ML042050495

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Safety Evaluation Report Related to the License Renewal of the Dresden Nuclear Power Station, Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2 (Exelon Generation Company, LLC (Exelon) - Cover Page
ML042050495
Person / Time
Site: Dresden, Quad Cities  Constellation icon.png
Issue date: 07/01/2004
From: Kuo P
Division of Regulatory Improvement Programs
To: Skolds J
Exelon Generation Co
BURTON WF, NRR/DRIP, 415-2853
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Download: ML042050495 (22)
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Safety Evaluation Report Related to the License Renewal of the Dresden Nuclear Power Station, Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2 Docket Nos. 50-237, 50-249, 50-254, and 50-265 Exelon Generation Company, LLC (Exelon)

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555-0001 July 2004

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-iii-ABSTRACT This safety evaluation report (SER) documents the technical review of the Dresden Nuclear Power Station (DNPS), Units 2 and 3 and Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, license renewal application (LRA) by the U.S. Nuclear Regulatory Commission (NRC) staff (staff). By letter dated January 3, 2003, Exelon Generation Company (Exelon or the applicant) submitted the LRA for Dresden and Quad Cities (D/QCNPS) in accordance with Title 10, Part 54 of the Code of Federal Regulations (10 CFR Part 54 or the Rule). Exelon requests renewal of the operating licenses for DNPS Unit 2 (License No. DRP-19), DNPS Unit 3 (License No. DRP-25), QCNPS Unit 1 (License No. DRP-29), and QCNPS Unit 2 (License No. DRP-30) for a period of 20 years beyond the current license expirations of midnight, December 22, 2009; January 12, 2011; December 14, 2012; and December 14, 2012, respectively.

DNPS is located in Grundy County, Illinois, on the shore of a man-made cooling lake, with the Illinois River to the north and the Kankakee River to the east. The QCNPS is located in Rock Island County, IL, on the east bank of the Mississippi River opposite the mouth of the Wapsipinicon River, and about 3 miles north of Cordova, IL. DNPS, Units 1 and 2, and QCNPS, Units 2 and 3, each consist of a General Electric boiling-water reactor (BWR/3) authorized to operate individually at a steady state reactor power level not to exceed 2957 megawatts-thermal, or approximately 850 megawatts-electric.

This SER presents the status of the staffs review of information submitted to the NRC through June 22, 2004, the cutoff date for consideration in the SER. The staff identified open items and confirmatory items that had to be resolved before the staff could make a final determination on the application. Sections 1.5 and 1.6 of this report summarize these items and their resolutions. Section 6 provides the staffs final conclusion of its review of the D/QCNPS LRA.

The NRC license renewal project manager is Mr. Rajender Auluck. Mr. Auluck may be reached at (301) 415-1025. Written correspondence should be addressed to the License Renewal and Environmental Impacts Program, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

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-v-TABLE OF CONTENTS Abstract

........................................................... -iii-Table of Contents.................................................... -v-Abbreviations...................................................... -xv-

1. Introduction and General Discussion........................................ 1-1 1.1 Introduction..................................................... 1-1 1.2 License Renewal Background...................................... 1-2 1.2.1 Safety Reviews.......................................... 1-3 1.2.2 Environmental Reviews.................................... 1-5 1.3 Principal Review Matters.......................................... 1-5 1.4 Interim Staff Guidance............................................ 1-6 1.5 Summary of Open Items........................................... 1-8 1.6 Summary of Confirmatory Items.................................... 1-14 1.7 Summary of Proposed License Conditions............................ 1-31
2. Scoping and Screening Methodology for Identifying Structures and Components Subject to Aging Management Review, and Implementation Results..................... 2-1 2.1 Scoping and Screening Methodology................................. 2-1 2.1.1 Introduction............................................. 2-1 2.1.2 Summary of Technical Information in the Application.............. 2-1 2.1.2.1 Scoping Methodology.............................. 2-1 2.1.2.2 Screening Methodology............................. 2-8 2.1.3 Staff Evaluation......................................... 2-10 2.1.3.1 Scoping Methodology............................. 2-11 2.1.3.2 Screening Methodology............................ 2-33 2.1.4 Conclusions............................................ 2-37 2.1.5 References

............................................ 2-38 2.2 Plant-Level Scoping Results....................................... 2-39 2.2.1 Summary of Technical Information in the Application............. 2-39 2.2.2 Staff Evaluation......................................... 2-40 2.2.3 Conclusions............................................ 2-41 2.3 Scoping and Screening Results: Mechanical Systems

.................. 2-41 2.3.1 Reactor Vessel, Internals, and Reactor Coolant System.......... 2-43 2.3.1.1 Reactor Vessel.................................. 2-43 2.3.1.2 Reactor Vessel Internals........................... 2-47 2.3.1.3 Reactor Coolant System........................... 2-53 2.3.1.4 Evaluation Findings............................... 2-66 2.3.1.5 References...................................... 2-66 2.3.2 Engineered Safety Features Systems........................ 2-67 2.3.2.1 High-Pressure Coolant Injection System............... 2-67 2.3.2.2 Core Spray System............................... 2-69 2.3.2.3 Containment Isolation Components and Primary Containment Piping System.................................... 2-72

-vi-2.3.2.4 Reactor Core Isolation Cooling SystemQuad Cities Only

............................................... 2-78 2.3.2.5 Isolation CondenserDresden Only.................. 2-81 2.3.2.6 Residual Heat Removal SystemQuad Cities Only...... 2-86 2.3.2.7 Low-Pressure Coolant Injection SystemDresden Only... 2-89 2.3.2.8 Standby Liquid Control System...................... 2-92 2.3.2.9 Standby Gas Treatment System..................... 2-95 2.3.2.10 Automatic Depressurization System................. 2-101 2.3.2.11 Anticipated Transient Without Scram System.......... 2-102 2.3.2.12 Evaluation Findings............................. 2-104 2.3.2.13 References.................................... 2-104 2.3.3 Auxiliary Systems....................................... 2-104 2.3.3.1 Refueling Equipment............................. 2-104 2.3.3.2 Shutdown Cooling System (Dresden only)............. 2-107 2.3.3.3 Control Rod Drive Hydraulic System................. 2-111 2.3.3.4 Reactor Water Cleanup System.................... 2-119 2.3.3.5 Fire Protection System........................... 2-122 2.3.3.6 Emergency Diesel Generator and Auxiliaries........... 2-128 2.3.3.6A Diesel Generator Room Ventilation................. 2-132 2.3.3.7 Main Control Room Heating, Ventilation, and Air Conditioning.................................... 2-134 2.3.3.8 HVACReactor Building.......................... 2-138 2.3.3.9 Emergency Core Cooling System Corner Room HVAC... 2-140 2.3.3.10 Station Blackout Building HVAC................... 2-143 2.3.3.11 Station Blackout System (Diesel and Auxiliaries)....... 2-146 2.3.3.12 Diesel Generator Cooling Water System............. 2-148 2.3.3.13 Diesel Fuel Oil System

.......................... 2-151 2.3.3.14 Process Sampling System........................ 2-154 2.3.3.15 Carbon Dioxide System.......................... 2-157 2.3.3.16 Service Water System........................... 2-159 2.3.3.17 Reactor Building Closed Cooling Water System....... 2-162 2.3.3.18 Turbine Building Closed Cooling Water System....... 2-165 2.3.3.19 Demineralizer Water Makeup System............... 2-168 2.3.3.20 Residual Heat Removal Service Water System (Quad Cities Only)........................................ 2-174 2.3.3.21 Containment Cooling Service Water System (Dresden Only).................................. 2-180 2.3.3.22 Ultimate Heat Sink System....................... 2-184 2.3.3.23 Fuel Pool Cooling and Filter Demineralizer System..... 2-187 2.3.3.24 Plant Heating System........................... 2-192 2.3.3.25 Containment Atmosphere Monitoring System......... 2-195 2.3.3.26 Nitrogen Containment Atmosphere Dilution System.... 2-197 2.3.3.27 Drywell Nitrogen Inerting System................... 2-199 2.3.3.28 Safe Shutdown Makeup Pump System (Quad Cities Only)........................................ 2-202 2.3.3.29 Circulating Water System........................ 2-208 2.3.3.30 Dresden Laundry Treatment System................ 2-210 2.3.3.31 Zinc Injection System

........................... 2-212 2.3.4 Steam and Power Conversion Systems...................... 2-214

-vii-2.3.4.1 Main Steam System

............................. 2-214 2.3.4.2 Feedwater System............................... 2-221 2.3.4.3 Condensate and Condensate Storage Systems........ 2-225 2.3.4.4 Main Condenser................................ 2-228 2.3.4.5 Main Turbine and Auxiliary Systems................. 2-232 2.3.4.6 Turbine Oil System.............................. 2-235 2.3.4.7 Main Generator and Auxiliary

...................... 2-238 2.3.4.8 Extraction Steam System (Quad Cities Only).......... 2-241 2.3.4.9 Feedwater Heater Drains and Vents System........... 2-243 2.4 Scoping and Screening Results: Structures.......................... 2-246 2.4.1 Primary Containment.................................... 2-250 2.4.1.1 Summary of Technical Information in the Application.... 2-250 2.4.1.2 Staff Evaluation................................. 2-253 2.4.1.3 Conclusions.................................... 2-257 2.4.2 Reactor Building........................................ 2-258 2.4.2.1 Summary of Technical Information in the Application.... 2-258 2.4.2.2 Staff Evaluation................................. 2-260 2.4.2.3 Conclusions.................................... 2-262 2.4.3 Main Control Room and Auxiliary Electric Equipment Room...... 2-262 2.4.3.1 Summary of Technical Information in the Application.... 2-262 2.4.3.2 Staff Evaluation................................. 2-264 2.4.3.3 Conclusions.................................... 2-265 2.4.4 Turbine Building........................................ 2-265 2.4.4.1 Summary of Technical Information in the Application.... 2-265 2.4.4.2 Staff Evaluation................................. 2-267 2.4.4.3 Conclusions.................................... 2-267 2.4.5 Diesel Generator Buildings................................ 2-268 2.4.5.1 Summary of Technical Information in the Application.... 2-268 2.4.5.2 Staff Evaluation................................. 2-270 2.4.5.3 Conclusions.................................... 2-270 2.4.6 Station Blackout Building and Yard Structures................. 2-270 2.4.6.1 Summary of Technical Information in the Application.... 2-270 2.4.6.2 Staff Evaluation................................. 2-272 2.4.6.3 Conclusions.................................... 2-274 2.4.7 Isolation Condenser Pump House (Dresden).................. 2-275 2.4.7.1 Summary of Technical Information in the Application.... 2-275 2.4.7.2 Staff Evaluation................................. 2-276 2.4.7.3 Conclusions.................................... 2-276 2.4.8 Makeup Demineralizer Building (Dresden).................... 2-277 2.4.8.1 Summary of Technical Information in the Application.... 2-277 2.4.8.2 Staff Evaluation................................. 2-278 2.4.8.3 Conclusions.................................... 2-278 2.4.9 Radwaste Floor Drain Surge Tank.......................... 2-278 2.4.9.1 Summary of Technical Information in the Application.... 2-278 2.4.9.2 Staff Evaluation................................. 2-279 2.4.9.3 Conclusions.................................... 2-280 2.4.10 Miscellaneous Foundations.............................. 2-280 2.4.10.1 Summary of Technical Information in the Application... 2-280 2.4.10.2 Staff Evaluation................................ 2-281

-viii-2.4.10.3 Conclusions................................... 2-282 2.4.11 Crib House........................................... 2-282 2.4.11.1 Summary of Technical Information in the Application... 2-282 2.4.11.2 Staff Evaluation................................ 2-284 2.4.11.3 Conclusions................................... 2-289 2.4.12 Unit 1 Crib House (Dresden)............................. 2-289 2.4.12.1 Summary of Technical Information in the Application... 2-289 2.4.12.2 Staff Evaluation................................ 2-290 2.4.12.3 Conclusions................................... 2-291 2.4.13 Station Chimney....................................... 2-292 2.4.13.1 Summary of Technical Information in the Application... 2-292 2.4.13.2 Staff Evaluation................................ 2-293 2.4.13.3 Conclusions................................... 2-293 2.4.14 Cranes and Hoists..................................... 2-293 2.4.14.1 Summary of Technical Information in the Application... 2-293 2.4.14.2 Staff Evaluation................................ 2-294 2.4.14.3 Conclusions................................... 2-300 2.4.15 Component Supports Commodity Group.................... 2-300 2.4.15.1 Summary of Technical Information in the Application... 2-300 2.4.15.2 Staff Evaluation................................ 2-301 2.4.15.3 Conclusions................................... 2-304 2.4.16 Insulation Commodity Group............................. 2-304 2.4.16.1 Summary of Technical Information in the Application... 2-304 2.4.16.2 Staff Evaluation................................ 2-305 2.4.16.3 Conclusions................................... 2-309 2.5 Scoping and Screening Results: Electrical and Instrumentation and Controls 2-309 2.5.1 Insulated Cables and Connections.......................... 2-310 2.5.1.1 Summary of Technical Information in the Application.... 2-310 2.5.1.2 Staff Evaluation................................. 2-310 2.5.1.3 Conclusions.................................... 2-311 2.5.2 Bus Duct............................................. 2-311 2.5.2.1 Summary of Technical Information in the Application.... 2-311 2.5.2.2 Staff Evaluation................................. 2-312 2.5.2.3 Conclusions.................................... 2-312 2.5.3 High Voltage Transmission Conductors and Insulators.......... 2-312 2.5.3.1 Summary of Technical Information in the Application.... 2-312 2.5.3.2 Staff Evaluation................................. 2-313 2.5.3.3 Conclusions.................................... 2-313 2.5.4 Electrical/I&C Penetration................................ 2-313 2.5.4.1 Summary of Technical Information in the Application.... 2-314 2.5.4.2 Staff Evaluation................................. 2-314 2.5.4.3 Conclusions.................................... 2-314 2.5.5 References

........................................... 2-315

3. Aging Management Review............................................... 3-1 3.0 Aging Management Review........................................ 3-1 3.0.1 The GALL Format for the LRA............................... 3-2 3.0.2 The Staffs Review Process for GALL......................... 3-3

-ix-3.0.3 Common Aging Management Programs....................... 3-5 3.0.3.1 ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD (B.1.1)................................... 3-8 3.0.3.2 Water Chemistry Program (B.1.2).................... 3-10 3.0.3.3 BWR Stress Corrosion Cracking (B.1.7)............... 3-14 3.0.3.4 Flow-Accelerated Corrosion (B.1.11).................. 3-16 3.0.3.5 Bolting Integrity (B.1.12)........................... 3-18 3.0.3.6 Open-Cycle Cooling Water System (B.1.13)............ 3-21 3.0.3.7 Closed-Cycle Cooling Water System (B.1.14)........... 3-27 3.0.3.8 Compressed Air Monitoring (B.1.16).................. 3-30 3.0.3.9 Above Ground Carbon Steel Tanks (B.1.20)............ 3-33 3.0.3.10 One Time Inspection (B.1.23)...................... 3-36 3.0.3.11 Selective Leaching of Materials (B.1.24).............. 3-40 3.0.3.12 Buried Piping and Tanks Inspection (B.1.25)........... 3-43 3.0.3.13 10 CFR Part 50, Appendix J (B.1.28)

................ 3-47 3.0.3.14 Structures Monitoring Program (B.1.30).............. 3-49 3.0.3.15 Heat Exchanger Test and Inspection Activities (B.2.6)... 3-53 3.0.3.16 Lube Oil Monitoring Activities (B.2.5)................. 3-58 3.0.3.17 Periodic Inspection of Ventilation System Elastomers (B.2.3).......................................... 3-62 3.0.3.18 Periodic Inspection of Components Subject to Moist Environments (B.2.9).............................. 3-67 3.0.4 Quality Assurance For Aging Management Programs............ 3-69 3.0.4.1 Summary of Technical Information in Application........ 3-70 3.0.4.2 Staff Evaluation.................................. 3-71 3.0.4.3 Conclusions..................................... 3-72 3.0.4.4 References..................................... 3-72 3.0.5 Aging Management Review for Additional In-Scope Components.... 3-72 3.1 Reactor Vessel, Internals, and Reactor Coolant System................. 3-86 3.1.1 Summary of Technical Information in the Application............. 3-86 3.1.2 Staff Evaluation......................................... 3-86 3.1.2.1 Aging Management Evaluations in the GALL Report That Are Relied on for License Renewal, Which Do Not Require Further Evaluation....................................... 3-89 3.1.2.2 Aging Management Evaluations in the GALL Report That Are Relied on for License Renewal, For Which GALL Recommends Further Evaluation................................. 3-90 3.1.2.3 Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Components................. 3-98 3.1.2.4 Aging Management Review of Reactor Vessel, Vessel Internals, and Reactor Coolant System....................... 3-124 3.2 Engineered Safety Features Systems

.............................. 3-175 3.2.1 Summary of Technical Information in the Application............ 3-176 3.2.2 Staff Evaluation........................................ 3-176 3.2.2.1 Aging Management Evaluations in the GALL Report That Are Relied on for License Renewal, Which Do Not Require Further Evaluation...................................... 3-179

-x-3.2.2.2 Aging Management Evaluations in the GALL Report That Are Relied on for License Renewal, for Which GALL Recommends Further Evaluation................................ 3-179 3.2.2.3 Aging Management Programs for ESF System Components

.............................................. 3-187 3.2.2.4 Aging Management Review of Plant-Specific Engineered Safety Features Systems Components..................... 3-191 3.3 Auxiliary Systems.............................................. 3-223 3.3.1 Summary of Technical Information in the Application............ 3-223 3.3.2 Staff Evaluation........................................ 3-224 3.3.2.1 Aging Management Evaluations in the GALL Report That Are Relied on for License Renewal, Which Do Not Require Further Evaluation...................................... 3-228 3.3.2.2 Aging Management Evaluations in the GALL Report That Are Relied on for License Renewal, For Which GALL Recommends Further Evaluation................................ 3-228 3.3.2.3 Aging Management Programs for Auxiliary System Components

.............................................. 3-234 3.3.2.4 Aging Management Review of Auxiliary Systems....... 3-257 3.3.2.5 General Aging Management Review Issues........... 3-356 3.4 Steam and Power Conversion Systems.............................. 3-367 3.4.1 Summary of Technical Information in the Application............ 3-368 3.4.2 Staff Evaluation........................................ 3-368 3.4.2.1 Aging Management Evaluations in the GALL Report That Are Relied on for License Renewal, Which Do Not Require Further Evaluation...................................... 3-370 3.4.2.2 Aging Management Evaluations in the GALL Report That Are Relied on for License Renewal, for Which GALL Recommends Further Evaluation................................ 3-370 3.4.2.3 Aging Management Program for the Steam and Power Conversion System Components.................... 3-372 3.4.2.4 Aging Management Review of the Steam and Power Conversion Systems....................................... 3-374 3.5 Containment, Structures, and Component Supports.................... 3-389 3.5.1 Summary of Technical Information in the Application............ 3-390 3.5.2 Staff Evaluation........................................ 3-390 3.5.2.1 Aging Management Evaluations in the GALL Report That Are Relied on for License Renewal, Which Do Not Require Further Evaluation...................................... 3-397 3.5.2.2 Aging Management Evaluations in the GALL Report That Are Relied on for License Renewal, for Which the GALL Report Recommends Further Evaluation.................... 3-397 3.5.2.3 Aging Management Programs for Containment, Structures, and Component Supports............................. 3-415 3.5.2.4 Aging Management Review of Plant-Specific Structures and Structural Components............................ 3-432 3.6 Electrical and Instrumentation and Controls.......................... 3-461 3.6.1 Summary of Technical Information in the Application............ 3-462 3.6.2 Staff Evaluation........................................ 3-462

-xi-3.6.2.1 Aging Management Evaluations in the GALL Report that Are Relied on for License Renewal, Which Do Not Require Further Evaluation...................................... 3-464 3.6.2.2 Aging Management Evaluation in the GALL Report That Are Relied on for License Renewal, For Which GALL Recommends Further Evaluation................................ 3-465 3.6.2.3 Aging Management Programs for Electrical and I&C Components

.............................................. 3-465 3.6.2.4 Aging Management of Plant-Specific Components...... 3-479 3.7 Conclusion for Aging Management................................. 3-492

4. Time-Limited Aging Analyses.............................................. 4-1 4.1 Identification of Time-Limited Aging Analyses

.......................... 4-1 4.1.1 Summary of Technical Information in the Application.............. 4-1 4.1.2 Staff Evaluation.......................................... 4-2 4.1.3 Conclusions............................................. 4-2 4.2 Reactor Vessel and Internals Neutron Embrittlement..................... 4-2 4.2.1 Summary of Technical Information in the Application.............. 4-3 4.2.1.1 Reactor Vessel Materials Upper-Shelf Energy Reduction Due to Neutron Embrittlement.............................. 4-4 4.2.1.2 Adjusted Reference Temperature for Reactor Vessel Materials Due to Neutron Embrittlement......................... 4-4 4.2.1.3 Reflood Thermal Shock Analysis of the Reactor Vessel.... 4-4 4.2.1.4 Reflood Thermal Shock Analysis of the Reactor Vessel Core Shroud and Repair Hardware......................... 4-5 4.2.1.5 Reactor Vessel Thermal Limit AnalysesOperating Pressure-Temperature Limits................................ 4-5 4.2.1.6 Reactor Vessel Circumferential Weld Examination Relief... 4-6 4.2.1.7 Reactor Vessel Axial Weld Failure Probability............ 4-6 4.2.2 Staff Evaluation.......................................... 4-7 4.2.2.1 Reactor Vessel Materials Upper-Shelf Energy Reduction Due to Neutron Embrittlement.............................. 4-7 4.2.2.2 Adjusted Reference Temperature for Reactor Vessel Materials Due to Neutron Embrittlement........................ 4-12 4.2.2.3 Reflood Thermal Shock Analysis of the Reactor Vessel... 4-13 4.2.2.4 Reflood Thermal Shock Analysis of the Reactor Vessel Core Shroud and Repair Hardware........................ 4-14 4.2.2.5 Reactor Vessel Thermal Limit AnalysesOperating Pressure-Temperature Limits................................ 4-15 4.2.2.6 Reactor Vessel Circumferential Weld Examination Relief.. 4-16 4.2.2.7 Reactor Vessel Axial Weld Failure Probability........... 4-18 4.2.3 Conclusions............................................ 4-21 4.3 Metal Fatigue.................................................. 4-22 4.3.1 Summary of Technical Information in the Application............. 4-22 4.3.1.1 Reactor Vessel Fatigue Analyses.................... 4-22 4.3.1.2 Fatigue Analysis of the Reactor Internals.............. 4-23 4.3.1.3 ASME Code,Section III, Class 1 Reactor Coolant Pressure Boundary Piping and Component Fatigue Analysis

....... 4-24

-xii-4.3.1.4 Reactor Coolant Pressure Boundary Piping and Components Designed to USAS B31.1, ASME Code,Section III, Classes 2 and 3, or ASME Code,Section VIII, Classes B and C......... 4-25 4.3.1.5 Fatigue Analysis of the Isolation Condenser............ 4-25 4.3.1.6 Effects of Reactor Coolant Environment on Fatigue Life of Components and Piping (Generic Safety Issue 190)

...... 4-26 4.3.2 Staff Evaluation......................................... 4-27 4.3.2.1 Reactor Vessel Fatigue Analysis..................... 4-27 4.3.2.2 Fatigue Analysis of the Reactor Internals.............. 4-27 4.3.2.3 ASME Code,Section III, Class 1 Reactor Coolant Pressure Boundary Piping and Component Fatigue Analysis

....... 4-28 4.3.2.4 Reactor Coolant Pressure Boundary Piping and Components Designed to USAS B31.1, ASME Code,Section III, Classes 2 and 3, or ASME Code,Section VIII, Classes B and C......... 4-29 4.3.2.5 Fatigue Analysis of the Isolation Condenser............ 4-29 4.3.2.6 Effects of Reactor Coolant Environment on Fatigue Life of Components and Piping (Generic Safety Issue 190)

...... 4-30 4.3.3 Conclusions............................................ 4-31 4.4 Environmental Qualification....................................... 4-32 4.4.1 Environmental Qualification Program TLAA.................... 4-32 4.4.1.1 Summary of Technical Information in the Application..... 4-32 4.4.1.2 Staff Evaluation.................................. 4-35 4.4.1.3 Conclusions..................................... 4-35 4.4.2 Generic Safety Issue 168, Environmental Qualification of Low-Voltage Instrumentation and Control (I&C) Cables.................... 4-35 4.4.2.1 Summary of Technical Information in the Application..... 4-35 4.4.2.2 Staff Evaluation.................................. 4-36 4.4.2.3 Conclusions..................................... 4-37 4.5 Loss of Prestress in Concrete Containment Tendons.................... 4-37 4.6 Fatigue of Primary Containment, Attached Piping, and Components........ 4-37 4.6.1 Summary of Technical Information in the Application............. 4-38 4.6.1.1 Fatigue Analysis of the Suppression Chamber, Vents and Downcomers..................................... 4-38 4.6.1.2 Fatigue Analysis of Safety Relief Valve Discharge Piping Inside the Suppression Chamber, External Suppression Chamber Attached Piping, and Associated Penetrations........... 4-39 4.6.1.3 Drywell-to-Suppression Chamber Vent Line Bellows Fatigue Analyses........................................ 4-40 4.6.1.4 Primary Containment Process Penetration Bellows Fatigue Analysis........................................ 4-40 4.6.2 Staff Evaluation......................................... 4-41 4.6.2.1 Fatigue Analysis of the Suppression Chamber, Vents, and Downcomers..................................... 4-41 4.6.2.2 Fatigue Analysis of Safety Relief Valve Discharge Piping Inside the Suppression Chamber, External Suppression Chamber Attached Piping, and Associated Penetrations........... 4-42 4.6.2.3 Drywell-to-Suppression Chamber Vent Line Bellows Fatigue Analyses........................................ 4-43

-xiii-4.6.2.4 Primary Containment Process Penetration Bellows Fatigue Analysis........................................ 4-43 4.6.3 Conclusions............................................ 4-43 4.7 Other Plant-Specific Time-Limited Aging Analyses...................... 4-44 4.7.1 Reactor Building Crane Load Cycles......................... 4-44 4.7.1.1 Summary of Technical Information in the Application..... 4-44 4.7.1.2 Staff Evaluation.................................. 4-44 4.7.1.3 Conclusions..................................... 4-45 4.7.2 Metal Corrosion......................................... 4-45 4.7.2.1 Corrosion Allowance for Power-Operated Relief Valves... 4-45 4.7.2.2 Degradation Rates of Inaccessible Exterior Drywell Plate Surfaces........................................ 4-46 4.7.2.3 Galvanic Corrosion in the Containment Shell and Attached Piping Components due to Stainless Steel Emergency Core Cooling System Suction Strainers........................... 4-52 4.7.3 Crack Growth Calculation of a Postulated Flaw in the Heat Affected Zone of an Arc Strike in the Suppression Chamber Shell............. 4-54 4.7.3.1 Summary of Technical Information in the Application..... 4-54 4.7.3.2 Staff Evaluation.................................. 4-55 4.7.3.3 Conclusions..................................... 4-55 4.7.4 Radiation Degradation of Drywell Shell Expansion Gap Polyurethane Foam................................................ 4-56 4.7.4.1 Summary of Technical Information in the Application..... 4-56 4.7.4.2 Staff Evaluation.................................. 4-56 4.7.4.3 Conclusions..................................... 4-57 4.7.5 High-Energy Line Break Postulation Based on Fatigue Cumulative Usage Factor.......................................... 4-57 4.8 Conclusion for Time-Limited Aging Analyses.......................... 4-57

5. Review by the Advisory Committee on Reactor Safeguards

...................... 5-1

6. Conclusions........................................................... 6-1 Appendices Appendix A: Commitments for License Renewal.................................. A-1 Appendix B: Chronology

................................................... B-1 Appendix C: Principal Contributors............................................ C-1 Appendix D: References................................................... D-1 Tables 3.0.3-1 Common Aging Management Programs................................. 3-5 3.0.3-2 System-Specific Management Programs................................. 3-6

-xiv-3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report................ 3-87 3.2-1 Staff Evaluation for Dresden and Quad Cities Engineered Safety Features System Components in the GALL Report...................................... 3-177 3.3-1 Summary of Aging Management for Auxiliary Systems Evaluated in Chapter VII of the GALL Report..................................................... 3-225 Fire Protection Portions of Table 3.3-3Staff Evaluation Table for Dresden and Quad Cities Auxiliary System Components Evaluated in the GALL Report........... 3-248 3.4-1 Summary of Aging Management Programs for Steam and Power Conversion Systems Evaluated in Chapter VII of the GALL Report............................ 3-369 3.5-1 Staff Evaluation for Structures and Structural Components in NUREG-1801 (GALL)Common Components of All Types of PWR and BWR Containment... 3-391 3.5-2 PWR Concrete (Reinforced and Prestressed) and Steel Containment, BWR Concrete (Mark II and III) and Steel (Mark I, II, and III) Containment.................. 3-392 3.5-3 Class I Structures................................................. 3-394 3.5-4 Component Supports............................................... 3-396 3.6-1 Staff Evaluation Table for Dresden and Quad Cities Electrical Components Evaluated in the GALL Report.................................................. 3-463 4.2-1 Comparison of Results from Staff and BWRVIP Calculations................. 4-19

-xv-ABBREVIATIONS A

ampacity AC alternating current ACAD atmospheric containment air dilution system ACI American Concrete Institute ACRS Advisory Committee on Reactor Safeguards ACSR aluminum conductor steel reinforced ADAMS Agencywide Document Access and Management System ADS Automatic depressurization system AEC Atomic Energy Commission AERM aging effect requiring management AFW auxiliary feedwater AFU air filtration unit AHU air handling unit AISC American Institute of Steel Construction ALARA as low as reasonably achievable AMG aging management guideline AMP aging management program AMR aging management review AMSAC ATWS mitigation system actuation circuitry ANS American Nuclear Society ANSI American National Standards Institute AOR abnormal occurrence report APCSB Auxiliary and Power Conversion System Branch AR action report ARI alternate rod insertion ART adjusted reference temperature ASME American Society of Mechanical Engineers ASNT American Society for Nondestructive Testing ASTM American Society for Testing and Materials ATWS anticipated transient without a scram AVT all-volatile treatment AWG American wire gauge BMI bottom mounted instrument BTP branch technical position B&W Babcock and Wilcox BWR boiling-water reactor BWROG Boiling Water Reactor Owners Group BWRVIP Boiling Water Reactor Vessel and Internals Project C

Celsius CAM containment atmospheric monitoring CAR corrective action report CASS cast austenitic stainless steel CBF cycle-based fatigue CCST contaminated condensate storage tank

-xvi-CCSW containment cooling service water CCW component cooling water CDF core damage frequency CDHR c ontainer hydrogen detectors and recombiner CD-ROM compact disk-read only memory CDWST clean demineralized water storage tank CF chemistry factor CFR Code of Federal Regulations CI confirmatory item CIC contaminant isolation component CLB current licensing basis CMAA Crane Manufactures Association of America CO carbon monoxide CO2 carbon dioxide ComEd Commonwealth Edison C-RAI clarification of request for additional information CRD control rod drive CRDA control rod drop accident CRD(H) control rod drive (hydraulic)

CRDM control rod drive mechanism CR HVAC air handles heating/cooling CS containment spray system or carbon steel CST condensate storage tank CUF cumulative usage factor CV check valve CW circulating water D/QCNPS Dresden/Quad Cities Nuclear Power Station DAM data acquisition modules DBA design-basis accident DBD design baseline document DBE design-basis event DC direct current DFO diesel fuel oil DG diesel generator DGB-HVAC diesel generator building heating, ventilation, and air conditioning DGCW diesel generator cooling water DGSW diesel generator service water DNI drywell nitrogen inerting DNPIS drywell nitrogen purge and inerting system DNPS Dresden Nuclear Power Station DPR developmental power reactor DR drywell-to-refueling D-RAI draft request for additional information DRS development requirements specification DSER draft safety evaluation report DWM demineralizer water makeup EC engineering change

-xvii-ECCS emergency core cooling system ECP electrochemical corrosion potential ECR-HVAC emergency core cooling system corner room heating, ventilation, and air conditioning EDG emergency diesel generator EDY effective degradation years EFPY effective full-power year EFWST emergency feedwater storage tank EGC Exelon Generation Company, LLC EHC electrohydraulic control EMA equivalent margin analysis EPN equipment part number EPR ethylene propylene rubber EPRI Electric Power Research Institute EPU extended power uprate EQ environmental qualification ESF engineered safety features ESS electronic switching system ESW electroslag weld EWCS electronic work control system F

Fahrenheit FAC flow-accelerated corrosion FCC Federal Communications Commission Fen environmental fatigue multiplier FERC Federal Energy Regulatory Commission FOI factor of improvement FP fire protection FPP fire protection program FRP fiberglass reinforced plastic FSAR final safety analysis report FSER final safety evaluation report FSSD fire safe shutdown ft foot, feet ft-lb foot-pound FWRV feedwater regulating valve GALL Generic Aging Lessons Learned (Report)

GE General Electric GEIS generic environmental impact statement GL generic letter GSI generic safety issue GTR generic technical report HCU hydraulic control unit HELB high energy line break HEPA high efficiency particulate air

-xviii-HIC high integrity container HPCI high pressure coolant injection system HRRM high-range radiation monitor HRSS high radiation sampling system HSLAS high strength low alloy steel HSO hydrogen seal oil HTK high temperature kerite HVAC heating, ventilation, and air conditioning HWC hydrogren water chemistry HX heat exchanger I&C instrumentation and control IA instrument air IASCC irradiation-assisted stress-corrosion cracking ID inner diameter IDR inspection discrepancy report IE Inspection and Enforcement, Office of (NRC)

IEB inspection and enforcement bulletin IGA intergranular attack IGSCC intergranular stress-corrosion cracking IHSI induction heat stress improvement ILRT integrated leak rate test in.

inch, inches IN information notice INPO Institute of Nuclear Power Operations IPA integrated plant assessment IR insulation resistance IRM intermediate range monitor ISG interim staff guidance ISI inservice inspection ISP Integrated Surveillance Program IST inservice testing J

joule Keff effective multiplication factor Kip one thousand pounds KV kilovolt LBB leak before break LER licensee event report LLRT local leak rate test LOCA loss-of-coolant accident LOOP loss of offsite power LPCI low pressure coolant injection system LPRM local power range monitor

-xix-LR license renewal LRA license renewal application LRTI license renewal technical instruction LTOP low-temperature over-pressurization LWR light water reactor m

margin M/G motor generator MCC motor control center MeV one million electron volts MG motor generator MIC microbiologically influenced corrosion mil/y mils per year MOV motor-operated valve MRV minimum required value MSIP mechanical stress improvement method MSIV main steam isolation valve MSL mean sea level MSV main stop valve MSIV main steam isolation valve MT magnetic particle test MWe megawatt-electric NACE National Association of Corrosion Engineers NaOH sodium hydroxide NBI nuclear boiler instrumentation NCAD nitrogen containment atmospheric dilution NCR nonconformance report NDE nondestructive examination ND-QAP Quality Assurance Program for Station Operation NEI Nuclear Energy Institute NEPA National Environmental Policy Act of 1969 NFPA National Fire Protection Association NIS nuclear instrumentation system NMCA noble metal chemical application NNS non-nuclear safety NPAR nuclear plant aging research NPS nominal pipe size NPSH net positive suction head NRC U.S. Nuclear Regulatory Commission NSR non-safety-related NSSS nuclear steam supply system NUMARC Nuclear Management and Resources Council (now NEI)

NUREG Nuclear Regulatory Commission technical report NUREG/CR NUREG contractor report

-xx-OBE operating based event OD outside diameter OE operating experience ODSCC outside diameter stress-corrosion cracking OI open item OPT operability test P(F/E)

Conditional failure probability P&I piping and instrumentation P&ID piping and instrumentation diagram PC primary containment PCIS primary containment isolation system PDT piping design table PLL predicted lower limit PM preventive maintenance PMT post-maintenance test PORV power operated relief valve ppm parts per million PS process sampling psi pounds per square inch psig pounds per square inch gauge PT penetrant test P-T pressure-temperature PTS pressurized thermal shock PUAR plant-unique analysis reports PVC polyvinyl chloride PWHT postweld heat treatment PWR pressurized-water reactor QA quality assurance QAP quality action plan QCNPS Quad Cities Nuclear Power Station RAI request for additional information RAT reserve auxiliary transformer RBCCW reactor building closed cooling water RBH-HVAC reactor building heating, ventilation, and air conditioning RCCA rod cluster control assembly RCIC reactor core isolation cooling RCP reactor coolant pump RCPB reactor coolant pressure boundary RCS reactor coolant system RCU refrigeration condensing unit RFP reactor feed pump RG regulatory guide RHR residual heat removal

-xxi-RHRSW residual heat removal service water RIS Regulatory Issue Summary RMS radiation monitoring system RPS reactor protection system RPV reactor pressure vessel RT radiographic test RTD resistance temperature detector RTPTS reference temperature for pressurized thermal shock RTNDT reference nil ductility transition temperature RTV room temperature vulcanizing RV reactor vessel RVI reactor vessel internals RVID Reactor Vessel Integrity Database RVWLIS reactor vessel water level instrumentation system RWCU reactor water cleanup RWM rod worth minimizer RWST refueling water storage tank Sa stress intensity SAFW standby auxiliary feedwater SBGT standby gas treatment SBLC standby liquid control SBO station blackout SC structures and components SC-1 safety class 1 SC-2 safety class 2 SC-3 safety class 3 SCC stress-corrosion cracking SDCS shutdown cooling system SE safety evaluation SER safety evaluation report SFP spent fuel pool SFC&FS spent fuel cooling and fuel storage SFR system function report SG steam generator SI safety injection SIT structural integrity test SJAERM steam jet air ejector radiation monitor SLC standby liquid control SOC Statements of Consideration SOER significant operating event report SPCS steam and power conversion systems SPING system-level particulate, iodine, and noble gas monitors SRM source range monitor SRP Standard Review Plan SRP-LR standard review planlicense renewal

-xxii-SRV safety relief valve SS safety significant or stainless steel SSC structure, system, and component SSEL safe-shutdown equipment list SSMP safe shutdown makeup pump system SV safety valve SW service water T

thickness, also temperature TASCS thermal stratification, cycling, and stripping TBCCW turbine building closed cooling water TCV turbine control valve TDAFW turbine-driven auxiliary feedwater TID total integrated dose TIP transverse incore probe TLAA time-limited aging analysis TR topical report TRM technical requirements manual TSC technical support center TS technical specification TT thermal transients UFSAR updated final safety analysis report USAS United States of America Standard USE upper-shelf energy UT ultrasonic testing UV ultraviolet V

volt VAC volts alternating current VT visual test wt.%

percent by weight

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