects. Continued implementation of the 10 CFR Part 50, Appendix J program will provide reasonable assurance that the agingeffects will be managed so that the structural components within the scope of theprogram will continue to perform their intended function consistent with thecurrent licensing basis for the period of extended operation.

122.ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD (B2.1.02)The American Society of Mechanical Engineers Boiler and Pressure VesselCode (the ASME Code)Section XI inservice inspection (ISI), subsections IWB, IWC, and IWD program is an existing program that is generally consistent with NUREG 1801,Section XI.M.1, "ASME Section XI Inservice Inspection,Subsections IWB, IWC, and IWD". However a number of exceptions were identified during or following the NRC aging management review audit. Thisprogram provides for condition monitoring of Class 1, 2, and 3 pressure retaining components and their integral attachments.The inspectors reviewed LR program basis documents, program health report,inservice inspection self-assessment report, nondestructive examination records focused on Section XI Code-required reactor vessel nozzle examinations, NRC inspection report findings, and licensee event reports (LERs) associated with pressure retaining Code components. The inspectors also searched the applicant's corrective action program records for degraded Code components to determine plant specific aging effects and to assess the program's effectiveness at detecting and monitoring for age related degradation. Based upon thesereviews, the inspectors did not identify any issues adversely affecting the applicant's AMP. Additionally, the inspectors did not identify any additional exceptions from the Section X1.M.1 program. Therefore, the inspectors concluded that the applicant's AMP should continue to assure the ASME Code pressure boundary function consistent with the current licensing basis for the period of extended operation.The inspectors identified a current plant issue associated with the applicant'sfailure to submit relief requests to the NRC for six reactor vessel nozzleexaminations with limited weld volumetric coverage completed during the 2000 refueling outage. The applicant captured this issue in AR 01013875 and the inspectors turned this issue over to the NRC resident inspector for followup.3.ASME Section XI, ISI Subsection IWF (B2.1.03)The ASME Code Section XI ISI, Subsection IWF program is an existing programthat is consistent with NUREG 1801,Section XI.S.3, "ASME Section XI InserviceInspection, Subsections IWF" with one exception and with enhancements as described in NRC audit and review report for plant aging management reviewsand programs (ML052850461). The program provides for condition monitoring of Class 1, 2, and 3 and MC component supports.The inspectors reviewed LR program basis documents, NRC inspection reportfindings, and LERs associated with ASME Code Class 1, 2, and 3 and MC component supports. The inspectors also searched the applicant's corrective action program records for degraded Code component supports to determine plant specific aging effects and to assess the program's effectiveness at detecting and monitoring for age related degradation. Based upon these reviews, the inspectors identified that the existing program procedures lacked 13specific guidance to confirm the acceptability of baseplate gaps (e.g., gapbetween the metal support base plate and the concrete support structure) for Code component supports and that no exception was taken for these gaps. The applicant wrote action requests (ARs) 00829856-02 and 00830329-02 to incorporate procedural guidance for support plate baseplate gaps. The inspectors concluded that when these requirements were incorporated into existing procedures, the applicant's AMP should have adequately managed current plant aging effects relating to Code Class 1, 2, and 3 and MC support integrity consistent with the current licensing basis for the period of extendedoperation.4.Bolting Integrity (B2.1.04)The bolting integrity program is an existing program which complies with therecommendations of NUREG-1801,Section XI.M18, "Bolting Integrity," withsome exceptions. The program manages the aging affects associated with bolting in the scope of LR through periodic inspection, material selection, thread lubricant control, assembly and torque requirements, and repair and replacement requirements. These activities are based on the applicable requirements of ASME Section XI and plant operating experience and includes consideration of the guidance contained in NUREG-1339, "Resolution of Generic Safety Issue 29: Bolting Degradation or Failure in Nuclear Power Plants," EPRI documents NP-5769, "Degradation and Failure of Bolting in Nuclear power Plants", TR-104213, "Bolted Joint Maintenance and Application Guide," and NP-5067, Volumes 1 and 2, "Good Bolting Practices." The program also credits other aging management programs for inspection of installed bolts; some of these programs require enhancement to include such inspections.The inspectors reviewed LR program basis documentation, aging managementreview documents, and existing plant procedures. The inspectors witnessed a bolt torquing activity and verified that the aging management attributes were employed. Additionally, the inspectors performed a detailed walkdown of the high pressure coolant injection system and the torus exterior to verify theadequacy of structural bolting. Finally, the inspectors performed numerous searches of the corrective action program to determine the acceptability of the applicant's current program. The inspectors concluded that the bolting integrity program effectively manages aging effects. Continued implementation of the bolting integrity program will provide reasonable assurance that the aging effectswill be managed so that bolted components and structures within the scope ofthe program will continue to perform their intended function consistent with thecurrent licensing basis for the period of extended operation.5.Buried Piping and Tanks Inspection (B2.1.05)The buried piping and tanks inspection program is an existing program that,when enhanced, will be comparable to NUREG-1801,Section XI.M34, "BuriedPiping and Tanks Inspection." The program manages the aging effects for buried piping, conduit, and tanks in-scope for license renewal. Preventive 14measures consist of preventive coatings and/or wraps on buried components. Condition monitoring consists of periodic inspections of buried components.The inspectors reviewed the applicable LR program basis documentation,existing inspection procedures, and confirmed that the applicant had commitments in place to enhance the program prior to the start of the period of extended operation. The inspectors also interviewed the buried piping and tanks inspection program owner, reviewed a diesel fuel oil storage tank inspection report, reviewed an underground piping inspection report, and reviewed applicant condition reports and operating experience to verify concerns related to buried piping and tanks are being addressed through the applicant's corrective action program.The inspectors noted that although the program basis documents indicated thatburied conduit would be managed by the buried piping and tanks inspection program, the existing inspections and related procedures were limited to buried piping and the diesel fuel oil storage tank. Since buried conduit is galvanized and not wrapped or coated similar to carbon steel piping or tanks, conduit aging could be different than that for the underground piping and tanks. The applicantcommitted to identify buried conduit as an enhancement to the scope of the buried piping and tanks inspection program to be included in the LRA annual

update.The inspectors concluded that the buried piping and tanks inspection program,when enhanced as described in the application and with the above stated changes, will adequately manage aging effects. Implementation of the buriedpiping and tanks inspection program will provide reasonable assurance that theaging effects will be managed so that in-scope buried components will continueto perform their intended function consistent with the current licensing basis for the period of extended operation.6.Bus Duct Inspection Program (B2.1.06)The bus duct inspection program will be a new program that, when implemented,will be consistent with the ten elements described in Appendix A of NUREG-1800, "Standard Review Plan for Review of LR Applications for Nuclear Power Plants." This AMP will demonstrate, for in scope non-segregated bus duct, thatthe aging effects caused by ingress of moisture or contaminants (dust debris),

insulation degradation and bolt relaxation will be adequately managed to providereasonable assurance that the non-segregated bus ducts will perform theirintended function, consistent with the current licensing basis, during the period ofextended operation.The inspectors reviewed aging management program related documentation,condition reports, preventive maintenance (PM), and LR procedures and activities, and confirmed that the applicant had NRC commitment numberM05019A in place to implement the program prior to the start of the period of extended operation. The inspectors also interviewed applicant engineers concerning the bus duct inspection program to determine how and when aging 15management program changes that are required to satisfy LR commitments andapplicable interim staff guidance (ISGs) will be developed and implemented.In addition, the inspectors conducted field inspections of accessible portions ofthe non-segregated bus duct and identified a number of anomalies concerning performance of past bus duct PM inspections and several material condition items. The applicant initiated AR 01013874, AR 01013360, and generic AR 008298888-02 to address the concerns noted. The applicant also planned to revise the PM procedures as part of the program development. The inspectors also verified that the applicant performed adequate historic reviews of plant specific and industry experience to determine aging effects.The inspectors concluded that the bus duct inspection program, whenimplemented as described, will effectively manage aging effects, since it willincorporate proven monitoring and testing techniques, acceptance criteria, corrective actions, and administrative controls. Implementation of this program will provide reasonable assurance that the effects of aging will be managed suchthat components within the scope of the program will perform their int endedfunctions consistent with the current licensing basis for the period of extended operation.7.BWR Control Rod Drive Return Line Nozzle (B2.1.07)The BWR control rod drive return line nozzle (CRDRL) program is an existingprogram that is consistent with NUREG 1801,Section XI.M.6, "BWR Control RodDrive Return Line Nozzle" with two exceptions. The program provides for condition monitoring of the BWR CRDRL nozzles.The inspectors reviewed LR program basis documents and LERs associatedwith the CRDRL nozzles. The inspectors also searched the applicant'scorrective action program records for issues related to CRDRL nozzle integrity to determine plant specific aging effects and to assess the program's effectiveness at detecting and monitoring for age related degradation. Based upon thesereviews, the inspectors did not identify any issues adversely affecting the applicant's AMP. The inspectors also did not identify any additional exceptions from the Section X1.M.6 program. Therefore, the applicant's AMP should continue to perform its intended function and maintain the integrity of the CRDRLnozzles consistent with the current licensing basis for the period of extendedoperation.8.BWR Feedwater Nozzle (B2.1.08)The BWR feedwater nozzle program is an existing program that is consistentwith NUREG 1801,Section XI.M.5, "BWR Feedwater Nozzle" with one exceptionand enhancement as described in NRC audit and review report for plant agingmanagement reviews and programs (ML052850461). The applicant's program provides for condition monitoring of the BWR feedwater nozzles.

16The inspectors reviewed LR program basis documents and LERs associatedwith feedwater nozzles. The inspectors also searched the applicant's corrective action program records for issues related to feedwater nozzle integrity to determine plant specific aging effects and to assess the program's effectiveness at detecting and monitoring for age related degradation. Based upon thesereviews, the inspectors did not identify any issues adversely affecting the applicant's AMP. The inspectors also did not identify any additional exceptions from the Section X1.M.5 program. Therefore, the applicant's AMP should continue to perform its intended function and maintain the integrity of the feedwater nozzles consistent with the current licensing basis for the period ofextended operation.9.BWR Penetrations (B2.1.09)The BWR Penetrations Program is an existing program that is consistent withNUREG 1801,Section XI.M.8, "BWR Penetrations" with three exceptions asdescribed in NRC audit and review report for plant aging management reviewsand programs (ML052850461). This program provides for condition monitoring of the reactor vessel penetrations.The inspectors reviewed LR program basis documents and LERs associatedwith BWR reactor vessel penetrations. The inspectors also searched the applicant's corrective action program records for issues related to vessel penetration integrity to determine plant specific aging effects and to assess the program's effectiveness at detecting and monitoring for age related degradation.

Based upon these reviews, the inspectors did not identify any issues adversely affecting the applicant's AMP nor did the inspectors identify any additional exceptions from the Section X1.M.8 program. Therefore, the applicant's AMP should continue to perform its intended function and maintain the integrity of the feedwater nozzles consistent with the current licensing basis for the period ofextended operation.10.BWR Stress Corrosion Cracking (B2.1.10)The BWR penetrations program is an existing program that is consistent withNUREG 1801,Section XI.M.7, "BWR Stress Corrosion Cracking" with twoexceptions as described in NRC audit and review report for plant agingmanagement reviews and programs (ML052850461). This program provides for condition monitoring of pressure boundary material susceptible to stress corrosion cracking.The inspectors reviewed LR program basis documents and LERs associatedwith pressure boundary components affected by stress corrosion cracking. The inspectors also searched the applicant's corrective action program records for issues related to stress corrosion cracking to determine plant specific aging effects and to assess the program's effectiveness at detecting and monitoring for age related degradation. Based upon these reviews, the inspectors did not identify any issues adversely affecting the applicant's AMP. The inspectors also did not identify any additional exceptions from the Section X1.M.7 program.

17Therefore, the applicant's AMP should continue to perform its intended functionand maintain the integrity of components susceptible to stress corrosion cracking consistent with the current licensing basis for the period of extended operation.11.BWR Vessel Inner Diameter Attachment Welds (B2.1.11)The BWR inner diameter (ID) attachment welds program is an existing programthat is consistent with NUREG 1801,Section XI.M.4, "BWR ID AttachmentWelds" with two exceptions as described in NRC audit and review report for plantaging management reviews and programs (ML052850461). The program provides for condition monitoring of the reactor vessel interior attachment welds within and beyond the beltline region.The inspectors reviewed LR program basis documents and videotaped visualexaminations of reactor vessel ID attachment welds. The inspectors also searched the applicant's corrective action program records for issues related to vessel ID attachment welds to determine plant specific aging effects and to assess the program's effectiveness at detecting and monitoring for age related degradation. Based upon these reviews, the inspectors did not identify anyissues adversely affecting the applicant's AMP. The inspectors also did not identify any additional exceptions from the Section X1.M.4 program. Therefore, the applicant's AMP should continue to perform its intended function and maintain the integrity of the vessel ID attachment welds consistent with the current licensing basis for the period of extended operation.12.BWR Vessel Internals (B2.1.12)The BWR vessel internals program is an existing program that is consistent withNUREG 1801,Section XI.M.9, "BWR Vessel Internals" with one exception andone enhancement as described in NRC audit and review report for plant agingmanagement reviews and programs (ML052850461). The program provides for condition monitoring of the reactor vessel internals components for crack initiation and growth.The inspectors reviewed LR program basis documents, videotaped visualexaminations of reactor vessel internals components and LERs associated with vessel internal components. The inspectors also searched the applicant's corrective action program records for issues related to vessel internal components to determine plant specific aging effects and to assess the program's effectiveness at detecting and monitoring for age related degradation.

Based upon these reviews, the inspectors identified that additional changes were necessary to ensure that the applicant adequately managed plant aging effectsrelating to reactor vessel internals in accordance with the AMP and the GALL.a)The inspector noted that in-core monitoring instrument dry tubes werewithin the scope of the applicant's vessels internals AMP. However, these tubes were not subject to periodic inspections under the applicant's AMP which credited the BWRVIP-130 BWR water chemistry guidelines and the ASME Section XI inspection programs. Because these tubes are 18subject to radiation induced damage above the threshold for irradiationassisted stress corrosion cracking (IASCC), they could crack and cause pressure boundary leakage. GALL Section X1.M.9 identified that cracking has been observed at other BWRs. Furthermore, General Electric (GE) service information letter, SIL 409, "Incore Dry Tube Cracks," recommended periodic (every other outage) visual examinationsfocused on the upper two feet of the tube to detect cracking. The applicant had voluntarily implemented these examinations for the currentlicense; however, the applicant had not committed to continue with theseexaminations during the period of extended operation within their AMP. The applicant stated that the GE SIL 409 recommended examinations ofin-core monitoring instrument dry tubes would be incorporated into their AMP during the next annual LRA revision.b)The steam dryer was within the scope of the applicant's vessels internalsAMP for its structural function and the applicant conducted periodic inspection of steam dryer welds potentially subject to cracking. The applicant submitted the LRA prior to the issuance of the BWRVIP inspection program guidance defined by BWRVIP-139 "Steam Dryer Inspection and Flaw Evaluation Guidelines." The applicant stated that the BWRVIP-139 steam dryer weld examinations would be incorporated into the AMP during the next annual LRA revision.c)In CAP 014359 (condition report 20000209), the applicant documentedthat during the 2000 refueling outage, areas of the steam dryer in closeproximity to the main steam nozzles appeared polished and that this wear could be caused by steam impingement. The applicant also documented in AR 000032000032that Vermont Yankee, with an identical steam dryer design, had observed evidence of steam erosion at the underside of the steam dryer. To evaluate if steam erosion was occurring, a degradation mechanism not identified for the steam dryer in GALL Section XI.M9, the applicant stated that the affected areas of the Monticello steam dryerwould be reinspected during the next refueling outage. The inspectors also forwarded this issue to the NRR technical staff for further evaluation.d)In two letters dated May 30, 1997, and October 30, 1997, from thechairman of the BWRVIP to the NRC, the BWRVIP committed member utilities (including Monticello) to implementation of the BWRVIP guidelines (applicable to vessel internals) to the extent possible. The applicant stated that a direct commitment to the NRC for implementation of the BWRVIP internals guidelines (similar to this existing third party commitment) would be included during the next annual LRA revision.e)In LRA Table B1.6-11 "Responses to BWRVIP-74-A for the MNGP[Monticello Nuclear Generating Plant], Table 4-1", the applicant stated that the internal core spray piping welds P1, 2, and 3 were not inspected in accordance with BWRVIP-18 "Core Spray Internals Inspection and Flaw Evaluation Guidelines" because mechanical clamps were installed to insure the structural integrity of the sparger T-box welds and that a 19visual inspection was conducted each outage to confirm that T-boxintegrity was maintained. Specifically, the applicant performed a general visual examination (VT-3) of the mechanical clamp type repair hardware installed around the welds instead of an enhanced visual examination (EVT-1) of the welds.BWRVIP-18 did not require examination of repaired core spray pipewelds unless the integrity of the repair depended upon these welds. The inspectors noted that if cracks develop in the non-inspected core spraypiping welds P1, 2, and 3, a cooling water flow diversion path would exist outside the core shroud which could adversely affect the applicants peak fuel clad temperature (PCT) analysis. Because the applicant's PCT analysis relied, to some extent, on the leakage integrity for these repaired welds, the inspectors determined that these welds should be inspected using EVT-1 methods to meet BWRVIP-18 requirements. Therefore, the inspectors concluded that the applicant had deviated from the BWRVIP-18 guidance and that this deviation should be identified as an exception from Section XI.M.9 of the GALL. The applicant stated that the LRA would be changed to remove statements about not inspecting these welds during the next annual LRA revision and that the applicable inspection procedures would be changed to implement enhanced visual examinations of these welds.Based on the above enhancements being implemented, the inspectorsconcluded that the applicant's AMP should perform its intended function to maintain the integrity of reactor vessel internals components consistent with the current licensing basis for the period of extended operation.13.Closed-Cycle Cooling Water System (B2.1.13)The closed-cycle cooling water (CCCW) system program is an existing programwhich will be comparable to NUREG-1801,Section XI.M21, "Closed-CycleCooling Water System." However, the applicant also identified some exceptions to the GALL program. The CCCW system surveillance program manages agingeffects in closed cycle cooling water systems that are not subject to significantsources of contamination, in which water chemistry is controlled and heat is not directly rejected to the ultimate heat sink. The program includes:

(1) preventive measures to minimize corrosion; and

(2) periodic system and componentperformance testing and inspection to monitor the effects of corrosion and confirm intended functions are met.The inspectors reviewed LR program basis documentation, aging managementreview documents, historical chemistry parameter trends, corrective action documents, and existing procedures and surveillance. The inspectors alsointerviewed the CCCW program owner, interviewed the reactor building closed cooling water system engineer, and conducted walkdowns to assess thecondition of CCCW systems within the plant. The inspectors verified that theapplicant performed adequate historic reviews of plant specific experience to 20determine aging effects and the exceptions to NUREG 1801 specified in the LRAare consistent with current industry practice.The inspectors concluded that the CCCW system program effectively m anagesaging effects. Continued implementation of the CCCW system program willprovide reasonable assurance that the aging effects will be managed so that theCCCW system components will continue to perform their intended functionconsistent with the current licensing basis for the period of extended operation.14.Compressed Air Monitoring (B2.1.14)The compressed air monitoring program is an existing program that, withenhancements, will be consistent, with exceptions, to NUREG-1801, SectionXI.M24, "Compressed Air Monitoring." The MNGP compressed air monitoring program consists of inspection, monitoring, and testing of the Instrument and service air systems to provide reasonable assurance that they will perform theirintended function for the duration of extended operation.The inspectors reviewed program documentation, aging management reviewdocuments, historical chemistry parameter trends, corrective actions documents, and existing procedures. The inspectors verified that the applicant performed adequate historic reviews of plant specific experience to determine aging effects and that the exceptions to the GALL, specified in the LRA, were consistent withcurrent industry practice.The inspectors concluded that the compressed air monitoring program effectivelymanages aging effects. Continued implementation of the compressed air monitoring program will provide reasonable assurance that the aging effects willbe managed so that the primar y and secondary system components will continueto perform their intended function consistent with the current licensing basis for the period of extended operation.15.Cables and Connections Not Subject to Environmental QualificationRequirements (B2.1.15)The electrical cables and connectors not subject to 10 CFR 50.49 EQrequirements is a new program that, when implemented, will be consistent withthe program described in NUREG-1801,Section XI.E1, "Electrical Cables andConnections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements," and guidance provided in the applicable ISGs. The electrical cables and connectors not subject to 10 CFR 50.49 EQ requirements program

will apply to accessible insulated cables and connections installed in structureswithin the scope of LR and prone to adverse localized environments.The inspectors reviewed aging management program related documentation,condition reports, self assessments, proposed and existing procedures, and confirmed that the applicant had NRC commitment number M05027A in place toimplement the program prior to the start of the period of extended operation. The inspectors also interviewed the applicant to determine how and when the 21aging management program changes that are required to satisfy LRcommitments and applicable ISGs will be developed and implemented. Inaddition, the inspectors conducted visual field inspections of electrical equipment, components and cables and verified that the applicant performed adequate historic reviews of plant specific and industry experience to determine aging effects.The inspectors concluded that the electrical cables and connectors not subject to10 CFR 50.49 EQ requirements program, when implemented as described, should effectively manage aging effects, as it will incorporate proven monitoringand testing techniques, acceptance criteria, corrective actions, and administrative controls. Implementation of this program will provide reasonableassurance that the intended functions of insulated cables and connections exposed to adverse localized equipment environments caused by heat, radiation, or moisture will be maintained and that the effects of aging will be managed suchthat components within the scope of the program will perform their int endedfunctions consistent with the current licensing basis for the period of extended operation.16.Instrument Cables Not Subject to EQ Requirements (B2.1.16)The electrical cables not subject to 10 CFR 50.49 EQ requirements used ininstrumentation circuits program is a new program that the applicant will implement prior to the period of extended operation. This program will beconsistent with the program described in NUREG-1801,Section XI.E2,"Electrical Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits," with some exceptions. The electrical cables included in the scope of this program are cables which are not required to be environmentally qualified under 10 CFR 50.49; are used in radiation monitoring and nuclear instrumentation circuits with sensitive, low-level signals that are within scope of LR and are installed in adverse localized environments caused by heat, radiation and moisture in the presence of oxygen.The inspectors reviewed aging management program related documentation,condition reports, self assessments, proposed and existing procedures, and confirmed that the applicant had NRC commitment number M05028A in place toimplement the program prior to the start of the period of extended operation. The inspectors also interviewed the applicant to determine required enhancements and their implementation schedule to periodically test sensitive instrumentation circuits to ensure that the circuit will perform its intended functionthrough the period of extended operation. The inspectors also verified that the applicant performed adequate historic reviews of plant specific and industry experience to determine aging effects and that the exceptions to the GALLspecified in the LRA are consistent with current industry practice and are acceptable.The inspectors concluded that the electrical cables not subject to 10 CFR 50.49EQ requirements used in instrumentation circuits program, when implemented as described, will effectively manage aging effects, since it will incorporate 22proven monitoring techniques, acceptance criteria, corrective actions, andadministrative controls. Implementation of the electrical cables not subject to 10 CFR 50.49 EQ requirements used in instrumentation circuits program willprovide reasonable assurance that the effects of aging will be managed suchthat components within the scope of the program will perform their int endedfunctions consistent with the current licensing basis for the period of extended operation.17.Fire Protection (B2.1.17)The fire protection (FP) program is an existing program which, with the proposedenhancements, will be comparable to Section XI.M26, "Fire Protection," of theGALL report and as clarified by ISG-04. However, the applicant also identified some exceptions to the GALL program.The fire protection program is credited for detecting and managing age relateddegradation of FP system components and structures. The FP program includesfire barrier visual inspections, motor and diesel-driven fire pump tests and inspections. Periodic testing and inspection of the diesel driven fire pump is performed to ensure that an adequate flow of fire water is supplied and that there is no degradation of the diesel fuel supply lines. Fire barrier inspections will be performed, consisting of periodic visual inspection of fire barrier penetration seals, fire dampers, fire barrier walls, ceilings and floors; andperiodic visual inspection and functional tests of fire-rated doors to ensure that their operability is maintained. The FP program also includes periodic inspectionand testing of the Halon fire suppression system.The inspectors reviewed fire protection aging management program relateddocumentation, condition reports, self assessments, procedures, required enhancements, commitments and implementing documents and confirmed that the applicant had NRC commitments M05029A and M05030A in place toimplement the program prior to the start of the period of extended operation. The inspectors interviewed applicant engineers to confirm the continuation of the existing program along with the implementation schedule of the required program enhancements. In addition, the inspectors verified that the applicant performed adequate historic reviews of plant specific and industry experience to determine aging effects and the exceptions to NUREG 1801 specified in the LRAare consistent with current industry practice. The inspectors conducted field walkdowns of the fire protection system and identified one case of extensivecorrosion on conductor termination lugs for the diesel driven fire pump. The applicant initiated AR 01012302 to evaluate this issue.The inspectors concluded that the fire protecti on system program, in general,effectively manages aging effects. With the enhancements to be incorporated prior to the period of extended operation, continued implementation of the fireprotection system program will provide reasonable assurance that the agingeffects will be managed so that the fire protection system components willcontinue to perform their intended function consistent with the current licensingbasis for the period of extended operation.

2318.Fire Water System (B2.1.18)The fire water system program is an existing program which, with the proposedenhancements, will be comparable to the program described in Section XI.M27,"Fire Water System," of the GALL report, and as clarified by ISG-04.The fire water system program consists of water-based fire protection systemsthat include components that are periodically inspected and tested in accordancewith the applicable National Fire Protection Association codes and standards and plant procedures. These activities include sprinkler system inspections, pipe wallthickness testing, hydrant inspections, fire main flushes, and flow tests.The inspectors reviewed fire water system aging management program relateddocumentation, condition reports, existing procedures, required enhancements and implementing documents and confirmed that the applicant had NRCcommitments M05031A, M05032A, and M05033A in place to implement the program prior to the start of the period of extended operation. The inspectors also conducted plant visual inspections to assess the condition of fire water system equipment, interviewed the fire water system engineer to confirm thecontinuation of the existing program along with the implementation schedule ofthe required program enhancements. In addition, the inspectors verified that the applicant performed adequate historic reviews of plant specific and industry experience to determine aging effects and the exceptions to the GALL specified in the LRA are consistent with current industry practice.The inspectors concluded that the fire water system program, in general,effectively manages aging effects. With the enhancements to be incorporated prior to the period of extended operation, continued implementation of the firewater system program will provide reasonable assurance that the aging effectswill be managed so that the fire water system components will conti nue toperform their intended function consistent with the current licensing basis for the period of extended operation.19.Flow-Accelerated Corrosion (B2.1.19)The flow-accelerated corrosion (FAC) aging management program is an existingprogram credited in the LRA as being consistent with NUREG-1801, SectionXI.M17, "Flow-Accelerated Corrosion." The ongoing program is used to predict, detect, and monitor FAC in plant piping and other components, such as valve bodies, elbows, and expanders. The program was credited with:

(1) conducting an analysis to determine critical locations;

(2) performing baseline inspections to determine the extent of thinning at these locations; and

(3) performing follow-up inspections to confirm the predictions, or repairing or replacing components as necessary.The inspectors reviewed the applicable LR program basis documentation,interviewed the FAC program owner, reviewed applicable procedures, reviewed the determination of systems susceptible to FAC, reviewed engineeringevaluations of localized pipe wall thinning, reviewed FAC self-assessment and 24program health reports, reviewed a condition report and an external operatingreport related to FAC, and reviewed the applicant's commitment to revise the corporate fleet procedure FP-FE-FAC-01 to include 87.5 percent nominal pipe wall thickness for non-safety-related piping as a trigger for engineering analysis before the start of the period of extended operation.The inspectors identified current plant engineering calculations for safety-relatedpiping designed in accordance with the ASME Code that used an alternative method to evaluate localized pipe wall thinning instead of ASME Code Case 597-1. Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," lists Code case 597-1 as aconditionally acceptable code case. The applicant committed to evaluate the 2003 alternative methodology, address the use of Code Case 597-1 in future calculations, and review the relationship between the FAC program and ASME Section XI requirements when evaluating wall thinning in AR 01013831.The inspectors concluded that the FAC program was in place, had beenimplemented, was an ongoing program subject to NRC review, and generallyincluded the elements identified in the LRA. As it is a current program subject to periodic NRC review and inspection, there is reasonable assurance thatadequate inspections required by the program will be performed thr ough theperiod of extended operation.20.Fuel Oil Chemistry (B2.1.20)The fuel oil chemistry program is an existing program that, with enhancements,will be consistent, with exceptions, to NUREG-1801,Section XI.M30, "Fuel OilChemistry." The fuel oil program mitigates and manages aging effects on the internal surfaces of diesel fuel oil storage tanks and associated components in systems that contain diesel fuel oil. The program includes: a) surveillance and monitoring procedures for maintaining diesel fuel oil quality by controllingcontaminants in accordance with applicable ASTM Standards; b) periodic draining of water from diesel fuel oil tanks, if water is present; c) periodic or conditional visual inspection of internal surfaces or wall thickness measurements from external surfaces of diesel fuel oil tanks; and d) one-time inspections of a representative sample of components in systems that contain diesel fuel oil.The inspectors reviewed LR program basis documentation, aging managementreview documents, and existing procedures. The inspectors also interviewed the program owner and conducted walkdowns of the emergency diesel generators, day tanks, and enclosures for the fuel oil storage tanks. The inspectors confirmed that the applicant had commitments in place to enhance the programprior to the start of the period of extended operation. The inspectors verified that the applicant performed adequate historic reviews of plant specific experience to determine aging effects. The inspectors confirmed that the exception to the GALL to not require addition of biocides, stabilizers, and corrosion inhibitors wasacceptable based on plant history and on the continuing requirement to sample the fuel oil and evaluate abnormal test results.

25The inspectors concluded that the fuel oil chemistry program effectivelymanages aging effects. Providing the enhancements are incorporated as specified by the applicant's application, continued implementation of the fuelprogram will provide reasonable assurance that the aging effects will bemanaged so that the fuel oil system components will continue to perform theirintended function consistent with the current licensing basis for the period ofextended operation.21.Inaccessible Medium Voltage Cables Not Subject to EQ (B2.1.21)The inaccessible medium-voltage cables not subject to 10 CFR 50.49requirements program is a new program that the applicant will implement prior tothe period of extended operation. The program, when implemented will beconsistent with that described in NUREG-1801,Section XI.E3, "InaccessibleMedium-Voltage Cables Not Subject to 10 CFR 50.49 Requirements." This program applies to inaccessible (e.g., in conduits, cable trenches, cable troughs, duct banks, underground vaults, or direct-buried) medium-voltage cables within the scope of LR that are exposed to significant moisture simultaneously with applied voltage. This program includes a commitment to periodically inspect and test these cables to provide an indication of the condition of the conductor insulation. The specific type of tests performed will be determi ned andimplemented prior to the expiration of the current license.The inspectors reviewed aging management program related documentation,condition reports, existing procedures, and conducted field inspections of a number of manholes in the plant and in the substation and confirmed that the applicant had NRC commitment M05037A in place to implement the programprior to the start of the period of extended operation. The inspectors also interviewed applicant engineers to determine, in general, how the program would be enhanced to include manhole inspections and additional "state of the art" cable tests. In addition, the inspectors verified that the applicant performed adequate historic reviews of plant specific and industry experience to determine aging effects.The inspectors concluded that the inaccessible medium-voltage cables notsubject to 10 CFR 50.49 requirements program, when implemented as described in the required enhancements, will effectively manage aging effects, since it willincorporate periodic inspections and "state of the art" testing techniques.

Implementation of this program will provide reasonable assurance that theeffects of aging will be managed such that components within the scope of theprogram will perform their intended functions consistent with the current licensingbasis for the period of extended operation.22.Inspection of Overhead Heavy and Light Load Handling Systems (B2.1.22)The inspection of overhead heavy load and light load (related to refueling)handling systems program is an existing program which, when enhanced, will becomparable to NUREG-1801,Section XI.M23, "Inspection of Overhead HeavyLoad and Light Load (Related to Refueling) Handling Systems." However, the 26applicant also identified an exception to the GALL program. The purpose ofoverhead heavy load and light load (related to refueling) handling systemsprogram is to identify component aging effects prior to loss of intended function.

The program manages aging effects of the structural components for the cranes, heavy loads, rigging (reactor handling equipment) systems, and structures. Crane rails and structural components are visually inspected for indications of degradation, such as corrosion, wear, or cracks.The inspectors reviewed the applicable LR program basis documentation,existing overhead crane inspection procedures, and confirmed that the applicant had commitments in place to enhance the program to specify a five-year inspection frequency for the fuel preparation machines prior to the start of the period of extended operation. The inspectors also interviewed the crane program owner and maintenance personnel that perform overhead crane structural inspections, reviewed documentation associated with "engineered lifts", and reviewed condition reports to verify identified crane structural concerns are being addressed through the applicant's corrective action program.The inspectors reviewed the applicant's exception to a statement in SectionXI.M23 of the GALL report. The applicant took exception to tracking the number and magnitude of lifts made by the crane, as the current program does not track the number and magnitude of lifts within the crane's rated capacity. The inspectors determined the applicant does track the number and magnitude of "engineered lifts" that exceed the crane's rated capacity, and these lifts are controlled by applicant procedures. Therefore, the inspectors concluded that this exception was acceptable.The inspectors concluded that the inspection of overhead heavy load and lightload handling systems program effectively manages aging effects. With theenhancement to be incorporated prior to the period of extended operation, continued implementation of the inspection of overhead heavy load and light load (related to refueling) handling systems program will provide reasonableassurance that the aging effects will be managed so that the monitored structuralcomponents will continue to perform their intended function consistent with thecurrent licensing basis for the period of extended operation.23.One-Time Inspection Program (B2.1.23)The one-time inspection program is a new program that will be generallyconsistent with NUREG-1801,Section XI.M32, "One-Time Inspection." Theprogram will include measures to verify the effectiveness of the plant and fuel oilchemistry aging management programs. Additionally, the program will confirmthe absence of age degradation in selected components such as flow restrictors, venturis, and small bore piping that are within the scope of license renewal. The program is to be implemented prior to the period of extended operation.The inspectors reviewed the applicant's LR documentation and the one-timeinspection program sampling method implementation procedure. The inspectors also discussed the planned scope and methodology for the program with 27applicant engineers and confirmed that the applicant had an existingcommitment to implement the program prior to the period of license renewal.

The inspectors concluded that, provided the program was implemented in accordance with the commitment and planned methodology, the program should provide reasonable assurance that the effects of aging would be assessed and managed such that components within the scope of the program will performtheir intended functions consistent with the current licensing basis for the periodof extended operation24.Open-Cycle Cooling Water System (B2.1.24)The open-cycle cooling water system program is an existing program which iscomparable to NUREG-1801,Section XI.M20, "Open-Cycle Cooling WaterSystem." T he open-cycle cooling water system program to ensure that theeffects of aging on the raw water service water systems will be managed for theperiod of extended operation. This program manages the aging effects of metallic components in water systems exposed to raw, untreated water. Theseaging effects are due to corrosion, erosion, and biofouling in systems, structuresand components serviced by the OCCW system. The program includes: a)surveillance and control of biofouling; b) tests to verify heat transfer; and c)routine inspection and maintenance.The inspectors reviewed program documentation, aging management reviewdocuments, corrective actions documents, and existing procedures and surveillance. The inspectors also interviewed the program owners, interviewed the service water system engineer and conducted walkdowns to assess thecondition of the service water system. The inspectors verified that the applicantperformed adequate historic reviews of plant specific experience to determine aging effects.During an OCCW system walkdown, the inspectors identified floor drain pipesthat did not appear on the LR boundary drawings and that were within the scopeof license renewal. These drains fell within the wells and domestic water system scoping and screening program. The applicant committed to revise the LRA and supporting documents to accommodate these drain lines as being within scope of license renewal.The inspectors concluded that the OCCW system program effectively managesaging effects. Continued implementation of the OCCW system program willprovide reasonable assurance that the aging effects will be managed so that theservice water system components will continue to perform their intended functionconsistent with the current licensing basis for the period of extended operation.25.Plant Chemistry Program (B2.1.25)The water chemistry program is an existing program that is consistent withNUREG-1801,Section XI.M2, "Water Chemistry." However, the applicant alsoidentified some exceptions to the GALL program. The plant chemistry program mitigates the aging effects on component surfaces that are exposed to water as 28the process fluid; chemistry programs are used to control water chemistry forimpurities (e.g., chloride and sulfate) that accelerate corrosion or crack initiation and growth and that cause heat transfer degradation due to fouling in select heat exchangers.The inspectors reviewed program documentation, aging management reviewdocuments, historical chemistry parameter trends, corrective actions documents, and existing procedures. The inspectors verified that the applicant performed adequate historic reviews of plant specific experience to determine aging effects and that the exceptions to NUREG 1801 specified in the LRA are consistent withcurrent industry practice.The inspectors concluded that the water chemistry program effectively managesaging effects. Continued implementation of the water chemistry program willprovide reasonable assurance that the aging effects will be managed so that theprimary and secondary system components will continue to perform theirintended function consistent with the current licensing basis for the period ofextended operation.26.Primary Containment Inservice Inspection Program (B2.1.26)The primary containment ISI program is an existing program which is consistentwith NUREG-1801,Section XI.S1, "ASME Section XI, Subsection IWE." Theprimary containment ISI program requires visual examinations of the accessible surfaces (base metal and welds) of the drywell, torus, vent lines, internal vent system, penetration assemblies and associated integral attachments. Theprogram also requires examination of pressure retaining bolting and the drywell interior slab moisture barrier. The primary containment ISI program manages the aging effects of corrosion, cracking, mechanical damage, discoloration and other phenomena that can potentially impact structural and/or leak tight integrity. Most examination surfaces are coated; therefore, examinations focus on coatingconditions that indicate the possible deterioration of the underlying metal.The inspectors reviewed LR program basis documentation, aging managementreview documents, existing plant procedures, and recently completed inspection results. The inspectors verified that the applicant performed adequate historic reviews of plant specific and industry experience to determine aging effects.The inspectors concluded that the primary containment ISI program effectivelymanages aging effects. Continued implementation of the primary containment ISI program will provide reasonable assurance that the aging effects will bemanaged so that the structural components within the scope of the program willcontinue to perform their intended function consistent with the current licensingbasis for the period of extended operation.27.Protective Coating Monitoring and Maintenance Program (B2.1.27)The protective coating monitoring and maintenance program is an existingprogram which, when enhanced, will be consistent with the recommendations of 29NUREG-1801,Section XI.S8, "Protective Coating Monitoring and MaintenanceProgram." The program provides for inspections of the drywell and torus interior coated surfaces. The torus inspections include both above- and below-water line surface inspections. The protective coating monitoring and maintenance program is not relied upon to manage loss of material due to corrosion of carbon steel but, rather, is primarily relied upon to ensure that degradation of protective coatings inside containment won't lead to clogging of the emergency core cooling system (ECCS) suction strainers.The protective coating monitoring and maintenance program requiresinspections to be performed each operating cycle for the torus above-water lineand drywell coated surfaces. The program further states that below-water line painted surfaces, including the areas near the ECCS suction strainers, are inspected at intervals not exceeding five years and that a VT-3 exam is performed for all coating degradation identified. All unacceptable areas are required to be repaired or evaluated.During a review of the applicant's procedures 0135, "Pressure-SuppressionChamber Painted Surface Internal Inspection," and 0140, "Drywell Interior Surface Inspection," the inspectors identified that, for the previous two performances of the inspections, the applicant removed the step which required VT-3 examination of unacceptable areas because the examination was not a Code requirement. The applicant issued AR 00830109-03 to revise the procedures to ensure the requirement to perform VT-3 examinations for coatings is implemented to meet LR aging management program criteria.In addition, the inspectors reviewed procedure 1367, "Pressure-SuppressionChamber Below Water Line Painted Surface Internal Inspection." The purpose section of the procedure stated that inspections were to be conducted at internals not to exceed five years. However, in another section, the frequency is stated as "approximately every

(5) years." The inspectors noted a six-year interval between the previous two inspections. The applicant issued AR 00830109-02 to revise procedure 1367 to reflect the maximum surveillancefrequency of five years in accordance with the aging management program.The inspectors reviewed LR program basis documentation, aging managementreview documents, existing plant procedures, and recently completed inspection results. The inspectors verified that the applicant performed adequate historic reviews of plant specific and industry experience to determine aging effects.The inspectors concluded that the protective coating monitoring andmaintenance program effectively manages aging effects. Continued implementation of the protective coating monitoring and maintenance programwill provide reasonable assurance that the aging effects will be managed so thatthe structural components within the scope of the program will conti nue toperform their intended function consistent with the current licensing basis for the period of extended operation.

3028.Reactor Head Closure Studs (B2.1.28)The reactor head closure studs aging management program is an existingprogram credited in the LRA which, when enhanced, will be comparable toNUREG-1801,Section XI.M3, "Reactor Head Closure Studs." The scope of theongoing program is part of the ASME Section XI inservice inspection program and provides condition monitoring of the reactor head closure studs, nuts, threads in flange, closure washers, and bushings. Preventative measures to mitigate cracking have been taken in accordance with Regulatory Guide (RG)1.65, "Materials and Inspections for Reactor Vessel Closure Studs."The inspectors reviewed the applicable LR program basis documentation,interviewed the reactor head closure studs program owner, inspected a sample of reports for ultrasonic examination of the reactor head closure studs, reviewed the applicant's evaluation of external operating experience related to reactor head closure stud cracking, and reviewed applicable program procedures.The inspectors noted that RG 1.65 recommended that reactor head closure studmaterial measured ultimate tensile strength not exceed 170 ksi [kilo-pounds per square inch] to minimize the likelihood of stress corrosion cracking. Hardness tests conducted on the installed reactor head studs showed that most studs to have greater than 170 ksi tensile stress. The applicant committed to document this exception to NUREG-1801 in the LRA. The exception will state that theaging management program does not incorporate the tensile strengthrequirement of RG 1.65 for existing reactor head closure studs. The inspectors concluded that this exception to NUREG-1801 was acceptable because theapplicant considers these studs to be susceptible to cracking, continues to manage the studs using the other preventative measures of RG 1.65, continuesto conduct ultrasonic testing and surface examinations on a ten year interval, and to date no apparent discontinuities have been identified.The inspectors concluded that inspection reactor head closure studs was part ofthe ISI - ASME Section XI program, the program had been implemented, was an ongoing program subject to NRC review, and included the elements identified inthe LRA. As it is a currently required program subject to periodic NRC reviewand inspection, there is reasonable assurance that adequate inspections required by ASME and the NRC will be performed through the period ofextended operation.29.Reactor Vessel Surveillance (B2.1.29)The reactor vessel surveillance program is an existing program which is part ofthe BWRVIP integrated surveillance program (ISP) that uses data from BWR member surveillance programs to select the "best" representative vessel materials to monitor radiation embrittlement for a particular plant. The applicant identified that the reactor vessel surveillance program would be enhanced to beconsistent with the recommendations of NUREG-1801, Chapter XI, ProgramXI.M.31 "Reactor Vessel Surveillance." The AMP enhancement was the applicant's commitment to follow BWRVIP-116 "BWR Vessel and Internals 31Project, Integrated Surveillance Program Implementation for License Renewal"after NRC approval. Because the NRC review of BWRVIP-116 was ongoing, theinspectors did not assess the program for compliance with this enhancement. In

NRC audit and review report for plant aging management reviews and programs(ML052850461), the NRC stated that this program would be reviewed by theNRR staff and issues addressed within Section 3 of the SE for the applicant'sLRA.The inspectors reviewed LR program basis documents associated with theapplicants reactor vessel surveillance program and the calculation which demonstrated that the vessel would maintain adequate toughness (equivalentmargins for upper shelf impact energy) to meet 10 CFR Part 50, Appendix G, requirements for the planned 60 year service life. The inspectors noted that the equivalent margins calculation supported the information for the limiting vessel materials identified in Table 4.2.1-1 "Equivalent Margin Analysis for MNGP Plate Material," of the LRA. The inspectors also searched the applicant's corrective action program records for issues related to reactor vessel surveillance to assess the program's effectiveness at detecting and monitoring for age related degradation (e.g., vessel embrittlement). Based upon these reviews, the inspectors did not identify any issues adversely affecting the applicant's AMP.

Therefore, the reactor vessel surveillance program should continue to perform its intended function consistent with the current licensing basis for the period ofextended operation.The inspectors identified a current plant issue associated with the applicant'splanned change to their vessel surveillance capsule withdraw schedule. InBWRVIP-86-A "BWR Vessel and Internals Project Updated BWR Integrated Surveillance Program Implementation Plan," the ISP schedule defined that asurveillance capsule specimen was due to be removed from Monticello in 2006.

The inspectors noted that the next Monticello surveillance capsule sample was scheduled to be withdrawn during the spring 2007 Monticello refueling outage.

This change in schedule was accepted by the BWRVIP, but had not been submitted and approved by the NRC as required by paragraph III.B.3 of 10 CFRPart 50, Appendix H. The applicant intended to work with the BWRVIP to address this change in schedule and to ensure that the schedule change was approved by the NRC.30.Selective Leaching of Materials (B2.1.30)The selective leaching of materials inspection activities is a new program which,when implemented, will be consistent, with exceptions, to Section XI.M33,"Selective Leaching of Materials," of the GALL report. The program will consistof one-time visual inspection and hardness measurement of selected components that are susceptible to selective leaching. In situations wherehardness testing is not practical, a qualitative method by non-destructive examination or other metallurgical methods will be used to determine thepresence and extent of selective leaching.

32The inspectors reviewed the LR evaluation, interviewed the selective leaching ofmaterials program owner, and reviewed the applicant's commitment to develop and implement a selective leaching of materials inspection program before the start of the period of extended operation. The inspectors concluded that, if the procedures are implemented as planned, there should be reasonable assurance that aging effects will be managed so that components susceptible to selectiveleaching of materials will continue to perform their intended function consistentwith the current licensing basis for the period of extended operation.31.Structures Monitoring Program (B2.1.31)The structural monitoring program is an existing program which, whenenhanced, will be comparable to NUREG-1801, Sections XI.S5, "Masonry WallProgram," XI.S6, "Structures Monitoring Program," and XI.S7, "RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants."

The structural monitoring program manages aging effects to ensure that structures and components within its scope retain the ability to perform theirintended function and is implemented through visual examination of the structures and components. The program is implemented as part of the structures monitoring performed under the provisions of the Maintenance Rule, 10 CFR 50.65, with additional inspections of the intake structure and diesel fuel oil transfer house. The program is based on guidance contained in RG 1.160, "Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," and NUMARC 93-01, "Industry Guidelines for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." The program, when enhanced, willinclude the monitoring and inspection parameters for structural components within the scope of license renewal.The inspectors reviewed the applicable LR program basis documentation,existing structural monitoring inspection procedures, and confirmed that the applicant had commitments in place to enhance the program prior to the start of the period of extended operation. The inspectors also interviewed the structural monitoring program owner, performed walkdowns of selected structures, reviewed a maintenance rule program health status report, a periodic assessment report, a structural monitoring inspection report, and various deficiency reports, condition reports, and external operating experience to verify identified structural concerns are being evaluated and corrected if necessary.The inspectors concluded that the structural monitoring inspection programeffectively manages aging effects. When enhanced as described in the application, continued implementation of the structural monitoring inspectionprogram will provide reasonable assurance that the aging effects will bemanaged so that the monitored components will continue to perform theirintended function consistent with the current licensing basis for the period ofextended operation.

3332.System Condition Monitoring Program (B2.1.32)The system monitoring program is an existing plant-specific program whichmanages aging effects for normally accessible, external surfaces of piping, tanks, and other components and equipment within the scope of license renewal.

These aging effects are managed through visual inspection and monitoring of external surfaces for leakage and evidence of material degradation. The program relies upon periodic system walkdowns to monitor degradation of theprotective paint or coating, and/or the exterior steel surface area (if no paint orcoatings exist, or if the existing protective paint and coatings are degraded to a point whereby the exterior steel surface is exposed). Although no credit is taken for any coating or paint, inspections of the above-ground coating or paint willprovide an indication of the condition of the material underneath the coating or paint. The program will monitor for loose, worn or missing parts, fluid leaks, bolting or fastener degradation, and evidence of corrosion and sealant degradation. The minimum walkdown frequency is once per year for those systems and components that are accessible during normal operation. Currentwalkdowns are conducted every four months. The inspection frequency may be increased based on the safety significance, production significance, or operating experience of each system. Systems and components that are only accessibleduring plant outages are inspected at least once per refueling interval. The applicant planned to enhance the curr ent system walkdown procedure todescribe specific age degradation parameters to be monitored and inspected.

Acceptance criteria will also be incl uded.The inspectors reviewed LR program basis documentation, aging managementreview documents, and the existing system engineering walkdown procedure. The inspectors also accompanied the reactor building closed cooling water system engineer on a routine system walkdown. The inspectors verified that theapplicant performed adequate historic reviews of plant specific experience to determine aging effects.The inspectors concluded that t he system monitoring program adequatelymanaged current plant aging effects. With the enhancements to be incorporated prior to the period of extended operation, continued implementation of thesystem monitoring program will provide reasonable assurance that the agingeffects will be managed so that the plant components will continue to performtheir intended function consistent with the current licensing basis for the period ofextended operation.33.Thermal Aging and Neutron Irradiation Embrittlement of Cast AusteniticStainless Steel (B2.1.33)The thermal aging and neutron irradiation embrittlement of cast austeniticstainless steel (CASS) program is an existing program that is consistent with NUREG 1801,Section XI.M.13, "Thermal Aging and Neutron IrradiationEmbrittlement of Cast Austenitic Stainless Steel" with one exception as described in NRC audit and review report for plant aging management reviewsand programs (ML052850461). The program provides for condition monitoring of the CASS components subject to thermal or irradiation induced embrittlement.

34The inspectors reviewed LR program basis documents and searched theapplicant's corrective action program records for issues related to CASS components to assess the program's effectiveness at detecting and monitoring for degraded CASS components. The applicant concluded that the CASS components (e.g., fuel support pieces) within the core which met the screening criteria for potentially being susceptible to neutron irradiation induced embrittlement did not require augmented inspections due to compressive loadings. The inspectors noted that this position was consistent with NUREG1801 Section XI.M.13 recommendations.The inspectors did not identify any issues adversely affecting the applicant'sAMP. The inspectors also did not identify any additional exceptions from the Section X1.M.13 program. Therefore, the applicant's AMP should continue to perform its intended function and maintain the integrity of CASS components consistent with the current licensing basis for the period of extended operation.34.Environmental Qualification (B3.1)The electrical equipment qualification program is an existing ongoing programwhich manages component thermal, radiation, and cyclical aging through the useof aging evaluations based on 10 CFR 50.49(f) qualification methods. As required by 10 CFR 50.49, EQ components not qualified for the current licenseterm are to be refurbished or replaced, or have their qualification extended prior to reaching the aging limits established in the evaluation. Aging evaluations forEQ components that specify a qualification of at least 40 years are considered time-limited aging analyses (TLAA) for license renewal. The EQ program ensures that these EQ components are maintained within the bounds of theirqualification bases. The program is consistent with NUREG-1801,Section X.E1,"Environmental Qualification of Electric Components."The inspectors reviewed aging management program related documentation,condition reports, self assessments and existing procedures to confirm that the applicant has been successful in effectively managing aging effects of EQ electric components. The inspectors also interviewed the EQ program owners to confirm that the applicant will continue to carry out the EQ program for theduration of the extended operation. The inspectors also verified that theapplicant performed adequate historic reviews of plant specific and industry experience to determine aging effects.The inspectors concluded that the applicant's existing EQ program has beeneffective in managing aging effects. The program has been subject to periodic internal and external assessments that facilitate continuous improvement. Withcontinued implementation and effective management, the EQ program willprovide reasonable assurance that the aging effects will be managed so that theenvironmentally qualified plant components will continue to perform theirintended function consistent with the current licensing basis for the period ofextended operation.

3535.Metal Fatigue of the Reactor Coolant Pressure Boundary (B3.2)The fatigue monitoring program is an existing program which, when enhanced,will be comparable to Section X.M1, "Metal Fatigue of the Reactor CoolantPressure Boundary," of the GALL report. The fatigue monitoring program is a confirmatory program that monitors loading cycles due to thermal and pressuretransients and cumulative fatigue usage for selected reactor coolant and other component locations. Metal fatigue analyses is considered a TLAA under license renewal. The program provides an analytical basis for confirming that the actual number of cycles does not exceed the number of cycles used in thedesign analysis and that the cumulative usage is maintained below the allowable limit, or that appropriate corrective actions are taken to maintain component cumulative fatigue usage below the allowable limit during the period of extended operation. Enhancements include the incorporation of fatigue sensitive locations for older vintage GE plants identified in NUREG/CR-6260, "Application ofNUREG/CR-5999 Interim Fatigue Curves of Selected Nuclear Power PlantComponents," into the fatigue monitoring program implementing procedures.The inspectors reviewed the applicable LR program basis documentation,interviewed the fatigue monitoring program owner, reviewed applicable procedures, confirmed commitments are in place to include the identified enhancement into implementing procedures, reviewed a thermal fatigue program health status report, reviewed a sample of fatigue monitoring analytical calculations to confirm that the evaluations included the period of extendedoperation and the effects of reactor water environment if applicable, and reviewed applicant corrective actions of external operating experience related to the fatigue monitoring program.The inspectors concluded that the fatigue monitoring program effectivelymanages aging effects. When enhanced as described, continued implementation of this program will provide reasonable assurance that the agingeffects will be managed so that the monitored components will conti nue toperform their intended function consistent with the current licensing basis for the period of extended operation.E.Exit Meeting SummaryThe results of this inspection were discussed on February 22, 2005, with Mr. R. Jacobs,and other members of the Nuclear Management Corporation staff in an exit meeting open for public observation at the Monticello Community Center in Monticello, Minnesota. The applicant acknowledged the inspection results and presented no dissenting comments. The slides used during this meeting are provided in the supplemental information.The inspectors noted that proprietary documents were reviewed during the course of theinspection. The applicant confirmed that all such proprietary documents were returned or the copies destroyed and that the likely content of the report would not involve theproprietary material.ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Applicant

M. Aleksey, License Renewal TLAA Lead

R. Baumer, Licensing

P. Burke, Manager of Projects

R. Dennis, License Renewal Civil Lead

J. Grubb, Engineering Director

R. Jacobs, Site Director for Operations

J. Pairitz, License Renewal Project Manager and Mechanical Lead

J. Rootes, License Renewal Programs Lead

B. Sawatzke, Plant Manager

R. Siepel, License Renewal Electrical LeadNuclear Regulatory Commission

A. Stone, Chief, Engineering Branch 2

B. Burgess, Chief, DRP Branch 4

S. Thomas, Senior Resident Inspector

R. Orlikowski, Resident Inspector

2

LIST OF DOCUMENTS REVIEWED

The following is a list of applicant documents reviewed during the inspection, includingdocuments prepared by others for the applicant.

Inclusion of a document on this list does not imply that NRC inspectors reviewed the entire documents, but, rather that selected sections or portions of the documents were evaluated as part of the overall inspection effort.

In addition, inclusion of a document on this list does not imply NRC acceptance of the document, unlessspecifically stated in the body of the inspection report.LICENSE RENEWAL DOCUMENTS

License Renewal Application*Application for Renewed Operating License; March 2005

License Renewal Action Items (Written as a Result of the Inspection)*00829856-02; Add Base Plate Gap Inspection Requirement and Acceptance Criteria toImplementing Procedure

PEI-02.05.02; dated February 3, 2006*00830392-02; Add Base Plate Gap Inspection Requirement and Acceptance Criteria toSystem Walkdown Guidelines; dated February 3, 2006*00829888-02; Sub Task to Track Detailed Inspection Period Requirements for BusDuct; dated February 3, 2006*00830109-02; Revise Procedure 1367 to Specify 5-Year Surveillance Frequency; datedFebruary 9, 2006*00830109-03; Ensure

VT-3 Exams Remain in Protective Coatings Procedures; datedFebruary 9, 2006*00874822-02; Evaluate Non-Safety Components < 87.5 percent T

nom Attributable toFlow Accelerated Corrosion; dated February 8, 2006License Renewal Aging Management Program Basis Documents*PBD/AMP-001; Plant Chemistry Program; Revision 1*PBD/AMP-002; Flow Accelerated Corrosion Program; Revision 2

PBD/AMP-004; Thermal Aging and Neutron Irradiation Embrittlement of Cast AusteniticStainless Steel Program; Revision 1*PBD/AMP-006; Bolting Integrity; Revision 2

PBD/AMP-007; Open Cycle Cooling Water System Program; Revision 1

PBD/AMP-008; Closed Cycle Cooing Water Program; Revision 1

PBD/AMP-010; Inspection of Overhead Heavy Load and Light Load (Related toRefueling) Handling Systems; Revision 1*PBD/AMP-011; Compressed Air Monitoring Program; Revision 1

PBD/AMP-013; Fire Protection; Revision 1

PBD/AMP-014; Fire Water System; Revision 1

PBD/AMP-017; Fuel Oil Chemistry; Revision 1

PBD/AMP-018; Reactor Vessel Surveillance; Revision 1*PBD/AMP-019; One-Time Inspection; Revision 1

PBD/AMP-020; Selective Leaching of Materials; Revision 1

PBD/AMP-021; Buried Piping and Tanks Inspection; Revision 1

PBD/AMP-022; Primary Containment In-Service Inspection Program; Revision 1

3*PBD/AMP-024; ASME Subsection XI, Subsection IWF; Revision 1*PBD/AMP-025; 10

CFR 50, Appendix J Program; Revision 2

PBD/AMP-027; Structures Monitoring Program; Revision 1

PBD/AMP-029; Protective Coating Monitoring and Maintenance Program; Revision 1

PBD/AMP-030; Electrical Cables and Connectors Not Subject to 10

CFR 50.49Environmental Qualification (EQ) Requirements; Revision 1*PBD/AMP-031; Electrical Cables Not Subject to 10

CFR 50.49 EQ Requirements Usedin Instrumentation Circuits; Revision 1*PBD/AMP-031; Boiling Water Reactor (BWR) Penetrations; Revision 1*PBD/AMP-032; Inaccessible Medium Voltage (2kV to 34.5kV) Cables Not Subject to 10CFR 50.49 EQ Requirements; Revision 1*PBD/AMP-033; ASME Section XI Inservice Inspection, Subsections IWB, IWC, andIWD; dated February 18, 2005*PBD/AMP-034; Reactor Head Closure Studs; Revision 1

PBD/AMP-035; BWR Inside Diameter Attachment Welds; Revision 1

PBD/AMP-036; BWR Feedwater Nozzle; Revision 1

PBD/AMP-037; BWR Control Rod Drive Return Nozzle; Revision 1

PBD/AMP-038; BWR Stress Corrosion Cracking; Revision 2

PBD/AMP-040; BWR Vessel Internals; Revision 1

PBD/AMP-042; Electrical Equipment Subject to 10

CFR 50.49 EQ Requirements;Revision 1*PBD/AMP-043; Metal Fatigue of the Reactor Coolant Pressure Boundary; Revision 1

PBD/AMP-044; System Condition Monitoring Program; Revision 0

PBD/AMP-045; Bus Duct Inspection Program; Revision 2License Renewal Aging Management Review Reports*AMR-AIR; Instrument and Service Air System; Revision 1*AMR-CABLES; Non-EQ Cables and Connectors Commodity Group; Revision 3

AMR-CDR; Main Condenser System; Revision 1*AMR-CFW; Condensate and Feedwater System; Revision 1

AMR-DGN; Emergency Diesel Generator System; Revision 1

AMR-FPC; Fuel Pool Cooling and Cleanup System; Revision 0

AMR-FUSEHOLD; Fuse Holders Commodity Group; Revision 1

AMR-HTV; Heating and Ventilation System; Revision 3

AMR-RAD; Liquid and Solid Radwaste System; Revision 1

AMR-REC; Reactor Recirculation System; Revision 2

AMR-RWC; Reactor Water Cleanup System; Revision 1

AMR-SLC; Standby Liquid Control System; Revision 2

AMR-SSW; Service and Seal Water System; Revision 1License Renewal Drawings*LR-119259; Zinc Injection Passivation System; Revision 0*LR-155483-1; Wells and Domestic Water; Revision 0

LR-36033; Main Steam; Revision 1

LR-36034; Turbine and Extraction Steam, Sheet 1; Revision 2

LR-36035; Turbine and Extraction Steam; Revision 1

LR-36035-2; Steam Jet Air Ejectors; Revision 1

LR-36036; Condensate and Feedwater; Revision 1

4*LR-36037; Condensate and Feedwater, Revision 1*LR-36038; Condensate Demineralizer System; Revision 0

LR-36038-2; Condensate Demineralizer System; Revision 0

LR-36039; Condensate and Demineralized Water Storage Systems; Revision 2

LR-36041; Service Water System, Sheet 1; Revision 1

LR-36041-2; Service Water System Sheet 2; Revision 1

LR-36042; Reactor Building Cooling Water System; Revision 1

LR-36043; Radwaste Sump System Open (Dirty); Revision 1

LR-36044; Radwaste Sump System Closed (Clean); Revision 1

LR-36045; Clean Radwaste System; Revision 1

LR-36046; Dirty Radwaste System; Revision 1

LR-36047-1; Radwaste Solids Handling System; Revision 1

LR-36048; Fire Protection System Interior Locations; Revision 1

LR-36049-1; Compressed Air System; Revision 0

LR-36049-4; Service Air System; Revision 1

LR-36049-10; Alternate Nitrogen Supply System; Revision 1

LR-36049-12; Reactor Building and Drywell Instrument Air; Revision 1

LR-36049-13; Instrument Air-Reactor Building; Revision 1

LR-36049-14; Instrument Air-Reactor Building; Revision 1

LR-36051; Diesel Oil System, Sheet 1; Revision 1

LR-36051-1; Diesel Fuel Oil System, Sheet 2; Revision 1

LR-36241; Nuclear Boiler System Steam Supply; Revision 1

LR-36242; Vessel Instrumentation Nuclear Boiler System; Revision 1

LR-36254; Reactor Water Cleanup System; Revision 1

LR-36256; Fuel Pool Cooling and Cleanup System; Revision 2

LR-36259; Auxiliary and Heating Steam System, Turbine Building; Revision 1*LR-36259-1; Auxiliary and Heating Steam System, Miscellaneous Areas; Revision 1*LR-36260; Auxiliary and Heating Steam System, Reactor and Radwaste Buildings;Revision 1*LR-36261; Chilled Water Piping and Miscellaneous, Reactor and Radwaste Building;Revision 0*LR-36263; Air Flow Diagram, Turbine Building; Revision 1

LR-36266; Air Flow Diagram, Radwaste Area; Revision 1

LR-36267; Plant Air Flow Diagram, Sheet 1; Revision 1

LR-36298; Electrical Load Flow One Line Diagram; Revision 2

LR-36348; Auxiliary Heating System Heating Boiler; Revision 1*LR-36516; Fire Protection System Yard Areas; Revision 1

LR-36664; Residual Heat Removal Service Water and Emergency Service WaterSystems; Revision 2*LR-36665; Service Water System and Makeup Intake Structure; Revision 1

LR-36665-2; Service Water System and Makeup Intake Structure; Revision 1

LR-36666; Screen Wash, Fire and Chlorination System Intake Structure; Revision 1

LR-36776; Control Diagram, Intake and Discharge Structures; Revision 1

LR-36807; Airflow Diagram Reactor Building Lower Part; Revision 1

LR-46162; Primary Containment Nitrogen Control System; Revision 1License Renewal Miscellaneous Documents*GE-NE-0000-0020-0279-01; Time-Limited Aging Analyses-Reactor Vessel andInternals; dated April 2004

5*LRPP 1-5; License Renewal Implementation Plan; Revision DRAFT*LRPP 1-3; Operating Experience Data Collection; dated June 29, 2004

Meeting Notes; License Renewal One Time Inspection Expert Panel; datedDecember 8, 2005*One Time Inspection Program Sampling Method; dated February 9, 2006License Renewal Scoping and Screening Reports*SSR-AIR; Instrument and Service Air System; Revision 1*SSR-CDR; Main Condenser; Revision 1*SSR-CFW; Condensate and Feedwater System; Revision 1

SSR-DAC; Drywell Atmospheric Cooling System; Revision 0

SSR-DGN; Emergency Diesel Generators; Revision 4

SSR-FPC; Fuel Pool Cooling and Cleanup System; Revision 2

SSR-HTV; Heating and Ventilation System; Revision 2

SSR-INT; Reactor Pressure Vessel Internals; Revision 1

SSR-NDG; Non-Essential Diesel Generator System; Revision 1

SSR-RAD; Radwaste Solid and Liquid System; Revision 1

SSR-RWC; Reactor Water Cleanup System; Revision 1

SSR-SSW; Service and Seal Water System; Revision 1License Renewal Operating Experience Review Output Reports*OE-CDR; Main Condenser System; Revision 0*OE-CFW; Condensate and Feedwater System; Revision 0

OE-FPC; Fuel Pool Cooling and Cleanup System; Revision 0

OE-HTV; Heating and Ventilation System; Revision 0License Renewal Technical Reports*TR-003; Component Identification and Data Processing for Systems, Structures andComponents (SSC) Within Scope of 10

CFR 54.4(a)(3) for Environmental Qualification;

Revision 1*TR-004; Component Identification for SSC within Scope of 10

CFR 54.4(a)(3) for FireProtection Program; Revision 2CURRENT PLANT DOCUMENTS

Calculations*CA-01-116; Documentation of Thermal Cycles, including Addendum 1; Revision 0*CA-03-163; Determination of Remaining Allowable Operating Time for TW34-10"-HE(Code); Revision 0*CA-03-170; Evaluation of TW24-10"/12"-HE in Corroded Condition; Revision 0

CA-04-142; Environmental Fatigue Calculation; Revision 0

CA-04-143; Environmental Fatigue Calculations for NUREG/CR-6260 Locations;Revision 0*CA-04-166; 12 Emergency Diesel Generator Essential Service Water HX PerformanceTest-Summer 2005; Revision 0*CA-96-086; 1996 Core Shroud Evaluation Analysis; Revision 7

6Corrective Action Requests (AR) Initiated As a Result of the Inspection*01011589; Cracks in High Pressure Coolant Injection Pedestal Grout Cap; datedJanuary 24, 2006*01012112; Degraded Components Not Expeditiously Repaired or Replaced; datedJanuary 27, 2006*01012190; Questionable Note in Station Blackout Procedure; dated January 27, 2006*01012302; Division 1 Emergency Diesel Generator C-91 Panel Wire BundleDiscoloration; dated January 28, 2006*01012303; #12 Emergency Diesel Generator Maintenance Concerns; datedJanuary 28, 2006*01012304; 1R Metal Enclosed Bus Duct Cover Bent; dated January 28, 2006

01013177; Procedure Change Request Past Due

01013360; Frequency of 1R Bus Inspection Does Not Meet Equipment Excellence-WasLast Performed in 1994; dated February 3, 2006*01013447; Swallows Nest on 1R Transformer and Duct Support; datedJanuary 25, 2006*01013635; Unreliable 1R Trouble Annunciator (Hot Oil Portion); dated February 6, 2006

01013672; Noted Previous Leakage from Wall by Torus 896'; dated February 6, 2006

01013829; Swallows Nest on 2R Transformer and Duct Support; datedJanuary 25, 2006*01013831; Calculation Method for Wall Thinning on Residual Heat Removal Piping in2003; dated February 7, 2006*01013858; Small Diesel Fuel Oil Leak From Diesel Fire Pump Engine; datedFebruary 7, 2006*01013874; Rust on Cathodic Protection Junction Box; dated February 7, 2006

01013875; Six Limited In-Service Inspection Exams not Included in Cycle 19 ReliefRequests; dated February 7, 2006*01013930; Seal Water Pressure Gauges Swinging +/-10 Psi; dated February 7, 2006

01014163; Intake Structure Ceiling Penetrations Material Condition; datedFebruary 9, 2006Corrective Action Documents Reviewed During the Inspection*000032; Self Assessment from Vermont Yankee for Steam Dryer; dated May 4, 2004*000137; Apparent Cause Evaluation:

Crack Indication on Control Rod Drive WithdrawalLine Found in Drywell*000173; Condition Evaluation:

Flow Accelerated Corrosion Program Fleet ProcedureIssue; dated March 28, 2005*000298; Root Cause Evaluation:

Possible Cracked Control Rod Drive Withdrawal Linein Drywell*000512; Corrective Action:

Flow Accelerated Corrosion Program Fleet Procedure Issue;dated May 20, 2005*000532; Flow Accelerated Corrosion Program Fleet Procedure Issue; datedMarch 24, 2005*001952; Apparent Cause Evaluation:

Emergency Diesel Mission Time and Criteria forInterpreting Lube Oil Trending; dated April 28, 2004*003752; Apparent Cause Evaluation:

Adverse Trend for Local Leak Rate Test Failuresfor Flowserve Parallel Double Disk Gate Valves; dated May 1, 2005

7*004267; Apparent Cause Evaluation:

Drywell Floor Drain Isolation Valve Failed to CloseDuring Test; dated September 30, 2004*007464; Ops Manual Implies That Diesel Generator 13 Required per NUREG 0737;dated April 27, 2004*007828; Reactor Building Crane Main Hoist Components Do Not Meet RecommendedFactors of Safety; dated April 29, 2004*010385; Emergency Diesel Mission Time and Criteria for Interpreting Lube Oil Trending;dated August 1, 2002*010841; Degraded Components Found During Underwater Inspections at IntakeStructure; dated November 11, 2002*012010; Dryer-Separator Storage Pool Leak Detection System Indicates Pool Leak;dated March 11, 2005*013061; Reactor Building Crane Main Hoist Components Do Not Meet RecommendedFactors of Safety; dated February 10, 2003*013780; Several Pipe Support Baseplates; dated February 24, 2003

014359; In-Vessel Inspections Found Indications on Jet Pump; dated January 14, 2000

018127;

IN 2003-20:

Derating Whiting Cranes Purchased Before 1980; datedOctober 29, 2003*019081;

OE 17615- Excessive Heatup and Cooldown of RPV; dated January 22, 2004

019655;

IN 2004-05:

Spent Fuel Pool Leakage to Onsite Groundwater; datedMarch 9, 2004*021183;

OE 17615-Excessive Heatup and Cooldown of RPV; dated February 18, 2004

021357; Self Assessment NDE Program; dated January 19, 2005

025290; Through Wall Corrosion on Floor of Turbine Lube Oil Storage Tank T-42B;dated September 28, 2005*026489; Documentation of in Core Dry Tubes Examinations; dated May 1, 2003

026496; License Renewal Aging Effects; dated May 2, 2003

026759; Cracks Found in Lug Attachment Welds to Hanger; dated May 2, 2003*026787; Cracks Found in Lug Attachment Welds to Hanger; dated May 7, 2003*026976; Leakage Identified on LPRM 36-29 During Vessel Hydro; dated May 18, 2003

027067; DAEC Indications in the Steam Dryer; dated April 30, 2003

027102; Weld Repair for Indication Identified; dated April 29, 2003

027162; Core Spray Ultrasonic Testing of Slip Joint Welds; dated May 10, 2003

027445; Cracked Tack Weld on Jet Pump; dated May 28, 2003

027639; Steam Erosion Found on Piping at Outlet of

CV-1243; dated May 16, 2003*027905; Piping andInstrumentation Drawing Contains Errors; dated June 19, 2003*028223; Dissimilar Metal Welds on Control Rod Drive Scram Header; datedJuly 15, 2003*029744; Inconsistent Bolting on #11 Core Spray Sparger; dated September 25, 2003*029949;

IN 2003-20:

Derating Whiting Cranes Purchased Before 1980; dated October 29, 2003*030859;

OE-17271 Stress Corrosion Cracking in Closed L; dated November 25, 2003

031740;

OE 17615-Excessive Heatup and Cooldown of RPV; dated January 22, 2004

031967; Identified Hot Spots on Y10-09, Y20-28, and Y-8; dated February 12, 2004*032556;

IN 2004-05:

Spent Fuel Pool Leakage to Onsite Groundwater; datedMarch 9, 2004*033788; Identified a Hot Spot on X60/X80TR-ADISC During Routine ThermographyInspection; dated June 23, 2004*034444; Identified Hot Spot in D10 (11 Bay Charger) During Routine ThermographyInspection; dated August 12, 2004

8*034456; Potential Adverse Trend in Deteriorating Control Rod Drive Instrumentation;dated August 13, 2004*034891; High Pressure Coolant Injection Isometric Drawing; dated September 20, 2004*034915; Support

FWH-83; dated September 21, 2004

035273; Bolts on EFT Charcoal Filter Access Doors are Deteriorating; datedOctober 15, 2004*036850; Most Limiting Core Component not Used for Determining Crack GrowthMitigation; dated January 27, 2005*037309; Instrument Air Filter Analysis From Test 1362 is Out of Spec; datedFebruary 25, 2005*037461; Under-Vessel Leakage Inspection Identifies Four Control Rod Drive FlangeLeaks; dated March 5, 2005*037477; Local Leak Rate Test Failure of

MO-2373 and

MO-2374 During As Found Test(88 scfh combined); dated March 6, 2005*037603; Crack Like Indication Found on Steam Dryer; dated March8, 2005

037632; Control Rod Drive Pipe Thru Wall Leakage; dated March 9, 2005

037635; Linear Indication was Found in Piping; dated March 9, 2005

037675; Dryer-Separator Storage Pool Leak Detection System Indicates Pool Leak;dated March 11, 2005*037922; Flow Accelerated Corrosion Damage Identified in Pipe After

MO-2008 wasRemoved; dated March 17, 2005*037956; Shroud Inspections Discovered Indications in New Regions; datedMarch 18, 2005*037973; ISI/IWE Indication, ID Pitting Reported on Piping Outboard of -x47 During

VT-3;dated March 18, 2005*038158; NDE Readings; dated March 24, 2005

038523; Shroud Head Bolts 4 and 32 are Degraded; dated April 6, 2005

038813; In-Service Inspection Indication,

FW-97-1, Erosion on Hinge Pin Plug HoleThreads; dated April 26, 2005*038834; Corroded Wires Found in Diesel Fire Pump Control Cabinet; datedApril 28, 2005*039593; Seat Leakage Identified on Withdrawal Scram Valve; dated June 21, 2005*039657; Degraded Wiring of Diesel Fire Pump System; dated June 25, 2005

039869; Service Water Piping in Reactor Building Showing Signs of Wear from Metal toMetal Contact; dated July 14, 2005*040897; Strap Type Support on Emergency Diesel Generator 11 Ductwork Missing Nuton Anchor Bolt; dated September 21, 2005*040967; Through Wall Corrosion on Floor of Turbine Lube Oil Storage Tank T-42B;dated September 27, 2005*00694385;

IN 2003-20:

Derating Whiting Cranes Purchased Before 1980; datedOctober 29, 2003*00708389; Revise UFSAR Fluence Calculation Methodology; dated May 2, 2003

00735368; Loose Hangers Found on

SW3-JF Piping near the Condensate Pumps;dated July 27, 2004*00800385; Roof Leaking onto Turbine Floor Above V-EF-9; dated January 25, 2005

00804228; Canal Sample Pump House Has Material Condition Issues; datedFebruary 2, 2005*00815635; Several 3" Chunks of Concrete Found Under the Turbine Stop Valves; datedMarch 5, 2005

9*00838288; Small Holes in Cable Spreading Room Ceiling and Wall, NotThrough-Barrier; dated April 26, 2004*00859716; NRC Commitment to Retain Capsules Removed from MNGP; datedJune 22, 2005*0088473-02; Root Cause Evaluation of 13 Diesel Generator Vibration Induced Failure;dated December 16, 2005*01007906; Small Areas of Coatings in Containment Did Not Meet Inspection ProcedureAcceptance Criteria; dated December 7, 2001*03005356; Small Areas of Coatings in Containment Did Not Meet Inspection ProcedureAcceptance Criteria, 2003 RFO; dated May 15, 2003*03012378;

SIL 564, "Verification of SRM, IRM or LPRM Detector Response; datedDecember 04, 2003*98000890; Small Areas of Coatings in Containment Did Not Meet the AcceptanceCriteria Specified in Inspection Procedures; dated April 6, 1998Procedures*0135; Pressure-Suppression Chamber Painted Surface Internal Inspection; Revision 8*0136; Integrated Primary Containment Leak Rate Test; Revision 13

0137; Master Local Leak Rate Test; Revision 25

0140; Drywell Interior Surface Inspection; Revision 5

0192; Diesel Fuel Quality Check; Revision 18

0268; Fire Protection System Flow Test; Revision 15

1253; Underground Piping Inspection; Revision 4

1350; Underground Storage Tank Liquid Level Correlation; Revision 6

1362; Air Quality Test for the Instrument Air System; Revision 7

1367; Pressure-Suppression Chamber Below Water Line Painted Surface InternalInspection; Revision 1*1385; Periodic Structural Inspection; Revision 3

1396; Equipment/Structures Settling Check; Revision 2

1404-01; Emergency Diesel Generator Essential Service Water Heat ExchangerPerformance Test; Revision 9*1435-1; Underground Storage Tank Quarterly Monitoring; Revision 2

1475; Equipment Cycles Surveillance; Revision 0*3263; Controlled Specification Data Sheet (Fuel Oil); dated June 20, 1985*3802; Visual Inspection of Heat Exchanger Condition; Revision 0

8095; Fill Diesel Oil Receiving Tank From Truck; Revision 14*8096; Fuel Transfer From Diesel Oil Receiving Tank to Diesel Oil Storage Tank;Revision 5*8199; Generator Rotor Lift and Associated Crane Inspection Procedure; Revision 4

8236; Application of Nuclear Coatings; Revision 8

8280; Torus Painting; Revision 6

20-A-10; Southwest Equipment Room V-AC-4 High Temperature; Revision 6

20-A-17; Southwest Equipment Room V-AC-5 High Temperature; Revision 6

4AWI-06.06.01; Material Handling and Control of Heavy Loads; Revision 11

4AWI-07.04.02; Plant Chemistry Program; Revision 3

4AWI-09.04.00; Inservice Inspection Licensee Control Program; Revision 3

4AWI-09.04.02; System and Component Pressure Testing Program; Revision 11

4058-05-PM; "A" RHR Room Air Cooling Unit V-AC-5 Internal Cleaning, ExternalCleaning and Visual Inspection; Revision 9

10*4125-PM; East Service Water Bay Inspection and Dredging; Revision 9*4126-PM; West Service Water Bay Inspection and Dredging; Revision 9

4159-PM; Instrument and Service Air Leak Survey; Revision 1

4250-01PM; Reactor Building Crane, Bridge Drive System; Revision 20

4250-02PM; Reactor Building Crane, Trolley Drive System; Revision 19

4260-PM; Refueling Platform Inspection and Lubrication; Revision 22

4270-01PM; Turbine Building Crane, Bridge Drive System; Revision 16

4270-02PM; Turbine Building Crane, Trolley Drive System; Revision 16

4361-PM; Reactor Building Crane Inspection Checklist; Revision 3

4858-48-PM; 2R Transformer and Associated Bus PM; Revision 6

4858-59-PM; 1R Transformer and Associated Bus Maintenance Procedure; Revision 8

4864-PM; Reactor Vessel Head Lifting Device Inspection Procedure; Revision 6

B.09-15.01; Non-Essential Diesel Generator; Revision 4

B.8.11-01; Diesel Oil System; Revision 1

B.9.15; Maintenance Rule Program, System Basis Document Non-Essential DieselGenerator; Revision 2*C.4-B.09.02.A; Station Blackout Abnormal Procedures; Revision 27*E.4-01; Backfeed Bus 13 from 13 Diesel Generator; Revision 2

CD 5.17; Flow Accelerated Corrosion and Service Water Inspection Program Standard;Revision 1*CD 5.24; Reactor Vessel Integrity Program Standard; Revision 0

CD 5.25; Generic Letter 89-13 Standard; Revision 0

CD 5.28; Conduct of Systems Engineering; Revision 2

ESM-02.02; Design Requirements, Practices and Topics (Mechanical); Revision 11

EWI-05.02.01; Monticello Maintenance Rule Program Document; Revision 7

EWI-08.01.01; Boiling Water Reactor Vessel Internals Project (BWRVIP) AdministrativeManual; Revision 5*EWI-08.01.02; BWRVIP Implementation Guidelines; Revision 2

EWI-08.06.01; Primary Containment Leakage Rate Testing Program; Revision 2

EWI-08.07.01; Thermal Fatigue Monitoring Program; Revision 0

EWI-08.19.01; Cable Condition Monitoring Program; Revision 0

EWI-08.22.01; Generic Letter 89-13; Revision 1

EWI-08.22.02; Heat Exchanger Condition Assessment Program; Revision 1

EWI-09.04.00; ASME Section XI Inservice Inspection Program; Revision 0

FP-IH-EXC-01; Excavation and Trenching Controls; Revision 0

FP-PE-FAC-01; Flow Accelerated Corrosion Inspection Program; Revision 1

I.05.30; Sampling Underground Fuel Oil Tanks Monitoring Points; Revision 3

II.01; Strategic Chemistry Plan; Revision 6

II.03; Control and Diagnostic Parameters; Revision 2

II.05; Chemistry Limits and Sampling Frequencies; Revision 15

MWI-8-M-4.10; Concrete Expansion Bolt Installation; Revision 8

PEI-02.03.06; Ultrasonic Examination of Bolts and Studs to Appendix VIII; Revision 0*PEI-02.05.01; Visual Examination; Revision 0

PEI-02.05.03; Visual Examination of Class MC Components (VT-3); Revision 0

PEI-02.05.07; Visual Examination (VT-1) of Class MC Components; Revision 0Procedure Change Requests*000511; Flow Accelerated Corrosion Program Fleet Procedure Issue:

FP-PE-FAC-01;dated May 20, 2005

11Reports*04-15FWH to RX; CHECWORKS Wear Rate Analysis:

Combined Summary Report;dated October 21, 2005*2000U041; Ultrasonic Examination of the Reactor Vessel Nozzle N-2F NV; datedJanuary 12, 2000*2001M330; Magnetic Particle Examination:

Reactor Pressure Vessel (RPV) Studs;dated November 30, 2001*2001U302; Ultrasonic Bolting /Stud Examination:

RPV Studs; dated November 9, 2001*2001U329; Ultrasonic Examination of the Reactor Vessel Nozzle N-1B NV; datedNovember 10, 2001*2001U331; Ultrasonic Examination of the Reactor Vessel Nozzle N-2A NV; datedNovember 11, 2001*2001U332; Ultrasonic Examination of the Reactor Vessel Nozzle N-2B NV; datedNovember 11, 2001*2001UT025; Ultrasonic Examination of the Reactor Vessel Nozzle N-4C NV; datedMarch 16, 2005*2001UT026; Ultrasonic Examination of the Inner Radius of Reactor Vessel Nozzle N-4CIR; dated March 16, 2005*2002-001-5-016; Observation Report:

NEC Generic Letter 89-13, Service WaterSystem/Heat Exchanger Program; dated February 28, 2002*2002-011; Structural Deficiency Identification and Evaluation Form 4266; datedAugust 19, 2002*2002-024; Structural Deficiency Identification and Evaluation Form 4266; datedAugust 16, 2002*2003-003-5-024; Nuclear Oversight Observation Report:

Inservice Inspection Program;dated September 4, 2003*2005UT016; Ultrasonic Examination of the Reactor Vessel Nozzle N-2E NV; datedMarch 19, 2005*2005UT042; Ultrasonic Examination of the Inner Radius of Reactor Vessel Nozzle N-1AIR; dated March 27, 2005*94-0088; Ultrasonic Examination Report:

RPV Studs; dated September 23, 1994

98-0009; Ultrasonic Examination Report:

RPV Studs; dated March 24, 1998

98-016; Structural Deficiency Identification and Evaluation; dated August 3, 1998*98-021; Structural Deficiency Identification and Evaluation; dated August 3, 1998*EHT-416-1; Hardness Test Report, Reactor Vessel Head Stud 47; dated April 16, 1991

EHT-416-2; Hardness Test Report, Reactor Vessel Head Stud 48; dated April 16, 1991

EHT-416-3; Hardness Test Report, Reactor Vessel Head Stud 49; dated April 16, 1991

EHT-416-4; Hardness Test Report, Reactor Vessel Head Stud 50; dated April 16, 1991

EHT-416-5; Hardness Test Report, Reactor Vessel Head Stud 4-M; dated April 16, 1991

NP-5911SP; Acceptance Criteria for Structural Evaluation of Erosion-Corrosion Thinningin Carbon Steel Piping; dated July 1988*SA

021435; Generic Letter 89-13 Program and Ultimate Heat Sink FocusedSelf-Assessment; dated June 15, 2004*SASR 89-77; Accumulated Fatigue Usage for Reactor Pressure Vessel; datedDecember 1989*UST T-44; Out of Service Inspection Report on Tank T-44; dated May 6, 2003

XOE02010809; External Operating Experience Notebook:

OE15015-InterconnectingPipe Not in the Flow Accelerated Corrosion Program; dated September 21, 2005

2*Containment Inspection Plan (IWE) 1

st Interval:

September 9, 1998 throughMay 8, 2008; Revision 1; dated August 2, 2004*Core Shroud Ultrasonic Examination Final Report; (draft)

Flow Accelerated Corrosion Master Plan; dated January 8, 2004

Focused Self-Assessment:

Reactor Vessel Thermal Cycles Program; datedMarch 31, 2004*Focused Self-Assessment:

Flow Accelerated Corrosion and Service Water/MICPrograms, October 27 -30, 2003; dated November 26, 2003*In Vessel Visual Inspection Final Report, 2005

RFO-22; dated March 28, 2005

Inservice Inspection Containment Examination Program (Class

MC-IWE) SummaryReport, Cycle 21 Refueling Outage; dated May 1, 2005*IWE Outage Summary Report, Refueling Outage 20; dated December 23, 2002

Maintenance Rule Program:

Periodic Assessment Report, June 2003 throughMay 2004; undated*Operating Experience Assessment:

RICSIL 055:

RPV Head Stud Cracking, Revision 1,Supplement 1; dated November 18, 1993*Out of Service Inspection Report:

Diesel Fuel Oil Storage Tank T-44; datedMay 6, 2003Surveillances*1253; Underground Piping Inspection; completed October 5, 1987*1362; Air Quality Test for the Instrument Air System; completed:

February 25, 2005,March 16, 2005, and August 1, 2005*1385; Periodic Structural Inspection; completed:

August 7, 1998 and August 21, 2002*5126; Instrument Air Particulate Sizing; completed:

February 25, 2005, March 20, 2005,and August 1, 2005System and Program Health and Status Reports*CDR; Main Condenser System; dated January 3, 2006*CFW; Condensate and Feedwater System; dated January 3, 2006

CRN; Cranes, Heavy Loads, Rigging System Health Report; dated April 25, 2005*FAC; Flow Accelerated Corrosion Program; dated January 1, 2006

FPC; Fuel Pool Cooling and Cleanup System; dated November 3, 2005

ISI; Inservice Inspection Program; dated November 1, 2005 and December 30, 2005

MR; Maintenance Rule Program; dated December 1, 2005

RWCU; Reactor Water Cleanup System; dated January 24, 2006

TF; Thermal Fatigue Program; dated October 5, 2005Work Order Written as a Result of the Inspection*00003054; Refurbish Support Between

SW-10 and

SW-11; dated January 26, 2006

Work Orders

Reviewed During the Inspection*9206184; Clean and Perform Internal Visual Inspection of T-44; datedSeptember 16, 1993*9501798; Perform 5 Year Inspection on Turbine Crane; dated November 13, 1995

13*9906790; Inspection of Underground Off Gas Piping Per Procedure 1253; datedSeptember 17, 1999*0202752; Clean and Inspect Diesel Fuel Oil Storage Tank; dated May 6, 2003

0203364; Replace Lube Oil Cooler Core, 12 Emergency Diesel Generator; datedMay 5, 2003*0205376; Adjust Reactor Building Crane North Rails; dated November 14, 2002

0306092; South Rails on Reactor Building Crane Has Too Big Gap; datedJanuary 24, 2003*0306207; Turbine Building Crane:

Replace Bolt on Main Hoist Gear Case; datedFebruary 5, 2003*0309622; Concrete on NW Corner of High Pressure Coolant Injection Building ChippedAway; dated June 17, 2003*0309628; Excessive Corrosion in Sodium Hypochlorite Room; dated June 17, 2003

0310892; Perform Electrical and Mechanical Preventive Maintenance on ReactorBuilding Crane Bridge, Trolley, Main and Auxiliary Hoist; dated October 1, 2003*0311446; Inspect and Repair Sheaves for Reactor Building Crane Auxiliary Hoist; datedNovember 26, 2003*0401172; Hydrolaze Division II FSW Piping; dated March 22, 2005

0402742; Tighten Hangers on Service Water System; dated August 3, 2004

0508437; Replace Missing Nuts on Fuel Pool Cooling Supports; dated October 31, 2005Work Requests*92-06184; Inspection and Cleaning of Diesel Fuel Oil Storage Tank; datedFebruary 11, 1993*92-06189; Flushing of 11 Emergency Diesel Generator Day Tank; dated March 8, 1993

92-06192; Flushing of 12 Emergency Diesel Generator Day Tank; datedFebruary 11, 1993

EXIT MEETING SLIDESAttachment

1Monticello Nuclear Generating PlantScoping, Screening and Aging ManagementLicense Renewal InspectionExit MeetingFebruary 22, 2006AgendaIntroductionsLicense Renewal ProcessNRC Inspection ResultsApplicant CommentsPublic Questions

EXIT MEETING SLIDESAttachment

2License Renewal ProcessFederal regulations (10 CFR

Part 54) allow for renewal of operating licenses for an additional 20 year periodMonticello submitted its application for a renewed license on March 24, 2005License Renewal Process, cont.Current licensegranted for 40 year period; it will expire on September 8, 2010Operating license would be extended to 2030Day-to-day operations under

10 CFR Part 50 not changed

EXIT MEETING SLIDESAttachment

3License Renewal Process, cont.License Renewal Process includesTechnical Review by Office of Nuclear Reactor Regulation (NRR)Onsite Scoping AuditOnsite Aging Management AuditCulminates in Safety Evaluation ReportLicense Renewal Process, contFinally, process includes onsite inspection by regional officeFocuses on implementation and commitment managementEmphasizes in-plant walk downsCulminates in an inspection report

EXIT MEETING SLIDESAttachment

4InspectionOnsite inspection performed in accordance with NRC Inspection

Procedure

71002Inspected scoping, screening and aging management programsInspection, cont.Inspection occurred from January 23 to February 10, 2005 Consisted of two weeks onsiteInspection team consisted of five experienced inspectors plus license renewal project manager

EXIT MEETING SLIDESAttachment

5Inspection, cont.Team reviewed electrical, mechanical and structural systems, structures and componentsNormally inaccessible areas inside containment will be examined by routine inspections during next refueling outageScoping and ScreeningReviewed 12 systems to verify appropriateness of scoping and screening effortsSupplemented NRR'sreviewEmphasized physical walk downs of the plantConcentrated on systems, structures and components notincluded within the applicant's license renewal scope

EXIT MEETING SLIDESAttachment

6Scoping and Screening, cont.Systems reviewed included:Condensate and Feedwater Diesel Generator Starting AirDrywell Atmosphere CoolingFuel Pool Cooling and CleanupHeating and VentilationInstrument AirScoping and Screening, cont.Conclusion

Systems generally appeared appropriately scoped and screenedSome minor inconsistencies identifiedIn some cases, the changes required a revision to the application; this will be part of the annual application updateScoping and screening acceptable for license renewal

EXIT MEETING SLIDESAttachment

7Aging Management ProgramsReviewed 33 aging management programs and 2 time limited aging analyses (TLAA) programsFollowed up on NRR review effortsPerformed plant walk downsReviewed existing plant documentation Aging Management Programs,contReviewed operational experience informationReviewed corrective actions to current plant issues andReviewed proposed enhancements and commitmentsVerified that TLAAswere appropriate

EXIT MEETING SLIDESAttachment

8Aging Management Programs,contThe programs reviewed included:ASME Section XI (2 programs)Buried Piping & Tanks Inspection Bus Duct Inspection Program BWR Nozzles and Penetrations (3 Programs) BWR Stress Corrosion Cracking & Neutron Irradiation Embrittlement(2 Programs)BWR Vessel Internals and Attachments(2 Programs)Aging Management Programs,contPrograms reviewed, cont.Compressed Air MonitoringOpen & Closed Cycle Cooling Water Systems (2 Programs)Plant Chemistry ProgramPrimary Containment Inspection &

Leak Rate Testing (2 Programs)Protective Coating Program

EXIT MEETING SLIDESAttachment

9Aging Management Programs,contPrograms reviewed, cont.Reactor Head Closure Studs & Bolting Integrity (2 Programs)Reactor Vessel SurveillanceStructures & System Condition Monitoring (2 Programs)Time Limited Aging AnalysesEnvironmental QualificationFatigue MonitoringAging Management Programs,contPrograms reviewed, cont.Non-EQ Cables (3 Programs)Fire Protection & Fire Water(2 Programs)Flow-Accelerated CorrosionFuel Oil ChemistryInspection of Overhead CranesOne-Time Inspection & Selective Leaching (2 Programs)

EXIT MEETING SLIDESAttachment

10Aging Management Programs,contConclusionExisting aging management programs generally implemented as described in the applicationEnhancements and exceptions appeared acceptable and were captured in commitment tracking databaseAging Management Programs,contConclusion, cont.Some minor inconsistencies identified which either required revision to the application or documentation in the corrective action program Application revisions will be part of the annual application updateAging Management Programs should be adequate for period of extended operation

EXIT MEETING SLIDESAttachment

11Inspection ConclusionsMonticello scoping, screening and aging management programs sufficient for extended operationRegion III does not see any inspection impediments to renewing the operating license

2

LIST OF ACRONYMS

USEDADAMSAgency Wide Access Management SystemAMPAging Management Program

AMSACATWS Mitigation System Actuating Circuitry

ARAction Request

ASMEAmerican Society of Mechanical Engineers

ASTMAmerican Society for Testing and Materials

ATWSAnticipated Transient Without Scram

BWRBoiling Water Reactor

BWRVIPBoiling Water Reactor Vessel Internals Project

CASSCast Austenitic Stainless Steel

CCCWClosed Cycle Cooling Water

CFRCode of Federal Regulations

CFWCondensate and Feedwater Systems

CRDRLControl Rod Drive Return Line

FDegree FahrenheitDRSDivision of Reactor Safety

ECCSEmergency Core Cooling Systems

EQEnvironmental Qualification

EVTEnhanced Visual (Test) Examination

FACFlow Accelerated Corrosion

FPFire Protection

GALLGeneric Aging Lessons Learned

GEGeneral Electric

HTVHeating and Ventilation

IASCCIrradiation Assisted Stress Corrosion Cracking

IDInside Diameter

ISGInterim Staff Guidance

ISIInservice Inspection

ISPIntegrated Surveillance Program

ksiKilo (1000) Pounds per Square Inch

kVkiloVolt (1000 Volts)

LERLicensee Event Report

LRLicense Renewal

LRALicense Renewal Application

MNGPMonticello Nuclear Generating Plant

NMCN uclear Management Company,

LLC [[]]

NRCNuclear Regulatory Commission

NRROffice of Nuclear Reactor Regulation

OCCWOpen Cycle Cooling Water

PARSPublically Available Records System

PCTPeak Clad Temperature

PMPreventive Maintenance

RGRegulatory Guide

RSWResidual Heat Removal Service Water

SILService Information Letter

SSCSystem, Structure, or Component

SSWService and Seal Water

TLAATime Limited Aging Analyses

13USARU pdated Safety Analysis ReportVTVisual (Testing) Examination

v t e Inspection Report - Monticello - 2006006
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2006 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 30(0) 8(0) 9(1) 11(0) 12(0) Security 401(0) Operator Exams Emergency Planning Other

IR 05000263/2006006

Text

Dear Mr. Conway:

Enclosure:

w/Attachment:

Enclosure:

w/Attachment:

SUMMARY OF FINDINGS

REPORT DETAILS

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

LIST OF DOCUMENTS REVIEWED

Work Orders

Procedure

LIST OF ACRONYMS

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