ML030370649
| ML030370649 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 02/05/2003 |
| From: | Exelon Generation Co |
| To: | Office of Nuclear Reactor Regulation |
| Solorio, D L, NRR/DRIP/RLEP, 415-1973 | |
| References | |
| Download: ML030370649 (22) | |
Text
Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3 Docket Nos. 50-277 and 50-278 Exelon Generation Company, LLC (Exelon)
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation February, 2003
-ii-ABSTRACT This document is a safety evaluation report regarding the application to renew the operating licenses for Peach Bottom Atomic Power Station, Units 2 and 3. The application was filed by the Exelon Generation Company LLC, (Exelon) by letter dated July 2, 2001. The Office of Nuclear Reactor Regulation has reviewed the Peach Bottom Atomic Power Station, Units 2 and 3, license renewal application for compliance with the requirements of Title 10 of the Code of Federal Regulations, Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, and prepared this report to document its findings.
In its submittal of July 2, 2001, the Exelon requested renewal of the Peach Bottom, Units 2 and 3, operating licenses (License Nos. DPR-44 and DPR-56, respectively), which were issued under Section 104b of the Atomic Energy Act of 1954, as amended, for a period of 20 years beyond the current license expiration dates of August 8, 2013, and July 2, 2014, respectively.
The Peach Bottom Atomic Power Station is a two-unit nuclear power plant located in York County and Lancaster County in southeastern Pennsylvania. Each unit consists of a General Electric boiling-water reactor nuclear steam supply system designed to generate 3458 megawatts thermal or 1093 megawatts electric.
The NRC license renewal project manager for Peach Bottom, Units 2 and 3, is David Solorio.
Mr. Solorio may be contacted by calling 301-415-1973 or by writing to the License Renewal and Environmental Impacts Program, U.S. Nuclear Regulatory Commission, Washington, DC 20555-001.
-iii-TABLE OF CONTENTS PEACH BOTTOM, UNITS 2 AND 3 SAFETY EVALUATION REPORT ABSTRACT............................................................... -ii-ACRONYMS............................................................ -viii-1 INTRODUCTION AND GENERAL DISCUSSION............................... 1-1 1.1 Introduction..................................................... 1-1 1.2 License Renewal Background...................................... 1-2 1.2.1 Safety Review........................................... 1-2 1.2.2 Environmental Review..................................... 1-4 1.3 Summary of the Principal Review Matters............................. 1-5 1.3.1 Boiling Water Reactor Vessel Internals Project (BWRVIP) Topical Reports
...................................................... 1-6 1.4 Summary of Open Items........................................... 1-7 1.5 Summary of Confirmatory Items..................................... 1-7 1.6 Summary of Proposed License Conditions............................. 1-8 2 STRUCTURES AND COMPONENTS SUBJECT TO AN AGING MANAGEMENT REVIEW
........................................................................ 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.............................. 2-2 2.1.2.2 Screening Methodology............................ 2-6 2.1.3 Staff Evaluation.......................................... 2-8 2.1.3.1 Evaluation of the Methodology for Identifying Systems, Structures, and Components Within the Scope of License Renewal. 2-8 2.1.3.2 Evaluation of the Methodology for Identifying Structures and Components Subject to an Aging Management Review
... 2-13 2.1.4 Conclusions............................................ 2-14 2.2 Plant-Level Scoping Results....................................... 2-15 2.2.1 Introduction............................................ 2-15 2.2.2 Summary of Technical Information in the Application............. 2-15 2.2.2.1 Systems, Structures, and Components Within the Scope of License Renewal............................................. 2-16 2.2.2.2 Systems and Structures Not Within the Scope of License Renewal............................................. 2-17 2.2.3 Staff Evaluation......................................... 2-18 2.2.4 Conclusions............................................ 2-23 2.3 System Scoping and Screening Results Mechanical.................... 2-23 2.3.1 Reactor Coolant System.................................. 2-23 2.3.1.1 Reactor Pressure Vessel and Internals................ 2-23 2.3.1.2 Fuel Assemblies................................. 2-27
-iv-2.3.1.3 Reactor Pressure Vessel Instrumentation System........ 2-28 2.3.1.4 Reactor Recirculation System....................... 2-29 2.3.2 Engineered Safety Features Systems........................ 2-31 2.3.2.1 High Pressure Coolant Injection System............... 2-31 2.3.2.2 Core Spray System............................... 2-33 2.3.2.3 Primary Containment Isolation System................ 2-34 2.3.2.4 Reactor Core Isolation Cooling System................ 2-36 2.3.2.5 Residual Heat Removal System..................... 2-38 2.3.2.6 Containment Atmosphere Control and Dilution System.... 2-40 2.3.2.7 Standby Gas Treatment System..................... 2-44 2.3.2.8 Secondary Containment........................... 2-48 2.3.3 Auxiliary Systems.................................. 2-51 2.3.3.1 Fuel Handling Systems............................ 2-51 2.3.3.2 Fuel Pool Cooling and Cleanup System
............... 2-53 2.3.3.3 Control Rod Drive System.......................... 2-56 2.3.3.4 Standby Liquid Control System...................... 2-58 2.3.3.5 High-Pressure Service Water System................. 2-59 2.3.3.6 Emergency Service Water System................... 2-62 2.3.3.7 Fire Protection System............................ 2-64 2.3.3.8 Control Room Ventilation System.................... 2-69 2.3.3.9 Battery and Emergency Switchgear Ventilation System
... 2-75 2.3.3.10 Diesel Generator Building Ventilation System.......... 2-79 2.3.3.11 Pump Structure Ventilation System.................. 2-81 2.3.3.12 Safety-Grade Instrument Gas System................ 2-84 2.3.3.13 Backup Instrument Nitrogen to ADS................. 2-88 2.3.3.14 Emergency Cooling Water System.................. 2-91 2.3.3.15 Condensate Storage System....................... 2-93 2.3.3.16 Emergency Diesel Generator (EDG)................. 2-96 2.3.3.17 Suppression Pool Temperature Monitoring System...... 2-99 2.3.3.18 Cranes and Hoists.............................. 2-101 2.3.3.19 Non-Safety-Related Systems Affecting Safety-Related Systems
.................................................... 2-105 2.3.4 Steam and Power Conversion Systems...................... 2-112 2.3.4.1 Main Steam System............................. 2-112 2.3.4.2 Main Condenser................................ 2-116 2.3.4.3 Feedwater System............................... 2-118 2.4 Scoping and Screening Results: Structures and Component Supports..... 2-121 2.4.1 Containment Structure................................... 2-122 2.4.2 Reactor Building Structure................................ 2-125 2.4.3 Radwaste Building and Reactor Auxiliary Bay................. 2-127 2.4.4 Turbine Building and Main Control Room Complex............. 2-129 2.4.5 Emergency Cooling Tower and Reservoir.................... 2-132 2.4.6 Station Blackout Structure and Foundations.................. 2-134 2.4.7 Yard Structures........................................ 2-136 2.4.8 Stack................................................ 2-140 2.4.9 Nitrogen Storage Building................................ 2-141 2.4.10 Diesel Generator Building
............................... 2-142 2.4.11 Circulating Water Pump Structure......................... 2-145
-v-2.4.12 Recombiner Building................................... 2-146 2.4.13 Component Supports................................... 2-148 2.4.14 Hazard Barriers and Elastomers.......................... 2-150 2.4.15 Miscellaneous Steel.................................... 2-152 2.4.16 Electrical and Instrumentation Enclosures and Raceways....... 2-153 2.4.17 Insulation............................................ 2-155 2.5 Scoping and Screening Results: Electrical and Instrumentation and Controls 2-156 3 AGING MANAGEMENT REVIEW RESULTS.................................. 3-1 3.0 Common Aging Management Programs............................... 3-1 3.0.1 Introduction............................................. 3-1 3.0.2 Program and Activity Attributes.............................. 3-2 3.0.3 Common Aging Management Programs and Activities
............ 3-3 3.0.3.1 Flow-Accelerated Corrosion Program.................. 3-3 3.0.3.2 Reactor Coolant System Chemistry Program............ 3-6 3.0.3.3 Closed Cooling Water Chemistry..................... 3-11 3.0.3.4 Demineralized Water and Condensate Storage Tank Chemistry Activities........................................ 3-15 3.0.3.5 Torus Water Chemistry Activities..................... 3-18 3.0.3.6 Inservice Inspection Program....................... 3-22 3.0.3.7 Primary Containment Inservice Inspection Program...... 3-28 3.0.3.8 Primary Containment Leakage Rate Testing Program..... 3-32 3.0.3.9 Reactor Pressure Vessel and Internals Inservice Inspection Program........................................ 3-36 3.0.3.10 Inservice Testing Program......................... 3-42 3.0.3.11 Maintenance Rule Structural Monitoring Program....... 3-45 3.0.3.12 Ventilation System Inspection and Testing Activities..... 3-51 3.0.3.13 Outdoor, Buried, and Submerged Component Inspection Activities........................................ 3-53 3.0.3.14 Door Inspection Activities
......................... 3-56 3.0.3.15 Generic Letter 89-13 Activities...................... 3-60 3.0.3.16 Fire Protection Activities.......................... 3-63 3.0.3.17 Heat Exchanger Inspection Activities................. 3-72 3.0.3.18 Lubricating and Fuel Oil Quality Testing Activities....... 3-75 3.0.3.19 One-Time Piping Inspection Activities................ 3-82 3.0.3.20 Reactor Materials Surveillance Program.............. 3-85 3.0.3.21 Torus Piping Inspection Activities................... 3-89 3.0.3.22 Fuel Pool Chemistry Activities...................... 3-91 3.0.4 Quality Assurance Program................................ 3-94 3.1 Aging Management of Reactor Coolant System........................ 3-97 3.1.1 Reactor Pressure Vessel and Internals....................... 3-97 3.1.2 Fuel Assemblies........................................ 3-104 3.1.3 Reactor Pressure Vessel Instrumentation System.............. 3-104 3.1.4 Reactor Recirculation System............................. 3-109 3.2 Aging Management of Engineered Safety Features Systems............. 3-115 3.2.1 High-pressure Coolant Injection............................ 3-115 3.2.2 Core Spray System..................................... 3-127 3.2.3 Primary Containment Isolation System
...................... 3-132
-vi-3.2.4 Reactor Core Isolation Cooling System...................... 3-135 3.2.5 Residual Heat Removal.................................. 3-139 3.2.6 Containment Atmosphere Control and Dilution System.......... 3-146 3.2.7 Standby Gas Treatment System........................... 3-148 3.2.8 Secondary Containment System........................... 3-151 3.3 Aging Management of Auxiliary Systems............................ 3-153 3.3.0 General.............................................. 3-153 3.3.1 Fuel Handling System................................... 3-157 3.3.2 Fuel Pool Cooling and Cleanup System...................... 3-158 3.3.3 Control Rod Drive System................................ 3-160 3.3.4 Standby Liquid Control System............................ 3-163 3.3.5 High-Pressure Service Water System....................... 3-166 3.3.6 Emergency Service Water System.......................... 3-169 3.3.7 Fire Protection System................................... 3-172 3.3.8 Control Room Ventilation System........................... 3-175 3.3.9 Battery and Emergency Switchgear Ventilation System.......... 3-178 3.3.10 Diesel Generator Building Ventilation System................ 3-180 3.3.11 Pump Structure Ventilation System........................ 3-182 3.3.12 Safety-Grade Instrument Gas System...................... 3-184 3.3.13 Backup Instrument Nitrogen to the Automatic Depressurization System
.................................................... 3-185 3.3.14 Emergency Cooling Water System
........................ 3-187 3.3.15 Condensate Storage System............................. 3-190 3.3.16 Emergency Diesel Generator............................. 3-192 3.3.17 Suppression Pool Temperature Monitoring System............ 3-199 3.3.18 Cranes and Hoists..................................... 3-200 3.4 Aging Management of Steam and Power Conversion Systems........... 3-204 3.4.1 Main Steam System..................................... 3-205 3.4.2 Main Condenser........................................ 3-208 3.4.3 Feedwater System...................................... 3-209 3.5 Aging Management of Structures and Component Supports............. 3-211 3.5.1 Containment Structure................................... 3-211 3.5.2 Reactor Building Structure................................ 3-217 3.5.3 Other Structures........................................ 3-223 3.5.4 Component Supports.................................... 3-227 3.5.5 Hazard Barriers and Elastomers........................... 3-231 3.5.6 Miscellaneous Steel..................................... 3-234 3.5.7 Electrical and Instrumentation Enclosures and Raceways........ 3-236 3.5.8 Insulation............................................. 3-238 3.6 Aging Management of Electrical and Instrumentation and Controls........ 3-239 3.6.1 Cables............................................... 3-240 3.6.2 Connectors, Splices, and Terminal Blocks.................... 3-259 3.6.3 Station Blackout System................................. 3-265 4 TIME-LIMITED AGING ANALYSES......................................... 4-1 4.1 Identification of Time-Limited Aging Analyses
.......................... 4-1 4.1.1 Introduction............................................. 4-1 4.1.2 Summary of Technical Information in the Application.............. 4-1
-vii-4.1.3 Staff Evaluation.......................................... 4-3 4.1.4 Conclusion.............................................. 4-5 4.2 Reactor Vessel Neutron Embrittlement................................ 4-6 4.2.1 10 CFR Part 50 Appendix G Reactor Vessel Rapid Failure Propagation and Brittle Fracture Considerations: Charpy Upper Shelf Energy (USE)
Reduction and RTNDT Increase, Reflood thermal shock analysis..... 4-6 4.2.2 Reactor Vessel Thermal Analyses: Operating Pressure-Temperature Limit (P-T Limit) Curves...................................... 4-10 4.2.3 Reactor Vessel Circumferential Weld Examination Relief......... 4-12 4.2.4 Reactor Vessel Axial Weld Failure Probability.................. 4-14 4.3 Metal Fatigue.................................................. 4-16 4.3.1 Summary of Technical Information in the Application............. 4-17 4.3.2 Staff Evaluation......................................... 4-18 4.3.3 Conclusions............................................ 4-25 4.4 Environmental Qualification....................................... 4-25 4.4.1 Electrical Equipment Environmental Qualification Analyses........ 4-26 4.4.2 GSI-168, Environmental Qualification of Low Voltage Instrumentation and Control (I&C) Cables.................................... 4-31 4.5 Reactor Vessel Internals Fatigue and Embrittlement.................... 4-32 4.5.1 Summary of Technical Information in the Application............. 4-32 4.5.2 Staff Evaluation......................................... 4-33 4.5.3 Conclusions............................................ 4-34 4.6 Containment Fatigue............................................ 4-35 4.6.1 Fatigue Analysis of Containment Pressure Boundaries: Analysis of Tori, Torus Vents, and Torus Penetrations........................ 4-35 4.6.2 Fatigue Analysis of SRV Discharge Lines and External Torus-Attached Piping................................................ 4-36 4.6.3 Expansion Joints and Bellows Fatigue Analyses: Drywell-to-Torus Vent Bellows............................................... 4-37 4.6.4 Expansion Joint and Bellows Fatigue Analyses: Containment Process Penetration Bellows..................................... 4-37 4.7 Other Plant-Specific TLAAs....................................... 4-38 4.7.1 Reactor Vessel Main Steam Nozzle Cladding Removal Corrosion Allowance............................................. 4-38 4.7.2 Generic Letter 81-11 Crack Growth Analysis to Demonstrate Conformance to the Intent of NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking......................... 4-39 4.7.3 Fracture Mechanics of ISI-Reportable Indications for Group 1 Piping: As-forged Laminar Tear in a Unit 3 Main Steam Elbow Near Weld 1-B-3BC-LDO Discovered During Preservice UT...................... 4-42 5 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
......... 5-1 6 CONCLUSIONS........................................................ 6-1 APPENDIX A, CHRONOLOGY.............................................. A - 1 APPENDIX B, REFERENCES.............................................. B - 1
-viii-APPENDIX C, PRINCIPAL CONTRIBUTORS................................. C - 1 APPENDIX D, COMMITMENT LISTING...................................... D - 1
-ix-ACRONYMS AAC alternate ac AASHTO American Association of State Highway and Transportation Official ACI American Concrete Institute ACSR aluminum conductor steel reinforced ADS automatic depressurization system AMP aging management program AMR aging management review ANL Argonne National Laboratory AO abnormal occurrence APCSB Auxiliary and Power Conversion Systems Branch ARI alternate rod insertion ART anticipatory reactor trip ASCO American Switch Company ASME American Society of Mechanical Engineers ATWS anticipated transient without scram BESVS Battery and Emergency Switchgear Ventilation Systems BPT Branch Technical Position BWR boiling water reactor BWROG boiling water reactor owners group BWRVIP Boiling Water Reactor Vessel and Internals Project CAC containment atmosphere control (system)
CAD containment atmospheric dilution (system)
CASS cast austenitic stainless steel CCW closed cooling water CDF core damage frequency CFR Code of Federal Regulations CLB current licensing basis CRD control rod drive CRDHS control rod drive housing supports CRL component record list CRVS control room ventilation system CST condensate storage tank CUF cumulative usage factor DBA design-basis accidents DBD design baseline document DBE design basis event DGBVS diesel generator building ventilation system DRF dose reduction factor ECCS emergency core cooling system ECP electrochemical potential ECT emergency cooling tower ECW emergency cooling water (system)
EDG emergency diesel generator EFPY effective full-power years EPDM ethylene propylene dienyl monomer EPRI Electric Power Research Institute
-x-EQ environmental qualification ESF engineered safety feature ESW emergency service water (system)
FAC flow-accelerated corrosion FERC Federal Energy Regulatory Commission FMP fatigue monitoring program FPP fire protection program FSAR final safety analysis report FSSD fire safe shutdown GDC general design criteria GL generic letter GSI Generic Safety Issues HEDL Hanford Engineering and Development Laboratory HELB high-energy line break HEPA high-efficiency particulate air HPCI high-pressure coolant injection (system)
HPSW high-pressure service water (system)
HVAC heating, ventilation, and air conditioning HWC hydrogen water chemistry HX heat exchanges I & C instrumentation and controls IASCC irradiation assisted stress corrosion cracking ICEA Insulated Cable Engineers Association ICM Instrument Control Monitor IGSCC intergranular stress corrosion cracking ILRT integrated leak rate test IN information notice INPO Institute of Nuclear Power Operations IPA integrated plant assessment IPE individual plant evaluation IPEEE individual plant examination of external events ISI inservice inspection IST inservice testing LEFM linear elastic fracture mechanics LER licensee event report LLRT local leak rate tests LMFBR Liquid Metal Fast Breeder Reactor LOCA loss of coolant accident LPCI low-pressure coolant injection (system)
LPRM local power range monitor LRA license renewal application LRC level recorder controller LWR light-water reactor MCC motor control center MCRE main control room envelope MCRE main control room envelope MIC microbiologically influenced corrosion MOV motor-operated valve
-xi-MR maintenance rule MSIV main steam isolation valve MSRV main steam relief valve NCR nonconformance report NDE nondestructive examination NEI Nuclear Energy Institute NEMA National Electrical Manufactures Associates NFPA National Fire Protection Association NMCA noble metals chemical addition NPAR nuclear plant aging research NRC Nuclear Regulatory Commission NSAC Nuclear Safety Analysis Center NSR non-safety related NSSS nuclear steam supply systems NSW normal service water NUMARC Nuclear Management and Resources Council OE operating experience OFS orificed fuel support ORNL Oak Ridge National Laboratory P&ID piping and instrumentation diagram PBAPS Peach Bottom Atomic Power Station PCIS primary containment isolation system PECO Philadelphia Electric Company PLI project level instruction PM preventive maintenance P-T pressure-temperature PSVS pump structure ventilation system PUA plant-unique analysis PWR pipe whip restraint QAP quality assurance procedure RAI request for additional information RBM rod block monitor RCIC reactor core isolation cooling (system)
RCS reactor coolant system RG Regulatory Guide RHR residual heat removal (system)
RMS radiation monitoring system RPS reactor protection system RPV reactor pressure vessel RRS reactor recirculation system RTNDT nil-ductility transition reference temperature RVID reactor vessel integrity database RWM rod worth minimizer RWCU reactor water cleanup RWST refueling water storage tank SBLC standby liquid control (system)
SBO station blackout SCC stress corrosion cracking
-xii-SE safety evaluation SECY Secretary of the Commission Office of the (NRC)
SER safety evaluation report SGIG safety grade instrument gas (system)
SGTS standby gas treatment system SIL Service Information Letter SLC standby liquid control SOER significant operating experience reports SPOTMOS suppression pool temperature monitoring system SRM source range monitor SRP-LR Standard Review Plan - license renewal SRV safety relief valve SCs structures and components SSCs systems, structures, and components SV safety valve SSWP Susquehanna Substation Wooden Pole TID total integrated dose TLAAs time-limited aging analyses TTA thenyltrifluoroacetone UFSAR updated final safety analysis report UL Underwriters Laboratories, Inc.
USAS United States of America Standards USE upper-shelf energy USI unresolved safety issue WRNM wide range neutron monitor XLPE cross-linked polyethylene XLPO cross-linked polyolefin
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1-1 1 INTRODUCTION AND GENERAL DISCUSSION 1.1 Introduction This document is a safety evaluation report (SER) on the application to renew the operating licenses for Peach Bottom Atomic Power Station, Units 2 and 3, filed by Exelon Generation Company, LLC, (Exelon) (hereafter referred to as Exelon or the applicant).
By letter dated July 2, 2001, Exelon submitted its application to the U.S. Nuclear Regulatory Commission (NRC) for renewal of the operating licenses for Peach Bottom Atomic Power Station, Units 2 and 3, for an additional 20 years. The NRC staff reviewed the Peach Bottom license renewal application (LRA) for compliance with the requirements of Title 10 of the Code of Federal Regulations, Part 54 (10 CFR Part 54), Requirements for Renewal of Operating Licenses for Nuclear Power Plants, and prepared this report to document its findings. The NRCs license renewal project manager for Peach Bottom Atomic Power Station, Units 2 and 3, is David Solorio. Mr. Solorio may be contacted by calling 301-415-1973 or by writing to the License Renewal and Environmental Impacts Program, U.S. Nuclear Regulatory Commission, Washington, DC 20555-001.
In its application, Exelon requested renewal of the operating licenses issued under Section 104b of the Atomic Energy Act of 1954, as amended, for Peach Bottom Atomic Power Station, Units 2 and 3 (License Nos. DPR-44 and DPR-56, respectively) for a period of 20 years beyond the current license expiration dates of August 8, 2013 and July 2, 2014, respectively. The Peach Bottom Atomic Power Station is a two-unit boiling water reactor located in York County and Lancaster County in southeastern Pennsylvania. Each unit consists of a General Electric boiling-water reactor nuclear steam supply system designed to generate 3458 megawatts thermal or 1093 megawatts electric. Details concerning the plant and the site are found in the updated final safety analysis report (UFSAR) for each unit.
The license renewal process proceeds along two tracks: a technical review of safety issues and an environmental review. The requirements for these two reviews are stated in NRC regulations 10 CFR Parts 54 and 51, respectively. The safety review is based on Exelons application for license renewal and on the applicants answers to requests for additional information (RAIs) from the NRC staff. Exelon has also supplemented its answers to the RAIs in meetings and docketed correspondence. The public can review the LRA and all pertinent information and material, including the UFSARs, at the NRC Public Document Room, 11555 Rockville Pike, Rockville, MD 20852-2738. In addition, the Peach Bottom Atomic Power Station, Units 2 and 3, LRA and significant information and material related to the license renewal review are available on the NRCs Website at www.nrc.gov through the NRCs electronic reading room.
This SER summarizes the findings of the staffs safety review of the Peach Bottom Atomic Power Station, Units 2 and 3, and describes the technical details considered in evaluating the safety aspects of its proposed operation for an additional 20 years beyond the term of the current operating licenses. The staff reviewed the LRA in accordance with the NRC regulations and the guidance presented in the NRC Standard Review Plan (SRP) for the Review of License Renewal Applications for Nuclear Power Plants, dated July 2001.
1-2 1.2 License Renewal Background Pursuant to the Atomic Energy Act of 1954, as amended, and NRC regulations, operating licenses for commercial power reactors are issued for 40 years. These licenses can be renewed for up to 20 additional years. The original 40-year license term was selected on the basis of economic and antitrust considerations, not technical limitations. However, some individual plant and equipment designs may have been engineered on the basis of an expected 40-year service life.
In 1982, the NRC anticipated interest in license renewal and held a workshop on nuclear power plant aging. That led the NRC to establish a comprehensive program plan for nuclear plant aging research (NPAR). On the basis of the results of that research, a technical review group concluded that many aging phenomena are readily manageable and do not involve technical issues that would preclude extending the life of nuclear power plants.
In 1986, the NRC published a request for comment on a policy statement that would address major policy, technical, and procedural issues related to life extension for nuclear power plants.
In 1991, the NRC published the license renewal rule in 10 CFR Part 54. The NRC participated in an industry-sponsored demonstration program to apply the rule to pilot plants and develop experience to establish implementation guidance. To establish a scope of review for license renewal, the rule defined age-related degradation unique to license renewal. However, during the demonstration program, the NRC found that many aging mechanisms occur and are managed during the period of the initial license. In addition, the NRC found that the scope of the review did not allow sufficient credit for existing programs, particularly for the implementation of the Maintenance Rule, which also manages plant aging phenomena.
As a result, in 1995 the NRC amended the license renewal rule in 10 CFR Part 54. The amended rule established a regulatory process that is simpler, more stable, and more predictable than the previous license renewal rule. In particular, 10 CFR Part 54 was clarified to focus on managing the adverse effects of aging rather than on identifying all aging mechanisms. The rule changes were intended to ensure that important systems, structures, and components (SSCs) will continue to perform their intended function in the period of extended operation. In addition, the integrated plant assessment (IPA) process was clarified and simplified to be consistent with the revised focus on passive, long-lived structures and components (SCs).
In parallel with these efforts, the NRC pursued a separate rulemaking effort to amend 10 CFR Part 51 to focus the scope of the review of environmental impacts of license renewal, and fulfill, in part, the NRC's responsibilities under the National Environmental Policy Act of 1969 (NEPA).
1.2.1 Safety Review License renewal requirements for power reactors are based on two key principles:
(1)
The regulatory process is adequate to ensure that the licensing basis of all currently operating plants maintains an acceptable level of safety, with the possible exception is the detrimental effects of aging on the functionality of certain SSCs during the period of
1-3 extended operation, and a few other safety issues may arise only during the period of extended operation (1)
The plant-specific licensing basis must be maintained during the renewal term in the same manner and to the same extent as during the original licensing term.
In implementing these two principles 10 CFR 54.4 defines the scope of license renewal as including those plant SSCs (a) that are safety-related, (b) whose failure could affect safety-related functions, (c) that are relied on to demonstrate compliance with the Commission's regulations for fire protection, environmental qualification, pressurized thermal shock, anticipated transients without scram, and station blackout.
Pursuant to 10 CFR 54.21(a)(1), the applicant must review all SSCs that are within the scope of the rule to identify SCs that are subject to an aging management review (AMR). SCs that are subject to an AMR are those that perform an intended function without moving parts or without a change in configuration or properties and that are not subject to replacement based on a qualified life or specified time period. As required by 10 CFR 54.21(a), the applicant must demonstrate that the effects of aging will be managed in such a way that the intended function or functions of the SCs that are within the scope of license renewal will be maintained, consistent with the current licensing basis, for the period of extended operation.
Active equipment, however, is considered to be adequately monitored and maintained by existing programs. The detrimental effects of aging on active equipment are more readily detectable and will be identified and corrected through routine surveillance, performance indicators, and maintenance. The surveillance and maintenance programs and activities for active equipment, as well as other aspects of maintaining the plant design and licensing basis, are required to continue throughout the period of extended operation.
Pursuant to 10 CFR 54.21(b), each year following submittal the LRA and at least 3 months before the scheduled completion of the NRC review, an amendment to the renewal application must be submitted that identifies any change to the CLB of the facility that materially affects the contents of the LRA, including the FSAR supplement.
Another requirement for license renewal is the identification and updating of time-limited aging analyses. During the design phase for a plant, certain assumptions are made about the initial operating term of the plant, and these assumptions are incorporated into design calculations for several of the plants SSCs. In accordance with 10 CFR 54.21(c)(1), these calculations must be shown to be valid for the period of extended operation or must be projected to the end of the period of extended operation, or the applicant must demonstrate that the effects of aging on these SSCs will be adequately managed for the period of extended operation. Pursuant to 10 CFR 54.21(c)(2), each applicant must provide a list of the exemptions granted pursuant to 10 CFR 50.12 and still in effect that are based on the TLAAs as defined in 10 CFR 54.3.
Pursuant to CFR 54.21(c)(2), each applicant must also provide an evaluation that justifies the continuation of these exemptions for the period of extended operation.
Pursuant to 10 CFR 54.21(d), each application is required to include a supplement to the FSAR. This supplement must contain a summary description of the programs and activities for managing the effects of aging, and the evaluation of TLAAs for the period of extended operation.
1-4 In July 2001, the NRC issued Regulatory Guide 1.188, Standard Format and Content for Applications to Renew Nuclear Power Plant Operating License; NUREG-1800, Standard Review Plan for the Review of License Renewal Application for Nuclear Power Plants (SRP-LR); and NUREG-1801, Generic Aging Lessons Learned (GALL) Report. These documents describe methods acceptable to the NRC staff for implementing the license renewal rule, as well as techniques used by the NRC staff in evaluating applications for license renewals. The draft versions of these documents were issued for public comment on August 31, 2000 (64 FR 53047). The staff assessment of public comments was issued as NUREG-1739, Analysis of Public Comments on the improved License Renewal Guidance Documents. The regulatory guide endorsed an implementation guideline prepared by the Nuclear Energy Institute (NEI) as an acceptable method of implementing the license renewal rule. The NEI guideline is NEI 95-10, "Industry Guideline for Implementing the Requirements of 10 CFR Part 54-The License Renewal Rule," Revision 3 issued in April 2001. The staff used the RG1.188, along with the SRP, to review this application and to assess topical reports on license renewal issues as submitted by industry groups.
1.2.2 Environmental Review In December 1996, the staff revised the environmental protection regulations in 10 CFR Part 51 to facilitate environmental reviews for license renewal. The staff prepared a Generic Environmental Impact Statement (GEIS) for License Renewal of Nuclear Plants (NUREG-1437) to document its evaluation of the possible environmental impacts associated with renewing licenses of nuclear power plants. For certain types of environmental impacts, the GEIS establishes generic findings that are applicable to all nuclear power plants. These generic findings are identified as Category 1 issues in 10 CFR Part 51, Subpart A, Appendix B.
Pursuant to 10 CFR 51.53(c)(3)(i), an applicant for license renewal may incorporate these generic findings in its environmental report. Analyses of environmental impacts of license renewal that must be evaluated on a plant-specific basis are identified as Category 2 issues in 10 CFR Part 51, Subpart A, Appendix B. Such analyses must be included in an environmental report in accordance with 10 CFR 51.53(c)(3)(ii).
In accordance with NEPA and the requirements of 10 CFR Part 51, the NRC performs a plant-specific review of the environmental impacts of license renewal, including whether there is new and significant information not considered in the GEIS. Two public meetings were held near the Peach Bottom site on November 7, 2001, as part of the NRC's scoping process to identify environmental issues specific to the plant. The results of the environmental review and a preliminary recommendation on the license renewal action were documented in NRC draft plant-specific Supplement 10 to the GEIS, dated June 2002. Two additional public meetings have been conducted near the site on July 31, 2002 (during the 75-day comment period for draft plant-specific Supplement 10 to the GEIS). At the meetings, the staff described the environmental review and answered questions from members of the public to help them formulate their comments on the review. The Final Supplement 10 to the GEIS was issued on January 22, 2003.
The Final Supplement 10 to the GEIS presents the NRCs environmental analysis of the effects of renewing the Peach Bottom Units 2 and 3 operating licenses for up to an additional 20 years.
The analysis considers and weighs the environmental effects and alternatives that are available
1-5 to avoid adverse environmental effects. On the basis of the analyses and findings in the GEIS, the environmental report submitted by the applicant, consultation with other Federal, State, and local agencies, its own independent review, and its consideration of public comments, the staff recommended in Supplement 10 that the Commission determine that the adverse environmental impacts of license renewal for Peach Bottom Units 2 and 3 are not so great that preserving the option of license renewal for energy planning decision-making would be unreasonable.
1.3 Summary of the Principal Review Matters The requirements for renewing operating licenses for nuclear power plants are described in 10 CFR Part 54. The staff performed its technical review of the Peach Bottom Atomic Power Station, Units 2 and 3, license renewal application in accordance with Commission guidance and the requirements of 10 CFR Part 54. The standards for renewing a license are contained in 10 CFR 54.29.
In 10 CFR 54.19(a), the Commission requires a license renewal applicant to submit general information. Exelon submitted this general information in an enclosure to its July 2, 2001, application for renewed operating licenses for Peach Bottom Atomic Power Station, Units 2 and
- 3. The applicant supplemented this information in a letter dated August 23, 2001. The staff reviewed the enclosure and the supplemental information.
In 10 CFR 54.19(b), the Commission requires that LRAs include conforming changes to the standard indemnity agreement, 10 CFR 140.92, Appendix B, to account for the expiration term of the proposed renewed license. The applicant stated the following in its renewal application regarding this issue:
The current indemnity agreement for Peach Bottom Atomic Power Station, Units 2 and 3 states in Article VII that the agreement shall terminate at the time of expiration of the license specified in Item 3 of the Attachment to the agreement.
Item 3 of the Attachment to the indemnity agreement, lists two license numbers, DRP-44 and DRP-56. Should the license numbers be changed upon issuance of the renewed licenses, Exelon requests that the conforming changes be made to Article VII and Item 3 of the Attachment, and to any other sections of the indemnity agreement as appropriate.
The staff will use the original license number for the renewed license. Therefore, there is no need to make conforming changes to the indemnity agreement, and the requirements of 10 CFR 54.19(b) have been met.
In 10 CFR 54.21, the Commission requires that each application for a renewed license for a nuclear facility must contain (a) an integrated plant assessment (IPA), (b) description of current licensing basis changes made during the NRC review of the application, (c) an evaluation of time-limited aging analyses (TLAAs), and (d) a final safety analysis report (FSAR) supplement.
On July 2, 2001, the applicant submitted the information required by 10 CFR 54.21(a) and (c) in the Enclosure of its LRA.
In 10 CFR 54.22, the Commission states requirements regarding technical specifications. The applicant did not request any changes to the plant technical specification in its LRA.
1-6 The staff evaluated the technical information required by 10 CFR 54.21 and 54.22 in accordance with the NRC's regulations and the guidance provided in the SRP. The staff's evaluation of this information is documented in Chapters 2, 3, and 4 of this SER.
The staff's evaluation of the environmental information required by 10 CFR 54.23 is documented in the draft plant-specific supplement to the GEIS (NUREG-1437, Supplement 10), which states the considerations related to renewing the licenses for Peach Bottom Atomic Power Station, Units 2 and 3.
1.3.1 Boiling Water Reactor Vessel Internals Project (BWRVIP) Topical Reports In accordance with 10 CFR 54.17(e), Exelon also incorporated by reference several BWRVIP topical reports into the Peach Bottom LRA. The purpose of the topical reports is to generically demonstrate that the aging effects for reactor coolant system components are adequately managed for the period of extended operation under a renewed license. Exelon incorporated the following BWRVIP topical reports into its application:
BWRVIP-05, BWR RPV Shell Weld Inspection Recommendations, September 1995 BWRVIP-18, Core Spray Internals Inspection and Flaw Evaluation Guidelines, July 1996 BWRVIP-25, BWR Core Plate Inspection and Flaw Evaluation Guidelines, October 1999 BWRVIP-26, Top Guide Inspection and Flaw Evaluation Guidelines, December 1996 BWRVIP-27, Standby Liquid Control System/Core Plate P Inspection and Flaw Evaluation Guidelines, April 1997 BWRVIP-38, Shroud Support Inspection and Flaw Evaluation Guidelines, September 1997 BWRVIP-41, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines, October 1997 BWRVIP-47, BWR Lower Plenum Inspection and Flaw Evaluation Guidelines, December 1997 BWRVIP-48, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines, March 1998 BWRVIP-49, Instrument Penetration Inspection and Flaw Evaluation Guidelines, March 1998 BWRVIP-74, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines, September 1999
1-7 BWRVIP-75, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (NUREG-0313), October 1999 BWRVIP-76, BWR Core Shroud Inspection and Flaw Evaluation Guidelines, December 1999 All the BWRVIP reports listed above have been approved by the staff with the exception of BWRVIP-76. The staff is presently reviewing the responses from the Owners Group, and is expected to issue a safety evaluation report by the end of 2003. Because the staffs review is not complete the license will be conditioned as discussed below in Section 1.6.
The applicant committed to follow the BWRVIP reports as approved by the staff. The staff finds this commitment to be acceptable for aging management of the systems and components addressed in the subject BWRVIP reports.
1.4 Summary of Open Items As a result of its review of the license renewal application for the Peach Bottom Atomic Power Station Units 2 & 3, including the additional information submitted to the NRC through May 22, 2002, the staff identified 15 issues that remained open at the time this report was published previously as an SER with Open Items on September 16, 2002. An issue was considered open if Exelon had not presented a sufficient basis for its resolution. Each Open Item was assigned a unique identifying number, which identified the section in this report in which the Open Item was described. For example, Open Item 3.0-1 was discussed in Section 3.0 of this report. By letters dated November 26 and December 19, 2002, January 14, and January 29, 2003, the applicant responded to these Open items. The staff reviewed the responses and has closed all of the Open Items. The base for closing the Open Items can be found in the following Sections:
2.3.2.7.2, 2.3.2.7.2, 2.3.3.8.2, 2.3.3.8.2, 2.3.3.9.2, 2.3.3.18.2, 2.3.3.19.2, 2.4.7.2, 3.0.3.6.2, 3.0.3.11.2, 3.0.3.16.2, 3.1.3.2.1, 3.6.1.2.1, 3.6.1.2.2, and 4.5.2.
1.5 Summary of Confirmatory Items As a result of the staffs review of Exelons application for license renewal, including the additional information and clarifications submitted subsequently, the staff identified the confirmatory items listed below, as of the time this report was published previously as an SER with Open Items on September 16, 2002. Confirmatory Items were those for which Exelon had not yet provided adequate documentation. In addition, confirmatory items may include significant matters that need to be considered as possible license conditions or technical specification requirements, depending on the form of the resolution. Each Confirmatory Item was assigned a unique identifying number, which identified the section in this report in which the Confirmatory Item was described. For example Confirmatory Item 3.0-1 was discussed in Section 3.0 of this report. By letters dated November 26 and December 19, 2002, January 14, and January 29, 2003, the applicant responded to these Confirmatory Items. The staff reviewed the responses and has closed all the Confirmatory Items. The base for closing the Confirmatory Items can be found in the following Sections: 3.0.3.3.2, 3.0.3.11.2, 3.0.3.13.2, 3.0.3.14.3, 3.0.3.17.2, 3.0.3.19.2, 3.0.3.20.3, 3.0.4, 3.2.1.2.2, 3.6.1.2.2, 3.6.2.2.2, 4.1.2, 4.1.3, 4.1.3, 4.2.1.2, 4.2.3.2, 4.2.4.2 and 4.3.2.
1-8 1.6 Summary of Proposed License Conditions As a result of the staffs review of Exelons application for license renewal, including the additional information and clarifications submitted subsequently, the staff identified 4 license conditions. The first license condition requires the applicant to include the UFSAR Supplement in the next UFSAR update required by 10 CFR 50.71 (e). The second license condition requires that, prior to operation in the renewal term, the applicant will notify the NRC of its decision to implement either the staff-approved reactor vessel integrated surveillance program, or a plant-specific program, and provide the appropriate revision to the UFSAR Supplement summary descriptions of the program. The third license condition requires that the future inspection activities identified in the UFSAR Supplement be completed before the beginning of the extended period of operation. The fourth license condition requires that, prior to operation in the renewal term, the applicant will notify the NRC of its decision to implement either the staff-approved core shroud inspection and evaluation guidelines program, or a plant specific program, and provide the appropriate revision to the UFSAR supplement summary description of the program.
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