RS-11-159, Quad Cities, Units 1 and 2 - Updated Final Safety Analysis Report (Ufsar), Revision 11, Appendix a

Quad Cities, Units 1 and 2 - Updated Final Safety Analysis Report (Ufsar), Revision 11, Appendix a

ML11305A071
Person / Time
Site:	Quad Cities
Issue date:	10/19/2011
From:	Exelon Generation Co, Exelon Nuclear
To:	Office of Nuclear Reactor Regulation
References
RS-11-159
Download: ML11305A071 (30)

Text

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Table of Contents Page A-i Revision 8, October 2005 Quad Cities Units 1 and 2 License Renewal UFSAR Supplement........................................A-1 Introduction.........................................................................................................

A-1 A.1 AGING MANAGEME NT PROGRAMS................................................................A-1 A.1.1 ASME Section XI Inservice Ins pection, Subsections IWB, IWC, and IWD..............................................................................................................A-1 A.1.2 Water Chemistry.................................................................................................A-1 A.1.3 Reactor Head Closure Studs..............................................................................A-1 A.1.4 BWR Vessel ID Attachment Welds.....................................................................A-2 A.1.5 BWR Feedwat er Nozzle......................................................................................A-2 A.1.6 BWR Control Rod Driv e Return Line Nozzle......................................................A-2 A.1.7 BWR Stress Corrosion Cracking........................................................................A-2 A.1.8 BWR Penet rations...............................................................................................A-3 A.1.9 BWR Vessel Internals.........................................................................................A-3 A.1.10 Thermal Aging and Neutron Irr adiation Embrittlement of Cast Austenitic Stainless Steel (CASS)......................................................................A-4 A.1.11 Flow-Accelerated Corrosion................................................................................A-4 A.1.12 Bolting Integrity...................................................................................................A-4 A.1.13 Open-Cycle Cooli ng Water System....................................................................

A-5 A.1.14 Closed-Cycle Coo ling Water System..................................................................A-5 A.1.15 Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems.........................................................................A-6 A.1.16 Compressed Ai r Monitoring................................................................................A-6 A.1.17 BWR Reactor Wa ter Cleanup System................................................................A-6 A.1.18 Fire Protection.....................................................................................................A-7 A.1.19 Fire Water System..............................................................................................A-7 A.1.20 Aboveground Carbon Steel Tanks......................................................................A-8 A.1.21 Fuel Oil Chemistry...............................................................................................A-8 A.1.22 Reactor Vessel Surveillance...............................................................................A-8 A.1.23 One-Time Inspection...........................................................................................A-9 A.1.24 Selective Leaching of Materials........................................................................A-10 A.1.25 Buried Piping and T anks Inspection.................................................................A-10 A.1.26 ASME Section XI, Subsection IWE...................................................................

A-10 A.1.27 ASME Section XI, Subsection IWF...................................................................

A-11 A.1.28 10 CFR Part 50, Appendix J.............................................................................A-11 A.1.29 Masonry Wa ll Program.....................................................................................A-11 A.1.30 Structures M onitoring Program.........................................................................

A-11 A.1.31 RG 1.127, Inspection of Wate r-Control Structures Associated with Nuclear Po wer Plants...............................................................................A-12 A.1.32 Protective Coating Moni toring and Maintenanc e Program...............................A-13 A.1.33 Electrical Cables and Connecti ons Not Subject to 10 CFR 50.49 Environmental Qualificat ion Requirements......................................................A-13 A.1.34 Metal Fatigue of Reactor Coolant Pressure Boundary.....................................A-13 A.1.35 Environmental Qualification (E Q) of Electrical Components............................A-13 A.1.36 Boraflex Monitoring...........................................................................................A-14 APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Table of Contents (Cont'd.)

Page A-ii Revision 8, October 2005 A.1.37 Electrical Cables Not Subj ect to 10 CFR 50.49 Environmental Qualification Requirements used in Instrument Circuits..................................A-14 A.2 PLANT-SPECIFIC AGING MA NAGEMENT PROGRAMS...............................A-15 A.2.1 Corrective Ac tion Program................................................................................A-15 A.2.2 Periodic Inspection of Non-EQ, Non-Segregated Electrical Bus Ducts.........................................................................................................A-15 A.2.3 Periodic Inspection of Ventilation System Elastomers.....................................A-15 A.2.4 Periodic Testing of Dryw ell and Torus Spra y Nozzles......................................A-16 A.2.5 Lubricating Oil Moni toring Activities..................................................................A-16 A.2.6 Heat Exchanger Test and Inspection Activities................................................A-16 A.2.7 Generator Stator Wate r Chemistry Ac tivities....................................................A-17 A.2.8 Periodic Inspection of Plant Heating System....................................................A-17 A.2.9 Periodic Inspection of Components Subject to Moist Air Environments...........A-17 A.2.10 Periodic Inspection of Steam Dryers------------------A-18

A.3 TIME-LIMITED AGING ANALYSIS SUMMARIES............................................A-19 A.3.1 Neutron Embrittlement of t he Reactor Vessel and Internals............................A-19 A.3.1.1 Reactor Vessel Materials Upper-Shelf Energy Reduction Due to Neutron Embr ittlement.................................................................................A-19 A.3.1.2 Adjusted Reference Temperature for Reactor Vessel Materials Due to Neut ron Embrittlement..........................................................

A-19 A.3.1.3 Reflood Thermal Shock Analys is of the Reac tor Vessel..................................A-19 A.3.1.4 Reflood Thermal Shock Analysis of the Reactor Vessel Core Shroud and Repai r Hardware..................................................................A-19 A.3.1.5 Reactor Vessel Thermal Limit Analyses: Operating Pressure - Temperature Limits...........................................................................................A-20 A.3.1.6 Reactor Vessel Circumferent ial Weld Examinat ion Relief................................A-20 A.3.1.7 Reactor Vessel Axial Weld Failure Probability..................................................A-20

A.3.2 Metal Fatigue....................................................................................................A-21 A.3.2.1 Reactor Vessel Fatigue.....................................................................................

A-21 A.3.2.2 Fatigue Analysis of R eactor Vessel Internals...................................................A-21 A.3.2.2.1 Low-Cycle Thermal Fatigue Anal ysis of the Core Shroud and Repair Hardware..........................................................................................................A-21 A.3.2.3 Reactor Coolant Pressure Boundary Piping and Component Fatigue Analysis............................................................................................................A-21 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 VIII

Class B and C...................................................................................................A-21 A.3.2.4 Effects of Reactor Coolant Environment on Fatigue Life of Components and Piping (Generic Safety Issue 190).......................................

A-22 A.3.3 Environmental Qualification Of Electrical Equipment.......................................A-22

A.3.4 Containment Fatigue.........................................................................................A-22 A.3.4.1 Fatigue Analysis of the Suppression Chamber, Vents, and Downcomers.....................................................................................................

A-23 APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Table of Contents (Cont'd.)

Page A-iii Revision 11, October 2011 A.3.4.2 Fatigue Analysis of SRV Di scharge Piping Inside the Suppression Chamber, External Suppression Chamber Attached Piping, and

Associated Penet rations...................................................................................A-23 A.3.4.3 Drywell to Suppression Chamber Vent Line Bellows Fatigue Analyses..............................................................................................A-23 A.3.4.4 Primary Containment Pr ocess Penetrations Bellows Fatigue Analysis...............................................................................................A-23

A.3.5 Other Plant-Specific TLAAs..............................................................................A-24 A.3.5.1 Reactor Buildi ng Crane Load Cycles................................................................

A-24 A.3.5.2 Metal Corrosi on Allowances.............................................................................A-24 A.3.5.2.1 Corrosion Allowance for Power Operated Re lief Valves..................................A-24 A.3.5.2.2 Degradation Rates of Inaccessible Exterior Dr ywell Plate Surfaces................A-24 A.3.5.2.3 Galvanic Corrosion in the C ontainment Shell and Attached Piping Components due to Stainless Steel ECCS Suction Strainers.........................A-25 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-25 A.3.5.4 Radiation Degradation of Drywell Shell Expansion Gap Polyurethane Foam..........................................................................................

A-25 A.3.6 References for Section A.3...............................................................................A-26

A.4 NEWLY IDENTIFIED SSCs (10 CFR 54.37(b))................................................A-27

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-1 Revision 11, October 2011 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 Quad

Cities Nuclear Power Station following issuance of the renewed operating license.

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 vo lumetric and visual examinations of components for assessment, identification of signs of degradation, and establishment of

corrective actions. The inspections will be implemented in accordance with 10 CFR

50.55a.

Quad Cities will implement the guidance of BW RVIP-74, "BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines," with the following exception. Exception: Risk

Informed Inservice Inspection is implemented in lieu of ASME Section XI requirements for portions of Class 1 and Class 2 systems.

Technical Specification revisions containing new P-T Curves were 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 vari ous contaminants below system-specific limits based on industry-recognized guidelines of EPRI 1016579, "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 guidanc e provided in BWRVIP-190, "BWR Water Chemistry Guidelines - 2008 Revision" (EPRI Report 1016579).

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.

The requirements of ASME Section XI will be implemented in accordance with 10 CFR

50.55a. The reactor head studs are not metal-plated, and have had manganese phosphate coatings applied.

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-2 Revision 11, October 2011 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 XI; and (b) monitoring and control of

reactor coolant water chemistry in accordanc e with industry-recognized guidelines of EPRI 1016579, "BWR Water Chemistry Guidelines." The requirements of ASME Section XI will

be implemented in accordance with 10 CFR 50.55a.

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-A, Revision 01, "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 XI 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. The requirements of ASME Section XI will be

implemented in accordance with 10 CFR 50.55a.

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 1016579, "BWR Water Chemistry Guidelines." The requirements of ASME Section XI will

be implemented in accordance with 10 CFR 50.55a.

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-3 Revision 11, October 2011 Quad Cities will implement the general guidance provided in 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 1016579, "BWR Water Chemistry Guidelines," to ensure the long-term

integrity and safe operation of boiling water reactor vessel internal components. The requirements of ASME Section XI will be implemented in accordance with 10 CFR 50.55a.

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 1016579, "BWR Water Chemistry Guidelines,"

to ensure the long-term integrity and safe operation of boiling water reactor vessel internal

components. The requirements of ASME Section XI will be implemented in accordance with 10 CFR 50.55a.

Quad Cities will implement the general guidanc e provided in BWRVIP-18, "BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines."

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 pr ovided 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 guidanc e 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 Gui delines." Quad Cities will perform the APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-4 Revision 11, October 2011 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 guidanc e provided in BWRVIP-47, "BWR Lower Plenum Inspection and Flaw Evaluation Guidelines."

Quad Cities will implement the general guidanc e provided in BWRVIP-76, "BWR Core Shroud Inspection and Flaw Evaluation Guidelines."

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 nec essary. 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 3 guidelines. Prior to the period of extended

operation the program will be revised to include main steam and reactor head vent 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 XI. The requirements of ASME Section XI will be implemented in accordance with

10 CFR 50.55a. 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-ring

girder bolting.

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-5 Revision 9, October 2007 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 co vered by the structures monitoring program. Aging management of ASME Section XI Class 1, 2, and 3 and Class MC (excluding Class

MC piping) support members, including mechanical connections, is covered by the "ASME Section XI, Subsection IWF" aging management program. Aging management of Class MC piping support members, including mechanical connections is covered by the

"Structures Monitoring" aging management program.

A.1.13 Open-Cycle Cooling Water System The open-cycle cooling water system aging m anagement program includes (a) surveillance and control of biofouling, (b) tests to verify heat transfer, (c) a routine inspection and

maintenance program, including system flus hing and chemical treatment, (d) periodic inspections for leakage, loss of material, and blockage, (e) engineering evaluations and

heat sink performance assessments, and (f) asse ssments 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-cycl e 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 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 agi ng management program relies on preventive measures to minimize corrosion by maintain ing inhibitors and by performing non-chemistry monitoring consisting of inspection and nondes tructive examinations (NDEs) based on industry-recognized guidelines of EPRI 1007820, "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 1007820 guidance.

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-6 Revision 10, October 2009 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 Le tter 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, "C ompressor and Instrument Air System Maintenance Guide."

A.1.17 BWR Reactor Water Cleanup System The BWR reactor water cleanup (RWCU) syst em aging management program monitors and controls reactor water chemistry based on industry-recognized guidelines of EPRI 1016579, "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 XI is not required.

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-7 Revision 8, October 2005 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; 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:

• 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 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, "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 pum ps, outdoor hydrants, and outdoor transformer deluge systems; and periodic non-intrusive wall th ickness measurements of selected portions of the fire water system at interv als that do not exceed every 10 years.

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-8 Revision 8, October 2005 A.1.20 Aboveground Carbon Steel Tanks The aboveground carbon steel tanks aging m anagement 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-foundation interfaces (including foundation coatings)

are periodically inspected 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 inspections of 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 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 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 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- 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-specific surveillance plan for the LR period in accordance with 10 CFR 50, Appendices G and H.

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-9 Revision 11, October 2011 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.

• 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 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, and a sample of compressed gas system flexible hoses for age-related degradation.

• 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 tr eatment 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.

Based on the completed Unit 1 inspection, these hoses are made of carbon steel, not

elastomer as originally assumed. Therefore, no additional inspections are required.

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-10 Revision 11, October 2011

• 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."

• Inspection of stainless steel clevis pins for crack initiation and growth due to SCC and the interfacing carbon steel support members for loss of material due to pitting and galvanic

corrosion in torus water environment.

A.1.24 Selective Leaching of Materials The selective leaching of materials aging m anagement program includes numerous one-time inspections of components of the different susceptible materials to determine if loss of

material due to selective leaching is occurring.

These inspections will consist of visual inspection consistent with ASME Section XI VT-1 visual inspection requirements. 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 extended

operation, and of the need to expand the test samp le. For systems subjected to environments where water is not treated (i.e., the open-cycle c ooling 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 hardne ss 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 m anagement 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, one-time visual

inspection of exterior surface of section of ductile iron fire protection piping, 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 will be performed.

A.1.26 ASME Section XI, Subsection IWE The ASME Section XI, 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 pr ogram covers steel containment shells and their integral attachments; containment hatches and airlocks; seals, gaskets and moisture APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-11 Revision 11, October 2011 barriers; and pressure-retaining bolting. The program includes assessment of damage and corrective actions. The requirements of ASME Section XI will be implemented in accordance with 10 CFR 50.55a.

A.1.27 ASME Section XI, Subsection IWF The ASME Section XI, Subsection IWF aging management program consists of periodic visual examination of ASME Section XI Class 1, 2, and 3 component and piping supports for

signs of degradation, evaluation, and establishment of corrective actions. The requirements of ASME Section XI will be implemented in accordance with 10 CFR 50.55a. Prior to the period of extended operation the program w ill include ASME Class MC non-piping component supports.

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," and an alternate test that was accepted by NRC granting the

exemption for primary containment penetration expansion bellows assemblies that use a two ply design.

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 XI, 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.

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-12 Revision 8, October 2005 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

• 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.

• 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.

• Addition of inspection criteria for standard components such as snubbers, struts, and spring cans.

• A VT-3 visual inspection of 15% of the non-exempt Class MC pipe supports once every 10 years.

• Management of the aging effects of those major components and larger piping headers that were credited as an anchor for non-safety related piping.

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.

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-13 Revision 8, October 2005 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-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 revi ew 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 Subject 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. All accessible electrical

cables and connections installed in adverse lo calized 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 fo r calculation of stress cycles and fatigue usage factors from operating cycles, automated c ounting 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 Equi pment 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 APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-14 Revision 8, October 2005 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, "BorafIex Degradation in Spent Fuel Pool Storage Racks." The Boraflex

monitoring activities are based on the maintenance 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 (SRM's), Intermediate Range Monitors (IRM's

), Local Power Range Monitors (LPRM's), 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.

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-15 Revision 8, October 2005 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 corre ctive action program applies to all plant systems, structures and components (both safe ty-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 4160V Engineering Safety Systems (ESS) buses for signs of aging degradation that indicate possible loss of intended function. This program will be enhanced

prior to the period of extended operation to inspect the bus bar insulation material at the

accessible bolted connections of the non-segregated bus ducts that connect the reserve

auxiliary transformer to the 4160V ESS buses and inspect 10% of the splice insulation

material 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

will also include inspections for the presence of dirt or moisture in the bus ducts. The visual

inspection will include all visible insulation in both directions beyond the location of the bolted

connection splice insulation inspected.

These bus ducts are in scope of license renewal but are not subject to 10 CFR 50.49

environmental qualification requirements. 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-109619, "Guideline for the

Management of Adverse Localized Equipment Environments."

A.2.3 Periodic Inspection of Ventilation System Elastomers The periodic inspection of ventilation syst em 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 r oom ventilation systems. Prior to the period of extended operation an existing program for ins pection 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 APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-16 Revision 8, October 2005 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 Monitoring Activities The lubricating oil monitoring activities aging management program manages loss of material, cracking, and elastomer hardening/loss of strength in lubricating oil heat

exchangers and other specific components in t he scope of license renewal by monitoring physical and chemical properties in lubricati ng oil and heat exchanger performance testing.

Sampling, testing, and trending verify lubricating oil properties and proper heat exchanger

operation. 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, HPCI system, electro-hydr aulic control (EHC) system, reactor core isolation cooling system, and generator hydr ogen seal oil system. The complete aging management program for 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 activi ties 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 APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-17 Revision 8, October 2005 provide inservice inspection and performanc e 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 activi ties 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. G enerator stator water is continuously monitored for conductivity and an alarm annunciates if conductivity increases to a predetermined limit.

A.2.8 Periodic Inspection of Plant Heating System The periodic inspection of plant heating syst em 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 Subject to Moist Air Environments The periodic inspection of components subjec t 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 consis t 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.

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-18 Revision 8, October 2005 A.2.10 Periodic Inspection of Steam Dryers Dresden and Quad Cities will perform periodic inspections of the steam dryer plate material and welds for evidence of cracking. These inspection will be performed in accordance with

an NRC approved BWRVIP. However, if the NRC has not approved a BWRVIP for dryer

inspection prior to entering the period of extended operation, Dresden and Quad Cities will

prepare and submit a site-specific inspection program for NRC review and approval.

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-19 Revision 8, October 2005 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 E nergy 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, RTNDT, 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.

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-20 Revision 9, October 2007 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 embri ttlement 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 (54 EFPY) have been submitted to and approved by the NRC. 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)(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.

The re-evaluation of the axial weld failure probability for 60 years depends on vessel RT NDT 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 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.

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-21 Revision 8, October 2005 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 Fatigue 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 fa tigue 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). 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 VIII 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, APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-22 Revision 8, October 2005 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

VIII require design for cyclic thermal loading. The number of thermal cycles assumed for design

of Class I and II piping 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 (F EN) 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 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 w ill be managed in the Environmental Qualification Program discussed in Section A.1.35, "Environm ental 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 si gnificant hydrodynamic loads ("new loads") caused by safety relief valve (SRV) and small, intermedi ate, 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 APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-23 Revision 8, October 2005 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 it s 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 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 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.

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-24 Revision 8, October 2005 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.3.5 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 A1. 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 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. 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.

APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-25 Revision 8, October 2005 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 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 repai red 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 Dryw ell 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 APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-26 Revision 8, October 2005 thermal expansion compressive loads therefore also remains valid for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(i).

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. Quad Cities Letter RS-03-099 from Patrick R. Simpson (Exelon) to USNRC, Relief Request for Alternative Reactor Pressure Ve ssel 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 Four th Interval Inservice Inspection Program, Letter dated July 7, 2003 APPENDIX A QUAD CITIES, UNITS 1 AND 2 LICENSE RENEWAL UFSAR SUPPLEMENT Page A-27 Revision 11, October 2011 A.4 NEWLY IDENTIFIED SSCs (10 CFR 54.37(b))

After the renewed license is issued, the UFSAR update required by 10 CFR 50.71(e) must include any systems, structures, and components newly identified that would have been subject to an aging management review or evaluation of time-limited aging analyses in accordance with

§54.21. This UFSAR update must describe how the effects of aging will be managed such that the intended function(s) in §54.4(b) will be effectively maintained during the period of extended operation.

No. Date Identified SSC Description Aging Management Review (AMR)

Conclusion Aging Management Program 1. 09/22/2011 The Environmental Qualification (EQ) Program was revised to reflect increases in the assumed zone temperatures and pressures in the first floor of the Reactor Building.

These changes resulted in the addition of several SSCs into the EQ Program. These SSCs were installed at the time of the license renewal review, so they are considered "newly identified" under the terms of 10 CFR 54.37(b).

The changes to the EQ Program zone assumptions associated with the Reactor Building components were noted in a revision to UFSAR Figure 3.11-1, Sheets 1a, 1b, 1c, 1d, 3a, 5a, 7a.

Specific SSCs impacted by this change are noted in EC 370997.

SSCs that are included in the scope of the EQ Program are subject to Time Limited Aging Analysis (TLAA).

Since these SSCs have been added to the scope of the EQ Program, they are within the scope of its TLAA. The existing EQ Program was credited without revision as Aging Management Program (AMP) B.1.35 in the Renewed License SER (NUREG 1796).

The addition of these newly identified SSCs to the scope of the EQ Program also included them in the scope of Aging Management Program (AMP) B.1.35.

These SSCs are now subject to the same aging management activities as those in the License Renewal Application.