ML030090655
| ML030090655 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Quad Cities  |
| Issue date: | 01/03/2003 |
| From: | Exelon Generation Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| Download: ML030090655 (55) | |
Text
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 IW D..........................................................................................
A-3 A.1.2 W ater Chemistry...................................................................................
A-3 A. 1.3 Reactor Head Closure Studs................................................................
A-3 A.1.4 BW R Vessel ID Attachment Welds.......................................................
A-3 A.1.5 BW R Feedwater Nozzle.......................................................................
A-3 A.1.6 BW R Control Rod Drive Return Line Nozzle.........................................
A-4 A.1.7 BW R Stress Corrosion Cracking............... %
........................................... A-4 A.1.8 BWR Penetrations................................................................................
A-4 A.1.9 BW R Vessel Internals.......................................................................
A-5 A.1.10 Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS)............................................
A-5 A. 1.11 Flow-Accelerated Corrosion.............................................................
A-5 A.1.12 Bolting Integrity.....................................................................................
A-5 A. 1.13 Open-Cycle Cooling W ater System......................................................
A-6 A. 1.14 Closed-Cycle Cooling W ater System....................................................
A-6 A. 1.15 Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems.............................................
A-7 A.1.16 Compressed Air Monitoring..................................................................
A-7 A.1.17 BW R Reactor W ater Cleanup System..................................................
A-7 A.1.18 Fire Protection......................................................................................
A-8 A.1.19 Fire W ater System...........................................................................
A-8 A.1.20 Aboveground Carbon Steel Tanks........................................................
A-9 A. 1.21 Fuel Oil Chemistry................................................................................
A-9 A.1.22 Reactor Vessel Surveillance.................................................................
A-9 A. 1.23 One-Time Inspection....................................................................
A-I0 A.1.24 Selective Leaching of Materials.....................................................
A-1 I A.1.25 Buried Piping and Tanks Inspection...............................................
A-11 A.1.26 ASME Section XI, Subsection IW E.................................................
A-11 A.1.27 ASME Section XI, Subsection IW F......................................................
A-12 A.1.28 10 CFR Part 50, Appendix J................................................................
A-12 A.1.29 Masonry W all Program........................................................................
A-12 A.1.30 Structures Monitoring Program......................................................
A-12 A. 1.31 RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants..........................................
A-13 A.1.32 Protective Coating Monitoring and Maintenance Program................... A-13 A.1.33 Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements..........................................
A-14 Dresden and Quad Cities Page A-i License Renewal Application
Table of Contents A.1.34 Metal Fatigue of Reactor Coolant Pressure Boundary....................
A-1 4 A.1.35 Environmental Qualification (EQ) of Electrical Components................ A-14 A.2 PLANT-SPECIFIC AGING MANAGEMENT PROGRAMS.............. A-15 A.2.1 Corrective Action 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 Drywell and Torus Spray Nozzles.......................... A-16 A.2.5 Lubricating Oil Monitoring Activities.....................................................
A-16 A.2.6 Heat Exchanger Test and Inspection Activities....................................
A-16 A.3 TIME-LIMITED AGING ANALYSIS SUMMARIES................................ A-18 A.3.1 Neutron Embrittlement of the Reactor Vessel and Internals............ A-18 A.3.1.1 Reactor Vessel Materials Upper-Shelf Energy Reduction Due to Neutron Embrittlement.................................................................
A-18 A.3.1.2 Adjusted Reference Temperature for Reactor Vessel Materials Due to Neutron Embrittlement......................................................
A-18 A.3.1.3 Reflood Thermal Shock Analysis of the Reactor Vessel.................
A-18 A.3.1.4 Reflood Thermal Shock Analysis of the Reactor Vessel Core Shroud and Repair Hardware.................................................................................
A-18 A.3.1.5 Reactor Vessel Thermal Limit Analyses: Operating Pressure Temperature Limits.............................................................................
A-19 A.3.1.6 Reactor Vessel Circumferential Weld Examination Relief.................... A-19 A.3.1.7 Reactor Vessel Axial Weld Failure Probability.....................................
A-19 A.3.2 M etal Fatigue.......................................................................................
A-20 A.3.2.1 Reactor Vessel Fatigue Analyses........................................................
A-20 A.3.2.2 Fatigue Analysis of Reactor Vessel Internals.......................................
A-20 A.3.2.2.1 High-Cycle Flow-Induced Vibration Fatigue Analysis of Jet Pump Riser Braces...............................................................................................
A -20 A.3.2.3 Reactor Coolant Pressure Boundary Piping and Component Fatigue Analysis.........................................................................................
A-20 A.3.2.3.1 ASME Section III Class I Reactor Coolant Pressure Boundary Piping and Component Fatigue Analysis..............................................................
A-20 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 VIII C lass B and C.....................................................................................
A-21 A.3.2.3.3 Fatigue Analysis of the Isolation Condenser........................................
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 Dow ncom ers.......................................................................................
A-23 A.3.4.2 Fatigue Analysis of SRV Discharge Piping Inside the Suppression Chamber, External Suppression Chamber Attached Piping, and Associated Penetrations.....................................................................
A-23 A.3.4.3 Drywell-to-Suppression Chamber Vent Line Bellows Fatigue AnalysesA-23 Dresden and Quad Cities Page A-ii License Renewal Application
Table of Contents A.3.4.4 Primary Containment Process Penetrations Bellows Fatigue Analysis A-23 A.3.5 Other Plant-Specific TLAAs.................................................................
A-24 A.3.5.1 Reactor Building Crane Load Cycles...................................................
A-24 A.3.5.2 Metal Corrosion Allowances................................................................
A-24 A.3.5.2.1 Degradation Rates of Inaccessible Exterior Drywell Plate Surfaces..... A-24 A.3.5.2.2 Galvanic Corrosion in the Containment Shell and Attached Piping Components due to Stainless Steel ECCS Suction Strainers.............. A-24 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 Quad Cities Units I and 2 Updated Final Safety Analysis Supplement.............. A-27 A.1 AGING MANAGEMENT PROGRAMS.................................................
A-28 A.1.1 ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD.......................................
A-28 A.1.2 W ater Chemistry..................................................................................
A-28 A.1.3 Reactor Head Closure Studs...............................................................
A-28 A.1.4 BW R Vessel ID Attachment W elds......................................................
A-28 A. 1.5 BWR Feedwater Nozzle......................................................................
A-28 A.1.6 BW R Control Rod Drive Return Line Nozzle........................................
A-29 A.1.7 BW R Stress Corrosion Cracking..........................................................
A-29 A.1.8 BW R Penetrations...............................................................................
A-29 A.1.9 BWR Vessel Internals..........................................................................
A-30 A.1.10 Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS).......................................................
A-30 A.1.11 Flow-Accelerated Corrosion.................................................................
A-30 A.1.12 Bolting Integrity....................................................................................
A-30 A.1.13 Open-Cycle Cooling W ater System.....................................................
A-31 A. 1.14 Closed-Cycle Cooling W ater System...................................................
A-31 A.1.15 Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems.........................................................
A-31 A. 1.16 Compressed Air Monitoring.................................................................
A-32 A. 1.17 BW R Reactor W ater Cleanup System.................................................
A-32 A.1.18 Fire Protection.....................................................................................
A-32 A. 1.19 Fire W ater System...............................................................................
A-33 A.1.20 Aboveground Carbon Steel Tanks.......................................................
A-33 A.1.21 Fuel Oil Chemistry...............................................................................
A-34 A.1.22 Reactor Vessel Surveillance................................................................
A-34 A.1.23 One-Time Inspection...........................................................................
A-34 A.1.24 Selective Leaching of Materials...........................................................
A-35 A.1.25 Buried Piping and Tanks Inspection.....................................................
A-36 A.1.26 ASME Section XI, Subsection IW E......................................................
A-36 A.1.27 ASME Section XI, Subsection IW F......................................................
A-36 A.1.28 10 CFR Part 50, Appendix J................................................................
A-36 A.1.29 Masonry W all Program........................................................................
A-37 Dresden and Quad Cities Page A-ilu License Renewal Application
Table of Contents A. 1.30 Structures Monitoring Program............................................................
A-37 A.1.31 RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants..................................................................
A-38 A.1.32 Protective Coating Monitoring and Maintenance Program................... A-38 A.1.33 Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements..........................................
A-38 A. 1.34 Metal Fatigue of Reactor Coolant Pressure Boundary......................... A-39 A.1.35 Environmental Qualification (EQ) of Electrical Components................ A-39 A.1.36 Boraflex Monitoring..............................................................................
A-39 A.2 PLANT-SPECIFIC AGING MANAGEMENT PROGRAMS................... A-40 A.2.1 Corrective Action Program...................................................................
A-40 A.2.2 Periodic Inspection of Non-EQ, Non-Segregated Electrical Bus D ucts...........................................................................................
A -40 A.2.3 Periodic Inspection of Ventilation System Elastomers.......................... A-40 A.2.4 Periodic Testing of Drywell and Torus Spray Nozzles.......................... A-41 A.2.5 Lubricating Oil Monitoring Activities.....................................................
A-41 A.2.6 Heat Exchanger Test and Inspection Activities....................................
A-41 A.2.7 Generator Stator Water Chemistry Activities........................................
A-42 A.3 Time-Limited Aging Analysis Summaries.............................................
A-43 A.3.1 Neutron Embrittlement of the Reactor Vessel and Internals................. A-43 A.3.1.1 Reactor Vessel Materials Upper-Shelf Energy Reduction Due to Neutron Embrittlement....................................................................
A-43 A.3.1.2 Adjusted Reference Temperature for Reactor Vessel Materials Due to Neutron Embrittlement..............................................
A-43 A.3.1.3 Reflood Thermal Shock Analysis of the Reactor Vessel...................... A-43 A.3.1.4 Reflood Thermal Shock Analysis of the Reactor Vessel Core Shroud and Repair Hardware.....................................................
A-43 A.3.1.5 Reactor Vessel Thermal Limit Analyses: Operating Pressure Temperature Umits......................................................................
A-44 A.3.2 Metal Fatigue......................................................................................
A-44 A.3.2.1 Reactor Vessel Fatigue......... #............................................................
A-44 A.3.2.2 Fatigue Analysis of Reactor Vessel Internals.......................................
A-44 A.3.2.2.1 Low-Cycle Thermal Fatigue Analysis of the Core Shroud and Repair Hardw are............................................................................................
A-44 A.3.2.3 Reactor Coolant Pressure Boundary Piping and Component Fatigue A nalysis.............................................................................................
A-45 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-45 A.3.2.4 Effects of Reactor Coolant Environment on Fatigue Life of Components and Piping (Generic Safety Issue 190)...........
A-45 A.3.3 Environmental Qualification of Electrical Equipment............................ A-45 A.3.4 Containment Fatigue...........................................................................
A-46 A.3.4.1 Fatigue Analysis of the Suppression Chamber, Vents, and Downcomers..................................................................................
A-46 Dresden and Quad Cities Page A-lv License Renewal Application
Table of Contents A.3.4.2 A.3.4.3 A.3.4.4 A.3.5 A.3.5.1 A.3.5.2 A.3.5.2.1 A.3.5.2.2 A.3.5.2.3 A.3.5.3 A.3.5.4 A.3.6 Dresden and Quad Cities Page A-v License Renewal Application Fatigue Analysis of SRV Discharge Piping Inside the Suppression Chamber, External Suppression Chamber Attached Piping, and Associated Penetrations.....................................................................
A-46 Drywell-to-Suppression Chamber Vent Line Bellows Fatigue Analyses................................................................................
A-46 Primary Containment Process Penetrations Bellows Fatigue Analysis..................................................................................
A-47 Other Plant-Specific TLAAs.................................................................
A-47 Reactor Building Crane Load Cycles...................................................
A-47 Metal Corrosion Allowances................................................................
A-47 Corrosion Allowance for Power Operated Relief Valves...................... A-47 Degradation Rates of Inaccessible Exterior Drywell Plate Surfaces..... A-47 Galvanic Corrosion in the Containment Shell and Attached Piping Components due to Stainless Steel ECCS Suction Strainers.............. A-48 Crack Growth Calculation of a Postulated Flaw in the Heat Affected Zone of an Arc Strike in the Suppression Chamber Shell...... A-48 Radiation Degradation of Drywell Shell Expansion Gap Polyurethane Foam.............................................................................
A-48 References for Section A.3..................................................................
A-50
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.
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.
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 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, Wessel ID Attachment Weld Inspection and Evaluation Guidelines," andlor 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."
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.
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 Dresden and Quad Cities Page A-3 License Renewal Application
Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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, "Altemate 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 1996 Addenda.
A.J.7 BWR Stress Corrosion Crackinq 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." 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.
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 XI, 1995 Edition through the 1996 Addenda.
Dresden and Quad Cities Page A-4 License Renewal Application
Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 inservice inspection program will be revised to be consistent with ASME Section XI, 1995 Edition through the 1996 Addenda.
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.I1 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, "ErosionlCorrosion-lnduced 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.
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.
Prior to the period of extended operation the inservice inspection program will be revised to be consistent with ASME Section XA, 1995 Edition through the 1996 Addenda.
In addition, the program will Dresden and Quad Cities Page A-5 License Renewal Application
Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT include inspections of bolted joints of diesel generator system components and of components in locations containing high humidity or moisture.
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. The aging management of structural bolting is covered by the structures monitoring program. 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" 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 Genera! 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 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 Syistem 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.
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.1.15 Inspection of Overhead Heavy Load and Litght 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 ANSIIISA-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 TR-103515, MBWR 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.
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.1.18 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 wraps, 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.1.19 Fire Water System The fire water system aging management program provides fire system header and hydrant flushing, system performance (flow land 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 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.
Dresden and Quad Cities Page A-8 License Renewal Application
Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 Above-ground Carbon Steel Tanks The aboveground carbon steel tanks aging management program manages corrosion of outdoor nitrogen 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. 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.
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 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. Emergency diesel generator fuel oil analysis acceptance criteria are contained in the'Technical Specifications and are based on industry-recognized ASTM methods and standards.
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, ORadiation Embrittlement of Reactor Vessel Materials,"
Revision 2.
Prior to the period of extended operation the program will be consistent with BWRVIP 78, Integrated Surveillance Program," and BWRVIP-86, "BWR Integrated Surveillance Program Implementation Plan.' The program will ensure coupon availability during the period of extended operation, and provide for saving withdrawn coupons for future reconstitution.
Dresden and Quad Cities Page A-9 License Renewal Application
Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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:
Inspection of a sample 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.
a Inspection a sample of piping exposed to the containment atmosphere (safety relief valve discharge piping and HPCI turbine exhaust sample locations) for loss of material.
0 Inspection of a sample of condensate and torus water components for corrosion in stagnant locations to verify effective water chemistry control.
a Inspection of a sample of compressed gas system piping components for corrosion and a sample of compressed gas system flexible hoses for elastomer degradation.
Inspection of a sample of lower sections of carbon steel fuel oil and lubricating oil tanks for reduced thickness.
0 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.
Dresden and Quad Cities Page A-10 License Renewal Application
Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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, NGeneric 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.
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 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.'
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, 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 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 program covers steel containment shells and their integral attachments; containment hatches and aidocks; 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 ASME Section XI in its entirety instead of the 1992 Edition and Addenda.
Dresden and Quad Cities Page A-1I License Renewal Application
Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 program is in accordance with ASME Section XI, Subsection IWF, 1989 Edition, and Code Case N-491-1. Prior to the period of extended operation the program will include ASME Class MC component supports consistent with NUREG-1801, "Generic Aging Lessons Learned (GALL) Report," Chapter III, Section B1.3.
A.1.28 10 CFR Part 60, 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 10CFR50 Appendix J Option B, Regulatory Guide 1.163, uPerformance-Based Containment Leak-Testing Program,"
NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50 Appendix J,' and ANSIIANS 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 XI, Subsection IWFU 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.
Dresden and Quad Cities Page A-12 License Renewal Application
Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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.
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 Unit 1 and Unit 2 and 3 crib houses and within the scope of license renewal. 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, 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 Monitorinq 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, Dresden and Quad Cities Page A-13 License Renewal Application
Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT "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 connections which might be susceptible to aging during the period of extended operation. A sample of 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 tows, 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.
Dresden and Quad Cities Page A-14 License Renewal Application
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 transformers to the 4160V ESF buses. They are 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 IE 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." Non-segregated bus duct internal components and materials will be inspected for signs of aging degradation that indicate possible loss of insulation function.
Repair or rework is initiated as required to maintain the operating functions of the bus ducts.
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 Dresden and Quad Cities Page A-15 License Renewal Application
Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 concem 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.2.6 Lubricating Oil Monltoring Activities The lubricating oil monitoring activities aging management program manages corrosion, loss of material, and cracking in lubricating oil heat exchangers 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, station blackout diesel generator, and HPCI oil coolers. The complete aging management program for these 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 blackou diesel generator oil coolers; and in 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. For the isolation condensers this program also includes the augmentation activities identified in NUREG-1801, "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.
Dresden and Quad Cities Page A-16 License Renewal Application
Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 IS! 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 tubes.
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 XI 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."
Dresden and Quad Cities Page A-17 License Renewal Application
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 altemative 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.
Dresden and Quad Cities Page A-18 License Renewal Application
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 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.
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 determine if the coverage was sufficient in the inspection of regions contributing to the majority of the risk.
Dresden and Quad Cities Page A-19 License Renewal Application
Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 A.3.2.3.1 ASME Section III Class I 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 Dresden and Quad Cities Page A-20 License Renewal Application
Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 I 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 VIII 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 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.3.2.3.3 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 Dresden and Quad Cities Page A-21 License Renewal Application
Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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.
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. 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).
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 loadso) 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-suppresslon chamber vents, SRV discharge piping, other piping attached to the suppression chamber and its penetrations, and the drywell-to-suppression chamber vent bellows.
Dresden and Quad Cities Page A-22 License Renewal Application
Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 > 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 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).
Dresden and Quad Cities Page A-23 License Renewal Application
Appendix A Dresden, Units 2 and 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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.
The calculation will be revised for the realistic environment and for a full 60-year design
- life, in accordance with 10CFR54.21(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.3.5.2.2 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 milslyear 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).
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.3.5.3 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).
Dresden and Quad Cities Page A-25 License Renewal Application
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, Ucense 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.
Dresden and Quad Cities Page A-26 License Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT Quad Cities Units I and 2 Updated Final Safety Analysis Supplement Dresden and Quad Cities Page A-27 License Renewal Application
Appendix A Quad Cities, Units I and 2 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, surface, 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 Xl, 1995 Edition through the 1996 Addenda.
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.
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 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 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 XI, 1995 Edition through the 1996 Addenda.
A.1.6 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 Dresden and Quad Cities Page A-28 License Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 XI requirements.
No augmented inspections in accordance with NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Une Nozzle Cracking," or the alternative recommendations of GE NE-523-A71-0594, "Altemate 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 1996 Addenda.
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.
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.
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 SystemlCore 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,3 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.
Dresden and Quad Cities Page A-29 License Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 inservice inspection program will be revised to be consistent with ASME Section XI, 1995 Edition through the 1996 Addenda.
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.
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.
Prior to the period of extended operation the inservice inspection program will be revised to be consistent with ASME Section XI, 1995 Edition through the 1996 Addenda.
In addition, the program will include inspections of bolted joints of diesel generator system components and of components in locations containing high humidity or moisture.
Dresden and Quad Cities Page A-30 License Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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. The aging management of structural bolting is covered by the structures monitoring program. Aging management of ASME Section XI Class 1, 2, and 3 and Class MC support members, including mechanical connections, is covered by the uASME Section XI, 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 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 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-1 07396, "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 Dresden and Quad Cities Page A-31 License Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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, "lnstrument 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.
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-103515, "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 CorrosiQn 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.
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 wraps, 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.
Dresden and Quad Cities Page A-32 License Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 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 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 Above-ground Carbon Steel Tanks The aboveground carbon steel tanks aging management program manages corrosion of outdoor nitrogen 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. 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.
Dresden and Quad Cities Page A-33 License Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 fuel oil is in accordance with industry-recognized ASTM methods and standards. Emergency diesel generator fuel oil analysis acceptance criteria are contained in the Technical Specifications and are based on industry-recognized ASTM methods and standards.
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, "Integrated Surveillance Program," and BWRVIP-86, "BWR Integrated Surveillance Program Implementation Plan." 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:
Inspection of a sample 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.
Dresden and Quad Cities Page A-34 License Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 0
Inspection a sample of piping exposed to the containment atmosphere (safety relief valve discharge piping and HPCI turbine exhaust sample locations) for loss of material.
Inspection of a sample of condensate and torus water components for corrosion 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 elastomer 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.
0 Inspection of a sample of standby gas treatment and ventilation system components for loss of material.
0 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.
0 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.
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 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, OService Water System Problems Affecting Safety-Related Equipment.
Dresden and Quad Cities Page A-35 License Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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, 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 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 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 XI Subsection IWE for steel containments (Class MC), 1992 Edition including 1992 Addenda.
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 program is in accordance with ASME Section XI, Subsection IWF, 1989 Edition, and Code Case N-491-1. Prior to the period of extended operation the program will include ASME Class MC component supports consistent with NUREG-1801, "Generic Aging Lessons Learned (GALL) Report,' Chapter III, Section B1.3.
A.1.28 10 CFR Part 60. 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 Dresden and Quad Cities Page A-36 License Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 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.
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 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.
Dresden and Quad Cities Page A-37 License Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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. 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-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 I 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 Sublect to 10 CFR 60.49 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 sample of 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.
Dresden and Quad Cities Page A-38 License Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 EPRl-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 Borafiex monitoring aging management program consists of (1) neutron attenuation testing ("blackness testingn) 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 Generc Letter 96-04, "Boraflex 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."
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Appendix A Quad Cities, Units I 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 4160V ESF buses. They are 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, uAging Management Guideline for Commercial Nuclear Power Plants - Electrical Cable and Terminations," and EPRI TR-109619, "Guideline for the Management of Adverse Localized Equipment Environments."
Non-segregated bus duct internal components and materials will be inspected for signs of aging degradation that indicate possible loss of insulation function.
Repair or rework is initiated as required to maintain the operating functions of the bus ducts.
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 Dresden and Quad Cities Page A-40 License Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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.2.6 Lubricating Oil Monltoring Activities The lubricating oil monitoring activities aging management program manages corrosion, loss of material, and cracking in lubricating oil heat exchangers 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.
These activities apply to the emergency diesel generator, station blackout diesel generator, and HPCI oil coolers. The complete aging management program for these 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.
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.
Dresden and Quad Cities Page A-41 Ucense Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 conductivity increases to a predetermined limit.
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Appendix A Quad Cities, Units I 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 fermtic 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-43 License Renewal Application
Appendix A Quad Cities, Units I 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 Umits 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.3.2 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 I 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).
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Appendix A Quad Cities, Units I 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 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, 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.
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. 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 will be managed in the Dresden and Quad Cities Page A-45 License Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 (unew 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 > 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 adeqdately 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 Dresden and Quad Cities Page A-46 License Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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.3.5 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, uSpecifications 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 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,"
Dresden and Quad Cities Page A-47 License Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 10CFR54.21(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 milslyear 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).
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 aro 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 Dresden and Quad Cities Page A-48 License Renewal Application
Appendix A Quad Cities, Units I and 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) SUPPLEMENT 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 Quad Cities, Units I 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 I and 2, Ucense 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.
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