ML040200982
| ML040200982 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 01/20/2004 |
| From: | Mitra S Office of Nuclear Reactor Regulation |
| To: | Moyer J Carolina Power & Light Co |
| Mitra, SK, NRR/DRIP/RLEP, 415-2783 | |
| Shared Package | |
| ML040200981 | List: |
| References | |
| Download: ML040200982 (14) | |
Text
Safety Evaluation Report Related to the License Renewal of the H. B. Robinson Steam Electric Plant, Unit 2 Docket No. 50-261 Carolina Power & Light Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation January 2004
ii ABSTRACT This safety evaluation report (SER) documents the technical review of the H.B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, known as Robinson Nuclear Plant (RNP), license renewal application (LRA) by the U.S. Nuclear Regulatory Commission (NRC) staff. By letter dated June 14, 2002, Carolina Power & Light Company (CP&L or the applicant) submitted the LRA for RNP in accordance with Title 10 of the Code of Federal Regulations Part 54 (10 CFR Part 54 or the Rule). RNP is requesting renewal of the operating license for Unit 2 (License Number DPR-23) for a period of 20 years beyond the current expiration date of midnight, July 31, 2010. The construction permit for RNP was issued by the NRC on April 13, 1967, and the operating license was issued September 23, 1970, pursuant to Section 104b of the Atomic Energy Act of 1954, as amended.
RNP is adjacent to Unit 1 of the H.B. Robinson Steam Electric Plant (SEP), a coal-fired steam power plant. The plant is located on the edge of Lake Robinson, a man-made lake in Darlington and Chesterfield Counties, South Carolina. RNP is a pressurized light-water moderated and cooled system. The nuclear power plant incorporates a three-loop closed-cycle, pressurized water, nuclear steam supply system (NSSS) Westinghouse Electrical Corporation and licensed to generate 2339 MW-thermal, or approximately 769 MW-electric.
This SER presents the status of the staffs review of information submitted to the NRC through January 21, 2004. In its SER issued August 25, 2003, the staff has identified open and confirmatory items that had to be resolved before the staff could make a final determination on the application. These items and their resolutions are summarized in Sections 1.5 and 1.6 of this report. The staffs final conclusion of its review of the RNP LRA can be found in Section 6 of this SER.
iii ABBREVIATIONS AC alternating current ACI American Concrete Institute ACRS Advisory Committee on Reactor Safeguards (NRC)
AFW auxiliary feedwater AISC American Institute of Steel Construction AISI American Iron and Steel Institute ALARA as low as reasonably achievable AMP aging management program AMR aging management review ANL Argonne National Laboratory ANSI American National Standards Institute API American Petroleum Institute ASA American Standards Association ASME American Society of Mechanical Engineers ASTM American Society for Testing and Materials ATWS anticipated transient without scram AWWA American Water Works Association B&PV boiler and pressure vessel BIT boron injection tank BMI bottom mounted instrumentation BTP branch technical position BWR boiling-water reactor CASS cast austenitic stainless steel CCMS core cooling monitor system CE Combustion Engineering, Inc.
CETS core exit thermocouple system CF chemistry factor CRD control rod drive CCW component cooling water CFR Code Of Federal Regulations CLB current licensing basis CMAA Crane Manufacturers Association of America CP&L Carolina Power & Light Co.
CRDM control rod drive mechanism CSS containment spray system CST condensate storage tank CUF cumulative usage factor CV containment vessel CVCS chemical and volume control system DBA design basis accident DBD design basis document DBE design basis earthquake dc direct current
iv DG diesel generator DOE Department of Energy, U.S.
DS dedicated shutdown DSDG dedicated shutdown diesel generator EAF environmentally assisted fatigue ECCS emergency core cooling system EDB equipment database EOCI electric overhead crane institute ESS extraction steam system ET eddy current testing (NRC List of Abbreviations has this as ECT)
EDG emergency diesel generator EFPY effective full-power years EHC electrohydraulic control EJMA expansion joint manufacturers association EMA equivalent margins analysis EOF emergency operations facility EPRI Electric Power Research Institute EQ environmental qualification ER environmental report ESF engineered safety features FAC flow-accelerated corrosion Fen environmental fatigue multiplier FERC Federal Energy Regulatory Commission FMP fatigue monitoring program FP fire protection FPC Florida Power Corporation FR Federal Register FHB fuel-handling building FSAR final safety analysis report FW Feedwater GALL Generic Aging Lessons Learned (GALL) Report, NUREG-1801 GEIS generic environmental impact statement GDC general design criterion/criteria GL generic letter GSI generic safety issue HAD heat-actuated device HBRNS H.B. Robinson Nuclear Station HEPA high-efficiency particulate air HELB high energy line break HPSI high pressure safety injection HVAC heating, ventilation, and air conditioning I&C Instrumentation and control IA instrument air IASCC irradiation assisted stress corrosion cracking IEEE Institute of Electrical and Electronic Engineers IGA intergranular attack IGSCC intergranular stress-corrosion cracking IR insulation resistance ISG interim staff guidance
v ILRT integrated leak rate test (Containment Type A Test)
IN Information Notice INPO Institute Of Nuclear Power Operations IPA integrated plant assessment ISI inservice inspection IVSW isolation valve seal water LBB Leak before break LOCA loss-of-coolant accident LR license renewal LRA license renewal application MRP Material Reliability Project (EPRI)
MSS main steam system MCC Motor control center MDAFW motor-driven auxiliary feedwater MIC microbiologically induced corrosion MOV motor-operated valve MSIV main steam isolation valve NACE National Association of Corrosion Engineers NDE nondestructive examination NEPA National Environmental Policy Act of 1969 NPAR nuclear plant aging research NPS nominal pipe size NUREG NRC technical report designation (Nuclear Regulatory Commission)
NEI Nuclear Energy Institute NEMA National Electrical Manufacturers Association NFPA National Fire Protection Association NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System OE operating experience OCB oil circuit breaker P&I piping and instrumentation PMAMP preventive maintenance aging management program PAP personnel access portal pH concentration of hydrogen ions PM preventive maintenance PORV power-operated relief valve PPS penetration pressurization system PRT pressurizer relief tank P-T pressure-temperature PTS pressurized thermal shock PVC polyvinyl Chloride PWR pressurized water reactor PWSCC primary water stress corrosion cracking PWST primary water storage tank PZR pressurizer QA quality assurance QC quality control RAB reactor auxiliary building RAI request for additional information
vi RCDT reactor coolant drain tank RCP reactor coolant pump RCPB reactor coolant pressure boundary RCS reactor coolant system REDS radioactive equipment drain system RG regulatory guide RHR residual heat removal RMS radiation monitoring system RNP Robinson Nuclear Plant RO/RFO refueling outage RPS reactor protection system RPV reactor pressure vessel RTPTS reference temperature, pressurized thermal shock RTS reactor trip system RV reactor vessel RWST refueling water storage tank RTD resistance temperature detector RVLIS reactor vessel level instrumentation system scss steam cycle sampling system SEP steam electric plant SGB steam generator blowdown SGBS steam generator blowdown system SGTR steam generator tube rupture SOC Statement of Consideration SRP-LR Standard Review Plan for License Renewal SUT startup transformer SAR safety analysis report SBO station blackout SCs structures and components SCC stress corrosion cracking SDAFW steam-driven auxiliary feedwater SER safety evaluation report SFP spent fuel pit SG steam Generator SI safety injection SIT structural integrity test SR silicone rubber SRP standard review plan SSCs systems, structures, and components SFPCS spent fuel pit cooling system SWS service water system TDR time domain reflectometry TGSCC transgranular stress corrosion cracking TID total integrated dose TLAA time-limited aging analysis TSC Technical Support Center UAT unit auxiliary transformer UV ultraviolet UFSAR Updated Final Safety Analysis Report
vii USAS United States Of America Standards USE Upper Shelf Energy UT ultrasonic test Vac volts alternating current Vdc volts direct current VHP vessel head penetration VCT volume control tank WCAP Westinghouse Commercial Atomic Power WDS waste disposal system WOG Westinghouse Owners Group
viii TABLE OF CONTENTS ABSTRACT
............................................................... ii ABBREVIATIONS.......................................................... iii 1 Introduction and General Discussion...................................... 1-1 1.1 Introduction..................................................... 1-1 1.2 License Renewal Background...................................... 1-2 1.2.1 Safety Review........................................... 1-3 1.2.2 Environmental Review..................................... 1-4 1.3 Principal Review Matters.......................................... 1-5 1.3.1 Westinghouse Topical Reports............................... 1-6 1.4 Interim Staff Guidance............................................. 1-7 1.5 Summary of Open Items........................................... 1-9 1.6 Summary of Confirmatory Items.................................... 1-10 1.7 Summary of Proposed License Conditions............................ 1-29 2 Scoping and Screening Methodology for Identifying Structures and Components Subject to an Aging Management Review, and Implementation Results................ 2-1 2.1 Scoping and Screening Methodology.............................. 2-2 2.1.1 Introduction............................................. 2-2 2.1.2 Summary of Technical Information in the Application............. 2-2 2.1.2.1 Scoping Methodology............................. 2-3 2.1.2.2 Screening Methodology............................ 2-6 2.1.3 Staff Evaluation......................................... 2-8 2.1.3.1 Scoping Methodology............................. 2-9 2.1.3.2 Screening Methodology........................... 2-15 2.1.4 Evaluation Findings..................................... 2-18 2.2 Plant-Level Scoping Results.................................... 2-18 2.2.1 Summary of Technical Information in the Application............ 2-18 2.2.2 Staff Evaluation........................................ 2-18 2.2.3 Evaluation Findings..................................... 2-20 2.3 Scoping and Screening Results: Mechanical Systems................ 2-21 2.3.1 Reactor Systems...................................... 2-22 2.3.1.1 Reactor Coolant System Piping..................... 2-22 2.3.1.2 Reactor Coolant Pumps........................... 2-25 2.3.1.3 Pressurizer.................................... 2-26 2.3.1.4 Reactor Pressure Vessel.......................... 2-30 2.3.1.5 Reactor Vessel Internals.......................... 2-31 2.3.1.6 Steam Generators............................... 2-32 2.3.1.7 Reactor Vessel Level Instrumentation................ 2-35 2.3.1.8 Evaluation Findings.............................. 2-36 2.3.2 Engineered Safety Features Systems...................... 2-36 2.3.2.1 Residual Heat Removal System.................... 2-36 2.3.2.2 Safety Injection System........................... 2-37 2.3.2.3 Containment Spray System........................ 2-38
ix 2.3.2.4 Containment Air Recirculation Cooling System......... 2-41 2.3.2.5 Containment Isolation System...................... 2-44 2.3.3 Auxiliary Systems..................................... 2-47 2.3.3.1 Sampling Systems............................... 2-47 2.3.3.2 Service Water System............................ 2-50 2.3.3.3 Component Cooling Water System.................. 2-52 2.3.3.4 Chemical and Volume Control System
............... 2-54 2.3.3.5 Instrument Air System............................ 2-55 2.3.3.6 Nitrogen Supply/Blanketing System.................. 2-56 2.3.3.7 Radioactive Equipment Drain...................... 2-57 2.3.3.8 Primary and Demineralized Water System............ 2-59 2.3.3.9 Spent Fuel Pool Cooling System.................... 2-62 2.3.3.10 Containment Purge System....................... 2-64 2.3.3.11 Rod Drive Cooling System........................ 2-66 2.3.3.12 Heating Ventilation and Air ConditioningAuxiliary Building
............................................... 2-70 2.3.3.13 Heating, Ventilation, and Air ConditioningControl Room Area
............................................... 2-74 2.3.3.14 Heating, Ventilation, and Air ConditioningFuel Handling Building......................................... 2-78 2.3.3.15 Fire Protection System.......................... 2-82 2.3.3.16 Diesel Generator System......................... 2-88 2.3.3.17 Dedicated Shutdown Diesel Generator.............. 2-89 2.3.3.18 Emergency Operations Facility/Technical Support Center (EOF/TSC) Security Diesel Generator............. 2-90 2.3.3.19 Fuel Oil System................................ 2-91 2.3.4 Steam and Power Conversion Systems.................... 2-92 2.3.4.1 Turbine System................................. 2-92 2.3.4.2 Electro-Hydraulic Control System................... 2-94 2.3.4.3 Turbine Generator Lube Oil System................. 2-95 2.3.4.4 Extraction Steam System
......................... 2-96 2.3.4.5 Main Steam System
............................. 2-98 2.3.4.6 Steam Generator Blowdown System................. 2-99 2.3.4.7 Steam Cycle Sampling
........................... 2-99 2.3.4.8 Feedwater System.............................. 2-101 2.3.4.9 Auxiliary Feedwater System
...................... 2-102 2.3.4.10 Condensate System
.............................................. 2-104 2.3.4.11 Steam Generator Chemical Addition............... 2-106 2.3.4.12 Circulating Water System....................... 2-107 2.4 Scoping and Screening Results: Structures...................... 2-108 2.4.1 Containment........................................ 2-108 2.4.1.1 Containment Structure........................... 2-109 2.4.1.2 Containment Internal Structural Components......... 2-110 2.4.1.3 Containment External Structural Components......... 2-111 2.4.1.4 Summary of Technical Information in the Application... 2-112 2.4.1.5 Staff Evaluation................................ 2-113 2.4.1.6 Conclusions................................... 2-114 2.4.2 Other Structures..................................... 2-114
x 2.4.2.1 Reactor Auxiliary Building........................ 2-115 2.4.2.2 Fuel Handling Building........................... 2-117 2.4.2.3 Turbine Building................................ 2-120 2.4.2.4 Dedicated Shutdown Diesel Generator Building....... 2-121 2.4.2.5 Radwaste Building.............................. 2-123 2.4.2.6 Intake Structures............................... 2-124 2.4.2.7 North Service Water Header Enclosure.............. 2-126 2.4.2.8 Emergency Operations Facility/Technical Support Center Security Diesel Generator Building.......................... 2-128 2.4.2.9 Discharge Structures............................ 2-129 2.4.2.10 Lake Robinson Dam........................... 2-130 2.4.2.11 Pipe Restraint Tower........................... 2-132 2.4.2.12 Yard Structures and Foundations................. 2-133 2.4.2.13 Refueling System............................. 2-136 2.4.3 Evaluation Findings................................. 2-137 2.5 Scoping and Screening Results: Electrical/Instrumentation and Control Systems
......................................................... 2-137 2.5.1 Bus Duct.......................................... 2-139 2.5.1.1 Summary of Technical Information in the Application.. 2-139 2.5.1.2 Staff Evaluation............................... 2-139 2.5.1.3 Conclusions.................................. 2-140 2.5.2 Insulated Cables and Connections....................... 2-140 2.5.2.1 Summary of Technical Information in the Application... 2-140 2.5.2.2 Staff Evaluation............................... 2-140 2.5.2.3 Conclusion................................... 2-141 2.5.3 Electrical/Instrumentation and Control Penetration Assemblies. 2-141 2.5.3.1 Summary of Technical Information in the Application... 2-142 2.5.3.2 Staff Evaluation............................... 2-142 2.5.3.3 Conclusions.................................. 2-142 2.5.4 Station Blackout..................................... 2-143 2.5.4.1 Summary of Technical Information in the Application... 2-143 2.5.4.2 Staff Evaluation............................... 2-143 2.5.4.3 Conclusions.................................. 2-144 2.5.5 Evaluation Findings.................................. 2-144 3 Aging Management Review.............................................. 3-1 3.0.1 The GALL Format for the License Renewal Application............ 3-2 3.0.2 The Staffs Review Process for GALL......................... 3-3 3.0.3 Aging Management Programs............................... 3-4 3.0.3.1 Metal Fatigue of Reactor Coolant Pressure Boundary (Fatigue Monitoring Program)................................ 3-9 3.0.3.2 ASME Section XI, Inservice Inspection, Subsections IWB, IWC, and IWD Program.................................... 3-12 3.0.3.3 Water Chemistry Program.......................... 3-13 3.0.3.4 Boric Acid Corrosion Program....................... 3-17 3.0.3.5 Flow Accelerated Corrosion Program................. 3-23 3.0.3.6 Bolting Integrity Program........................... 3-28 3.0.3.7 Open-Cycle Cooling Water System Program........... 3-31
3.0.3.8 Closed-Cycle Cooling Water System Program.......... 3-33 3.0.3.9 One-Time Inspection Program....................... 3-34 3.0.3.10 Selective Leaching of Materials Program.............. 3-36 3.0.3.11 Systems Monitoring Program
...................... 3-37 3.0.3.12 Preventive Maintenance Program.................... 3-40 3.0.4 RNP Quality Assurance Program Attributes Integral to Aging Management Programs............................................. 3-43 3.0.4.1 Summary of Technical Information in Application........ 3-44 3.0.4.2 Staff Evaluation.................................. 3-44 3.0.4.3 Conclusions..................................... 3-45 3.1 Reactor Systems............................................... 3-46 3.1.1 Summary of Technical Information in the Application............. 3-47 3.1.2 Staff Evaluation......................................... 3-47 3.1.2.1 Aging Management Evaluations in the GALL Report That Are Relied On for License Renewal, and Which Do Not Require Further Evaluation....................................... 3-52 3.1.2.2 Aging Management Evaluations in the GALL Report That Are Relied On for License Renewal, For Which GALL Recommends Further Evaluation................................. 3-74 3.1.2.3 Aging Management Programs (System-Specific)....... 3-102 3.1.2.4 Aging Management of Plant-Specific Components...... 3-136 3.1.3 Evaluation Findings..................................... 3-175 3.2 Engineered Safety Features Systems
.............................. 3-176 3.2.1 Summary of Technical Information in the Application............ 3-176 3.2.2 Staff Evaluation........................................ 3-176 3.2.2.1 Aging Management Evaluations in the GALL Report That Are Relied On for License Renewal, Which Do Not Require Further Evaluation...................................... 3-179 3.2.2.2 Aging Management Evaluations in the GALL Report That Are Relied On for License Renewal, For Which GALL Recommends Further Evaluation................................ 3-180 3.2.2.3 Aging Management Program for ESF System Components
.............................................. 3-184 3.2.2.4 Aging Management Review of Plant-Specific ESF System Components.................................... 3-184 3.2.3 Evaluation Findings..................................... 3-198 3.3 Auxiliary Systems............................................. 3-200 3.3.1 Summary of Technical Information in the Application............ 3-200 3.3.2 Staff Evaluation........................................ 3-201 3.3.2.1 Aging Management Evaluations in the GALL Report That Are Relied On for License Renewal, Which Do Not Require Further Evaluation...................................... 3-206 3.3.2.2 Aging Management Evaluations in the GALL Report That Are Relied On for License Renewal, For Which GALL Recommends Further Evaluation................................ 3-206 3.3.2.3 Aging Management Programs (System-Specific)....... 3-213 3.3.2.4 Aging Management of Plant-Specific Components...... 3-250 3.3.2.5 General AMR Issues............................. 3-294 3.3.3 Evaluation Findings..................................... 3-298 3.4 Steam and Power Conversion Systems............................. 3-299
xii 3.4.1 Summary of Technical Information in the Application............ 3-299 3.4.2 Staff Evaluation........................................ 3-299 3.4.2.1 Aging Management Evaluations in the GALL Report that Are Relied on for License Renewal, Which Do Not Require Further Evaluation
.............................................. 3-303 3.4.2.2 Aging Management Evaluations in the GALL Report That Are Relied On for License Renewal, For Which GALL Recommends Further Evaluation............................... 3-303 3.4.2.3 Aging Management Programs for Steam and Power Conversion Systems....................................... 3-309 3.4.2.4 Aging Management of Plant-Specific Components...... 3-309 3.4.3 Evaluation Findings..................................... 3-338 3.5 Containments, Structures, and Component Supports................ 3-339 3.5.1 Summary of Technical Information in the Application........... 3-339 3.5.2 Staff Evaluation....................................... 3-340 3.5.2.1 Aging Management Evaluations in the GALL Report That Are Relied On for License Renewal, Which Do Not Require Further Evaluation...................................... 3-347 3.5.2.2 Aging Management Evaluations in the GALL Report That Are Relied On for License Renewal, For Which GALL Recommends Further Evaluation............................... 3-347 3.5.2.3 Aging Management Programs for Containment, Structures, and Component Supports............................. 3-364 3.5.2.4 Aging Management Review of Plant-Specific Structures and Structural Components............................ 3-386 3.5.3 Evaluation Findings...................................... 3-403 3.6 Electrical and Instrumentation and Controls
.......................... 3-404 3.6.1 Summary of Technical Information in the Application............ 3-404 3.6.2 Staff Evaluation......................................... 3-405 3.6.2.1 Aging Management Evaluations in the GALL Report That Are Relied On for License Renewal, Which Do Not Require Further Evaluation
.............................................. 3-407 3.6.2.2 Electrical Equipment Subject to Environmental Qualification
.............................................. 3-408 3.6.2.3 Aging Management Programs for Electrical and Instrumentation and Controls Components......................... 3-408 3.6.2.4 Aging Management of Plant-Specific Components....... 3-420 3.6.3 Evaluation Findings..................................... 3-426 4 TIME-LIMITED AGING ANALYSES......................................... 4-1 4.1 Identification of Time-Limited Aging Analyses
.......................... 4-1 4.1.1 Summary of Technical Information in the Application.............. 4-1 4.1.2 Staff Evaluation.......................................... 4-2 4.1.3 Conclusions............................................. 4-3 4.2 Reactor Vessel Neutron Embrittlement................................ 4-3 4.2.1 Summary of Technical Information in the Application.............. 4-4 4.2.1.1 Pressurized Thermal Shock.......................... 4-4 4.2.1.2 Reactor Vessel Upper-Shelf Energy................... 4-4
xiii 4.2.1.3 Plant Heatup/Cooldown (Pressure/Temperature) Curves/Low-Temperature Overpressure Protection Power-Operated Relief Valve Setpoints......................................... 4-5 4.2.2 Staff Evaluation.......................................... 4-5 4.2.2.1 Pressurized Thermal Shock.......................... 4-6 4.2.2.2 Reactor Vessel Upper-Shelf Energy................... 4-7 4.2.2.3 Plant Heatup and Cooldown (Pressure/Temperature) Curves/Low-Temperature Overpressure Protection Power-Operated Relief Valves Setpoints................................... 4-9 4.2.3 UFSAR Supplement...................................... 4-10 4.2.4 Conclusions............................................ 4-12 4.3 Metal Fatigue.................................................. 4-12 4.3.1 Summary of Technical Information in the Application............. 4-13 4.3.2 Staff Evaluation......................................... 4-15 4.3.2.1 Explicit Fatigue Analysis (ASME Section III, Class A)..... 4-15 4.3.2.2 Implicit Fatigue Design (ASME Section III, Class C, ANSI B31.1)
............................................... 4-20 4.3.2.3 Environmentally Assisted Fatigue Evaluation
........... 4-21 4.3.3 Conclusions............................................ 4-23 4.3.4 Reactor Vessel Underclad Cracking.......................... 4-23 4.3.4.1 Summary of Technical Information in the Application..... 4-23 4.3.4.2 Staff Evaluation.................................. 4-23 4.3.4.3 Updated Final Safety Analysis Report Supplement....... 4-24 4.3.4.4 Conclusions..................................... 4-24 4.3.5 Containment Penetration Bellows Fatigue..................... 4-25 4.3.5.1 Summary of Technical Information in the Application..... 4-25 4.3.5.2 Staff Evaluation.................................. 4-25 4.3.5.3 Conclusions..................................... 4-26 4.3.6 Crane Cycle Load Limits.................................. 4-26 4.3.6.1 Summary of Technical Information in The Application..... 4-26 4.3.6.2 Staff Evaluation.................................. 4-27 4.3.6.3 Conclusions..................................... 4-28 4.4 Environmental Qualification of Electrical Equipment..................... 4-28 4.4.1 Electrical and I&C Component Environmental Qualification Analyses
..................................................... 4-29 4.4.1.1 Summary of Technical Information in the Application...... 4-29 4.4.1.2 Staff Evaluation................................. 4-30 4.4.1.3 Limitorque SBM Motor-Operated Valve ActuatorsOutside Containment..................................... 4-36 4.4.1.3 Conclusions.................................... 4-37 4.4.2 GSI-168, Environmental Qualification of Electrical Components.... 4-37 4.4.2.1 Summary of Technical Information in the Application..... 4-37 4.4.2.2 Staff Evaluation.................................. 4-38 4.4.2.3 Conclusions..................................... 4-38 4.5 Concrete Containment Tendon Loss of Prestress....................... 4-38 4.5.1 Summary of Technical Information in the Application............. 4-38 4.5.2 Staff Evaluation......................................... 4-39 4.5.3 Conclusions............................................ 4-41 4.6 Other TLAAs................................................... 4-42
xiv 4.6.1 Thermal Aging Embrittlement............................... 4-42 4.6.1.1 Summary of Technical Information in the Application..... 4-42 4.6.1.2 Staff Evaluation.................................. 4-43 4.6.1.3 Updated Final Safety Analysis Report Supplement....... 4-46 4.6.1.4 Conclusions..................................... 4-47 4.6.2 Foundation Pile Corrosion................................. 4-47 4.6.2.1 Summary of Technical Information in the Application..... 4-47 4.6.2.2 Staff Evaluation.................................. 4-48 4.6.2.3 Conclusions..................................... 4-48 4.6.3 Elimination of Containment Penetration Coolers................ 4-49 4.6.3.1 Summary of Technical Information in the Application..... 4-49 4.6.3.2 Staff Evaluation.................................. 4-49 4.6.3.3 Conclusions..................................... 4-50 4.6.4 Aging of Boraflex....................................... 4-50 4.6.4.1 Summary of Technical Information in the Application...... 4-50 4.6.4.2 Staff Evaluation.................................. 4-51 4.6.4.3 Updated Final Safety Analysis Report Supplement....... 4-52 4.6.4.4 Conclusions..................................... 4-52 5 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS......... 5-1 6 CONCLUSIONS........................................................ 6-1 Appendix A: COMMITMENT LISTINGS....................................... A-1 Appendix B: CHRONOLOGY............................................... B-1 Appendix C: PRINCIPAL CONTRIBUTORS................................... C-1 Appendix D: REFERENCES
............................................... D-1