View source for RS-04-039, Amendment to the Application for Renewed Operating Licenses for Dresden and Quad Cities Nuclear Power Stations

{{Adams
| number = ML040711186
| issue date = 03/05/2004
| title = Amendment to the Application for Renewed Operating Licenses for Dresden and Quad Cities Nuclear Power Stations
| author name = Simpson P
| author affiliation = Exelon Generation Co, LLC, Exelon Nuclear
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000237, 05000249, 05000254, 05000265
| license number = DPR-019, DPR-025, DPR-029, DPR-030
| contact person = 
| case reference number = RS-04-039
| document type = Letter type:RS
| page count = 103
}}

=Text=
{{#Wiki_filter:Exe krn.M I Exelon Generation 4300 Winfield Road www.exeloncorp.coM Nuclear Warrenville, IL 60555 10 CFR 54 RS-04-039 March 05, 2004 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001 Dresden Nuclear Power Station, Units 2 and 3 Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 Quad Cities Nuclear Power Station, Units 1 and 2 Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

==Subject:==
Amendment to the Application for Renewed Operating Licenses for Dresden and Quad Cities Nuclear Power Stations

==Reference:==
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 the Dresden Nuclear Power Station (DNPS), Units 2 and 3 and Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. Exelon Generation Company, LLC (EGC) is submitting the annual amendment to the DNPS and QCNPS License Renewal Application (LRA) in accordance with 10 CFR 54.21(b). This amendment identifies changes to the current licensing basis (CLB) that materially affect the contents of the DNPS and QCNPS LRA, including the UFSAR supplement. This amendment is required to be submitted each year following submittal of the LRA and at least 3 months before scheduled completion of the LRA review by the NRC.
Attachment 1 provides a description of the changes and replacement pages of the DNPS and QCNPS LRA that have been materially affected by changes to the CLB.
f\Ci1 9o3

March 05, 2004 U. S. Nuclear Regulatory Commission Page 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 q                                                    I' Executed on                           Patrick R. Simpson Manager - Licensing : Amendment to the Application for Renewed Operating Licenses for Dresden and Quad Cities Nuclear Power Stations cc:     Regional Administrator- NRC Region III NRC Senior Resident Inspector - Dresden Nuclear Power Station NRC Senior Resident Inspector - Quad Cities Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1 Amendment to the Application for Renewed Operating Licenses for Dresden and Quad Cities Nuclear Power Stations

==1.0 INTRODUCTION==

The License Renewal Rule, 10CFR54.21 (b), requires that each year following submittal of a license renewal application (LRA), an amendment must be submitted to identify changes to the facility current licensing basis (CLB) that materially impact the content of the LRA. In accordance with this requirement, Exelon Generation Company, LLC (Exelon) has completed the review of Dresden Nuclear Power Station (DNPS), Units 2 and 3 and Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, CLB changes since the submittal of the LRA.
Additionally, Exelon has identified some areas that required correction. This attachment provides 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 the LRA. The CLB changes resulted from both plant changes and analysis changes. Additionally, the UFSAR Supplement and Aging Management Program Descriptions were updated. The review 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 in the 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 the Shroud Support Structure This additional aging management reference was needed to 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 Groups Requiring Aging Management Review - Reactor Internals has been updated to add Reference 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 this TLAA is documented in Section 4.3.2. (This item was a correction.)
* High Pressure Coolant Iniection Pump Material Change at Dresden The HPCI cast iron pumps 2-2320-GSLO and 3-2320-GSLO were replaced with stainless steel pumps. 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 other valid 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 LRA Table 2.3.2-1, Component Groups Requiring Aging Management Review - High Pressure Coolant Injection System, has been revised to add LRA References 3.2.1.13 and 3.2.2.22. (This was a CLB change.)
* Installation of Tie-ins to Isolation Condenser Makeup Pump Suction to Allow for Temp Make-Up Capabilities at Dresden This installation of stainless steel valves in the isolation condenser makeup pump suction to allow temporary makeup capabilities. The LRA already contained Reference 3.2.2.22 (external) for the new stainless steel valves, which is included in LRA Table 2.3.2-5, Component Groups Requiring Aging Management Review - Isolation Condenser (Dresden only), for the "valves" line item. LRA Reference 3.2.1.13, which addresses the internal environment of the new stainless steel valves has been added to the 'valves" line item of Table 2.3.2-5. (This item was a CLB change.)
* Exelon Identified the Need to Add Debris Screens for Dresden's Emerqency Diesel Generator System. The addition of the Dresden debris screens required a change to 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 Steel Filters at Dresden The cast iron filters 2-3999-38, 3-3999-380, and 2/3-3999-380 were replaced with cast steel filters. The cast iron filters, covered by LRA References 3.3.2.208 (internal) and 3.3.2.31 (external), need to remain as they still apply 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 Generator Cooling Water System, has been revised to add LRA References 3.3.1.15, and 3.3.1.5. (This item was a CLB change.)
* Diesel Generator Cooling Water Pump Material Change at Dresden This change approved stainless steel as an acceptable material for pumps 2-3903, 3-3903, and 2/3-3903, which had been previously approved for carbon steel only. The pumps are currently covered by References 3.3.1.15 (internal) and 3.3.2.26 (external). LRA Reference 3.3.1.15 is adequate for both carbon steel and stainless steel internal environments. An additional reference for the external environment (air moisture, humidity, and leaking fluid) for stainless steel pumps has been added to LRA Table 2.3.3-12, Component Groups Requiring Aging Management Review- Diesel Generator Cooling Water System. LRA Reference 3.3.2.41 has been modified to include pumps. Reference 3.3.2.41 has been added to the line item "pumps" in Table 2.3.3-12. (This item is a potential CLB change, the design has been approved and changes could occur prior to the new license being issued)
* Exelon Identified the Need to Revise the External Environment for Ultimate Heat Sink Discharge Piping The Quad Cities station 16' discharge piping was assigned the wrong external environment link (3.3.2.28). This piping is evaluated in the Ultimate Heat Sink System. This pipe is carbon steel buried in soil. The correct link is 3.3.1.16. LRA Table 2.3.3-22, Component Groups Requiring Aging Management Review - 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. The modification also added a strainer. The addition of strainer 1-8741 during this modification resulted in a new component type for Quad Cities in LRA Section 2.3.3.27, 'Drywell Nitrogen Inerting System," Table 2.3.3-27. Table 2.3.3-27, contains Component Groups of "Filter/Strainers (Dresden Only).' The LRA change needed is to remove the "Dresden Only" notation from Table 2.3.3-27, Component Groups of "Filter/Strainers (Dresden Only)," as identified in the attachment below.
Table 2.3.3-27, component groups of "Filter/Strainers (Dresden Only)." contain the proper links to Table 3.3-2, lines 3.3.2.25 and 3.3.2.51 for the external and internal surfaces of the brass or bronze filter/strainers. (This was a CLB change.)
* Exelon Identified the Need to Change the Classification of Quad Cities Station Crib House Fire Doors The fire rated door component scoping was evaluated and its classification was changed to be in the scope of the Rule requiring aging management, similar to Dresden. This required revision to the License Renewal Application 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 LRA submittal because Dresden had applied for and received approval for relief from

reactor vessel circumferential weld inspections. After submittal of the LRA, Exelon discovered that Dresden LRA Tables 4.2.6-1 and 4.2.7-1 contained errors requiring correction as described in the Exelon letter to USNRC dated April 17, 2003.
Exelon subsequently applied for similar relief from reactor vessel circumferential weld inspections for Quad Cities by letter to USNRC dated May 16, 2003. Therefore this topic became a TLAA for Quad Cities, and these LRA sections have been updated to reflect applicability to both stations. The updated sections also incorporate 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 Vessel Closure Studs LRA Section 4.3.1 reported on the results of a reanalysis performed for Extended Power Uprate (EPU) which concluded that the current bounding analysis for the reactor vessel closure studs listed a value of 0.750 for the 40-year cumulative usage factor (CUF). However, a subsequent analysis indicates that the reactor vessel closure studs will have a bounding value of < 1.0 for the 40-year cumulative usage factor (CUF). The studs will be monitored under the Fatigue Monitoring Program (B.1.34), which will adequately manage their aging. (This was a CLB change)
* Appendix A, Updated Final Safety Analysis Report (UFSAR) Supplement This 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 identify activities that are credited for compliance with the License Renewal Rule for aging management of passive, long-lived components and structures within the scope of License Renewal. Aging Management Reviews (AMRs) developed in support of the DNPS/QCNPS LRA provided further detail regarding the aging management activities identified in the LRA appendices.
ASME Section Xl programs are credited in the aging management activities for a number of systems. ASME Section Xl programs have had enhancements made to the Edition and Addenda implemented for these programs to comply with 10 CFR 50.55(a). The LRA originally indicated that the current Code of record for Dresden and Quad Cities was the 1989 Edition of ASME Section XI, and committed to update the program to be consistent with the requirements of the 1995 Edition through the 1996 Addenda of ASME Section Xl, as recommended by NUREG-1801.
DNPS/QCNPS have updated the ASME Section XI programs as committed. Future updates of the ASME Section Xl programs will be in accordance with 10 CFR 50.55(a).

3.0 REVISIONs To LRA Revised pages to the LRA that reflect the changes described in Section 2.0 above are provided here. The italic and bold text identifies the required additions and the strikethrough text identifies the required deletions to the LRA.
The pages revised as a result of this annual update also reflect those changes due to RAI responses that affected the same pages. Because Appendix A is provided in its entirety, all RAI related changes are included in that appendix. However, changes to other LRA pages that resulted 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 I            . (Continued)
Component Group Coe S y LeFunction Component Intended 3
Aging Management Ref 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, 3.1.1.17 Tubes and Guide Tubes Jet Pump Assemblies (Does not Pressure Boundary        3.1.1.1, 3.1.1.17, 3.1.1.19 include Sensing Lines)        l                          _
et Pump Assemblies (Does not tructural Support       3.1.1.1, 3.1.1.17 include 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      3.1.1.17 Modification/Repair Hardware
* Core Spray Clamp
* Jet Pump Riser Clamp (Quad Cities only)
* Jet Pump Riser Brace Clamp (Quad Cities only)
* 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.

LRA Replacement Page 2-67 Table 2.3.2-1 Component Groups Requiring Aging Management Review - High Pressure ll Coolant Injection System (Continued)
Component Group                Component Intended Function             I Aging Management Ref             ]
Filters/Strainers (includes       Filter                           3.2.1.2, 3.2.1.4, 3.2.1.13, 3.2.2.32 separators)
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, 3.2.2.136 alves (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, 3.2.2.137 Piping 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, 3.2.2.137 support)                                               .
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

LRA Replacement Page 2-81 Table 2.3.2-5 Component Groups Requiring Aging Management Review - Isolation Condenser (Dresden only) (Continued)
Component Group          l   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 the reactor coolant pressure boundary functions and Section 3.2 for the additional isolation condenser functions.

LRA Replacement Page 2-118 System Intended Functions Provide emergency AC power - provides independent power source to assure safe reactor shutdown under emergency conditions on a total loss of offsite power concurrent with a design basis accident.
Credited in regulated event(s) - credited in support of fire protection (10CFR50.48). The system 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 be a hazard to safety related SSCs maintain sufficient integrity so that the intended function of safety related SSCs is not adversely affected.
Component Groups Requiring Aging Management Review Table 2.3.3-6 Component Groups Requiring Aging Management Review - Emergency Diesel Generator and Auxiliaries Component Group             Component Intended           Aging Management Ref Function Air Accumulator Vessels        Pressure Boundary          3.3.1.5,3.3.2.6 Closure 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             3.3.1.5 Frames (includes dampers, duct, and housings)
Duct Fittings, Hinges, Latches Pressure Boundary          3.3.2.49 (includes anchors, bolts, and fasteners)
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          3.3.1.2 Damper 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, 3.3.2.99 coolers)
Lubricators                    Pressure Boundary          3.3.1.5

LRA Replacement Page 2-137 Table 2.3.3-12Component Groups Requiring Aging Management Review-Diesel Generator Coolina Water System (Continued)
                   - - - - -
* _* _I _- _I .I                                                             .    \ _
l      Component Group              Component Intended                 Aging Management Ref l    _   _  _   _   __                     Functio n            [                                       l Piping 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, 3.3.1.15 support)
Pulsation Dampeners            Structural Integrity (attached) 3.3.1.15, 3.3.2.41 (attached support) (Quad Cities 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                          3.3.1.15 only)
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 ;~cauil1 3.3.

LRA Replacement Page 2-166 Component Groups Requiring Aging Management Review Table 2.3.3-22Component Groups Requiring Aging Management Review - Ultimate Heat Sink Component Group       l    Component Intended         l      Aging Management Ref I           Function            I                                       l Closure Bolting (Dresden     Pressure Boundary               3.3.1.22 only)                       l                                 _       _ll Concrete Slabs (Dresden     Structural Pressure Barrier      3.5.1.22 only)                       I.I Concrete Walls (Dresden      Structural Pressure Barrier     3.5.1.22 only)                       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.

LRA Replacement Page 2-178 Component Groups Requiring Aging Management Review Table 2.3.3-27        Component Groups Requiring Aging Management Review - Drywell Nitrogen Inerting System Component Group               Component Intended l       Aging Management Ref l   _    _   _   _    _  _                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 provided in Section 3.3.

LRA Replacement Page 2-231 Component Groups Requiring Aging Management Review 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, 3.5.1.22 Steel Panels and Cabinets (Quad Cities Non-S/R Structural Support 3.5.1.20 only)
Steel Plates (Dresden only)             Direct Flow                    3.5.1.22

LRA Replacement Page 3-7 Table 3.1-1    Aging management programs evaluated in NUREG-1801 that are relied on for license renewal for the reactor vessel, internals, and reactor coolant system Ref No Component Components Evaluated               Aging              Aging            Further                       Discussion Effect/Mechanism      Management        Evaluation Program       Recommended        _     __              __
3.1.1.1  Reactor      NUREG-1801               Cumulative fatigue TLAA, evaluated in Yes, TLAA        Further Evaluation of cumulative fatigue coolant      Components               damage             accordance with                     damage is described in Section 3.1.1.1.1, pressure       Closure Bolting                           10 CFR 54.21(c)                     Section 4.3.1(reactor pressure vessel),
boundary       Core Plates                                                                   Section 4.3.2 (reactor vessel internals),
components     Core Spray Lines and.                                                         Section 4.3.3.1 (Dresden 3 RCPB piping),
Spargers                                                                        Section 4.3.3.2 (RCPB piping) and Section Incore Instrumentation                                                        4.3.4 (environmental effects of fatigue).
Dry Tubes and Guide Tubes Jet Pump Assemblies (Does not include Sensing Lines)
Nozzle Safe Ends Nozzles Orificed Fuel Supports Penetrations Piping and Fittings Pumps Support Skirts and Attachment Welds Top Guides Top Head Enclosure (Head Flanges)
Valves Vessel Bottom Heads Vessel Shells Evaluated with NUREG-1801 Components Shroud Support Structures

LRA Replacement Page 3-97 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                        Discussion Management Program 3.3.2.41 Component External Stainless Steel     Air, moisture,  Loss of material/ Open-Cycle        NUREG-1 801 does not address auxiliary Surfaces                               humidity, and   Pitting and crevice Cooling Water   system stainless steel components in a high
(
(restricting o orifices,                 leaking fluid   corrosion           System (B.1.13) moisture (pump vault) indoor environment.
valves, tubing, pulsation dampeners, pumps) 3.3.2.42 Component External Stainless Steel     Containment     None                None            NUREG-1 801 does not address stainless steel Surfaces                               Nitrogen                                            components in a containment nitrogen (piping and fittings,                                                                      environment. Containment nitrogen is not valves, 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 lined Surfaces                               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 the fittings)                                                                                  components is unlined carbon steel.
3.3.2.45 Component External Titanium            Air, moisture,  None                None            NUREG-1 801 does not address titanium Surfaces                               humidity, and                                       components. The high moisture (pump vault)
(valves, piping and                    leaking fluid                                       indoor environment does not promote aging fittings)                                                                                  degradation of these titanium components as they are not exposed to high chloride concentrations 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 or air             Pitting and crevice Inspection      bronze components in a warm, moist air corrosion           (B.1.23)        environment.

LRA Replacement Page4-24 4.2.6 Reactor Vessel Circumferential Weld Examination Relief Applicability This section applies to Dresden and Quad Cities.
Summary Description Relief from reactor vessel circumferential weld examination requirements under Generic Letter 98-05 is based on probabilistic assessments that predict an acceptable probability of failure per reactor operating year. The analysis is based on reactor vessel metallurgical conditions as well as flaw indication sizes and frequencies of occurrence that are expected at the end of a licensed operating period.
Dresden has received this relief for the remainder of the 40 year licensed operating period. Quad Cities has submitted a relief request for the remainder of the 40 year licensed operating period. (Reference 4.27) The circumferential weld examination relief analysis meets the requirements of 10CFR54.3(a) and is a TLAA.
Analysis Dresden received NRC approval for a technical alternative which eliminated the reactor vessel circumferential shell weld inspections for the current license term. Quad Cities has submitted a request for similarrelief. The basis for this relief request was an analysis that satisfied the limiting conditional failure probability for the circumferential welds at the expiration of the current license, based on BWRVIP-05 and the extent of neutron embrittlement. The anticipated changes in metallurgical conditions expected over the extended licensed operating period require an additional analysis for 54 EFPY and 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 a probabilistic fracture mechanics (PFM) analysis to estimate the RPV shell weld failure probabilities. Three key assumptions of the PFM analysis are: 1) the neutron fluence that was the estimated end-of-life mean fluence, 2) the chemistry values are mean values based on vessel types, and 3) the potential for beyond-design-basis events is considered.
Dresden Table 4.2.6-1 provides a comparison of the Dresden reactor vessel limiting circumferential weld parameters to those used in the NRC analysis for the first two key assumptions. Data provided in Table 4.2.6-1 was supplied from Tables 2.6-4 and 2.6-5 of the Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report.
Although the chemistry composition and chemistry factor for unit 3 are higher than the limits of the NRC Analysis; the 54 EFPY fluence results are considerably lower for both Dresden Units 2 and 3. As a result, the shifts in reference temperature for both units are lower than the 54 EFPY shift from the NRC analysis. In addition, the unirradiated reference temperatures for both Dresden units are lower. The combination of unirradiated reference temperature     (RTNDT(U)) and shift (ARTNDTW/O margin) yields adjusted reference temperatures that are considerably lower than the NRC mean analysis values. Therefore, the RPV shell weld embrittlement due to fluence has a

LRA Replacement Page 4-25 negligible effect on the probabilities of RPV shell weld failure. The Mean RTNDTvalues for both units at 54 EFPY are bounded by the 64 EFPY Mean RTNDT provided by the NRC. Although a conditional failure probability has not been calculated, the fact that the Dresden 54 EFPY values are less than the 64 EFPY value provided by the NRC leads to the conclusion that the Dresden RPV conditional failure probability is bounded by the NRC analysis.
The procedures and training used to limit cold over-pressure events will be the same as those approved by the NRC when Dresden requested the BWRVIP-05 technical alternative be used for the current term (Reference 4.14). An extension of this relief for Dresden for the 60-year period will be submitted to the NRC for approval prior to the period of extended operation.
Table 4.2.6-1 Effects for Irradiation on RPV Circumferential Weld Properties Dresden Units 2 & 3 B&W               Dresden Unit 2        Dresden Unit 3 Group                 64 EFPY                 54 EFPY              54 EFPY 0.31                   0.23                  0.34 Cu%
Ni%                     0.59                   0.59                  0.68 CF                   196.7                   168                   221 Fluence at                 0.19                  0.042                 0.041 clad/weld interface 2
(10'9 n/cm )
ARTNDTW/O     margin           109.4                    44                    58 (OF)
RTNDT(U)         (OF)            20                      10                   -5 Mean   RTNDT      (IF)         129.4                    54                    53 P(FIE)          NRC         4.83 x 104                   ---      _---
P(FIE)      BWRVIP               ---                    ---
Quad Cities Table 4.2.6-2 provides a comparison of the limiting RPV circumferential weld parameters for each QCNPS unit to those used in the NRC analysis for the first two key assumptions.
The chemistry composition and chemistry factor for QCNPS Unit 1 are less than or equal to the limits of the NRC analysis. While the nickel content for Unit 2 is higher than the value utilized in the NRC analysis, the Unit 2 copper content and the chemistry factor are considerably lower than the values utilized in the NRC analysis. Additionally, the unirradiated reference temperatures for both QCNPS units are lower than the NRC limits. The combination of unirradiated reference temperature and embrittlement shift yields adjusted reference temperatures considerably 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 fluences for QCNPS, Units 1 and 2, are also considerably lower than the NRC estimated fluence. Therefore, the RPVshell weld embrittlement due to fluence has a negligible effect on the probabilities of RPVshell weld failure. Each unit's RPV shell circumferential weld failure probabilities are bounded by the conditional failure 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 Properties Quad Cities Units I & 2 B&W              Quad Cities Unit 1      Quad Cities Unit 2 roup64                  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 at clad/weld                    0.19                    0.041                  0.041 interface (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

LRA Replacement Page 4-26 4.2.7            Reactor Vessel Axial Weld Failure Probability Applicability This section applies to Dresden and Quad Cities.
Summary Description The Boiling Water Reactor Owner's Group Vessel and Internals Program recommendations 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 per reactor year. BWRVIP-05 showed that this axial weld failure rate of 5 x 104 per reactor year is orders of magnitude greater than the 40-year end-of-life circumferential weld failure probability, and used this analysis to justify relief from inspection of the circumferential welds as described in Section 42.6 Dresden received relief from the circumferential weld inspections for the remaining 40 year licensed operating period. Quad Cities has also submitted a relief request for the remaining 40-year license operating period.
Analysis As stated in Section 4.2.6, Dresden Station received NRC approval for a technical alternative which eliminated the reactor vessel circumferential shell weld inspections for the current license term. Quad Cities has applied for a similarrelief request. The basis for both relief requests was an analysis that satisfied the limiting conditional failure probability for the circumferential welds at the expiration of the current license, based on BWRVIP-05 and the extent of neutron embrittlement. The NRC SER associated with BWRVIP-05 (Reference 4.15) concluded that the reactor vessel failure frequency due to failure of the limiting axial welds in the BWR fleet at the end of 40 years of operation is less than 5 x 104 per reactor year. This failure frequency is dependent upon given assumptions of flaw density, distribution, and location. The failure frequency also assumes that "essentially 100%" of the reactor vessel axial welds will be inspected.
Due to various obstructions within the reactor vessel, Dresden and Quad Cities have not been able to meet the "essentially 100%" inspection requirement. For Dresden, an analysis was performed to assess the effect on the probability of fracture due to the actual inspection performed on the vessel axial welds and to determine if the coverage was sufficient in the inspection of regions contributing to the majority of the risk. The analysis included an estimate and comparison of the probability of failure for the cases of uessentially 100%' inspection and the limited scope inspections on the Dresden 2 and 3 vessel axial welds. The analysis concluded that the conditional probabilities of failure due to a low temperature over pressurization event are very small, 2.96 x 1O8 and 3.15 x 10i10 for Dresden Unit 2 and Unit 3, respectively. The conditional probability of failure with the 'essentially 100%" inspections were an order of magnitude lower than that for the actual inspection coverage. However, the Dresden analysis only applies to the current 40-year operating period. The anticipated changes in metallurgical conditions expected over the extended licensed operating period require an additional

LRA Replacement Page 4-27 analysis for 54 EFPY and approval by the NRC to extend the reactor vessel circumferential 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 2 and 3 against the values taken from Table 2.6-5 found in the NRC SER for BWRVIP-05 and associated supplement to the SER (Reference 4.16). The SER supplement required the limiting axial weld to be compared with data found in Table 3 of the document. Table 4.2.7-2 compares the limitingaxial weld 54 EFPY properties for Quad Cities Units 1 and 2 against the values taken from Table 2.6-5 found in the NRC SER. For Dresden 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 limiting welds for the BWR fleet and vessel failure probability calculations determined for Clinton should bound those for the BWR fleet.
The limiting axial welds at both Dresden and Quad Cities are all electroslag welds with similar chemistry. The Dresden and Quad Cities limiting weld chemistry, chemistry factor (CF), and 54 EFPY mean RTNDT values are within the limits of the values assumed in the analysis performed by the NRC staff in the March 7, 2000 BWRVIP-05 SER supplement and the 64 EFPY limits and values obtained from Table 2.6.5 of the SER.
As stated above, the probability of a failure event PFE calculated by the NRC BWRVIP-05 SER and its supplements depends in part on an assumption that 90 per cent of axial welds can be inspected. Less than 90 per cent of axial welds can be examined at Dresden and Quad Cities. As such, an analysis was performed for 54 EFPY to assess the effect on the probability of fracture due to the actual inspection performed on the vessel axial welds and to determine if the coverage was sufficient in the inspection of regions contributing to the majority of the risk. The analysis included the estimate and comparison of the probability of failure for both the case of "essentially 100%" inspection and the actual limited scope inspections on the Dresden Unit 2 and Unit 3 and Quad Cities Unit 1 and Unit 2 vessel axial welds. The analysis concluded that the conditional probabilities of failure due to a low temperature over pressurization event are very small: 3.89 x 104 and 5.07 x 10.8 for Dresden Unit 2 and Unit 3, and 2.08xIOi-and 5.277xIU7 for Quad Cities Unit 1 and Unit 2, respectively. The evaluation shows that the calculated unit-specific axial weld conditional failure probabilities at 54 EFPY for Dresden and Quad Cities are less than the failure probabilities calculated by the NRC staff in the SER at 64 EFPY and the limiting Clinton values found in Table 3 of the SER supplement. The probability of failure of an axial weld at Dresden or Quad Cities will therefore provide adequate margin above the probability of failure of a circumferential weld, in support of relief from inspection of circumferential welds, for the extended licensed operating period.

LRA Replacement Page 4-28 Table 4.2.7-1 Effects for Irradiation on RPV Axial Weld Properties Dresden Units 2 & 3 Value    -          B&W               SER            DRE 2           DRE 3 Supplement        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           0.057 n/cm ARTNDT               88.9               121              44              44 OF RTNDT(U)               10               -30              23              23 OF 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 10 BWRVIP

LRA Replacement Page 4-28a Table 4.2.7-2, Effects for Irradiation on RPVAxial Weld Properties Quad Cities Units 1 & 2 Parameter             B&W              SER        QCNPS Unit I  QCNPS Unit 2 Description          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       0.057 X 1019  n/cm2 ARTNDT,   OF           88.9             121              44          44 ARTNDT(U),  OF           10              -30              23          23 Mean RTNDT, OF           98.9              91              67          67 P(FIE) NRC         1.87x 10-'       2.73x 10f        2.08x 10-7  5.27x 107 P(FIE) BWR VIP                        1.52 x  10"f         ...         ---

LRA Replacement Page 4-30 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 lower heads, closure flanges, nozzles and penetrations, nozzle safe ends, and closure studs, depend on assumed numbers and severity of normal and upset-event pressure and thermal operating cycles to predict end-of-life fatigue usage factors.
These assumed cycle counts and fatigue usage factors are based on 40 years of operation. Calculation of fatigue usage factors is part of the current licensing basis and is used to support safety determinations. The reactor vessel fatigue analyses are TLAAs.
Analysis The original reactor pressure vessel stress report included a fatigue analysis for the reactor vessel components based on a set of design basis duty cycles. These duty cycles are listed in Table 3.9-1 of the Dresden and Quad Cities UFSARs. The original 40-year analyses demonstrated that the cumulative usage factors (CUF) for the critical components 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 a part of Extended Power Uprate (EPU) implementation at all four Dresden and Quad Cities units. (References 4.10 and 4.11) A subset of the bounding reactor vessel components was evaluated as a part of this re-analysis. The resulting fatigue cumulative usage factors (CUFs) for these limiting components supersede the values determined 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 CUFs for limiting components.
Shroud Support                          0.820 Support Skirt                           0.862 Feedwater Nozzle (Safe End)             0.748 Closure Studs                        < 1.000 The original code analysis of the reactor vessel included fatigue analysis of the feedwater and control rod drive hydraulic system return line nozzles. After several years of operation, it was discovered that both the control rod drive hydraulic system return line nozzles and the feedwater nozzles were subject to cracking caused by a number of factors including rapid thermal cycling. Consequently, the control rod drive hydraulic system return line nozzles were capped and removed from service. As such, they are no longer subject to rapid thermal aging. A reanalysis was later performed on the

LRA Replacement Page 4-32 extended operation. Dresden and Quad Cities have programs in place to track operating thermal and pressure cycles and to assess their effect on vessel fatigue. The requirements from these procedures will be incorporated into the Metal Fatigue of Reactor Coolant Pressure Boundary (B.1.34) aging management program.
Table 4.3.1-1 Fatigue Monitoring Locations for Reactor Pressure Vessel Components Component            Computed Fatigue      Computed Fatigue         Monitoring Usage Factor         Usage Factor           Technique (Pre-EPU)             (Post EPU)
Recirculation Outlet Nozzle        0.270                 Note 4         CBF (NUREG/CR-6260 component)
Recirculation Inlet Nozzle         0.301                 Note 4         CBF (NUREG/CR-6260 component)
Feedwater Nozzle                   0.538                  .748          SBF (NUREGICR-6260 component)
Core Spray Nozzle                  0.079                 Note 4         CBF (NUREG/CR-6260 component)
Support Skirt                      0.940                 0.862          SBF Shroud Support                     0.630                 0.820          CBF Closure Stud Bolts                  0.79                  < 1.0         CBF Vessel Shell                       0.141                 Note 4         CBF (NUREG/CR-6260 component)
Notes:
: 1. CBF = Cycle Based Fatigue and SBF = Stress Based Fatigue
: 2. EPU = Extended Power Uprate
: 3. The components listed as a "NUREG/CR-6260 component" will be monitored for GSI -190. See Section 4.3.4
: 4. Only locations with 40-year CUF expected to exceed 0.400 have been computed.

LRA Replacement Page 4-72 4.17 S&L Report EMD-033967 for Quad Cities, Piping Stress Analysis Report, Mark I Program Plant Unique Analysis of the SRV Discharge Piping System in the Suppression 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 I Program Plant Unique Analysis of the SRV Discharge Piping System in the Suppression 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 Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants."
4.23 DOR Guidelines, 'Guidelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors," U.S. Nuclear Regulatory Commission, June 1979.
4.24 NUREG-0588, 'Interim Staff Position on Environmental Qualification of Safety Related Electrical Equipment," U.S. Nuclear Regulatory Commission, July 1981.
4.25 Regulatory Guide 1.89, Revision 1, "Environmental Qualification of Certain Electrical Equipment Important to Safety for Nuclear Power Plants," U.S. Nuclear Regulatory Commission, June 1984.
4.26 C.l. Grimes (NRC) letter to D. Walters (NEI), "Guidance on Addressing GSI-168 for License Renewal," Project 690, June 1998.
4.27 USNRC Safety Evaluation by the Office of Nuclear Reactor Regulations, Related to Alternative to Inspection of Reactor Pressure Vessel Circumferential Welds, Dresden Nuclear 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 Weld Examinations (TAC NOS. MA6228 and MA6229) dated February 25, 2000.
4.28 Letter from Patrick R. Simpson (Exelon) to USNRC, "Corrected Fluence Tables for Dresden Nuclear Power Station, Units 2 and 3 License Renewal Application," dated April 17, 2003

Table of Contents APPENDIX A UPDATED 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-3 A.1.1         ASME Section Xi Inservice Inspection, Subsections IWB, IWC, and IWD .A-3 A.1.2         Water Chemistry .A-3 A.1.3         Reactor Head Closure Studs .A-3 A.1.4         BWR Vessel IDAttachment Welds .                                                         A-3 A.1.5         BWR Feedwater Nozzle .                                                                  A-4 A.1.6         BWR Control Rod Drive Return Line Nozzle.                                               A4 A.1.7         BWR Stress Corrosion Cracking .A-4 A.1.8         BWR Penetrations .                                                                      A-5 A.1.9         BWR Vessel Internals .A-5 A.1.10        Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) .                                               A-6 A.1.11        Flow-Accelerated Corrosion .A-6 A.1.12        Bolting Integrity ............                                                          A-7 A.1.13        Open-Cycle Cooling Water System .A-7 A.1.14        Closed-Cycle Cooling Water System .                                                     A-8 A.1.15        Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems .A-8 A.1.16        Compressed Air Monitoring .A-8 A.1.17        BWR Reactor Water Cleanup System .                                                      A-8 A.1.18        Fire Protection .                                                                       A-9 A.1.19        Fire Water System .A-9 A.1.20        Aboveground Carbon Steel Tanks .                                                       A-10 A.1.21        Fuel Oil Chemistry .A-10 A.1.22        Reactor Vessel Surveillance .A-11 A.1.23        One-Time Inspection .A-11 A.1.24        Selective Leaching of Materials .A-12 A.1.25        Buried Piping and Tanks Inspection .A-13 A.1.26        ASME Section Xl, Subsection IWE .A-13 A.1.27        ASME Section Xl, Subsection IWF .A-13 A.1.28        10 CFR Part 50, Appendix J.A-14 A.1.29        Masonry Wall Program .A-14 A.1.30        Structures Monitoring Program .A-14 A.1.31        RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants.A-15 A.1.32        Protective Coating Monitoring and Maintenance Program .A-15 A.1.33        Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements .A-15 A.1.34        Metal Fatigue of Reactor Coolant Pressure Boundary .A-16 A.1.35        Environmental Qualification (EQ) of Electrical Components .A-16 A.1.36        Not Used.A-16 A.1.37        ElectricalCables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements used in Instrument Circuits .A-16 A.1.38        Inaccessible Medium-Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements .A-17 Dresden and Quad Cities                                                                   Page A-i License Renewal Application

Table of Contents A.2           PLANT-SPECIFIC AGING MANAGEMENT PROGRAMS ................................ A-18 A.2.1         Corrective Action Program .............................................A-18 A.2.2         Periodic Inspection of Non-EQ, Non-Segregated Electrical Bus Ducts .......................................                                                          A-18 A.2.3         Periodic Inspection of Ventilation System Elastomers ...................................... A-1 9 A.2.4         Periodic Testing of Drywell and Torus Spray Nozzles ....................................... A-19 A.2.5         Lubricating Oil Monitoring Activities .......................................                               A-1 9 A.2.6         Heat Exchanger Test and Inspection Activities ..                    ....................................... A-20 A.2.7         Not Used .......................................                                                            A-20 A.2.8         Periodic Inspection of Plant Heating System .......................................                         A-20 A.2.9         PeriodicInspection of Components Subject to Moist Air Environments. A-21 A.3           TIME-LIMITED AGING ANALYSIS SUMMARIES ........................................... A-22 A.3.1         Neutron Embrittlement of the Reactor Vessel and Internals ............................. A-22 A.3.1.1       Reactor Vessel Materials Upper-Shelf Energy Reduction Due to Neutron Embrittlement .                                                                                    A-22 A.3.1.2       Adjusted Reference Temperature for Reactor Vessel Materials Due to Neutron Embrittlement ........................................                                      A-22 A.3.1.3       Reflood Thermal Shock Analysis of the Reactor Vessel ................................... A-22 A.3.1.4       Reflood Thermal Shock Analysis of the Reactor Vessel Core Shroud and Repair Hardware .                                                                                                 A-22 A.3.1.5       Reactor Vessel Thermal Limit Analyses: Operating Pressure - Temperature Limits .........................................                                                           A-23 A.3.1.6       Reactor Vessel Circumferential Weld Examination Relief ................................ A-23 A.3.1.7       Reactor Vessel Axial Weld Failure Probability .........................................                     A-23 A.3.2         Metal Fatigue ...........................................                                                   A-24 A.3.2.1       Reactor Vessel Fatigue Analyses.........................................                                    A-24 A.3.2.2       Fatigue Analysis of Reactor Vessel Internals ..                   ......................................... A-24 A.3.2.2.1     High-Cycle Flow-Induced Vibration Fatigue Analysis of Jet Pump Riser Braces ...                                                                                                  A-24 A.3.2.3       Reactor Coolant Pressure Boundary Piping and Component Fatigue Analysis .                                                                                                  A-24 A.3.2.3.1     ASME Section III Class 1 Reactor Coolant Pressure Boundary Piping and Component Fatigue Analysis .                                                                                A-25 A.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 VilI Class B and C ...............................                                                               A-25 A.3.2.3.3     Fatigue Analysis of the Isolation Condenser ...............................                                 A-26 A.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-26 A.3.4         Containment Fatigue ......................................                                                  A-27 A.3.4.1       Fatigue Analysis of the Suppression Chamber, Vents, and Downcomers .                                                                                                A-27 A.3.4.2       Fatigue Analysis of SRV Discharge Piping Inside the Suppression Chamber, External Suppression Chamber Attached Piping, and Associated Penetrations .......                                                                                        A-27 A.3.4.3       Drywell to Suppression Chamber Vent Line Bellows Fatigue Analyses ............ A-27 A.3.4.4       Primary Containment Process Penetrations Bellows Fatigue Analysis ............. A-28 A.3.5         Other Plant-Specific TLAAs ...................................................                              A-28 A.3.5.1       Reactor Building Crane Load Cycles ...................................................                      A-28 A.3.5.2       Metal Corrosion Allowances ...................................................                              A-28 A.3.5.2.1     Degradation Rates of Inaccessible Exterior Drywell Plate Surfaces ................. A-28 Dresden and Quad Cities                                                                                       Page A-ii License Renewal Application

Table of Contents A.3.5.2.2     Galvanic Corrosion in the Containment Shell and Attached Piping Components due to Stainless Steel ECCS Suction Strainers .......................... A-29 A.3.5.3       Crack Growth Calculation of a Postulated Flaw in the Heat Affected Zone of an Arc Strike in the Suppression Chamber Shell ..................A-29 A.3.5.4       Radiation Degradation of Drywell Shell Expansion Gap Polyurethane 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-33 A.1.1         ASME Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD .A-33 A.1.2         Water Chemistry .A-33 A.1.3         Reactor Head Closure Studs .A-33 A.1.4         BWR Vessel IDAttachment Welds .                                                      A-33 A.1.5         BWR Feedwater Nozzle .A-34 A.1.6         BWR Control Rod Drive Return Line Nozzle .A-34 A.1.7         BWR Stress Corrosion Cracking .A-34 A.1.8         BWR Penetrations .                                                                   A-35 A.1.9         BWR Vessel Internals .A-35 A.1.10        Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) .A-36 A.1.11        Flow-Accelerated Corrosion .A-36 A.1.12        Bolting Integrity .                                                                  A-37 A.1.13        Open-Cycle Cooling Water System .                                                    A-37 A.1.14        Closed-Cycle Cooling Water System .                                                  A-38 A.1.15        Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems .A-38 A.1.16        Compressed Air Monitoring .A-38 A.1.17        BWR Reactor Water Cleanup System .                                                   A-39 A.1.18        Fire Protection .                                                                    A-39 A.1.19        Fire Water System .A-39 A.1.20        Aboveground Carbon Steel Tanks .                                                     A-40 A.1.21        Fuel Oil Chemistry.                                                                  A-40 A.1.22        Reactor Vessel Surveillance.                                                         A-41 A.1.23        One-Time Inspection .A-41 A.1.24        Selective Leaching of Materials.                                                     A-42 A.1.25        Buried Piping and Tanks Inspection.                                                  A-43 A.1.26        ASME Section Xl, Subsection IWE .A-43 A.1.27        ASME Section Xl, Subsection IWF ........................                             A-43 A.1.28        10 CFR Part 50, Appendix J.                                                          A-44 A.1.29        Masonry Wall Program .........................                                       A-44 A.1.30        Structures Monitoring Program.                                                       A-44 A.1.31        RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants ..................                                        A-45 A.1.32        Protective Coating Monitoring and Maintenance Program.                               A-45 A.1.33        Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements.                                           A-46 A.1.34        Metal Fatigue of Reactor Coolant Pressure Boundary.                                  A-46 A.1.35        Environmental Qualification (EQ) of Electrical Components.                           A-46 A.1.36        Boraflex Monitoring.                                                                 A-46 A.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-iii License Renewal Application

Table of Contents A.2           PLANT-SPECIFIC AGING MANAGEMENT PROGRAMS ................................ A-48 A.2.1         Corrective Action Program .............................................A-48 A.2.2         Periodic Inspection of Non-EQ, Non-Segregated Electrical Bus Ducts .......................................                                                         A-48 A.2.3         Periodic Inspection of Ventilation System Elastomers ...................................... A-49 A.2.4         Periodic Testing of Drywell and Torus Spray Nozzles ....................................... A-49 A.2.5         Lubricating Oil Monitoring Activities .......................................                              A-49 A.2.6         Heat Exchanger Test and Inspection Activities ..                   ....................................... A-50 A.2.7         Generator Stator Water Chemistry Activities .......................................                        A-50 A.2.8         Periodic Inspection of Plant Heating System ....................................... A-51 A.2.9         Periodic Inspection of Components Subject to Moist Air Environments. A-51 A.3           Time-Limited Aging Analysis Summaries ...........................................                          A-52 A.3.1         Neutron Embrittlement of the Reactor Vessel and Internals ............................. A-52 A.3.1.1       Reactor Vessel Materials Upper-Shelf Energy Reduction Due to Neutron Embrittlement .                                                                                 A-52 A.3.1.2       Adjusted Reference Temperature for Reactor Vessel Materials Due to Neutron Embrittlement ........................................                            A-52 A.3.1.3       Reflood Thermal Shock Analysis of the Reactor Vessel ................................... A-52 A.3.1.4       Reflood Thermal Shock Analysis of the Reactor Vessel Core Shroud and Repair Hardware .                                                                          A-52 A.3.1.5       Reactor Vessel Thermal Limit Analyses: Operating Pressure - Temperature Limits ..                                                                                                  A-53 A.3.1.6       Reactor Vessel Circumferential Weld Examination Relief ......................... A-53 A.3.1.7       Reactor Vessel Axial Weld Failure Probability ............................................ A-53 A.3.2         Metal Fatigue ..............................................                                               A-54 A.3.2.1       Reactor Vessel Fatigue ..............................................                                      A-54 A.3.2.2       Fatigue Analysis of Reactor Vessel Internals .............................................. A-54 A.3.2.2.1     Low-Cycle Thermal Fatigue Analysis of the Core Shroud and Repair Hardware .                                                                                                 A-54 A.3.2.3       Reactor Coolant Pressure Boundary Piping and Component Fatigue Analysis ...                                                                                               A-55 A.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 VilI Class B and C.                                                                                             A-55 A.3.2.4       Effects of Reactor Coolant Environment on Fatigue Life of Components and Piping (Generic Safety Issue 190) ..................................... A-55 A.3.3         Environmental Qualification Of Electrical Equipment .....................................                 A-56 A.3.4         Containment Fatigue .....................................                                                 A-56 A.3.4.1       Fatigue Analysis of the Suppression Chamber, Vents, and Downcomers .                                                                                              A-56 A.3.4.2       Fatigue Analysis of SRV Discharge Piping Inside the Suppression Chamber, External Suppression Chamber Attached Piping, and Associated Penetrations .                                                                                 A-56 A.3.4.3       Drywell to Suppression Chamber Vent Line Bellows Fatigue Analyses .                                                                                        A-57 A.3.4.4       Primary Containment Process Penetrations Bellows Fatigue Analysis ........................                                                                 A-57 A.3.5         Other Plant-Specific TLAAs ........................                                                       A-57 A.3.5.1       Reactor Building Crane Load Cycles ........................                                               A-57 A.3.5.2       Metal Corrosion Allowances ........................                                                       A-57 A.3.5.2.1     Corrosion Allowance for Power Operated Relief Valves ................................... A-57 A.3.5.2.2     Degradation Rates of Inaccessible Exterior Drywell Plate Surfaces ................. A-58 Dresden and Quad Cities                                                                                      Page A-iv License Renewal Application

Table of Contents A.3.5.2.3     Galvanic Corrosion in the Containment Shell and Attached Piping Components due to Stainless Steel ECCS Suction Strainers .......................... A-58 A.3.5.3       Crack Growth Calculation of a Postulated Flaw in the Heat Affected Zone of an Arc Strike in the Suppression Chamber Shell ..................A-59 A.3.5.4       Radiation Degradation of Drywell Shell Expansion Gap Polyurethane Foam ................................................                        A-59 A.3.6         References for Section A.3 ................................................               A-60 Dresden and Quad Cities                                                                           Page A-v License Renewal Application

Appendix A UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT APPENDIX A              UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT Introduction The summary descriptions of aging management program activities presented in this Appendix A represent commitments for managing aging of the systems, structures and components within the scope of license renewal during the period of extended operation. This appendix also provides summary descriptions of time-limited aging analyses. These summary descriptions of aging management program activities and time-limited aging analyses will be incorporated in the Updated Final Safety Analysis Reports for the Dresden Nuclear Power Station and the Quad Cities Nuclear Power Station following issuance of the renewed operating license.
A separate Appendix A Updated Final Safety Analysis Supplement has been provided or Dresden, Units 2 and 3 and for Quad Cities, Units 1 and 2.
Dresden and Quad Cities                                                       Page A-1 I License Renewal Application

Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT Dresden Units 2 and 3 Updated Final Safety Analysis Supplement Dresden and Quad Cities                                        Page A-2 License Renewal Application

Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 aging management program consists of periodic volumetric and visual examinations of components for assessment, identification of signs of degradation, and establishment of corrective 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. The inspections will be implemented in accordance with 10 CFR 50.55(a).
Dresden will implement the guidance of BWRVIP-74, "BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines," with the followingexception.
Exception: Risk Informed Inservice Inspectionis implemented in lieu of ASME Section Xl requirements for portions of Class 1 and Class 2 systems. Technical Specification revisions containingnew P-T Curves willbe submittedprior to the term of extended operation.
A.1.2 Water Chemistry The water chemistry aging management program consists of monitoring and control of water chemistry to keep peak levels of various contaminants below system-specific limits based on industry-recognized guidelines of EPRI TR-103515, UBWR Water Chemistry Guidelines." To mitigate aging effects on component surfaces that are exposed to water as process fluid, the chemistry programs are used to control water chemistry for impurities (e.g., chlorides, and sulfates) that accelerate corrosion.
Dresden will implement the general guidance providedin BWRVIP-79, "EPRI Report TR-103515-R2."
A.1.3 Reactor Head Closure Studs The reactor head closure studs aging management program includes inservice inspection (ISI). This program also includes preventive actions and inspection techniques 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 10 CFR 50.55(a). The reactor head studs are not metal-plated, and have had manganese phosphate 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-approved Boiling Water Reactor Vessel and Internals Project BWRVIP-48, 'Vessel ID Attachment Weld Inspection and Evaluation Guidelines," and/or ASME Section Xl; and (b) monitoring and control of reactor coolant water chemistry in accordance with industry-recognized guidelines of EPRI TR-103515, 'BWR Water Chemistry Dresden and Quad Cities                                                           Page A-3 License Renewal Application

Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT Guidelines." Prior to the period of extended operation the program will be revised to be consistent with ASME Section Xl, 1995 Edition through the 1996 Addenda. The requirements of ASME Section Xl will be implemented in accordance with 10 CFR 50.55(a).
A.1.5    BWR Feedwater Nozzle The BWR feedwater nozzle aging management program includes enhancing the inservice inspections (ISI) specified in the ASME Code, Section Xi, with the recommendation of General Electric (GE) NE-523-A71-0594, "Alternate BWR Feedwater Nozzle Inspection Requirements," to perform periodic ultrasonic testing inspection 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 of previously implemented system modifications and inservice inspections that manage the aging effect of cracking in the control rod drive return line nozzles. The control rod drive return line nozzles have been capped. Inservice inspections are performed consistent with ASME Section Xl requirements. No augmented inspections in accordance with NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," or the alternative recommendations of GE NE-523-A71-0594, 'Alternate BWR Feedwater Nozzle Inspection Requirements," are required. Prior to the period of extended 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 will be 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 manage intergranular stress corrosion cracking (IGSCC) in boiling water reactor coolant pressure boundary piping four inches and larger nominal pipe size made of stainless steel (SS) is delineated, in part, in NUREG-0313, "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," Revision 2, BWRVIP 75, 'Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules," and Nuclear Regulatory Commission (NRC) Generic Letter (GL) 88-01, "NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping," and its Supplement 1. The program includes (a) replacements and preventive measures to mitigate IGSCC and (b) inspections to monitor IGSCC and its effects. Water chemistry is monitored and maintained in accordance with industry-recognized guidelines in EPRI TR-103515, "BWR Water Chemistry Guidelines." P-ior-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 10 CFR 50.55(a).
Dresden will implement the general guidance providedin BWRVIP-75, "Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules, " with Exception Dresden and Quad Cities                                                          Page A-4 License Renewal Application

Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT
- The Relief Request submittedfor the implementation of RISI indicates the Category A Welds are "subsumed into the RISI program."
A.1.8 BWR Penetrations The BWR penetrations aging management program includes (a) inspection and flaw evaluation in conformance with the guidelines of staff-approved Boiling Water Reactor Vessel and Internals Project (BWRVIP)-49, "Instrument Penetration Inspection and Flaw Evaluation Guidelines," and BWRVIP-27, 'BWR Standby Liquid Control System/Core Plate Delta-P Inspection and Flaw Evaluation Guidelines," documents and (b) monitoring and control of reactor coolant water chemistry in accordance with industry-recognized guidelines of EPRI TR-103515, "BWR Water Chemistry Guidelines," to ensure the long-term integrity and safe operation of boiling water reactor vessel internal components. Prior to the period of extended operation the program will be revised to be consistent with ASME Section Xl, 1895 Edition through the 1996 Addenda-The requirements of ASME Section Xl will be implemented in accordance with 10 CFR 50.55(a).
Dresden will implement the guidance provided in BWRVIP-27, "BWR Standby Liquid Control System/Core Plate AP Inspection and Flaw Evaluation Guidelines,"
with the followingexception. Dresden has implemented approved iSI Relief Requests, which provide a VT-2 examination of the standby liquid controlsystem nozzle inner radius in lieu of the Code required volumetric examination.
Dresden will implement the guidance provided in BWRVIP-49, "Instrument Penetration Inspection and Flaw Evaluation Guidelines."
A.1.9   BWR Vessel Internals The BWR vessel internals aging management program includes (a) inspection and flaw evaluation in conformance with the guidelines of applicable and staff-approved Boiling Water Reactor Vessel and Internals Project (BWRVIP) documents, and with ASME Section Xl; and (b) monitoring and control of reactor coolant water chemistry in accordance with industry-recognized guidelines of EPRI TR-103515, UBWR Water Chemistry Guidelines," to ensure the long-term integrity and safe operation of boiling water reactor vessel internal components. Prior to the period of extended operation the pFrogramn will be revised to be conRsitent with ASME Section Xl, 1995 Edition through the 18996 Addenda. The requirements of ASME Section Xl willbe implemented in accordance with 10 CFR 50.55(a).
Dresden will implement the general guidance provided in BWRVIP-18, "BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines."
Dresden will implement the general guidance providedin BWRVIP-25, "BWR Core Plate Inspection and Flaw Evaluation Guidelines."
Dresden will implement the guidance provided in BWRVIP-26, "BWR Top Guide Inspection and Flaw Evaluation Guidelines." Additionally, Dresden willperform Dresden and Quad Cities                                                         Page A-5 License Renewal Application

Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT augmented inspections for the top guide similar to the inspections of control rod drive housing (CRDH) guide tubes.
Dresden will implement the general guidance provided in BWRVIP-38, "BWR Shroud Support Inspection an Flaw Evaluation Guidelines." Dresden will perform the additional inspections in the lower plenum (i.e. shroud support leg welds) when new inspection techniques and tooling are developed, incorporated into the applicable BWRVIP document(s), and approved by NRC SER.
Dresden will implement the general guidance provided in BWRVIP-41, "BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines." Dresden will perform the additional inspections of the inaccessible thermal sleeve welds when new inspection techniques and tooling are developed, incorporated into the applicable BWRVIP document(s), and approved by NRC SER.
Dresden will implement the general guidance provided in BWRVIP-47, "BWR Lower Plenum Inspection and Flaw Evaluation Guidelines."
Dresden will implement the general guidance provided in BWRVIP-76, "BWR Core Shroud Inspection and Flaw Evaluation Guidelines."
Dresden will implement the general guidance provided in BWRVIP-104, "Evaluation and Recommendations to Address Shroud Support Cracking in BWRs. "
A.1.10 Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless 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 of cast 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 is being implemented prior to the period of extended operation.
A.1.11 Flow-Accelerated Corrosion The flow-accelerated corrosion aging management program consists of (1) appropriate analysis and baseline inspections, (2) determination of the extent of thinning, and replacement or repair of components, and (3) follow-up inspections to confirm or quantify effects, and to take longer-term corrective actions. This program is in response to NRC Generic Letter 89-08, "Erosion/Corrosion-Induced Pipe Wall Thinning." The program relies on implementation of the EPRI NSAC-202L, "Recommendations for an Effective Flow Accelerated Corrosion Program," Revision 2 guidelines. Prior to the period of extended operation the program will be revised to include main steam piping within the scope of license renewal.
Dresden and Quad Cities                                                          Page A-6 License Renewal Application

Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.1.12 Bolting Integrity This bolting integrity aging management program incorporates industry recommendations of EPRI NP-5769, 'Degradation and Failure of Bolting in Nuclear Power Plants," and includes periodic visual inspections for external surface degradation that may be caused by loss of material or cracking of the bolting, or by an adverse environment. Inspection of inservice inspection Class 1, 2, and 3 components is conducted in accordance with ASME Section Xl. Prior to the period of extended operation the program will be revised to be consistent with A.SME Section Xl, 1995 Edition through the 1996 Addenda. The requirements ofASME Section XI will be implemented in accordance with 10 CFR 50.55(a). The program will also include inspections of bolted joints of diesel generator system components and of components in locations containing high humidity or moisture. In addition, the program will include inspections of the reactor vessel-to-ringgirder bolting.
Program activities address the guidance contained in EPRI TR-104213, 'Bolted Joint Maintenance and Applications Guide," but do not specifically identify its use.' Non-safety component inspections rely on detection of visible leakage during preventive maintenance and routine observation. The program does not address structural and component support bolting with the exception of the reactor vessel-to-ring girder bolting. The aging management of all other structural bolting is covered by the structures monitoring program. Aging management of ASME Section Xl Class 1, 2, and 3 and Class MC support members, including mechanical connections is covered by the "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 routine inspection and maintenance program, including system flushing and chemical treatment, (d) periodic inspections for leakage, loss of material, and blockage, (e) engineering evaluations and heat sink performance assessments, and (f) assessments of the overall heat sink program. These evaluations and assessments produced specific component and programmatic corrective actions. The program provides assurance that the open-.
cycle cooling water system is in compliance with General Design Criteria, and with quality assurance requirements, to ensure that the open-cycle cooling water system can be managed for an extended period of operation. This program is in response to and uses the test and inspection guidelines of NRC Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment." Prior to the period of extended operation, the scope of the program will be increased to include inspection of an additionalstrainer, additional heat exchangers and sub-components, external surfaces of various submerged pumps and piping, cooling water pump linings, and components in the pump vaults that have a high humidity or moisture environment.
Dresden and Quad Cities                                                              Page A-7 License Renewal Application

Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.1.14 Closed-Cycle Cooling Water System The closed-cycle cooling water system aging management program relies on preventive measures 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 Water Chemistry Guidelines," for closed-cycle cooling water systems. Station maintenance inspections and NDE provide condition monitoring of heat exchangers exposed to closed-cycle cooling water environments. Prior to the period of extended operation, the program will be enhanced to include procedure revisions that provide for monitoring of specific 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) handling systems aging management program confirms the effectiveness of the maintenance monitoring program and the effects of past and future usage on the structural reliability of cranes and hoists. Administrative controls ensure that only allowable loads are handled, and fatigue failure of structural elements is not expected. A time-limited aging analysis concludes that there are no fatigue concerns for reactor building overhead cranes during the period of extended operation. The bridge, trolley, and other structural components are visually inspected on a routine basis for degradation. These cranes are included in the corporate structural monitoring program (which complies with the 10 CFR 50.65 maintenance rule) and in various station procedures. Prior to the period of extended operation, the program will be enhanced to include inspections for rail wear and 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, visual inspections, and walkdowns of various system locations; and (2) preventive monitoring that checks air quality at various locations in the system to ensure that dewpoint, particulates, and suspended hydrocarbons are kept within the specified limits. This program is consistent with responses to NRC Generic Letter 88-14, "Instrument Air Supply 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 include inspections of instrument air distribution piping based on EPRI TR-108147,
'Compressor and Instrument Air System Maintenance Guide," and blowdown of instrument air distribution piping.
A.1.17 BWR Reactor Water Cleanup System The BWR reactor water cleanup (RWCU) system aging management program monitors and controls reactor water chemistry based on industry-recognized guidelines of EPRI Dresden and Quad Cities                                                         Page A-8 License Renewal Application

Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT TR-1 03515, UBWR Water Chemistry Guidelines," to reduce the susceptibility of RWCU piping to stress corrosion cracking (SCC) and intergranular stress corrosion cracking (IGSCC). RWCU system piping has been replaced with piping that is resistant to intergranular stress corrosion cracking, in response to NRC Generic Letter 88-01, "NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping," concerns. In addition, all actions requested in NRC Generic Letter 89-10, 'Safety-Related Motor-Operated Valve Testing and Surveillance,' have been completed. Therefore, inservice inspection in accordance with ASME Section Xl is not required.
A.1.18 Fire Protection The fire protection aging management program includes a fire barrier inspection program and a diesel-driven fire pump inspection program. The fire barrier inspection program requires periodic visual inspection of fire barrier penetration seals; and fire wraps and fire proofing; fire barrier walls, ceilings, and floors; flood barrier penetration seals that also serve as fire barrier seals; and periodic visual inspection and functional tests of fire rated doors to ensure that their operability is maintained. The program includes surveillance tests of fuel oil systems for the diesel-driven fire pumps and isolation condenser diesel-driven makeup pumps to ensure that the fuel supply lines can perform intended functions. The program also includes visual inspections and periodic operability tests of halon and carbon dioxide fire suppression systems based on NFPA codes.
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 dioxide system
* Periodic capacity tests of the isolation condenser makeup pumps
* Specific fuel supply leak inspection criteria for fire pumps and isolation condenser 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 and hydrant flushing, system performance (flow and pressure) testing, and inspections, on a periodic basis, and for injection of chemical agents during or subsequent to flushing to minimize biofouling. System performance tests measure hydraulic resistance and compare results with previous testing. This approach eliminates the need for tests at maximum design flow and pressure. Internal inspections are conducted on system Dresden and Quad Cities                                                            Page A-9 License Renewal Application

Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT components when disassembled to identify evidence of corrosion or biofouling. Fire header pressure is maintained through a crosstie with the service water system.
Significant leakage (exceeding the capacity of this line) would be identified by automatic start of the fire pumps, which would initiate immediate investigation and corrective action. Inspection and surveillance testing is performed in accordance with procedures based on applicable NFPA codes. Where code deviations are required or desirable, the intent of the code is maintained by documented technical justifications.
Sprinkler test requirements will be modified prior to the period of extended operation to include sprinkler sampling in accordance with NFPA 25, Inspection, Testing and Maintenance of Water-Based Fire Protection Systems," Section 2-3.1. Samples will be submitted to a testing laboratory prior to being in service 50 years. This testing will be repeated at intervals not exceeding 10 years.
Prior to the period of extended operation the program will be revised to include external surface inspections of submerged fire pumps, outdoor hydrants, and outdoor transformer deluge systems; and periodic non-intrusive wall thickness measurements of selected 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 of outdoor nitrogen tanks and aluminum storage tanks. Paint is a corrosion preventive measure, and periodic visual inspections monitor degradation of the paint and any resulting metal degradation. Carbon steel tanks in the scope of license renewal are above ground and not directly supported by earthen or concrete foundations. Therefore, inspection of the sealant or caulking at the tank-foundation interface, and inspection of inaccessible tank locations and on-grade tank bottoms do not apply.
Aluminum storage tanks included within the scope of license renewal are supportedby earthen/concrete foundations. Sealants at the tank-foundation interfaces for these tanks are periodicallyinspected for degradation. Periodic internal/externalinspectionsof the aluminum tanks for pittingand crevice corrosion willbe performed at a frequency not to exceed once every five years.
UT wall thickness inspections willbe performed on the tank bottoms of all aluminum tanks included within the scope of license renewal at a frequency not to exceed once every 10 years.
Prior to the period of extended operation the program will be revised to include documentation of results of periodic system engineer walkdowns of the nitrogen tanks and periodic visual and ultrasonicinspectionsof the internal/external surfaces of the aluminum storage tanks.
A.1.21 Fuel Oil Chemistry The fuel oil chemistry aging management program relies on a combination of surveillance and maintenance procedures. Monitoring and controlling fuel oil contamination maintains the fuel oil quality. Exposure to fuel oil contaminants such as Dresden and Quad Cities                                                          Page A-10 License Renewal Application

Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT water and microbiological organisms is minimized by routine draining and cleaning of fuel oil tanks, and by fuel oil sampling and analysis, including analysis of new fuel before its introduction into the storage tanks. A biocide is added to the fuel oil storage tanks during each new fuel delivery. Sampling and testing of diesel fuel oil is in accordance with industry recognized ASTM methods and standards ASTM D2709, ASTM D4057 and ASTM D5452. Emergency diesel generator fuel oil analysis acceptance criteria are contained in the Technical Specifications and are based on the requirements of ASTM methods and standards D975.
A.1.22 Reactor Vessel Surveillance The reactor vessel surveillance aging management program includes periodic testing of metallurgical surveillance samples to monitor the progress of neutron embrittlement of the reactor pressure vessel as a function of neutron fluence, in accordance with Regulatory Guide (RG) 1.99, "Radiation Embrittlement of Reactor Vessel Materials,"
Revision 2.
Prior to the period of extended operation the program will be consistent with BWRVIP-78, uintegrated Surveillance Program," and 86-A, "BWR Integrated Surveillance Program Implementation Plan," for 32 EFPY only.
Dresden willimplement BWRVIP-116 "Integrated Surveillance Program (ISP)
Implementation for License Renewal," if approved by the NRC. If BWRVIP-116 is not approved, Exelon will provide a plant-specificsurveillance plan for the LR periodin accordance with 10 CFR 50, Appendices G and H.
The program will ensure coupon availability during the period of extended operation, and provide for saving withdrawn coupons for future reconstitution.
A.1.23 One-Time Inspection The one-time inspection aging management program includes inspections of a number of samples of the piping and components listed below. The inspections are scheduled for implementation prior to the period of extended operation to manage aging effects of selected components within the scope of license renewal. The purpose of the inspection is to determine if a specified aging effect is occurring. If the aging effect is occurring, an evaluation is performed to determine the effect it will have on the ability of affected components to perform their intended functions for the period of extended operation, and appropriate corrective action is taken. The program includes the following one-time inspections:
* Volumetric examination of 10% of the high and medium risk butt welds of Class I piping less than four inch nominal pipe size (NPS) exposed to reactor coolant for cracking.
* Inspection of a sample of torus saddle Lubrite baseplates for galvanic corrosion, wear, and lockup to confirm the condition of the inaccessible drywell radial beam Lubrite baseplates.
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Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT
* Inspection of a sample of spent fuel pool cooling and demineralizer system components for corrosion in stagnant locations to verify effective water chemistry controls.
* Inspection a samle of piping exposed to the Gontainment atmosphere (6afety relief valve discharge piping and HPCI turbine exhaust sample locations) for loss of mateial.
* Inspection of a sample of condensate and torus water components for corrosion and/or stress corrosion cracking in stagnant locations to verify effective water chemistry control.
* Inspection of a sample of compressed gas system piping components for corrosion aRd a sample of compressed gas system flexible hoses.
* Inspection of a sample of lower sections of carbon steel fuel oil and lubricating oil tanks for reduced thickness.
* Inspection of a sample of fuel oil and lubricating oil piping and components for corrosion.
* Inspection of a sample of standby gas treatment and ventilation system components for loss of material.
* Inspection of a sample of stainless steel standby liquid control (SBLC) system components not in the reactor coolant pressure boundary of the SBLC system for cracking, to verify effective water chemistry control.
* Inspection of a sample of HPCI turbine lubricating oil hoses for age-related degradation.
* Inspection of a sample of non-safety related vents and drains including their valves and associated piping, for age-related degradation leading to a loss of structural integrity.
* Inspection of a sample of 10 CFR 54.4(a)(2) components for corrosion for which the component, material, environment, aging effect, or their combination is not specifically identified in NUREG-1801, "Generic Aging Lessons Learned (GALL)
Report."
A.1.24 Selective Leaching of Materials The selective leaching of materials aging management program includes numerous one-time inspections of components of the different susceptible materials selected from each of the applicable environments to determine if loss of material due to selective leaching is occurring. These inspections will consistof visual inspection consistentwith ASME Section Xl VT-1 visual inspectionrequirements. If selective leaching is occurring the program requires evaluation of the effect it will have on the ability of the affected components to perform their intended functions for the period of Dresden and Quad Cities                                                           Page A-12 License Renewal Application

Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT extended operation, and of the need to expand the test sample. For systems subjected to environments where water is not treated (i.e., the open-cycle cooling water system) the program also follows the guidance of NRC Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment." NUREG-1801 indicates that the selective leaching of materials aging management program includes one-time hardness measurements of a selectedset of components. Visual inspections supplemented by other examinations in lieu of hardness tests of the selected set of components will be performed.
A.1.25 Buried Piping and Tanks Inspection The buried piping and tanks inspection aging management program includes (1) preventive measures to mitigate corrosion, and (2) periodic inspection to manage the effects of corrosion on the pressure-retaining capacity of buried carbon steel piping and tanks. The program includes the use of piping and component coatings and wrappings, periodic pressure testing, buried tank leakage checks, inspections of buried tank interior surfaces, and inspections of the ground above buried tanks and piping.
Prior to the period of extended operation a one-time visual inspection of the external surface of a buried piping section, and a one-time internal ultrasonic inspection of a sampling of the buried steel tanks, and a one-time internal ultrasonic inspection on the bottom of an outdoor aluminum storage tank will be performed.
A.1.26 ASME Section Xl, Subsection IWE The ASME Section Xl, Subsection IWE aging management program consists of periodic visual examination for signs of degradation, and limited surface or volumetric examination when augmented examination is required. The program covers steel containment shells and their integral attachments; containment hatches and airlocks; seals, gaskets and moisture barriers; and pressure-retaining bolting. The program includes assessment of damage and corrective actions. The program utilizes an approved relief request that permits utilization of the 1998 Edition of Subsection IWE of ASIVE Section XI in its entirety instead of the 1992 Edition and Addenda.- The requirements of ASME Section Xi willbe implemented in accordance with 10 CFR 50.55(a).
A.1.27 ASME Section Xl, Subsection IWF The ASME Section Xl, Subsection IWF aging management program consists of periodic visual examination of ASME Section Xl Class 1, 2, and 3 component and piping supports for signs of degradation, evaluation, and establishment of corrective actions.
The program is in accordance with ASSME Section XI, Subsection IWF, 1989 Edition, and Code Case N 191 1. The requirements of ASME Section Xl will be implemented in accordance with 10 CFR 50.55(a). Prior to the period of extended operation the program will include ASME Class MC component supports consistent with NUREG-1 801, "Generic Aging Lessons Learned (GALL) Report," Chapter 1II, Section 61.3.
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Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.1.28 10 CFR Part 50, Appendix J The 10 CFR Part 50, Appendix J aging management program monitors leakage rates through the containment pressure boundary, including the drywell and torus, penetrations, fittings, and other access openings, in order to detect degradation of containment pressure boundary. Corrective actions are taken if leakage rates exceed acceptance criteria. The Appendix J program also manages changes in material properties of gaskets, o-rings, and packing materials for the containment pressure boundary access points. The containment leak rate tests are performed in accordance with the regulations and guidance provided in 10 CFR 50 Appendix J Option B, Regulatory Guide 1.163, 'Performance-Based Containment Leak-Testing Program,"
NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50 Appendix J," and ANSI/ANS 56.8, "Containment System Leakage Testing Requirements."
A.1.29 Masonry Wall Program This masonry wall aging management program consists of inspections, based on IE Bulletin 80-11, "Masonry Wall Design," and plant-specific monitoring proposed by IN 87-67, "Lessons Learned from Regional Inspections of Licensee Actions in Response to IE Bulletin 80-11," for managing cracking of masonry walls. This program is part of the structures monitoring program.
A.1.30 Structures Monitoring Program The structures monitoring aging management program includes periodic inspection and monitoring of the condition of structures; supports not included in the "ASME Section Xl, Subsection IWF" aging management program; and external surfaces of mechanical and electrical components. The program ensures that aging degradation leading to loss of intended functions will be detected and that the extent of degradation can be determined. This program was developed under 10 CFR 50.65 and is based on NUMARC 93-01, 'Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Revision 2 and Regulatory Guide 1.160, 'Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Revision 2.
Prior to the period of extended operation the program will be revised to include:
* Inspections of structural steel components in secondary containment, flood barriers, electrical panels and racks, junction boxes, instrument panels and racks, and offsite power structural components and their foundations.
* Periodic reviews of chemistry data on below-grade water to confirm that the environment remains non-aggressive for aggressive chemical attack of concrete or corrosion of embedded steel.
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Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT
* Inspection of a sample of non-insulated indoor piping external surfaces at locations immediately adjacent to periodically inspected piping supports.
* Reference to specific insulation inspection criteria for existing cold weather preparation and inspection procedures for outdoor insulation, and the establishment of new inspections for various indoor area piping and equipment insulation.
* Addition of specific inspection parameters for non-structural joints, roofing, grout pads and isolation gaps.
* Extension of inspection criteria to the structural steel, concrete, masonry walls, equipment foundations, and component support sections of the program.
A.1.31     RG 1.127, lnspection of Water-Control Structures Associated with Nuclear Power Plants The RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants," aging management program consists of inspection and surveillance of structural steel elements (exposed to raw water) and concrete (exposed and not exposed to raw water) that are in the Unit 1 and Unit 2 and 3 crib houses, the discharge outfall structure and within the scope of license renewal and the earthen walls of the intake and discharge canals. The activities are based on Regulatory Guide 1.127, Revision 1, and are part of the structures monitoring program. Prior to the period of extended operation the program will be revised to include monitoring discharge outfall structure concrete walls and slabs and crib house concrete walls and slabs with opposing sides in contact with river water, to emphasize inspection for structural integrity of concrete and steel components, and to identify specific types of components to be inspected.
A.1.32 Protective Coatinq Monitoring and Maintenance Program The protective coating monitoring and maintenance aging management program consists of guidance for selection, application, inspection, and maintenance of Service Level I protective coatings. This program is implemented in accordance with Regulatory Guide 1.54, "Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants," Revision 0, ANSI N101 4-1972, "Quality Assurance for Protective Coatings Applied to Nuclear Facilities," and the guidance of EPRI TR-109937, "Guidelines on Nuclear Safety-Related Coating." Prior to the period of extended operation the program will be revised to include thorough visual inspection of Service Level 1 coatings near sumps or screens for the emergency core cooling system, pre-inspection review of previous reports so that trends can be identified, and analysis of suspected causes of any coating failures.
A.1.33 Electrical Cables and Connections Not Subiect to 10 CFR 50.49 Environmental Qualification Requirements The electrical cables and connections not subject to 10 CFR 50.49 environmental qualification requirements aging management program manages aging of cables and Dresden and Quad Cities                                                           Page A-15 License Renewal Application

Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT connections which might be susceptible to aging during the period of extended operation. A            t6ample-o All accessible electrical cables and connections installed in adverse localized environments are visually inspected at least once every 10 years for indications of accelerated insulation aging. An adverse localized environment is a condition in a limited plant area that is significantly more severe than the specified service environment for a subject electrical cable or connection. This is a new program initiated prior to the period of extended operation.
A.1.34 Metal Fatigue of Reactor Coolant Pressure Boundary The metal fatigue and reactor coolant pressure boundary aging management program ensures that the design fatigue usage factor limit will not be exceeded during the period of extended operation. The program will be enhanced prior to the period of extended operation. The enhanced program calculates and tracks cumulative usage factors for bounding locations in the reactor coolant pressure boundary (reactor pressure vessel and Class I piping), containment torus, torus vents, and torus attached piping and penetrations. The program also tracks isolation condenser fatigue stress cycles. The enhanced program uses the EPRI-licensed FatiguePro cycle counting and fatigue usage factor tracking computer program, which provides for calculation of stress cycles and fatigue usage factors from operating cycles, automated counting of fatigue stress cycles, and automated calculation and tracking of fatigue cumulative usage factors.
A.1.35 Environmental Qualification (EQ) of Electrical Components The effects of aging on the intended functions will be adequately managed per the requirements of 10 CFR 54.21 (c)(1)(iii). The existing environmental qualification (EQ) program will manage aging of electrical equipment within the scope of 10 CFR 50.49, "Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants," for the period of extended operation. The program establishes, demonstrates, and documents the level of qualification, qualified configurations, maintenance, surveillance and replacements necessary to meet 10 CFR 50.49. A qualified life is determined for equipment within the scope of the program and appropriate actions such as replacement or refurbishment are taken prior to or at the end of the qualified life of the equipment so that the aging limit is not exceeded.
A.1.36 Not Used A.1.37 Electrical Cables Not Subiect to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrument Circuits The cables of the Nuclear Instrumentationsystems which include Source Range Monitors (SRMs), Intermediate Range Monitors (IRMs), Local Power Range Monitors (LPRMs), and the Radiation Monitoringsystems which include Drywell High Range Radiation Monitors, Main Steam Line Radiation Monitors, and the Steam Jet Air Ejector Radiation Monitors are sensitive instrumentationcircuits with low-levelsignals and are located in areas where the cables could be exposed to adverse localized environments caused by heat, radiation, or moisture. These Dresden and Quad Cities                                                            Page A-16 License Renewal Application

Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT adverse localized environments can result in reduced insulation resistance causing increases in leakage currents. Calibration testing, cable testing or surveillance testing is performed to ensure that the cable insulation resistance is adequate for the instrumentation circuits to perform their intended functions.
This provides sufficient indication of the need for corrective actions based on acceptance criteria related to instrumentation loop performance and cable testing. This aging management program is a new program. The calibration testing, cable testing and surveillance testing that will be used for this program are performed currently, and are effective in identifying the existence of age related degradation. The program will be implementedprior to the period of extended operation and will include a review of the calibration and surveillance results for cable aging degradation.
A.1.38 Inaccessible Medium-Voltage Cables Not Subiect to 10 CFR 50.49 Environmental Qualification Requirements Five inaccessible medium-voltage cables not subject to 10 CFR 50.49 environmental qualification requirements that feed the Dresden service water pumps will be managed by this program. These cables may at times be exposed to moisture and are subjected to system voltage for more than 25% of the time.
The cables will be tested at least once every 10 years to provide an indication of the condition of the conductor insulation. The first tests will be completed prior to the period of extended operation. The cables will be tested with a proven test for detecting deterioration of the insulation system due to wetting, such as power factor, partial discharge, or polarization index, as described in EPRI TR-103834-P1-2, or other testing that is state-of-the-art at the time the test is performed. The end of the duct bank at the crib house will be inspected annually to verify that the crib house end of the duct run is not plugged with debris.
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Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.2      PLANT-SPECIFIC AGING MANAGEMENT PROGRAMS A.2.1 Corrective Action Program The 10 CFR Part 50, Appendix B program provides corrective actions, confirmation processes, and administrative controls for aging management programs for license renewal. Prior to the period of extended operation the scope of the program will be expanded to include non-safety related structures and components that are subject to an aging management review for license renewal. The corrective action program applies to all plant systems, structures and components (both safety related and non-safety related) within the scope of license renewal. Administrative controls are in place for existing aging management programs and activities. Administrative controls will also be applied to new and enhanced programs and activities as they are implemented. As a minimum, these programs and activities are or will be performed in accordance with written procedures that are or will be reviewed and approved in accordance with the Quality Assurance Program.
A.2.2 Periodic Inspection of Non-EQ, Non-Segregated Electrical Bus Ducts This program inspects the non-segregated bus ducts that connect the reserve auxiliary transformer to the 4160V ESF Engineering Safety Systems (ESS) buses for signs of aging degradation that indicate possible loss of intended function. This program willbe enhanced prior to the period of extended operation to inspect the bus bar insulationmaterial at the accessible bolted connections of the non-segregated bus ducts that connect the reserve auxiliarytransformer to the 4160V ESS buses and inspect 10% of the splice insulation material at the bolted connectionsfor the non-segregated bus ducts that connect the EDG to the ESS buses and the non-segregated bus duct that connects the ESS buses (cross-tie) for signs of aging degradation that indicate possible loose connections. For non-segregated bus ducts that connect the EDGs to the ESS buses and the non-segregated bus duct that connects the ESS buses, the enhancement will also include inspections for the presence of dirt or moisture in the bus ducts. The visualinspection will include all visible insulationin both directionsbeyond the location of the bolted connectionsplice insulationinspected. They arc normally energized, and therefore the bus duct insulation material will experience temperature rise due to energization, which may cause age related degradation during the period of extended operation.
These bus ducts are in scope of license renewal but are not subject to 10 CFR 50.49 environmental qualification requirements.
An inspection program will be established prior to the period of extended operation. The program will provide for inspection of the bus ducts. This inspection program considers the technical information and guidance provided in IEEE Standard P1205, 'IEEE Guide for Assessing, Monitoring and Mitigating Aging Effects on Class 1E Equipment Used in Nuclear Power Generating Stations," SAND 96-0344, "Aging Management Guideline for Commercial Nuclear Power Plants - Electrical Cable and Terminations," and EPRI TR-1 09619, "Guideline for the Management of Adverse Localized Equipment Environments." Non segregated bus duct internal components and materials will be inspected for sign of aging degradation that indicate possible loss of insulation function.
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Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT Repair or rework is initiated as required to maintain the operating functions of the bus dUGtS.
A.2.3 Periodic Inspection of Ventilation System Elastomers The periodic inspection of ventilation system elastomers aging management program provides for routine inspections of certain elastomers in the standby gas treatment, reactor building ventilation, station blackout diesel generator building ventilation, and main control room ventilation systems. Prior to the period of extended operation an existing program for inspection of ventilation system elastomers will be enhanced. The program will include inspections for cracking, loss of material, or other evidence of aging of all flexible boots, access door seals and gaskets, and filter seals and gaskets in the components of these systems that are within the scope of license renewal. The scope of inspections will also include RTV silicone used as a duct sealant, in systems within the scope of license renewal.
A.2.4 Periodic Testing of Drvwell and Torus Spray Nozzles Carbon steel piping upstream of the drywell and torus spray nozzles is subject to possible general corrosion. The periodic flow tests of drywell and torus spray nozzles address a concern that rust from the possible general corrosion may plug the spray nozzles. These periodic tests verify that the drywell and torus spray nozzles are free from plugging that could result from corrosion product buildup from upstream sources.
A.2.5 Lubricating Oil Monitoring Activities The lubricating oil monitoring activities aging management program manages corrosion, loss of material, and cracking in lubricating oil heat exchangers and other specific components in the scope of license renewal by monitoring physical and chemical properties in lubricating oil. Sampling, testing, and monitoring verify lubricating oil properties. Oil analysis permits identification of specific wear mechanisms, contamination, and oil degradation within operating machinery.
These activities apply to the emergency diesel generator system, station blackout diesel generator system, and-HPCI system oil oolers, and electro-hydrauliccontrol (EHC) system. The complete aging management program for there-the emergency diesel generator oil coolers, station blackout diesel generator oil coolers, and HPCI oil coolers also includes secondary-side (heat sink) chemistry controls, performance monitoring, and inspections. Those portions of the lubricating oil heat exchanger management program are described in:
* Section A.1.14, 'Closed-Cycle Cooling Water System," for the diesel generator and station blackout diesel generator oil coolers; and in
* Section A.2.6, "Heat Exchanger Test and Inspection Activities," for the HPCI oil coolers.
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Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.2.6 Heat Exchanger Test and Inspection Activities The heat exchanger test and inspection activities aging management program provides condition monitoring, inspection, and performance testing to manage loss of material, cracking, and buildup of deposits in heat exchangers in the scope of license renewal, that are not tested and inspected under 'Open-Cycle Cooling Water" or "Closed-Cycle Cooling Water" aging management programs. For the isolation condensers this program also includes the augmentation activities identified in NUREG-1 801, 'Generic Aging Lessons Learned (GALL) Report," lines IV.Cl .4-a and IV.C1.4-b.
These are new activities that will be implemented prior to the period of extended operation.
The isolation condenser test and inspection augmentation activities detect cracking due to stress corrosion cracking or cyclic loading, and detect loss of material due to pitting and crevice corrosion. These are ISI augmentation activities, outside the ISI program, not augmented ISI activities within the ISI program. These augmentation activities verify that significant degradation is not occurring, and therefore that the intended function of the isolation condenser is maintained during the extended period of operation. These augmentation activities consist of temperature and radioactivity monitoring of the shell-side (cooling) water, and eddy current testing of the tubes, and visual inspectionsof the channel head, tube sheets, and internal surfaces of the shell.
These activities include tests, inspections, and monitoring and trending of test results to confirm that aging effects are managed. To ensure that system and component functions are maintained, these components are also being included in the scope of other activities, which provide inservice inspection and performance monitoring, and primary and secondary-side (water and oil) chemistry controls.
* Inservice inspection is described in Section A.1.1, "ASME Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD."
* Management of water chemistry is described in Section A.1.2, "Water Chemistry."
* Management of the primary, oil side of the HPCI lubricating oil coolers is described in Section A.2.5, "Lubricating Oil Monitoring Activities."
A.2.7 Not Used A.2.8 Periodic Inspection of Plant Heating System The periodic inspection of plant heating system aging management program provides for routine inspectionsof selected components in the plant heating system. Prior to the period of extended operation, a new program for periodic inspection ofselected components in the plant heating system will be implemented. The selected components will be inspected to ensure they are free Dresden and Quad Cities                                                          Page A-20 License Renewal Application

Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT of cracking, loss of material and leakage. The inspection will consist of a visual inspection for the presence of general, crevice, galvanic, and pitting corrosion.
Where practical, the selected components will include components in stagnant flow areas that are most susceptible to loss of material. The inspection will be performed in accordance with ASME Code requirements. Certified NDE examiners will conduct a VT-3 visual inspection.
A.2.9 Periodic Inspection of Components Subject to Moist Air Environments.
The periodic inspection of components subject to moist air environments aging management program provides for periodic inspections of selected components exposed to moist air environments and subject to wetting conditions based on system operation. Prior to the period of extended operation, a new program for periodic inspection of selected components will be implemented. The inspection will consist of UT examinations of components with interior surfaces that are inaccessible and visual inspection (VT-3) of components with accessible interior surfaces for the presence of loss of material due to general corrosion, pitting and crevice corrosion. The inspection will be performed in accordance with ASME Code requirements. Certified NDE examiners will conduct the UT and VT-3 visual inspection. In addition, visual inspection of flexible hoses will determine any age-related degradation prior to loss of function.
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Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.3 Time-Limited Aging Analysis Summaries In the descriptions of this section, Class I and Class II are the Dresden safety classifications described in UFSAR Section 3.2.
A.3.1     Neutron Embrittlement of the Reactor Vessel and Internals The ferritic materials of the reactor vessel are subject to embrittlement due to high energy neutron exposure. Reactor vessel neutron embrittlement is a TLAA.
A.3.1.1      Reactor Vessel Materials Upper-Shelf Energy Reduction Due to Neutron Embrittlement The reactor vessel end-of-life neutron fluence has been recalculated for a 60-year (54 EFPY) extended licensed operating period.
The 54 EFPY USE was evaluated by an equivalent margin analysis (EMA) using the 54 EFPY calculated fluence and the Dresden surveillance capsule results, in accordance with the requirements of 10 CFR 54.21 (c)(1)(ii).
A.3.1.2      Adjusted Reference Temperature for Reactor Vessel Materials Due to Neutron Embrittlement The reactor vessel materials peak fluence, ARTNDT, and ART values for the 60-year (54 EFPY) license operating period were calculated in accordance with the requirements of 10 CFR 54.21(c)(1)(ii).
A.3.1.3      Reflood Thermal Shock Analysis of the Reactor Vessel The effects of a reflood thermal shock described in UFSAR Section 3.9.5.3.3 were examined. An alternative analysis confirms that the effects remain acceptable for the period of extended operation, in accordance with the requirements of 10 CFR 54.21 (c)(1)(i).
A.3.1.4      Reflood Thermal Shock Analysis of the Reactor Vessel Core Shroud and Repair Hardware Radiation embrittlement may affect the ability of reactor vessel internals, particularly the core shroud and repair hardware, to withstand a low-pressure coolant injection (LPCI) thermal shock transient. Embrittlement effects are evaluated for the maximum-fluence beltline region of the core shroud, where the maximum event strain is about 0.57 percent [UFSAR Section 3.9.5.3.2], and design of the core shroud repair tie rod stabilizer assemblies included an investigation of possible embrittlement effects.
The effects of the increase in neutron fluence with a 54 EFPY life at uprated power were evaluated, and the allowable strain for this faulted event remains a considerable margin above the expected strain.
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Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT The core shroud repair tie rod stabilizer assemblies were designed for a 40-year life, which will not be exceeded at the end of the extended licensed operating period.
The existing analyses of the effects of embrittlement in the internals have been evaluated and remain valid for the period of extended operation, in accordance with the requirements of 10 CFR 54.21 (c)(1)(i).
A.3.1.5       Reactor Vessel Thermal Limit Analyses: Operating Pressure - Temperature Limits Revised pressure-temperature (P-T) limits for a 60-year licensed operating life will be prepared using available Dresden capsule data and submitted to the NRC for approval prior to the start of the extended period of operation, in accordance with the requirements of 10 CFR 54.21(c)(1)(ii).
A.3.1.6       Reactor Vessel Circumferential Weld Examination Relief Relief has been granted from the requirements for inspection of RPV circumferential welds for the remainder of the current 40-year licensed operating period. The justification for relief is consistent with Boiling Water Reactor Vessel and Internals Program BWRVIP-05, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations," guidelines. Application for an extension of this relief for the 60-year period of extended operation has been will be submitted. prior to the end of the current operating license term.
The procedures and training that will be used to limit the frequency of cold over-pressure events to the number specified in the SER for the RPV circumferential weld relief request extension, during the license renewal term, are the same as those approved for use in the current period (Ref. 3).
The analyses associated with reactor vessel circumferential weld examination relief will be projected to the end of the period of extended operation, in accordance with the requirements of 10 CFR 54.21(c)(1)(ii).
A.3.1.7      Reactor Vessel Axial Weld Failure Probability BWRVIP-05, 'BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations," estimated the 40-year end-of-life failure probability of a limiting reactor vessel axial weld, showed that it was orders of magnitude greater than the 40-year end-of-life circumferential weld failure probability, and used this analysis to justify relief from inspection of the circumferential welds, as described in Section A.3.1.6 above.
The re-evaluation of the axial weld failure probability for 60 years depends on vessel ARTNDT calculations. The NRC staff review and the NRC staff and BWRVIP calculations of the test-case failure probabilities assume that 90 percent of axial welds will be inspected. At Dresden, less than 90 percent of axial welds can be inspected. As such, an analysis was performed for 54 EFPY to assess the effect on the probability of fracture due to the actual inspection performed on the vessel axial welds and to Dresden and Quad Cities                                                           Page A-23 License Renewal Application

Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT determine if the coverage was sufficient in the inspection of regions contributing to the majority of the risk.
The evaluation shows that the calculated unit-specific axial weld conditional failure probabilities at 54 EFPY for Dresden are less than the failure probabilities calculated by the NRC staff in the NRC BWRVIP-05 SER at 64 EFPY and the limiting Clinton values found in Table 3 of the SER supplement. The projected probability of failure of an axial weld at Dresden will therefore provide adequate margin above the probability of failure of a circumferential weld, in support of relief from inspection of circumferential welds, for the extended licensed operating period, in accordance with the requirements of 10 CFR 54.21 (c)(1)(ii).
A.3.2    Metal Fatigue The thermal and mechanical fatigue analyses of mechanical components have been identified as TLAAs for Dresden. Specific components have been designed considering transient cycle assumptions, as listed in vendor specifications and the Dresden UFSAR.
A.3.2.1     Reactor Vessel Fatigue Analyses Unit 2 and Unit 3 reactor vessel fatigue analyses depend on cycle count assumptions that assume a 40-year operating period. The effects of fatigue in the reactor vessel will be managed for the period of extended operation by the fatigue management program for cycle counting and fatigue usage factor tracking, as described in Section A.1.34.
This aging management program will ensure that fatigue effects in vessel pressure boundary components will be adequately managed and will be maintained within code design limits for the period of extended operation, in accordance with the requirements of 10 CFR 54.21 (c)(1)(iii).
A.3.2.2     Fatigue Analysis of Reactor Vessel Internals A.3.2.2.1   High-Cycle Flow-Induced Vibration Fatigue Analysis of Jet Pump Riser Braces The original design addressed high-cycle fatigue in the internals. Except for fatigue in the Dresden Unit 2 jet pump riser braces, the original evaluation of specific components such as the core plate, top guide, jet pump assemblies, fuel supports, or control rod drive assemblies used a displacement criterion, which is not time-limited.
The Dresden Unit 2 riser braces will be repaired or replaced prior to the period of extended operation and will be qualified for the extended licensed operating period, in accordance with the requirements of 10 CFR 54.21 (c)(1)(ii).
A.3.2.3      Reactor Coolant Pressure Boundary Piping and Component Fatigue Analysis Dresden and Quad Cities                                                           Page A-24 License Renewal Application

Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.3.2.3.1    ASME Section III Class 1 Reactor Coolant Pressure Boundary Piping and Component Fatigue Analysis Other than special cases under the Mark I containment 'New Loads" program, the only piping which has received a fatigue analysis is the Dresden Unit 3 recirculation piping replaced under the IGSCC mitigation program, including some connected shutdown cooling, low pressure coolant injection, isolation condenser, and reactor water cleanup piping.
The effects of fatigue in Class I primary system piping, including the piping analyzed to ASME Section III Class 1 criteria, will be managed for the period of extended operation by the fatigue management cycle counting and fatigue usage factor tracking program, as part of the Metal Fatigue of Reactor Coolant Pressure Boundary aging management program described in Section A.1.34. The fatigue management cycle counting and fatigue usage tracking program will apply to piping whose calculated usage factor exceeds 0.4. This aging management activity will ensure that fatigue effects in pressure boundary components will be adequately managed and will be maintained within code design limits for the period of extended operation, in accordance with the requirements of 10 CFR 54.21(c)(1)(iii).
A.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 Vil Class B and C Except for the Dresden Unit 3 recirculation piping described in A.3.2.3.1, all other primary system or reactor coolant pressure boundary (RCPB) piping systems were designed to USAS B31.1, 1967 Edition, as were the safety relief valve (SRV) discharge lines inside the drywell. Neither the USAS B31.1 piping design nor the additional nuclear code and code case rules applied to this piping invoke a fatigue analysis, but USAS B31.1 does apply a stress range reduction factor based on an assumed finite number of equivalent full-range thermal cycles for the design life. The B31.1 designs are therefore TLAAs because they are part of the current licensing basis, are used to support a safety determination, and depend on a specific number of cycles which might change with a change in licensed operating life.
The assumed number of design lifetime equivalent full-range thermal cycles determines the allowable stress range (the stress range reduction factor) for design of all Class I and Class II USAS B31.1 or ASME Class 2 or 3 piping. With the exception of containment vent and process bellows, no components in the scope of license renewal designed to ASME Section III or Section Vil require design for cyclic thermal loading.
The number of thermal cycles assumed for design of Class I and 11piping has been evaluated and the existing stress range reduction factor remains valid for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(i).
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Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.3.2.3.3        Fatigue Analysis of the Isolation Condenser The isolation condensers and the supporting system piping and components were specified for 250 shutdown depressurization cycles and for 250 thermal shock events in 40 years.
A review of isolation condenser operations since 1990 and a conservative estimate of earlier condenser operations based on number of unit scrams concluded that the projected total cycle count for 60 years is well below the number of design cycles.
The analyses of the effects of thermal cycle and thermal shock events on the Dresden isolation condenser systems and components have been evaluated and remain valid for the period of extended operation, in accordance with the requirements of 10 CFR 54.21 (c)(1)(i).
A.3.2.4     Effects of Reactor Coolant Environment on Fatigue Life of Components and Piping (Generic Safety Issue 190)
Generic Safety Issue (GSI) 190 was identified by the NRC because of concerns about potential effects of reactor water environments on component fatigue life during the period of extended operation.
Prior to the period of extended operation, Exelon will perform plant-specific calculations for the applicable locations identified in NUREG/CR 6260, "Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components," for older-vintage BWR plants, to assess the potential effects of reactor coolant on component fatigue life in accordance with 10 CFR 54.21 (c)(1)(ii). The calculations of current and projected cumulative usage factors (CUFs) under this program will include appropriate environmental fatigue effect (FEN) factors from NUREG/CR 6583 and NUREG/CR 5704. Appropriate corrective action will be taken if the resulting projected end-of-life CUF values exceed 1.0.
Exelon reserves the right to modify this position in the future based on the results of industry activities currently underway, or based on other results of improvements in methodology, subject to NRC approval prior to changes in this position.
A.3.3 Environmental Qualification of Electrical Equipment Electrical equipment included in the Dresden Environmental Qualification Program, which has a specified qualified life of at least 40 years involves time-limited aging analyses for license renewal. The aging effects of this equipment will be managed in the Environmental Qualification Program discussed in Section A.1.35, 'Environmental Qualification (EQ) of Electrical Components," in accordance with the requirements of 10 CFR 54.21 (c)(1)(iii).
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Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.3.4 Containment Fatigue The Dresden Mark I containments were originally designed to stress limit criteria without fatigue analyses. However, the discovery of significant hydrodynamic loads ('new loads") caused by safety relief valve (SRV) and small, intermediate, and design basis pipe break discharges into the suppression pool required the reanalysis of the suppression chamber, vents, and attached piping and internal structures, including some fatigue analyses at limiting locations. These fatigue analyses of the suppression chamber, and its internals, and vents in each unit include assumed pressure, temperature, seismic, and SRV cycles, and combinations thereof. The scope of the analyses included the suppression chamber, the drywell-to-suppression chamber vents, SRV discharge piping, other piping attached to the suppression chamber and its penetrations, and the drywell-to-suppression chamber vent bellows.
A.3.4.1      Fatigue Analysis of the Suppression Chamber, Vents, and Downcomers For low cumulative usage factor (CUF) locations (40-year CUF < 0.4) the Dresden new loads analyses of each suppression chamber and its associated vents and downcomers have been evaluated and remain valid for the period of extended operation, in accordance with the requirements of 10 CFR 54.21 (c)(1)(i).
For higher cumulative usage factor locations in the analyses of the suppression chamber and suppression chamber vents and downcomers (40-year CUF 2 0.4) the effects of fatigue will be managed for the period of extended operation by the fatigue management cycle counting and fatigue usage factor tracking program, as described in Section A.1.34.
The fatigue management activities will ensure that fatigue effects in containment pressure boundary components are adequately managed and are maintained within code design limits for the period of extended operation, in accordance with the requirements of 10 CFR 54.21 (c)(1)(iii).
The fatigue management activities will ensure that fatigue effects in containment pressure boundary components are adequately managed and are maintained within code design limits for the period of extended operation, in accordance with the requirements of 10 CFR 54.21 (c)(1)(iii).
A.3.4.2      Fatigue Analysis of SRV Discharge Piping Inside the Suppression Chamber, External Suppression Chamber Attached Piping, and Associated Penetrations SRV discharge lines and external suppression chamber attached piping and associated penetrations were analyzed separately from the suppression chamber, vents and downcomers.
The disposition of these analyses is the same as described for the suppression chamber, vents and downcomers in Section A.3.4.1 above.
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Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.3.4.3       Drywell-to-Suppression Chamber Vent Line Bellows Fatigue Analyses A fatigue analysis of the drywell-to-suppression chamber vent line bellows was performed assuming 150 thermal and internal pressure load cycles for the 40-year life of the plant. The drywell-to-suppression chamber vent line bellows have a rated capacity of 1,000 cycles at maximum displacement.
The Dresden new loads fatigue analysis of the drywell-to-suppression chamber vent line bellows have been evaluated and remain valid for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(i).
A.3.4.4       Primary Containment Process Penetrations Bellows Fatigue Analysis The only containment process piping expansion joints subject to significant thermal expansion and contraction are those between the drywell shell penetrations and process piping. These are designed for a stated number of operating and thermal cycles.
The thermal cycle designs of Dresden containment process penetration bellows have been evaluated and remain valid for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(i).
A.3.5       Other Plant-Specific TLAAs A.3.5.1     Reactor Building Crane Load Cycles The reactor building overhead cranes in Dresden were designed to meet or exceed the design criteria of the Crane Manufacturers Association of America (CMAA)
Specification 70, 'Specifications for Electric Overhead Traveling Cranes," Class Al.
These cranes are capable of a minimum of 100,000 cycles at the full rated load of 125 tons. Correspondence with the NRC stated that over their 40-year life these cranes would most probably see fewer than 5,000 cycles at a maximum of 100 tons and a larger number of cycles at significantly less than 100 tons.
The load cycle designs of Dresden reactor building cranes have been evaluated and remain valid for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(i).
A.3.5.2     Metal Corrosion Allowances A.3.5.2.1   Degradation Rates of Inaccessible Exterior Drywell Plate Surfaces In its response to Generic Letter 87-05, 'Request for Additional Information Assessment of Licensee Measures to Mitigate and/ or Identify Degradation of Mark I Drywells,"
Commonwealth Edison evaluated the potential effects of corrosion on exterior drywell steel surfaces in the "sand pockets" of Dresden Unit 3 and found that 27 years of service remained before corrosion at the assumed rate would have a significant adverse effect on design basis stresses. The evaluation concluded that the findings were applicable to Dresden Unit 2 and Quad Cities Units I and 2 as well.
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Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT The calculation will be revised for the realistic environment and for a full 60 year design life, in accordance with 10 CFR 54.21(c)(1)(ii). A UT inspection will validate assumptions used in the calculation. These actions will be completed before the period of extended operation.
A program will be institutedfor the Dresden Unit 3 inaccessibleannulus areasto monitorpotential corrosion. Dresden Unit 3 is consideredthe limiting case for potential drywell corrosionamong the four Dresden and Quad Cities units.
The program will inspect a sample of locations in the cylindricaland upper sphericalareasof the drywell, using ultrasonicmeasurements of the drywell shell thickness made from accessible areas of the drywell interior. A baseline inspection will be performedpriorto the period of extended operation. A follow-up inspection consistingpartly of the same locations andpartly of variable locations will be conducted by the thirdrefueling outage after the baseline inspection. The follow-up inspection will be used to determine whether any corrosionis occurringand that any observed corrosionrate will not threaten drywell integrity during the extended 60-yearplant life. These augmented inspections will be added to LR program B.1.26, ASME Section Xl, Subsection IWE, for Dresden 3.
This aging managementactivity will be added to ensure that potential corrosion of the drywell liner will be adequatelymanaged for the period of extended operation, in accordancewith the requirements of 10 CFR 54.21(c)(1)(iii).
A.3.5.2.2 Galvanic Corrosion in the Containment Shell and Attached Piping Components due to Stainless Steel ECCS Suction Strainers The Dresden ECCS suction strainers have been replaced with larger strainers. The replacement strainers are stainless steel. The modification included drilling new bolt holes and enlarging the existing bolt holes in each of the existing carbon steel strainer support flanges to provide sufficient bolting for the larger replacement strainers. The holes in the carbon steel flanges are not coated to protect them from corrosion. The calculation of corrosion effects assumes a corrosion allowance of 4 mils/year and assumes a design life of 33 years, which is just short of the 60-year extended operating period.
The corrosion rate assumptions used in the calculation will be confirmed by an ultrasonic inspection prior to the period of extended operation. Based on the results of the inspection, a revised galvanic corrosion calculation will be performed to validate acceptable wall thickness to the end of the 60-year licensed operating period, in accordance with 10 CFR 54.21(c)(1)(ii). In the event that the measuredgalvanic corrosion rate will not ensure acceptable thickness to the end of the 60-year licensed operatingperiod, appropriatecorrective action will be identified and implemented to maintain the structuralintegrity of the strainerflanges.
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Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.3.5.3     Crack Growth Calculation of a Postulated Flaw in the Heat Affected Zone of an Arc Strike in the Suppression Chamber Shell The Dresden Unit 3 torus contains an area that was damaged by an arc strike. The flaw was ground smooth and evaluated by a calculation. .This calculation showed that the Dresden flaw was bounded by a Quad Cities Unit 2 arc strike, and therefore by its analysis. The Quad Cities analysis included a crack growth calculation. A further evaluation was performed in 1997 and it was determined that the flaw depth of the arc strike at Dresden was not of sufficient depth to warrant any final repairs.
The supporting Quad Cities crack growth calculation has been evaluated and remains valid for Dresden for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(i).
A.3.5.4     Radiation Degradation of Drywell Shell Expansion Gap Polyurethane Foam The steel drywell shell is largely enclosed within the structural and shielding concrete of the reactor containment building. To accommodate thermal expansion, compressible foam was used to form an expansion gap between the concrete and the drywell shell. A confirming analysis contained in the UFSAR evaluates the increase in external compressive loads on the drywell exterior, due to additional compression of this foam, for accident-condition thermal expansion of the drywell. The load depends on the stress-strain curve of the foam, and the validity of this confirming analysis of the Dresden drywells therefore depends on the stiffness of the polyurethane foam. The analysis would require validation if the foam became stiffer (higher compressive stress for the same strain) as a result of increased radiation exposure from extended plant operation.
The expected radiation exposure of the foam has been evaluated and remains below the significant damage threshold at the end of the period of extended operation. The evaluation of thermal expansion compressive loads therefore also remains valid for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(i).
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Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.3.6 References for Section A.3
: 1.      Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2, License Renewal Project, TLAA Technical Report.
Revision 0, June 2002. Prepared by Parsons Energy and Chemicals, Inc. for the General Electric Company.
: 2.      Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2, License Renewal Project, Potential TLAA Review Results Package. Revision 0, June 2002. Prepared by Parsons Energy and Chemicals, Inc. for the General Electric Company.
: 3.      Dresden Letter JMHLTR 99-0078 from J. M. Heffley (ComEd) to USNRC, Relief Request for Alternative Weld Examination of Circumferential Reactor Pressure Vessel Shell Welds. July 26, 1999. Attached to Dresden ISI Relief Request No. CR-18.
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Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT Quad Cities Units 1 and 2 Updated Final Safety Analysis Supplement Dresden and Quad Cities                                              Page A-32 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.1     AGING MANAGEMENT PROGRAMS A.1.1 ASME Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD The ASME Section XI Inservice Inspection, Subsections IWB, IWC and IWD aging management program consists of periodic volumetric and visual examinations of components for assessment, identification of signs of degradation, and establishment of corrective actions. Prior to the period of extended operation the program will be revised to be conuistent w'ith ASME SectioR Xl, 1095 Edition through the 196 .Addenda. The inspectionswill be implemented in accordance with 10 CFR 50.55(a).
Quad Cities willimplement the guidance of BWRVIP-74, "BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines," with the followingexception.
Exception: Risk Informed Inservice Inspection is implemented in lieu of ASME Section Xl requirements for portions of Class I and Class 2 systems. Quad Cities implements a staffapproved code case for inspection of the Reactor Vessel Leak Detection Line.
Technical Specificationrevisionscontainingnew P-T Curves will be submitted prior to the term of extended operation.
A.1.2 Water Chemistry The water chemistry aging management program consists of monitoring and control of water chemistry to keep peak levels of various contaminants below system-specific limits based on industry-recognized guidelines of EPRI TR-103515, 'BWR Water Chemistry Guidelines.' To mitigate aging effects on component surfaces that are exposed to water as process fluid, the chemistry programs are used to control water chemistry for impurities (e.g., chlorides, and sulfates) that accelerate corrosion.
Quad Cities will implement the general guidance provided in BWRVIP-79, "EPRI Report TR-103515-R2."
A.1.3 Reactor Head Closure Studs The reactor head closure studs aging management program includes inservice inspection (ISI). This program also includes preventive actions and inspection techniques for BWRs. Prior to the period of extended operation the program will be revised to be coRsistent with ASME Section XI, 1095 Edition through the 1996 Addenda-The requirements of ASME Section Xl willbe implemented in accordance with 10 CFR 50.55(a). The reactor head studs are not metal-plated, and have had manganese phosphate 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-approved Dresden and Quad Cities                                                           Page A-33 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT Boiling Water Reactor Vessel and Internals Project BWRVIP-48, "Vessel ID Attachment Weld Inspection and Evaluation Guidelines," and/or ASME Section XI; and (b) monitoring and control of reactor coolant water chemistry in accordance with industry-recognized guidelines of EPRI TR-1 03515, "BWR Water Chemistry Guidelines." Prior to the period of extended operation the program will be revised to be consistent with ASME Section ;l, 1995 Edition through the 1996 Addenda. The requirements of ASME Section Xl will be implemented in accordance with 10 CFR 50.55(a).
A.1.5 BWR Feedwater Nozzle The BWR feedwater nozzle aging management program includes enhancing the inservice inspections (ISI) specified in the ASME Code, Section Xl, with the recommendation of General Electric (GE) NE-523-A71-0594, "Alternate BWR Feedwater Nozzle Inspection Requirements," to perform periodic ultrasonic testing inspection 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 of previously implemented system modifications and inservice inspections that manage the aging effect of cracking in the control rod drive return line nozzles. The control rod drive return line nozzles have been capped. Inservice inspections are performed consistent with ASME Section Xl requirements. No augmented inspections in accordance with NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," or the alternative recommendations of GE NE-523-A71-0594, "Alternate BWR Feedwater Nozzle Inspection Requirements," are required. Prior to the period of extenRded operatioR the program will be revised to be consistent ws;ith ASME Section Xl, 1905 Edition through the 1996 Addenda. The requirements of ASME Section Xl will be 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 manage intergranular stress corrosion cracking (IGSCC) in boiling water reactor coolant pressure boundary piping four inches and larger nominal pipe size made of stainless steel (SS) is delineated, in part, in NUREG-0313, "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," Revision 2, BWRVIP 75, "Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules," and Nuclear Regulatory Commission (NRC) Generic Letter (GL) 88-01, "NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping", and its Supplement 1. The program includes (a) replacements and preventive measures to mitigate IGSCC and (b) inspections to monitor IGSCC and its effects. Water chemistry is monitored and maintained in accordance with industry-recognized guidelines in EPRI TR-103515, "BWR Water Chemistry Guidelines." P~i9F to the period of extennded operation the program will be revised to be consistent with ASME Section Xl, 1995 Edition through the 1906 Addenda. The requirements of ASME Section Xl will be implemented in accordance with 10 CFR 50.55(a).
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Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT Quad Cities will implement the general guidance providedin BWRVIP-75, "Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules,"
with Exception - The Relief Request submitted for the implementation of RISI indicates the Category A Welds are "subsumed into the RISI program."
A.1.8 BWR Penetrations The BWR penetrations aging management program includes (a) inspection and flaw evaluation in conformance with the guidelines of staff-approved Boiling Water Reactor Vessel and Internals Project (BWRVIP)-49, Instrument Penetration Inspection and Flaw Evaluation Guidelines," and BWRVIP-27, 'BWR Standby Liquid Control System/Core Plate Delta-P Inspection and Flaw Evaluation Guidelines," documents and (b) monitoring and control of reactor coolant water chemistry in accordance with industry-recognized guidelines of EPRI TR-103515, "BWR Water Chemistry Guidelines," to ensure the long-term integrity and safe operation of boiling water reactor vessel internal components. Prior to the period of extended operation the program will be revised to be consistent with ASME Section Xl, 1995 Edition through the 1996 Addenda. The requirements of ASME Section Xl will be implemented in accordance with 10 CFR 50.55(a).
Quad Cities will implement the guidance providedin BWRVIP-27, "BWR Standby Liquid Control System/Core Plate AP Inspection and Flaw Evaluation Guidelines,"
with the followingexception. Quad Cities implements approved ISI Relief Requests, which provides a VT-2 examination of the standby liquidcontrol system nozzle inner radius in lieu of the Code required volumetric examination.
Quad Cities will implement the guidance provided in BWRVIP-49, "Instrument Penetration Inspection and Flaw Evaluation Guidelines."
A.1.9 BWR Vessel Internals The BWR vessel internals aging management program includes (a) inspection and flaw evaluation in conformance with the guidelines of applicable and staff-approved Boiling Water Reactor Vessel and Internals Project (BWRVIP) documents, and with ASME Section XI; and (b) monitoring and control of reactor coolant water chemistry in accordance with industry-recognized guidelines of EPRI TR-103515, "BWR Water Chemistry Guidelines," to ensure the long-term integrity and safe operation of boiling water reactor vessel internal components. 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 Xi will be implemented in accordance with 10 CFR 50.55(a).
Quad Cities will implement the general guidance providedin BWRVIP-18, "BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines."
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Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT Quad Cities will implement the general guidance provided in BWRVIP-25, "BWR Core Plate Inspection and Flaw Evaluation Guidelines."
Quad Cities will implement the guidance provided in BWRVIP-26, "BWR Top Guide Inspection and Flaw Evaluation Guidelines." Additionally, Quad Cities will perform augmented inspections for the top guide similar to the inspections of control rod drive housing (CRDH) guide tubes.
Quad Cities will implement the general guidance provided in BWRVIP-38, "BWR Shroud Support Inspection an Flaw Evaluation Guidelines." Quad Cities will perform the additional inspections of the lower plenum (i.e. shroud support leg welds) when new inspection techniques and tooling are developed, incorporated into the applicable BWRVIP document(s), and approved by NRC SER.
Quad Cities will implement the general guidance provided in BWRVIP-41, "BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines." Quad Cities will perform the additional inspections of the inaccessible thermal sleeve welds when new inspection techniques and tooling are developed, incorporated into the applicable BWRVIP document(s), and approved by NRC SER.
Quad Cities will implement the general guidance provided in BWRVIP-47, "BWR Lower Plenum Inspection and Flaw Evaluation Guidelines. "
Quad Cities will implement the general guidance provided in BWRVIP-76, "BWR Core Shroud Inspection and Flaw Evaluation Guidelines."
Quad Cities will implement the general guidance provided in BWRVIP-104, "Evaluation and Recommendations to Address Shroud Support Cracking in BWRs."
A.1.10   Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless 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 of cast 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 is being implemented prior to the period of extended operation.
A.1.11    Flow-Accelerated Corrosion The flow-accelerated corrosion aging management program consists of (1) appropriate analysis and baseline inspections, (2) determination of the extent of thinning, and replacement or repair of components, and (3) follow-up inspections to confirm or quantify effects, and to take longer-term corrective actions. This program is in response to NRC Generic Letter 89-08, "Erosion/Corrosion-Induced Pipe Wall Thinning." The Dresden and Quad Cities                                                          Page A-36 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT program relies on implementation of the EPRI NSAC-202L, "Recommendations for an Effective Flow Accelerated Corrosion Program," Revision 2 guidelines. Prior to the period of extended operation the program will be revised to include main steam piping within the scope of license renewal.
A.1.12    Bolting Integrity This bolting integrity aging management program incorporates industry recommendations of EPRI NP-5769, 'Degradation and Failure of Bolting in Nuclear Power Plants," and includes periodic visual inspections for external surface degradation that may be caused by loss of material or cracking of the bolting, or by an adverse environment. Inspection of inservice inspection Class 1, 2, and 3 components is conducted in accordance with ASME Section Xl. 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 10 CFR 50.55(a). The program will also include inspections of bolted joints of diesel generator system components and of components in locations containing high humidity or moisture. In addition, the program will include inspectionsof the reactor vessel-to-ring girderbolting.
Program activities address the guidance contained in EPRI TR-104213, 'Bolted Joint Maintenance and Applications Guide," but do not specifically identify its use. Non-safety component inspections rely on detection of visible leakage during preventive maintenance and routine observation. The program does not address structural and component support bolting with the exception of the reactor vessel-to-ringgirder bolting. The aging management of all other structural bolting is covered by the structures monitoring program. Aging management of ASME Section Xl Class 1, 2, and 3 and Class MC support members, including mechanical connections, is covered by the "ASME Section Xl, Subsection IWF" aging management program.
A.1.13    ODen-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 routine inspection and maintenance program, including system flushing and chemical treatment, (d) periodic inspections for leakage, loss of material, and blockage, (e) engineering evaluations and heat sink performance assessments, and (f) assessments of the overall heat sink program. These evaluations and assessments produced specific component and programmatic corrective actions. The program provides assurance that the open-cycle cooling water system is in compliance with General Design Criteria, and with quality assurance requirements, to ensure that the open-cycle cooling water system can be managed for an extended period of operation. This program is in response to and uses the test and inspection guidelines of NRC Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment." Prior to the period of extended operation, the scope of the program will be increased to include inspection of additional heat exchangers and sub-components, external surfaces of various submerged pumps Dresden and Quad Cities                                                              Page A-37 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT and piping, cooling water pump linings, and components in 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 preventive measures 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 Water Chemistry Guidelines," for closed-cycle cooling water systems. Station maintenance inspections and NDE provide condition monitoring of heat exchangers exposed to closed-cycle cooling water environments. Prior to the period of extended operation, the program will be enhanced to include procedure revisions that provide for monitoring of specific chemistry parameters in order to meet EPRI TR-107396 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) handling systems aging management program confirms the effectiveness of the maintenance monitoring program and the effects of past and future usage on the structural reliability of cranes and hoists. Administrative controls ensure that only allowable loads are handled, and fatigue failure of structural elements is not expected. A time-limited aging analysis concludes that there are no fatigue concerns for reactor building overhead cranes during the period of extended operation. The bridge, trolley, and other structural components are visually inspected on a routine basis for degradation. These cranes are included in the corporate structural monitoring program (which complies with the 10 CFR 50.65 maintenance rule) and in various station procedures. Prior to the period of extended operation, the program will be enhanced to include inspections for rail wear and 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, visual inspections, and walkdowns of various system locations; and (2) preventive monitoring that checks air quality at various locations in the system to ensure that dewpoint, particulates, and suspended hydrocarbons are kept within the specified limits. This program is consistent with responses to NRC Generic Letter 88-14, "Instrument Air Supply 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 include inspections of instrument air distribution piping based on EPRI TR-108147,
'Compressor and Instrument Air System Maintenance Guide."
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Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.1.17     BWR Reactor Water Cleanup System The BWR reactor water cleanup (RWCU) system aging management program monitors and controls reactor water chemistry based on industry-recognized guidelines of EPRI TR-1 03515, "BWR Water Chemistry Guidelines," to reduce the susceptibility of RWCU piping to stress corrosion cracking (SCC) and intergranular stress corrosion cracking (IGSCC). RWCU system piping has been replaced with piping that is resistant to intergranular stress corrosion cracking, in response to NRC Generic Letter 88-01, "NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping," concerns. In addition, all actions requested in NRC Generic Letter 89-10, "Safety-Related Motor-Operated Valve Testing and Surveillance," have been completed. Therefore, inservice inspection in accordance with ASME Section Xl is not required.
A.1.18     Fire Protection The fire protection aging management program includes a fire barrier inspection program and a diesel-driven fire pump inspection program. The fire barrier inspection program requires periodic visual inspection of fire barrier penetration seals; and-fire wraps and fire proofing; fire barrier walls, ceilings, and floors; flood barrier penetration seals that also serve as fire barrier seals; and periodic visual inspection and functional tests of fire rated doors to ensure that their operability is maintained. The program includes surveillance tests of fuel oil systems for the diesel-driven fire pumps to ensure that the fuel supply line can perform intended functions. The program also includes visual inspections and periodic operability tests of the carbon dioxide fire suppression system based on NFPA codes.
Prior to the period of extended operation, the program will be revised to include:
I nspeioen of oil rspill barriers
* Inspection of external surfaces of the carbon dioxide systems
* Specific fuel supply leak inspection criteria for fire pumps
* Specific inspection criteria for fire doors A.1.19     Fire Water System The fire water system aging management program provides fire system header and hydrant flushing, system performance (flow and pressure) testing, and inspections, on a periodic basis; and for injection of chemical agents during or subsequent to flushing to minimize biofouling. System performance tests measure hydraulic resistance and compare results with previous testing. This approach eliminates the need for tests at maximum design flow and pressure. Internal inspections are conducted on system Dresden and Quad Cities                                                            Page A-39 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT components when disassembled to identify evidence of corrosion or biofouling. Fire header pressure is maintained through a crosstie with the service water system.
Significant leakage (exceeding the capacity of this line) would be identified by automatic start of the fire pumps, which would initiate immediate investigation and corrective action. Inspection and surveillance testing is performed in accordance with procedures based on applicable NFPA codes. Where code deviations are required or desirable, the intent of the code is maintained by technical justifications.
Sprinkler test requirements will be modified prior to the period of extended operation to include sprinkler sampling in accordance with NFPA 25, uInspection, Testing and Maintenance of Water-Based Fire Protection Systems," Section 2-3.1. Samples will be submitted to a testing laboratory prior to being in service 50 years. This testing will be repeated at intervals not exceeding 10 years.
Prior to the period of extended operation the program will be revised to include external surface inspections of submerged fire pumps, outdoor hydrants, and outdoor transformer deluge systems; and periodic non-intrusive wall thickness measurements of selected 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 of outdoor nitrogen tanks and aluminum storage tanks. Paint is a corrosion preventive measure, and periodic visual inspections monitor degradation of the paint and any resulting metal degradation. Carbon steel tanks in the scope of license renewal are above ground and not directly supported by earthen or concrete foundations. Therefore, inspection of the sealant or caulking at the tank-foundation interface, and inspection of inaccessible tank locations and on-grade tank bottoms do not apply.
Aluminum storage tanks within the scope of license renewal are supported by earthen/concrete foundations. The tank-foundationinterfaces (including foundation coatings) are periodicallyinspected for degradation. Periodic visual inspections of the internal/external surfaces of the aluminum storage tanks are conducted.
Prior to the period of extended operation, the program will be revised to include documentation of results of periodic system engineer walkdowns of the nitrogen tanks, periodic visual inspectionsof the internal/external surfaces of aluminum tanks, and a one-time internal ultrasonic inspection of the bottom of one aluminum storage tank.
A.1.21     Fuel Oil Chemistry The fuel oil chemistry aging management program relies on a combination of surveillance and maintenance procedures. Monitoring and controlling fuel oil contamination maintains the fuel oil quality. Exposure to fuel oil contaminants such as water and microbiological organisms is minimized by routine draining and cleaning of fuel oil tanks, and by fuel oil sampling and analysis, including analysis of new oil before Dresden and Quad Cities                                                          Page A-40 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT its introduction into the storage tanks. A biocide is added to the fuel oil storage tanks during each new fuel delivery. Sampling and testing of diesel fuel oil is in accordance with industry recognized ASTM methods and 6tandards D2709, ASTM D4057 and ASTM D5452. Emergency diesel generator fuel oil analysis acceptance criteria are contained in the Technical Specifications and are based on industry recognized the requirements of ASTM methods and standards D975.
A.1.22     Reactor Vessel Surveillance The reactor vessel surveillance aging management program includes periodic testing of metallurgical surveillance samples to monitor the progress of neutron embrittlement of the reactor pressure vessel as a function of neutron fluence, in accordance with Regulatory Guide (RG) 1.99, "Radiation Embrittlement of Reactor Vessel Materials,"
Revision 2.
Prior to the period of extended operation the program will be consistent with BWRVIP-78, uintegrated Surveillance Program," and 86-A, 'BWR Integrated Surveillance Program Implementation Plan," for 32 EFPY only.
Quad Cities will implement BWRVIP-116 "Integrated Surveillance Program (ISP)
Implementation for License Renewal," if approved by the NRC. If BWRVIP-116 is not approved, Exelon will provide a plant-specificsurveillance plan for the LR period in accordance with 10 CFR 50, Appendices G and H.
The program will ensure coupon availability during the period of extended operation, and provide for saving withdrawn coupons for future reconstitution.
A.1.23     One-Time Inspection The one-time inspection aging management program includes inspections of a number of samples of the piping and components listed below. The inspections are scheduled for implementation prior to the period of extended operation to manage aging effects of selected components within the scope of license renewal. The purpose of the inspection is to determine if a specified aging effect is occurring. If the aging effect is occurring, an evaluation is performed to determine the effect it will have on the ability of affected components to perform their intended functions for the period of extended operation, and appropriate corrective action is taken. The program includes the following one-time inspections:
* Volumetric examination of 10% of the high and medium risk butt welds of Class I piping less than four inch nominal pipe size (NPS) exposed to reactor coolant for cracking.
* Inspection of a sample of torus saddle Lubrite baseplates for galvanic corrosion, wear, and lockup to confirm the condition of the inaccessible drywell radial beam Lubrite baseplates.
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Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT
* Inspection of a sample of spent fuel pool cooling and demineralizer system components for corrosion in stagnant locations to verify effective water chemistry controls.
* InspeGtion a sample of piping cxposed to the containment atmosphere (safety relief valve discharge piping and HPCI turbine exhaust sample locations) for loss Of nateal.
* Inspection of a sample of condensate and torus water components for corrosion and/orstress corrosion cracking in stagnant locations to verify effective water chemistry control.
* Inspection of a sample of compressed gas system piping components for corrosion and a sample of compressed gas system flexible hoses.
* Inspection of a sample of lower sections of carbon steel fuel oil and lubricating oil tanks for reduced thickness.
* Inspection of a sample of fuel oil and lubricating oil piping and components for corrosion.
* Inspection of a sample of standby gas treatment and ventilation system components for loss of material.
* Inspection of a sample of stainless steel standby liquid control (SBLC) system components not in the reactor coolant pressure boundary of the SBLC system for cracking, to verify effective water chemistry control.
* Inspection of a sample of HPCI turbine lubricating oil hoses for age-related degradation.
* Inspection of a sample of non-safety related vents and drains including their valves and associated piping, for age-related degradation leading to a loss of structural integrity.
* Inspection of a sample of 10 CFR 54.4(a)(2) components for corrosion for which the component, material, environment, aging effect, or their combination is not specifically identified in NUREG-1 801, 'Generic Aging Lessons Learned (GALL)
Report."
A.1.24     Selective Leaching of Materials The selective leaching of materials aging management program includes numerous one-time inspections of components of the different susceptible materials selected from each of the applicable environments to determine if loss of material due to selective leaching is occurring. These inspections will consist of visual inspection consistent with ASME Section Xl VT-1 visual inspection requirements. If selective leaching is Dresden and Quad Cities                                                           Page A-42 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT occurring the program requires evaluation of the effect it will have on the ability of the affected components to perform their intended functions for the period of extended operation, and of the need to expand the test sample. For systems subjected to environments where water is not treated (i.e., the open-cycle cooling water system) the program also follows the guidance of NRC Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment." NUREG-1801 indicates that the selective leaching of materials aging management program includes one-time hardness measurements of a selected set of components. Visual inspections supplemented by other examinations in lieu of hardness tests of the selected components will be performed.
A.1.25     Buried Piping and Tanks Inspection The buried piping and tanks inspection aging management program includes (1) preventive measures to mitigate corrosion, and (2) periodic inspection to manage the effects of corrosion on the pressure-retaining capacity of buried carbon steel piping and tanks. The program includes the use of piping and component coatings and wrappings, periodic pressure testing, buried tank leakage checks, inspections of buried tank interior surfaces, and inspections of the ground above buried tanks and piping.
Prior to the period of extended operation a one-time visual inspection of the external surface of a buried piping section, and a one-time internal ultrasonic inspection of a sampling of the buried steel tanks, and a one time internal ultrasonic inspcction on the bottom of an outdoor aluminum storage tank will be performed.
A.1.26     ASME Section Xl, Subsection IWE The ASME Section Xl, Subsection IWE aging management program consists of periodic visual examination for signs of degradation, and limited surface or volumetric examination when augmented examination is required. The program covers steel containment shells and their integral attachments; containment hatches and airlocks; seals, gaskets and moisture barriers; and pressure-retaining bolting. The program includes assessment of damage and corrective actions. The program complies with ASME Section Xl, Subsection PVVE for steel containmcnts (Class MC), 1902 Edition including 1992 Addenda. The requirements of ASME Section Xl willbe implemented in accordance with 10 CFR 50.55(a).
A.1.27    ASME Section Xl, Subsection IWF The ASME Section Xl, Subsection IWF aging management program consists of periodic visual examination of ASME Section Xl Class 1, 2, and 3 component and piping supports for signs of degradation, evaluation, and establishment of corrective actions.
The program is in accordance with ASME Section Xl, Subsection WIF, 1989 Edition, and Code Case N 191 1. The requirements of ASME Section Xl will be implemented in accordance with 10 CFR 50.55(a). Prior to the period of extended operation the program will include ASME Class MC component supports consistent with Dresden and Quad Cities                                                           Page A-43 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT NUREG-1801, 'Generic Aging Lessons Learned (GALL) Report," Chapter III, Section B1.3.
A.1.28     10 CFR Part 50, Appendix J The 10 CFR Part 50, Appendix J aging management program monitors leakage rates through the containment pressure boundary, including the drywell and torus, penetrations, fittings, and other access openings; in order to detect degradation of containment pressure boundary. Corrective actions are taken if leakage rates exceed acceptance criteria. The Appendix J program also manages changes in material properties of gaskets, o-rings, and packing materials for the containment pressure boundary access points. The containment leak rate tests are performed in accordance with the regulations and guidance provided in 10 CFR 50 Appendix J Option B, Regulatory Guide 1.163, 'Performance-Based Containment Leak-Testing Program,"
NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50 Appendix J," and ANSI/ANS 56.8, 'Containment System Leakage Testing Requirements."
A.1.29     Masonry Wall Program This masonry wall aging management program consists of inspections, based on IE Bulletin 80-11, "Masonry Wall Design," and plant-specific monitoring proposed by IN 87-67, "Lessons Learned from Regional Inspections of Licensee Actions in Response to IE Bulletin 80-11," for managing cracking of masonry walls. This program is part of the structures monitoring program.
A.1.30     Structures Monitoring Program The structures monitoring aging management program includes periodic inspection and monitoring of the condition of structures; supports not included in the "ASME Section Xl, Subsection IWF" aging management program; and external surfaces of mechanical and electrical components. The program ensures that aging degradation leading to loss of intended functions will be detected and that the extent of degradation can be determined. This program was developed under 10 CFR 50.65 and is based on NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Revision 2 and Regulatory Guide 1.160, "Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Revision 2.
Prior to the period of extended operation the program will be revised to include:
* Inspections of structural steel components in secondary containment, flood barriers, electrical panels and racks, junction boxes, instrument panels and racks, offsite power structural components and their foundations, and the discharge canal weir as part of the ultimate heat sink.
* Periodic reviews of chemistry data on below-grade water to confirm that the environment remains non-aggressive for aggressive chemical attack of concrete or Dresden and Quad Cities                                                         Page A-44 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT corrosion of embedded steel.
* Inspection of a sample of non-insulated indoor piping external surfaces at locations immediately adjacent to periodically inspected piping supports.
* Reference to specific insulation inspection criteria for existing cold weather preparation and inspection procedures for outdoor insulation, and the establishment of new inspections for various indoor area piping and equipment insulation.
* Addition of specific inspection parameters for non-structural joints, roofing, grout pads and isolation gaps.
* Extension of inspection criteria to the structural steel, concrete, masonry walls, equipment foundations, and component support sections of the program.
A.1.31    RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants The RG 1.127, 'Inspection of Water-Control Structures Associated with Nuclear Power Plants," aging management program consists of inspection and surveillance of structural steel elements (exposed to raw water) and concrete (exposed and not exposed to raw water) that are in the crib house and discharge canal weir structure supporting the ultimate heat sink and within the scope of license renewal and the earthen walls of the intake and discharge flumes/canals. The activities are based on Regulatory Guide 1.127, Revision 1, and are part of the structures monitoring program. Prior to the period of extended operation the program will be revised to include monitoring crib house concrete walls and slabs with opposing sides in contact with river water, and the discharge canal weir supporting the ultimate heat sink; to emphasize inspection for structural integrity of concrete and steel components; and to identify specific types of components to be inspected.
A.1.32    Protective Coating Monitoring and Maintenance Program The protective coating monitoring and maintenance aging management program consists of guidance for selection, application, inspection, and maintenance of Service Level I protective coatings. This program is implemented in accordance with Regulatory Guide 1.54, "Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants," Revision 0, ANSI N101 4-1972, "Quality Assurance for Protective Coatings Applied to Nuclear Facilities," and the guidance of EPRI TR-1 09937, "Guidelines on Nuclear Safety-Related Coating." Prior to the period of extended operation the program will be revised to include thorough visual inspection of Service Level 1 coatings near sumps or screens for the emergency core cooling system, pre-inspection review of previous reports so that trends can be identified, and analysis of suspected causes of any coating failures.
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Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.1.33      Electrical Cables and Connections Not Subiect to 10 CFR 50.49 Environmental Qualification Requirements The electrical cables and connections not subject to 10 CFR 50.49 environmental qualification requirements aging management program manages aging of cables and connections which might be susceptible to aging during the period of extended operation. A 6ample e All accessible electrical cables and connections installed in adverse localized environments are visually inspected at least once every 10 years for indications of accelerated insulation aging. An adverse localized environment is a condition in a limited plant area that is significantly more severe than the specified service environment for a subject electrical cable or connection. This is a new program initiated prior to the period of extended operation.
A.1.34      Metal Fatigue of Reactor Coolant Pressure Boundary The metal fatigue and reactor coolant pressure boundary aging management program ensures that the design fatigue usage factor limit will not be exceeded during the period of extended operation. The program will be enhanced prior to the period of extended operation. The enhanced program calculates and tracks cumulative usage factors for bounding locations in the reactor coolant pressure boundary (reactor pressure vessel and Class I piping), containment torus, torus vents, and torus attached piping and penetrations. The enhanced program uses the EPRI-licensed FatiguePro cycle counting and fatigue usage factor tracking computer program, which provides for calculation of stress cycles and fatigue usage factors from operating cycles, automated counting of fatigue stress cycles, and automated calculation and tracking of fatigue cumulative usage factors.
A.1.35      Environmental Qualification (EQ) of Electrical Components The effects of aging on the intended functions will be adequately managed per the requirements of 10 CFR 54.21 (c)(1)(iii). The existing environmental qualification (EQ) program will manage aging of electrical equipment within the scope of 10 CFR 50.49, "Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants," for the period of extended operation. The program establishes, demonstrates, and documents the level of qualification, qualified configurations, maintenance, surveillance and replacements necessary to meet 10 CFR 50.49. A qualified life is determined for equipment within the scope of the program and appropriate actions such as replacement or refurbishment are taken prior to or at the end of the qualified life of the equipment so that the aging limit is not exceeded.
A.1.36      Boraflex Monitoring The Boraflex monitoring aging management program consists of (1) neutron attenuation testing ("blackness testing") to determine gap formation, (2) sampling for the presence of silica in the spent fuel pool along with boron loss, and (3) analysis of criticality to assure that the required 5% subcriticality margin is maintained. This program is implemented in response to Generic Letter 96-04, "Boraflex Degradation in Spent Fuel Pool Storage Racks." The Boraflex monitoring activities are based on the maintenance Dresden and Quad Cities                                                             Page A-46 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT rule and on EPRI TR-108761, "A Synopsis of the Technology Developed to Address the Boraflex Degradation Issue."
A.1.37    Electrical Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrument Circuits The cables of the Nuclear Instrumentation systems which include Source Range Monitors (SRMs), Intermediate Range Monitors (IRMs), Local Power Range Monitors (LPRMs), and the Radiation Monitoring systems which include Drywell High Range Radiation Monitors, Main Steam Line Radiation Monitors, and the Steam Jet Air Ejector Radiation Monitors are sensitive instrumentation circuits with low-level signals and are located in areas where the cables could be exposed to adverse localized environments caused by heat, radiation, or moisture. These adverse localized environments can result in reduced insulation resistance causing increases in leakage currents. Calibration testing, cable testing or surveillance testing is performed to ensure that the cable insulation resistance is adequate for the instrumentation circuits to perform their intended functions.
This provides sufficient indication of the need for corrective actions based on acceptance criteria related to instrumentation loop performance and cable testing. This aging management program is a new program. The calibration testing, cable testing and surveillance testing that will be used for this program are performed currently, and are effective in identifying the existence of age related degradation. The program will be implemented prior to the period of extended operation and will include a review of the calibration and surveillance results for cable aging degradation.
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Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.2      Plant-Specific Aging Management Programs A.2.1    Corrective Action Program The 10 CFR Part 50, Appendix B program provides corrective actions, confirmation processes, and administrative controls for aging management programs for license renewal. Prior to the period of extended operation the scope of the program will be expanded to include non-safety-related structures and components that are subject to an aging management review for license renewal. The corrective action program applies to all plant systems, structures and components (both safety-related and non-safety-related) within the scope of license renewal. Administrative controls are in place for existing aging management programs and activities. Administrative controls will also be applied to new and enhanced programs and activities as they are implemented. As a minimum, these programs and activities are or will be performed in accordance with written procedures that are or will be reviewed and approved in accordance with the Quality Assurance Program.
A.2.2 Periodic Inspection of Non-EQ, Non-Segregated Electrical Bus Ducts This program inspects the non-segregated bus ducts that connect the reserve auxiliary transformers to the 41 60V ESF-Engineering Safety Systems (ESS) buses for signs of aging degradation that indicate possible loss ofintended function. This program will be enhanced prior to the period of extended operation to inspect the bus bar insulationmaterial at the accessible bolted connections of the non-segregated bus ducts that connect the reserve auxiliarytransformer to the 4160V ESS buses and inspect 10% of the splice insulationmaterial at the bolted connections for the non-segregated bus ducts that connect the EDGs to the ESS buses for signs of aging degradation that indicate possible loose connections.
For non-segregated bus ducts that connect the EDGs to the ESS buses, the enhancement willalso include inspectionsfor the presence of dirt or moisture in the bus ducts. The visual inspection will include all visible insulationin both directions beyond the location of the bolted connectionsplice insulation inspected. They arc normally energized, and therefore the bus duct insulation material will expcriencc temperatre risc   due to energization, which mnay ause age related degradation during the period of extended operation.
These bus ducts are in scope of license renewal but are not subject to 10 CFR 50.49 environmental qualification requirements. An inspection program will be established prior to the period of extended operation. Th program Will provide for inspectien of the-bus duGts. This inspection program considers the technical information and guidance provided in IEEE Standard P1205, "IEEE Guide for Assessing, Monitoring and Mitigating Aging Effects on Class 1E Equipment Used in Nuclear Power Generating Stations,"
SAND 96-0344, 'Aging Management Guideline for Commercial Nuclear Power Plants -
Electrical Cable and Terminations," and EPRI TR-1 09619, "Guideline for the Management of Adverse Localized Equipment Environments." Non segregated bur duct internal components and materials will be inspected for signs of aging degradation Dresden and Quad Cities                                                         Page A-48 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT that indicate possible loss of insulation function. Repair or rework is initiated as required to maintain the operating functions of the bus duct&.
A.2.3 Periodic Inspection of Ventilation System Elastomers The periodic inspection of ventilation system elastomers aging management program provides for routine inspections of certain elastomers in the standby gas treatment, reactor building ventilation, emergency diesel generator building ventilation, station blackout diesel generator building ventilation, and main control room ventilation systems. Prior to the period of extended operation an existing program for inspection of ventilation system elastomers will be enhanced. The program will include inspections for cracking, loss of material, or other evidence of aging of all flexible boots, access door seals and gaskets, and filter seals and gaskets in the components of these systems that are within the scope of license renewal. The scope of inspections will also include RTV silicone used as a duct sealant, in systems within the scope of license renewal.
A.2.4 Periodic Testing of Drywell and Torus Spray Nozzles Carbon steel piping upstream of the drywell and torus spray nozzles is subject to possible general corrosion. The periodic flow tests of drywell and torus spray nozzles address a concern that rust from the possible general corrosion may plug the spray nozzles. These periodic tests verify that the drywell and torus spray nozzles are free from plugging that could result from corrosion product buildup from upstream sources.
A.2.5 Lubricating Oil Monitorinq Activities The lubricating oil monitoring activities aging management program manages corrosion, loss of material, and cracking in lubricating oil heat exchangers and other specific components in the scope of license renewal by monitoring physical and chemical properties in lubricating oil. Sampling, testing, and trending verify lubricating oil properties. Oil analysis permits identification of specific wear mechanisms, contamination, and oil degradation within operating machinery.
Dresden and Quad Cities                                                            Page A49 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT These activities apply to the emergency diesel generator system, station blackout diesel generator system, and-HPCI system, electro-hydrauliccontrol (EHC) system, reactor core isolation cooling system, oil Gooler, and generator hydrogen seal oil system. The complete aging management program for these-the emergency diesel generator oil coolers, station blackout diesel generator oil coolers, and HPCI oil coolers also includes secondary-side (heat sink) chemistry controls, performance monitoring, and inspections. Those portions of the lubricating oil heat exchanger management program are described in:
* Section A.1.14, Closed-Cycle Cooling Water System," for the diesel generator and station blackout diesel generator oil coolers;
* Section A.2.6, 'Heat Exchanger Test and Inspection Activities," for the HPCI oil coolers.
A.2.6 Heat Exchanger Test and Inspection Activities The heat exchanger test and inspection activities aging management program provides condition monitoring, inspection, and performance testing to manage loss of material, cracking, and buildup of deposits in heat exchangers in the scope of license renewal, that are not tested and inspected under "Open-Cycle Cooling Water" or "Closed-Cycle Cooling Water" aging management programs.
These are new activities that will be implemented prior to the period of extended operation.
These activities include tests, inspections, and monitoring and trending of test results to confirm that aging effects are managed. To ensure that system and component functions are maintained, these components are also being included in the scope of other activities, which provide inservice inspection and performance monitoring, and primary and secondary-side (water and oil) chemistry controls.
* Management of water chemistry is described in Section A.1.2, "Water Chemistry."
* Management of the primary, oil side of the HPCI lubricating oil coolers is described in Section A.2.5, "Lubricating Oil Monitoring Activities."
A.2.7 Generator Stator Water Chemistry Activities The generator stator water chemistry activities aging management program manages loss of material and cracking aging effects by monitoring and controlling water chemistry. Generator stator water chemistry control maintains high purity water in accordance with General Electric guidelines for stator cooling water systems. Generator stator water is continuously monitored for conductivity and an alarm annunciates if Dresden and Quad Cities                                                         Page A-50 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT conductivity increases to a predetermined limit.
A.2.8 Periodic Inspection of Plant Heatinq System The periodic inspection of plant heating system aging management program provides for routine inspections of selected components in the plant heating system. Prior to the period of extended operation, a new program for periodic inspection of selected components in the plant heating system will be implemented. The selected components will be inspected to ensure they are free of cracking, loss of material and leakage. The inspection will consist of a visual inspection for the presence of general, crevice, galvanic, and pitting corrosion.
Where practical, the selected components will include components in stagnant flow areas that are most susceptible to loss of material. The inspection will be performed in accordance with ASME Code requirements. Certified NDE examiners will conduct a VT-3 visual inspection.
A.2.9    Periodic Inspection of Components Subiect to Moist Air Environments The periodic inspection of components subject to moist air environments aging management program provides for periodic inspections of selected components exposed to moist air environments and subject to wetting conditions based on system operation. Prior to the period of extended operation, a new program for periodic inspection of selected components will be implemented. The inspection will consist of UT examinations of components with interior surfaces that are inaccessible and visual inspection (VT-3) of components with accessible interior surfaces for the presence of loss of material due to general corrosion, pitting and crevice corrosion. The inspection will be performed in accordance with ASME Code requirements. Certified NDE examiners will conduct the UT and VT-3 visual inspection. In addition, visual inspection of flexible hoses will determine any age-related degradation prior to loss of function.
Dresden and Quad Cities                                                    Page A-51 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.3      Time-Limited Aging Analysis Summaries In the descriptions of this section, Class I and Class II are the Quad Cities safety classifications described in UFSAR Section 3.2.
A.3.1     Neutron Embrittlement of the Reactor Vessel and Internals The ferritic materials of the reactor vessel are subject to embrittlement due to high energy neutron exposure. Reactor vessel neutron embrittlement is a TLAA.
A.3.1.1    Reactor Vessel Materials Upper-Shelf Energy Reduction Due to Neutron Embrittlement The reactor vessel end-of-life neutron fluence has been recalculated for a 60-year (54 EFPY) extended licensed operating period.
The 54 EFPY USE was evaluated by an equivalent margin analysis (EMA) using the 54 EFPY calculated fluence and the Quad Cities surveillance capsule results in accordance with the requirements of 10 CFR 54.21(c)(1)(ii).
A.3.1.2      Adjusted Reference Temperature for Reactor Vessel Materials Due to Neutron Embrittlement The reactor vessel materials peak fluence, ARTNDT, and ART values for the 60-year (54 EFPY) license operating period were calculated in accordance with the requirements of 10 CFR 54.21 (c)(1)(ii).
A.3.1.3      Reflood Thermal Shock Analysis of the Reactor Vessel The effects of a reflood thermal shock described in UFSAR Section 3.9.5.3.3 were examined. An alternative analysis confirms that the effects remain acceptable for the period of extended operation, in accordance with the requirements of 10 CFR 54.21(c)(1)(i).
A.3.1.4      Reflood Thermal Shock Analysis of the Reactor Vessel Core Shroud and Repair Hardware Radiation embrittlement may affect the ability of reactor vessel internals, particularly the core shroud and repair hardware, to withstand a low-pressure coolant injection (LPCI) thermal shock transient. Embrittlement effects are evaluated for the maximum-fluence beltline region of the core shroud, where the maximum event strain is about 0.57 percent [UFSAR Section 3.9.5.3.21, and design of the core shroud repair tie rod stabilizer assemblies included an investigation of possible embrittlement effects.
The effects of the increase in neutron fluence with a 54 EFPY life at uprated power were evaluated, and the allowable strain for this faulted event remains a considerable margin above the expected strain.
Dresden and Quad Cities                                                          Page A-52 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT The core shroud repair tie rod stabilizer assemblies were designed for a 40-year life, which will not be exceeded at the end of the extended licensed operating period.
The existing analyses of the effects of embrittlement in the internals have been evaluated and remain valid for the period of extended operation, in accordance with the requirements of 10 CFR 54.21(c)(1)(i).
A.3.1.5      Reactor Vessel Thermal Limit Analyses: Operating Pressure - Temperature Limits Revised pressure-temperature (P-T) limits for a 60-year licensed operating life will be prepared using available Quad Cities capsule data and submitted to the NRC for approval prior to the start of the extended period of operation, in accordance with the requirements of 10 CFR 54.21 (c)(1)(ii).
A.3.1.6      Reactor Vessel Circumferential Weld Examination Relief Relief has been requested from the requirements for inspection of RPV circumferential welds for the remainder of the current 40-year licensed operating period. The justification for relief is consistent with Boiling Water Reactor Vessel and Internals Program BWRVIP-05, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations," guidelines. Application for an extension of this relief for the 60-year period of extended operation has been submitted.
The procedures and training that will be used to limit the frequency of cold over-pressure events to the number specified in the SER for the RPV circumferential weld relief request extension, during the license renewal term, are the same as those approved for use in the current period (Ref. 3 and 4).
The analyses associated with reactor vessel circumferential weld examination relief will be projected to the end of the period of extended operation, in accordance with the requirements of 10 CFR 54.21(c)(1)(il).
A.3.1.7      Reactor Vessel Axial Weld Failure Probability BWRVIP-05, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations," estimated the 40-year end-of-life failure probability of a limiting reactor vessel axial weld, showed that it was orders of magnitude greater than the 40-year end-of-life circumferential weld failure probability, and used this analysis to justify relief from inspection of the circumferential welds, as described in Section A.3.1.6.
The re-evaluation of the axial weld failure probability for 60 years depends on vessel ARTNDT calculations. The NRC staff review and the NRC staff and BWRVIP calculations of the test-case failure probabilities assume that 90 percent of axial welds will be inspected. At Quad Cities, less than 90 percent of axial welds can be inspected. As such, an analysis was performed for 54 EFPY to assess the effect Dresden and Quad Cities                                                         Page A-53 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT on the probability of fracture due to the actual inspection performed on the vessel axial welds and to determine if the coverage was sufficient in the inspection of regions contributing to the majority of the risk.
The evaluation shows that the calculated unit-specific axial weld conditional failure probabilities at 54 EFPY for Quad Cities are less than the failure probabilities calculated by the NRC staff in the NRC BWRVIP-05 SER at 64 EFPY and the limiting Clinton values found in Table 3 of the SER supplement. The projected probability of failure of an axial weld at Quad Cities will therefore provide adequate margin above the probability of failure of a circumferential weld, in support of relief from inspection of circumferential welds, for the extended licensed operating period, in accordance with the requirements of 10 CFR 54.21 (c)(1)(ii).
A.3.2 Metal Fatique The thermal and mechanical fatigue analyses of mechanical components have been identified as TLAAs for Quad Cities. Specific components have been designed considering transient cycle assumptions, as listed in vendor specifications and the Quad Cities UFSAR.
A.3.2.1      Reactor Vessel Fatigue Unit 1 and Unit 2 reactor vessel fatigue analyses depend on cycle count assumptions that assume a 40-year operating period. The effects of fatigue in the reactor vessel will be managed for the period of extended operation by the fatigue management program for cycle counting and fatigue usage factor tracking, as described in Section A.1.34.
This aging management program will ensure that fatigue effects in vessel pressure boundary components will be adequately managed and will be maintained within code design limits for the period of extended operation, in accordance with the requirements of 10 CFR 54.21(c)(1)(iii).
A.3.2.2      Fatigue Analysis of Reactor Vessel Internals A.3.2.2.1    Low-Cycle Thermal Fatigue Analysis of the Core Shroud and Repair Hardware Only one Quad Cities analysis of low-cycle fatigue in RPV internals exists: the evaluation of a standard design for repair of the core shroud. This analysis is a TLAA.
The calculated fatigue effects are not significant.
The fatigue analysis of the core shroud repair has been evaluated and remains valid for the period of extended operation, in accordance with the requirements of 10 CFR 54.21 (c)(1)(i).
Dresden and Quad Cities                                                         Page A-54 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.3.2.3      Reactor Coolant Pressure Boundary Piping and Component Fatigue Analysis A.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 Vil Class B and C All primary system and other reactor coolant pressure boundary (RCPB) piping systems were designed to USAS B31.1, 1967 Edition, as were the safety relief valve (SRV) discharge lines inside the drywell. The USAS B31.1 piping design does not invoke a fatigue analysis, but USAS B31.1 does apply a stress range reduction factor based on an assumed finite number of equivalent full-range thermal cycles for the design life. The B31.1 designs are therefore TLAAs because they are part of the current licensing basis, are used to support a safety determination, and depend on a specific number of cycles which might change with a change in licensed operating life.
The assumed number of design lifetime equivalent full-range thermal cycles determines the allowable stress range (the stress range reduction factor) for design of all Class I and Class II USAS B31.1 or ASME Class 2 or 3 piping. With the exception of containment vent and process bellows, no components in the scope of license renewal designed to ASME Section III or Section Vil require design for cyclic thermal loading.
The number of thermal cycles assumed for design of Class I and 11piping has been evaluated and the existing stress range reduction factor remains valid for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(i).
A.3.2.4      Effects of Reactor Coolant Environment on Fatigue Life of Components and Piping (Generic Safety Issue 190)
Generic Safety Issue (GSI) 190 was identified by the NRC because of concerns about potential effects of reactor water environments on component fatigue life during the period of extended operation.
Prior to the period of extended operation, Exelon will perform plant-specific calculations for the applicable locations identified in NUREG/CR 6260, "Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components," for older-vintage BWR plants, to assess the potential effects of reactor coolant on component fatigue life in accordance with 10 CFR 54.21(c)(1)(ii). The calculations of current and projected cumulative usage factors (CUFs) under this program will include appropriate environmental fatigue effect (FEN) factors from NUREGICR 6583 and NUREGICR 5704. Appropriate corrective action will be taken if the resulting projected end-of-life CUF values exceed 1.0.
Exelon reserves the right to modify this position in the future based on the results of industry activities currently underway, or based on other results of improvements in methodology, subject to NRC approval prior to changes in this position.
Dresden and Quad Cities                                                          Page A-55 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.3.3         Environmental Qualification of Electrical Equipment Electrical equipment included in the Quad Cities Environmental Qualification Program which has a specified qualified life of at least 40 years involves time-limited aging analyses for license renewal. The aging effects of this equipment will be managed in the Environmental Qualification Program discussed in Section A.1.35, "Environmental Qualification (EQ) of Electrical Components," in accordance with the requirements of 10 CFR 54.21 (c)(1 )(iii).
A.3.4        Containment Fatigue The Quad Cities Mark I containments were originally designed to stress limit criteria without fatigue analyses. However, the discovery of significant hydrodynamic loads
("new loads") caused by safety relief valve (SRV) and small, intermediate, and design basis pipe break discharges into the suppression pool required the reanalysis of the suppression chamber, vents, and attached piping and internal structures, including some fatigue analyses at limiting locations. These fatigue analyses of the suppression chamber, and its internals, and vents in each unit include assumed pressure, temperature, seismic, and SRV cycles, and combinations thereof. The scope of the analyses included the suppression chamber, the drywell-to-suppression chamber vents, SRV discharge piping, other piping attached to the suppression chamber and its penetrations, and the drywell-to-suppression chamber vent bellows.
A.3.4.1       Fatigue Analysis of the Suppression Chamber, Vents, and Downcomers For low cumulative usage factor (CUF) locations (40-year CUF < 0.4) the Quad Cities new loads analyses of each suppression chamber and its associated vents and downcomers have been evaluated and remain valid for the period of extended operation, in accordance with the requirements of 10 CFR 54.21 (c)(1)(i).
For higher cumulative usage factor locations in the analyses of the suppression chamber and suppression chamber vents and downcomers (40-year CUF 2 0.4) the effects of fatigue will be managed for the period of extended operation by the fatigue management cycle counting and fatigue usage factor tracking program, as described in Section A.1.34.
The fatigue management activities will ensure that fatigue effects in containment pressure boundary components are adequately managed and are maintained within code design limits for the period of extended operation, in accordance with the requirements of 10 CFR 54.21 (c)(1)(iii).
A.3.4.2       Fatigue Analysis of SRV Discharge Piping Inside the Suppression Chamber, External Suppression Chamber Attached Piping, and Associated Penetrations SRV discharge lines and external suppression chamber attached piping and associated penetrations were analyzed separately from the suppression chamber, vents and Dresden and Quad Cities                                                           Page A-56 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT downcomers. The disposition of these analyses is the same as described for the suppression chamber, vents and downcomers in Section A.3.4.1 above.
A.3.4.3     Drywell-to-Suppression Chamber Vent Line Bellows Fatigue Analyses A fatigue analysis of the drywell-to-suppression chamber vent line bellows was performed assuming 150 thermal and internal pressure load cycles for the 40-year life of the plant. The drywell-to-suppression chamber vent line bellows have a rated capacity of 1,000 cycles at maximum displacement.
The Quad Cities new loads fatigue analysis of the drywell-to-suppression chamber vent line bellows have been evaluated and remain valid for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(i).
A.3.4.4     Primary Containment Process Penetrations Bellows Fatigue Analysis The only containment process piping expansion joints subject to significant thermal expansion and contraction are those between the drywell shell penetrations and process piping. These are designed for a stated number of operating and thermal cycles.
The thermal cycle designs of Quad Cities containment process penetration bellows have been evaluated and remain valid for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(i).
A.3.6       Other Plant-Specific TLAAs A.3.5.1     Reactor Building Crane Load Cycles The reactor building overhead cranes in Quad Cities were designed to meet or exceed the design criteria of the Crane Manufacturers Association of America (CMAA)
Specification 70, 'Specifications for Electric Overhead Traveling Cranes," Class Al.
These cranes are capable of a minimum of 100,000 cycles at the full rated load of 125 tons. Correspondence with the NRC stated that over their 40-year life these cranes would most probably see fewer than 5,000 cycles at a maximum of 100 tons, and a larger number of cycles at significantly less than 100 tons.
The load cycle designs of Quad Cities reactor building cranes have been evaluated and remain valid for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(i).
A.3.5.2     Metal Corrosion Allowances A.3.5.2.1      Corrosion Allowance for Power Operated Relief Valves GE specification 25A5508, 'Relief Valve, Power Operated," for the Quad Cities Unit 2 replacement PORVs prescribes a corrosion allowance of 0.002 inches for stainless steel Dresden and Quad Cities                                                        Page A-57 License Renewal Application

Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT and 0.120 inches for carbon steel for a design life of 40 years. The specification is cited in Quad Cities UFSAR Section 5.2.2.
The corrosion allowance for the Quad Cities Unit 2 replacement PORVs has been evaluated and remains valid for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(i).
A.3.5.2.2         Degradation Rates of Inaccessible Exterior Drywell Plate Surfaces In its response to Generic Letter 87-05, 'Request for Additional Information Assessment of Licensee Measures to Mitigate and/ or Identify Degradation of Mark I Drywells,"
Commonwealth Edison evaluated the potential effects of corrosion on exterior drywell steel surfaces in the "sand pockets" of Dresden Unit 3 drywell and found that 27 years of service remained before corrosion at the assumed rate would have a significant adverse effect on design basis stresses. The evaluation concluded that the findings were applicable to Dresden Unit 2 and Quad Cities Units 1 and 2 as well.
The calculation will be revised for the realistic environment and for a full 60-year design life, in accordance with 10 CFR 54.21(c)(1)(ii). A UT inspection will validate assumptions used in the calculation. These actions will be completed before the period of extended operation.
A.3.5.2.3        Galvanic Corrosion in the Containment Shell and Attached Piping Components due to Stainless Steel ECCS Suction Strainers The Quad Cities ECCS suction strainers have been replaced with larger strainers. The replacement strainers are stainless steel. The modification included drilling new bolt holes and enlarging the existing bolt holes in each of the existing carbon steel strainer support flanges to provide sufficient bolting for the larger replacement strainers. The holes in the carbon steel flanges are not coated to protect them from corrosion. The calculation of corrosion effects assumes a corrosion allowance of 4 mils/year and assumes a design life of 33 years, which is just short of the 60-year extended operating period.
The corrosion rate assumptions used in the calculation will be confirmed by an ultrasonic inspection prior to the period of extended operation. Based on the results of the inspection, a revised galvanic corrosion calculation will be performed to validate acceptable wall thickness to the end of the 60-year licensed operating period, in accordance with 10 CFR 54.21(c)(1)(ii). In the event that the measured galvanic corrosion rate will not ensure acceptable thickness to the end of the 60-year licensed operating period, appropriate corrective action will be identified and implemented to maintain the structural integrity of the strainer flanges.
A.3.5.3          Crack Growth Calculation of a Postulated Flaw in the Heat Affected Zone of an Arc Strike in the Suppression Chamber Shell Dresden and Quad Cities                                                           Page A-58 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A calculation provides technical justification for continued operation of the Quad Cities Unit 2 torus which was damaged by an arc strike. The flaw has been ground smooth and NDE tested. It was initially assumed the damaged area would be repaired after two fuel cycles of operation. This time limit has been extended with appropriate NDE being performed to assure no cracks or other linear flaws exist in the affected area.
The crack growth calculation has been evaluated and remains valid for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(i).
A.3.5.4       Radiation Degradation of Drywell Shell Expansion Gap Polyurethane Foam The steel drywell shell is largely enclosed within the structural and shielding concrete of the reactor containment building. To accommodate thermal expansion, compressible foam was used to form an expansion gap between the concrete and the drywell shell. A confirming analysis contained in the UFSAR evaluates the increase in external compressive loads on the drywell exterior, due to additional compression of this foam, for accident-condition thermal expansion of the drywell. The load depends on the stress-strain curve of the foam, and the validity of this confirming analysis of the Quad Cities drywells therefore depends on the stiffness of the polyurethane foam. The analysis would require validation if the foam became stiffer (higher compressive stress for the same strain) as a result of increased radiation exposure from extended plant operation.
The expected radiation exposure of the foam has been evaluated and remains below the significant damage threshold at the end of the period of extended operation. The evaluation of thermal expansion compressive loads therefore also remains valid for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(i).
Dresden and Quad Cities                                                            Page A-59 License Renewal Application

Appendix A Quad Cities, Units 1 and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT A.3.6 References for Section A.3
: 1. Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2, License Renewal Project, TLAA Technical Report.
Revision 0, June 2002. Prepared by Parsons Energy and Chemicals, Inc. for the General Electric Company.
: 2. Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units I and 2, License Renewal Project, Potential TLAA Review Results Package. Revision 0, June 2002. Prepared by Parsons Energy and Chemicals, Inc. for the General Electric Company.
: 3. Quad Cities Letter RS-03-099 from Patrick R. Simpson (Exelon) to USNRC, Relief Request for Alternative Reactor Pressure Vessel Circumferential Weld Examinations for Fourth Interval Inservice Inspection Program, Letter dated May 16, 2003.
: 4. Quad Cities Letter RS-03-131 from Patrick R. Simpson (Exelon) to USNRC, Additional Information Supporting the Relief Request for Alternative Reactor Pressure Vessel Circumferential Weld Examinations for Fourth Interval Inservice Inspection Program, Letter dated July 7, 2003.
Dresden and Quad Cities                                                      Page A-60 License Renewal Application

Appendix B Aging Management Programs B.1      AGING MANAGEMENT PROGRAMS EVALUATED IN NUREG-1 801 B.1.1 ASME Section Xi Inservice Inspection, Subsections IWB, IWC, and IWD Description The ASME Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD aging management program is part of the inservice inspection (ISI) program and provides for condition monitoring of reactor coolant pressure retaining piping and components within the scope of license renewal, except for the reactor pressure vessel. It also provides for condition monitoring of reactor internal components within the scope of license renewal, and the Dresden isolation condenser. The program is implemented through procedures that require examinations consistent with the 1995 Edition through the 1996 Addenda of ASME Section Xl and for Quad Cities an approved relief request PR-02.
The program includes:
* Cracking monitoring for susceptible inservice inspection Class 1 components subject to a steam or reactor water environment, through volumetric examinations of pressure retaining welds and their heat affected zones in piping components.
* Cracking monitoring of the Quad Cities reactor vessel flange leak detection line through monitoring for leaks during reactor vessel flood-up in accordance with relief request PR-02.
* Loss of fracture toughness monitoring for susceptible inservice inspection Class 1 components in reactor recirculation and reactor water cleanup systems through visual inspections of reactor recirculation and reactor water cleanup valves and reactor recirculation pumps for signs of this aging effect.
* Cracking monitoring for susceptible reactor internal components subject to a reactor water environment, through surface and volumetric examinations.
* Cracking monitoring of the Dresden isolation condenser through surface and volumetric examinations of pressure retaining nozzle welds and their heat affected zones that are subject to a steam or reactor water environment.
Dresden and Quad Cities                                                     LRA Replacement Page B-6 License Renewal Application

Appendix B Aging Management Programs
* Loss of material monitoring of portions of the Dresden isolation condenser subject to a steam or reactor water environment, through system pressure tests.
NUREG-1801 Consistency With enhancement, The ASME Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD aging management program is consistent with the ten elements of aging management program XL.M1, "ASME Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD," specified in NUREG-1801 with the following exceptions.
Exceptions to NUREG-1801 NUREG-1801 indicates that the aging of the isolation condenser is to be managed by ASME Section Xl Inservice Inspection (ISI) Subsection IWB (for Class I components). However, the Dresden isolation condenser is ISI Class 2 on the tube side and ISI Class 3 on the shell side. Therefore, Subsections IWC and IWD are used, as Class I requirements do not apply.
Enhancements
* NlUREG 1801 indicates this program is to use the 1995 Edition through the 1996 Addenda of ASME Section Xi. The current Code Ofrecord for Dresden and Quad Cities is the 1989 Edition of ASME SeGtion Xi. The Program will be revised to be consistent with the requirements of the 1995 Edition through the 1996 Addenda, of ASME Section XI.
The enhancement is scheduled for implementation prior to the period of extended operation.
Operating Experience Dresden and Quad Cities have both successfully identified indications of age-related degradation prior to the loss of the intended functions of the components, and have taken appropriate corrective actions through evaluation, repair or replacement of the components in accordance with ASME Section Xl and station implementing procedures.
Conclusion The ASME Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD aging management program provides reasonable assurance that aging effects are adequately managed so that the intended functions of components within the scope of license renewal that are covered by this program are maintained during the period of extended operation.
Dresden and Quad Cities                                                      LRA Replacement Page B-7 License Renewal Application

Appendix B Aging Management Programs B.1.3 Reactor Head Closure Studs Description The reactor head closure studs aging management program provides for condition monitoring and preventive activities to manage stud cracking and loss of material. The program is implemented through station procedures based on the examination and inspection requirements specified in the 1995 Edition through the 1996 Addenda of ASME Section Xl, Table IWB-2500-1 and preventive measures described in Regulatory Guide 1.65, 'Materials and Inspection for Reactor Vessel Closure Studs." The reactor head studs at Dresden and Quad Cities are not metal-plated, and have had manganese phosphate coatings applied.
NUREG-1801 Consistency With enhancement, The reactor head closure studs aging management program is consistent with the ten elements of aging management program XI.M3, "Reactor Head Closure Studs," specified in NUREG-1801 with the following exceptions.
Exceptions to NUREG-1801 NUREG-1801 indicates that program inspections are in accordance with the requirements of ASME Section Xl, Subsection IWB, Table IWB 2500-1. The Dresden and Quad Cities programs utilize relief requests CR-13 and CR-1I respectively. These relief requests provide for a VT-1 visual inspection instead of a surface examination of reactor closure head nuts.
NUREG-1801 indicates that the reactor closure head studs receive volumetric examinations. Quad Cities reactor closure head studs are examined by end shot UT as approved in relief request CR-12.
Ennancremems 0     NUREG 1801 indicates this program is to use the 1005 Edition through the 1096 Addenda of ASME Section XI. The current Code of record for Dresden and Quad Cities is the 1989 Edition of ASME Section Xl. The Program will be revised to be consistent with the requirements of the 1995 Edition through the 1996 Addenda, of ASME Section Xl.
Try_
InR .. _ _
ecnnalencement
__L IS GRedculeo for implementation prior to the period of extended operation A
_J.
Dresden and Quad Cities                                                                  LRA Replacement Page B-11 License Renewal Application

Appendix B Aging Management Programs B.1.4 BWR Vessel ID Attachment Welds Description The BWR vessel ID attachment welds aging management program activities incorporate the inspection and evaluation recommendations of BWRVIP-48, "Vessel ID Attachment Weld Inspection and Evaluation Guidelines," as well as the water chemistry recommendations of EPRI TR-103515-R2, 'BWR Water Chemistry Guidelines." The program is implemented through station procedures that provide for mitigation of cracking through water chemistry and monitoring for cracking through invessel examinations. Reactor vessel attachment weld inspections are implemented through station procedures that are part of inservice inspection and incorporate the requirements of the 1995 Edition through.the 1996 Addenda of ASME Section Xl. The ASME inspections are enhanced with inspections consistent with BWRVIP-48.
NUREG-1801 Consistency With enhancement, The BWR vessel ID attachment welds aging management program is consistent with the ten elements of aging management program XL.M4, "BWR Vessel ID Attachment Welds," specified in NUREG-1 801 with the following exceptions.
Exceptions to NUREG-1801 NUREG-1801 indicates that water chemistry control is in accordance with BWRVIP-29, "BWR Water Chemistry Guidelines." BWRVIP-29 references the 1993 revision of EPRI TR-1 03515, "BWR Water Chemistry Guidelines." The Dresden and Quad Cities water chemistry programs are based on EPRI TR-103515-R2, which is the 2000 revision. Section B.1.2 presents the Water Chemistry aging management program and the exceptions to the program as specified in NUREG-1801.
* NUREG 1801 indicates this program is to use the 1995 Edition through the 1996 Addenda of ASME SectioR Xl. The cUrRt Code of record for Dresden and Quad Cities is the 1980 Edition of ASMEl Section XI. The Program will be revised to be consistent with the requirements of the 1995 Edition through the 1096 Addenda, of ASME Section XI.
The enhanGement is scheduled for implementation prior to the period of extended operation.
Operating Experience The Dresden and Quad Cities inspection and testing methodologies have not detected cracking attachment welds. However, the same inspection and testing methodologies have detected cracking on the vessel internals. Section B.1.9 presents the BWR Vessel Internals aging management program.
Dresden and Quad Cities                                                    LRA Replacement Page B-13 License Renewal Application

Appendix B Aging Management Programs B.1.6 BWR Control Rod Drive Return Line Nozzle Description The control rod drive return line nozzle aging management program consists of inservice inspections and previously implemented system modifications to manage the aging effect of cracking in the control rod drive return line nozzles. Dresden and Quad Cities have capped the control rod drive return line nozzles. Inservice inspections are performed consistent with the 1995 Edition through the 1996 Addenda of ASME Section Xl requirements. No augmented inspections are required.
NUREG-1801 Consistency With enhancement, The control rod drive return line nozzle aging management program is consistent with the ten elements of aging management program XI.M6, "BWR Control Rod Drive Return Line Nozzle," specified in NUREG-1801 with the following exceptions.
Exceptions to NUREG-1801 NUREG-1801 indicates that the program includes system modifications and enhanced inservice inspections.
The Dresden and Quad Cities programs do not provide for the augmented inspections specified in NUREG-0619, `BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," because the control rod drive return line nozzles have been capped. Inservice inspections are performed consistent with ASME Section Xl requirements.
En hanccments
* NUREG 1801 indicates this program is to use the 1995 Edition through the 1996 Addenda of ASIVIE Section XI. The current Code of record for Dresden and Quad Cities isthe 1989 Edition of ASME Section Xl. The Program will be revised to be consistent with the requiremnents of the 1 995 Edition through the 1996 Addenda, of ASME Section XI.
The enhancement is scheduled for implementation prior to the period of extended operation.
Operating Experience The Dresden and Quad Cities inspection and testing methodologies have detected indication of cracking aging effects on control rod drive nozzles prior to loss of their intended functions. These indications were repaired by flaw removal or weld overlay as determined appropriate by engineering evaluation in accordance with ASME Code and station procedure requirements.
Dresden and Quad Cities                                                              LRA Replacement Page B-16 License Renewal Application

Appendix B Aging Management Programs B.1.7 BWR Stress Corrosion Cracking Description BWR stress corrosion cracking aging management program mitigates Intergranular stress corrosion cracking (IGSCC) in stainless steel reactor coolant pressure boundary components and piping four inches and greater nominal pipe size. Preventive measures include monitoring and controlling of water impurities by water chemistry program activities and providing replacement stainless steel components in the solution annealed condition with a maximum carbon content of 0.035 wt. % and a minimum ferrite level of 7.5 wt. %. Inspection and flaw evaluation are conducted in accordance with the inservice inspection program plan for the station.
The program is implemented through station procedures based on NUREG-0313, "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," GL 88-01, "NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping," and its Supplement 1, BWRVIP-75, "Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules,"
EPRI TR-103515-R2, "BWR Water Chemistry Guidelines," and the 1995 Edition through the 1996 Addenda of ASME Section XI.
NUREG-1801 Consistency With enhancement, The BWR stress corrosion cracking aging management program is consistent with the ten elements of aging management program XI.M7, "BWR Stress Corrosion Cracking," specified in NUREG-1801 with the following exceptions.
Exceptions to NUREG-1801 NUREG-1801 indicates that water chemistry control is in accordance with BWRVIP-29 for water chemistry in BWRs. BWRVIP-29 references the 1993 revision of EPRI TR- 103515, "BWR Water Chemistry Guidelines."
The Dresden and Quad Cities water chemistry programs are based on EPRI TR-103515-R2, which is the 2000 revision. Section B.1.2 presents the Water Chemistry aging management program and the exceptions to the program as specified in NUREG-1801.
Enhanccmcnts
* NUREG 1801 indicates this program is to usec the 1995 Edition through the 1996 Addenda of ASME Section Xl. The current Code of Fereod for Dereden and Quad Cities is the 1989 Edition of ASIME Section XI. The Program will be revised to be consistent with the requirements of the 1905 Edition through the 1906 Addenda, of ASME Section XI.
The enhancement is scheduled for implementation prior to the period of extended operation.
Dresden and Quad Cities                                                    LRA Replacement Page B-18 License Renewal Application

Appendix B Aging Management Programs B.1.8 BWR Penetrations Description The BWR penetrations aging management program activities incorporate the inspection and evaluation recommendations of BWRVIP-27, 'BWR Standby Liquid Control System/Core Plate Delta-P Inspection and Flaw Evaluation Guidelines," and BWRVIP-49, Instrument Penetration Inspection and Flaw Evaluation Guidelines," as well as the water chemistry recommendations of EPRI TR-1 03515-R2, "BWR Water Chemistry Guidelines." The program is implemented through station procedures that provide for mitigation of cracking through the water chemistry and monitoring for cracking through inservice inspection examinations.
Penetration inspections are implemented through station procedures that are part of inservice inspection and incorporate the requirements of the 1995 Edition through the 1996 Addenda of ASME Section Xl.
NUREG-1801 Consistency With enhancement, The BWR penetrations aging management program is consistent with the ten elements of aging management program XL.M8, "BWR Penetrations," specified in NUREG-1801 with the following exceptions.
Exceptions to NUREG-1801 NUREG-1801 indicates that water chemistry control is in accordance with BWRVIP-29 for water chemistry in BWRs. BWRVIP-29 references the 1993 revision of EPRI TR-103515, "BWR Water Chemistry Guidelines."
The Dresden and Quad Cities water chemistry programs are based on EPRI TR-103515-R2, which is the 2000 revision. Section B.1.2 presents the Water Chemistry aging management program and the exceptions to the program as specified in NUREG-1801.
NUREG-1801 indicates that the standby liquid control system nozzles are inspected in accordance with the requirements of ASME Section Xl, Subsection IWB. The Dresden and Quad Cities programs utilize relief request ISI CR-01 that provides for inspection of the inner radius of the standby liquid control system nozzle by a VT-2 examination instead of the normal volumetric inspection required by the ASME Code.
Ena cements NUREG 1801 indicates this progran is to use the 1995 Edition through the 1906 Addenda of ASME Section MI. The current Code of record for Dresden and Quad Cities is the 1989 Edition of ASME Section XI. The Program will be revised to be consistent with the requirements of the 1995 Edition through the 1996 Addenda, of ASME Scction XI.
The enhancement is 6cheduled for implementatieRon prwior oth period of extended operation.
Dresden and Quad Cities                                                     LRA Replacement Page B-20 License Renewal Application

Appendix B Aging Management Programs B.1.9 BWR Vessel Internals Description The BWR vessel internals aging management program mitigates the effects of SCC, IGSCC, and IASCC in reactor pressure vessel internals through water chemistry activities that are implemented through station procedures and are consistent with the guidelines of EPRI TR-103515-R2, "BWR Water Chemistry Guidelines," 2000 Revision. The program also manages cracking of reactor pressure vessel internals through condition monitoring activities that consist of examinations implemented through station procedures consistent with the recommendations of the BWRVIP guidelines, as well as the requirements of the 1995 Edition through the 1996 Addenda of ASME Section XI.
NUREG-1801 Consistency With enhancement, The BWR vessel internals aging management program is consistent with the ten elements of aging management program XI.M9, "BWR Vessel Internals," specified in NUREG-1801 with the following exceptions.
Exceptions to NUREG-1801 NUREG-1801 indicates that water chemistry control is in accordance with BWRVIP-29, "BWR Water Chemistry Guidelines." BWRVIP-29 references the 1993 revision of EPRI TR-103515, "BWR Water Chemistry Guidelines." The Dresden and Quad Cities water chemistry programs are based on EPRI TR-103515-R2, which is the 2000 revision. Section B.1.2 presents the Water Chemistry aging management program and the exceptions to the program as specified in NUREG-1 801.
Enhanercmcnts
* NUREG 1801 indicates this program is to use the 1995 Edition through the 1906 Addenda of ASME Section Xl. The current Code of record for Dresden and Quad Cities is the 1989 Edition of ASME Section Xl. The Program will be revised to be consistent with the requirements of the 1095 Edition-through the 1996 AddendaI of ASME Section Xl.
The enhancement is scheduled for implementation prior to the period of extended operation.
Operating Experience BWR vessel internals aging management activities have detected aging degradation and implemented appropriate corrective actions to maintain system and component intended functions including prompt repair of degraded components prior to failure. This includes cracking adjacent to the Quad Cities access hole cover and cracking of the reactor internal portion of the core spray piping at Dresden Unit 3. After evaluation by engineering, these cracks were repaired in accordance with ASME Section Xl and station procedure requirements.
Jet Pump Beam Assembly #20 failed at Quad Cities Unit 1 in January 2002. All similar beams have been replaced with improved heat treatment beams.
Dresden and Quad Cities                                                     LRA Replacement Page B-22 License Renewal Application

Appendix B Aging Management Programs B.1.12 Bolting Integrity Description The bolting integrity aging management program provides for condition monitoring of selected pressure retaining bolted joints and external surfaces for piping and components within the scope of license renewal.
The bolting integrity program incorporates industry recommendations addressed in EPRI NP-5769, "Degradation and Failure of Bolting in Nuclear Power Plants,' as part of the comprehensive corporate component pressure retaining bolting program. The program consists of visual inspections for external surface degradation that may be caused by loss of material or cracking of the bolting, or by an adverse environment.
The activities are implemented through station procedures and predefined tasks. Inspection of inservice inspection Class 1, 2 and 3 components is conducted in accordance with the 1995 Edition through the 1996 Addenda of ASME Section XI. Nonsafety component inspections rely on detection of visible leakage during preventive maintenance and routine observation activities.
NUREG-1801 Consistency With enhancements the bolting integrity aging management program is consistent with the ten elements of aging management program XI.M18, "Bolting Integrity," specified in NUREG-1801 with the following exceptions.
Exceptions to NUREG-1801 NUREG-1801 indicates that EPRI TR-104213, "Bolted Joint Maintenance and Applications Guide," is used as a basis for evaluation of the structural integrity of nonsafety related bolting. NUREG-1801 also indicates that the bolting integrity programs developed and implemented in accordance with commitments made in response to NRC communications on bolting events have provided an effective means of ensuring bolting reliability.
These programs are documented in EPRI NP-5769 and TR-1 04213 and represent industry consensus. The Dresden and Quad Cities programs address the guidance contained in EPRI TR-104213 but do not specifically cite its use.
NUREG-1801 indicates that the program covers all bolting within the scope of license renewal including structural bolting. The Dresden and Quad Cities bolting integrity programs do not address structural bolting.
The Structures Monitoring Program (B.1.30) covers aging management of structural bolting.
NUREG-1801 indicates that the program covers all bolting within the scope of license renewal including bolting for Class 1 NSSS component supports. The Dresden and Quad Cities bolting integrity programs do not address Class 1 NSSS component support bolts. Aging management of ASME Section XI Class 1, 2, and 3, and Class MC support members, including mechanical connections, is covered by the ASME Section XI, Subsection IWF (B.1.27) aging management program.
Dresden and Quad Cities                                                        LRA Replacement Page B-27 License Renewal Application

Appendix B Aging Management Programs NUREG 1801 indicates that the programn geerally includes pcriodic inspeRtion for loss of prelead. The Dresden and Quad Cities programs do not include inspections for loss of prelead because loss of prelead in a mechanical joint is a design driven effect and not an aging effect.
Enhancements 4-  The Program will be rovised as necessary to be inxaGG     9Fdance with the 1905 Edition through the 1996 Addenda, and approved relief requests instead of-the 1989 Edition with approved relief Fequests, of ASME Section XL for inspertion of inere      inpee ,tiGnn   laSS 1, 2 and 3 co mponnt.
* The program will provide for formal inspections of bolted joints of diesel generator system components and performing periodic component bolted joint inspections in high-humidity/moisture areas (pump vaults).
The enhancements are is scheduled for implementation prior to the period of extended operation.
Operating Experience Dresden and Quad Cities have experienced isolated cases of bolting degradation attributed to loss of material and cracking aging effects. No reactor coolant pressure boundary leakage due to boric acid induced degradation has been noted since both stations are BWRs. System engineer walkdowns have also identified incidental surface rust on exterior component surfaces. In all cases, the existing inspection and testing methodologies have discovered the deficiencies and corrective actions were implemented prior to loss of system or component intended functions.
Conclusion The bolting integrity aging management program provides reasonable assurance that loss of material and cracking aging effects are adequately managed so that the intended functions of bolting for pressure retaining joints and external surfaces for piping and components within the scope of license renewal are maintained during the period of extended operation.
Dresden and Quad Cities                                                        LRA Replacement Page B-28 License Renewal Application}}