exe krn.m imarch 05, 2004 u. s. nuclear regulatory commission page 2 should you have any questions,...

Exe I krn.MExelon Generation4300 Winfield RoadWarrenville, IL 60555

www.exeloncorp.coM Nuclear

10 CFR 54

RS-04-039

March 05, 2004

U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555-001

Dresden Nuclear Power Station, Units 2 and 3Facility Operating License Nos. DPR-19 and DPR-25NRC Docket Nos. 50-237 and 50-249

Quad Cities Nuclear Power Station, Units 1 and 2Facility Operating License Nos. DPR-29 and DPR-30NRC Docket Nos. 50-254 and 50-265

Subject:

Reference:

Amendment to the Application for Renewed Operating Licenses forDresden and Quad Cities Nuclear Power Stations

Letter from J. A. Benjamin (Exelon Generation Company, LLC) to U. S.NRC, "Application for Renewed Operating Licenses," dated January 3,2003

The reference letter submitted an Application for Renewed Operating Licenses for theDresden Nuclear Power Station (DNPS), Units 2 and 3 and Quad Cities Nuclear PowerStation (QCNPS), Units 1 and 2. Exelon Generation Company, LLC (EGC) is submittingthe annual amendment to the DNPS and QCNPS License Renewal Application (LRA) inaccordance with 10 CFR 54.21(b). This amendment identifies changes to the currentlicensing basis (CLB) that materially affect the contents of the DNPS and QCNPS LRA,including the UFSAR supplement. This amendment is required to be submitted eachyear following submittal of the LRA and at least 3 months before scheduled completionof the LRA review by the NRC.

Attachment 1 provides a description of the changes and replacement pages of theDNPS and QCNPS LRA that have been materially affected by changes to the CLB. f\Ci1

9o3

March 05, 2004U. S. Nuclear Regulatory CommissionPage 2

Should you have any questions, please contact Al Fulvio at 610-765-5936.

I declare under penalty of perjury that the foregoing is true and correct.

Respectfully,

My" 5, eo qExecuted on Patrick R. Simpson

Manager - Licensing

I'

Attachment 1: Amendment to the Application for Renewed Operating Licenses forDresden and Quad Cities Nuclear Power Stations

cc: Regional Administrator- NRC Region IIINRC Senior Resident Inspector - Dresden Nuclear Power StationNRC Senior Resident Inspector - Quad Cities Nuclear Power StationIllinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1

Amendment to the Application for Renewed Operating Licenses for Dresden and QuadCities Nuclear Power Stations

1.0 INTRODUCTION

The License Renewal Rule, 1 OCFR54.21 (b), requires that each year following submittal of alicense renewal application (LRA), an amendment must be submitted to identify changes to thefacility current licensing basis (CLB) that materially impact the content of the LRA. Inaccordance with this requirement, Exelon Generation Company, LLC (Exelon) has completedthe review of Dresden Nuclear Power Station (DNPS), Units 2 and 3 and Quad Cities NuclearPower Station (QCNPS), Units 1 and 2, CLB changes since the submittal of the LRA.Additionally, Exelon has identified some areas that required correction. This attachmentprovides its results, and required revisions and corrections to the LRA.

2.0 REVIEW RESULTS

The review identified seven (7) CLB changes and four (4) additional corrections that impact theLRA. The CLB changes resulted from both plant changes and analysis changes. Additionally,the UFSAR Supplement and Aging Management Program Descriptions were updated. Thereview did not identify new operating experience that affects the content of the LRA.

Each change and its impact on the LRA are briefly discussed below in the order they appear inthe LRA and the same order they will be found on the attached marked up changes.

* Exelon Identified the Need for an Additional Aqing Manaqement Reference for theShroud Support Structure This additional aging management reference was neededto provide a link from the Component Description to the Time Limited Aging Analysis(TLAA) documented in Chapter 4. LRA Table 2.3.1-2, Component GroupsRequiring Aging Management Review - Reactor Internals has been updated to addReference 3.1.1.1 to Component Group Shroud Support Structures. Additionally,LRA Table 3.1-1, Item 3.1.1 was revised to add Shroud Support Structures, as thisTLAA is documented in Section 4.3.2. (This item was a correction.)

* High Pressure Coolant Iniection Pump Material Change at Dresden The HPCI castiron pumps 2-2320-GSLO and 3-2320-GSLO were replaced with stainless steelpumps. The pumps are currently covered by LRA References 3.2.2.69 (internal)and 3.2.2.17 (external). These references need to remain as they still cover othervalid components. LRA References 3.2.1.13 (internal) and 3.2.2.22 (external)already exist to cover these new materials. However, the "pumps" line item of LRATable 2.3.2-1, Component Groups Requiring Aging Management Review - HighPressure Coolant Injection System, has been revised to add LRA References3.2.1.13 and 3.2.2.22. (This was a CLB change.)

* Installation of Tie-ins to Isolation Condenser Makeup Pump Suction to Allow forTemp Make-Up Capabilities at Dresden This installation of stainless steel valves inthe isolation condenser makeup pump suction to allow temporary makeupcapabilities. The LRA already contained Reference 3.2.2.22 (external) for the newstainless steel valves, which is included in LRA Table 2.3.2-5, Component GroupsRequiring Aging Management Review - Isolation Condenser (Dresden only), for the"valves" line item. LRA Reference 3.2.1.13, which addresses the internalenvironment of the new stainless steel valves has been added to the 'valves" lineitem of Table 2.3.2-5. (This item was a CLB change.)

* Exelon Identified the Need to Add Debris Screens for Dresden's Emerqency DieselGenerator System. The addition of the Dresden debris screens required a changeto LRA Section 2.3.3.6, "Emergency Diesel Generator and Auxiliaries," Table 2.3.3-6, Component Group of "Debris Screens (Quad Cities Only)" to remove the "Quad

Cities Only" notation. (This item was a correction.)

* Replacement Diesel Generator Service Water Cast Iron Filters with Cast SteelFilters at Dresden The cast iron filters 2-3999-38, 3-3999-380, and 2/3-3999-380were replaced with cast steel filters. The cast iron filters, covered by LRAReferences 3.3.2.208 (internal) and 3.3.2.31 (external), need to remain as they stillapply to other valid components. LRA Reference 3.3.1.15 (internal) and 3.3.1.5(external) already exist to cover the new material. Filters are covered by line item"strainer bodies" in Table 2.3.3-12, the 'strainer bodies" line item in LRA Table 2.3.3-12, Component Groups Requiring Aging Management Review - Diesel GeneratorCooling Water System, has been revised to add LRA References 3.3.1.15, and3.3.1.5. (This item was a CLB change.)

* Diesel Generator Cooling Water Pump Material Change at Dresden This changeapproved stainless steel as an acceptable material for pumps 2-3903, 3-3903, and2/3-3903, which had been previously approved for carbon steel only. The pumpsare currently covered by References 3.3.1.15 (internal) and 3.3.2.26 (external). LRAReference 3.3.1.15 is adequate for both carbon steel and stainless steel internalenvironments. An additional reference for the external environment (air moisture,humidity, and leaking fluid) for stainless steel pumps has been added to LRA Table2.3.3-12, Component Groups Requiring Aging Management Review- DieselGenerator Cooling Water System. LRA Reference 3.3.2.41 has been modified toinclude pumps. Reference 3.3.2.41 has been added to the line item "pumps" inTable 2.3.3-12. (This item is a potential CLB change, the design has been approvedand changes could occur prior to the new license being issued)

* Exelon Identified the Need to Revise the External Environment for Ultimate HeatSink Discharge Piping The Quad Cities station 16' discharge piping was assignedthe wrong external environment link (3.3.2.28). This piping is evaluated in theUltimate Heat Sink System. This pipe is carbon steel buried in soil. The correct linkis 3.3.1.16. LRA Table 2.3.3-22, Component Groups Requiring Aging ManagementReview - Ultimate Heat Sink, has been revised, adding AMR Reference 3.3.1.16(external) to Component Group Piping and Fittings. (This item was a correction.)

* A modification replaced the electric nitrogen vaporizer at Quad Cities. Themodification also added a strainer. The addition of strainer 1-8741 during thismodification resulted in a new component type for Quad Cities in LRA Section2.3.3.27, 'Drywell Nitrogen Inerting System," Table 2.3.3-27. Table 2.3.3-27,contains Component Groups of "Filter/Strainers (Dresden Only).' The LRA changeneeded is to remove the "Dresden Only" notation from Table 2.3.3-27, ComponentGroups of "Filter/Strainers (Dresden Only)," as identified in the attachment below.Table 2.3.3-27, component groups of "Filter/Strainers (Dresden Only)." contain theproper links to Table 3.3-2, lines 3.3.2.25 and 3.3.2.51 for the external and internalsurfaces of the brass or bronze filter/strainers. (This was a CLB change.)

* Exelon Identified the Need to Change the Classification of Quad Cities Station CribHouse Fire Doors The fire rated door component scoping was evaluated and itsclassification was changed to be in the scope of the Rule requiring agingmanagement, similar to Dresden. This required revision to the License RenewalApplication Table 2.4-11, to delete "(Dresden only)" from component 'Fire Doors(Dresden only)." (This item was a correction.)

* Changes to 4.2.6 and 4.2.7 These sections applied only to Dresden in the LRAsubmittal because Dresden had applied for and received approval for relief from

reactor vessel circumferential weld inspections. After submittal of the LRA, Exelondiscovered that Dresden LRA Tables 4.2.6-1 and 4.2.7-1 contained errors requiringcorrection as described in the Exelon letter to USNRC dated April 17, 2003.

Exelon subsequently applied for similar relief from reactor vessel circumferentialweld inspections for Quad Cities by letter to USNRC dated May 16, 2003. Thereforethis topic became a TLAA for Quad Cities, and these LRA sections have beenupdated to reflect applicability to both stations. The updated sections alsoincorporate the Dresden table corrections previously submitted.

These sections were replaced in their entirety. (These were CLB Changes)

Change of the Cumulative Fatigue Usage Factor for Quad Cities Reactor VesselClosure Studs LRA Section 4.3.1 reported on the results of a reanalysis performedfor Extended Power Uprate (EPU) which concluded that the current boundinganalysis for the reactor vessel closure studs listed a value of 0.750 for the 40-yearcumulative usage factor (CUF). However, a subsequent analysis indicates that thereactor vessel closure studs will have a bounding value of < 1.0 for the 40-yearcumulative usage factor (CUF). The studs will be monitored under the FatigueMonitoring Program (B.1.34), which will adequately manage their aging. (This was aCLB change)

* Appendix A, Updated Final Safety Analysis Report (UFSAR) SupplementThis section has been updated to reflect the following changes:* ASME Section Xl will be implemented in accordance with 10 CFR 50.55(a).* Specific commitments and exceptions to various BWRVIP documents have been

incorporated into the applicable programs.* Updated the status of the reactor pressure vessel circumferential weld

examination relief request.* Added drywell corrosion monitoring program for Dresden Unit 3.* Incorporates changes resulting from RAI Responses.

Appendix A has been replaced in its entirety.

* Changes to ASME Section Xl Programs. Appendices A and B of the LRA identifyactivities that are credited for compliance with the License Renewal Rule for agingmanagement of passive, long-lived components and structures within the scope ofLicense Renewal. Aging Management Reviews (AMRs) developed in support of theDNPS/QCNPS LRA provided further detail regarding the aging managementactivities identified in the LRA appendices.

ASME Section Xl programs are credited in the aging management activities for anumber of systems. ASME Section Xl programs have had enhancements made tothe Edition and Addenda implemented for these programs to comply with 10 CFR50.55(a). The LRA originally indicated that the current Code of record for Dresdenand Quad Cities was the 1989 Edition of ASME Section XI, and committed to updatethe program to be consistent with the requirements of the 1995 Edition through the1996 Addenda of ASME Section Xl, as recommended by NUREG-1801.DNPS/QCNPS have updated the ASME Section XI programs as committed. Futureupdates of the ASME Section Xl programs will be in accordance with 10 CFR50.55(a).

3.0 REVISIONs To LRA

Revised pages to the LRA that reflect the changes described in Section 2.0 above are providedhere. The italic and bold text identifies the required additions and the strikethrough textidentifies the required deletions to the LRA.

The pages revised as a result of this annual update also reflect those changes due to RAIresponses that affected the same pages. Because Appendix A is provided in its entirety, allRAI related changes are included in that appendix. However, changes to other LRA pages thatresulted only from submitted RAI responses are not included in the annual update.

LRA Replacement Page 2-47

Table 2.3.1-2 Component Groups Requiring Aging Management Review - Reactor Internals. (Continued)I Component Group Component Intended Aging Management Ref

Coe S y LeFunction 3 3

Core Spray Lines and Spargers Pressure Boundary 3.1.1.1, 3.1.1.17

Core Spray Lines and Spargers Spray 3.1.1.1, 3.1.1.17

Core Spray Lines and Spargers Structural Support 3.1.1.1, 3.1.1.17

Incore Instrumentation Dry Pressure Boundary 3.1.1.1, 3.1.1.17Tubes and Guide Tubes

Jet Pump Assemblies (Does not Pressure Boundary 3.1.1.1, 3.1.1.17, 3.1.1.19include Sensing Lines) l _

et Pump Assemblies (Does not tructural Support 3.1.1.1, 3.1.1.17include Sensing Lines)

Orificed Fuel Support Pieces Structural Support 3.1.1.17, 3.1.1.19

Orificed Fuel Supports Structural Support 3.1.1.1

Reactor Internals Structural Support 3.1.1.17Modification/Repair Hardware

* Core Spray Clamp

* Jet Pump Riser Clamp(Quad Cities only)

* Jet Pump Riser BraceClamp (Quad Citiesonly)

* Shroud Repair

Shroud Support Structures Structural Support 3.1.1.1, 3.1.1.17

Top Guides Structural Support 3.1.1.1, 3.1.1.17

Aging management review results for the reactor internals are provided in Section 3.1.


Table 2.3.2-1 Component Groups Requiring Aging Management Review - High PressureCoolant Injection System (Continued)

Component Group Component Intended Aging Management Ref ]ll Function I

Filters/Strainers (includes Filter 3.2.1.2, 3.2.1.4, 3.2.1.13, 3.2.2.32separators)

Flexible Hoses Pressure Boundary 3.2.2.33, 3.2.2.34

Flow Orifices Pressure Boundary 3.2.1.2, 3.2.1.4, 3.2.2.137

Heat Exchangers (includes Pressure Boundary 3.2.2.40, 3.2.2.41, 3.2.2.42, 3.2.2.43,condensers) 3.2.2.137

Heat Exchangers Heat Transfer 3.2.2.38, 3.2.2.39

NSR Vents or Drains, Piping and Structural Integrity (attached) 3.2.2.10, 3.2.2.55, 3.2.2.136alves (attached support)

Piping and Fittings (includes Pressure Boundary 3.1.1.1, 3.1.1.11, 3.2.1.1, 3.2.1.2,thermowells) 3.2.1.3, 3.2.1.4, 3.2.1.5, 3.2.1.13,

.2.2.2, 3.2.2.9, 3.2.2.10, 3.2.2.13,3.2.2.14, 3.2.2.22, 3.2.2.23, 3.2.2.24,3.2.2.25, 3.2.2.26, 3.2.2.27, 3.2.2.28,3.2.2.56, 3.2.2.58, 3.2.2.59, 3.2.2.64,3.2.2.65, 3.2.2.68, 3.2.2.126, 3.2.2.127,3.2.2.137

Piping and Fittings (attached Structural Integrity (attached) 3.2.1.13, 3.2.2.10, 3.2.2.14, 3.2.2.14,support) _ 3.2.2.22, 2.2.2.56, 3.2.2.57, 3.2.2.137Piping and Fittings (small bore) Pressure Boundary 3.1.1.5, 3.2.2.9, 3.2.2.10, 3.2.2.13,

3.2.2.14, 3.2.2.22, 3.2.2.23, 3.2.2.24,3.2.2.25, 3.2.2.26, 3.2.2.27, 3.2.2.28,3.2.2.137,

Pumps Pressure Boundary 3.2.1.2, 3.2.1.4, 3.2.1.13, 3.2.2.17,3.2.2.22, 3.2.2.69, 3.2.2.70, 3.2.2.71,3.2.2.137

Restricting Orifices Pressure Boundary 3.2.1.2, 3.2.1.4, 3.2.2.13, 3.2.2.128,3.2.2.137

Restricting Orifices Throttle 3.2.1.2, 3.2.1.4, 3.2.2.128

Restricting Orifices (attached Structural Integrity (attached) 3.2.2.72, 3.2.2.13, 3.2.2.128, 3.2.2.137support) .Rupture Discs Pressure Boundary 3.2.2.22, 3.2.2.129

Rupture Discs (attached support) Structural Integrity (attached) 3.2.2.22, 3.2.2.129

Sight Glasses (attached support) Structural Integrity (attached) 3.2.2.20, 3.2.2.75, 3.2.2.76, 3.2.2.137

Sight Glasses (Quad Cities only) Pressure Boundary 3.2.2.5, 3.2.2.20, 3.2.2.76

Tanks Pressure Boundary 3.2.1.2, 3.2.1.4, 3.2.2.13, 3.2.2.83,.__ .3.2.2.84, 3.2.2.130, 3.2.2.137


Table 2.3.2-5 Component Groups Requiring Aging Management Review - Isolation Condenser(Dresden only) (Continued)

Component Group | Component Intended Aging Management Ref_ _ _ _ _ _ _ lFunction l

Piping and Fittings (Dresden Pressure Boundary 3.1.1.1, 3.1.1.15, 3.2.1.2, 3.2.1.4,only) 3.2.1.13, 3.2.2.10, 3.2.2.13, 3.2.2.14,

3.2.2.15, 3.2.2.22, 3.2.2.23, 3.2.2.24,3.2.2.25, 3.2.2.26, 3.2.2.56,.2.2.137

Piping and Fittings (attached Structural Integrity (attached) .2.2.13, 3.2.2.14, 3.2.2.15, 3.2.2.22,support) (Dresden only) 3.2.2.23, 3.2.2.24, 3.2.2.25, 3.2.2.61,

3.2.2.137

Piping and Fittings (small Pressure Boundary 3.1.1.5, 3.2.2.13, 3.2.2.14, 3.2.2.15,bore) (Dresden only) 3.2.2.22, 3.2.2.23, 3.2.2.24, 3.2.2.25,

3.2.2.137

Pumps (Dresden only) Pressure Boundary 3.2.1.13, 3.2.2.22

Flow Elements (Dresden only) Pressure Boundary 3.1.1.15, 3.2.2.24, 3.2.1.13, 3.2.2.22

Sight Glasses (Dresden only) Pressure Boundary 3.2.2.20, 3.2.2.76

Tanks (Dresden only) Pressure Boundary 3.2.2.10, 3.2.2.82

Thermowells (Dresden only) Pressure Boundary 3.1.1.15, 3.2.2.22

Tubing (Dresden only) Pressure Boundary 3.2.2.22, 3.2.2.23, 3.2.2.24, 3.2.2.97

Valves (Dresden only) Pressure Boundary 3.1.1.1, 3.1.1.15, 3.2.1.2, 3.2.1.4,3.2.1.12, 3.2.1.13, 3.2.2.13, 3.2.2.14,3.2.2.15, 3.2.2.22, 3.2.2.23, 3.2.2.24,3.2.2.25, 3.2.2.122, 3.2.2.137

Valves (attached support) Structural Integrity (attached) 3.2.2.22, 3.2.2.111, 3.2.2.116,(Dresden only) 3.2.2.137

Aging management review results for the isolation condenser system are provided in Section 3.1 for thereactor coolant pressure boundary functions and Section 3.2 for the additional isolation condenserfunctions.


System Intended Functions

Provide emergency AC power - provides independent power source to assure safe reactorshutdown under emergency conditions on a total loss of offsite power concurrent with a designbasis accident.

Credited in regulated event(s) - credited in support of fire protection (10CFR50.48). Thesystem contains components that are relied upon for compliance with 10 CFR 50.49, (EQ).

Preclude adverse effects on safety related SSCs - Non-safety related components that could bea hazard to safety related SSCs maintain sufficient integrity so that the intended function ofsafety related SSCs is not adversely affected.

Component Groups Requiring Aging Management Review

Table 2.3.3-6 Component Groups Requiring Aging Management Review - Emergency DieselGenerator and Auxiliaries

Component Group Component Intended Aging Management RefFunction

Air Accumulator Vessels Pressure Boundary 3.3.1.5,3.3.2.6Closure Bolting Pressure Boundary 3.3.1.22, 3.3.2.18

Debris Screens (Quad Cities- Filter 3.3.2.59

Doors, Closure Bolts, Equip Pressure Boundary 3.3.1.5Frames (includes dampers,duct, and housings)

Duct Fittings, Hinges, Latches Pressure Boundary 3.3.2.49(includes anchors, bolts, andfasteners)

Filters/Strainers Pressure Boundary 3.3.1.5, 3.3.2.29, 3.3.2.55, 3.3.2.58,3.3.2.60

Filters/Strainers Filter 3.3.1.5, 3.3.2.55, 3.3.2.58, 3.3.2.60

Flex Collars, Doors and Pressure Boundary 3.3.1.2Damper Seals

Flexible Hoses Pressure Boundary 3.3.2.65, 3.3.2.66

Heat Exchangers (includes Pressure Boundary 3.3.1.5, 3.3.2.34, 3.3.2.94, 3.3.2.95,coolers) 3.3.2.100, 3.3.2.101, 3.3.2.108,

3.3.2.109, 3.3.2.110, 3.3.2.111

Heat Exchangers (includes Heat Transfer 3.3.2.93, 3.3.2.96, 3.3.2.99coolers)

Lubricators Pressure Boundary 3.3.1.5


Table 2.3.3-12Component Groups Requiring Aging Management Review-Diesel GeneratorCoolina Water System (Continued)- - - - - * _ * _ I _ - _I .I . \ _ _

| Component Group Component Intended Aging Management Refl _ _ _ _ __ F u n c t i o n [ lPiping and Fittings (includes Pressure Boundary 13.3.1.5, 3.3.1.15, 3.3.2.26, 3.3.2.45,flow elements 3.3.2.141, 3.3.2.169

Piping and Fittings (attached Structural Integrity (attached) 3.3.1.5, 3.3.1.15support)

Pulsation Dampeners Structural Integrity (attached) 3.3.1.15, 3.3.2.41(attached support) (QuadCities only)

Pumps Pressure Boundary 3.3.1.15, 3.3.2.26, 3.3.2.41,3.3.2.173, 3.3.2.183, 3.3.2.184

Strainer Bodies (Dresden only) Pressure Boundary 3.3.1.5, 3.3.1.15, 3.3.2.31, 3.3.2.208

Strainer Screens (Dresden Filter 3.3.1.15only)

Thermowells Pressure Boundary 3.3.1.15, 3.3.2.26

Tubing Pressure Boundary 3.3.1.15, 3.3.2.40, 3.3.2.41

Valves Pressure Boundary 3.3.1.5, 3.3.1.15, 3.3.1.27, 3.3.2.23,3.3.2.24, 3.3.2.26, 3.3.2.31,3.3.2.40, 3.3.2.41, 3.3.2.45,3.3.2.279, 3.3.2.280, 3.3.2.298,3.3.2.300

Valves (attached support) Structural Integrity (attached) 3.3.1.15, 3.3.1.27, 3.3.2.23,3.3.2.24, 3.3.2.40, 3.3.2.41,3.3.2.279, 3.3.2.280, 3.3.2.300

AA .Aging management review results ior tne diesel generator cooling water system are provideu In ;~cauil13.3.



Table 2.3.3-22Component Groups Requiring Aging Management Review - Ultimate Heat Sink

Component Group | Component Intended | Aging Management RefI Function I l

Closure Bolting (Dresden Pressure Boundary 3.3.1.22only) | _ _llConcrete Slabs (Dresden Structural Pressure Barrier 3.5.1.22only) I.I

Concrete Walls (Dresden Structural Pressure Barrier 3.5.1.22only) l ll

Earthen Structures Structural Pressure Barrier 3.5.1.22

Piping and Fittings Pressure Boundary 3.3.1.5, 3.3.1.15, 3.3.1.16, 3.3.2.28,3.3.2.141

Pump Casings (Dresden only) Pressure Boundary 3.3.2.172, 3.3.2.300

Stop Logs (Dresden only) Structural Pressure Barrier 3.3.2.304

alves Pressure Boundary 3.3.2.278, 3.3.2.300

Aging management review results for the ultimate heat sink are provided in Section 3.3.



Table 2.3.3-27 Component Groups Requiring Aging Management Review - DrywellNitrogen Inerting System

Component Group Component Intended | Aging Management Refl _ _ _ _ _ _ Function l l

Closure Bolting Pressure Boundary 3.3.1.22

Filters/Strainers (Dresden Pressure Boundary 3.3.2.25, 3.3.2.51

Filters/Strainers (Resde Filter 3.3.2.51

Flow Elements Pressure Boundary 3.3.1.5, 3.3.2.40, 3.3.2.69, 3.3.2.71

Isolation Barriers Pressure Boundary 3.3.2.124, 3.3.2.125

Piping and Fittings Pressure Boundary 3.3.1.5, 3.3.2.27, 3.3.2.138,3.3.2.161

Tanks (includes vaporizers) Pressure Boundary 3.3.1.5, 3.3.2.21, 3.3.2.40,3.3.2.210, 3.3.2.212

Thermowells Pressure Boundary 3.3.1.5, 3.3.2.223

raps (Quad Cities only) Pressure Boundary 3.3.1.5, 3.3.2.228

Tubing Pressure Boundary 3.3.2.22, 3.3.2.34, 3.3.2.35,3.3.2.40, 3.3.2.43, 3.3.2.231,3.3.2.239, 3.3.2.248

Valves Pressure Boundary 3.3.1.5, 3.3.2.23, 3.3.2.25,3.3.2.40, 3.3.2.260, 3.3.2.262,3.3.2.268, 3.3.2.273, 3.3.2.289,3.3.2.295

Aging management review results for the drywell nitrogen inerting system are providedin Section 3.3.



Table 2.4-11 Component Groups Requiring Aging Management Review - Crib House

Component Component Intended Aging Management Ref.__Function

Concrete & Grout Structural Support 3.5.1.29

Concrete & Grout Non-S/R Structural Support 3.5.1.29

Concrete Canal Weirs (Quad Cities only) Heat Sink 3.5.1.22

Concrete Curbs Direct Flow 3.5.1.22

Concrete Slabs Structural Support 3.5.1.22, 3.5.1.26

Concrete Slabs Shutdown Cooling Water 3.5.1.22

Concrete Slabs Heat Sink 3.5.1.22, 3.5.1.26

Concrete Stairs Structural Support 3.5.1.22

Concrete Stairs Non-S/R Structural Support 3.5.1.22

Concrete Walls Structural Support 3.5.1.22

Concrete Walls Non-S/R Structural Support 3.5.1.22

Concrete Walls Shutdown Cooling Water 3.5.1.22

Concrete Walls Heat Sink 3.5.1.22

Fire Doors (Dresden Only) Fire Barrier 3.3.1.18

Foundations Structural Support 3.5.1.22, 3.5.1.26

Foundations Non-S/R Structural Support 3.5.1.22

Masonry Walls Structural Support 3.5.1.24

Masonry Walls Shelter, Protection, Shielding 3.5.1.24

Metal Siding (Dresden only) Shelter, Protection, Shielding 3.5.1.22

Misc. Steel (Dresden only) Non-S/R Structural Support 3.5.1.22

Misc. Steel (Dresden only) Direct Flow 3.5.1.22

Precast Concrete Panels Structural Support 3.5.1.22

Precast Concrete Panels Shelter, Protection, Shielding 3.5.1.22

Roofing Shelter, Protection, Shielding 3.5.2.11

Steel Embedments Structural Support 3.5.1.20, 3.5.1.22

Steel Embedments (Dresden only) Non-S/R Structural Support 3.5.1.22

Steel Panels and Cabinets Structural Support 3.5.1.20, 3.5.1.22Steel Panels and Cabinets (Quad Cities Non-S/R Structural Support 3.5.1.20only)Steel Plates (Dresden only) Direct Flow 3.5.1.22


Table 3.1-1 Aging management programs evaluated in NUREG-1801 that are relied on for license renewal for the reactor vessel, internals, andreactor coolant system

Ref No Component Components Evaluated Aging Aging Further DiscussionEffect/Mechanism Management Evaluation

Program Recommended _ __ __

3.1.1.1 Reactorcoolantpressureboundarycomponents

NUREG-1801Components

Closure BoltingCore PlatesCore Spray Lines and.

SpargersIncore Instrumentation

Dry Tubes and GuideTubes

Jet Pump Assemblies(Does not includeSensing Lines)

Nozzle Safe EndsNozzlesOrificed Fuel SupportsPenetrationsPiping and FittingsPumpsSupport Skirts and

Attachment WeldsTop GuidesTop Head Enclosure

(Head Flanges)ValvesVessel Bottom HeadsVessel Shells

Evaluated withNUREG-1801Components

Shroud SupportStructures

Cumulative fatiguedamage

TLAA, evaluated inaccordance with10 CFR 54.21(c)

Yes, TLAA Further Evaluation of cumulative fatiguedamage is described in Section 3.1.1.1.1,Section 4.3.1(reactor pressure vessel),Section 4.3.2 (reactor vessel internals),Section 4.3.3.1 (Dresden 3 RCPB piping),Section 4.3.3.2 (RCPB piping) and Section4.3.4 (environmental effects of fatigue).


Table 3.3-2 Aging management review results for the auxiliary systems that are not addressed in NUREG-1 801 (Continued)

Ref No Component Group Material Environment Aging Effect Aging DiscussionManagement

Program

3.3.2.41 Component External Stainless Steel Air, moisture, Loss of material/ Open-Cycle NUREG-1 801 does not address auxiliarySurfaces humidity, and Pitting and crevice Cooling Water system stainless steel components in a high

( o leaking fluid corrosion System (B.1.13) moisture (pump vault) indoor environment.(restricting orifices,valves, tubing,pulsationdampeners, pumps)

3.3.2.42 Component External Stainless Steel Containment None None NUREG-1 801 does not address stainless steelSurfaces Nitrogen components in a containment nitrogen(piping and fittings, environment. Containment nitrogen is notvalves, tubing) conducive to promoting aging degradation of

stainless steel.3.3.2.43 Component External Stainless Steel Outdoor None None NUREG-1801 does not address stainless steel

Surfaces ambient in the plant outdoor environment. Stainless(p an conditions steel materials are not subject to any viable(piping and Fatingsaging mechanism in the absence of aggressive

tubing) chemical species.

3.3.2.44 Component External Steel Saran Lined Air, moisture, Loss of material/ Open-Cycle NUREG-1801 does not address saran linedSurfaces humidity, and General pitting Cooling Water steel components in a plant indoor(valves, piping and leaking fluid crevice corrosion System (B.1.13) environment. The external surface of thefittings) components is unlined carbon steel.

3.3.2.45 Component External Titanium Air, moisture, None None NUREG-1 801 does not address titaniumSurfaces humidity, and components. The high moisture (pump vault)(valves, piping and leaking fluid indoor environment does not promote agingfittings) degradation of these titanium components as

they are not exposed to high chlorideconcentrations at high temp.

3.3.2.46 Dampeners Brass or Bronze Warm, moist Loss of material/ One-Time NUREG-1 801 does not address brass orair Pitting and crevice Inspection bronze components in a warm, moist air

corrosion (B.1.23) environment.


4.2.6 Reactor Vessel Circumferential Weld Examination Relief

ApplicabilityThis section applies to Dresden and Quad Cities.

Summary DescriptionRelief from reactor vessel circumferential weld examination requirements under GenericLetter 98-05 is based on probabilistic assessments that predict an acceptable probabilityof failure per reactor operating year. The analysis is based on reactor vesselmetallurgical conditions as well as flaw indication sizes and frequencies of occurrencethat are expected at the end of a licensed operating period.Dresden has received this relief for the remainder of the 40 year licensed operatingperiod. Quad Cities has submitted a relief request for the remainder of the 40 yearlicensed operating period. (Reference 4.27) The circumferential weld examination reliefanalysis meets the requirements of 10CFR54.3(a) and is a TLAA.

AnalysisDresden received NRC approval for a technical alternative which eliminated the reactorvessel circumferential shell weld inspections for the current license term. Quad Citieshas submitted a request for similar relief. The basis for this relief request was ananalysis that satisfied the limiting conditional failure probability for the circumferentialwelds at the expiration of the current license, based on BWRVIP-05 and the extent ofneutron embrittlement. The anticipated changes in metallurgical conditions expectedover the extended licensed operating period require an additional analysis for 54 EFPYand approval by the NRC to extend this relief request.

Disposition: Revision, 10 CFR 54.21 (c)(1)(ii)The USNRC evaluation of BWRVIP-05 utilized the FAVOR code to perform aprobabilistic fracture mechanics (PFM) analysis to estimate the RPV shell weld failureprobabilities. Three key assumptions of the PFM analysis are: 1) the neutron fluencethat was the estimated end-of-life mean fluence, 2) the chemistry values are meanvalues based on vessel types, and 3) the potential for beyond-design-basis events isconsidered.

DresdenTable 4.2.6-1 provides a comparison of the Dresden reactor vessel limitingcircumferential weld parameters to those used in the NRC analysis for the first two keyassumptions. Data provided in Table 4.2.6-1 was supplied from Tables 2.6-4 and 2.6-5of the Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05Report.

Although the chemistry composition and chemistry factor for unit 3 are higher than thelimits of the NRC Analysis; the 54 EFPY fluence results are considerably lower for bothDresden Units 2 and 3. As a result, the shifts in reference temperature for both units arelower than the 54 EFPY shift from the NRC analysis. In addition, the unirradiatedreference temperatures for both Dresden units are lower. The combination ofunirradiated reference temperature (RTNDT(U)) and shift (ARTNDTW/O margin) yieldsadjusted reference temperatures that are considerably lower than the NRC meananalysis values. Therefore, the RPV shell weld embrittlement due to fluence has a


negligible effect on the probabilities of RPV shell weld failure. The Mean RTNDTvaluesfor both units at 54 EFPY are bounded by the 64 EFPY Mean RTNDT provided by theNRC. Although a conditional failure probability has not been calculated, the fact that theDresden 54 EFPY values are less than the 64 EFPY value provided by the NRC leads tothe conclusion that the Dresden RPV conditional failure probability is bounded by theNRC analysis.

The procedures and training used to limit cold over-pressure events will be the same asthose approved by the NRC when Dresden requested the BWRVIP-05 technicalalternative be used for the current term (Reference 4.14). An extension of this relief forDresden for the 60-year period will be submitted to the NRC for approval prior to theperiod of extended operation.

Table 4.2.6-1 Effects for Irradiation on RPV Circumferential Weld PropertiesDresden Units 2 & 3

B&W Dresden Unit 2 Dresden Unit 3Group 64 EFPY 54 EFPY 54 EFPY

0.31 0.23 0.34Cu%Ni% 0.59 0.59 0.68CF 196.7 168 221

Fluence at 0.19 0.042 0.041clad/weld interface

(10'9 n/cm2 )

ARTNDTW/O margin 109.4 44 58(OF)

RTNDT(U) (OF) 20 10 -5Mean RTNDT (IF) 129.4 54 53P(FIE) NRC 4.83 x 104 --- _---P(FIE) BWRVIP --- ---

Quad CitiesTable 4.2.6-2 provides a comparison of the limiting RPV circumferential weldparameters for each QCNPS unit to those used in the NRC analysis for the firsttwo key assumptions.

The chemistry composition and chemistry factor for QCNPS Unit 1 are less thanor equal to the limits of the NRC analysis. While the nickel content for Unit 2 ishigher than the value utilized in the NRC analysis, the Unit 2 copper content andthe chemistry factor are considerably lower than the values utilized in the NRCanalysis. Additionally, the unirradiated reference temperatures for both QCNPSunits are lower than the NRC limits. The combination of unirradiated referencetemperature and embrittlement shift yields adjusted reference temperaturesconsiderably lower than the NRC mean analysis values.

LRA Replacement Page 4-25a

The end of life (i.e., 54 effective full power year (EFPY)) inside diameter fluencesfor QCNPS, Units 1 and 2, are also considerably lower than the NRC estimatedfluence. Therefore, the RPVshell weld embrittlement due to fluence has anegligible effect on the probabilities of RPVshell weld failure. Each unit's RPVshell circumferential weld failure probabilities are bounded by the conditionalfailure probability, P(FIE), in the NRC's limiting plant specific analysis (64 EFPY)through the projected additional license operating period.

Table 4.2.6-2 Effects for Irradiation on RPV Circumferential Weld PropertiesQuad Cities Units I & 2

B&W Quad Cities Unit 1 Quad Cities Unit 2roup64 EFPY 54 EFPY 54 EFPY

Cu% 0.31 0.27 0.05

Ni% 0.59 0.59 0.96

CF 196.7 183 68

Fluence atclad/weld 0.19 0.041 0.041interface

(10"9 n/cm2 ) .

ARTNDTW/O margin 109.4 48 18(9F:) 109_ _ _ 4_ _ _48 _ _ _ _ _

RTNDT(U) 20 -5 -32( CF) _ _ _ _ _ _ _ _ _ _ _ _

Mean RTNDT 129.4 43 -14( 'F)_ ___ ___ __

P(F/E) 4.83 x 10 -- --NRC

P(F/E)BWRVIP


4.2.7 Reactor Vessel Axial Weld Failure Probability

ApplicabilityThis section applies to Dresden and Quad Cities.

Summary DescriptionThe Boiling Water Reactor Owner's Group Vessel and Internals Programrecommendations for inspection of reactor vessel shell welds (BWRVIP-05,Reference 4.14) contain generic analyses supporting an NRC SER (Reference 4.15)conclusion that the generic-plant axial weld failure rate is no more than 5 x 104 perreactor year. BWRVIP-05 showed that this axial weld failure rate of 5 x 104 per reactoryear is orders of magnitude greater than the 40-year end-of-life circumferential weldfailure probability, and used this analysis to justify relief from inspection of thecircumferential welds as described in Section 42.6

Dresden received relief from the circumferential weld inspections for the remaining 40year licensed operating period. Quad Cities has also submitted a relief request for theremaining 40-year license operating period.

AnalysisAs stated in Section 4.2.6, Dresden Station received NRC approval for a technicalalternative which eliminated the reactor vessel circumferential shell weld inspections forthe current license term. Quad Cities has applied for a similar relief request. Thebasis for both relief requests was an analysis that satisfied the limiting conditionalfailure probability for the circumferential welds at the expiration of the current license,based on BWRVIP-05 and the extent of neutron embrittlement. The NRC SERassociated with BWRVIP-05 (Reference 4.15) concluded that the reactor vessel failurefrequency due to failure of the limiting axial welds in the BWR fleet at the end of 40years of operation is less than 5 x 104 per reactor year. This failure frequency isdependent upon given assumptions of flaw density, distribution, and location. Thefailure frequency also assumes that "essentially 100%" of the reactor vessel axial weldswill be inspected.

Due to various obstructions within the reactor vessel, Dresden and Quad Cities havenot been able to meet the "essentially 100%" inspection requirement. For Dresden, ananalysis was performed to assess the effect on the probability of fracture due to theactual inspection performed on the vessel axial welds and to determine if the coveragewas sufficient in the inspection of regions contributing to the majority of the risk. Theanalysis included an estimate and comparison of the probability of failure for the casesof uessentially 100%' inspection and the limited scope inspections on the Dresden 2and 3 vessel axial welds. The analysis concluded that the conditional probabilities offailure due to a low temperature over pressurization event are very small, 2.96 x 1O8and 3.15 x 10i10 for Dresden Unit 2 and Unit 3, respectively. The conditional probabilityof failure with the 'essentially 100%" inspections were an order of magnitude lower thanthat for the actual inspection coverage. However, the Dresden analysis only applies tothe current 40-year operating period. The anticipated changes in metallurgicalconditions expected over the extended licensed operating period require an additional


analysis for 54 EFPY and approval by the NRC to extend the reactor vesselcircumferential weld inspection relief request.

Disposition: Revision, 10 CFR 54.21(c)(1)(ii)Table 4.2.7-1 compares the limiting axial weld 54 EFPY properties for Dresden Units 2and 3 against the values taken from Table 2.6-5 found in the NRC SER for BWRVIP-05and associated supplement to the SER (Reference 4.16). The SER supplement requiredthe limiting axial weld to be compared with data found in Table 3 of the document. Table4.2.7-2 compares the limiting axial weld 54 EFPY properties for Quad Cities Units1 and 2 against the values taken from Table 2.6-5 found in the NRC SER. ForDresden and Quad Cities, the comparison was made to the Clinton plant information.The supplemental SER stated that the axial welds for the Clinton plant are the limitingwelds for the BWR fleet and vessel failure probability calculations determined for Clintonshould bound those for the BWR fleet.

The limiting axial welds at both Dresden and Quad Cities are all electroslag welds withsimilar chemistry. The Dresden and Quad Cities limiting weld chemistry, chemistryfactor (CF), and 54 EFPY mean RTNDT values are within the limits of the valuesassumed in the analysis performed by the NRC staff in the March 7, 2000 BWRVIP-05SER supplement and the 64 EFPY limits and values obtained from Table 2.6.5 of theSER.

As stated above, the probability of a failure event PFE calculated by the NRCBWRVIP-05 SER and its supplements depends in part on an assumption that 90 percent of axial welds can be inspected. Less than 90 per cent of axial welds can beexamined at Dresden and Quad Cities. As such, an analysis was performed for 54EFPY to assess the effect on the probability of fracture due to the actual inspectionperformed on the vessel axial welds and to determine if the coverage was sufficient inthe inspection of regions contributing to the majority of the risk. The analysis includedthe estimate and comparison of the probability of failure for both the case of "essentially100%" inspection and the actual limited scope inspections on the Dresden Unit 2 andUnit 3 and Quad Cities Unit 1 and Unit 2 vessel axial welds. The analysis concludedthat the conditional probabilities of failure due to a low temperature over pressurizationevent are very small: 3.89 x 104 and 5.07 x 10.8 for Dresden Unit 2 and Unit 3, and2.08xIOi-and 5.277xIU7 for Quad Cities Unit 1 and Unit 2, respectively. The evaluationshows that the calculated unit-specific axial weld conditional failure probabilities at54 EFPY for Dresden and Quad Cities are less than the failure probabilities calculatedby the NRC staff in the SER at 64 EFPY and the limiting Clinton values found in Table 3of the SER supplement. The probability of failure of an axial weld at Dresden or QuadCities will therefore provide adequate margin above the probability of failure of acircumferential weld, in support of relief from inspection of circumferential welds, for theextended licensed operating period.

LRA Replacement Page 4-28Table 4.2.7-1 Effects for Irradiation on RPV Axial Weld Properties

Dresden Units 2 & 3

Value - B&W SER DRE 2 DRE 3Supplement 54 EFPY 54 EFPY

64 EFPY (Clinton)

Cu% 0.25 0.10 0.24 0.24

Ni% 0.35 1.08 0.37 0.37

CF 142.5 --- 141 141

Fluence x 0.35 0.69 0.057 0.057n/cm

ARTNDT 88.9 121 44 44OF

RTNDT(U) 10 -30 23 23OF

Mean RTNDT 98.9 91 67 67

P(FIE) 1.87 x 10 ' 2.73 x 103 3.89 x 108 5.07 x 10-8

NRC

P(RIE) 1.52 x 10BWRVIP

LRA Replacement Page 4-28a

Table 4.2.7-2, Effects for Irradiation on RPVAxial Weld PropertiesQuad Cities Units 1 & 2

Parameter B&W SER QCNPS Unit I QCNPS Unit 2Description 64 EFPY Supplement 54 EFPY 54 EFPY

(Clinton)Cu % 0.25 0.10 0.24 0.24

Ni % 0.35 1.08 0.37 0.37

CF 142.5 -- 141 141

Fluence, 0.25 0.69 0.057 0.057X 1019 n/cm2

ARTNDT, OF 88.9 121 44 44

ARTNDT(U), OF 10 -30 23 23

Mean RTNDT, OF 98.9 91 67 67P(FIE) NRC 1.87x 10-' 2.73x 10f 2.08x 10-7 5.27x 107

P(FIE) BWR VIP 1.52 x 10"f ... ---


4.3.1 Reactor Vessel Fatigue Analyses

Applicability

This section applies to Dresden and Quad Cities.

Summary Description

Reactor vessel fatigue analyses of the vessel support skirt, shell, upper and lowerheads, closure flanges, nozzles and penetrations, nozzle safe ends, and closure studs,depend on assumed numbers and severity of normal and upset-event pressure andthermal operating cycles to predict end-of-life fatigue usage factors.These assumed cycle counts and fatigue usage factors are based on 40 years ofoperation. Calculation of fatigue usage factors is part of the current licensing basis andis used to support safety determinations. The reactor vessel fatigue analyses areTLAAs.

AnalysisThe original reactor pressure vessel stress report included a fatigue analysis for thereactor vessel components based on a set of design basis duty cycles. These dutycycles are listed in Table 3.9-1 of the Dresden and Quad Cities UFSARs. The original40-year analyses demonstrated that the cumulative usage factors (CUF) for the criticalcomponents would remain below the ASME Code Section III allowable value of 1.0.A reanalysis was performed for reactor vessel cumulative fatigue usage factors as apart of Extended Power Uprate (EPU) implementation at all four Dresden and QuadCities units. (References 4.10 and 4.11) A subset of the bounding reactor vesselcomponents was evaluated as a part of this re-analysis. The resulting fatiguecumulative usage factors (CUFs) for these limiting components supersede the valuesdetermined in the original reactor vessel analyses. The current bounding-case analysis(worst CUFs for all four reactor vessels) lists the following values for the 40-year CUFsfor limiting components.

Shroud Support 0.820Support Skirt 0.862Feedwater Nozzle (Safe End) 0.748Closure Studs < 1.000

The original code analysis of the reactor vessel included fatigue analysis of thefeedwater and control rod drive hydraulic system return line nozzles. After several yearsof operation, it was discovered that both the control rod drive hydraulic system returnline nozzles and the feedwater nozzles were subject to cracking caused by a number offactors including rapid thermal cycling. Consequently, the control rod drive hydraulicsystem return line nozzles were capped and removed from service. As such, they areno longer subject to rapid thermal aging. A reanalysis was later performed on the


extended operation. Dresden and Quad Cities have programs in place to track operatingthermal and pressure cycles and to assess their effect on vessel fatigue. Therequirements from these procedures will be incorporated into the Metal Fatigue ofReactor Coolant Pressure Boundary (B.1.34) aging management program.

Table 4.3.1-1 Fatigue Monitoring Locations for Reactor Pressure VesselComponents

Component Computed Fatigue Computed Fatigue MonitoringUsage Factor Usage Factor Technique

(Pre-EPU) (Post EPU)

Recirculation Outlet Nozzle 0.270 Note 4 CBF

(NUREG/CR-6260component)

Recirculation Inlet Nozzle 0.301 Note 4 CBF


Feedwater Nozzle 0.538 .748 SBF

(NUREGICR-6260component)

Core Spray Nozzle 0.079 Note 4 CBF


Support Skirt 0.940 0.862 SBF

Shroud Support 0.630 0.820 CBF

Closure Stud Bolts 0.79 < 1.0 CBFVessel Shell 0.141 Note 4 CBF


Notes:1. CBF = Cycle Based Fatigue and SBF = Stress Based Fatigue2. EPU = Extended Power Uprate3. The components listed as a "NUREG/CR-6260 component" will be monitored for

GSI -190. See Section 4.3.44. Only locations with 40-year CUF expected to exceed 0.400 have been computed.


4.17 S&L Report EMD-033967 for Quad Cities, Piping Stress Analysis Report, Mark IProgram Plant Unique Analysis of the SRV Discharge Piping System in theSuppression Chamber, Quad Cities Nuclear Power Station, Units 1&2. Revision 0,March 23, 1983.

4.18 S&L Report EMD-036594 for Dresden, Piping Stress Analysis Report, Mark IProgram Plant Unique Analysis of the SRV Discharge Piping System in theSuppression Chamber, Dresden Nuclear Power Station, Units 2&3. Revision 0,February 16,1983.

4.19 Letter from G. A. Abrell (ComEd Nuclear Licensing Administrator) to NRC,"Supplement to Dresden Special Report No. 41 and Quad Cities Special Report No.16, ... " December 8,1975.

4.20 Letter from G. A. Abrell (CoinEd Nuclear Licensing Administrator) to NRC,"Supplement to Dresden Special Report No. 41 and Quad Cities Special Report No.16, ... " February 9, 1976.

4.21 Quad Cities SER 77012701, "Am Nos. 37/35 Modified Crane Handling System,"January 27, 1977.

4.22 Title 10 US Code, Part 50, Section 49 (10 CFR 50.49), "Environmental Qualificationof Electrical Equipment Important to Safety for Nuclear Power Plants."

4.23 DOR Guidelines, 'Guidelines for Evaluating Environmental Qualification of Class 1 EElectrical Equipment in Operating Reactors," U.S. Nuclear Regulatory Commission,June 1979.

4.24 NUREG-0588, 'Interim Staff Position on Environmental Qualification of SafetyRelated Electrical Equipment," U.S. Nuclear Regulatory Commission, July 1981.

4.25 Regulatory Guide 1.89, Revision 1, "Environmental Qualification of Certain ElectricalEquipment Important to Safety for Nuclear Power Plants," U.S. Nuclear RegulatoryCommission, June 1984.

4.26 C.l. Grimes (NRC) letter to D. Walters (NEI), "Guidance on Addressing GSI-168 forLicense Renewal," Project 690, June 1998.

4.27 USNRC Safety Evaluation by the Office of Nuclear Reactor Regulations, Related toAlternative to Inspection of Reactor Pressure Vessel Circumferential Welds, DresdenNuclear Power Station, Units 2 and 3. Attached to NRC letter from Anthony J.Mendiola, Chief Section 2 Project Directorate IlIl to Oliver D. Kingsley, Dresden-Authorization for Proposed Alternative Reactor Pressure Vessel Circumferential WeldExaminations (TAC NOS. MA6228 and MA6229) dated February 25, 2000.

4.28 Letter from Patrick R. Simpson (Exelon) to USNRC, "Corrected Fluence Tablesfor Dresden Nuclear Power Station, Units 2 and 3 License RenewalApplication," dated April 17, 2003

Table of Contents

APPENDIX AUPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR)SUPPLEMENT

UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT ...... A-1

Dresden Units 2 and 3 Updated Final Safety Analysis Supplement ...................................... A-2

A.1 AGING MANAGEMENT PROGRAMS . A-3A.1.1 ASME Section Xi Inservice Inspection, Subsections IWB, IWC,

and IWD .A-3A.1.2 Water Chemistry .A-3A.1.3 Reactor Head Closure Studs .A-3A.1.4 BWR Vessel ID Attachment Welds . A-3A.1.5 BWR Feedwater Nozzle . A-4A.1.6 BWR Control Rod Drive Return Line Nozzle. A4A.1.7 BWR Stress Corrosion Cracking .A-4A.1.8 BWR Penetrations . A-5A.1.9 BWR Vessel Internals .A-5A.1.10 Thermal Aging and Neutron Irradiation Embrittlement of

Cast Austenitic Stainless Steel (CASS) . A-6A.1.11 Flow-Accelerated Corrosion .A-6A.1.12 Bolting Integrity ............ A-7A.1.13 Open-Cycle Cooling Water System .A-7A.1.14 Closed-Cycle Cooling Water System . A-8A.1.15 Inspection of Overhead Heavy Load and Light Load

(Related to Refueling) Handling Systems .A-8A.1.16 Compressed Air Monitoring .A-8A.1.17 BWR Reactor Water Cleanup System . A-8A.1.18 Fire Protection . A-9A.1.19 Fire Water System .A-9A.1.20 Aboveground Carbon Steel Tanks . A-10A.1.21 Fuel Oil Chemistry .A-10A.1.22 Reactor Vessel Surveillance .A-11A.1.23 One-Time Inspection .A-11A.1.24 Selective Leaching of Materials .A-12A.1.25 Buried Piping and Tanks Inspection .A-13A.1.26 ASME Section Xl, Subsection IWE .A-13A.1.27 ASME Section Xl, Subsection IWF .A-13A.1.28 10 CFR Part 50, Appendix J .A-14A.1.29 Masonry Wall Program .A-14A.1.30 Structures Monitoring Program .A-14A.1.31 RG 1.127, Inspection of Water-Control Structures

Associated with Nuclear Power Plants .A-15A.1.32 Protective Coating Monitoring and Maintenance Program .A-15A.1.33 Electrical Cables and Connections Not Subject to 10 CFR 50.49

Environmental Qualification Requirements .A-15A.1.34 Metal Fatigue of Reactor Coolant Pressure Boundary .A-16A.1.35 Environmental Qualification (EQ) of Electrical Components .A-16A.1.36 Not Used .A-16A.1.37 Electrical Cables Not Subject to 10 CFR 50.49 Environmental Qualification

Requirements used in Instrument Circuits .A-16A.1.38 Inaccessible Medium-Voltage Cables Not Subject to 10 CFR 50.49

Environmental Qualification Requirements .A-17

Dresden and Quad Cities Page A-iLicense Renewal Application

Table of Contents

A.2 PLANT-SPECIFIC AGING MANAGEMENT PROGRAMS ................................ A-18A.2.1 Corrective Action Program ............................................. A-18A.2.2 Periodic Inspection of Non-EQ, Non-Segregated Electrical

Bus Ducts ....................................... A-18A.2.3 Periodic Inspection of Ventilation System Elastomers ...................................... A-1 9A.2.4 Periodic Testing of Drywell and Torus Spray Nozzles ....................................... A-19A.2.5 Lubricating Oil Monitoring Activities ....................................... A-1 9A.2.6 Heat Exchanger Test and Inspection Activities .. ....................................... A-20A.2.7 Not Used ....................................... A-20A.2.8 Periodic Inspection of Plant Heating System ....................................... A-20A.2.9 Periodic Inspection of Components Subject to Moist Air Environments. A-21

A.3 TIME-LIMITED AGING ANALYSIS SUMMARIES ........................................... A-22A.3.1 Neutron Embrittlement of the Reactor Vessel and Internals ............................. A-22A.3.1.1 Reactor Vessel Materials Upper-Shelf Energy Reduction Due to

Neutron Embrittlement . A-22A.3.1.2 Adjusted Reference Temperature for Reactor Vessel Materials

Due to Neutron Embrittlement ........................................ A-22A.3.1.3 Reflood Thermal Shock Analysis of the Reactor Vessel ................................... A-22A.3.1.4 Reflood Thermal Shock Analysis of the Reactor Vessel Core Shroud and Repair

Hardware . A-22A.3.1.5 Reactor Vessel Thermal Limit Analyses: Operating Pressure - Temperature

Limits ......................................... A-23A.3.1.6 Reactor Vessel Circumferential Weld Examination Relief ................................ A-23A.3.1.7 Reactor Vessel Axial Weld Failure Probability ......................................... A-23

A.3.2 Metal Fatigue ........................................... A-24A.3.2.1 Reactor Vessel Fatigue Analyses ......................................... A-24A.3.2.2 Fatigue Analysis of Reactor Vessel Internals .. ......................................... A-24A.3.2.2.1 High-Cycle Flow-Induced Vibration Fatigue Analysis of Jet Pump Riser

Braces ... A-24A.3.2.3 Reactor Coolant Pressure Boundary Piping and Component Fatigue

Analysis . A-24A.3.2.3.1 ASME Section III Class 1 Reactor Coolant Pressure Boundary Piping and

Component Fatigue Analysis . A-25A.3.2.3.2 Reactor Coolant Pressure Boundary Piping and Components Designed to

USAS B31.1, ASME Section III Class 2 and 3, or ASME Section VilIClass B and C ............................... A-25

A.3.2.3.3 Fatigue Analysis of the Isolation Condenser ............................... A-26A.3.2.4 Effects of Reactor Coolant Environment on Fatigue Life of

Components and Piping (Generic Safety Issue 190) ...................................... A-26

A.3.3 Environmental Qualification Of Electrical Equipment ...................................... A-26A.3.4 Containment Fatigue ...................................... A-27A.3.4.1 Fatigue Analysis of the Suppression Chamber, Vents, and

Downcomers . A-27A.3.4.2 Fatigue Analysis of SRV Discharge Piping Inside the Suppression Chamber,

External Suppression Chamber Attached Piping, and AssociatedPenetrations ....... A-27

A.3.4.3 Drywell to Suppression Chamber Vent Line Bellows Fatigue Analyses ............ A-27A.3.4.4 Primary Containment Process Penetrations Bellows Fatigue Analysis ............. A-28

A.3.5 Other Plant-Specific TLAAs ................................................... A-28A.3.5.1 Reactor Building Crane Load Cycles ................................................... A-28A.3.5.2 Metal Corrosion Allowances ................................................... A-28A.3.5.2.1 Degradation Rates of Inaccessible Exterior Drywell Plate Surfaces ................. A-28

Dresden and Quad Cities Page A-iiLicense Renewal Application

Table of Contents

A.3.5.2.2 Galvanic Corrosion in the Containment Shell and Attached PipingComponents due to Stainless Steel ECCS Suction Strainers .......................... A-29

A.3.5.3 Crack Growth Calculation of a Postulated Flaw in the HeatAffected Zone of an Arc Strike in the Suppression Chamber Shell .................. A-29

A.3.5.4 Radiation Degradation of Drywell Shell Expansion GapPolyurethane Foam .A-30

A.3.6 References for Section A.3 .A-31

Quad Cities Units I and 2 Updated Final Safety Analysis Supplement ............................... A-32

A.1 AGING MANAGEMENT PROGRAMS .A-33A.1.1 ASME Section Xl Inservice Inspection, Subsections IWB, IWC,

and IWD .A-33A.1.2 Water Chemistry .A-33A.1.3 Reactor Head Closure Studs .A-33A.1.4 BWR Vessel ID Attachment Welds . A-33A.1.5 BWR Feedwater Nozzle .A-34A.1.6 BWR Control Rod Drive Return Line Nozzle .A-34A.1.7 BWR Stress Corrosion Cracking .A-34A.1.8 BWR Penetrations . A-35A.1.9 BWR Vessel Internals .A-35A.1.10 Thermal Aging and Neutron Irradiation Embrittlement of Cast

Austenitic Stainless Steel (CASS) .A-36A.1.11 Flow-Accelerated Corrosion .A-36A.1.12 Bolting Integrity . A-37A.1.13 Open-Cycle Cooling Water System . A-37A.1.14 Closed-Cycle Cooling Water System . A-38A.1.15 Inspection of Overhead Heavy Load and Light Load (Related

to Refueling) Handling Systems .A-38A.1.16 Compressed Air Monitoring .A-38A.1.17 BWR Reactor Water Cleanup System . A-39A.1.18 Fire Protection . A-39A.1.19 Fire Water System .A-39A.1.20 Aboveground Carbon Steel Tanks . A-40A.1.21 Fuel Oil Chemistry. A-40A.1.22 Reactor Vessel Surveillance. A-41A.1.23 One-Time Inspection .A-41A.1.24 Selective Leaching of Materials. A-42A.1.25 Buried Piping and Tanks Inspection. A-43A.1.26 ASME Section Xl, Subsection IWE .A-43A.1.27 ASME Section Xl, Subsection IWF ........................ ; A-43A.1.28 10 CFR Part 50, Appendix J. A-44A.1.29 Masonry Wall Program ......................... A-44A.1.30 Structures Monitoring Program. A-44A.1.31 RG 1.127, Inspection of Water-Control Structures Associated

with Nuclear Power Plants .................. A-45A.1.32 Protective Coating Monitoring and Maintenance Program. A-45A.1.33 Electrical Cables and Connections Not Subject to 10 CFR 50.49

Environmental Qualification Requirements. A-46A.1.34 Metal Fatigue of Reactor Coolant Pressure Boundary. A-46A.1.35 Environmental Qualification (EQ) of Electrical Components. A-46A.1.36 Boraflex Monitoring. A-46A.1.37 Electrical Cables Not Subject to 10 CFR 50.49 Environmental

Qualification Requirements used in Instrument Circuits .A-47

Dresden and Quad Cities Page A-iiiLicense Renewal Application

Table of Contents

A.2 PLANT-SPECIFIC AGING MANAGEMENT PROGRAMS ................................ A-48A.2.1 Corrective Action Program ............................................. A-48A.2.2 Periodic Inspection of Non-EQ, Non-Segregated Electrical

Bus Ducts ....................................... A-48A.2.3 Periodic Inspection of Ventilation System Elastomers ...................................... A-49A.2.4 Periodic Testing of Drywell and Torus Spray Nozzles ....................................... A-49A.2.5 Lubricating Oil Monitoring Activities ....................................... A-49A.2.6 Heat Exchanger Test and Inspection Activities .. ....................................... A-50A.2.7 Generator Stator Water Chemistry Activities ....................................... A-50A.2.8 Periodic Inspection of Plant Heating System ....................................... A-51A.2.9 Periodic Inspection of Components Subject to Moist Air Environments. A-51

A.3 Time-Limited Aging Analysis Summaries ........................................... A-52A.3.1 Neutron Embrittlement of the Reactor Vessel and Internals ............................. A-52A.3.1.1 Reactor Vessel Materials Upper-Shelf Energy Reduction Due

to Neutron Embrittlement . A-52A.3.1.2 Adjusted Reference Temperature for Reactor Vessel

Materials Due to Neutron Embrittlement ........................................ A-52A.3.1.3 Reflood Thermal Shock Analysis of the Reactor Vessel ................................... A-52A.3.1.4 Reflood Thermal Shock Analysis of the Reactor Vessel

Core Shroud and Repair Hardware . A-52A.3.1.5 Reactor Vessel Thermal Limit Analyses: Operating Pressure - Temperature

Limits .. A-53A.3.1.6 Reactor Vessel Circumferential Weld Examination Relief ......................... A-53A.3.1.7 Reactor Vessel Axial Weld Failure Probability ............................................ A-53

A.3.2 Metal Fatigue .............................................. A-54A.3.2.1 Reactor Vessel Fatigue .............................................. A-54A.3.2.2 Fatigue Analysis of Reactor Vessel Internals .............................................. A-54A.3.2.2.1 Low-Cycle Thermal Fatigue Analysis of the Core Shroud and Repair

Hardware . A-54A.3.2.3 Reactor Coolant Pressure Boundary Piping and Component Fatigue

Analysis ... A-55A.3.2.3.1 Reactor Coolant Pressure Boundary Piping and Components Designed to

USAS B31.1, ASME Section III Class 2 and 3, or ASME Section VilIClass B and C . A-55

A.3.2.4 Effects of Reactor Coolant Environment on Fatigue Life ofComponents and Piping (Generic Safety Issue 190) ..................................... A-55

A.3.3 Environmental Qualification Of Electrical Equipment ..................................... A-56

A.3.4 Containment Fatigue ..................................... A-56A.3.4.1 Fatigue Analysis of the Suppression Chamber, Vents, and

Downcomers . A-56A.3.4.2 Fatigue Analysis of SRV Discharge Piping Inside the Suppression

Chamber, External Suppression Chamber Attached Piping, andAssociated Penetrations . A-56

A.3.4.3 Drywell to Suppression Chamber Vent Line BellowsFatigue Analyses . A-57

A.3.4.4 Primary Containment Process Penetrations BellowsFatigue Analysis ........................ A-57

A.3.5 Other Plant-Specific TLAAs ........................ A-57A.3.5.1 Reactor Building Crane Load Cycles ........................ A-57A.3.5.2 Metal Corrosion Allowances ........................ A-57A.3.5.2.1 Corrosion Allowance for Power Operated Relief Valves ................................... A-57A.3.5.2.2 Degradation Rates of Inaccessible Exterior Drywell Plate Surfaces ................. A-58

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A.3.5.2.3

A.3.5.3

A.3.5.4

A.3.6

Galvanic Corrosion in the Containment Shell and Attached PipingComponents due to Stainless Steel ECCS Suction Strainers .......................... A-58Crack Growth Calculation of a Postulated Flaw in the HeatAffected Zone of an Arc Strike in the Suppression Chamber Shell .................. A-59Radiation Degradation of Drywell Shell Expansion GapPolyurethane Foam ................................................ A-59

References for Section A.3 ................................................ A-60

Dresden and Quad CitiesLicense Renewal Application

Page A-v

Appendix AUPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT

APPENDIX A UPDATED FINAL SAFETY ANALYSIS REPORT(UFSAR) SUPPLEMENT

IntroductionThe summary descriptions of aging management program activities presented in thisAppendix A represent commitments for managing aging of the systems, structures andcomponents within the scope of license renewal during the period of extendedoperation. This appendix also provides summary descriptions of time-limited aginganalyses. These summary descriptions of aging management program activities andtime-limited aging analyses will be incorporated in the Updated Final Safety AnalysisReports for the Dresden Nuclear Power Station and the Quad Cities Nuclear PowerStation following issuance of the renewed operating license.

A separate Appendix A Updated Final Safety Analysis Supplement has been provided orDresden, Units 2 and 3 and for Quad Cities, Units 1 and 2.

Dresden and Quad CitiesI License Renewal Application

Page A-1

Appendix ADresden, Units 2 and 3

UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT

Dresden Units 2 and 3

Updated Final Safety Analysis Supplement

Dresden and Quad CitiesLicense Renewal Application

Page A-2



A.1 AGING MANAGEMENT PROGRAMS

A.1.1 ASME Section Xi Inservice Inspection, Subsections IWB, IWC, and IWD

The ASME Section Xi Inservice Inspection, Subsections IWB, IWC and IWD agingmanagement program consists of periodic volumetric and visual examinations ofcomponents for assessment, identification of signs of degradation, and establishment ofcorrective actions. Prior to the period of extended operation the program will be revised-to be consistent with ASME Section XI, 1995 Edition through the 1996 Addenda. Theinspections will be implemented in accordance with 10 CFR 50.55(a).

Dresden will implement the guidance of BWRVIP-74, "BWR Reactor PressureVessel Inspection and Flaw Evaluation Guidelines," with the following exception.Exception: Risk Informed Inservice Inspection is implemented in lieu of ASMESection Xl requirements for portions of Class 1 and Class 2 systems. TechnicalSpecification revisions containing new P-T Curves will be submitted prior to theterm of extended operation.

A.1.2 Water Chemistry

The water chemistry aging management program consists of monitoring and control ofwater chemistry to keep peak levels of various contaminants below system-specificlimits based on industry-recognized guidelines of EPRI TR-103515, UBWR WaterChemistry Guidelines." To mitigate aging effects on component surfaces that areexposed to water as process fluid, the chemistry programs are used to control waterchemistry for impurities (e.g., chlorides, and sulfates) that accelerate corrosion.

Dresden will implement the general guidance provided in BWRVIP-79, "EPRIReport TR-103515-R2."

A.1.3 Reactor Head Closure Studs

The reactor head closure studs aging management program includes inserviceinspection (ISI). This program also includes preventive actions and inspectiontechniques for BWRs. Prior to the period of extended operation the program will be-revised to be consistent with ASME Section Xi, 1995 Edition through the 1996 Addenda.The requirements of ASME Section Xl will be implemented in accordance with 10CFR 50.55(a). The reactor head studs are not metal-plated, and have had manganesephosphate coatings applied.

A.1.4 BWR Vessel ID Attachment Welds

The BWR vessel ID attachment welds aging management program includes(a) inspection and flaw evaluation in conformance with the guidelines of staff-approvedBoiling Water Reactor Vessel and Internals Project BWRVIP-48, 'Vessel ID AttachmentWeld Inspection and Evaluation Guidelines," and/or ASME Section Xl; and(b) monitoring and control of reactor coolant water chemistry in accordance withindustry-recognized guidelines of EPRI TR-103515, 'BWR Water Chemistry

Dresden and Quad Cities Page A-3License Renewal Application



Guidelines." Prior to the period of extended operation the program will be revised to beconsistent with ASME Section Xl, 1995 Edition through the 1996 Addenda. Therequirements of ASME Section Xl will be implemented in accordance with 10 CFR50.55(a).

A.1.5 BWR Feedwater Nozzle

The BWR feedwater nozzle aging management program includes enhancing theinservice inspections (ISI) specified in the ASME Code, Section Xi, with therecommendation of General Electric (GE) NE-523-A71-0594, "Alternate BWRFeedwater Nozzle Inspection Requirements," to perform periodic ultrasonic testinginspection of critical regions of the BWR feedwater nozzles.

A.1.6 BWR Control Rod Drive Return Line Nozzle

The BWR control rod drive return line nozzle aging management program consists ofpreviously implemented system modifications and inservice inspections that manage theaging effect of cracking in the control rod drive return line nozzles. The control rod drivereturn line nozzles have been capped. Inservice inspections are performed consistentwith ASME Section Xl requirements. No augmented inspections in accordance withNUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line NozzleCracking," or the alternative recommendations of GE NE-523-A71-0594, 'AlternateBWR Feedwater Nozzle Inspection Requirements," are required. Prior to the period ofextended operation the program will be revised to be consistent with ASME Section XI,1995 Edition through the 1096 Addenda. The requirements of ASME Section Xl willbe implemented in accordance with 10 CFR 50.55(a).

A.1.7 BWR Stress Corrosion Cracking

The BWR stress corrosion cracking aging management program to manageintergranular stress corrosion cracking (IGSCC) in boiling water reactor coolant pressureboundary piping four inches and larger nominal pipe size made of stainless steel (SS) isdelineated, in part, in NUREG-0313, "Technical Report on Material Selection andProcessing Guidelines for BWR Coolant Pressure Boundary Piping," Revision 2,BWRVIP 75, 'Technical Basis for Revisions to Generic Letter 88-01 InspectionSchedules," and Nuclear Regulatory Commission (NRC) Generic Letter (GL) 88-01,"NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR AusteniticStainless Steel Piping," and its Supplement 1. The program includes (a) replacementsand preventive measures to mitigate IGSCC and (b) inspections to monitor IGSCC andits effects. Water chemistry is monitored and maintained in accordance with industry-recognized guidelines in EPRI TR-103515, "BWR Water Chemistry Guidelines." P-ior-tOthe period of extended operation the program will be revised to be consistent with ASMESection XI, 1995 Edition through the 1996 Addenda. The requirements of ASMESection Xl will be implemented in accordance with 10 CFR 50.55(a).

Dresden will implement the general guidance provided in BWRVIP-75, "TechnicalBasis for Revisions to Generic Letter 88-01 Inspection Schedules, " with Exception




- The Relief Request submitted for the implementation of RISI indicates theCategory A Welds are "subsumed into the RISI program."

A.1.8 BWR Penetrations

The BWR penetrations aging management program includes (a) inspection and flawevaluation in conformance with the guidelines of staff-approved Boiling Water ReactorVessel and Internals Project (BWRVIP)-49, "Instrument Penetration Inspection and FlawEvaluation Guidelines," and BWRVIP-27, 'BWR Standby Liquid Control System/CorePlate Delta-P Inspection and Flaw Evaluation Guidelines," documents and(b) monitoring and control of reactor coolant water chemistry in accordance withindustry-recognized guidelines of EPRI TR-103515, "BWR Water ChemistryGuidelines," to ensure the long-term integrity and safe operation of boiling water reactorvessel internal components. Prior to the period of extended operation the program willbe revised to be consistent with ASME Section Xl, 1895 Edition through the 1996Addenda-The requirements of ASME Section Xl will be implemented inaccordance with 10 CFR 50.55(a).

Dresden will implement the guidance provided in BWRVIP-27, "BWR StandbyLiquid Control System/Core Plate AP Inspection and Flaw Evaluation Guidelines,"with the following exception. Dresden has implemented approved iSI ReliefRequests, which provide a VT-2 examination of the standby liquid control systemnozzle inner radius in lieu of the Code required volumetric examination.

Dresden will implement the guidance provided in BWRVIP-49, "InstrumentPenetration Inspection and Flaw Evaluation Guidelines."

A.1.9 BWR Vessel Internals

The BWR vessel internals aging management program includes (a) inspection and flawevaluation in conformance with the guidelines of applicable and staff-approved BoilingWater Reactor Vessel and Internals Project (BWRVIP) documents, and with ASMESection Xl; and (b) monitoring and control of reactor coolant water chemistry inaccordance with industry-recognized guidelines of EPRI TR-103515, UBWR WaterChemistry Guidelines," to ensure the long-term integrity and safe operation of boilingwater reactor vessel internal components. Prior to the period of extended operation thepFrogramn will be revised to be conRsitent with ASME Section Xl, 1995 Edition through the18996 Addenda. The requirements of ASME Section Xl will be implemented inaccordance with 10 CFR 50.55(a).

Dresden will implement the general guidance provided in BWRVIP-18, "BWR CoreSpray Internals Inspection and Flaw Evaluation Guidelines."

Dresden will implement the general guidance provided in BWRVIP-25, "BWR CorePlate Inspection and Flaw Evaluation Guidelines."

Dresden will implement the guidance provided in BWRVIP-26, "BWR Top GuideInspection and Flaw Evaluation Guidelines." Additionally, Dresden will perform




augmented inspections for the top guide similar to the inspections of control roddrive housing (CRDH) guide tubes.

Dresden will implement the general guidance provided in BWRVIP-38, "BWRShroud Support Inspection an Flaw Evaluation Guidelines." Dresden will performthe additional inspections in the lower plenum (i.e. shroud support leg welds)when new inspection techniques and tooling are developed, incorporated into theapplicable BWRVIP document(s), and approved by NRC SER.

Dresden will implement the general guidance provided in BWRVIP-41, "BWR JetPump Assembly Inspection and Flaw Evaluation Guidelines." Dresden willperform the additional inspections of the inaccessible thermal sleeve welds whennew inspection techniques and tooling are developed, incorporated into theapplicable BWRVIP document(s), and approved by NRC SER.

Dresden will implement the general guidance provided in BWRVIP-47, "BWRLower Plenum Inspection and Flaw Evaluation Guidelines."

Dresden will implement the general guidance provided in BWRVIP-76, "BWR CoreShroud Inspection and Flaw Evaluation Guidelines."

Dresden will implement the general guidance provided in BWRVIP-104,"Evaluation and Recommendations to Address Shroud Support Cracking inBWRs. "

A.1.10 Thermal Aging and Neutron Irradiation Embrittlement of Cast AusteniticStainless Steel (CASS)

The thermal aging and neutron irradiation embrittlement of cast austenitic stainless steel(CASS) aging management program consists of (1) determination of the susceptibility ofcast austenitic stainless steel components to thermal aging embrittlement, (2)accounting for the synergistic effects of thermal aging and neutron irradiation, and (3)implementing a supplemental examination program, as necessary. The program isbeing implemented prior to the period of extended operation.

A.1.11 Flow-Accelerated Corrosion

The flow-accelerated corrosion aging management program consists of (1) appropriateanalysis and baseline inspections, (2) determination of the extent of thinning, andreplacement or repair of components, and (3) follow-up inspections to confirm orquantify effects, and to take longer-term corrective actions. This program is in responseto NRC Generic Letter 89-08, "Erosion/Corrosion-Induced Pipe Wall Thinning." Theprogram relies on implementation of the EPRI NSAC-202L, "Recommendations for anEffective Flow Accelerated Corrosion Program," Revision 2 guidelines. Prior to theperiod of extended operation the program will be revised to include main steam pipingwithin the scope of license renewal.




A.1.12 Bolting Integrity

This bolting integrity aging management program incorporates industryrecommendations of EPRI NP-5769, 'Degradation and Failure of Bolting in NuclearPower Plants," and includes periodic visual inspections for external surface degradationthat may be caused by loss of material or cracking of the bolting, or by an adverseenvironment. Inspection of inservice inspection Class 1, 2, and 3 components isconducted in accordance with ASME Section Xl. Prior to the period of extendedoperation the program will be revised to be consistent with A.SME Section Xl, 1995Edition through the 1996 Addenda. The requirements of ASME Section XI will beimplemented in accordance with 10 CFR 50.55(a). The program will also includeinspections of bolted joints of diesel generator system components and of componentsin locations containing high humidity or moisture. In addition, the program willinclude inspections of the reactor vessel-to-ring girder bolting.

Program activities address the guidance contained in EPRI TR-104213, 'Bolted JointMaintenance and Applications Guide," but do not specifically identify its use.' Non-safetycomponent inspections rely on detection of visible leakage during preventivemaintenance and routine observation. The program does not address structural andcomponent support bolting with the exception of the reactor vessel-to-ring girderbolting. The aging management of all other structural bolting is covered by thestructures monitoring program. Aging management of ASME Section Xl Class 1, 2,and 3 and Class MC support members, including mechanical connections is covered bythe "ASME Section Xl, Subsection IWF" aging management program.

A.1.13 Open-Cycle Cooling Water System

The open-cycle cooling water system aging management program includes(a) surveillance and control of biofouling, (b) tests to verify heat transfer, (c) a routineinspection and maintenance program, including system flushing and chemical treatment,(d) periodic inspections for leakage, loss of material, and blockage, (e) engineeringevaluations and heat sink performance assessments, and (f) assessments of the overallheat sink program. These evaluations and assessments produced specific componentand programmatic corrective actions. The program provides assurance that the open-.cycle cooling water system is in compliance with General Design Criteria, and withquality assurance requirements, to ensure that the open-cycle cooling water system canbe managed for an extended period of operation. This program is in response to anduses the test and inspection guidelines of NRC Generic Letter 89-13, "Service WaterSystem Problems Affecting Safety-Related Equipment." Prior to the period of extendedoperation, the scope of the program will be increased to include inspection of anadditional strainer, additional heat exchangers and sub-components, external surfacesof various submerged pumps and piping, cooling water pump linings, and componentsin the pump vaults that have a high humidity or moisture environment.




A.1.14 Closed-Cycle Cooling Water System

The closed-cycle cooling water system aging management program relies on preventivemeasures to minimize corrosion by maintaining inhibitors and by performing non-chemistry monitoring consisting of inspection and nondestructive examinations (NDEs)based on industry-recognized guidelines of EPRI TR-107396, "Closed Cooling WaterChemistry Guidelines," for closed-cycle cooling water systems. Station maintenanceinspections and NDE provide condition monitoring of heat exchangers exposed toclosed-cycle cooling water environments. Prior to the period of extended operation, theprogram will be enhanced to include procedure revisions that provide for monitoring ofspecific chemistry parameters in order to meet EPRI TR-1 07396 guidance.

A.1.15 Inspection of Overhead Heavy Load and Light Load (Related to Refueling)Handling Systems

The inspection of overhead heavy load and light load (related to refueling) handlingsystems aging management program confirms the effectiveness of the maintenancemonitoring program and the effects of past and future usage on the structural reliabilityof cranes and hoists. Administrative controls ensure that only allowable loads arehandled, and fatigue failure of structural elements is not expected. A time-limited aginganalysis concludes that there are no fatigue concerns for reactor building overheadcranes during the period of extended operation. The bridge, trolley, and other structuralcomponents are visually inspected on a routine basis for degradation. These cranes areincluded in the corporate structural monitoring program (which complies with the10 CFR 50.65 maintenance rule) and in various station procedures. Prior to the periodof extended operation, the program will be enhanced to include inspections for rail wearand proper crane travel on rails, and corrosion of crane structural components.

A.1.16 Compressed Air Monitoring

The compressed air monitoring aging management program consists of inspection,monitoring, and testing of the entire system, including (1) pressure decay testing, visualinspections, and walkdowns of various system locations; and (2) preventive monitoringthat checks air quality at various locations in the system to ensure that dewpoint,particulates, and suspended hydrocarbons are kept within the specified limits. Thisprogram is consistent with responses to NRC Generic Letter 88-14, "Instrument AirSupply Problems," and ANSI/ISA-S7.3-1975, "Quality Standard for Instrument Air."Prior to the period of extended operation, the program will be enhanced to includeinspections of instrument air distribution piping based on EPRI TR-108147,'Compressor and Instrument Air System Maintenance Guide," and blowdown ofinstrument air distribution piping.

A.1.17 BWR Reactor Water Cleanup System

The BWR reactor water cleanup (RWCU) system aging management program monitorsand controls reactor water chemistry based on industry-recognized guidelines of EPRI




TR-1 03515, UBWR Water Chemistry Guidelines," to reduce the susceptibility of RWCUpiping to stress corrosion cracking (SCC) and intergranular stress corrosion cracking(IGSCC). RWCU system piping has been replaced with piping that is resistant tointergranular stress corrosion cracking, in response to NRC Generic Letter 88-01, "NRCPosition on Intergranular Stress Corrosion Cracking (IGSCC) in BWR AusteniticStainless Steel Piping," concerns. In addition, all actions requested in NRC GenericLetter 89-10, 'Safety-Related Motor-Operated Valve Testing and Surveillance,' havebeen completed. Therefore, inservice inspection in accordance with ASME Section Xl isnot required.

A.1.18 Fire Protection

The fire protection aging management program includes a fire barrier inspectionprogram and a diesel-driven fire pump inspection program. The fire barrier inspectionprogram requires periodic visual inspection of fire barrier penetration seals; and firewraps and fire proofing; fire barrier walls, ceilings, and floors; flood barrier penetrationseals that also serve as fire barrier seals; and periodic visual inspection and functionaltests of fire rated doors to ensure that their operability is maintained. The programincludes surveillance tests of fuel oil systems for the diesel-driven fire pumps andisolation condenser diesel-driven makeup pumps to ensure that the fuel supply lines canperform intended functions. The program also includes visual inspections and periodicoperability tests of halon and carbon dioxide fire suppression systems based on NFPAcodes.

Prior to the period of extended operation, the program will be revised to include:

* Inspection of oil spill barriers

* Inspection of external surfaces of the halon system and the carbon dioxidesystem

* Periodic capacity tests of the isolation condenser makeup pumps

* Specific fuel supply leak inspection criteria for fire pumps and isolationcondenser makeup pumps during tests

* Specific inspection criteria for fire doors

* Inspection frequencies for fire doors and spill barriers

A.1.1 9 Fire Water System

The fire water system aging management program provides fire system header andhydrant flushing, system performance (flow and pressure) testing, and inspections, on aperiodic basis, and for injection of chemical agents during or subsequent to flushing tominimize biofouling. System performance tests measure hydraulic resistance andcompare results with previous testing. This approach eliminates the need for tests atmaximum design flow and pressure. Internal inspections are conducted on system




components when disassembled to identify evidence of corrosion or biofouling. Fireheader pressure is maintained through a crosstie with the service water system.Significant leakage (exceeding the capacity of this line) would be identified by automaticstart of the fire pumps, which would initiate immediate investigation and correctiveaction. Inspection and surveillance testing is performed in accordance with proceduresbased on applicable NFPA codes. Where code deviations are required or desirable, theintent of the code is maintained by documented technical justifications.

Sprinkler test requirements will be modified prior to the period of extended operation toinclude sprinkler sampling in accordance with NFPA 25, Inspection, Testing andMaintenance of Water-Based Fire Protection Systems," Section 2-3.1. Samples will besubmitted to a testing laboratory prior to being in service 50 years. This testing will berepeated at intervals not exceeding 10 years.

Prior to the period of extended operation the program will be revised to include externalsurface inspections of submerged fire pumps, outdoor hydrants, and outdoortransformer deluge systems; and periodic non-intrusive wall thickness measurements ofselected portions of the fire water system at intervals that do not exceed every 10 years.

A.1.20 Aboveground Carbon Steel Tanks

The aboveground carbon steel tanks aging management program manages corrosion ofoutdoor nitrogen tanks and aluminum storage tanks. Paint is a corrosion preventivemeasure, and periodic visual inspections monitor degradation of the paint and anyresulting metal degradation. Carbon steel tanks in the scope of license renewal areabove ground and not directly supported by earthen or concrete foundations. Therefore,inspection of the sealant or caulking at the tank-foundation interface, and inspection ofinaccessible tank locations and on-grade tank bottoms do not apply.

Aluminum storage tanks included within the scope of license renewal aresupported by earthen/concrete foundations. Sealants at the tank-foundationinterfaces for these tanks are periodically inspected for degradation. Periodicinternal/external inspections of the aluminum tanks for pitting and crevicecorrosion will be performed at a frequency not to exceed once every five years.UT wall thickness inspections will be performed on the tank bottoms of allaluminum tanks included within the scope of license renewal at a frequency notto exceed once every 10 years.

Prior to the period of extended operation the program will be revised to includedocumentation of results of periodic system engineer walkdowns of the nitrogen tanksand periodic visual and ultrasonic inspections of the internal/external surfaces ofthe aluminum storage tanks.

A.1.21 Fuel Oil Chemistry

The fuel oil chemistry aging management program relies on a combination ofsurveillance and maintenance procedures. Monitoring and controlling fuel oilcontamination maintains the fuel oil quality. Exposure to fuel oil contaminants such as




water and microbiological organisms is minimized by routine draining and cleaning offuel oil tanks, and by fuel oil sampling and analysis, including analysis of new fuel beforeits introduction into the storage tanks. A biocide is added to the fuel oil storage tanksduring each new fuel

exe krn.m imarch 05, 2004 u. s. nuclear regulatory commission page 2 should you have any questions,...

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