Related to the License Renewal of Fermi 2 · 2016-07-14 · Related to the License Renewal of Fermi...

Safety Evaluation Report Related to the License Renewal of Fermi 2

Docket No. 50-341

DTE Electric Company

United States Nuclear Regulatory Commission

Office of Nuclear Reactor Regulation

July 2016

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ABSTRACT

This safety evaluation report (SER) documents the technical review of the Fermi 2 Nuclear Power Plant (Fermi 2) license renewal application (LRA) by the U.S. Nuclear Regulatory Commission (NRC) staff (the staff). By letter dated April 24, 2014, DTE Electric Company (DTE or the applicant) submitted the LRA in accordance with Title 10 of the Code of Federal Regulations Part 54, “Requirements for Renewal of Operating Licenses for Nuclear Power Plants” (10 CFR Part 54). DTE requests renewal of the Fermi 2 operating license (Operating License No. NPF-43) for a period of 20 years beyond the current expiration at midnight on March 20, 2025.

Fermi 2 is located on the western shore of Lake Erie at Lagoona Beach, Frenchtown Township, in Monroe County, Michigan. The NRC issued the operating license on March 20, 1985. Fermi 2 is a single-cycle, forced-circulation boiling water reactor (GE-BWR 4). General Electric Company (GE) furnished the nuclear steam supply system. Fermi 2’s licensed power output is 3,486 megawatts thermal with a turbine-generator net electrical output of approximately 1,170 megawatts electric.

The decommissioned Enrico Fermi Atomic Power Plant (Fermi 1) is within the Fermi 2 Owner Controlled Area. Fermi 1 (Operating License No. DPR-9) was a sodium-cooled fast breeder reactor. It is permanently shut down, in SAFSTOR status. The nuclear fuel has been shipped offsite.

This SER presents the status of the staff’s review of information submitted through May 30, 2016. The one open item identified in the SER with open items, issued January 28, 2016, has been closed (see Section 1.5); therefore, no open items remain to be resolved. On the basis of its review of the LRA, the staff determines that the requirements of 10 CFR 54.29(a) have been met (see Section 5).

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TABLE OF CONTENTS

ABSTRACT .......................................................................................................................... iii TABLE OF CONTENTS ............................................................................................................ v APPENDICES ........................................................................................................................ viii LIST OF TABLES ................................................................................................................... viii ABBREVIATIONS .................................................................................................................... xi

INTRODUCTION AND GENERAL DISCUSSION ............................................ 1-1 1.1 Introduction .................................................................................................................... 1-1 1.2 License Renewal Background ........................................................................................ 1-2

1.2.1 Safety Review ..................................................................................................... 1-3 1.2.2 Environmental Review ........................................................................................ 1-4

1.3 Principal Review Matters ................................................................................................ 1-5 1.4 Interim Staff Guidance .................................................................................................... 1-6 1.5 Summary of Open Items ................................................................................................. 1-7 1.6 Summary of Confirmatory Items ..................................................................................... 1-8 1.7 Summary of Proposed License Conditions ..................................................................... 1-9

STRUCTURES AND COMPONENTS SUBJECT TO 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 2.1.3 Scoping and Screening Program Review ............................................................ 2-1 2.1.4 Plant Systems, Structures, and Components Scoping Methodology ................... 2-6 2.1.5 Screening Methodology .................................................................................... 2-19 2.1.6 Summary of Evaluation Findings ...................................................................... 2-24

2.2 Plant-Level Scoping Results ......................................................................................... 2-24 2.2.1 Introduction ....................................................................................................... 2-24 2.2.2 Summary of Technical Information in the Application ........................................ 2-24 2.2.3 Staff Evaluation ................................................................................................. 2-25 2.2.4 Conclusion ........................................................................................................ 2-25

2.3 Scoping and Screening Results: Mechanical Systems ................................................ 2-25 2.3.1 Reactor Coolant System ................................................................................... 2-26 2.3.2 Engineered Safety Features ............................................................................. 2-34 2.3.3 Auxiliary Systems ............................................................................................. 2-42 2.3.4 Steam and Power Conversion .......................................................................... 2-83

2.4 Scoping and Screening Results: Structures ................................................................. 2-92 2.4.1 Reactor/Auxiliary Building and Primary Containment ........................................ 2-93 2.4.2 Water-Cooled Structures .................................................................................. 2-97 2.4.3 Turbine Building, Process Facilities, and Yard Structures ................................. 2-99

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2.4.4 Bulk Commodities ........................................................................................... 2-101 2.4.5 Fire Barriers Portion of Bulk Commodities ...................................................... 2-102

2.5 Scoping and Screening Results: Electrical and Instrumentation and Control Systems ..................................................................................................................... 2-107 2.5.1 Electrical and Instrumentation and Control Components and Commodity

Groups ............................................................................................................ 2-108 2.6 Conclusion for Scoping and Screening ....................................................................... 2-110

AGING MANAGEMENT REVIEW RESULTS .................................................. 3-1 3.0 Applicant’s Use of the Generic Aging Lessons Learned Report ...................................... 3-1

3.0.1 Format of the License Renewal Application ........................................................ 3-2 3.0.2 Staff’s Review Process ....................................................................................... 3-3 3.0.3 Aging Management Programs ............................................................................ 3-7 3.0.4 QA Program Attributes Integral to Aging Management Programs ................... 3-192 3.0.5 Operating Experience for Aging Management Programs ................................ 3-195

3.1 Aging Management of Reactor Vessel, Internals, and Reactor Coolant System ......... 3-200 3.1.1 Summary of Technical Information in the Application ...................................... 3-200 3.1.2 Staff Evaluation ............................................................................................... 3-201 3.1.3 Conclusion ...................................................................................................... 3-247

3.2 Aging Management of Engineered Safety Features Systems ..................................... 3-247 3.2.1 Summary of Technical Information in the Application ...................................... 3-247 3.2.2 Staff Evaluation ............................................................................................... 3-247 3.2.3 Conclusion ...................................................................................................... 3-277

3.3 Aging Management of Auxiliary Systems ................................................................... 3-277 3.3.1 Summary of Technical Information in the Application ...................................... 3-277 3.3.2 Staff Evaluation ............................................................................................... 3-277 3.3.3 Conclusion ...................................................................................................... 3-352

3.4 Aging Management of Steam and Power Conversion Systems .................................. 3-352 3.4.1 Summary of Technical Information in the Application ...................................... 3-353 3.4.2 Staff Evaluation ............................................................................................... 3-353 3.4.3 Conclusion ...................................................................................................... 3-379

3.5 Aging Management of Structures and Component Supports ...................................... 3-379 3.5.1 Summary of Technical Information in the Application ...................................... 3-380 3.5.2 Staff Evaluation ............................................................................................... 3-380 3.5.3 Conclusion ...................................................................................................... 3-436

3.6 Aging Management of Electrical and Instrumentation and Controls System ............... 3-436 3.6.1 Summary of Technical Information in the Application ...................................... 3-437 3.6.2 Staff Evaluation ............................................................................................... 3-437 3.6.3 Conclusion ...................................................................................................... 3-452

3.7 Conclusion for Aging Management Review Results ................................................... 3-452 TIME-LIMITED AGING ANALYSES ................................................................ 4-1

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

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4.1.1 Summary of Technical Information in the Application .......................................... 4-2 4.1.2 Staff Evaluation ................................................................................................... 4-3 4.1.3 Conclusion ........................................................................................................ 4-20

4.2 Reactor Vessel Neutron Embrittlement ......................................................................... 4-20 4.2.1 Reactor Vessel Fluence .................................................................................... 4-20 4.2.2 Adjusted Reference Temperatures (ARTs) ....................................................... 4-22 4.2.3 Pressure-Temperature Limits ............................................................................ 4-27 4.2.4 Upper-Shelf Energy .......................................................................................... 4-29 4.2.5 Reactor Vessel Circumferential Weld Inspection Relief ..................................... 4-34 4.2.6 Reactor Vessel Axial Weld Failure Probability .................................................. 4-39 4.2.7 Reactor Pressure Vessel Core Reflood Thermal Shock Analysis ...................... 4-42

4.3 Metal Fatigue ............................................................................................................... 4-45 4.3.1 Class 1 Fatigue Analyses ................................................................................. 4-45 4.3.2 Non-Class 1 Fatigue Analyses .......................................................................... 4-56 4.3.3 Effects of Reactor Water Environment on Fatigue Life ...................................... 4-60

4.4 Environmental Qualification (EQ) Analyses of Electric Equipment ................................ 4-70 4.4.1 Summary of Technical Information in the Application ........................................ 4-70 4.4.2 Staff Evaluation ................................................................................................. 4-71 4.4.3 UFSAR Supplement.......................................................................................... 4-72 4.4.4 Conclusion ........................................................................................................ 4-72

4.5 Concrete Containment Tendon Prestress Analyses ..................................................... 4-72 4.5.1 Summary of Technical Information in the Application ........................................ 4-72 4.5.2 Staff Evaluation ................................................................................................. 4-73 4.5.3 UFSAR Supplement.......................................................................................... 4-73 4.5.4 Conclusion ........................................................................................................ 4-73

4.6 Containment Liner Plate, Metal Containment, and Penetrations Fatigue Analyses ....... 4-73 4.6.1 Primary Containment ........................................................................................ 4-73 4.6.2 Vent Line Bellows ............................................................................................. 4-78 4.6.3 Refueling and Drywell Seal Bellows .................................................................. 4-79 4.6.4 Traversing Incore Probe Penetration Bellows ................................................... 4-81 4.6.5 Containment Penetrations ................................................................................ 4-82

4.7 Other Plant-Specific TLAAs .......................................................................................... 4-86 4.7.1 Erosion of the Main Steam Line Flow Restrictors .............................................. 4-86 4.7.2 Determination of High-Energy Line Break Locations ......................................... 4-88 4.7.3 Jet Pump Auxiliary Spring Wedge Assembly .................................................... 4-90 4.7.4 Jet Pump Slip Joint Repair Clamps ................................................................... 4-93 4.7.5 Flaw Evaluations for the Reactor Pressure Vessel ............................................ 4-95 4.7.6 Main Steam Bypass Lines Cumulative Operating Time ..................................... 4-97 4.7.7 Crane (Heavy Load) Cycles .............................................................................. 4-99

4.8 Conclusion for TLAAs ................................................................................................. 4-100

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REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS ................................................................................................ 5-1 CONCLUSION ................................................................................................. 6-1

APPENDICES

FERMI 2 LICENSE RENEWAL COMMITMENTS ............................................ A-1 CHRONOLOGY ............................................................................................... B-1 PRINCIPAL CONTRIBUTORS ........................................................................ C-1 REFERENCES ................................................................................................ D-1

LIST OF TABLES

Table 1.4-1 Current Interim Staff Guidance ......................................................................... 1-6 Table 2.3-1 Service Water System Piping Continuations Not within the Scope of

License Renewal............................................................................................ 2-48 Table 2.3-2 Emergency Diesel Generator System Piping Continuations Not within

the Scope of License Renewal ....................................................................... 2-66 Table 2.3-3 Plant Drains System Piping Continuations Not within the Scope of

License Renewal............................................................................................ 2-71 Table 2.3-4 Plant Drains System Seismic or Equivalent Anchors Not within the

Scope of License Renewal ............................................................................. 2-71 Table 2.3-5 Primary Containment Monitoring and Leakage Detection System Piping

Continuations Not within the Scope of License Renewal ................................ 2-76 Table 2.3-6 Miscellaneous Auxiliary Systems in Scope for 10 CFR 54.4(a)(2) Piping

Continuations Not within the Scope of License Renewal ................................ 2-79 Table 2.3-7 Miscellaneous Auxiliary Systems in Scope for 10 CFR 54.4(a)(2) Piping

Continuations Not within the Scope of License Renewal ................................ 2-81 Table 2.3-8 Feedwater and Standby Feedwater System Piping Continuations Not

within the Scope of License Renewal ............................................................. 2-86 Table 2.3-9 Miscellaneous Steam and Power Conversion Systems in Scope for

10 CFR 54.4(a)(2) Piping Continuations Not within the Scope of License Renewal ......................................................................................................... 2-88

Table 2.3-10 Miscellaneous Steam and Power Conversion Systems in Scope for 10 CFR 54.4(a)(2) Piping Continuations Not within the Scope of License Renewal ......................................................................................................... 2-90



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Table 3.0-1 Fermi 2 Aging Management Programs ............................................................. 3-7 Table 3.1-1 Staff Evaluation for Reactor Vessel, Internals, and Reactor Coolant

System Components in the GALL Report .................................................... 3-202 Table 3.2-1 Staff Evaluation for Engineered Safety Features Systems Components

in the GALL Report ...................................................................................... 3-248 Table 3.3-1 Staff Evaluation for Auxiliary Systems Components in the GALL Report ...... 3-279 Table 3.4-1 Staff Evaluation for Steam and Power Conversion Systems

Components in the GALL Report ................................................................. 3-354 Table 3.5-1 Staff Evaluation for Structures and Component Supports Components

in the GALL Report ...................................................................................... 3-381 Table 3.6-1 Staff Evaluation for Electrical and Instrumentation and Controls in the

GALL Report ................................................................................................ 3-438 Table 4.6.5-1 Transients Considered for Feedwater A/B Penetrations X-9A/B ..................... 4-83 Table A.1-1 Appendix A: Fermi 2 License Renewal Commitments ..................................... A-2 Table B.1-1 Appendix B: Chronology ................................................................................. B-1 Table C.1-1 Appendix C: Principal Contributors .................................................................. C-1 Table D.1-1 References ...................................................................................................... D-1

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ABBREVIATIONS

AAI applicant action item AC alternating current ACI American Concrete Institute ACRS Advisory Committee on Reactor Safeguards ADAMS Agencywide Documents Access and Management System AERM aging effect requiring management AFW auxiliary feedwater AISC American Institute of Steel Construction AMP aging management program AMR aging management review ANSI American National Standards Institute APCSB Auxiliary Power Conversion System Branch ART adjusted reference temperature ASME American Society of Mechanical Engineers ASTM American Society for Testing and Materials ATWS anticipated transient without scram

BDBEE beyond design basis external events BTP Branch Technical Position BWR boiling water reactor

°C degrees Celsius CARD Condition Assessment Resolution Document CASS cast austenitic stainless steel CBF cycle-based fatigue CCHVAC Control Center heating, ventilation, and air conditioning CE Combustion Engineering Company CECO Central Component (database) CFR Code of Federal Regulations CLB current licensing basis CO2 carbon dioxide CRD control rod drive CRT condensate return tank CST condensate storage tank CTG combustion turbine generator Cu copper CUF cumulative usage factor CUFen environmentally assisted fatigue cumulative usage factor

DBA design basis accident DBE design basis event DC direct current DG diesel generator DO dissolved oxygen DTE DTE Electric Company

EAF environmentally assisted fatigue ECCS emergency core cooling system EDG emergency diesel generator

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EDGSW emergency diesel generator service water EECW emergency equipment cooling water EESW emergency equipment service water EFPY effective full-power year EPRI Electric Power Research Institute EPU extended power uprate EQ environmental qualification ESF engineered safety features

°F degrees Fahrenheit Fen environmentally assisted fatigue correction factor FERC Federal Energy Regulatory Commission Fermi 2 Fermi 2 Nuclear Power Plant FOST fuel oil storage tank FPCCS fuel pool cooling and cleanup system FPEE fire protection engineering evaluation FR Federal Register FSAR final safety analysis report FSER final safety evaluation report ft foot (feet) ft-lb foot-pound FW feedwater

GALL Generic Aging Lessons Learned (Report) GE General Electric Company GEH General Electric-Hitachi GEIS Generic Environmental Impact Statement GL generic letter gpm gallons per minute GSI generic safety issue GSW general service water

HELB high-energy line break HPCI high-pressure coolant injection HPSI high-pressure safety injection HVAC heating, ventilation, and air conditioning

I&C instrumentation and controls I&E inspection and evaluation I&FE inspection and flaw evaluation IASCC irradiation-assisted stress corrosion cracking ID inside diameter IER Institute of Nuclear Power Operations Event Report IGSCC intergranular stress corrosion cracking ILRT integrated leak rate testing IN information notice INPO Institute of Nuclear Power Operations IPA integrated plant assessment ISG interim staff guidance ISI inservice inspection ISP integrated surveillance program

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ksi kilopound per square inch KV or kV kilovolt

lb pound(s) LBB leak-before-break LCO limiting condition of operation(s) LLRT local leakage rate testing LOCA loss-of-coolant accident LPCI low-pressure coolant injection LRA license renewal application LR-ISG license renewal interim staff guidance LTOP low-temperature overpressure protection µg/cm2 micrograms per centimeter squared MC metal containment MDCT mechanical draft cooling tower MEB metal enclosed bus MeV million electron volt(s) MoS2 molybdenum disulfide mpy mil(s) per year MSIV main steam isolation valve MUR measurement uncertainty recapture MUR/TPO measurement uncertainty recapture/thermal power optimization mV millivolt MWe megawatts electric MWt megawatts thermal

n/cm2 neutrons per square centimeter NACE National Association of Corrosion Engineers NDE nondestructive examination NEI Nuclear Energy Institute NFPA National Fire Protection Association Ni nickel NPS nominal pipe size NRC U.S. Nuclear Regulatory Commission

O2 oxygen OBE operating-basis earthquake ODSCC outside-diameter stress corrosion cracking

PCAC primary containment atmosphere cooling PCM primary containment monitoring PCP primary containment pneumatics pH potential of hydrogen PM preventative maintenance PORV power-operated relief valve ppm parts per million psi pounds per square inch psig pound(s) per square inch gauge PSPM Periodic Surveillance and Preventive Maintenance P-T pressure-temperature

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PTLR pressure-temperature limits report PTS pressurized thermal shock PVC polyvinyl chloride PWR pressurized water reactor PWSCC primary water stress corrosion cracking

QA quality assurance

radwaste radioactive waste RAI request for additional information RBCCW reactor building closed cooling water RCIC reactor core isolation cooling RCP reactor coolant pump RCPB reactor coolant pressure boundary RCS reactor coolant system RFO refueling outage RG regulatory guide RHR residual heat removal RHRSW residual heat removal service water RPV reactor pressure vessel RTNDT reference temperature nil ductility transition RVI reactor vessel internal RWCU reactor water cleanup

SA stress allowables SBA small break accident SBO station blackout SBF stress-based fatigue SC structure and component SCC stress corrosion cracking scfh standard cubic feet per hour SE safety evaluation SER safety evaluation report SC structure and component SGTS standby gas treatment system SLC standby liquid control SLC/core ΔP standby liquid control system/core ΔP S-N stress-number S&PC steam and power conversion SRP Standard Review Plan SRP-LR Standard Review Plan for Review of License Renewal Applications for

Nuclear Power Plants SRV safety relief value SSC system, structure, and component SSE safe-shutdown earthquake

TIP traversing incore probe TLAA time-limited aging analysis TR Technical Report TRM Technical Requirements Manual TS Technical Specification(s)

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UFSAR updated final safety analysis report USE upper-shelf energy UT ultrasonic testing UUSE unirradiated upper-shelf energy UV ultraviolet

V volt(s) VAC volts alternating current VDC volts direct current

yr year

Zn zinc

1/4 T one-fourth of the way through the vessel wall measured from the internal surface of the vessel

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INTRODUCTION AND GENERAL DISCUSSION

1.1 Introduction

This document is a safety evaluation report (SER) on the license renewal application (LRA) for Fermi 2 Nuclear Power Plant (Fermi 2) as filed by DTE Electric Company (DTE, or the applicant). By letter dated April 24, 2014, DTE submitted its application to the U.S. Nuclear Regulatory Commission (NRC) for renewal of the Fermi 2 operating license for an additional 20 years. The NRC staff (the staff) prepared this report to summarize the results of its safety review of the LRA for compliance with Title 10, Part 54, “Requirements for Renewal of Operating Licenses for Nuclear Power Plants,” of the Code of Federal Regulations (10 CFR Part 54). The NRC project managers for the license renewal review are Ms. Daneira Meléndez-Colón and Ms. Lois James. Ms. Meléndez-Colón may be contacted by telephone at 301-415-3301 or by electronic mail at [email protected]. Ms. James may be contacted by telephone at 301-415-3306 or by electronic mail at [email protected]. Alternatively, written correspondence may be sent to the following address:

Division of License Renewal U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001 Attention: Daneira Meléndez-Colón, Mail Stop O11-F1

Lois James, Mail Stop O11-F1

In its April 24, 2014, submission letter, as amended, the applicant requested renewal of the operating license issued under Section 103 (Operating License No. NPF-43) of the Atomic Energy Act of 1954, as amended, for Fermi 2 for a period of 20 years beyond the current expiration at midnight on March 20, 2025. Fermi 2 is located on the western shore of Lake Erie at Lagoona Beach, Frenchtown Township, in Monroe County, Michigan. The NRC issued the operating license on March 20, 1985. Fermi 2 is a single-cycle, forced-circulation boiling water reactor (GE-BWR 4). General Electric Company (GE) furnished the nuclear steam supply system. Fermi 2’s licensed power output is 3,486 megawatts thermal with a turbine-generator net electrical output of approximately 1,170 megawatts electric. The updated final safety analysis report (UFSAR) shows details of the plant and the site.

The license renewal process consists of two concurrent reviews, a review of safety issues and an environmental review. The NRC regulations in 10 CFR Part 54 and 10 CFR Part 51, “Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions,” respectively, set forth requirements for these reviews. The safety review for the Fermi 2 license renewal is based on the applicant’s LRA and responses to the staff’s requests for additional information (RAIs). The applicant supplemented the LRA and provided clarifications through its responses to the staff’s RAIs in audits, meetings, and docketed correspondence. The staff reviewed and considered information submitted through May 30, 2016. The public may view the LRA and all pertinent information and materials, including the UFSAR, at the NRC Public Document Room located on the first floor of One White Flint North, 11555 Rockville Pike, Rockville, MD 20852-2738 (301-415-4737/800-397-4209), and at the Ellis Library and Reference Center, 3700 South Custer Road, Monroe, MI 48161. In addition, the public may find the LRA, as well as materials related to the license renewal review, on the NRC website at http://www.nrc.gov.

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This SER summarizes the results of the staff’s safety review of the LRA and describes the technical details considered in evaluating the safety aspects of the unit’s proposed operation for an additional 20 years beyond the term of the current operating license. The staff reviewed the LRA in accordance with NRC regulations and the guidance in NUREG-1800, Revision 2, “Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants” (SRP-LR), dated December 2010.

SER Sections 2 through 4 address the staff’s evaluation of license renewal issues considered during the review of the application. SER Section 5 is reserved for the report of the Advisory Committee on Reactor Safeguards (ACRS). The conclusions of this SER are in Section 6.

SER Appendix A is a table showing the applicant’s commitments for renewal of the operating license. SER Appendix B is a chronology of the principal correspondence between the staff and the applicant regarding the LRA review. SER Appendix C is a list of principal contributors to the SER and Appendix D is a bibliography of the references in support of the staff’s review.

In accordance with 10 CFR Part 51, the staff prepared a draft plant-specific supplement to NUREG-1437, “Generic Environmental Impact Statement for License Renewal of Nuclear Plants (GEIS)” (“Generic Environmental Impact Statement for License Renewal of Nuclear Plants: Supplement 56 Regarding Fermi 2 Nuclear Power Plant,” issued October 27, 2015). This supplement discusses the environmental considerations for license renewal for Fermi 2. The final, plant-specific GEIS Supplement 56 is scheduled to be issued in 2016.

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 and can be renewed for up to 20 additional years. The original 40-year license term was selected based on economic and antitrust considerations rather than on technical limitations; however, some individual plant and equipment designs may have been engineered for an expected 40-year service life.

In 1982, the staff anticipated interest in license renewal and held a workshop on nuclear power plant aging. This workshop led the NRC to establish a comprehensive program plan for nuclear plant aging research. From the results of that research, a technical review group concluded that many aging phenomena are readily manageable and pose no technical issues precluding life extension for nuclear power plants. In 1986, the staff published a request for comment on a policy statement that would address major policy, technical, and procedural issues related to license renewal for nuclear power plants.

In 1991, the staff published 10 CFR Part 54, the License Renewal Rule (Volume 56, page 64943, of the Federal Register (56 FR 64943), dated December 13, 1991). The staff participated in an industry-sponsored demonstration program to apply 10 CFR Part 54 to a pilot plant and to gain the experience necessary to develop implementation guidance. To establish a scope of review for license renewal, 10 CFR Part 54 defined age-related degradation unique to license renewal; however, during the demonstration program, the staff found that adverse aging effects on plant systems and components are managed during the period of initial license and that the scope of the review did not allow sufficient credit for management programs, particularly the implementation of 10 CFR 50.65, “Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants,” which regulates management of plant-aging phenomena. As a result of this finding, the staff amended 10 CFR Part 54 in 1995. As published May 8, 1995, in 60 FR 22461, amended 10 CFR Part 54 establishes a regulatory

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process that is simpler, more stable, and more predictable than the previous 10 CFR Part 54. In particular, as amended, 10 CFR Part 54 focuses on the management of adverse aging effects rather than on the identification of age-related degradation unique to license renewal. The staff made these rule changes to ensure that important systems, structures, and components (SSCs) will continue to perform their intended functions during the period of extended operation. In addition, the amended 10 CFR Part 54 clarifies and simplifies the integrated plant assessment process to be consistent with the revised focus on passive, long-lived structures and components (SCs).

Concurrent with these initiatives, the staff pursued a separate rulemaking effort (61 FR 28467, June 5, 1996) and amended 10 CFR Part 51 to focus the scope of the review of environmental impacts of license renewal in order to fulfill NRC responsibilities under the National Environmental Policy Act of 1969.

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 bases of all currently operating plants maintain an acceptable level of safety with the possible exceptions of the detrimental aging effects on the functions of certain SSCs, as well as a few other safety-related issues, during the period of extended operation.

(2) 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, “Scope,” defines the scope of license renewal as including those SSCs that (1) are safety-related, (2) whose failure could affect safety-related functions, or (3) are relied on to demonstrate compliance with the NRC’s regulations for fire protection, environmental qualification (EQ), pressurized thermal shock (PTS), anticipated transient without scram (ATWS), and station blackout (SBO).

In accordance with 10 CFR 54.21(a), a license renewal applicant must review all SSCs within the scope of 10 CFR Part 54 to identify SCs subject to an aging management review (AMR). Those SCs subject to an AMR perform an intended function without moving parts or without change in configuration or properties and are not subject to replacement based on a qualified life or specified time period. In accordance with 10 CFR 54.21(a), a license renewal applicant must demonstrate that the aging effects will be managed so that the intended function(s) of those SCs will be maintained consistent with the current licensing basis (CLB) for the period of extended operation. However, active equipment is considered to be adequately monitored and maintained by existing programs. In other words, detrimental aging effects that may affect active equipment can be readily identified and corrected through routine surveillance, performance monitoring, and maintenance. Surveillance and maintenance programs for active equipment, as well as other maintenance aspects of plant design and licensing basis, are required throughout the period of extended operation.

In accordance with 10 CFR 54.21(d), the LRA is required to include a UFSAR supplement with a summary description of the applicant’s programs and activities for managing the effects of aging and an evaluation of time-limited aging analyses (TLAAs) for the period of extended operation.

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License renewal also requires TLAA identification and updating. During the plant design phase, certain assumptions about the length of time the plant can operate are incorporated into design calculations for several plant SSCs. In accordance with 10 CFR 54.21(c)(1), the applicant must either show that these calculations will remain valid for the period of extended operation, project the analyses to the end of the period of extended operation, or demonstrate that the aging effects on these SSCs will be adequately managed for the period of extended operation.

In 2005, the NRC revised Regulatory Guide (RG) 1.188, “Standard Format and Content for Applications to Renew Nuclear Power Plant Operating Licenses.” This RG endorses Nuclear Energy Institute (NEI) 95-10, Revision 6, “Industry Guideline for Implementing the Requirements of 10 CFR Part 54 – The License Renewal Rule,” dated June 2005. NEI 95-10 details an acceptable method of implementing 10 CFR Part 54. The staff used the SRP-LR to review the LRA.

In the LRA, the applicant stated that it used the process defined in NUREG-1801, Revision 2, “Generic Aging Lessons Learned (GALL) Report,” dated December 2010. The GALL Report summarizes staff-approved aging management programs (AMPs) for many SCs subject to an AMR. If an applicant commits to implementing these staff-approved AMPs, the time, effort, and resources for LRA review can be greatly reduced, improving the efficiency and effectiveness of the license renewal review process. The GALL Report summarizes the aging management evaluations, programs, and activities credited for managing aging for most of the SCs used throughout the industry. The report is also a quick reference for both applicants and staff reviewers to AMPs and activities that can manage aging adequately during the period of extended operation.

1.2.2 Environmental Review

Part 51 of 10 CFR contains environmental protection regulations. In December 1996, the staff revised the environmental protection regulations to facilitate the environmental review for license renewal. The staff prepared the GEIS to document its evaluation of possible environmental impacts associated with nuclear power plant license renewals. For certain types of environmental impacts, the GEIS contains generic findings that apply to all nuclear power plants and are codified in Appendix B, “Environmental Effect of Renewing the Operating License of a Nuclear Power Plant,” to Subpart A, “National Environmental Policy Act – Regulations Implementing Section 102(2),” of 10 CFR Part 51. In accordance with 10 CFR 51.53(c)(3)(i), a license renewal applicant may incorporate these generic findings in its environmental report. In accordance with 10 CFR 51.53(c)(3)(ii), an environmental report also must include analyses of environmental impacts that must be evaluated on a plant-specific basis (i.e., Category 2 issues).

In June 2013, the NRC staff issued a final rule revising 10 CFR Part 51 to update the potential environmental impacts associated with the renewal of an operating license for a nuclear power reactor for an additional 20 years. Revision 1 to the GEIS was issued concurrently with the final rule. The revised GEIS specifically supports the revised list of environmental issues identified in the final rule. Revision 1 to the GEIS and the 2013 final rule reflect lessons learned and knowledge gained during previous license renewal environmental reviews.

In accordance with the National Environmental Policy Act of 1969 and 10 CFR Part 51, the staff reviewed the plant-specific environmental impacts of license renewal, including whether there was new and significant information not considered in the GEIS. As part of its scoping process, the staff held a public meeting on July 24, 2014, at the Monroe County Community College, in Monroe, Michigan, to identify plant-specific environmental issues. The draft, plant-specific GEIS

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Supplement 56 documents the results of the environmental review and makes a preliminary recommendation on the license renewal action. The staff held another public meeting on December 2, 2015, at the Monroe County Community College in Monroe, Michigan, to discuss the draft, plant-specific GEIS Supplement 56. After considering comments on the draft, the staff will publish the final, plant-specific GEIS Supplement 56 separately from this report.

1.3 Principal Review Matters

Part 54 of 10 CFR describes the requirements for renewal of operating licenses for nuclear power plants. The staff’s technical review of the LRA was in accordance with NRC guidance and 10 CFR Part 54 requirements. Section 54.29, “Standards for Issuance of a Renewed License,” of 10 CFR sets forth the license renewal standards. This SER describes the results of the staff’s safety review.

In accordance with 10 CFR 54.19(a), the NRC requires a license renewal applicant to submit general information, which the applicant provided in LRA Section 1. The staff reviewed LRA Section 1 and finds that the applicant has submitted the required information.

In accordance with 10 CFR 54.19(b), the NRC requires that the LRA 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.” On this issue, the applicant stated in the LRA:

10 CFR 54.19(b) requires that license renewal applications “... 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.” Item 3 of the Attachment to the current indemnity agreement (No. B-20) for Fermi 2, as revised by Amendment No. 27, lists Fermi 2 facility operating license number NPF-43 with no expiration date for the license. Therefore, no changes to the indemnity agreement are deemed necessary as part of this application. Should the license number be changed by NRC upon issuance of the renewed license, DTE requests that NRC amend the indemnity agreement to include conforming changes to Item 3 of the attachment and other affected sections of the agreement.

The staff intends to maintain the original license number upon issuance of the renewed license, if approved. Therefore, conforming changes to the indemnity agreement need not be made and the 10 CFR 54.19(b) requirements have been met.

In accordance with 10 CFR 54.21, “Contents of Application – Technical Information,” the NRC requires that the LRA contain (a) an integrated plant assessment, (b) a description of any CLB changes during the staff’s review of the LRA, (c) an evaluation of TLAAs, and (d) a UFSAR supplement. LRA Sections 3 and 4 and Appendix B address the license renewal requirements of 10 CFR 54.21(a), (b), and (c). LRA Appendix A satisfies the license renewal requirements of 10 CFR 54.21(d).

In accordance with 10 CFR 54.21(b), the NRC requires that, each year following submission of the LRA and at least 3 months before the scheduled completion of the staff’s review, the applicant submit an LRA amendment identifying any CLB changes to the facility that affect the contents of the LRA, including the UFSAR supplement. By letter dated May 9, 2016, the

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applicant submitted an LRA update that summarizes the CLB changes that have occurred during the staff’s review of the LRA. This submission satisfies 10 CFR 54.21(b) requirements.

In accordance with 10 CFR 54.22, “Contents of Application – Technical Specifications,” the NRC requires that the LRA include changes or additions to the Technical Specifications (TS) that are necessary to manage aging effects during the period of extended operation. In LRA Appendix D, the applicant stated that it had not identified any TS changes necessary for issuance of the renewed Fermi 2 operating license. This statement adequately addresses the 10 CFR 54.22 requirement.

The staff evaluated the technical information required by 10 CFR 54.21 and 10 CFR 54.22 in accordance with NRC regulations and SRP-LR guidance. SER Sections 2, 3, and 4 document the staff’s evaluation of the LRA technical information.

As required by 10 CFR 54.25, “Report of the Advisory Committee on Reactor Safeguards,” the ACRS will issue a report documenting its evaluation of the staff’s LRA review and SER. SER Section 5 is reserved for the ACRS report when it is issued. SER Section 6 documents the findings required by 10 CFR 54.29.

1.4 Interim Staff Guidance

License renewal is a living program. The staff, industry, and other interested stakeholders gain experience and develop lessons learned with each renewed license. The lessons learned address the staff’s performance goals of maintaining safety, improving effectiveness and efficiency, reducing regulatory burden, and increasing public confidence. Interim staff guidance (ISG) is documented for use by the staff, industry, and other interested stakeholders until incorporated into such license renewal guidance documents as the SRP-LR and GALL Report.

Table 1.4-1 shows the current set of ISGs, as well as the SER sections in which the staff addresses them.

Table 1.4-1 Current Interim Staff Guidance ISG Issue

(Approved ISG Number) Purpose SER Section

“Aging Management of Stainless Steel Structures and Components in Treated Borated Water,” Revision 1 (LR-ISG-2011-01)

This LR-ISG clarifies the staff’s existing position on aging management in treated borated water environments.

Not applicable for the SER

“Aging Management Program for Steam Generators” (LR-ISG-2011-02)

This LR-ISG evaluates the suitability of using Revision 3 of NEI 97-06 for implementing the licensee’s steam generator aging management program.

Not applicable to BWRs

“Changes to the Generic Aging Lessons Learned (GALL) Report Revision 2 AMP XI.M41, ‘Buried and Underground Piping and Tanks’” (LR-ISG-2011-03)

This LR-ISG gives additional guidance on managing the effects of aging on buried and underground piping and tanks.

SER Sections 3.0.3.1.2 and 3.0.3.2.7

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ISG Issue (Approved ISG Number) Purpose SER Section

“Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors” (LR-ISG-2011-04)

This LR-ISG updates the GALL Report, Revision 2, and SRP-LR, Revision 2, to ensure consistency with MRP-227-A for the aging management of age-related degradation for components of pressurized water reactor vessel internal components during the term of a renewed operating license.

Not applicable to BWRs

“Ongoing Review of Operating Experience” (LR-ISG-2011-05)

This LR-ISG clarifies the staff’s existing position in the SRP-LR that acceptable license renewal AMPs should be informed and enhanced when necessary, based on the ongoing review of both plant-specific and industry operating experience.

SER Section 3.0.5.2

“Wall Thinning Due to Erosion Mechanisms” (LR-ISG-2012-01)

This LR-ISG gives additional guidance on managing the effects of wall thinning due to erosion mechanisms.

SER Section 3.0.3.2.11

“Aging Management of Internal Surfaces, Fire Water Systems, Atmospheric Storage Tanks, and Corrosion Under Insulation” (LR-ISG-2012-02)

This LR-ISG gives guidance on managing the effects of aging for internal surfaces, fire water system, atmospheric storage tanks, and corrosion under insulation.

SER Sections 3.0.3.1.1, 3.0.3.1.11, 3.0.3.2.7, 3.0.3.2.10, 3.0.3.3.1, 3.3.2.1.14, and 3.3.2.3.3

“Aging Management of Loss of Coating or Lining Integrity for Internal Coatings/Linings on In-Scope Piping, Piping Components, Heat Exchangers, and Tanks” (LR-ISG-2013-01)

This LR-ISG gives guidance on aging management for coating or lining integrity for internal coatings/linings on in-scope piping, piping components, heat exchangers, and tanks.

SER Sections 3.0.3.2.24, 3.2.2.3.2, 3.3.2.1.1, 3.3.2.3.3, and 3.4.2.3.3

“Changes to Buried and Underground Piping and Tank Recommendations” (LR-ISG-2015-01)

This LR-ISG replaces GALL Report AMP XI.M41, “Buried and Underground Piping and Tanks,” and the associated final safety analysis report summary description. The LR-ISG provides revised guidance on managing aging effects associated with buried and underground piping and tanks.

SER Section 3.0.3.1.2

1.5 Summary of Open Items

As a result of its review of the LRA, including additional information submitted through May 30, 2016, the staff closed the following open item previously identified in the “Safety Evaluation Report with Open Items Related to the License Renewal of Fermi 2,” dated January 28, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16020A440). No other open items remain to be addressed. An item is considered open if, in the staff’s judgment, it does not meet all applicable regulatory requirements at the time of the issuance of this SER. A summary of the basis for the open item closure is presented here.

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Open Item 4.3.3-1 Effects of Reactor Water Environment on Fatigue Life

By letter dated September 24, 2015, the applicant provided its response to RAI 4.3.3-3. In this letter, the applicant stated that there are locations where the environmentally assisted fatigue (EAF) correction factors (Fen) were recalculated using average transient temperatures or maximum operating temperatures. The RAI 4.3.3-3 response also states that these Fen factors were recalculated in a manner consistent with NUREG/CR-6909, “Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials.”

The staff and applicant held a telephone conference call on December 15, 2015, to discuss and clarify the applicant’s response to RAI 4.3.3-3. Specifically, the topic of the telephone conference was to clarify the manner in which the average temperatures were calculated to confirm consistency with NUREG/CR-6909. During the telephone conference call, the applicant described the methods used to calculate average temperatures. Based on these descriptions, the staff determined that the applicant’s average temperature calculations were not always consistent with the guidance in NUREG/CR-6909 when the minimum temperature is below the temperature threshold for a given material. This may result in the underestimation of both the Fen factors and the resulting EAF cumulative usage factors (CUFen) for some locations. A summary of the telephone conference is provided in the NRC’s letter dated January 8, 2016 (ADAMS Accession No. ML16005A399). By letter dated January 14, 2016, the staff issued RAI 4.3.3-3a requesting that the applicant assess the impact of revising the evaluations to use the correct determination of average temperature in a manner consistent with NUREG/CR-6909 and submit a description of the impact of this revision to the previous screening and Fen evaluation results for staff review. The staff stated that the assessment should include a description of whether the revised average temperature calculations impact the selection of sentinel locations.

By letter dated March 10, 2016, the applicant provided its response to RAI 4.3.3-3a. The NRC staff finds that the applicant has demonstrated that, pursuant to 10 CFR 54.21(c)(1)(iii), the effects of environmentally assisted fatigue due to the reactor water environment on the intended functions of the American Society of Mechanical Engineers (ASME) Class 1 reactor pressure vessel boundary will be adequately managed for the period of extended operation. Additionally, it meets the acceptance criteria in SRP-LR Section 4.3.2.1.3 because the applicant has addressed the staff recommendation for the closure of GSI-190, “Fatigue Evaluation of Metal Components for 60-Year Plant Life.” The applicant has identified high-fatigue usage locations, including those in NUREG/CR-6260, “Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components,” and has evaluated these locations using the formulas, fatigue curves, and guidance in NUREG/CR-6909. The applicant is managing the effects of cumulative fatigue damage on the intended functions of the applicable components using the Fatigue Monitoring Program. The staff’s review of the Fatigue Monitoring Program appears in SER Section 3.0.3.2.8. Open Item 4.3.3-1 is closed.

1.6 Summary of Confirmatory Items

As a result of its review of the LRA, including additional information submitted through May 30, 2016, the staff determines that no confirmatory items exist that would require a formal response from the applicant.

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1.7 Summary of Proposed License Conditions

Following the staff’s review of the LRA, including subsequent information and clarifications from the applicant, the staff identified three proposed license conditions.

License Condition No. 1:

The information in the UFSAR supplement, submitted pursuant to 10 CFR 54.21(d), as revised during the license renewal application review process, and licensee commitments as listed in Appendix A to the “Safety Evaluation Report Related to the License Renewal of Fermi 2,” are collectively the “License Renewal UFSAR Supplement.” This supplement is henceforth part of the UFSAR which will be updated in accordance with 10 CFR 50.71(e). As such, the licensee may make changes to the programs, activities, and commitments described in this Supplement, provided the licensee evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59, “Changes, Tests and Experiments,” and otherwise complies with the requirements in that section.


The License Renewal UFSAR Supplement, as updated by license condition [1] above, describes certain programs to be implemented and activities to be completed before the period of extended operation, as follows:

(a) The applicant shall implement those new programs and enhancements to existing programs no later than 6 months prior to the period of extended operation [PEO].

(b) The applicant shall complete those activities by the 6-month date before the PEO or the end of the last refueling outage prior to the PEO, whichever occurs later.

The applicant shall notify the NRC in writing within 30 days after having accomplished item (a) above and include the status of those activities that have been or remain to be completed in item (b) above.


DTE shall fully implement the Boraflex rack replacement approved in Amendment No. 141 before the period of extended operation (i.e., March 20, 2025), so that the Boraflex material in the spent fuel pool will not be required to perform a neutron absorption function. DTE shall submit a letter to the NRC, within 60 days following completion of the removal of the Boraflex material and installation of the Boral material, as described in Amendment No. 141, confirming the removal of the Boraflex material and discontinued reliance on its neutron absorption function.

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STRUCTURES AND COMPONENTS SUBJECT TO AGING MANAGEMENT REVIEW

2.1 Scoping and Screening Methodology

2.1.1 Introduction

Title 10, Section 54.21, “Contents of Application – Technical Information,” of the Code of Federal Regulations (10 CFR 54.21), requires the applicant to identify the structures, systems, and components (SSCs) within the scope of license renewal in accordance with 10 CFR 54.4(a). In addition, the license renewal application (LRA) must contain an integrated plant assessment (IPA) that identifies and lists those structures and components (SCs), contained in the SSCs identified to be within the scope of license renewal, that are subject to an aging management review (AMR).

2.1.2 Summary of Technical Information in the Application

LRA Section 2.0, “Scoping and Screening Methodology for Identifying Structures and Components Subject to Aging Management Review and Implementation Results,” provides the technical information required by 10 CFR 54.21(a). LRA Section 2.0 states, in part, that the applicant had considered the following in developing the scoping and screening methodology described in LRA Section 2.0:

• Part 54 of 10 CFR, “Requirements for Renewal of Operating Licenses for Nuclear Power Plants” (the Rule)

• Nuclear Energy Institute (NEI) 95-10, Revision 6, “Industry Guideline for Implementing the Requirements of 10 CFR Part 54 – The License Renewal Rule,” dated June 2005 (NEI 95-10)

LRA Section 2.1, “Scoping and Screening Methodology,” describes the methodology used by DTE Electric Company (DTE or the applicant) to identify the SSCs at Fermi 2 within the scope of license renewal (scoping) and the SCs subject to an AMR (screening).

2.1.3 Scoping and Screening Program Review

The U.S. Nuclear Regulatory Commission (NRC or the staff) evaluated the applicant’s scoping and screening methodology in accordance with the guidance in NUREG-1800, Revision 2, “Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants” (SRP-LR), Section 2.1, “Scoping and Screening Methodology.” The following regulations provide the basis for the acceptance criteria that the staff used to assess the adequacy of the scoping and screening methodology the applicant used to develop the LRA:

• 10 CFR 54.4(a), as it relates to the identification of SSCs within the scope of the Rule

• 10 CFR 54.4(b), as it relates to the identification of the intended functions of SSCs within the scope of the Rule

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• 10 CFR 54.21(a), as it relates to the methods used by the applicant to identify plant SCs subject to an AMR

The staff reviewed the information in LRA Section 2.1 to confirm that the applicant described a process for identifying SSCs that are within the scope of license renewal in accordance with 10 CFR 54.4(a) and for identifying SCs that are subject to an AMR in accordance with 10 CFR 54.21(a).

In addition, the staff conducted a scoping and screening methodology audit at the Fermi 2 facility located in Monroe County, Michigan, during the week of August 4-7, 2014. The audit focused on ensuring that the applicant had developed and implemented adequate guidance to conduct the scoping and screening of SSCs in accordance with the methodology described in the LRA and with the requirements of the Rule. The staff reviewed the project-level guidelines, technical basis documents, and implementing procedures that described the applicant’s scoping and screening methodology. The staff conducted detailed discussions with the applicant on the implementation and control of the license renewal methodology, the quality practices used by the applicant during the LRA development, and the training of the applicant’s staff that participated in the LRA development.

On a sampling basis, the staff performed a review of scoping and screening results reports and supporting current licensing basis (CLB) information for portions of the emergency equipment service water (EESW) system and residual heat removal (RHR) complex support equipment and corresponding structures. In addition, the staff performed walkdowns of selected portions of those systems and structures as a part of the sampling review of the implementation of the applicant’s 10 CFR 54.4(a)(2) scoping methodology.

2.1.3.1 Implementation Procedures and Documentation Sources Used for Scoping and Screening

Summary of Technical Information in the Application

The applicant had developed implementing procedures used to identify SSCs within the scope of license renewal and SCs subject to an AMR to implement the processes described in LRA Sections 2.0 and 2.1. Additionally, the applicant’s implementing procedures provided guidance on the review and consideration of CLB documentation sources relative to the requirements in 10 CFR 54.4, “Scope” and 10 CFR 54.21.

LRA Section 2.1 listed the following information sources for the license renewal scoping and screening process:

• Fermi 2 Central Component (CECO) database • updated final safety analysis report (UFSAR) • maintenance rule basis documents • design basis documents • fire hazards analysis • safe shutdown analysis • station drawings

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Staff Evaluation

Scoping and Screening Implementation Procedures. The staff reviewed the applicant’s scoping and screening methodology implementing procedures, including license renewal guidelines, documents, and reports, as documented in the staff’s audit report, to ensure that the guidance is consistent with the requirements of the Rule, the SRP-LR, and Regulatory Guide (RG) 1.188, “Standard Format and Content for Applications to Renew Nuclear Plant Operating Licenses,” Revision 1, dated September 2005, which endorses the use of NEI 95-10. The staff determined that the overall process used to implement the 10 CFR Part 54 requirements described in the implementing procedures, including license renewal guidelines, documents, and reports, is consistent with the Rule, the SRP-LR, and the endorsed industry guidance.

The applicant’s implementing procedures contain guidance for determining plant SSCs within the scope of the Rule and for determining SCs contained in systems within the scope of license renewal that are subject to an AMR. During the review of the implementing procedures, the staff focused on the consistency of the detailed procedural guidance with information in the LRA, including the implementation of the staff’s positions documented in the SRP-LR, and with information in the applicant’s responses, dated November 18, 2014, to the staff’s requests for additional information (RAIs), dated October 20, 2014. After reviewing the LRA and supporting documentation, the staff determined that the scoping and screening methodology instructions are consistent with the methodology description provided in LRA Section 2.1. The staff also determined that the methodology is sufficiently detailed in the implementing procedures to provide concise guidance on the scoping and screening process to be followed during the LRA activities.

Sources of Current Licensing Basis Information. The regulation at 10 CFR Part 54.21(a)(3) requires, for each SC determined to be subject to an AMR, the applicant to demonstrate that the effects of aging will be adequately managed so that the intended functions will be maintained consistent with the CLB for the period of extended operation. The regulation at 10 CFR Part 54.3(a) defines the CLB, in part, as the set of NRC requirements applicable to a specific plant and a licensee’s written commitments for ensuring compliance with, and operation within, applicable NRC requirements and the plant-specific design bases that are docketed and in effect. The CLB includes applicable NRC regulations, orders, license conditions, exemptions, technical specifications, and design basis information (documented in the most recent UFSAR). The CLB also includes licensee commitments remaining in effect that were made in docketed licensing correspondence, such as licensee responses to NRC bulletins, generic letters, and enforcement actions, and licensee commitments documented in NRC safety evaluations or licensee event reports. The staff considered the scope and depth of the applicant’s CLB review to verify that the methodology is sufficiently comprehensive to identify SSCs within the scope of license renewal and as SCs that are subject to an AMR.

During the scoping and screening methodology audit, the staff confirmed that the applicant’s detailed license renewal program guidelines specified the use of the CLB source information in developing scoping evaluations. The staff reviewed pertinent information sources used by the applicant including the Fermi 2 CECO database, the UFSAR, maintenance rule basis documents, design basis documents, fire hazards analysis, safe shutdown analysis, and station drawings.

During the audit, the staff discussed the applicant’s administrative controls for the CECO database and the other information sources used to verify system information. These controls are described and implemented by plant procedures. Based on a review of the administrative

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controls and on a sample of the system classification information contained in the applicable documentation, the staff determined that the applicant has established adequate measures to control the integrity and reliability of system identification and safety classification data; therefore, the staff determined that the information sources used by the applicant during the scoping and screening process provided a controlled source of system and component data to support scoping and screening evaluations.

In addition, the staff reviewed the implementing procedures and results reports used to support identification of SSCs that the applicant relied on to demonstrate compliance with the requirements of 10 CFR 54.4(a). The applicant’s license renewal program guidelines provided a listing of documents used to support scoping evaluations. The staff determined that the design documentation sources, required to be used by the applicant’s implementing procedures, provided sufficient information to ensure that the applicant identified SSCs to be included within the scope of license renewal consistent with the plant’s CLB.

Conclusion

Based on its review of LRA Sections 2.0 and 2.1, the scoping and screening implementing procedures, and the results from the scoping and screening audit, the staff concludes that the applicant’s use of implementing procedures and consideration of document sources, including CLB information, is consistent with the Rule, the SRP-LR, and NEI 95-10 guidance and, therefore, is acceptable.

2.1.3.2 Quality Controls Applied to License Renewal Application Development

Staff Evaluation

The staff reviewed the quality controls used by the applicant to ensure that the scoping and screening methodology used to develop the LRA were adequate for the activity. The applicant used the following quality control processes during the LRA development:

• The license renewal team coordinated and reviewed all license renewal activities.

• Subject matter experts, supervisors, and managers prepared and reviewed basis documents, reports, and the LRA.

• The nuclear quality assurance organization performed a surveillance of LRA development activities.

• Industry peers reviewed the draft LRA.

• The onsite safety review organization and nuclear safety review group reviewed the LRA.

The staff performed a review of implementing procedures and guides, examined the applicant’s documentation of activities in reports, reviewed the applicant’s activities performed to assess the quality of the LRA, and held discussions with the applicant’s license renewal management and staff. The staff determines that, through its activities, the applicant has provided assurance that the LRA was developed consistent with its license renewal program requirements.

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Conclusion

Based on its review of pertinent LRA development guidance and review of the applicant’s documentation of the activities performed to assess the quality of the LRA, the staff concludes that the applicant’s quality assurance activities are adequate to ensure that LRA development activities were performed in accordance with the applicant’s license renewal program requirements.

2.1.3.3 Training

Staff Evaluation

The staff reviewed the training process used by the applicant for license renewal project personnel to confirm that it was appropriate for the activity. As specified by the license renewal implementing procedures, the applicant has required training and qualification of personnel performing activities supporting the development of the LRA, including identification of SSCs within the scope of license renewal, identification of SCs subject to an AMR, and documentation of the information in reports.

Training included the following topics and activities:

• Fermi 2 License Renewal Project Plan • license renewal overview • operating experience review • industry guidelines for implementation of 10 CFR Part 54 • NRC SRP-LR • NUREG-1801, Revision 2, “Generic Aging Lessons Learned (GALL) Report” • system and structure scoping • mechanical system screening and AMR • structural screening and AMR • electrical system screening and AMR • evaluation of aging management programs (AMPs) • time-limited aging analyses and exemptions evaluation • LRA development

The staff discussed training activities with the applicant’s management and license renewal project personnel and performed a sampling review of applicable documentation. The staff determines that the applicant has developed and implemented adequate controls for the training of personnel performing LRA activities.

Conclusion

Based on discussions with the applicant’s license renewal personnel responsible for the scoping and screening process and its review of selected documentation in support of the process, the staff concludes that the applicant has developed and implemented adequate procedures to train personnel to implement the scoping and screening methodology described in its implementing procedures and the LRA.

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2.1.3.4 Conclusion of Scoping and Screening Program Review

Based on its review of information provided in LRA Sections 2.0 and 2.1, review of the applicant’s scoping and screening implementing procedures, discussions with the applicant’s license renewal personnel, review of the quality controls applied to the LRA development, training of personnel participating in the LRA development, and the results from the scoping and screening methodology audit, the staff concludes that the applicant’s scoping and screening program is consistent with the SRP-LR and the requirements of 10 CFR Part 54 and, therefore, is acceptable

2.1.4 Plant Systems, Structures, and Components Scoping Methodology

LRA Section 2.1 describes the applicant’s methodology used to identify SSCs within the scope of license renewal pursuant to the criteria in 10 CFR 54.4(a). The LRA states that the scoping process identified the SSCs that (1) are safety related, (2) perform and support an intended function for responding to a design basis event (DBE), (3) are nonsafety related whose failure could prevent accomplishment of a safety-related function, or (4) support a specific requirement for one of the regulated events applicable to license renewal. In addition, the LRA states that the scoping methodology used was consistent with 10 CFR Part 54 and with the industry guidance in NEI 95-10.

2.1.4.1 Application of the Scoping Criteria in 10 CFR 54.4(a)(1)


The applicant addressed the methods used to identify SSCs included within the scope of license renewal in accordance with 10 CFR 54.4(a)(1) in LRA Section 2.1.1.1, “Application of Safety-Related Scoping Criteria,” which states:

A system or structure is within the scope of license renewal if it performs a safety function during and following a design basis event as defined in 10 CFR 50.49(b)(1). Design basis events are defined in 10 CFR 50.49(b)(1) as conditions of normal operation, including anticipated operational occurrences, design basis accidents, external events, and natural phenomena for which the plant must be designed to ensure functions identified in 10 CFR 54.4(a)(1)(i) through (iii). A Fermi 2 engineering procedure provides the criteria and methodology for determining and evaluating the safety and quality classification of systems, structures, and components.

Staff Evaluation

Pursuant to 10 CFR 54.4(a)(1), the applicant must consider all safety-related SSCs that are relied on to remain functional during and following a DBE to ensure the following functions: (1) the integrity of the reactor coolant pressure boundary (RCPB), (2) the ability to shut down the reactor and maintain it in a safe shutdown condition, or (3) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to those referred to in 10 CFR 50.34(a)(1); 10 CFR 50.67(b)(2); or 10 CFR 100.11, “Determination of Exclusion Area, Low Population Zone, and Population Center Distance,” as applicable.

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With regard to identification of DBEs, SRP-LR Section 2.1.3, “Review Procedures,” states:

The set of design basis events as defined in the rule is not limited to Chapter 15 (or equivalent) of the UFSAR. Examples of design basis events that may not be described in this chapter include external events, such as floods, storms, earthquakes, tornadoes, or hurricanes, and internal events, such as a high energy line break. Information regarding design basis events as defined in 10 CFR 50.49(b)(1) may be found in any chapter of the facility UFSAR, the Commission’s regulations, NRC orders, exemptions, or license conditions within the CLB. These sources should also be reviewed to identify systems, structures, and components that are relied upon to remain functional during and following design basis events (as defined in 10 CFR 50.49(b)(1)) to ensure the functions described in 10 CFR 54.4(a)(1).

During the audit, the applicant stated that it evaluated the types of events listed in NEI 95-10 (anticipated operational occurrences, design basis accidents (DBAs), external events, and natural phenomena) that were applicable to Fermi 2. The staff reviewed the applicant’s basis documents, which described design basis conditions in the CLB, and addressed events defined by 10 CFR 50.49(b)(1) and 10 CFR 54.4(a)(1). The UFSAR and basis documents discussed events, such as internal and external flooding, tornados, and missiles. The staff concludes that the applicant’s evaluation of DBEs was consistent with the SRP-LR.

The staff determined that the applicant has performed scoping of SSCs for the 10 CFR 54.4(a)(1) criterion in accordance with the license renewal implementing procedures that provide guidance for the preparation, review, verification, and approval of the scoping evaluations to ensure the adequacy of the results of the scoping process. The staff reviewed the implementing procedures governing the applicant’s evaluation of safety-related SSCs and sampled the applicant’s reports of the scoping results to ensure that the applicant applied the methodology in accordance with the implementing procedures. In addition, the staff discussed the methodology and results with the applicant’s personnel who were responsible for these evaluations.

The staff reviewed the applicant’s evaluation of the Rule and CLB definitions pertaining to 10 CFR 54.4(a)(1). The staff noted the applicant’s CLB definition of safety related met the definition of safety related as specified in the Rule with the exception that it did not include a reference to the 10 CFR 50.67(b)(2) alternate source term related to potential offsite exposure criteria. The staff reviewed LRA Section 2.1.1, “Scoping Methodology,” which states that the applicant had reviewed the applicability of 10 CFR 50.67, “Accident Source Term,” and determined that the applicant had credited the use of alternate source term in the fuel-handling accident and loss-of-coolant accident (LOCA) analyses and had identified one additional SSC, the standby liquid control (SLC) system, which was included within the scope of license renewal. The staff determined that the applicant had reconciled the difference between the CLB definition of safety related and the definition in 10 CFR 54.4(a)(1).

The staff reviewed a sample of the license renewal scoping results for portions of the EESW system and RHR complex support equipment and corresponding structures to provide additional assurance that the applicant adequately implemented its scoping methodology with respect to 10 CFR 54.4(a)(1).

The staff verified that the applicant had developed the scoping results for each of the sampled systems consistently with the methodology, identified the SSCs credited for performing intended

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functions, and adequately described the basis for the results and the intended functions. The staff also confirmed that the applicant had identified and used pertinent engineering and licensing information to identify the SSCs required to be within the scope of license renewal in accordance with the criteria in 10 CFR 54.4(a)(1).

Conclusion

Based on its review of the LRA, the applicant’s implementing procedures and reports, and the review of two systems on a sampling basis, the staff concludes that the applicant’s methodology for identifying safety-related SSCs relied on to remain functional during and following DBEs and including them within the scope of license renewal is consistent with the SRP-LR and 10 CFR 54.4(a)(1) and, therefore, is acceptable.

2.1.4.2 Application of the Scoping Criteria in 10 CFR 54.4(a)(2)


The applicant addressed the methods that it used to identify SSCs included within the scope of license renewal in accordance with 10 CFR 54.4(a)(2).

LRA Section 2.1.1.2, “Application of Criterion for Nonsafety-Related SSCs Whose Failure Could Prevent the Accomplishment of Safety Functions,” states, in part:

Functional Support for Safety-Related SSC 10 CFR 54.4(a)(1) Functions. At Fermi 2, systems and structures required to perform a function to support a safety function are classified as safety-related and have been included in the scope of license renewal per Section 2.1. For the exceptions where nonsafety-related equipment is required to remain functional to support a safety function (e.g., support the fuel pool cooling and cleanup system in removing decay heat from fuel assemblies stored in the fuel pools), the system containing the equipment has been included in scope, and the function is listed as an intended function for 10 CFR 54.4(a)(2) for the system.

Connected to and Provide Structural Support for Safety-Related SSCs. For nonsafety-related SSCs directly connected to safety-related SSCs (typically piping systems), components within the scope of license renewal include the connected piping and supports up to and including the first seismic or equivalent anchor beyond the safety-nonsafety interface, or up to a point determined by alternative bounding criteria (such as a base-mounted component, flexible connection, or the end of a piping run).

Potential for Spatial Interactions with Safety-Related SSCs. Moderate- and low-energy systems have the potential for spatial interactions of spray and leakage. Nonsafety-related systems and nonsafety-related portions of safety-related systems with the potential for spray or leakage that could prevent safety-related SSCs from performing their required safety function are within the scope of license renewal and subject to aging management review.

The review used a spaces approach for scoping of nonsafety-related systems with potential spatial interaction with safety-related SSCs. The spaces approach focuses on the interaction between nonsafety-related and safety-related SSCs

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that are located in the same space. A space is defined as a room or cubicle that is separated from other spaces by substantial objects (such as wall, floors, or ceilings). The space is defined such that any potential interaction between nonsafety-related and safety-related SSCs, including flooding, is limited to the space. Nonsafety-related systems that contain water, oil, or steam with components located inside structures containing safety-related SSCs are potentially in scope for possible spatial interaction under criterion 10 CFR 54.4(a)(2). These systems were evaluated further to determine if system components were located in a space such that safety-related equipment could be affected by a component failure.

Staff Evaluation

RG 1.188, Revision 1, endorses the use of NEI 95-10, Revision 6, which discusses the implementation of the staff’s position on 10 CFR 54.4(a)(2) scoping criteria, to include nonsafety-related SSCs that may have the potential to prevent satisfactory accomplishments of safety-related intended functions. Such SSCs include nonsafety-related SSCs connected to safety-related SSCs, nonsafety-related SSCs in proximity to safety-related SSCs, and mitigative and preventive options related to nonsafety-related and safety-related SSCs interactions. LRA Section 2.0 states that the applicant’s methodology is consistent with the guidance in Appendix F to NEI 95-10, Revision 6.

In addition, the staff’s position (as discussed in the SRP-LR Section 2.1.3.1.2) is that the applicant should not consider hypothetical failures; instead, it should base its evaluation on the plant’s CLB, engineering judgment and analyses, and relevant operating experience. NEI 95-10 further describes operating experience as all documented plant-specific and industry-wide experience that can be used to determine the plausibility of a failure. Documentation would include NRC generic communications and event reports; plant-specific condition reports; industry reports, such as safety operational event reports; and engineering evaluations. The staff reviewed LRA Section 2.1.1.2 and subsections in which the applicant described the scoping methodology for nonsafety-related SSCs pursuant to 10 CFR 54.4(a)(2). In addition, the staff reviewed the applicant’s implementing procedure and results report, which documented the guidance and corresponding results of the applicant’s scoping review pursuant to 10 CFR 54.4(a)(2).

Nonsafety-Related SSCs Required to Perform a Function That Supports a Safety-Related SSC. The staff reviewed LRA Section 2.1.1.2.1, “Functional Failures of Nonsafety-Related SSCs,” and the applicant’s 10 CFR 54.4(a)(2) implementing procedure that describes the method used to identify those nonsafety-related SSCs required to perform a function that supports a safety-related SSC intended function, within the scope of license renewal in accordance with 10 CFR 54.4(a)(2). The staff confirmed that the applicant had reviewed the UFSAR, plant drawings, the CECO database, and other CLB documents to identify the nonsafety-related systems and structures that function to support a safety-related system whose failure could prevent the performance of a safety-related intended function. The staff determined that the applicant had identified the nonsafety-related SSCs that performed a safety function or supported a safety system that would require the nonsafety-related SSC to be included within the scope of license renewal in accordance with 10 CFR 54.4(a)(2).

The staff determined that the applicant’s methodology for identifying nonsafety-related systems that perform functions that support safety-related intended functions for inclusion within the

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scope of license renewal was in accordance with the guidance in SRP-LR and the requirements of 10 CFR 54.4(a)(2).

Nonsafety-Related SSCs Directly Connected to Safety-Related SSCs. The staff reviewed LRA Section 2.1.1.2.2, “Physical Failures of Nonsafety-Related SSCs,” and the applicant’s 10 CFR 54.4(a)(2) implementing procedure that describes the method used to identify nonsafety-related SSCs, directly connected to safety-related SSCs, within the scope of license renewal in accordance with 10 CFR 54.4(a)(2). The applicant had reviewed the safety-related to nonsafety-related interfaces for each mechanical system in order to identify the nonsafety-related components located between the safety- to nonsafety-related interface and license renewal structural boundary.

The staff determined that the applicant had used a combination of the following to identify the portion of nonsafety-related piping systems to include within the scope of license renewal:

• seismic anchors

• equivalent anchors

• bounding conditions described in Appendix F to NEI 95-10, Revision 6 (i.e., base-mounted component, flexible connection, inclusion to the free end of nonsafety-related piping, inclusion of the entire piping run, or a branch line off of a header where the moment of inertia of the header is greater than seven times the moment of inertia of the branch)

The staff determined that the applicant’s methodology for identifying and including nonsafety-related SSCs, directly connected to safety-related SSCs, within the scope of license renewal was in accordance with the guidance in SRP-LR and the requirements of 10 CFR 54.4(a)(2).

The staff determined that additional information would be required to complete its review. RAI 2.1-1, dated October 20, 2014, states, in part:

During the on-site scoping and screening methodology audit the staff reviewed the implementing document used by the applicant to identify nonsafety-related SSCs with the potential to affect safety-related SSCs, for inclusion within the scope of license renewal. The applicant’s implementing document states that only nonsafety-related SSCs that had not been included within the scope of license renewal based on the potential for spray or leakage needed to be reviewed for nonsafety-related SSCs directly attached to safety-related SSCs to identify the portion of the nonsafety-related SSC up to the first anchor, equivalent anchor, or bounding condition, past the safety-related/nonsafety-related interface. The staff required additional information to determine that the applicant had identified the portion of the nonsafety-related SSC, past the safety-related/nonsafety-related interfac

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