ML25080A241
| ML25080A241 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/21/2025 |
| From: | Hammock J NRC/NRR/DNRL/NLRP |
| To: | Erb D Tennessee Valley Authority |
| References | |
| EPID L-2023-SLR-0000 | |
| Download: ML25080A241 (1) | |
Text
March 21, 2025 Delson C. Erb Vice President, OPS Support Tennessee Valley Authority 1101 Market Street, LP 4A-C Chattanooga, TN 37402-2801
SUBJECT:
BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3 - REPORT FOR THE AGING MANAGEMENT AUDIT REGARDING THE SUBSEQUENT LICENSE RENEWAL APPLICATION REVIEW (EPID NO. L-2023-SLR-0000)
Dear Delson C. Erb:
By Subsequent License Renewal Application Appendix E, Applicant’S Environmental Report-Operating License Renewal Stage|letter dated January 19, 2024]] (Agencywide Documents Access and Management System (ADAMS) Package Accession No.ML24019A009), as supplemented by letters dated January 22, 2024 (ML24022A292), October 9, 2024 (ML24283A091), November 1, 2024 (ML24306A203), December 17, 2024 (ML24352A216), January 8, 2025 (ML25008A150),
February 12, 2025 (ML25043A270 & ML25043A035), and March 4, 2025 (ML25063A184),
Tennessee Valley Authority (TVA, the applicant) submitted an application for subsequent license renewal (SLR) of Renewed Facility Operating License Nos.DPR-33, DPR-52, and DPR-68 for Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, respectively, to the U.S. Nuclear Regulatory Commission (NRC). TVA submitted the application pursuant to Title 10 of the Code of Federal Regulations part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, for SLR.
The NRC staff completed its aging management audit from March 18, 2024 - January 29, 2025, in accordance with the audit plan (ML24072A028). The audit report is enclosed. If you have any questions, please contact me by email at Jessica.Hammock@nrc.gov.
Sincerely,
/RA/
Jessica Hammock, Project Manager License Renewal Projects Branch Division of New and Renewed Licenses Office of Nuclear Reactor Regulation Docket Nos: 50-259, 50-260, and 50-296
Enclosure:
Audit Report cc w/encl: Listserv
Enclosure Aging Management Audit Report Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application March 18, 2024 - January 29, 2025 Division of New and Renewed Licenses Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission
ii U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION DIVISION OF NEW AND RENEWED LICENSES Docket No:
50-259, 50-260, 50-296 License No:
DPR-33, DPR-52, DPR-68 Licensee:
Tennessee Valley Authority (TVA)
Facility:
Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3 Location:
Rockville, Maryland Athens, Alabama Dates:
March 18, 2024 - January 29, 2025 Approved By:
Steven D. Bloom, Chief Corrosion and Steam Generator Branch Division of New and Renewed Licenses Angie R. Buford, Chief Vessels and Internals Branch Division of New and Renewed Licenses Ian Tseng, Chief Structural, Civil, Geotech Engineering Branch B Division of Engineering and External Hazards Bill Rogers, Acting Chief License Renewal Projects Branch Division of New and Renewed Licenses Jason Paige, Acting Chief Long Term Operations and Modernization Branch Division of Engineering and External Hazards Philip Sahd, Chief Nuclear Systems Performance Branch Division of Safety Systems Matthew A. Mitchell, Chief Piping and Head Penetrations Branch Division of New and Renewed Licenses Milton Valentin, Chief Containment & Plant Systems Branch Division of Safety Systems NRC Staff Reviewers:
iv Seung Min NPHP/DNRL Varoujan Kalikian NPHP/DNRL Carol Moyer Vessels and Internals (NVIB)/DNRL Carolyn Fairbanks NVIB/DNRL Cory Parker NVIB/DNRL David Dijamco NVIB/DNRL Emma Haywood NVIB/DNRL Eric Palmer NVIB/DNRL Jim Medoff NVIB/DNRL John Tsao NVIB/DNRL Mike Benson NVIB/DNRL On Yee NVIB/DNRL Steven Levitus NVIB/DNRL Derek (Rick) Scully Containment and Plant Systems (SCPB)/Division of Safety Systems (DSS)
Nicholas Soliz SCPB/DSS Raul Hernandez SCPB/DSS Rob Atienza SCPB/DSS Fred Forsaty Nuclear Systems Performance (SNSB)/DSS Santosh Bhatt SNSB/DSS Kevin Heller Nuclear Methods and Fuels (SFNB)/DSS Jessica Hammock License Renewal Projects (NLRP)/DNRL Austin Im NLRP/DNRL Chris Tyree NLRP/DNRL Bill Rogers NLRP/DNRL
v TABLE OF CONTENTS TABLE OF CONTENTS.............................................................................................................. v ACRONYMS.............................................................................................................................. ix
- 1. Introduction.......................................................................................................................... 1
- 2. Audit Activities...................................................................................................................... 2 SLRA Section 2.1, Scoping and Screening Methodology..................................................... 3 SLRA AMP B.2.1.1, ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD............................................................................................................................... 4 SLRA AMP B.2.1.2, Water Chemistry.................................................................................. 5 SLRA AMP B.2.1.3, Reactor Head Closure Stud Bolting...................................................... 6 SLRA AMP B.2.1.4, Boiling Water Reactor (BWR) Vessel ID Attachment Welds................. 7 SLRA AMP B.2.1.5, BWR Stress Corrosion Cracking Program............................................ 9 SLRA AMP B.2.1.6, BWR Penetrations.............................................................................. 10 SLRA AMP B.2.1.7, BWR Vessel Internals........................................................................ 11 SLRA AMP B.2.1.8, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)............................................................................................................................... 14 SLRA AMP B.2.1.9, Flow-Accelerated Corrosion............................................................... 15 SLRA AMP B.2.1.10, Bolting Integrity................................................................................ 18 SLRA AMP B.2.1.11, Open Cycle Cooling Water............................................................... 19 SLRA AMP B.2.1.12, Closed Treated Water Systems........................................................ 19 SLRA AMP B.2.1.13, Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems............................................................................................. 20 SLRA AMP B.2.1.14, Compressed Air Monitoring.............................................................. 22 SLRA AMP B.2.1.15, Fire Protection.................................................................................. 23 SLRA AMP B.2.1.16, Fire Water System............................................................................ 25 SLRA AMP B.2.1.17, Outdoor and Large Metallic Storage Tanks Program........................ 27 SLRA AMP B.2.1.18, Fuel Oil Chemistry............................................................................ 28 SLRA AMP B.2.1.19, Reactor Vessel Material Surveillance............................................... 29 SLRA AMP B 2.1.20, One-Time Inspection........................................................................ 30 SLRA AMP B.2.1.21, Selective Leaching........................................................................... 31 SLRA AMP B.2.1.22, ASME Code Class 1 Small-Bore Piping Program............................. 32 SLRA AMP B.2.1.23, External Surfaces Monitoring of Mechanical Components................ 34 SLRA AMP B.2.1.24, Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components.......................................................................................................... 34 SLRA AMP B.2.1.25, Lubricating Oil Analysis.................................................................... 35
vi SLRA AMP B.2.1.26, Monitoring of Neutron-Absorbing Materials Other Than Boraflex.............................................................................................................................. 36 SLRA AMP B.2.1.27, Buried and Underground Piping and Tanks...................................... 37 SLRA AMP B.2.1.28, Internal Coatings/Linings for In-Scope Piping, Piping Components, Tanks, and Heat Exchangers....................................................................... 39 SLRA AMP B.2.1.29, ASME Section XI, Subsection IWE................................................... 40 SLRA AMP B.2.1.30, ASME Section XI, Subsection IWF................................................... 41 SLRA AMP B.2.1.31, 10 CFR Part 50, Appendix J............................................................. 44 SLRA AMP B.2.1.32, Masonry Walls.................................................................................. 46 SLRA AMP B.2.1.33, Structures Monitoring....................................................................... 46 SLRA AMP B.2.1.34, Inspection of Water-Control Structures Associated with Nuclear Power Plants...................................................................................................................... 49 SLRA AMP B.2.1.35, Protective Coating Monitoring and Maintenance.............................. 50 SLRA AMP B.2.1.36, Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements................................. 51 SLRA AMP B.2.1.37, Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits...................................................................................................... 52 SLRA AMP B.2.1.38, Electrical Insulation for Inaccessible Medium-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements............... 54 SLRA AMP B.2.1.39, Electrical Insulation for Inaccessible Instrument and Control Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements............... 55 SLRA AMP B.2.1.40, Electrical Insulation for Inaccessible Low-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements............... 57 SLRA AMP B.2.1.41, Metal Enclosed Bus.......................................................................... 58 SLRA AMP B.2.1.42, Fuse Holders.................................................................................... 59 SLRA AMP B.2.1.43, Electrical Cable Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements......................................................................... 60 SLRA AMP B.3.1.1, Fatigue Monitoring.............................................................................. 60 SLRA AMP B.3.1.3, Environmental Qualification of Electric Equipment.............................. 63 SLRA TLAA Section 4.1, TLAA and Exemption Identification............................................. 64 SLRATLAA Section 4.2.2, Reactor Vessel Upper-Shelf Energy (USE) Analyses and TLAA Section 4.2.3, Reactor Vessel Adjusted Reference Temperature (ART)
Analyses............................................................................................................................ 65 SLRATLAA Section 4.2.4, Reactor Vessel Pressure-Temperature (P-T) Limits................. 66 SLRA TLAA Section 4.2.5 Reactor Vessel Circumferential Weld Failure Probability Analyses............................................................................................................................ 66 SLRA TLAA Section 4.2.6, Reactor Vessel Axial Weld Failure Probability Analyses.......... 68 SLRA TLAA Section 4.2.7, Reactor Vessel Reflood Thermal Shock Analysis.................... 69 SLRA TLAA Section 4.2.8, Core Shroud Reflood Thermal Shock Analysis........................ 70
vii SLRA TLAA Section 4.2.9, Core Plate Hold-Down Bolt Loss of Preload Analysis.............. 71 SLRA TLAA Section 4.2.10, Jet Pump Slip Joint Repair Clamp Loss of Preload Analysis.............................................................................................................................. 71 SLRA TLAA Section 4.2.11, Jet Pump Auxiliary Spring Wedge Assembly Loss of Preload Analysis................................................................................................................ 72 SLRA TLAA Section 4.2.12, Jet Pump Riser Repair Clamp Loss of Preload Analysis........ 73 SLRA TLAA Section 4.2.13, Replacement Core Support Plate Plug Extended Life Irradiation - Enhanced Stress Relaxation Analysis............................................................ 74 SLRA TLAA Section 4.2.14, Irradiation Assisted Stress Corrosion Cracking (IASCC) of Reactor Vessel Internals................................................................................................ 75 SLRA TLAA Section 4.2.15, Core Spray Replacement Piping Bolting Loss of Preload Evaluation.......................................................................................................................... 76 SLRA TLAA Section 4.2.16, Core Spray Sparger Repair Clamp Loss of Preload Evaluation.......................................................................................................................... 76 SLRA TLAA Section 4.2.17, Access Hole Cover Repair Loss of Preload Evaluation.......... 78 SLRA TLAA Section 4.2.18, Jet Pump Hold-Down Beam Assembly Loss of Preload Analysis.............................................................................................................................. 79 SLRA TLAA Section 4.2.19, Jet Pump Sensing Line Clamps Loss of Preload Analysis.............................................................................................................................. 79 SLRA TLAA Section 4.3.1, Transient Cycle and Cumulative Usage Projections for 80 Years................................................................................................................................. 80 SLRA TLAA Section 4.3.2, Metal Fatigue of Class 1 Components..................................... 81 SLRA TLAA Section 4.3.3, Class 1 Fatigue Waivers.......................................................... 83 SLRA TLAA Section 4.3.4, Metal Fatigue of Non-Class 1 Components.............................. 84 SLRA TLAA Section 4.3.5, Environmental Fatigue Analyses for Reactor Vessel and Class 1 Piping.................................................................................................................... 86 SLRA TLAA Section 4.3.6, Replacement Steam Dryer Stress Report and Fatigue Evaluation.......................................................................................................................... 87 SLRA TLAA Section 4.3.7, Emergency Equipment Cooling Water (EECW) System Weld Flaws Evaluation....................................................................................................... 88 SLRA TLAA Section 4.3.8, Core Shroud Support Fatigue Analysis Reevaluation.............. 89 SLRA TLAA Section 4.3.9, BFN Unit 3 Core Spray T-Box Repair Fatigue Evaluation........ 90 SLRA TLAA Section 4.3.10, BFN Unit 3 Core Spray Lower Line Section Replacement Fatigue......................................................................................................... 91 SLRA TLAA Section 4.3.11, Jet Pump to Core Shroud Support Plate Fatigue Evaluation.......................................................................................................................... 93 SLRA TLAA Section 4.4, Environmental Qualification of Electric Equipment...................... 94 SLRA TLAA Section 4.6, Primary Containment Fatigue Analyses...................................... 95 SLRA TLAA Section 4.7.1, Crane Load Cycle Limits.......................................................... 96 SLRA TLAA Section 4.7.2, Radiation Degradation of Drywell Expansion Gap Foam Analysis.............................................................................................................................. 96
viii SLRA TLAA Section 4.7.3, BFN Unit 2 Reactor Vessel Axial Weld Flaw............................ 97 AMR Items Not Associated with an AMP............................................................................ 98 SLRA AMR 3.6.2.3.1, High-Voltage Electrical Insulators.................................................... 98 SLRA AMR 3.6.2.2.3, Loss of Material Due to Wind-Induced Abrasion, Loss of Conductor Strength Due to Corrosion, and Increased Resistance of Connection Due to Oxidation or Loss of Preload for Transmission Conductors, Switchyard Bus, and Connections..................................................................................................................... 100 SLRA Further Evaluations for Stress Corrosion Cracking and Loss of Material for Stainless Steel, Nickel Alloys, and Aluminum Alloys......................................................... 101 SLRA Section 3.1.2.2.1, Cumulative Fatigue Damage..................................................... 102 SLRA Section 3.5 Irradiated Concrete.............................................................................. 102
- 3. Supplements to the SLRA................................................................................................. 110
- 4. Audit Questions Provided to TVA...................................................................................... 110
- 5. Applicant Personnel Contacted During Audit..................................................................... 113
- 6. Exit Meeting...................................................................................................................... 114
ix ACRONYMS AND ABBREVIATIONS ADAMS Agencywide Documents Access and Management System AMP aging management program AMR aging management review ART Adjusted reference temperature ASME American Society of Mechanical Engineers BFN BWR Browns Ferry Nuclear Plant boiling water reactor CAP corrective action program CASS Cast Austenitic Stainless Steel CFR Code of Federal Regulations CLB current licensing basis CR change request CUF cumulative usage factor CUFen environmentally adjusted cumulative usage factor EAF Environmentally Assisted Fatigue EFPY effective full-power years EOI end of interval EPRI Electric Power Research Institute FAC Flow-Accelerated Corrosion FE Further Evaluation FSAR final safety analysis report GALL-SLR NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal GE General Electric HELB high energy line break HPCI High-Pressure Coolant Injection IASCC irradiation assisted stress corrosion cracking IPA Integrated Plant Assessment ISG interim staff guidance ISI inservice inspection ISP Integrated Surveillance Program kV kilovolts LR license renewal
x NEI Nuclear Energy Institute NRC U.S. Nuclear Regulatory Commission OE operating experience PBD program basis document P-T pressure-temperature PWR pressurized-water reactor RAI request for additional information RCI request for confirmation of information RCP RHR reactor coolant pressure residual heat removal RHRSW residual heat removal service water RPV reactor pressure vessel RTMAX maximum reference temperature RTNDT reference temperature nil ductility RV reactor vessel RVI reactor vessel internals SBO station black out SE safety evaluation SLR Subsequent License Renewal SLRA subsequent license renewal application SPEO SRV Subsequent period of extended operation Safety Relief Valves SSCs systems, structures, and components TLAA Time Limited Aging Analysis TGS Turbine Generator System USAR Updated Safety Analysis Report USE upper-shelf energy UT ultrasonic testing VT visual examination WO work order WOL weld overlay
1 Report for the Aging Management Audit Browns Ferry Nuclear Plant, Units 1, 2, and 3 Subsequent License Renewal Application
- 1. Introduction The U.S. Nuclear Regulatory Commission (NRC) staff conducted an aging management audit of Tennessee Valley Authority (TVA, the applicant) of (1) plant specific operating experience (OE),
(2) methodology to identify the systems, structures, and components (SSCs) to be included within the scope of license renewal (LR) and subject to an aging management review (AMR)
(Scoping and Screening Portion), and (3) aging management programs (AMPs), AMR items, Time Limited Aging Analyses (TLAAs), and associated bases and documentation as applicable (AMP and TLAA Portion) for the subsequent license renewal (SLR) of Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 for Browns Ferry Nuclear Plant (BFN),
Units 1, 2, and 3, respectively.
The purpose of the plant specific OE portion of the audit is to identify examples of age-related degradation, as documented in the applicants corrective action program (CAP) database, and to provide a basis for the staffs conclusions on the ability of the applicants proposed AMPs to manage the effects of aging in the period of extended operation. TVA searched their OE database and provided the results for the associated AMPs and TLAAs for NRC staff review.
Additional word searches were performed by TVA upon the NRC staffs request, and the results were provided to the NRC staff for review.
The purpose of the Scoping and Screening portion of the audit is to evaluate the scoping and screening process as documented in the SLR application, implementing procedures, reports, and drawings, such that the NRC staff:
Obtain an understanding of the process used to identify the SSCs within the scope of LR and to identify the structures and components subject to an AMR; and Have sufficient docketed information to allow the staff to reach a conclusion on the adequacy of the scoping and screening methodology as documented and applied.
The purpose of the AMP and TLAA Portion of the audit is to:
Examine TVAs AMPs, AMR items, and TLAAs for BFN; Verify TVAs claims of consistency with the corresponding NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, issued in July 2017, AMPs, and AMR items; and Assess the adequacy of the TLAAs.
Enhancements and exceptions will be evaluated on a case-by-case basis. The NRC staffs review of enhancements and exceptions will be documented in the safety evaluation (SE).
The regulatory basis for the audit was Title 10 of the Code of Federal Regulations (10 CFR) part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants. The staff also considered the guidance contained in NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR), dated July 2017, and NUREG-2191. The SRP-SLR allows an applicant to reference in its SLR application the AMPs described in the GALL-SLR Report. By referencing the GALL-SLR Report
2 AMPs, the applicant concludes that its AMPs correspond to those AMPs reviewed and approved in the GALL-SLR Report and that no further staff review is required. If an applicant credits an AMP for being consistent with a GALL-SLR Report program, it is incumbent on the applicant to ensure that the plant program contains-all of the elements of the referenced GALL-SLR Report program. The applicant should document this determination in an auditable form and maintain-the documentation onsite.
- 2. Audit Activities A regulatory audit is a planned, license-related activity that includes the examination and evaluation of primarily non-docketed information. A regulatory audit is conducted to gain greater understanding of an application, to verify information, and, if applicable, to identify information that will require docketing to support the staffs conclusions that form the basis of the licensing or regulatory decision.
The subsequent license renewal application (SLRA) states that every AMP in the SLRA is consistent with the program elements of the GALL-SLR Report. To verify this claim of consistency, the staff audited each AMP, including any enhancements or exceptions associated with an AMP.
The SLRA discusses each TLAA, the disposition of the TLAA in accordance with 10 CFR 54.21(c)(1), and the basis for that disposition. To verify that the applicant provided a basis to support its disposition of the TLAA, the staff audited each TLAA.
The staff also audited AMR items not associated with an AMP to determine if the information in the SLRA is consistent with the further evaluation (FE) information in the SRP-SLR.
Furthermore, the staff audited the final safety analysis report (FSAR) supplement descriptions for each AMP and TLAA for consistency with the SRP-SLR. During its audit, the staff interviewed the applicants staff and reviewed documentation contained in the SLRA and provided by the applicant via the ePortal.
For the OE review, the applicant made a presentation on the process used to identify and evaluate the pertinent OE. Afterwards, the staff conducted its review of the applicants methodology and OE by reviewing documentation contained in the SLRA and ePortal.
From April 29 to May 3, 2024, the staff participated in an onsite audit at BFN in Athens, AL to gain a general overview of current conditions of the structures compared to the provided OE.
While onsite, the staff engaged with the applicant staff, conducted walkdowns, and reviewed documents provided by the applicant.
For the OE review, the staff conducted its review of the applicants methodology and OE by reviewing documentation contained in the SLRA and ePortal.
Licensing conclusions or staff findings are not made in the audit reports since licensing and regulatory decisions cannot be made solely based on an audit. Therefore, items identified but not resolved within the scope of the audit will be followed using other NRC processes, such as requests for additional information (RAIs), requests for confirmation of information (RCIs), and public meetings. Licensing conclusions, staff findings, staff review of enhancement and exceptions, and resolution of audit items will be documented in the staffs SE. The following sections discuss the SLRA areas reviewed by the staff.
8 BWR Vessel Inside Diameter Attachment Welds Program 0-TI-592 Aging Management Program Basis Document - BWR Vessel ID Attachment Welds, Revision 1 6/24/2014 EPRI Report 3002018321 BWRVIP-48, Revision 2, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines 9/2021 CRP-ENG-FSA-15-003 Final Report - BFN RPVII Program Focused Self-Assessment 5/27/2016 1-TI-365 Unit 1 Reactor Pressure Vessel Internals Inspection (RPVII)
Revision 0030 12/05/2022 During the audit, the staff verified the applicants claim that the preventive actions, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP. The staff also verified the applicants claim that aspects of the scope of program, parameters monitored or inspected, detection of aging effects and monitoring and trending, program elements not associated with the exception identified in the SLRA are consistent with the corresponding program elements in the GALL-SLR Report AMP.
During the audit, the staff made the following observations:
The SLRA describes an AMP consistent with the AMP specified in the GALL-SLR Report, with one exception. The BFN BWR Vessel ID Attachment Welds AMP is based on the inspection, evaluation, and repair guidelines contained in BWRVIP-48 Revision 2.
o Use of BWRVIP-48, Revision 2, rather than BWRVIP-48-A, is an exception from the guidance specified in the GALL-SLR Report. The applicant asserts that Revision 2 of the report is not less conservative than the guidance in the staff-approved report.
o The AMP described in GALL-SLR Report also assumes compliance with the requirements of ASME Code,Section XI. According to the AMP basis document, SLR-BFN-0034, Revision 1 (Dec. 2023), the BFN program substitutes the inspection and evaluation recommendations within BWRVIP-48 Revision 2, BWR Vessel and Internals Project, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines, for the requirements of ASME Code,Section XI, Table IWB-2500-1, Examination Category B-N-2. The program elements affected are scope of the program, parameters monitored or inspected, detection of aging effects, and monitoring and trending. Although the licensee currently has a staff-approved alternative under 10 CFR 50.55a(z) to use the guidance in BWRVIP-48, Revision 2 in place of ASME Code,Section XI requirements for inservice inspection, sufficient justification has not been docketed to establish this report as adequate to manage aging effects through the SLR period of extended operation. The staff will consider issuing RAIs to obtain the necessary information.
OE reports included no instances of stress corrosion cracking in BWR ID attachment welds.
12 BWRVIP-97 Guidelines for Performing Weld Repairs to Irradiated BWR Internals Revision 1 GEH Report No.
007N4785 TVA BFNS Recommendations for Future Inspections
- Replacement Steam Dryer Revision 0 November 2022 BWRVIP-181 Steam Dryer Repair Design Criteria Revision 2 BWRVIP-315 Reactor Internals Aging Management Evaluation for Extended Operations Revision 0 ML22025A113 EPRI, Response to BWRVIP-315 Draft Safety Evaluation January 20, 2022 ML23251A069 Final Safety Evaluation By the Office of Nuclear Reactor Regulation Topical Report BWRVIP-315 Boiling Water Reactor Vessel and Internals Project:
Reactor Internals Aging Management Evaluation for Extended Operations October 31, 2023 0-TPP-ENG-467 Browns Ferry Nuclear Plant Technical Program Procedures: Inservice Inspection Program Revision 0001 Structural Integrity Associates (SIA)
Report 2200107.406P Stress Relaxation of Core Spray and Jet Pump Component Hardware for Subsequent License Renewal February 23, 2023 BWRVIP-139 Steam Dryer Inspection and Flaw Evaluation Guidelines Revision 1-A BWRVIP-180 Access Hole Cover Inspection and Flaw Evaluation Guidelines Revision 1 BWRVIP Letter 2021-030 (ML21084A164)
Potential Non-Conservatism in EPRI Report, BWRVIP-100, Rev. 1-A, 3002008388 and Impacted BWRVIP Reports March 22, 2021 Condition Report 1672927 10 CFR Part 21 - Transfer of Information Notice -
Potential Non-Conservatism in EPRI Software November 30, 2022 (Date Closed)
During the audit, the staff verified the applicants claim that the preventive actions program element of the SLRA AMP is consistent with the corresponding element of the GALL-SLR Report AMP. The staff also verified the applicants claim that aspects of the scope of program, parameters monitored or inspected, detection of aging effects, monitoring and trending, and acceptance criteria program elements not associated with the exceptions identified in the SLRA or by the staff during the audit will be consistent with the corresponding program elements in the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
In addition, the staff found that for the corrective actions program element, sufficient information associated with Exception 6 and Enhancement 6 was not available to determine whether it was consistent with the corresponding program element of the GALL-SLR Report AMP. The staff will consider issuing RAIs to obtain the necessary information.
13 During the audit, the staff made the following observations:
The staff reviewed and verified that BWRVIP-139 Revision 1-A does not apply to the replacement steam dryers installed at BFN, Units 1, 2, and 3 as noted in Exception 1.
The staff reviewed the guidance in GEH Report No. 007N4785 which forms the basis for managing age-related degradation of the replacement steam dryers, along with BWRVIP-181 Revision 2 which the applicant cites for reference when preparing a repair plan for the replacements steam dryers in lieu of BWRVIP-181-A that is cited in the GALL-SLR Report AMP. The staff verified that GEH Report No. 007N4785 is the basis for the long term inspection plan sent to the NRC to satisfy Unit 1 Operating License Condition 2.C(18)(i), Unit 2 Operating License Condition 2.C(18)(i) and Unit 3 Operating License Condition 2.c(14)(g) to provide guidance for the aging management of the replacement steam dryers.
The staff reviewed the eight BWRVIP reports discussed in Exception 2 and verified the following:
BWRVIP-03 Revision 20 provides standards for demonstration of nondestructive evaluation techniques and does not change any component specific technical criteria that would impact aging management.
BWRVIP-62 Revision 2 contains the primary and secondary parameters for monitoring water chemistry that are consistent with the parameters cited in BWRVIP-62-A.
BWRVIP-180 Revision 1 contains guidance for inspecting and evaluating access hole covers consistent with the guidance provided in BWRVIP-180 Revision 0 which is cited in the GALL-SLR AMP.
The staff reviewed Exception 4 to verify that the applicant intends to use BWRVIP-100 Revision 2 in lieu of the GALL-SLR AMP guidance to use BWRVIP-100-A. The staff noted that the applicant addressed the non-conservatism reported in BWRVIP-100 Revision 1-A by BWRVIP Letter 2021-030 (ML21084A164) via Condition Report 1672927 which describes the applicants plant-specific response to BWRVIP Letter 2021-030. However, the staff noted several instances in the application where BWRVIP-100-A was cited in lieu of BWRVIP-100 Revision 2. The staff will consider issuing an RAI to confirm that the applicant intended to use BWRVIP-100 Revision 2 in all places where BWRVIP-100-A was cited.
The staff reviewed Exception 5 and verified that BWRVIP-97 Revision 1 provides guidance for performing weld repairs to irradiated internals that are consistent with, or more conservative, than the guidance provided in BWRVIP-97-A, which is cited by the GALL-SLR AMP.
The staff reviewed Exception 6 and found that sufficient information was not available to determine if BWRVIP-84 Revision 3 sufficiently manages the repair and replacement of aging reactor vessel internals as discussed in the corrective actions element of the GALL-SLR AMP guidance. Additionally, the staff reviewed Enhancement 6 and found insufficient information to determine if BWRVIP-84 Revision 3 provides sufficient guidance regarding the repair and replacement of the components covered by the BWR Vessel Internals AMP.
The staff reviewed the discussion regarding BWRVIP-315 and noted that the application provides responses to the two license conditions and four applicant action items
16 CR 468792 - CR 315818 Apparent Cause Evaluation Bypass Flow in 2C RHR Heat Exchanger Lower Tier Apparent Cause 3/4/2011 CR 468792 - WO 90-721841-000 Disassembly, clean heat exchanger BFR-1-HEX-074-0900A, U1C8 8/3/2010 CR948584 Replace gamma (RT) plug due to hole in weld 10/22/2014 CR959675 Minium Wall issues on RHRSW piping at IPS 11/21/2014 CR1028757 Minimum wall thickness reading on B2 RHRSW piping 5/21/2015 CR 1112822 Program Engr FAC inspection on the 2C3 FWH shell found water spitting out of a shell hole 12/8/2015 CR 1112822 - WO 117410109 Perform Eddy current test. FAC inspection on 2C3 FWH shell found water spitting out of the shell hole 12/8/2015 CR 1294482 Replace Section of 4 steam drain piping before it enters the A condenser during U2R20 RFO 5/11/2017 CR 1294482 - WO 118730362 Replace Section of 4 steam drain piping before it enters the A condenser during U2R20 RFO 7/17/2018 CR 1342473 Tube thinning in 3B DG Coolers 9/28/2017 CR 1342473 - WO 118125762 Eddy current test the coolers 3B1 and 3B2 9/25/2017 CR 1342473 - WO 119145468 Replace 3B1 and 3B2 DG Heat Exchangers 5/13/2019 CR 1591931 3A3 LP Feedwater Heater (FWHTR) FAC inspection near location of 3 sheared tubes 3/3/2020 CR 1591931 - WO 121155847 Perform FAC inspection on 3A3 LP FWHTR and 3A3 Heater 3/7/2020 CR 1537212 OE review of Susquehanna - wall thinning identified in bottom head drain line 8/1/2019 OE Review of Susquehanna BFN Reactor Bottom head drain piping and flow-accelerated corrosion review 8/1/2019 CR 1769308 Replace unit 3, 4 elbow in the moisture separator room during U3R21 outage 4/13/2022 CR 1786005 Replace the 4 Alt Leak Path 3MSZ-HDRDRVNT-06E elbow during the U3R21 outage 6/27/2022 CR 1786005 WO 23014359 Replace the 4 Alt Leak Path 3MSZ-HDRDRVNT-06E elbow during the U3R21 outage and the 12 of upstream and downstream pipe 3/10/2024 CR 1808953 U1R15 1MSZ-HDRDR-19E proactive replacement 10/12/2022 0-TI-140_003947906 Aging Management Program Basis Document - Flow-Accelerated Corrosion Revision 1 NPG-SPP-09.7.2 NPG Standard Program and Processes Revision 0005 DS-M4.2.1_
PROCEDURE _
003680051 Engineering Design Guide/Standards Flow-Accelerated Corrosion Program Methods Revision 0010
17 17-0291-TR-006 BFN1 SSE Updates Full Report Flow-Accelerated Corrosion Program - System Susceptibility Evaluation Unit 1 Revision 2 17-0291-TR-011 BFN2 SSE Updates Full Report Flow-Accelerated Corrosion Program - System Susceptibility Evaluation Unit 2 Revision 2 BFN-3-FAC-SSE ENG DOC BFN-3 System Susceptibility Evaluation for Flow-Accelerated Corrosion Revision 0 17-0291-TR-007 BFN1 SNM Updates Full Report Flow-Accelerated Corrosion Program -
Susceptible Non-Modeled Analysis Unit 1 Revision 2 17-0291-TR-012 BFN2 SNM Updates Full Report Flow-Accelerated Corrosion Program -
Susceptible Non-Modeled Analysis Unit 2 Revision 2 BFN-3-FAC-SNM BFN-3 SUSCEPTIBLE NON-MODELED (SNM)
ANALYSIS FOR FAC BFN-3 Susceptible Non-Modeled Analysis for Flow-Accelerated Corrosion Revision 0 U1 REACTOR DRAIN LINE INSPECTION -
BOP-38 Inspection report from 2004 12/11/2004 During the audit, the staff verified the applicants claim that the preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
In addition, the staff found that for the scope of program program element, sufficient information was not available to determine whether it is consistent with the corresponding program element of the GALL-SLR Report AMP. The staff will consider issuing RAIs in order to obtain the information necessary to verify the program element is consistent with the corresponding program element of the GALL-SLR Report AMP.
During the audit, the staff made the following observation:
The BFN Flow-Accelerated Corrosion Program System Susceptibility Evaluation reports use System Code "UN" and System Name "Unspecified," which are database artifacts from the vendor and the applicant does not use them.
The staff also audited the description of the SLRA Flow-Accelerated Corrosion program provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
24 CR1770355 Damaged Albi Clad April 19, 2022 WO 115813466 Albi Clad missing March 2015 WO 122900665 Damaged Thermolag due to water exposure September 2022 WO 115813466 Albi Clad fire proofing insulation is missing from structure beam in ceiling March 2015 SLR-BFN-0060 Aging Management Program Basis Document - Fire Protection Revision 2 0-SI-4.11.D.1.b Control Bay CO2 System Functional Test Revision 0028 SSP-9.3 Plant Modifications and Design Change Control June 2, 1994 1/2-SI-4.11.D.1.b Unit 1 & 2 Diesel Generator Building CO2 System Functional Test Revision 0031 3-SI-4.11.D.1.b Unit 3 Diesel Generator Building CO2 System Functional Test Revision 0033 0-SI-4.11.G.1.A Visual Inspection of Risk Significant Fire Rated Barriers (Floors, Walls & Ceiling)
Revision 0029 0-SI-4.11.G.1.a(1)
Visual Inspection of Electrical Raceway Fire Barrier Systems (ERFBS)
Revision 0018 0-SI-4.11.G.1.a(2)
Increased Frequency Visual Inspection of Electrical Raceway Fire Barrier Systems (ERFBS)
Revision 0000 0-SI-4.11.G.1.A(b)
Visual Inspection of Non-Risk Significant Fire Rated Barriers (Floors, Walls & Ceiling)
Revision 0000 0-SI.4.11.G.1.A Visual Inspection of Fire Rated Barriers (Unit 1 Steam Tunnel Floors, Walls & Ceiling)
Revision 004 2-SI.4.11.G.1.a Visual Inspection of Fire Rated Barriers (Unit 2 Steam Tunnel Floors, Walls & Ceilings)
Revision 0007 3-SI.4.11.G.1.a Visual Inspection of Fire Rated Barriers (Unit 3 Steam Tunnel Floors, Walls & Ceilings)
Revision 0007 0-SI-4.11.G.1.c(2)
Visual Inspection of Cable Tray Penetrations in Fire Rated Barriers and Cable Tray Covers Revision 027 0-SI-4.11.G.1.b(1)
Visual Inspection of First Period Fire Dampers Revision 0017 0-SI-4.11.G.1.b(2)
Visual Inspection of Second Period Fire Dampers Revision 0020 0-SI-4.11.G.1.b(4)
Visual Inspection of Fourth Period Fire Dampers Revision 0019 0-SI-4.11.G.1.b(5)
Visual Inspection of Fifth Period Fire Dampers Revision 0013 0-SI-4.11.G.1.b(6)
Visual Inspection of Third Period Fire Dampers Revision 0013 0-SI-4.11.G.2 Semiannual Fire Door Inspection Revision 0032 0-SI-4.11.G.2.A Monthly Functional Test of Fire Door Supervision Circuits Revision 0017 0-SI-4.11.G.2.b Fire Door Inspection Revision 0026 0-SI-4.11.G.2.e Inspection of YARD / SWITCH Interface Doors Revision 0002
26 0-SI-4.11.E.1.A Outside Fire Hose Replacement Revision 0021 0-SI-4.11.E.1.b(1)
Fire Hose Station Functionality/Flow Test Revision 0017 0-SI-4.11.E.1.B(2)
Safety Related Fire Hose Replacement Revision 0015 0-SI-4.11.F.1.a Safe Shutdown Fire Hydrant Inspection, Flush and Flow Check Revision 0014 FP-0-026-INS002 Fire Hose Hydrostatic Test Revision 0013 FP-0-026-INS003 Inside Loop Header Flushes and Fire Hydrant Inspection, Lubrication and Flush Revision 0028 CR-382514 Fire hose station out of service 6/6/2011 Wo112341021 Fire hose station out of service 7/22/2011 CR-347005 Fire protection piping leak 3/31/2011 CR-384687 Claim shells in fire protection system 6/7/2011 PER 384675 Apparent cause evaluation report - clams in fire protection system 7/29/2011 CR-391395 Broken drain line 6/22/2011 CR-433961 Clogged bearing lube water strainers 9/16/2011 WO 112385345 Disassemble and clean bearing lube water strainer 9/28/2011 CR-437195 Through wall leak at a fire protection header elbow 9/24/2011 WO 112744671 Pin hole piping leak in fire protection header at an elbow 2/2/2012 CR-468986 Thru wall corrosion on fire protection piping upstream of hose station valve 3/20/2015 CR-508930 Fire header pipe break 2/21/2012 CR-797148 Hydrant failures on inside loop 10/22/2013 CR-1082715 Clogged sprinkler piping 9/16/2015 WO 117168765 Clogged sprinkler piping 8/8/2018 CR-1140288 Clogged spray nozzles 2/21/2016 WO 118246142 2-year functional test of transformer sprinkler system 4/10/2017 CR-1283472 12 clogged sprinkler heads 4/12/2017 CR-1495117 Clogged deluge system nozzles 3/3/2019 CR-1707469 Water seeping through rusted spot on bottom of pipe 7/13/2021 WO 122247414 Leak on fire protection piping 10/23/2023 CR-1770323 Leak on 4 fire header 4/18/2022 WO 117361299 Repair/replace 12 fire protection piping 11/11/2016 CR-1774216 Three underground leaks on fire protection piping outside loop 5/5/2022 CR-1794800 6 fire protection pipe low wall thickness readings 8/8/2022 CR-307631 Pin hole leak in 4 Fire protection piping 1/6/2011
27 WO 111481963 Replace 4 and 2.5 piping with pin holes 4/2011 CR-1233411 Replace Channel Diesel Fire Pump Cooling Reservoir 11/16/2016 CR-1233430 Channel Diesel Fire Pump Failed to Start 11/16/2016 CR-1235141 Channel Diesel Fire Pump Cooling Pump Leak 11/16/2016 WO 118322106 Walkdown to Document Channel Diesel Fire Pump Coolant Leak 12/2016 CR 1234856 Functionality Evaluation for Channel Diesel Fire Pump 11/22/2016 WO 124062752 Repair Channel Diesel Fire Pump Coolant Leak 10/2023 WO 118309899 Replace Channel Diesel fire Pump Engine Coolant Pump 12/2016 WO 118317776 Replace Coolant Fill Neck on Channel Diesel Fire Pump 11/2016 CR-1887411 Channel Diesel Fire Pump Coolant Leak 10/19/2023 During the audit, the staff verified the applicants claim that the preventive actions, detection of aging effects, monitoring and trending, acceptance, and corrective actions program elements of the SLRA AMP were consistent with the corresponding elements of the GALL-SLR Report AMP or will be consistent after implementation of the identified enhancements.
In addition, the staff found that for the scope of program and parameters monitored or inspected, program elements, sufficient information was not available to determine whether they are consistent with the corresponding program elements of the GALL-SLR Report AMP.
The staff will consider issuing RAIs in order to obtain the information necessary to verify whether these program elements are consistent with the corresponding program elements of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA Fire Water System program provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B.2.1.17, Outdoor and Large Metallic Storage Tanks Program Summary of Information in the Application The SLRA states that AMP B.2.1.17, Outdoor and Large Metallic Storage Tanks Program, is an existing program that with enhancements and exception will be consistent with the program elements in GALL-SLR Report AMP XI.M29, Outdoor and Large Metallic Storage Tanks. At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this enhanced program, and the staffs audit addressed only the program elements described in the applicants basis document.
Audit Activities The table below lists the documents that were reviewed by the staff and were found relevant to the program. The staff will document its review of this information in the SE.
33 SLR-BFN-0082 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Subsequent License Renewal:
License Renewal Aging Management Program Implementation Effectiveness Report - AMP O.1.4, ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program Revision 0 09/2023 SLR-BFN-0067 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Subsequent License Renewal:
License Renewal Aging Management Program Basis Document - ASME Code Class 1 Small-Bore Piping Revision 1 06/2023 CR1277160 Through Wall Leak on Instrumentation Line RTV-001-0251A During Unit 2 Hydro 03/26/2017 SLR-BFN-0072 Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Subsequent License Renewal:
Quality Assurance, Administrative Controls, and Operating Experience Input to Subsequent License Renewal Application Revision 1 06/2023 W47 200828 001 TVA Inspection and Examination Procedure:
Guideline for Conducting Ultrasonic Examinations of Dissimilar Metal Welds Revision 4 08/31/2020 PER157918 Unit 1 ASME Code Class Leak in the Base Metal of N-11B Instrumentation Nozzle Safe End 11/23/2008 CR1106738 Leak Discovered on Unit 1 A Recirc Seal Purge Line Socket Weld Union 11/20/2015 CR1644945 During UT Examination (R-057) of the N10-1 Safe End to Pipe Weld a Circumferential Indication Was Identified 10/15/2020 PER171239 Planar Flaw Exceeded Acceptance Criteria, Unit 2 Instrumentation Nozzle N12A Safe End to Pipe Weld 05/13/2009 PER130777 Pressure Boundary and Code Class Piping Leakage 09/22/2007 PER916833 Unit 2 RHR System Leak Identified Down Stream of 2-FCV-074-0048 08/03/2014 CR550940 Apparent Cracked Weld at 3-TV-74-639A 05/13/2012 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP.
The staff also audited the description of the SLRA ASME Code Class 1 Small-Bore Piping program provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
38 BP-2023-0027-03-TR Browns Ferry License Renewal Buried Piping Cathodic Protection Review Revision 3 16342-FOR-01-1 Browns Ferry Nuclear Plant Corrosion Risk and Condition Assessment Considerations 08/25/2023 SLR-BFN-0105 License Renewal Aging Management Program Implementation Effectiveness Report - AMP O.1.28 -
Buried Piping and Tanks Inspection Program Revision 0 CR 793555 Underground Air Leak Near Intake 10/15/2013 CR 705057 Stds Team Obs. Outside U1&2 diesel generators area. Southside on the grounds steam leak 04/01/2013 CR 1120048 Possible fire header leader leak between the intake and 0-26-528 valve 12/30/2015 CR 681335 Air Leak Causing Sink Hole at Base of Berm on East Side of Plant 02/14/2013 CR 743618 Air leak in yard creating safety concern 06/20/2013 CR 898407 Degraded Service Air Piping 06/13/2014 CR 589483 Service Air Piping Leak 08/02/2012 CR 828934 100+ GPM leak in FP system identified at cooling tower 2 01/06/2014 The staff found that for the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements, sufficient information was not available to determine whether they were consistent with the corresponding program elements of the GALL-SLR Report AMP. The staff will consider issuing RAIs in order to obtain the information necessary to verify whether these program elements are consistent with the corresponding program elements of the GALL-SLR Report AMP.
During the audit, the staff made the following observations:
The staff reviewed CR 793555 and noted the suspected cause of the leak was corrosion similar to that seen in two service air piping leaks earlier this year.
The staff reviewed CR 1120048 and noted the source of the leak was determined to be service air and not fire protection.
The staff reviewed CR 898407 and noted (a) a section of service air piping had no visible exterior tape coating or polyethylene coating as required by design drawings; and (b) the uncoated section of piping had multiple thru wall holes.
The staff reviewed CR 828934 and noted it was indeterminate whether the leak was coming from a slipjoint between sections of pipe or a hole in the pipe wall.
The staff also audited the description of the SLRA Buried and Underground Piping and Tanks program provided in the FSAR supplement. The staff found that sufficient information was not available to determine whether the description provided in the FSAR supplement was an adequate description of the SLRA Buried and Underground Piping and Tanks program. The staff will consider issuing RAIs in order to obtain the information necessary to verify the
42 NPG-SPP-09.1.3 ASME Section XI Repair/Replacement Activities Program Revision 6 NPG-SPP-7.0 Work Management Revision 4 N-VT-1 Visual Examination Procedure for ASME Section XI Pre-service and Inservice Revision 49 N-VT-25 Visual Examination of IWE Class MC Components Revision 1 MAI-5.2 Bolting and Structural Connections Revision 18 SLR-BFN-0107 LR AMP Implementation Effectiveness Report - AMP O.1.30 ASME Section XI Subsection IWF Program Revision 1 SSA1805 Site Audit Report SSA1805 Engineering Programs BFNP 6/18/18-6/29/18 N/A SSA1606 Site Audit Report SSA1606 Engineering Programs BFNP 5/31/16-6/10/16 N/A CR 944463 NOI U1R10-001: Support 1-47B452S0183 Loose Structural Bolting 10/12/2014 CR 1676573 NOI U2R21-001, variable Spring Setting out of range 3/5/2021 CR 1759027 BFN NOI U3R20-001 3-47B455-629 3/3/2022 CR 946497 NOI U1R10-004: HPCI Support 1-47B455-2134 Loose Bolting and Displacement 10/16/2014 WO 116239248 Tighten Bolt(s) listed on NOI-001 per MAI-5.1A 10/19/2014 WO 116248564 NOI U1R10-004: HPCI Support 1-47B455-2134 Loose Bolting and Displacement 10/20/2014 Drawing 2-47B452-1325-1/2 Mechanical Residual Heat Removal System Pipe Support Revision 0 PM 37817 Perform a visual inspection for structural integrity and evidence of loss of material due to corrosion 4/10/2024 BFN-0-37W205-5 Mechanical Pumping Station & Water Treatment -
Piping & Equipment Revision 17 BFN-0-37W205-10-2 Seismic Restraint for RHRSW Pumps Option A Sheet 1
Revision 6 BFN-0-37W205-10-3 Seismic Restraint for RHRSW Pumps Option A Sheet 2
Revision 6 BFN-0-37W205-10-6 Seismic Restraint for RHRSW Pump B3 Option A Sheet 1 Revision 1 BFN-0-37W205-10-7 Seismic Restraint for RHRSW Pump B3 Option A Sheet 2 Revision 1 BFN-1-47E858-1-ISI ASME Section XI RHR Service Water System Code Class Boundaries Revision 36 NOI U1R10-001 U1R10 NOI for Pipe Support 1-47B452S0183 10/24/2014 NOI U1R10-004 U1R10 NOI for Pipe Support 1-47B455-2134 10/23/2014 NOI U2C15-030 U2C15 NOI for Pipe Support 2-47B452S0234 5/21/2009
43 NOI U2RF16-007 U2RF16 NOI for Pipe Support 2-47B400S00002 3/26/2011 NOI U1C7-008 U1C7 NOI for Pipe Support 1-47B400-2035 11/9/2008 CR 1005978 U2 Existing Spring Can Setting Does Not Meet Design Output Drawing 8/17/2015 CR 946069 Dual Spring Support 1-47B401 Loose Bolt on Riser Clamp 10/19/2014 CR 704155 Loose Bolts 4/1/2013 During the audit, the staff verified the applicants claim that the parameters monitored or inspected, monitoring and trending, and corrective actions program elements of the SLRA AMP will be consistent with the corresponding elements of the GALL-SLR Report AMP after implementation of the identified enhancements. In addition, the staff found that for scope of program, preventive actions, detection of aging effects, and acceptance criteria, program elements, sufficient information was not available to determine whether they were consistent with the corresponding program elements of the GALL-SLR Report AMP. The staff will consider issuing RAIs in order to obtain the information necessary to verify whether these program elements are consistent with the corresponding program elements of the GALL-SLR Report AMP.
During the audit, the staff made the following observations:
During the review of SLRA B.2.1.30 Program Description, the staff noted that the SLRA stated the program will also perform inspections of the seismic restraints in the RHRSW pump pit. The staff further noted that the inspection of the seismic restraints in the RHRSW pump pits is one of the initial LR commitments and it can be inferred that the applicant plans to continue the LR commitment made for the PEO into the SPEO.
However, SLRA AMP B.2.1.30 does not discuss such commitment and the reason of changing the AMP to manage the aging effect of the seismic restraints in the RHRSW pump pits.
During the review of SLRA Table 3.5.2-2, the staff noted that Molykote@321 or equal is identified as a material for sliding surfaces subject to loss of mechanical function aging effect. The staff also noted that Molykote D-321 is not a long-lived material having a usable life of 24 months from the date of production. It is not clear whether 10 CFR Part 54 provides the regulatory framework for managing effects of aging on the lubricant having a usable life of 24 months.
During the review of SLRA B.2.1.30, the staff noted that it does not appear to discuss the reactor vessel (RV) supports, stabilizer, and star truss. Specifically, there is no discussion on their condition and Industry OE as it is applied to BFN.
During the review of the OE, the staff noted that following the identification of the loose bolts the issues were entered into the CAP database and engineering evaluations were performed. It is not clear whether the safety factor consistent with BFN current licensing basis (CLB) was used in the evaluation to determine the safety margin.
During the audit meeting, the applicant claimed that in addition to ASME Section XI, Subsection IWF program they also use other inspection programs and processes to inspect the conditions of IWF component supports.
45 U1R13 LLRT Summary Spreadsheet 4/4/2024 U2R21 LLRT Summary Spreadsheet 4/4/2024 U3R20 LLRT Summary Spreadsheet 4/4/2024 During the audit, the staff verified the applicants claim that the preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP. In addition, the staff found that for the scope of program program element, sufficient information was not available to verify whether it was consistent with the corresponding program element of the GALL-SLR Report AMP. The staff will consider issuing RAIs to obtain the information necessary to verify whether this program element is consistent with the corresponding program element of the GALL-SLR Report AMP.
During the audit, the staff made the following observation:
The staff reviewed SLRA B.2.1.31 and its Program Basis Document (PBD) TVA SLR-BFN-0042, and noted that Section 3.1.2 of PBD states, in part: Components required to be leak rate tested or excluded from testing are identified in UFSAR Table 5.2-2. The one-time inspection (SLR-BFN-0065) program, water chemistry (SLR-BFN-0040) program, inspection of internal surfaces in miscellaneous piping and ducting components (SLR-BFN-0045) program, external surfaces monitoring of mechanical components SLR-BFN-044) program, BWR stress corrosion cracking (SLR-BFN-0041) program, ASME Section XI inservice inspection, Subsection IWB, IWC, and IWD (SLR-BFN-0032) program and the Flow-Accelerated Corrosion (SLR-BFN-0063) program are included among the aging management programs that manage the aging effects associated with components excluded from leak rate testing. While SLRA Section B.2.1.31 states, in part: Components required to be leak rate tested or excluded from testing are identified in FSAR Table 5.2-2., it does not include or state the AMPs, as identified in the PBD, that will be used to manage the aging effects of components excluded from testing. The GALL-SLR Report AMP XI.S4 states, The aging effects associated with containment pressure-retaining boundary components within the scope of subsequent license renewal and excluded from Type B or C Appendix J testing must still be managed. Other programs may be credited for managing the aging effects associated with these components; however, the component and the proposed AMP should be clearly identified. It is not clear for those components (valves, penetrations, and others) that have been excluded from 10 CFR Part 50 Appendix J program whether they are included in the scope and which AMPs will be used to manage the aging effects for each of the excluded components.
The staff also audited the description of the SLRA 10 CFR Part 50, Appendix J program provided in the FSAR supplement. The staff verified that this description is consistent with the description provided in the GALL-SLR Report.
47 SLR-BFN-0029 Aging Management Program Basis Documents -
Structures Monitoring Program Project Task Report SLR-BFN-0110 License Renewal Aging Management Program Implementation Effectiveness Report - AMP O.1.33, Structures Monitoring Program Revision 0 Project Task Report SLR-BFN-009 Tennessee Valley Authority, Browns Ferry Nuclear Plant Units 1, 2, and 3 Subsequent License Renewal, Screening/Aging Management Review -
Containments, Structures and Component Supports Revision 01 0-TI-346 Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting - 10CFR 50.65 Revision 55 CI-421 Browns Ferry Nuclear Plant Chemistry Instruction, Wells Sampling and Maintenance Revision 21 LCEI-CI-09 Tennessee Valley Authority Browns Ferry Nuclear Plant Lead Civil Engineer Instruction, Procedure for walkdown of Structures for Maintenance Rule and License Renewal Revision 11 0-TI-651 Browns Ferry Nuclear Plant Technical Instruction, License Renewal Screening/Aging Management Review Reports Revision 01 G-29B - P.S.4.M.4.4 ASME Section III and Non-ASME Section III (Including AISC, ANSI B31.1, and ANSI 31.5) Bolting Material Revision 07 MAI-5.2 Browns Ferry Nuclear Plant Modification and Addition Instruction, Bolting and Structural Connections Revision 18 G-2 General Engineering Specification G-2 for Plain and Reinforced Concrete Revision 08 G-66 General Engineering Specification G-66 for Requirements for Use of Undercut Anchors Set in Hardened Concrete During Construction, Modification, and Maintenance Revision 08 G-32 General Engineering Specification G-2 for Bolt Anchors Set in Harden Concrete Revision 26 CR 1749223 Proposed Sampling of Monitoring Wells 01/21/2022 CR 1596974 Gate Structure No. 2 Concrete Spall 03/24/2020 CR 3399927 Indications of a Through Wall Leak 03/16/2011 CR 594952 Evaluate the Radwaste Tunnel Expansion Joints for Integrity 08/14/2012 CR 1842341 Concrete Spall on Floor Elevation 557 feet, Condenser Narrow Side 03/16/2023 CR 1680436 Water Eroding Through the Wall in Bay 9 of Unit 3 Reactor Building Torus Area 03/23/2021 CR 1028902 Minor Cracks in 3C Main Bank Transfer Foundation -
BFN-3-OXF-236-0010 05/21/2015
48 CR 1887849 Unit 2 NE Quad Elevation 541 feet, Groundwater Seepage from Concrete Wall 10/23/2023 CR 653303 Unit 3 Elevation 541 feet Structural Beam Connection 12/14/2012 CR 1034064 Concrete Spalling and Scaling on Unit 3 Inside Wall of Torus Room, Bay 8 06/05/2015 CR 1839302 Large Crack in Ceiling of Moisture Separator Room 03/06/2023 CR 1187058 Unit 3 T11/B - Line Column Spall 06/29/2016 CR 1093916 Minor Damage on Grout Pad BFN-1/2/3-027-0141 10/19/2015 CR 1187552 Concrete Cracks in Off-Gas Recombiner Room Roof 06/30/2016 Calculation CDQ0003032023000000 Attachment - Groundwater Well Low pH Revision 02 During the audit, the staff verified the applicants claim that the preventive actions, monitoring and trending, and corrective actions program elements of the SLRA AMP are consistent with the corresponding elements of the GALL-SLR Report AMP. In addition, the staff found that for the scope of program, parameters monitored or inspected, detection of aging effects, and acceptance criteria program elements, sufficient information was not available to determine whether they were consistent with the corresponding program elements of the GALL-SLR Report AMP. The staff will consider issuing RAIs in order to obtain the information necessary to verify whether these program elements are consistent with the corresponding program elements of the GALL-SLR Report AMP.
During the audit, the staff made the following observations:
During the review of the aging management of the Earth Berm, the staff noted that the Earth Berm is a rock and earthfill embankment, and the Earth Berm has the following structures located within it: Equipment Access Lock, the RHRSW Tunnels, Vent Vaults, and Standby Gas Treatment Building. The Structures Monitoring AMP consists of periodic visual inspections, which cannot adequately detect aging effects of the Earth Berm without taking other measures. The staff finds that SLRA does not make clear how the Structures Monitoring AMP can adequately manage aging effects of loss of material and loss of form due to erosion for the Earth Berm during the SPEO.
During the review of test results for groundwater and raw water samples in the attachment F, Groundwater Low pH Evaluation, of the calculation CDQ0003032023000000, Revision 2, the staff noted that the pH limits are below the lower threshold limit for pH of 5.5 per NUREG-1557 at monitoring wells MW-10 and MW-11, therefore, the plant has aggressive ground water at these two locations. The staff also noted that the plant has non-aggressive groundwater at other monitoring wells. The applicant is requested to provide specific and detailed information for actions that will be taken during the SPEO to adequately address aging management of both accessible and inaccessible areas exposed to potentially aggressive groundwater/soil environment.
During the review of operating experiences described in SLRA Section B.2.1.33, the staff noted canceled or delayed work orders without sufficient explanation.
52 NPG-SPP-31.0 Records Management and Document Control Programs Revision 0007 0-TI-566 Accessible Non-Environmental Qualification Cables and Connections Inspection Program Revision 0003 PM#40329 Perform an Inspection of Accessible Non-Environmental Qualification Cables and Connections Evaluation Unit 2 System 265 N/A PM#40327 Perform an Inspection of Accessible Non-Environmental Qualification Cables and Connections Evaluation Unit 1 System 265 N/A PM#40330 Perform an Inspection of Accessible Non-Environmental Qualification Cables and Connections at a freq Evaluation Unit 3 System 265 N/A WO# 111369931 License Renewal Accessible Non EQ Cable U1R8 Outage Inspection 03/21/2013 WO# 111762782 License Renewal Accessible Non EQ Cable U2R16 Outage Inspection 02/25/2011 WO# 112222366 License Renewal Accessible Non EQ Cable U2R15 Outage Inspection 02/21/2012 During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP will be consistent with the corresponding elements of the GALL-SLR Report AMP after implementation of the identified enhancements.
During the onsite audit, the staff conducted walkdowns of the cable raceways and cables and connections associated with onsite power systems.
The staff also audited the description of the SLRA Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CR 50.49 Environmental Qualification Requirements program provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report.
SLRA AMP B.2.1.37, Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits Summary of Information in the Application The SLRA states that AMP B.2.1.37, Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits, is an existing program that with enhancements will be consistent with the program elements in GALL-SLR Report AMP XI.E2, Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits. At the time of the audit, the applicant had not yet fully developed the documents necessary to implement this enhanced program, and the
55 RIMS No: R27 090812 306 Cable, Okonite N/A During the audit, the staff verified the applicants claim that the scope of program, preventive actions, parameters monitored or inspected, detection of aging effects, monitoring and trending, acceptance criteria, and corrective actions program elements of the SLRA AMP will be consistent with the corresponding elements of the GALL-SLR Report AMP after implementation of the identified enhancements.
During the audit, the staff made the followingobservation:
The staff identified that the applicant did not mention consistency with SLR-ISG-2021-04-ELECTRICAL, Updated Aging Management Citeria for Electrical Portions of the Subsequent License Renewal Guidance, in SLRA AMP B.2.1.38. The applicant acknowledged this omission and agreed to modify the application in a future supplement.
The staff also audited the description of the SLRA Electrical Insulation for Inaccessible Medium-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements program provided in the FSAR supplement. The staff verified this description is consistent with the description provided in the GALL-SLR Report as modified by SLR-ISG-2021-04-ELECTRICAL.
During the onsite audit, the staff made the following observation during walkdowns:
During the walkdown of the manholes for inaccessible cables, the staff noted that handhole 15 (HH15) containing inaccessible cables for the RHRSW system, the cables are being supported by nylon ratchet straps. These cables had previously been exposed to excessive wetting and submergence, resulting in an electrical short to ground in 2012, and a modification performed to lift the cables above the expected water line. The staff reviewed the 2012 corrective actions taken in 119773 which included replacing the cables and modifying the cable supports to raise the cables above the expected water line. The permanent modification to lift the cables above the expected water line used nylon ratchet straps anchored to steel supports. The modification also used plastic tie wraps for securing the ratcheting mechanism of the ratchet strap. The staff requested details of the aging management program to monitor the nylon ratchet straps for age-related degradation of the strap material and purpose/function of the plastic tie wraps.
GALL-SLR section XI.E3A Evaluation and Technical Basis Item 2, Preventive Actions, directs that the aging management of the physical structures, including cable support structures of cable vaults/manholes is managed by GALL-SLR Report AMP XI.S6, Structures Monitoring. NRR/DEX/ESEB will be addressing these ratchet straps and tie wraps during their aging management review.
SLRA AMP B.2.1.39, Electrical Insulation for Inaccessible Instrument and Control Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Summary of Information in the Application The SLRA states that AMP B.2.1.39, Electrical Insulation for Inaccessible Instrument and Control Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements, is a
62 The staff found that, for the parameter monitored or inspected and acceptance criteria program elements, sufficient information was not available to verify whether they were consistent with the corresponding program elements of the GALL-SLR Report AMP. The staff will consider issuing RAIs in order to obtain the information necessary to verify whether these program elements are consistent with the corresponding program elements of the GALL-SLR Report AMP.
During the audit, the staff made the following observations:
Enhancement 4 regarding the parameters monitored or inspected program element indicates that analysis has been completed to reevaluate the cumulative fatigue limit for the recirculation inlet nozzle safe ends, and the limits will be revised in FatiguePro (fatigue monitoring software) prior to entry into the SPEO. However, the following items related to this enhancement are not clear to the staff: (1) the meaning of the re-evaluated cumulative usage limit for the recirculation inlet nozzle safe ends in their fatigue reevaluation; and (2) the meaning of the limits that will be revised in the fatigue monitoring software. These items are also related to the observation regarding Enhancement 7, as further discussed below. In addition, the staff needs clarification on whether the enhancement is applied to the recirculation inlet nozzle.
Enhancement 7 regarding the acceptance criteria program element is to revise implementing procedures to reflect the re-evaluated cumulative fatigue values for the Units 1, 2, and 3 recirculation inlet nozzle safe end. However, it is not clear to the staff which aspect of the acceptance criteria program element will be enhanced in Enhancement 7. In addition, the staff needs clarification on whether the enhancement is applied to the recirculation inlet nozzle.
Enhancement 5 regarding the parameters monitored or inspected program element indicates that FatiguePro (fatigue monitoring software), Version 4, will be implemented prior to entry into the SPEO. In comparison, the following reference suggests that the applicants fatigue monitoring may be already using FatiguePro, Revision 4 (
Reference:
SIA Report 1801147.301, Revision 2, BFN 80 Year Cycle Projection Update for SLR, Section 4.0 and Figure 1). The staff needs clarification on the following items related to the use of the fatigue monitoring software: (1) currently used version of the FatiguePro software; (2) differences between the currently used version of FatiguePro and Revision 4; (3) whether the stress-based fatigue monitoring is performed on any components by using the fatigue monitoring software; and (4) how the applicant evaluated and addressed the potential concern discussed in NRC Regulatory Issue Summary 2008-30, Fatigue Analysis of Nuclear Power Plant Components (i.e., the use of only one value of stress in the fatigue analysis rather than considering the six components of the stress).
The staff also audited the description of the SLRA Fatigue Monitoring program in the FSAR supplement. The staff found that sufficient information was not available to determine whether the description provided in the FSAR supplement was an adequate description of the SLRA Fatigue Monitoring program especially regarding the description of the program enhancements as discussed above. The staff will consider issuing an RAI in order to obtain the information necessary to verify the sufficiency of the FSAR supplement program description.
85 information was not available to complete its review of the applicants basis for the TLAA disposition. In order to obtain the necessary information, the staff will consider issuing RAIs.
During the audit, the staff made the following observations:
The TLAAs regarding allowable stress analyses rely on the implicit fatigue analysis provisions in the ANSI B31.1 code. These provisions allow no reduction in the allowable stress range for thermal expansion stresses if the number of equivalent full temperature cycles does not exceed 7000 cycles. SLRA Tables 4.3.4-3 and 4.3.4-4 describe the 80-year projected cycles for the non-Class 1 piping systems other than the high temperature process sample system lines and the 80-year projected cycles for the high temperature process sample system lines, respectively. However, SLRA Section 4.3.4 does not clearly describe how the 80-year projected cycles were determined (e.g., based on piping system design information, plant operation procedures, test requirements, FSAR information and specific system-level knowledge).
SLRA Section 4.3.4 explains that the non-Class 1 piping portions of the following piping systems are only affected by the same pressure and temperature transients as the RCS transients that are listed in SLRA Table 4.3.4-2: (1) control rod drive, (2) core spray, (3) feedwater, (4) main steam, (5) containment atmosphere dilution, (6) RHR (including RHRSW since transients are bounded the RHR transients), and (7) standby liquid control piping systems. However, SLRA Section 4.3.4 does not clearly discuss why the non-Class 1 piping portions of these piping systems are not subject to piping-specific transients other than the RCS transients listed in SLRA Table 4.3.4-2.
SLRA Section 4.3.2 explains that all BFN piping lines were originally designed and evaluated in accordance with USAS B31.1-1967 design requirements and the design code requirements did not include an explicit fatigue analysis that involves CUF calculations. In comparison, the implicit fatigue analysis in SLRA Section 4.3.4, which involves the comparison of 80-year transient cycles to 7000 cycles, addresses only non-Class 1 piping but not Class 1 piping. Given that SLRA Section 4.3.2 does not address the explicit fatigue analysis (involving CUF calculations) for Class 1 piping, the staff needs clarification on whether SLRA Section 4.3.4 includes the Class 1 piping in its scope.
SLRA Section 4.3 does not include a TLAA related to high energy line break (HELB) analysis. In comparison, FSAR Appendix M, Section M1 indicates that the BFN HELB analysis uses the provisions in NRC GL 87-11 for the selection of pipe break locations.
The discussion in the FSAR also refers to Browns Ferry Design Criteria BFN-50-C-7105.
GL 87-11 includes Branch Technical Position (BTP) MEB 3-1, which includes the HELB postulation criterion for Class 1 piping based on the CUF (ADAMS Accession No. ML19137A335). MEB 3-1 also includes the HELB postulation criterion for non-Class 1 piping based on the allowable stress range for expansion stress (SA). SA may need to be adjusted based on the time-dependent cycles. FSAR Section M.6.9 regarding the additional HELB analysis also indicates that break locations were chosen based on pipe stress and routing. Therefore, the staff needs clarification on whether the HELB analysis is a TLAA given that the MEB 3-1 referenced in FSAR Appendix M includes the time-dependent criteria for break location postulations (e.g., criteria based on CUF or SA).
The staff also audited the description of the SLRA Metal Fatigue of Non-Class 1 Components TLAAs for the non-Class 1 piping systems provided in the FSAR supplement. The staff found that sufficient information was not available to determine whether the description provided in the FSAR supplement was an adequate description of the fatigue TLAAs for the non-Class 1
90 During the audit of the TLAA, the staff verified that the applicant has provided its basis that supports its disposition of 10 CFR 54.21(c)(1)(ii). However, the staff found that sufficient information was not available to complete its review of the applicants basis for the TLAA disposition. In order to obtain the necessary information, the staff will consider issuing RAIs.
During the audit, the staff made the following observations:
SLRA Section 4.3.8 addresses the 80-year CUF and CUFen calculations for the core shroud support. However, SLRA Section 4.3.8 does not clearly describe the following items: (1) the material of the core shroud location; (2) whether the analysis is bounding for BFN, Units 1, 2, and 3; (3) whether the evaluated core shroud location is a reactor coolant pressure boundary location; and (4) whether the CUFen calculations are performed in accordance with the guidance in NUREG/CR-6909, Revision 1.
The staff also noted that that the 80-year CUFen value (0.256) described in SLRA Section 4.3.8 is identical to the CUFen value for the location 21 (RV Region C/shroud support low-alloy steel location) described in SLRA Table 4.3.1-4. Therefore, the staff needs clarification on whether the location 21 in SLRA Table 4.3.1-4 is identical to the core shroud location evaluated in SLRA Sections 4.3.1 and 4.3.5 (EAF analysis).
In comparison with SLRA Section 4.3.8, SLRA Section 4.3.5 indicates that the aging effects of EAF for the core shroud support will be managed by using the Fatigue Monitoring AMP in accordance with 10 CFR 54.21(c)(1)(iii). Therefore, the staff also needs resolution of the potential inconsistency between the TLAA dispositions for the core shroud location (i.e., 10 CFR 54.21(c)(1)(ii) in SLRA Section 4.3.8 and 10 CFR 54.21(c)(1)(iii) in SLRA Section 4.3.5) regarding the fatigue analysis for the core shroud support.
The staff also audited the description of the SLRA Core Shroud Support Fatigue Analysis Reevaluation TLAA for the core shroud support provided in the FSAR supplement. The staff found that sufficient information was not available to determine whether the description provided in the FSAR supplement was an adequate description of the SLRA fatigue TLAA for the core shroud support. The staff will consider issuing an RAI in order to obtain the information necessary to verify the sufficiency of the FSAR supplement program description.
SLRA TLAA Section 4.3.9, BFN Unit 3 Core Spray T-Box Repair Fatigue Evaluation Summary of Information in the Application SLRA Section 4.3.10, BFN Unit 3 Core Spray T-Box Repair Fatigue Evaluation, discusses the fatigue analysis for the repaired T-box of BFN, Unit 3 core spray line. The applicant dispositioned the TLAA in accordance with 10 CFR54.21(c)(1)(iii).
Audit Activities The table below lists documents that were reviewed by the staff and were found relevant to the TLAA. The staff will document its review of this information in the SE.
98 During the audit, the staff made the following observations:
The staff confirmed that the applicant conservatively considered the effects of embrittlement on changes in toughness properties for the Unit 2 RV axial weld using the embrittlement prediction methods in RG 1.99, Revision 2 even though the threshold fluence level of 1E17 n/cm2 from Appendix H to 10 CFR Part 50 was not exceeded.
The staff verified that the maximum neutron fluence at the axial weld V-3-A is 1.92E15 n/cm2 as shown in BFN Unit 2 Reactor Pressure Vessel Fluence Evaluation Subsequent License Renewal, TransWare Enterprises, Inc, which is consistent with the information in SLRA Section 4.7.3.
The staff verified the ART of 24.2 0F for the axial weld at 64 EFPY as shown in Calculation File No. 2200107.307, Structural Integrity Associates, Inc, which was calculated consistent with RG 1.99, Revision 2 using available information from SLRA Section 4.2.
The staff verified the allowable flaw size of 1.7156 inches and initial flaw size of 1.6 inches as shown in Request for Alternative, BFN-21-ISI-02 (ML22014A344).
The staff also audited the description of the SLRA BFN Unit 2 Reactor Vessel Axial Weld Flaw TLAA for the RV axial weld provided in Section A.4.7.3 of the FSAR supplement. The staff verified this summary description is consistent with the generic description provided in the SRP-SLR.
AMR Items Not Associated with an AMP SLRA AMR 3.6.2.3.1, High-Voltage Electrical Insulators Summary of Information in the Application During the audit, the staff reviewed plant documentation associated with the AMR discussed in SLRA Section 3.6.2.3.1 and the following:
SLRA Table 3.6-1, Summary of Aging Management Evaluations for Electrical Commodities, Item Number 3.6-1,002-High-voltage electrical insulators composed of porcelain; malleable iron; aluminum; galvanized steel; cement; toughened glass; polymers; silicone rubber; fiberglass; aluminum alloy exposed to air - outdoor.
SLRA Table 3.6-1, Summary of Aging Management Evaluations for Electrical Commodities, Item Number 3.6-1,003-High-Voltage electrical insulators composed of porcelain; malleable iron; aluminum; galvanized steel; cement, toughened glass; polymers silicone rubber; fiberglass, aluminum alloy exposed to air-outdoor.
The SLRA states the in-scope high-voltage insulators (HVIs) provide electrical insulation for switchyard bus, transmission conductors, switchyard active components, and associated connections that are part of the circuits that supply power from electric utility transmission system to plant buses, including connecting the alternate AC source in the event of a station black out (SBO). The HVIs evaluated for the BFN SLRA are located in the offsite power source circuits and the alternate AC source for the SBO. The in-scope insulators include 500 kilovolt (kV), and 161 kV post insulators in offsite source paths.
104 Report No.
1900225.4/01 Assessment of Time Limited Aging Analysis for the Browns Ferry Nuclear lant Units 1, 2 and 3 for Subsequent License Renewal Revision 0 06/07/2019 Material Certification Reports Certified Material Test Reports of ASTM A36 Structural Steel Various: 1968 to 1969 N/A NRC Onsite Audit Items 1-15 Response Revision 0 N/A References in Support of Item #203 Related to TRP-076 Revision 0 N/A TRP-076-Question F5 Response Revision 0 N/A TRP-076_20240819(1).pdf August 2024 N/A TRP-076_20240918_1302.pdf September 2024 SLR-BFN-0016 Tennessee Valley Authority Browns Ferry Nuclear Plants Unit 1, 2, and 3 Subsequent License Renewal.
Further Evaluations - Containments, Structures and Component Support Revision 1 MAI-5.2 Browns Ferry Nuclear Plant Unit 0. Modification and Additions Instruction. Bolting and Structural Connections Revision 0018 NDQ000020030019 Dose to Polyurethane Foam Between the Drywell Steel and the Reactor Building Concrete Revision 1 12/14/2021 BFN-50-C-7100 QA Record. General Design Criteria Document.
Browns Ferry Nuclear Plant. Design of Civil Structures.
Revision 14 09/22/2022 G-29-B-P.S.4.M.4.4 ASME Section III and non ASME Section III (including AISC, ANSI, B 31.1, and ANSI B 31.5) Bolting Material Revision 0007 B44 060413 002 BFN Units 1, 2, and 3 Insulation Report 04/12/2006 0-TI-586 Aging Management Program Basis Document ASME Section XI, Subsection IWF Program Revision 2 0-TI-617 Aging Management Program Basis Document Structures Monitoring Program Revision 3 0-TPP-ENG-376 Containment Inservice Inspection Program Revision 3 0-TPP-ENG-417 Inservice Inspection Program Revision 0 NDQ0064980007 Primary Containment Analysis (Temperature)
Revision 8 CDQ0003032012000048 2012 Maintenance Rule Inspections Revision 0 CDQ0003032022000000 2022 Structural Maintenance Rule and License Renewal Inspections (includes Pace Analytical Report on effluent)
Revision 4 CDQ0303885672 Reactor Building Units 1, 2 and 3. Structural Steel Design for the Sacrificial Shield Wall inside the Drywell Revision 5 01/15/2016 CDQ0303885705 Reactor Building Powerhouse Seismic Analysis for Reactor Pressure Vessel Pedestal 07/16/1969
105 CDQ0303884304 Reactor Building-Vessel Anchor Bolts (Girder, Anchor Bolt, Vessel for BFN Units 1, 2, 3) 09/08/1969 CDQ2303900651 Sacrificial Shield Wall - Evaluation of Header Beams and 1/4 Plate 12/31/1990 B41 050414 002 General Engineering Specification G-2 for Plain and Reinforced Concrete Revision 8 CR-538994 Spalling of concrete on inner wall under Torus (Bay 16 of U3 just below the drywell shield wall of U3) 05/08/2012 CR-1680436 Water eroding through the wall in Bay 9 of U3 reactor building torus area 03/30/2021 CR-1679622 NOI U2R21-R004: Missing Washers from RPV Support RPV-SUPP-2-1 03/17/2021 WO 120323318 Magnetic Particle Examination, R-087, BFN 3 ISI Outage U3R19. ASME Section XI 2007 Ed 2008 Add.
Item B-K, Location annulus 03/04/2020 WO 1203/23318 U3R19 Visual Examination Record, R-082, BFN 3 ISI Outage U3R19. ASME Section XI 2007 Ed 2008 Add.
Item 03/04/2020 BFN-FLU-001-T-001 Displacements per Atom in the Steel Cladding of the Biological Shield at 64 Effective Full Power Years Revision 0 BFN-FLU-001-R-001 Browns Ferry Nuclear Plant Fluence Methodology Report. Attachment 2. Qualification of the Browns Ferry Unit 1 Reactor Fluence Model - Cycles 1 to 13 Revision 0 BFN-FLU-001-R-001 Browns Ferry Nuclear Plant Fluence Methodology Report. Attachment 1. Qualification of the Browns Ferry Unit 2 Reactor Fluence Model - Cycles 1 to 21 Revision 0 BFN-FLU-001-R-001 Browns Ferry Nuclear Plant Fluence Methodology Report. Attachment 3. Qualification of the Browns Ferry Unit 3 Reactor Fluence Model - Cycles 1 to 20 Revision 0 BFN-FLU-001-R-001 Browns Ferry Nuclear Plant Fluence Methodology Report.
Revision 0 TVA Report No.63-200-1 Browns Ferry Nuclear Plant - Final Design Report.
09/1974 TVA Report No.63-300-VII-I Browns Ferry Nuclear Plant - Final Construction Report. Volume I 06/1976 TVA/BFN 47E225-100 Harsh Environmental Data Drawing Series Index, Notes and References 03/18/2003 TVA/BFN 1-47E225-102 Harsh Environmental Data Elevation View 04/19/2019 TVA/BFN 2-47E225-101-2 Harsh Environmental Data Figures (LOCA Temperature Response) 07/10/1987 TVA/BFN 0-48N435 Structural Steel Framing at El. 541-6. Plan and Details 02/28/1968 TVA/BFN 0-48E445 Structural Steel. Sacrificial Shield Wall. Plans and Elevation 02/03/1969 TVA/BFN 3-47E865-10 Flow Diagram-Cooling Air Flow (Unit 3) 03/21/2004
106 TVA/BFN 3-47E610 1 Mechanical Control Diagram Primary Containment Cooling Temperature Monitoring System (Unit 3) 11/142011 TVA/BFN 0-47W481-11 Mechanical. Drains and Embedded Piping. Stage III and IV 04/22/1968 TVA/BFN 0-47W922-4 Mechanical. Cooling and Ventilating. Drywell Revision 4 02/20/1986 TVA/BFN 0-47B435-1 Mechanical. General Notes. Pipe Supports Revision 10 TVA/BFN 0-48B805-1 Miscellaneous Steel Seismic Conduit Supports Typical Supports Revision11 TVA-0/BFN 48N1017 BFN Reactor Building Units 1 & 3, Miscellaneous Steel Drywell Support Framing Pedestal and Sacrificial Shield Wall 04/271970 TVA/BFN 0-48N449 Reactor Building Units 1, 2 & 3. Structural Steel Drywell Seismic Bracing EL 624-8 & Bulkhead Plates EL 639-11 08/01/1980 TVA/BFN 0-48N446 BFN Reactor Building Units 1, 2 & 3, Structural Steel Sacrificial Shield Wall Sections & Details - SH 1 02/03/1969 TVA/BFN 0-48N447 BFN Reactor Building Units 1, 2 & 3, Structural Steel Sacrificial Shield Wall Sections & Details - SH 2 02/03/1969 TVA/BFN 0-48N448 BFN Reactor Building Units 1, 2 & 3, Structural Steel Sacrificial Shield Wall Sections & Details - SH 3 02/03/1969 TVA/BFN 48N978 Miscellaneous Steel Vessel Support Pedestal Embedded Parts - sheet 1 (BFN Units 1, 2, 3) 09/25/1972 TVA/BFN 48N1017 BFN Reactor Building Units 1 & 3, Miscellaneous Steel Drywell - Support Framing Pedestal & Sacrificial Shield Wall 10/29/1987 TVA/BFN 2-ISI-0415-C BFN Unit 2, Reactor Vessel, Vessel Supports 02/16/2005 TVA/BFN 3-ISI-0416-C BFN Unit 3, Reactor Vessel, Vessel Supports 02/16/2005 TVA/BFN 3-ISI-0414-C BFN Unit 1, Reactor Vessel, Vessel Supports 02/16/2005 TVA/BFN ISI-0415-C BFN Unit 2, Reactor Vessel, Vessel Supports 02/17/1993 GE-730E844 Reactor Vessel Support Nuclear Boiler (Drawing Sheets 1 and 2) 12/07/1967 GE-730E853 Reactor Vessel Stabilizer (including description of materials used)
Revision 10 GE-719E531 Primary Containment Loading 01/05/1972 B&W-122875E-2 Refueling Containment Skirt 10/28/1970 B&W-122672E BFN Units 1 & 2, Support Skirt Assy and Detail Revision 8 B&W-122872E BFN Units 1 & 2, Support Skirt Assy and Detail Revision 8 During the audit, the staff made the following observations:
Electric Power Research Institute (EPRI) Report 3002018400, Revision 1, September 2020, Basis for Evaluation of Concrete Biological Shield Wall for Aging Management
108 o The NRC staff noted the benchmark results consider secondary gamma sources from activated ex-core materials, ((
)) However, the NRC staff could not determine whether the results consider secondary gamma sources from hematite grout, which was used in the construction of the concrete biological shield walls for all units of the Brown Ferry Nuclear Generating Station. This was discussed in the audit, and the applicant identified ((
))
o Based on these observations, the NRC staff intends to submit several requests for clarification of information to the applicant to ensure the staffs assessment as documented in the safety evaluation will accurately reflect the manner in which the neutron fluence and gamma dose estimates were determined.
Structural Concrete, as stated in CDQ0303885672:
o is designed to ACI 318-63; o has Reinforcing steel with a yield strength of 36,000 psi; and o has the embedded anchor bolt tensile strengths determined by ACI 505-54 (Chimney Code).
Structural Concrete, as stated in TVA DWG 41N1080 and Concrete General Specs G-2:
o has a design compressive strength of 4,000 psi at 28 days, and o contains 1.5 limestone aggregates and fly ash.
Structural Concrete of the sacrificial shield wall (SSW) extends from elevation 575 and 8-5/8 to approximately 586 and 2-5/8. There are no process piping penetrations in the SSW structural concrete region. Above that elevation process piping is insulated using metal reflecting insulation per TVA procedure N1M-001 and age managed via SLRA AMP B.2.1.20.
TVA/BFN drawing No. 0-48E445, modified for SLRA shows the limestone concrete fill elevations to be from 575.38 to 595.27 and from 609.11 to 624.67. All other SSW fill is hematite and is exposed to core midplane at elevation 602.53.
Regarding effects of gamma radiation on the concrete infill of the biological shield wall, SOCOTEC report no. LA245188-R-001 states that This approach utilizing only the prompt secondary gamma exposure has been compared to a prior TransWare Enterprises Report (2013) [35] that considered both prompt secondary gamma and delayed gamma sources that were combined to determine a total gamma exposure. The delayed gamma sources included gamma from the decay of fission products in the fuel and activation of materials outside of the core. The analysis results that compared the prompt secondary gamma assessment only (i.e., as described in [2]) to that of the results that included prompt secondary and delayed gamma [35] were considered a favorable comparison relative to predicting temperature in the shield concrete.
Welding is in accordance with AWS D1-0 Specifications for Building Construction.
Implementing procedure 0-TTP-ENG-467, as supplemented, clarifies SLRA AMP B.2.1.1 and B.2.1.30 jurisdictions for inspection of the RV knuckle region and skirt.
109 IWF VT-3 Visual Examination of the support skirt includes the outside surface of the skirt, the ring girder assembly, girder bolting, and anchor bolting. The inside surface of the support skirt is not 100% accessible for VT-3 examination. It is examined to the maximum extent practical.
SSW access doors self-lubricating bearings that include graphite bronze bushings and graphite bronze thrust washer identified in TVA Calculation CDQ0303885672 do not serve an SLR intended function.
There are no process piping penetrations in the structural concrete region of the SSW.
Process piping penetrating the SSW above the 10.5 region are assumed to operate at operating temperatures that do not exceed 550°F. Process piping penetrating the SSW is insulated with metal reflective insulation age managed via the SLRA One-Time inspection AMP B.2.1.20.
SLRA AMP B.2.130, ASME Section XI, Subsection IWF, is enhanced to include volumetric examination of high strength bolting of 1 or grater nominal size and 150 ksi or more yield strength. Determination of whether the skirt to ring girder bolts will be included in the sample size requiring volumetric examination will be determined during implementation of the AMP.
GE drawing 730E853 specifies the yield stress of RV Stabilizer Rod Steel to be 136 ksi.
The certified material test reports for ASTM A36 structural steel did not include nil-ductility transition temperature values of the steel.
TVA/BFN drawing No. 0-48E445 specifies the structural steel framing of the SSW shall be ASTM A36.
TVA/BFN drawing Nos. 47E225-100 and 1-47E225-102 indicate an average drywell temperature of 136°F, a maximum drywell temperature of 145°F, and a minimum drywell temperature of 60°F in the core region elevation.
TVA/BFN drawing Nos. 0-48E445 and 1-47E225-102 show elevations of the top and bottom of the biological shield wall, and TVA drawing No. 0-48E445 shows elevations of the beltline region, including the core midplane.
B&W drawing No. 122872E shows dimensions of the reactor vessel bottom head and the attachment location of reactor vessel skirt.
Sargent & Lundy Report No. SL-017102 (Letter No. SL-TVA-1750) states that the average normal operating temperature in the drywell is 136°F based on the environmental data elevation drawings for all BFN units.
With regard to the SSW welds, in the August 2024 responses to TRP-076 breakout questions (file TRP-076_20240819(1).pdf), the applicant stated in part in response to Question F5 that the BFN SLRA will be updated via supplement to include an AMR item for steel in a concrete environment with the aging effect of distortion and cracking of welds resulting from potential expansion of concrete due to irradiation to be managed by the Structures Monitoring program.
111 Fire Water System 6/25/2024 9/23/2024 Atmospheric Metallic Tanks 7/3/2024 7/23/2024 One-Time Inspection 6/28/2024 Selective Leaching 6/5/2024 7/12/2024 8/27/2024 10/24/2024 11/18/2024 ASME Code Class 1 Small-Bore Piping 5/30/2024 6/13/2024 Monitoring of Neutron-Absorbing Materials Other Than Boraflex 5/29/2024 6/11/2024 ASME Section XI, Subsection IWF 7/9/2024 8/8/2024 10 CFR Part 50, Appendix J 6/18/2024 Masonry Wall Program 6/12/2024 Structures Monitoring Program 7/10/2024 10/16/2024 10/23/2024 12/16/2024 Non-EQ Electrical Cable and Connection Insulation Material 5/16/2024 Fuse Holders 5/28/2024 Fatigue Monitoring Program, GALL X.M1 5/16/2024 7/24/2024 Concrete 7/2/2024 7/12/2024 8/19/2024 10/24/2024 Irradiation-Structural 6/14/2024 6/27/2024 7/1/2024 7/9/2024 7/16/2024 8/7/2024 8/8/2024 8/9/2024 8/15/2024 9/20/2024 11/13/2024 12/6/2024 12/11/2024
112 Recurring Internal Corrosion 6/12/2024 8/13/2024 9/23/2024 SS Nickel Alloy Aluminum Alloy Further Evaluations 6/11/2024 No Aging Effects - Mechanical Components 6/6/2024 11/13/2024 12/4/2024 Non-GALL AMR - Structural Components 7/2/2024 10/24/2024 Jet Pump Auxiliary Spring Wedge Assembly Loss of Preload Analysis 5/23/2024 Jet Pump Riser Repair Clamp Loss of Preload Analysis 5/23/2024 Replacement Core Support Plate Plug Extended Life Irradiation - Enhanced Stress Relaxation Analysis 5/23/2024 Irradiation Assisted Stress Corrosion Cracking (IASCC) of Reactor Vessel Internals 5/23/2024 Core Spray Replacement Piping Bolting Loss of Preload Evaluation 6/13/2024 Core Spray Sparger Repair Clamp Loss of Preload Evaluation 5/29/2024 Access Hole Cover Repair Loss of Preload Evaluation 5/23/2024 Jet Pump Hold-Down Beam Assembly Loss of Preload Analysis 5/23/2024 Jet Pump Sensing Line Clamps Loss of Preload Analysis 5/23/2024 Elimination of Boiling Water Reactor Circumferential Weld Inspections 5/22/2024 Boiling Water Reactor Axial Welds 5/22/2024 Metal Fatigue, TLAA 4.3 5/23/2024 12/16/2024 Transient Cycle and Cumulative Usage Projections For 80 Years 5/16/2024 Jet Pump to Core Shroud Support Plate Fatigue Evaluation 5/16/2024 5/23/2024 Metal Fatigue of Class 1 Components 5/16/2024 11/4/2024 Class 1 Fatigue Waivers 5/16/2024 6/25/2024 Metal Fatigue of Non-Class 1 Components 5/22/2024 7/24/2024 Environmentally Assisted Fatigue 5/16/2024 7/24/2024 Replacement Steam Dryer Stress Report and Fatigue Evaluation 5/16/2024 Emergency Equipment Cooling Water System Weld Flaws Evaluation 5/16/2024 7/24/2024 Core Shroud Support Fatigue Analysis Reevaluation 5/16/2024
114 Thomas Earl Braudt TVA Travis Lamarr Shults TVA Victor D. Schiavone TVA William Andrew Whitener TVA William J. Baker TVA William Ross Victor TVA Zita I. Martin TVA Annessa Lippincott GSE Solutions Ethan Guio GSE Solutions Gregory Lupia GSE Solutions Jessie Jennings GSE Solutions Alexsandar Milicevic Sargent & Lundy David C. Skiba Sargent & Lundy Ed J. Englert Sargent & Lundy Jack F. Wakeland Sargent & Lundy Joseph Levine Sargent & Lundy Joseph Styczynski Sargent & Lundy Dan Denis Structural Integrity Associates Keith Evon Structural Integrity Associates Laura Perez Structural Integrity Associates Terry Herrmann Structural Integrity Associates