ML14010A314
| ML14010A314 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 12/30/2013 |
| From: | Energy Northwest |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML14010A476 | List:
|
| References | |
| GO2-13-174 | |
| Download: ML14010A314 (44) | |
Text
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 Appendix A SUPPLEMENT AGING MANAGEMENT PROGRAMS AND ACTIVITIES CREDITED FOR COLUMBIA LICENSE RENEWAL TABLE OF CONTENTS Section Page LDCN-09-035 A-i A.0 FINAL SAFETY ANALYSIS REPORT SUPPLEMENT ..............................
A.1-1 A.1 INTRODUCTION ..............................................................................
A.1-1 A.2 AGING MANAGEMENT PROGRAMS AND ACTIVITIES
..........................
A.1-1 A.2.1 AGING MANAGEME NT PROGRAMS ................................................... A.2-2 A.2.1.1 Aboveground Steel Tanks Inspect ion ...................................................
A.2-2 A.2.1.2 Air Quality Sa mpling Program ........................................................... A.2-2 A.2.1.3 Appendix J Program ....................................................................... A.2-3 A.2.1.4 Bolting Integr ity Program ................................................................. A.2-3 A.2.1.5 Buried Piping and Tank s Inspection Program ......................................... A.2-3 A.2.1.6 BWR Feedwater Nozzle Program ....................................................... A.2-4 A.2.1.7 BWR Penetrati ons Program .............................................................. A.2-4 A.2.1.8 BWR Stress Corrosion Cracking Program ............................................. A.2-5 A.2.1.9 BWR Vessel ID Attachment Welds Program .......................................... A.2-5 A.2.1.10 BWR Vessel Internals Program .......................................................... A.2-5 A.2.1.11 BWR Water Chem istry Program ........................................................ A.2-6 A.2.1.12 Chemistry Program Effectiveness Inspection ..........................................
A.2-6 A.2.1.13 Closed Cooling Wate r Chemistry Program ............................................ A.2-7 A.2.1.14 Cooling Units In spection Program ...................................................... A.2-7 A.2.1.15 Control Rod Drive Retu rn Line Nozzle Program ..................................... A.2-7 A.2.1.16 Diesel Starti ng Air Inspec tion ............................................................
A.2-8 A.2.1.17 Diesel Systems Inspection Program ..................................................... A.2-8 A.2.1.18 Diesel-Driven Fire Pumps Inspection Program ....................................... A.2-8 A.2.1.19 Electrical Cabl es and Connections Not Subj ect to 10 CFR 50.49 EQ Requirements Pr ogram ....................................................................
A.2-9 A.2.1.20 Electrical Cables and Connections Not Subj ect to 10 CFR 50.49 EQ Requirements Used in Instrumentation Circuits Program ........................... A.2-9 A.2.1.21 Electrical Cable Connections Not Subject to 10 CFR 50.49 EQ Requirements Inspection
.................................................................................. A.2-10 A.2.1.22 EQ Pr ogram ...............................................................................
A.2-10 A.2.1.23 External Surfaces Monitoring Program
............................................... A.2-11 A.2.1.24 Fatigue Monito ring Program
........................................................... A.2-11 A.2.1.25 Fire Pr otection Program
................................................................. A.2-11 C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 Appendix A SUPPLEMENT AGING MANAGEMENT PROGRAMS AND ACTIVITIES CREDITED FOR COLUMBIA LICENSE RENEWAL TABLE OF CONTENTS (continued)
Section Page LDCN-09-035 A-ii A.2.1.26 Fire Wate r Program
...................................................................... A.2-12 A.2.1.27 Flexible Connection Inspection Program ............................................. A.2-12 A.2.1.28 Flow-Accelerated Corro sion (FAC) Program ....................................... A.2-12 A.2.1.29 Fuel Oil Chem istry Program ........................................................... A.2-13 A.2.1.30 Heat Exchange rs Inspection ............................................................ A.2-13 A.2.1.31 High-Voltage Porcelain Insulators Aging Management Program ................ A.2-14 A.2.1.32 Inaccessible Power Cables Not Subject to 10 CFR 50.49 EQ Requirements Program .................................................................................... A.2-14 A.2.1.33 Inservice Inspect ion (ISI) Program .................................................... A.2-15 A.2.1.34 Inservice Inspection (ISI) Program - IWE ........................................... A.2-15 A.2.1.35 Inservice Inspection (ISI) Program - IWF ........................................... A.2-16 A.2.1.36 Lubricating Oil An alysis Program ..................................................... A.2-16 A.2.1.37 Lubricating Oil Inspection .............................................................. A.2-17 A.2.1.38 Masonry Wall Inspection
................................................................ A.2-17 A.2.1.39 Material Handling System Inspection Program ..................................... A.2-17 A.2.1.40 Metal-Enclosed Bus Program .......................................................... A.2-17 A.2.1.41 Monitoring and Collection Systems Inspection Program .......................... A.2-18 A.2.1.42 Open-Cycle Cooli ng Water Program ................................................. A.2-18 A.2.1.43 Potable Water M onitoring Program ................................................... A.2-19 A.2.1.44 Preventive Maintenance -
RCIC Turbine Casing ................................... A.2-19 A.2.1.45 Reactor Head Clos ure Studs Program ................................................ A.2-19 A.2.1.46 Reactor Vessel Surveillance Program ................................................. A.2-19 A.2.1.47 Selective Leach ing Inspection .......................................................... A.2-20 A.2.1.48 Service Air System Inspection Program .............................................. A.2-20 A.2.1.49 Small Bore Class 1 Piping Program ................................................... A.2-21 A.2.1.50 Structures Mon itoring Program ........................................................ A.2-21 A.2.1.51 Supplemental Piping/Tank Inspection
................................................. A.2-21 A.2.1.52 Thermal Aging and Neutron Embrittlement of Cast Austenitic Stainless Steel (CASS) Program
.......................................................................... A.2-22 A.2.1.53 Water Control Stru ctures Inspec tion .................................................. A.
2-22 A.2.1.54 Boron Carbide M onitoring Progra m .................................................. A.2-22 A.2.1.55 Service Level 1 Protective Coatings Program ....................................... A.2-23
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 Appendix A SUPPLEMENT AGING MANAGEMENT PROGRAMS AND ACTIVITIES CREDITED FOR COLUMBIA LICENSE RENEWAL TABLE OF CONTENTS (continued)
Section Page LDCN-09-035 A-iii A.2.2 EVALUATION OF TIME-LIMITE D AGING ANALYSES ......................... A.2-23 A.2.2.1 Reactor Vessel Neutron Embrittlement ............................................... A.2-23 A.2.2.1.1 Neutron Fluence ................................................................. A.2-24 A.2.2.1.2 Upper Shelf En ergy Evaluation ............................................... A.2-25 A.2.2.1.3 Adjusted Reference Temperature Analysis ................................. A.2-26 A.2.2.1.4 Pressure-Temperature Limits
.................................................. A.2-26 A.2.2.1.5 Reactor Vessel Circumferential Weld Inspection Relief .................. A.2-27 A.2.2.1.6 Reactor Vessel Axial Weld Failure Probability ............................ A.2-28 A.2.2.2 Metal Fatigue
.............................................................................. A.2-28 A.2.2.2.1 Reactor Pressure Vessel Fatigue Analyses .................................. A.2-29 A.2.2.2.2 Reactor Pressure Vessel Internals ............................................ A.2-29 A.2.2.2.3 Reactor Coolant Pre ssure Boundary Pi ping and Piping Component Fatigue Analyses ................................................. A.2-31 A.2.2.3 Non-Class 1 Component Fatigue Analyses .......................................... A.2-32 A.2.2.4 Effects of Reactor Coolan t Environment on Fatigue Life of Components and Piping ...................................................................................... A.2-33 A.2.2.5 Environmental Qualification of Electrical Equipment
.............................. A.2-34 A.2.2.6 Fatigue of Primary Containmen t, Attached Piping, and Components
........... A.2-34 A.2.2.6.1 Primary C ontainment
........................................................... A.2-34 A.2.2.6.2 ASME Class MC Components ................................................ A.2-35 A.2.2.6.3 Downcomers ..................................................................... A.2-35 A.2.2.6.4 Safety Relief Valv e Discharge Piping ....................................... A.2-36 A.2.2.6.5 Diaphragm Floor Seal
.......................................................... A.2-36 A.2.2.6.7 ECCS Sucti on Strainers
........................................................ A.2-37 A.2.2.7 Other Plant-Specific Time-Limited Aging Analyses ............................... A.2-37 A.2.2.7.1 Reactor Vessel Shell Indications .............................................. A.2-37 A.2.2.7.2 Sacrificial Shield Wall .......................................................... A.2-38 A.2.2.7.3 Main Steam Flow Restri ctor Erosion Anal yses ............................ A.2-38 A.2.2.7.4 Core Plate Rim Hold-Down Bolts ............................................ A.2-39 A.2.2.7.5 Crane Load Cycle Limit ....................................................... A.2-39 A.3 REFERENCES .................................................................................. A.3-1
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.1-1 APPENDIX A SUPPLEMENT AGING MANAGEMENT PROGRAMS AND ACTIVITIES CREDITED FOR CO LUMBIA LICENSE RENEWAL A.0 FINAL SAFETY ANALYSIS REPORT SUPPLEMENT A.1 INTRODUCTION This appendix provides the information submitted for the Final Safety Analysis Report (FSAR) Supplement as required by 10 CFR 54.21(d) for the License Renewal A pplication (LRA). The programs and activities credited to manage the effects of aging are described in LRA Appendix B. Section 4 of the LRA documents the evaluations of time-limited aging analyses for the period of extended operation. LR A Section 3, Section 4, and Appendix B have been used to prepare the program and activity descriptions that are contained in this appendix.
A.2 AGING MANAGEMENT PROGRAMS AND ACTIVITIES The license renewal integrated plant assessment identified existing and new ag ing management programs (AMPs) necessary to provide reasonabl e assurance that compone nts within the scope of license renewal will continue to perform their intended functions consistent with the current licensing basis (CLB) for the peri od of extended operation. This section describes the aging management programs a nd activities identified during the integrated plant assessment. The aging management programs will be implemented prior to the pe riod of extended operation. One-time inspections will be conducted within the 10-year period prior to beginning the period of extended operation. The aging management programs identified as necessary in association with the evaluation of time-limited aging analyses (TLAAs) are described in Section A.2.2. Three elements of an effective aging manage ment program that are common to each of the aging management programs are corrective actions, confirmation process, and administrative controls. These elements are included in the Operational Quality Assurance Program Description (OQAPD) for Columbia, which im plements the require ments of 10 CFR 50, Appendix B. Prior to the period of extended operation, the elements of corrective actions, confirmation process, and administrative controls in the OQAPD will be applied to required aging management programs for both safety-related and non-safety related structures and components determined to require ag ing management during the period of extended operation.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-1 The existing Corrective Action Program and the Operating Experience Program ensure, through the continual review of both plant-specific and industr y operating experience, that the license renewal aging management programs are effective to manage the aging effects for which they are credited. The aging management programs are either enhanced or new programs are developed when the review of operating experience indicates that the aging management programs may not be effective. For each ag ing management program listed in this section, operating experience is reviewed on a continuing basis.
The processes and procedures for the review of operating experience address the following points: All operating experience is sc reened for aging of long liv ed passive structures or components and further evaluati on as applicable is perfor med by personne l trained in the requirements of license re newal scoping, screening, an d aging management reviews (aging effects and mechanisms). The evaluation is completed and prioritized commensurate with the potential significance of the issue. Such evaluations are documented and retained in an a uditable and retrievable form.
Periodic training for system engineers, e quipment operators and maintenance personnel specific to identifying aging issues.
The License Renewal program lead is traine d in the requirements of license renewal scoping, screening, and aging management reviews (aging effects and mechanisms).
Aging management program owners are trained in the requirements of license renewal scoping, screening, and agi ng management reviews (aging effects and mechanisms) associated with their particul ar aging management program.
When it is determined that enhancements are necessary to adequately manage the effects of aging, the enhancements are entere d into and implemented consistent with the plant corrective action program or operating experience program, as applicable.
Enhancements can include, as appropriate , modifications to aging management programs or the creation and implementation of new AMPs.
Operating experience that is related to aging of long lived passive structures or components is keyword tagged "Aging."
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-2 The processes are adequate so as to not preclude the consideration of operating experience related to agi ng management. The proce sses appropriately gather information on all structures and components within the scope of license renewal, and their materials, environments, aging effects, and aging mechanisms. In addition, the processes include the AMPs credited for managing the effects of aging, and the activities under these AMPs (e.g., inspection methods, pr eventive actions , evaluation techniques, etc.).
While the programs and procedures may specify reviews of certain sources of information, such as NRC generic communi cations and Institute of Nuclear Power Operations reports, they allow for any poten tial source of relevant plant specific or industry operating expe rience information.
AMP owners review data collected by the AMPs, utilize the corr ective action program for any conditions that are unsatisfactory to ensure they will be addressed and corrected, maintain required records for the program and maintain the program current and implement revisions as needed based on program results and internal or external operating experience.
Provide guidance on sharing internal operating experience related to license renewal issues with the industry.
A.2.1 AGING MANAGEMENT PROGRAMS A.2.1.1 ABOVEGROUND STEEL TANKS INSPECTION The Aboveground Steel Tanks In spection detects and characterizes the conditions on the bottom surfaces of the condensat e storage tanks. The inspecti on provides direct evidence through volumetric examination as to whether, and to what exte nt, a loss of material due to corrosion has occurred in inaccessible areas (i.e., tank base and bottom surface).
The Aboveground Steel Tanks Insp ection is a new inspection program that will be implemented prior to the period of extended operation. The inspection activ ities will be conducted within the 10-year period prior to the period of extended operation.
A.2.1.2 AIR QUALITY SAMPLING PROGRAM The Air Quality Sampling Program is an existing prevention and condition monitoring program that manages loss of material due to corrosion for Diesel Star ting Air (DSA) components that contain compressed air through periodic sampling of the air for hydrocarbons, dewpoint, and C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-3 particulates and periodic ultras onic inspection of the DSA System air receivers. In addition, the Air Quality Sampling Program ensures that th e Control Air System remains dry and free of contaminants, such that no agi ng effects require management.
The Air Quality Sampling Program is supplemen ted by the Diesel Starting Air Inspection, which provides verification of the effectiveness of the program in mitigating the effects of aging in the DSA System dryers and the downstream piping and components (excluding the DSA System air receivers).
A.2.1.3 APPENDIX J PROGRAM The Appendix J Program is an existing monitoring program that detects degradation of the Primary Containment and systems penetrati ng the Primary Containment, which are the containment shell and primary containment pene trations including (but not limited to) the personnel airlock, equipment hatch, control rod drive hatch, and drywell head. The Appendix J Program provides assurance th at leakage from th e Primary Containment will not exceed maximum values for containment leakage.
A.2.1.4 BOLTING INTEGRITY PROGRAM The Bolting Integrity Program is a combination of existing activiti es that, in conjunction with other credited programs, address the management of aging for the bolting of mechanical components and structural connections within the scope of license re newal. The Bolting Integrity Program relies on manufacturer and vendor information and industry recommendations for the proper selection, assembly, and mainte nance of bolting for pressure-retaining closures and structur al connections. The Bolting Inte grity Program includes, through the Inservice Inspection (ISI) Pr ogram, Inservice Inspection (ISI)
Program - IWF, Structures Monitoring Program, and Extern al Surfaces Monitoring Program, the periodic inspection of bolting for indications of degradation such as leakage, loss of material due to corrosion, loss of pre-load, and cracking due to stress corrosion cracking (SCC) and fatigue.
A.2.1.5 BURIED PIPING AND TANK S INSPECTION PROGRAM The Buried Piping and Tanks Insp ection Program manages the effects of loss of material due to corrosion on the external surfaces of metallic pipi ng and tanks that are buried or underground. The program also manages the e ffects of cracking, loss of mate rial (and loss of pre-load) for bolting that is buried. In addition, the program also verifies that ag ing degradation is not occurring for concrete and poly mer piping that is buried.
The Buried Piping and Tanks Inspection Program is a combina tion of a mitigation program (cons isting of protective coatings.
cathodic protection, and backfill quality) and a condition monitoring pr ogram (consisting of electrochemical verification of cathodic protection, confirmation of backfill quality, visual C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-4 inspections of pipe or tank exte rnal surfaces, and non-destruc tive evaluation of pipe or tank wall thickness as needed).
Inspection of buried and underground piping will be performed within the 10-year period prior to entering the period of ex tended operation. Additional inspections of buried and underground piping and buried tanks will be performed within 10 years after entering the period of extended operation, and in each 10 year period thereafter.
The Buried Piping and Tanks Inspection Program is an existing program that requires enhancement prior to the pe riod of extended operation.
A.2.1.6 BWR FEEDWATER NOZZLE PROGRAM The BWR Feedwater Nozzle Program is an existing program th at manages cracking due to stress corrosion cracking and intergranular attack (SCC/IGA) and flaw growth of the feedwater nozzles. The BWR Feedwater Nozzle Program is in accordance with ASME Section XI and NRC augmented requirements. The BWR Feedwater Nozzle Program consists of: (a) enhanced inservice inspection in accordance with the requirements of the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWB, Table IWB 2500-1 (2001 editio n including the 2002 an d 2003 Addenda) and the recommendations of General Electri c report NE-523-A71-0594-A (Reference A.3-1), and (b) system modifications, as described in FSAR Section 5.3.3.1.4.5 , to mitigate cracking. The program specifies periodic ultr asonic inspection of critical regi ons of the feedwater nozzles. The BWR Feedwater Nozzle Program credits por tions of the Inservice Inspection (ISI)
Program. A.2.1.7 BWR PENETRATIONS PROGRAM The BWR Penetrations Program is an existing condition monitoring pr ogram that manages cracking due to SCC or intergranular stress corrosion cracking (IGSCC) of reactor vessel instrument penetrations, jet pump instrument penetrations, control rod drive penetrations, and incore instrument penetrations. The BWR Penetrations Program detects and sizes cracks in accordance with the guidelin es of approved Boiling Water Reactor Vessel and Internals Project (BWRVIP) documents and the requirements of the ASME Boiler and Pressure Vessel Code,Section XI. The BWR Water Chemistry Program monitors and controls reactor coolant water chemistry in accordance with BWRV IP guidelines to ensure the long-term integrity and safe operation of the vessel components.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-5 The program credits portions of the Inservice Inspection (ISI) Program and the BWR Vessel Internals Program.
A.2.1.8 BWR STRESS CORROSION CRACKING PROGRAM The BWR Stress Corrosion Cracking Program is an existing condition mon itoring program that manages cracking due to SCC/IGA for stainless steel and nickel alloy reactor coolant pressure boundary piping, nozzle safe e nds, nozzle thermal sl eeves, valve bodies, flow elements, and pump casings. The BWR Stress Corrosion Cracking Program consis ts of (a) preventive measures to mitigate SCC/IGA, and (b) inspection and flaw evaluatio n to monitor SCC/IGA and its effects. The BWR Water Chemistry Program monitors and controls reactor coolant water chemistry in accordance with BWRVIP guidelines to ensure the long-term mitigation of SCC/IGA. The program includes the scope of the Generic Letter 88-01 program, as modified by the staff-approved BWRVIP-75 report. The program credits portions of the Inservice Inspection (ISI) Program and the BWR Water Chemistry Program.
A.2.1.9 BWR VESSEL ID ATTACHMENT WELDS PROGRAM The BWR Vessel ID Attachment Welds Program is an existing program that manages cracking due to SCC/IGA of the welds for internal attachments to the reactor vessel. The BWR Vessel ID Attachment Welds Program performs examin ations and inspections as required by ASME Section XI, augmented by BWRVIP-48-A. These inspections include enhanced visual inspections with resolution to the guidelines in BWRVIP-03. The BWR Water Chemistry Program monitors and controls reactor coolant water chemistry in accordance with BWRVIP guidelines to ensure the long-term integrity and safe operation of the vesse l internal attachment welds. The BWR Vessel ID Attachment Welds Program credits portions of the BWR Vessel Internals Program and the Inservice Inspection (ISI) Program.
A.2.1.10 BWR VESSEL INTERNALS PROGRAM The BWR Vessel Internals Program is an exis ting condition monitoring program that manages cracking due to stress corrosion cracking and irradiation assisted stress corrosion cracking (SCC/IASCC), SCC/IG A, flaw growth, and flow-induced vibration for various components and subcomponents of the reactor vessel internals. The BWR Vessel Internals Program consists of mitigation, inspection, flaw evaluation, and repair in accordance with the guidelines C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-6 of BWRVIP reports and the requirements of the ASME Boiler and Pr essure Vessel Code,Section XI. The BWR Water Chemistry Program monitors and controls reactor coolant water chemistry in accordance with BWRV IP guidelines to ensure the long-term integrity and safe operation of the vessel internal components. The BWR Vessel Internals Program credits porti ons of the Inservice Inspection (ISI) Program.
A.2.1.11 BWR WATER CHEMISTRY PROGRAM The BWR Water Chemistry Program is an existing program that mitigates degradation of components that are within the sc ope of license renewa l and contain or are exposed to treated water, treated water in the steam phase, reactor coolant, or treated water in a sodium pentaborate solution. The progra m manages the relevant conditions that could lead to the onset and propagation of a loss of material due to corrosion or erosion, cracking due to SCC, or reduction in heat transfer due to fouling through proper monitoring and control of chemical concentrations consistent with BWRVIP water chemistry guidelines.
The BWR Water Chemistry Prog ram is supplemented by the Ch emistry Program Effectiveness Inspection and the Heat Exchangers Inspection, to provide verification of the effectiveness of the program in managing the effects of aging. Additionally, the BWR Water Chemistry Program is supplemented by the BWR Feedwa ter Nozzle Program, BWR Stress Corrosion Cracking Program, BWR Penetrations Program, BWR Vessel ID Attachment Welds Program, BWR Vessel Internals Program, Inservice Insp ection (ISI) Program, and Small Bore Class 1 Piping Program to provide verification of the program's effectiveness in managing the effects of aging for reactor pressure vessel, reactor vessel internals, and reactor coolant pressure boundary components.
A.2.1.12 CHEMISTRY PROGRAM EFFECTIVENESS INSPECTION The Chemistry Program Effectiveness Inspection detects and characterizes the condition of materials in representa tive low flow and stagnant areas of systems with water chemistry controlled by the BWR Water Chemistry Progra m or the Closed Cooling Water Chemistry Program, and with fuel oil chemistry contro lled by the Fuel Oil Chemistry Program. The inspection provides direct evidence as to whether, and to what ex tent, a loss of material due to corrosion has occurred. The inspection also determines whether cracking due to SCC of susceptible materials in suscep tible locations has occurred.
The Chemistry Program Effectiveness Inspection is a new one-time inspection that will be implemented prior to the period of extended operation. The inspection activities will be conducted within the 10-year period prio r to the period of extended operation.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-7 A.2.1.13 CLOSED COOLING WATER CHEMISTRY PROGRAM The Closed Cooling Water Chemistry Program m itigates degradation of components that are within the scope of license re newal and contain clos ed cooling water.
The program manages the relevant conditions th at could lead to the onset and propa gation of a loss of material due to corrosion or erosion, cracking due to SCC, or reduction in heat transfer due to fouling through proper monitoring and control of corrosion inhibitor concentra tions consistent with EPRI closed cooling water chemistry guidelines. The Closed Cooling Water Chemistry Program includes corrosion rate measurement in reactor building closed cooling water locations a nd is supplemented by the one-time Chemistry Program Effectiveness Inspection and Heat Excha ngers Inspection, which provide verification of the effectiveness of the program in managing the effects of aging.
The Closed Cooling Water Chemistry Program is an existing program that requires enhancement prior to the pe riod of extended operation.
A.2.1.14 COOLING UNITS INSPECTION PROGRAM The Cooling Units Inspection Program manages the effect of loss of material for aluminum, steel, copper alloy, and stainless steel cooling unit components that are exposed to condensation. The inspection also manages the effects of a reduc tion in heat transfer due to fouling of heat excha nger tubes and fins, or cracking due to SCC of aluminum components exposed to condensation.
The Cooling Units Inspection is a new program that will be implemented via baseline inspection of a sample popula tion followed by opport unistic inspections when components are opened for periodic maintenance, repair, and surveillance activities when surfaces are made available for inspection. Thes e inspections ensure that the existing environmental conditions are not causing material degrada tion that could result in a loss of component intended function during the period of extended ope ration. Inspection of a sample population will be conducted within the 10-year period prior to the period of extended operation and serve as a baseline for future inspections.
A.2.1.15 CONTROL ROD DRIVE RETURN LINE NOZZLE PROGRAM The Control Rod Drive Return Line (CRDRL) Nozzle Program is an existing mitigation and condition monitoring program that manages cracking due to flaw growth of the control rod drive return line nozzle, safe end, cap, a nd connecting welds. The CRDRL Nozzle Program consists of a) mitigation activi ties, and b) inspection, flaw ev aluation, and repair in accordance with the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWB, Table IWB C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-8 2500-1 (2001 Edition th rough 2003 Addenda) and the r ecommendations of NUREG-0619. System modifications were implemented by the original equipmen t manufacturer prior to initial startup to mitigate cracking. The BWR Wate r Chemistry Program mo nitors and controls reactor coolant water chemistry in accordance with BWRVIP guidelines to ensure the long-term integrity and safe operation of the critical regions of the CRDRL nozzle.
The CRDRL Nozzle Program credits portions of the Inservice Inspection (ISI) Program.
A.2.1.16 DIESEL STARTING AIR INSPECTION The Diesel Starting Air Inspection detects and characterizes the condition of materials for the DSA System air dryers and downstream piping and components (excluding the DSA System air receivers). The inspection provides direct evidence as to whether, and to what extent, a loss of material due to corrosion has occurred. The Diesel Starting Air Inspection is a new one-time inspection that will be implemented prior to the period of extended operation. The inspec tion activities will be c onducted within the 10-year period prior to the period of extended operation.
A.2.1.17 DIESEL SYSTEMS INSPECTION PROGRAM The Diesel Systems Inspection Program manages the effects of the loss of material due to corrosion and cracking due to stress corrosion cracking of materials for the interior of the steel and stainless steel exhaust piping for the Division 1, 2, a nd 3 diesels in the Diesel Engine Exhaust System, including the l oop seal drains from the exha ust piping. The inspection provides direct evidence as to whether, and to what extent, a loss of material due to corrosion has occurred.
The Diesel Systems Inspection is a new program that will be implemented via baseline inspection of a sample popula tion followed by opport unistic inspection when components are opened for periodic maintenance, repair, or surveillance activities when surfaces are made available for inspection. Thes e inspections ensure that the existing environmental conditions are not causing material degrada tion that could result in a loss of component intended function during the period of extended ope ration. Inspection of a sample population will be conducted within the 10-year period prior to the period of extended operation and will serve as a baseline for future inspections.
A.2.1.18 DIESEL-DRIVEN FIRE PUMPS INSPECTION PROGRAM The Diesel-Driven Fire Pumps Inspection Program manages the effects of the loss of material, due to corrosion or erosion, and reduction in heat transfer of the interior of the Fire Protection C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-9 System diesel engine exhaust piping, and of Fire Protection System diesel heat exchangers exposed to a raw water environment. The in spection also manages cracking due to SCC of susceptible materials.
The Diesel-Driven Fire Pumps Inspection is a new program that will be implemented via baseline inspection of a sa mple population followed by oppo rtunistic inspection when components are opened for periodi c maintenance, repair, or surveillance activities when surfaces are made available for inspection. These inspections ensu re that the existing environmental conditions are not causing material degradation that could result in a loss of component intended function during the period of extended operation. In spection of a sample population will be conducted within the 10-year period prior to the period of extended operation and will serve as a ba seline for future inspections.
A.2.1.19 ELECTRICAL CABLES AND CONNECTIONS NOT SUBJECT TO 10 CFR 50.49 EQ REQUIREMENTS PROGRAM The Electrical Cables and Connections Not Subject to 10 CFR 50.49 EQ Requirements Program is an inspection program that detects degradation of electrical cables and connections that are not environmentally qualified and are within the scope of license renewal. The program provides for periodic visual inspecti on of accessible, non- e nvironmentally qualified cables and connections in orde r to determine if age-relate d degradation is occurring, particularly in plant areas w ith adverse localized environments. An adverse localized environment is a condition in a limited plant area that is significantly more severe than the specified design or bounding plant environment for the general area. The Electrical Cables and Connections Not Subject to 10 CFR 50.49 EQ Requirements Program is a new aging management program that will be implemented prior to the period of extended operation. The inspection frequency of the program will be once every 10 years, with the initial inspection to be performed prior to the period of extended operation.
A.2.1.20 ELECTRICAL CABLES AND CONNECTIONS NOT SUBJECT TO 10 CFR 50.49 EQ REQUIREMENTS USED IN INSTRUMENTATION CIRCUITS PROGRAM The Electrical Cables and Connections Not Subject to 10 CFR 50.49 EQ Requirements Used in Instrumentation Circuits Program is a mon itoring program that detects degradation of electrical cables and connections that are not environmen tally qualified and used in circuits with sensitive, low-current applica tions (such as radiation monito ring and nuclear instrumentation loops). The program provides for a review of calibration records for the low-current instruments, in order to dete ct and identify degradation of the cable system insulation resistance. The program retains the option to perform direct cable testing.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-10 The Electrical Cables and Connections Not Subject to 10 CFR 50.49 EQ Requirements Used in Instrumentation Circuits Program is a ne w aging management program that will be implemented prior to the period of extended operation. The frequency of the program will be once every 10 years, with the initial review to be performed prior to the period of extended operation.
A.2.1.21 ELECTRICAL CABLE CONNECTIONS NO T SUBJECT TO 10 CFR 50.49 EQ REQUIREMENTS INSPECTION The Electrical Cable Connections Not Subject to 10 CFR 50.49 EQ Requirements Inspection detects and characterizes the ma terial condition of me tallic electrical conn ections within the scope of license renewal. Th e inspection uses thermography (augmented by contact resistance testing) to detect loose or degraded connections that lead to increased resistance for a representative sample of metallic electrical connections in vari ous plant locations. The Electrical Cable Connections Not Subject to 10 CFR 50.49 EQ Requi rements Inspection is a new one-time inspection that will be implemented prior to the period of extended operation.
The inspection activities will be conducted within the 10-year period prior to the period of extended operation.
A.2.1.22 EQ PROGRAM Environmental qualification (EQ) analyses for electri cal components with a qualified life of 40 years or greater are identified as TLAAs; therefore, the effects of aging must be addressed for license renewal. NRC regulation 10 CFR 50.49, "Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants," requires licensees to identify electrical equipment covered under this regulation and to maintain a qua lification file demonstrating that the equipment is qualified for its application and will perform its safety function up to the end of its qualified life. The EQ Program is an existing program that implements the requirements of 10 CFR 50.49 (as further defined by the Division of Operating Reactor Guidelines, NUREG-0588, and Regulatory Guide 1.89 Revision 1). In accordance with 10 CFR 54.21(c)(1)(iii), the EQ Program will be used to manage the effects of aging on the intended functions of the components associated with EQ TLAAs for the period of extended operation, because equipment will be repl aced prior to reaching the end of its qualified life. Reanalysis addresses attributes of analytical methods , data collection and reduction methods, underlying assumptions, accep tance criteria, and corrective actions if acceptance criteria are not met. Reanalysis of aging evaluations to extend the qualification of C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-11 components is performed on a routine basis pursuant to 10 CFR 50.49(e) as part of the Columbia EQ Program.
A.2.1.23 EXTERNAL SURFACES MONITORING PROGRAM The External Surfaces M onitoring Program consists of obs ervation and surveillance activities intended to detect degradation resulting from loss of material due to corrosion and cracking due to SCC for mechanical components, as well as hardening and loss of strength for elastomers. The External Surfaces Monitoring Progr am is a condition-monitoring program.
The External Surfaces M onitoring Program is an existing program that requires enhancement prior to the period of extended operation.
A.2.1.24 FATIGUE MONITORING PROGRAM Fatigue evaluations for mechanic al components are identified as TLAAs; therefor e, the effects of fatigue have been addressed for license renewal.
Energy Northwest monitors fati gue of various components (inc luding ASME Class 1 reactor coolant pressure boundary, high energy line break locations, and Primary Containment) via the Fatigue Monitoring Program, which tracks transient cycles and calculates fatigue usage. Energy Northwest has assessed the impact of th e reactor coolant environment on the sample of critical components identified in NUREG/
CR-6260 and other limiti ng components beyond those locations identified in NUREG/CR-6260. Ca lculation of fatigue usage values is not required for non-Class 1 SSCs. Instead, stress intensifica tion factors and lower stress allowables are used to ensure compone nts are adequately designed for fatigue. In accordance with 10 CFR 54.21(c)(1)(iii), the Fatigue Monitori ng Program will be used to manage the effects of aging due to fatigue on the intended functions of the components associated with fatigue TLAAs for the period of extended operation.
The Fatigue Monitoring Program is an existing program that requires enhancement prior to the period of extended operation.
A.2.1.25 FIRE PROTECTION PROGRAM The Fire Protection Program is an existing program, described in Appendix F of the FSAR, that detects degradation of components in the scope of license re newal that have fire barrier functions. Periodic visual inspect ions and functional te sts are performed of fire dampers, fire barrier walls, ceilings and floors, fire-rated penetration seals, fire wraps, fire proofing, and fire doors to ensure that functionality and operability are mainta ined. In addition, the Fire C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-12 Protection Program supplements the Fuel Oil Chemistry Program and External Surfaces Monitoring Program through performance monitoring of the diesel-drive n fire pump fuel oil supply components and testing and inspection of the halon and carbon dioxide suppression systems, respectively. The Fire Protection Program is a condition monitoring program, comprised of tests and inspec tions based on National Fire Protection Association (NFPA) recommendations.
A.2.1.26 FIRE WATER PROGRAM The Fire Water Program (sub-prog ram of the overall Fire Protec tion Program) is described in Appendix F of the FSAR, and is credited with mana ging loss of material due to corrosion, erosion, macrofouling, and selective leaching, cracking due to SCC/IGA of susceptible water-based fire suppression components in the scope of license rene wal. Periodic inspection and testing of the water-based fire suppression systems provides reasonable assura nce that the systems will remain capable of performing their intended func tion. Periodic inspection and testing activities include hydrant and hose station inspections, fire main flushing, flow tests, and sprinkler inspections. The Fire Water Program is a condition monitoring program, comprised of tests and inspecti ons based on NFPA recommendations. The Fire Water Program is an existing program that requires enhancement prior to the period of extended operation.
A.2.1.27 FLEXIBLE CONNECTION INSPECTION PROGRAM The Flexible Connection Inspection Program manages degradation, including the effects of the loss of material due to wear and hardening and loss of strength of elastomer components exposed to treated water, dried air, gas, and indoor air environments. The Flexible Connection Inspection Program is a new plant-specific program that will be implemented via baseline in spection of a sample populatio n followed by opportunistic inspection when components are opened for peri odic maintenance, repair, or surveillance activities when surfaces are made available for inspection.
These inspections ensure that the existing environmental conditions are not causing material degradation that could result in a loss of component intended function during the period of extended operation. Inspection of a sample population will be conducted within the 10-year period prior to the period of extended operation and will serve as a ba seline for future inspections.
A.2.1.28 FLOW-ACCELERATED CORROSION (FAC) PROGRAM The Flow-Accelerated Corrosion (FAC) Program ma nages loss of material for steel and gray cast iron components located in the treated water environment of sy stems that are susceptible to C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-13 FAC, also called erosion-corrosion. The FAC Program combines the elements of predictive analysis; inspections (to baseline and monitor wall-thinning), industry experience, station information gathering and communication, and engineering judgm ent to monitor and predict FAC wear rates. The program is a condition monitoring program that implements the recommendations of NRC Generic Letter 89-08, and follows the guidance and recommendations of EPRI NSAC-202L (Reference A.3-2), to ensure that the integrity of piping systems susceptible to FAC is maintained. The FAC Program is an existing program that requires enhancement prior to the period of extended operation.
A.2.1.29 FUEL OIL CHEMISTRY PROGRAM The Fuel Oil Chemistry Program is an existing program that maintains fuel oil quality in order to mitigate degradation of the storage tanks and associated components c ontaining fuel oil that are within the scope of license renewal. The program includes diesel fuel oil testing for emergency diesel generator and diesel-driven fire pump fuel. The Fuel Oil Chemistry Program manages the relevant conditions that could lead to the onset and propagation of loss of material due to corrosion, or cracking due to SCC of susceptib le copper alloys, through proper monitoring and control of fuel oil contamination consistent with plant technical specifications and American Society for Testing and Materials (ASTM) standards for fuel oil. The relevant conditions are specific contaminants such as water or microbiologi cal organisms in the fuel oil that could lead to corrosion of susceptible mate rials. Exposure to these contaminants is minimized by verifying the quality of new fuel oil before it enters the emergency diesel generator storage tanks and by periodic sampling to ensure th at both the emergency diesel generator tanks and fire protec tion tanks are free of water and particulates. The Fuel Oil Chemistry Program is a mitigation program.
The Fuel Oil Chemistry Program is supplem ented by the Chemistry Program Effectiveness Inspection, which provides veri fication of the effec tiveness of the program in mitigating the effects of aging.
A.2.1.30 HEAT EXCHANGERS INSPECTION The Heat Exchangers Inspection detects and characterizes the surface conditions with respect to fouling of heat exchangers and coolers that are in the scope of the inspection and exposed to indoor air or to water with the chemistry c ontrolled by the BWR Water Chemistry Program or the Closed Cooling Water Chemis try Program. The inspection pr ovides direct evidence as to whether, and to what extent, a reduction of heat transfer due to fouling has occurred on the heat transfer surfaces of heat exchangers and coolers.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035.12-020 A.2-14 The Heat Exchangers Inspection is a new one-time inspection that will be implemented prior to the period of extended operation. The inspectio n activities will be conducted within the 10-year period prior to the period of extended operation.
A.2.1.31 HIGH-VOLTAGE PORCELAIN INSULATORS AGING MANAGEMENT PROGRAM The High-Voltage Porcelain Insulators Aging Management Program is an existing program that manages the build-up of contamination (har d water residue) on the surfaces of the 115-kV high-voltage insulators located in the transformer yard and th e 230-kV high voltage insulators located in the Ashe substation.
The program provides for pe riodic cleaning or recoating of insulators and visual inspecti on of the coating (if present) on the high-voltage station post insulators between the 115-kV backup transfor mer and circuit breaker E-CB-TRB located in the station transformer yard. Te sting for contamination, and cleaning if required, is conducted on the high voltage station post insulators between the 230-kV overhead line running to Columbia and circuit breaker E-CB-TRS , located in the Ashe substation.
The High-Voltage Porcelain Insulators Aging Management Program is a preventive maintenance program cons isting of activities to mitigate poten tial degradation of the insulation function due to hard wate r deposits. Uncoated insulators located in the transformer yard are inspected and cleaned every two years. Coated insulators are visually inspected for damage every two years and are re-coated every 10 years. The program requires enhancement prior to the period of extended operation to have the insulators located in the Ashe substation tested for contamination, and cleaned if required, every 8 years.
A.2.1.32 INACCESSIBLE POWER CABLES NOT SUBJECT TO 10 CFR 50.49 EQ REQUIREMENTS PROGRAM The Inaccessible Power Cables Not Subject to 10 CFR 50.49 EQ Requirements Program will manage the aging of in-scope, power cables ( 400V) exposed to signifi cant moisture. First tests or first inspection for license renewal will be completed before the period of extended operation. These cables will be tested at least once every 6 year s to provide an indication of the condition of the conductor insulation. Th e specific type of test performed will be determined prior to the initial test, and is to be a proven test for detecting deterioration of the insulation system due to wetting, such as power factor, part ial discharge, or polarization index, as described in EPRI TR-103834-P1-2 (Reference A.3-3), or other testing that is state-of-the-art at the time the test is performed. Significant mo isture is defined as periodic exposures that last more than a few days (e.g., cable in sta nding water). Pe riodic exposures that last less than a few days (e.g., normal rain and drain) ar e not significant. In addition, inspection for water collection in electrical manholes will be performed based on actual plant experience with water accumulation in the manholes. However, the inspection frequency will C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-15 be at least annually. Manhole inspection will al so be performed periodi cally, in response to event-driven occurrences (such as heavy rain or flooding). Th e inspection will include direct observation that cables are not wetted or submerged, that cables/splices a nd cable support structures are intact, and sump pump systems and associated alarms operate properly. In addition, sump pumps will be inspected and opera tion verified prior to any known or predicted heavy rain or flooding events which could require the sump pump to operate.
A.2.1.33 INSERVICE INSPECTI ON (ISI) PROGRAM The Inservice Inspection (ISI)
Program is an existing condition monitoring program that manages cracking due to SCC/IGA and flaw growth of multiple reactor coolant system pressure boundary components, including the reactor vessel, a limited number of internals components, and the reactor coolant system pr essure boundary. The In service Inspection (ISI)
Program also manages loss of material due to corrosion for reactor vesse l internals components and reduction of fracture toughness due to thermal embrittlement of cast austenitic stainless steel pump casings and valve bodies.
The Inservice Inspection (ISI) Pr ogram details the requirements for the examination, testing, repair, and replacement of components specified in ASME Section XI for Class 1, 2, or 3 components. The Inservice Inspection (ISI) Program complies with the ASME Code requirements. The program scope has been augmented to incl ude additional requireme nts, and components, beyond the ASME requirements. Examples incl ude the augmentation of ISI to expand reactor vessel feedwater nozzle examinations, examina tions of high energy line piping systems that penetrate containment, examina tions per Generic Lett er 88-01, and examinations of shroud support plate access hole covers pe r BWRVIP guidance. Such augm entation is consistent with the ISI program description in NUREG-1801,Section XI.M1.
A.2.1.34 INSERVICE INSPECTION (ISI) PROGRAM - IWE The Inservice Inspection (ISI) Program - IWE is an existi ng program that establishes responsibilities and requirem ents for conducting IWE in spections as required by 10 CFR 50.55a. The Inservice Inspection (ISI) Program - IWE includes visual examination of all accessible surface areas of the steel containment and its integral attachments, and containment pressure-retaining bolting in accordance with the requirements of the ASME Boiler and Pressure Ve ssel Code,Section XI, Subsection IWE. The inservice examinations conducted throughout the service life of Columbia will comply with the requirements of the ASME Section XI Edition and Addenda inco rporated by reference in 10 CFR 50.55a(b) twelve months prior to the st art of the inspection interval, subject to prior C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-16 approval of the edition and addenda by the NRC. This is consis tent with NRC statements of consideration for 10 CFR 54 associated with the adoption of new editions and addenda of the ASME Code in 10 CFR 50.55a.
A.2.1.35 INSERVICE INSPECTION (ISI) PROGRAM - IWF The Inservice Inspection (ISI) Program - IWF is an existing progra m that establishes responsibilities and requirements for conducting IWF Inspections for ASME Class 1, 2, and 3 component supports as required by 10 CFR 50.55a. The In service Inspection (ISI) Program -
IWF performs visual examination of supports based on sampling of the total support population. The sample size varies depending on the ASME Class. The largest sample size is specified for the most critical supports (ASME Class 1 and those other than piping supports (Class 1, 2, 3, and MC)). Th e sample size decreases for the less critical supports (ASME Class 2 and 3). The primary inspection method employed is visual exam ination. Degradation that potentially compromises suppo rt function or load capacity is identified for evaluation. Supports requiring corrective actions are re-e xamined during the next inspection period in accordance with the requirements of the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWF. The inservice examinations conducted throughout the service life of Columbia will comply with the requirements of the ASME Section XI Edition and Addenda inco rporated by reference in 10 CFR 50.55a(b) twelve months prior to the st art of the inspection interval, subject to prior approval of the edition and addenda by the NRC. This is consis tent with NRC statements of consideration for 10 CFR 54 associated with the adoption of new editions and addenda of the ASME Code in 10 CFR 50.55a.
A.2.1.36 LUBRICATING OIL ANALYSIS PROGRAM The Lubricating Oil Analysis Pr ogram manages loss of material due to corrosion or selective leaching of susceptible material s and reduction of heat transfer due to fouling for plant components that are within the scope of license re newal and exposed to a lubricating oil environment. The Lubricating Oil Analysis Program is a mitigation program.
The Lubricating Oil Analysis Program is s upplemented by the Lubricating Oil Inspection, which provides verification of the effectiveness of the program in mitigating the effects of aging. The Lubricating Oil Analysis Program is an existing program that requires enhancement prior to the period of extended operation.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-17 A.2.1.37 LUBRICATING OIL INSPECTION The Lubricating Oil Inspection detects and characterizes the condition of materials in systems and components for which the Lubricating Oil Analysis Program is credited with aging management. The inspection provides direct evidence as to whethe r, and to what extent, a loss of material due to corrosion or selective leaching has occurred. The inspection also determines whether a reduction in heat transf er due to fouling has occurred.
The Lubricating Oil Inspection is a new one-time inspection that will be implemented prior to the period of extended operation. The inspectio n activities will be conducted within the 10-year period prior to the period of extended operation.
A.2.1.38 MASONRY WALL INSPECTION The Masonry Wall Inspection consists of inspection activities to detect cracking of masonry walls within the scope of license renewal. Ma sonry walls that perform a fire barrier intended function are also managed by the Fire Protect ion Program. The Masonry Wall Inspection is implemented as part of the Structures Mon itoring Program. The Masonry Wall Inspection performs visual inspection of exte rnal surfaces of masonry walls. The Masonry Wall Inspection is an existing prog ram that requires enhancement prior to the period of extended operation.
A.2.1.39 MATERIAL HANDLING SYSTEM INSPECTION PROGRAM The Material Handling System Inspection Program manages loss of material for cranes (including bridge, trolley, rails, and girders), m onorails, and hoists within the scope of license renewal. The Material Handling System Inspect ion Program is based on guidance contained in ANSI B30.2 for overhead and gantry cran es, ANSI B30.11 for monorail systems and underhung cranes, and ANSI B30.16 for overhead hoists.
A.2.1.40 METAL-ENCLOSED BUS PROGRAM The Metal-Enclosed Bus Program is an inspection program that detects degradation of metal-enclosed bus within the scope of license renewal. The program provides for the visual inspection of interior sections of bus, and an inspection of the elastomeric seals at the joints of the duct sections. The program also makes pr ovision for thermographic inspection of bus bolted connections. The Metal-Enclosed Bus Program is a new aging management program that will be implemented prior to the period of extended operation. The ther mography portion of the C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-18 program will be performed once ev ery 10 years, with the initial inspections to be performed prior to the period of extended operation. The visual inspec tion portion of the program will also be performed once every 10 years, with the first inspections to be performed prior to the period of extended operation.
A.2.1.41 MONITORING AND COLLECTION SYSTEMS INSPECTION PROGRAM The Monitoring and Collection Systems Inspection Program manages the effects of the loss of material due to corrosion or er osion for the internal surfaces of subject mechanical components that are exposed to equipment or area draina ge water and ot her potential contaminants and fluids. The inspection also manages cracking due to SCC of susceptible materials.
The Monitoring and Collection Systems Inspec tion Program is a program that will be implemented via baseline in spection of a sample populatio n followed by opportunistic inspection when components are opened for peri odic maintenance, repair, or surveillance activities when surfaces are made available for inspection.
These inspections ensure that the existing environmental conditions are not causing material degradation that could result in a loss of component intended function during the period of extended operation. Inspection of a sample population will be conducted within the 10-year period prior to the period of extended operation and will serve as a ba seline for future inspections.
A.2.1.42 OPEN-CYCLE COOLING WATER PROGRAM The Open-Cycle Cooling Water Program manages loss of mate rial due to corrosion and erosion for components located in the Standby Se rvice Water and Plant Se rvice Water systems, and for components connected to or serviced by those systems. The program manages fouling due to particulates (e.g., corrosion products) and biological material (micro- or macro-organisms) resulting in reduction in heat transfer for heat ex changers (including condensers, coolers, cooling coils, and evaporators) within the scope of the program. The Open-Cycle Cooling Water Program also manage s loss of material for components associated with the feed-and-bleed mode for emergency makeup water to the spray pond. The Open-Cycle Cooling Water Pr ogram consists of inspections, surveillances, and testing to detect the presence, and assess the extent of fouling and loss of material. The inspection activities are combined with chemical treatments and cleaning activities to minimize the effects of aging. The program is a combination c ondition monitoring and m itigation program that implements the recommendations of NRC Generic Letter 89-13 for safety-related equipment in the scope of the program. The scope of th e program also includes non-safety related components containing either service water or spray pon d makeup water.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-19 The Open-Cycle Cooling Water Program is an existing progra m that requires enhancement prior to the period of extended operation.
A.2.1.43 POTABLE WATER MON ITORING PROGRAM The Potable Water Monitoring Program is a m itigation program that, by means of chemical water treatment, manages loss of material due to corrosion a nd erosion for components that contain potable water.
The Potable Water Monitoring Program is an existing program that requires enhancement prior to the period of extended operation. At least one inspection will be conducted within the 10-year period prior to the period of extended operation.
A.2.1.44 PREVENTIVE MAINTENANCE - RCIC TURBINE CASING Preventive Maintenance - RCIC Turbine Casing is an existing program that manages loss of material due to corrosion for the reactor core isolation cooling (RCIC) pump turbine casing and associated piping components downstream from the steam admission valve. These components are exposed to steam during RCIC system operation and testing, but are empty during normal plant operating conditions. Prev entive Maintenance - RC IC Turbine Casing is a condition monitoring program comprised of periodic inspection and surveillance activities to detect aging and age-related degradation.
A.2.1.45 REACTOR HEAD CLOSURE STUDS PROGRAM The Reactor Head Closure Studs Program is an existing program that manages cracking due to SCC and loss of material due to corrosion for the reactor head closure stud assemblies (studs, nuts, washers, and bushings). The Reactor Head Closure Studs Program examines reactor vessel stud assemblies in accordance with the examination and inspection requirements specified in the ASME Boiler and Pressure Ve ssel Code,Section XI, Subsection IWB (edition and addenda described in the Inservice In spection (ISI) Program), Table IWB 2500-1. The Reactor Head Closure Studs Program include s preventive measures in accordance with Regulatory Guide 1.65 to mitigate cracking.
The Reactor Head Closure Studs Program cred its portions of the Inservice (ISI) Inspection Program. A.2.1.46 REACTOR VESSEL SURVEILLANCE PROGRAM The Reactor Vessel Surveillance Program is an existing condition monitoring program that manages reduction of fracture toughness due to radiation embrittle ment for the low alloy steel C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-20 reactor vessel shell and welds in the beltline regi on. The Reactor Vessel Surveillance Program incorporates the BWRVIP Inte grated Surveillance Program (IS P), as described in reports BWRVIP-86-A and BWRVIP-116. Energy Northwest follows the re quirements of the BWRVIP ISP and applies the ISP data to Columbia. The NRC has approved the use of the BWRVIP ISP in place of a unique plant program for Columbia. The provisions of 10 CFR 50 Appendix G require Columbia to operate within the currently licensed pressure-temperature (P-T) limit curves, and to update these curves as necessary. The P-T limit curves, as contained in plant technical specifications, will be updated as necessary through the period of ex tended operation as part of the Re actor Vessel Surveillance Program. Reactor vessel P-T limits will thus be mana ged for the period of extended operation.
A.2.1.47 SELECTIVE LEACHING INSPECTION The Selective Leaching Inspection detects and characterizes the conditions on internal and external surfaces of subject com ponents exposed to raw water, trea ted water, fuel oil, soil, and moist air (including condensation) environmen ts. The inspection provides direct evidence through a combination of vi sual examination and hardne ss testing, or NRC-approved alternative, as to whether, and to what extent, a loss of materi al due to selective leaching has occurred.
The Selective Leaching In spection is a new one-time inspecti on that will be implemented prior to the period of extended operation. The inspec tion activities will be c onducted no earlier than 5 years prior to the peri od of extended operation.
A.2.1.48 SERVICE AIR SYSTEM INSPECTION PROGRAM The Service Air System Inspection Program manages the effects of the loss of material due to corrosion of steel piping and valve bodies exposed to an "air (int ernal)" (i.e., compressed air) environment within the license renewal boundary of the Service Air System.
The Service Air System Inspection Program is a new plant-specific program that will be implemented via baseline in spection of a sample populatio n followed by opportunistic inspection when components are opened for peri odic maintenance, repair, or surveillance activities when surfaces are made available for inspection.
These inspections ensure that the existing environmental conditions are not causing material degradation that could result in a loss of component intended function during the period of extended operation. Inspection of a sample population will be conducted within the 10-year period prior to the period of extended operation and will serve as a ba seline for future inspections.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-21 A.2.1.49 SMALL BORE CLASS 1 PIPING PROGRAM The Small Bore Class 1 Piping Program will de tect and characterize cracking of small bore Class 1 piping components that are exposed to reactor coolant. This periodic program will provide physical evidence as to whether, and to what extent, cracking due to SCC or to thermal or mechanical loading has occurred in small bore Class 1 piping components. The Small Bore Class 1 Piping Progra m will be a condition monitoring program with no actions to prevent or mitigate agi ng effect. The program will include visual and volumetric inspection of a representative sample of small bore Class 1 piping, including bu tt welds and socket welds.
The Small Bore Class 1 Piping Program is a new program that will be implemented prior to the period of extended operation.
Inspection activities will start during the fourth 10-year inservice inspection interval and continue through the period of extended operation. The Small Bore Class 1 Piping Inspection w ill credit portions of the Inservice Inspection (ISI) Program.
The Small Bore Class 1 Piping Inspection will verify the effectiveness of the BWR Water Chemistry Program in mitigating cracking of small bore piping and piping components.
A.2.1.50 STRUCTURES MONITORING PROGRAM The Structures Monitoring Progra m manages age-related degrada tion of plant structures and structural components within its scope to ensure that each stru cture or structural component retains the ability to perform its intended function. Aging e ffects are detected by visual inspection of external surfaces prior to the loss of the stru cture's or component's intended function. The Structures Monitoring Program encompasses and implements the Water Control Structures Inspection and the Masonry Wall Insp ection. This program implements provisions of the Maintenance Rule, 10 CFR 50.65, that relate to structures, masonry walls, and water control structures. C oncrete and masonr y walls that perform a fire barrier intended function are also managed by the Fire Protection Program. The Structures Monitoring Program is an existing program that requires enhancement prior to the period of extended operation.
A.2.1.51 SUPPLEMENTAL PIPING/TANK INSPECTION The Supplemental Piping/Tank Inspection detects and characterizes the material condition of steel, gray cast iron, and stainless steel components exposed to moist air environments. The inspection provides direct evidence as to whether, and to what ex tent, a loss of material due to corrosion has occurred.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-22 The Supplemental Piping/Tank Inspection is a new one-time inspection that will be implemented prior to the period of extended operation. The inspection activities will be conducted within the 10-year period prio r to the period of extended operation.
A.2.1.52 THERMAL AGING AND NEUTRON EMBRITTLEMENT OF CAST AUSTENITIC STAINLESS STEEL (CASS) PROGRAM The Thermal Aging and Neutron Embrittlement of Cast Austenitic Stainless Steel (CASS)
Program will manage reduction of fracture toughness due to therma l aging and neutron irradiation embrittlement of CA SS reactor vessel internals. The program includes: (a) identification of susceptible compone nts determined to be limiting from the standpoint of thermal aging or neutron irradiation embrittlement (neutron fluence), (b) a component-specific evaluation to determine each identified component's susceptibility to reduction of fracture toughness, and (c) a supplemental examin ation of any component not eliminated by the component-specific evaluation. The program credits portions of the Inservice Inspection (ISI) Program and the BWR Vessel Internals Program.
The Thermal Aging and Neutron Embrittlement of Cast Austenitic Stainless Steel (CASS)
Program is a new aging management program that will be implemented prior to the period of extended operation.
A.2.1.53 WATER CONTROL STRUCTURES INSPECTION The Water Control Structures In spection, implemented as part of the Structures Monitoring Program, consists of inspection activities to detect aging and age-related degradation. The Water Control Structures Inspection ensures the structural integrity and operational adequacy of the spray ponds, standby service water pump houses, circulating water pump house (including circulating water basi n), makeup water pump house, cooling towe r basins, and those structural components w ithin the structures.
The Water Control Structures Inspection is an existing program that requires enhancement prior to the period of extended operation. A.2.1.54 BORON CARBIDE MONITORING PROGRAM The Boron Carbide Monitoring Program detects degradatio n of the Boron Carbide (B 4 C) neutron absorbers in the spent fuel storage racks by monitoring spent fuel racks for potential off-gassing, by in situ testing of the sp ent fuel racks, or by inspecting the B 4 C coupons. From C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035,12-020 A.2-23 the monitoring data, the stability and integrity of Boron Carb ide in the storage cells are assessed. Periodic monitoring of B 4 C coupons permits early de termination of aging degradation, but may be discontinued based on in situ testing results. A.2.1.55 SERVICE LEVEL 1 PR OTECTIVE COATINGS PROGRAM The Service Level 1 Protective Coating Program monitors the performance of Service Level 1 coatings inside containment th rough periodic coating examina tions, condition assessments, and remedial actions, includi ng repair or testing. The program establishes roles, responsibilities, controls and deliverables for the Service Level 1 Protective Coatings Pr ogram. This program also ensures the Design Basis Accident (DBA) analys is limits with regard to coating will not be exceeded for the suction strainers.
A.2.2 EVALUATION OF TIME-LIMITED AGING ANALYSES In accordance with 10 CFR 54.21(c
), an application for a renewed operating license requires an evaluation of TLAAs for the period of ex tended operation. Th e following TLAAs have been identified and evaluated to meet this requirement.
A.2.2.1 REACTOR VESSEL NEUTRON EMBRITTLEMENT Neutron embrittlement is the change in mechanical properties of reactor vessel materials resulting from exposure to fast neutron flux (E>1.0 MeV) in the beltline region of the reactor core. The most pronounced mate rial change is a reduction in fr acture toughness. As fracture toughness decreases with cumulative fast neutron exposure, the material's resistance to crack propagation decreases. Fracture toughness is also dependent on temperature. The reference temperature for nil-ductility transition (RT NDT) is the temperature a bove which the material behaves in a ductile manner and below which the material behaves in a brittle manner. As fluence increases, RTNDT increases, and higher temperatures are required for the material to continue to act in a ductile manner.
Requirements associated with fracture toughness, pressure-tempe rature limits, and material surveillance programs for the reactor coolant pressure boundary are cont ained in Appendices G and H of 10 CFR 50. The analyses associated with evaluation of the effect of neutron embrittlement on the reactor pressure vessel for 40 years are TLAAs. Neutron fluence, upper shelf energy, adjusted reference temperature (ART), and vessel P-T limits are time dependent parameters associated with fracture toughness (embrittlemen t) of reactor vessel materials.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-24 A.2.2.1.1 Neutron Fluence EFPY Projection To evaluate the effects of radiation on reactor pressure vessel material embrittlement, the results of analyses were projected to determine neutron fluence out to 54 effective full power years (EFPY). Using actual reactor core power histories through 2007 a nd conservative estimates of future core desi gns, extended operation to 60 years was determ ined to be bounded by 54 EFPY. Fluence Projection Analyzed fluence values at 51.6 EFPY of reactor operation are addressed in FSAR Section 4.3.2.8 and FSAR Table 4.3-1. These fluence analyses are based on the original licensed thermal power of 3323 mega-watt thermal (MWt) through fuel cycle 10, and the currently licensed thermal power uprated to 3486 MWt from cycle 11 through the end of operation.
These fluence analyses use NRC-approved methodolo gy based on the guid ance of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence." The fluence analyses were projected to 54 EFPY for the extended operating period of 60 years.
Beltline Evaluation For the extended operating period, all ferritic materials for vessel beltline shells, welds, nozzles and the associated nozzle to vessel welds, and assembly components are required to be evaluated for neutron irradiation embrittlement if high energy neut ron fluence is greater than a threshold value of 1E+17 n/cm 2 (E >1 MeV) at the end of the 60 years. The only vessel assembly items, other than the shells and weld s of the beltline region that would experience neutron fluence greater than 1E+17 n/cm 2 during the period of extended operation are instrumentation nozzle N12 and residual heat removal/low pressure coolant injection (RHR/LPCI) nozzle N6 (and the associated nozzle-to-vessel welds).
Instrumentation nozzle N12 has a thickness less than 2.5 inches and was not originally evaluated for fracture toughness per ASME Code Appendix G, Section G2223. Nozzle N12 is not limiting for P-T curves as discussed in Section A.2.2.1.4; however, as nozzle N12 was evaluated for impact on the P-T curves it meets the definition of a beltline component per 10 CFR 50, Appendix G. The asso ciated nozzle-to-vessel weld is an austenitic weld and, therefore, is not subject to the fracture toughness requirements of 10 CFR 50, Appendix G. Nozzle N6 is included in the evaluation for USE in Section A.2.2.1.2. Nozzle N6 is evaluated for ART in Section A.2.2.1.3 below. Nozzle N6 is not the limiting material for the vessel. However, as nozzle N6 was evaluated for ART it meets the definition of a beltline component C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-25 per 10 CFR 50, Appendix G. The associated no zzle-to-vessel weld is a ferritic weld and, therefore, is subject to the fracture toughness requirements of 10 CFR 50, Appendix G. The nozzle-to-vessel weld for nozzle N6 is also included in the evaluation for USE in Section A.2.2.1.2 and is evaluated for ART in Section A.2.2.1.3.
The beltline definition for the period of exte nded operation includes the lower shell (Course #1 / Ring #21), lower-intermediate shell (Course #2 / Ring #22), asso ciated vertical (longitudinal) welds, the girth (circumferential) weld that connects the lower and lower-intermediate shells, and nozzles N6 (and its associated nozzle-to-vessel weld) and nozzle N12.
Disposition Neutron fluence has been projected to the end of the period of extended operation.
A.2.2.1.2 Upper Shelf Energy Evaluation Appendix G of 10 CFR 50 requires the upper shelf energy (USE) of the vessel beltline materials to remain above 50 ft-l b at all times during plant oper ation, including the effects of neutron radiation. If USE cannot be shown to remain above th is limit, then an equivalent margin analysis (EMA) must be performed to sh ow that the margins of safety against fracture are equivalent to those required by Appendix G of Section XI of the ASME Code. The initial (unirradiated) USE is not known for all the Columbia vessel pl ates and welds. For those plates and welds for which the initial USE is known, USE was projected using Regulatory Guide 1.99, Revision 2 methods. For the vessel plat es and welds for which the initial USE is not known, USE equivalent margin analyses we re performed using the Boiling Water Reactor Owners Group (BW ROG) equivalent margin an alysis (EMA) methodology. Results from the testing and analysis of surveillance materials were used in the EMA analyses.
All of the projected USE values for the vesse l beltline plates, nozzle forgings, and welds for which the initial USE is known re main above 50 ft-lbs through the end of the period of extended operation (54 EFPY). For the vessel beltline plates and welds, for which the initial USE is not known, the maximum decrease in US E was found to be le ss than the assumed decrease in the associated generic equivalent margin analyses. The maximum predicted decreases in USE for 54 EFPY for these beltline plates and welds are bounded by the generic equivalent margin analyses. Therefore, the projected USE for the vessel beltline plates and welds is acceptable for the period of extended operation.
Energy Northwest agrees that all beltline materials, including the N12 instrumentation nozzles, must be considered when the licensee develops pressure-temperature limits for Columbia in accordance with 10 CFR Part 50, Appendix G a nd ASME Code, Secti on XI, Appendix G.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-26 Energy Northwest will continue to develop future pressu re-temperature limit curves considering all beltline plates, welds, and nozzles.
Disposition Upper shelf energy TLAAs have been projected to the end of the peri od of extended operation for all reactor vessel beltline materials. Additionally, a specif ic 54 EFPY equivalent margins analysis will be performed for the N12 nozzle forgings prior to the period of extended operation.
A.2.2.1.3 Adjusted Reference Temperature Analysis In addition to USE, the other key parameter that characterizes the fracture toughness of a material is the RT NDT. This reference temper ature changes as a function of exposure to neutron radiation resulting in an adjusted reference temperature, ART.
The initial RT NDT is the reference temperature for the unirradiated material. The change due to neutron radiation is referred to as RT NDT. The ART is calculated by adding the initial RT NDT , the RT NDT, and a margin to account for uncertainties as prescribed in Regulatory Guide 1.99, Revision 2. The ART evaluations of record for the vessel beltline plates, nozzl e forgings, and welds for the currently licensed period (33.1 EFPY) include power uprate conditions. Based on projected fluence values, the methodology in Regulatory Guide 1.99 was us ed to project the ART for 54 EFPY. The ART values projected to 54 EFPY are used to develop P-T limit curves.
Projected ART values are well below the 200°F end of life ART suggested in Section 3 of Regulatory Guide 1.99 and are, thus, acceptable for the period of exte nded operation.
Disposition Reactor vessel adjusted reference temperature TLAAs have been projected to the end of the period of extended operation.
A.2.2.1.4 Pressure-Temperature Limits
To ensure that adequate margins of safety are maintained for various modes of reactor operation, 10 CFR 50, Appendix G specifies pressure a nd temperature requirements for affected materials for the service life of the reactor vessel. The basis for these fracture toughness requirements is ASME Section XI, Appendix G.
The ASME Code requires P-T limits be established for hydrostatic pressure tests and leak tests; for operation with the core not critical during heatup and cooldown; and for co re critical operation.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-27 The Columbia P-T lim it curves were revised in 2005 to incl ude the effects of power uprate to 3486 MWt. The P-T limits are valid for 33.1 EF PY through the end of the currently licensed period. The curves were reviewed in 2009 to assure that the N12 in strumentation nozzle did not affect the existing curves. P-T limits for the period of extended operation will be calculated using the most accurate fluence projections available at the time of the recalculation.
The projections may be adjusted if there are changes in core de sign or if additional surveillance capsule results show the need for an adjustment. The projected ART for the period of extended operation above gives confidence that future P-T cu rves will provide adequate operating margin.
Energy Northwest will continue to develop pressure-temperature limits in accordance with the Title 10 of the Code of Federal Regulations Part 50, Appendi x G (10 CFR Part 50, Appendix G) and ASME Code,Section XI, Appendix G, considering all beltline plates, welds, and nozzles.
License amendment requests to revise the P-T limits will be submitted to the NRC for approval, when necessary to co mply with 10 CFR 50 Appendix G, as part of the Reactor Vessel Surveillance Program.
Disposition The TLAA for P-T limits will be adequately managed for the pe riod of extended operation as part of the Reactor Vessel Surveillance Program.
A.2.2.1.5 Reactor Vessel Circumferential Weld Inspection Relief BWRVIP-74-A, "BWR Vessel and Internals Proj ect, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines for License Renewal," reiterated the recommendation of BWRVIP-05, "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations," that vessel circ umferential welds could be exempted from examination. The NRC safety evaluation report (SER) for BWRVIP-74 agreed, but required that plants apply for this re lief request individually. The re lief request is required to demonstrate that at the expiration of the current license, the circumferential welds will satisfy the limiting conditional failure probability in the (BWRVIP-05) evaluation. Energy Northwest requested and received permanent relief from vessel shell circumferential (girth) weld volumetric examinations through 33.1 EFPY.
The reactor pressure vessel circumferential weld parameters at 54 EFPY have been projected to remain within the bounding (64 EFPY) vessel parameters from the BWRVIP-05 SER. As such, the conditional probability of failure for circumferential welds remains below the limits contained in the SER for BWRVIP-05.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-28 Disposition The TLAA for reactor vessel circumferential weld examination relief has been projected to the end of the period of extended operation.
A.2.2.1.6 Reactor Vessel Axial We ld Failure Probability The NRC SER for BWRVIP-74-A evaluated the failu re frequency of axially oriented welds in BWR reactor vessels, and determined failure frequency acceptance criteria for 40 years of reactor operation. Applicants for license renewal are required to evaluate axially oriented vessel welds to show that their failure frequency remains below the acceptance criteria in the SER for BWRVIP-74. An acceptable way to do this is to show that the mean RTNDT of the limiting axial beltline weld at the end of the peri od of extended operation is less than the values specified in the SER. The Columbia limiting axial weld mean RT NDT at 54 EFPY is projected to remain well below the RT NDT from the SER for BWRVIP-74, thus the Columbia axial weld failure frequency meets the acceptable criteria.
Disposition The TLAA for the reactor vessel axial weld failure probability has been projected to the end of the period of extended operation.
A.2.2.2 METAL FATIGUE Fatigue evaluations for mechanic al components are identified as TLAAs; therefor e, the effects of fatigue must be addressed for license renewal. Fatigue is an age-related degradation mechanism caused by cyclic duty on a component by either mechanical or thermal loads. The ASME Boiler and Pressure Vessel Code requires evaluation of transient thermal and mechanical load cycles for Class 1 components. Cumulative usage factors for Class 1 components are calculated base d on normal and upset design transient definitions. The design transients used to generate cumulative usage factors for Class 1 compon ents are contained in FSAR Section 3.9.1.1. Energy Northwest is required to m onitor design transients listed in FSAR Table 3.9-1 to ensure that plant components ar e maintained within the design limits.
Calculation of fatigue usage values is not required for non-Class 1 SSCs. Instead, stress intensification factors and lo wer stress allowables are used to ensure components are adequately designed for fatigue.
The reactor coolant environmenta l effects of fatigue on plant co mponents were also evaluated.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-29 The design cycles for Columbia are summarized in FSAR Section 3.9 and FSAR Table 3.9-1. Energy Northwest counts all fatigue significant cycles, not only fo r the design transients listed in FSAR Table 3.9-1 but also for the analysis of other pl ant components. The events listed in FSAR Table 3.9-1 have been evaluated and in some cases regrouped for easier counting.
Faulted conditions listed in the FSAR are not used in the fatigue analyses and are not counted.
Additional transients determined to be fatigue significant after the original design have been added to the counting procedure, while FSAR Table 3.9-1 lists the original design cycles. The projected number of occurrences of design transients to 60 years determined that some analyzed numbers of transients may be exceeded. These proj ections were done using linear extrapolation from the beginning of plant lif
- e. Recent operating e xperience suggests lower projections and as addi tional operating data is accumulated, subsequent projections will refine the number of cycles expected in 60 years. Energy Northwest mana ges fatigue using the Fatigue Monitoring Program to track transient cy cles and require correc tive action before any analyzed number of cycles is reached.
A.2.2.2.1 Reactor Pressure Vessel Fatigue Analyses The reactor vessel assembly consists of the reactor pressure vessel (RPV), the vessel support skirt, the shroud support, nozzles , penetrations, stub t ubes, head closure fl anges, head closure studs, refueling bellows support, and stabilizer brackets. Design cumulative usage factors (CUFs) for the limiting RPV assembly locations are contained in design reports and were ca lculated based on the design transients. Energy Northwest manages fatigue for the RPV as sembly components using the Fatigue Monitoring Program to track transient cycles and requires corrective ac tion before any analyzed number of cycles is reached. Disposition The effects of aging on the intended functions of the RPV will be adequately managed for the period of extended operation by th e Fatigue Monitoring Program.
A.2.2.2.2 Reactor Pressure Vessel Internals
Fatigue analyses of the overall RPV internals (including the jet pump assemblies) were performed pre-startup as part of the plant desi gn. Component specific fa tigue analyses of the jet pumps were performed more recently to bound actual plant operati on. Each of these analyses is di scussed below.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-30 Reactor Vessel Internals Fatigue Analyses The RPV internals are describe d in terms of two assemblies: core support structures and reactor internals. Core support structures incl ude the shroud, shroud support (included as part of the reactor vessel for fatigue), core plate with core plate hold-down bolts, top guide, fuel supports, and control rod guide tubes. Reactor internals include the jet pump assemblies, jet pump instrumentation, feedwater spargers, vessel head spray nozzl e, differential pressure line, incore flux monitor guide tubes, surveillance sa mple holders, core spra y line (in-vessel) and spargers, incore instrument hous ings, low pressure coolant inj ection coupling, steam dryer, shroud head and steam separator assembly, guide rods, and cont rol rod drive thermal sleeves. The normal, test, and upset service load cycles used for the design and fa tigue analysis for the core support structures and reactor internals are shown in FSAR Table 3.9-1. Calculation of CUFs for the reactor internals was perfor med as part of a NSSS design evaluation. Review of the RPV internals in association with power uprate determined that stresses on the vessel internals remained well below all limits. No recalculation of cumulative usage factors was determined to be required. Energy Northwest manage s fatigue using the Fatigue Monitoring Program to track transient cycles and require corrective action before any analyzed number of cycles is reached.
Disposition The effects of aging on the intende d functions of the RPV internals will be adequately managed for the period of extended operation by the Fatigue Monitoring Program.
Jet Pump Fatigue Analyses In August 2000, Columbia operate d for a period of time with the recirculation pumps in an unbalanced mode (pump speeds different by more than 50 percent). The effect of that flow imbalance on the jet pumps was an additional accumulation of fatigue usage. As a result of inspections during the Spring 2001 outage (R-15), a fatigue analysis of the jet pumps was performed and cumulative usage factors were determined.
Jet pump clamps were installed during the 20 05 outage (R-17) to minimize flow induced vibration. These clamps gr eatly reduced the future potential for riser brace fatigue. As a result of evaluations after the 2007 outage the usage factor s were extended to 60 years.
The maximum CUF of the jet pump risers for 60 years of operation is projected to remain below the fatigue limit. Energy Northwest ma nages fatigue using th e Fatigue Monitoring Program to track transient cycles and require corrective ac tion before any analyzed number of cycles is reached. The Fati gue Monitoring Program credits the BWR Vessel Internals Program C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-31 to monitor the jet pump gaps. Together, these ac tions effectively manage the fatigue of the jet pumps through the period of extended operation.
Disposition The effects of aging on the inte nded functions of the jet pumps will be adequate ly managed for the period of extended operation by the Fatigue Monitoring Program.
A.2.2.2.3 Reactor Coolant Pressure Boundary Piping and Piping Co mponent Fatigue Analyses The Class 1 boundary encompasse s all reactor coolant pressure boundary piping (pipe and fittings) and in-line components subject to ASME Section XI, Subsection IWB, inspection requirements. Fatigue analyses of Class 1 piping are based on the transients found in the Columbia piping specifications that are in turn based on the design transients listed in FSAR Section 3.9. Potential high energy line break (HELB) intermediate locations can be eliminated based on CUFs of less than 0.1 if other stress criteria are also met. The usage factors, as calculated in the design fatigue analyses, account for the desi gn transients assumed fo r the original 40-year life of the plant. Therefore, th e determination of CUFs used in the selection of postulated high energy line intermediate break locations are TLAAs. The Fa tigue Monitoring Program will identify when the transients fo r piping systems are approaching their analyzed number of cycles. Prior to any transient exceeding its analyzed number of cycles for a piping system, the associated analyses will be reviewed to determine whether any additional locations need to be designated as postulated HELB locations. All Class 1 piping was reviewed for the power up rate. The evaluation determined that there was adequate margin in each system to accomm odate the power uprate. Design fatigue usage for 40 years of operation and projected fatigue usage for the period of extended operation are established for the limiting reactor coolant pressure boundary components.
A review of documentation foun d several fatigue analyses fo r Class 1 valve stress reports found fatigue analyses that were TLAAs. The fatigue usage for thos e valves is based on transients that are tracked by the Fatigue Monitoring Program.
Metal fatigue for all Class 1 re actor coolant pressure boundary piping and in-line components is managed by the Fatigue M onitoring Program. The Fati gue Monitoring Program will identify when the transients fo r piping systems are approaching their analyzed numbers of cycles. Prior to any transient exceeding its analyzed number of cycles for a piping system, the design calculations for that system will be revi ewed and appropriate actions will be taken.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-32 Disposition The effects of aging on the inte nded functions of the reactor coolant pressure boundary piping and components will be adequate ly managed for the period of extended operation by the Fatigue Monitoring Program.
A.2.2.3 NON-CLASS 1 COMPONENT FATIGUE ANALYSES The non-Class 1 mechanical components susceptible to fati gue fit into one of two major categories: (1) piping and in-lin e components (piping, valves, t ubing, traps, thermowells, etc.)
or (2) non-piping components (vessels, he at exchangers, tanks, pumps, etc.). Non-Class 1 components that are Quality Group B or C are designed and constructed as ASME Section III Code Class 2 and 3, respectively. The design of ASME Class 2 and 3 piping systems incorporates a stress range reduction factor for determining acceptability of piping design with respect to thermal stresses. Non-Class 1 component s designated as Quality Class D are designed to ANSI B31.1, which also incorporates stress range reduction factors based upon the number of thermal cycles. In general, a stress range reduction factor of 1.0 in the stress analyses applies for up to 7,000 thermal cycles. The al lowable stress range is reduced by the stress range reduction factor if the number of thermal cycles ex ceeds 7,000. If fewer than 7,000 cycles are expected through the period of extende d operation, then the fatigue analysis (stress range reduction factor) of record will remain valid through the period of extended operation.
Because none of the non-Class 1 vessels, heat exchangers, storage tanks, or pumps were designed to ASME Section VIII, Division 2 or ASME Section III, Subsection NC-3200, no fatigue evaluation is re quired. Therefore, th ere are no fatigue TLAAs for these components.
The fatigue evaluation of non-Cla ss 1 piping and in-line compone nts evaluated the associated operating temperature against the threshold temperature value for fatigue of the material. If the threshold temperature value was exceeded, then the number of transient cycles for the piping or in-line component was projected. In each case, the num ber of projected cycles for 60 years was found to be less than 7,000 for piping and in-line components whose temperatures exceed threshold values. Theref ore, fatigue for non-Cla ss 1 piping and in-line components remains valid for the period of extended operation.
Disposition The TLAA for non-Class 1 component fatigue an alyses remains valid for the period of extended operation.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-33 A.2.2.4 EFFECTS OF REACTOR COOLANT ENVIRONMENT ON FATIGUE LIFE OF COMPONENTS AND PIPING Applicants for license renewal are required to address the reactor coolant environmental effects on fatigue of plant components. The minimum set of components for a BWR of Columbia's vintage is derived from NUREG/CR-6260, "Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components," as follows: 1. Reactor vessel shell and lower head 2. Reactor vessel feedwater nozzle 3. Reactor recirculation piping (including inlet and outlet nozzles) 4. Core spray line reactor vessel nozzle and associated Class 1 piping 5. Residual heat removal return line Class 1 piping 6. Feedwater line Class 1 piping Energy Northwest has analyzed these locations for the effe cts of the reactor coolant environment on fatigue in support of license renewal. Energy Northwest has also analyzed other limiting components beyond those locati ons indentified in NUREG/CR-6260 for the effects of the reactor coolant environment. Original fatigue usage calculations were reviewed, and the transient groupings and load pairs used in those analyses were carried over to the environmentally-assisted fatigue analyses, with revised non-e nvironmentally assisted usage factors determined. An effective fatigue life adjustment factor, F en, that considers a time weighted average of operation with normal water chemistry and hydrogen water chemistry over 60 years of operation, was determined for each load pair an alyzed for the components. The fatigue life adjustment factors were applied to the revised component load pair usage factors, and the environmentally-adjusted usage fa ctors were summed to obtain environmentally-adjusted CUFs to verify acceptability of the components for the period of extended operation.
Using fatigue data projecte d by the Fatigue Monitoring Program and the methodology summarized above, the limiting locations were evaluated. None of the locations evaluated have an environmentally adjusted CUF of gr eater than 1.0 during the period of extended operation.
For the period of extended operati on, on an ongoing basis, ensure that all the limiting locations in Class 1 components and Class 1 systems have been evaluated for the effect of reactor water environment.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-34 The aging effect of fatigue, including consideration of the environmental effects, will be adequately managed for the period of extended operation using the Fatigue Monitoring Program. Disposition The effects of environmentally-assisted fatigue on the intended functions of the NUREG/CR-6260 and other limiting locations will be adequately managed for the period of extended operation using the Fatigue Monitoring Program.
A.2.2.5 ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT Environmental qualification analyses for electrical equipment are identified as TLAAs. NRC regulation 10 CFR 50.49, "Environmental Qualific ation of Electric E quipment Important to Safety for Nuclear Power Plants," requires licensees to identify electrical equipment covered under this regulation and to maintain a qualification file demons trating that the equipment is qualified for its application and will perform its safety function up to the end of its qualified life. The EQ Program implements the requi rements of 10 CFR 50.49 and will be used to manage the effects of aging on the intended functions of the components associated with environmental qualification TLAAs for the period of extended operation.
Disposition The effects of aging on the inte nded functions of the environm entally qualified components will be adequately managed for the period of extended operation by the EQ Program.
A.2.2.6 FATIGUE OF PRIMARY CONTAINMENT, ATTACHED PIPING, AND COMPONENTS The Primary Containment and atta ched piping and components sus ceptible to fatigue resulting from the effects of plant transients are evaluated below.
A.2.2.6.1 Primary Containment The cycles used in the fatigue evaluation of the containment components are provided in FSAR Table 3A.4.1-3. No operating basis earthquakes have been experienced by Columbia through 2007, and the containment analys is for five operati ng basis earthquakes remains valid for 60 years of plant operation. The safe shutdown earthquake and post-loss of coolant accident (LOCA) chugging are once in a li fetime events and are not pr ojected to occur during the extended period of operation.
Safety relief valve actuations have been projected through 60 years of operation based on the number of actual events through 2007. The fatigue analyses performed using these events will remain valid for the pe riod of extended operation.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-35 As the cycles on which the containment fatigue analysis is based will not be exceeded for 60 years of operation, these analyses will remain valid for the period of extended operation.
Disposition The TLAA associated with fatigue of the containment remains valid for the period of extended operation.
A.2.2.6.2 ASME Class MC Components Class MC components include th e primary containment vessel shell, large openings (equipment hatch, personnel hatche s, and access hatch), penetrations (all except the large openings), and attachments (pipe supports in the wetwell, welding pads in the drywell, supports for the stabilizer truss, seal and shear lugs at the drywell floor, supports for the downcomer bracing system, pipe whip supports, radial beam suppor ts, cap truss supports, ca twalks, monorail, and platforms). The Class MC components were analyzed for fatigue using the transients listed in FSAR Table 3A.4.1-3. As these cycles will not be exce eded for 60 years of operation, the Class MC component fatigue analysis will remain valid for the period of extended operation. A specific fatigue analysis was performed for the main steam penetrations using the transients listed in FSAR Table 3A.4.1-3. This analysis will remain valid for the period of extended operation as these cycles will not be exceeded for 60 years of operation.
The effects of power uprate on th e containment system response we re reviewed and determined to be negligible. The contai nment peak pressure values re main virtually unaffected by the power uprate and extended lo ad line limit. The LOCA cont ainment dynamic loads are not affected by power uprate, and sa fety relief valve containment lo ads will remain below their design allowables. (See FSAR Section 3A
.) All events, including safety relief valve actuations, for 60 year s of operation are projected to remain below the containment cyclic basis from FSAR Table 3A.4.1-3. Consequently, the analysis of the Class MC containment compon ents remains valid for the period of extended operation.
Disposition The TLAAs for fatigue of the ASME Class MC components remain valid through the end of the period of extended operation.
A.2.2.6.3 Downcomers Although not an ASME Code requirement, a fatigue evalua tion of the downcomers was performed. The fatigue eval uation of the downcomer lines in the wetwell air volume was C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-36 based on the number of cycl es presented in FSAR Table 3A.4.1-3. The maximum fatigue usage factor for the downcomers is provided in FSAR Table 3A.4.2-4 and FSAR Table 3A.4.2-5. All events, including safety relief valve actuations, for 60 year s of operation are projected to remain below the containment cyclic basis from FSAR Table 3A.4.1-3. Consequently, the analysis of the downcomers remains valid for the period of extended operation.
Disposition The TLAA for fatigue of the downcomers rema ins valid through the e nd of the period of extended operation.
A.2.2.6.4 Safety Relief Valve Discharge Piping Although not an ASME Code requirement, a fatigue evaluation of the safety relief valve (SRV) discharge piping was performed.
The fatigue evaluation used the number of cycles as presented in FSAR Table 3A.4.1-3. The maximum fatigue usage factor for all 18 SRV discharge lines in the wetwell air volume is below the ASME allowable limits per FSAR Section 3A.4.2.4.6. The SRV actuations for 60 years of operation are projected to remain below the containment cyclic basis from FSAR Table 3A.4.1-3. Consequently, the analys is of the SRV discharge piping remains valid for the period of exte nded operation.
Disposition The TLAA for fatigue of the SRV discharge piping remains valid through the end of the period of extended operation.
A.2.2.6.5 Diaphragm Floor Seal The diaphragm floor seal is located at the in side surface of the primary containment vessel periphery. It provides a flexib le, pressure tight seal between the primary containment vessel and the diaphragm floor and is capable of accommodating differentia l thermal expansion between them.
The fatigue evaluation was perfor med using the cycles in FSAR Table 3A.4.1-3. The maximum cumulative usage factor is less than the fatigue limit per FSAR Table 3A.4.1-5. All events, including SRV actuations, for 60 years of operation are pr ojected to remain below the containment cyclic basis from FSAR Table 3A.4.1-3. Consequently, the analysis of the diaphragm floor seal remains valid for the period of extended operation.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-37 Disposition The TLAA for fatigue of the containment diaphragm floor seal remains valid through the end of the period of extended operation.
A.2.2.6.6 ECCS Suction Strainers The original Columbia ECCS suction strainers were replace d with a new strainer design constructed from cold-worked au stenitic stainless steel. A linear elasti c fracture mechanics analysis was performed to bound all the martensitic material in the suction strainer screens. A crack depth was assumed based on the depth of the Alpha Prime martens ite in the strainer screen material. Cyclic stresses were considered in the crack growth analysis of the suction strainers. The fatigue crack evaluation determined that the assumed cracks will not propagate to a critical size for the remaining life of the plan
- t. The maximum computed stress intensity value (K) was less than that required to cause cracking in Alpha ma rtensite formed in aust enitic stainless steel.
The stress value conservatively included direct pressure and inertia l components from SRV actuation, operating basis eart hquake (OBE) loads, and SRV steam chugging. (See FSAR Table 3A.4.1-3
.) All events, including safety relief valve actuations, for 60 year s of operation are projected to remain below the containment cyclic basis from FSAR Table 3A.4.1-3. Consequently, the analysis of the ECCS suction strainers remains valid for the period of extended operation.
Disposition The TLAA for crack growth of the ECCS suction strainers remains valid through the end of the period of extended operation.
A.2.2.7 OTHER PLANT-SPECIFIC TIME-LIMITED AGING ANALYSES The TLAAs that do not fit into any of the pr evious major categories are evaluated below.
A.2.2.7.1 Reactor Vessel Shell Indications Two indications in the reactor vessel shell were identified using ultras onic inspection methods during the 2005 inservice inspec tions. The indications were present in past inservice inspection examinations, but became rejectable under current ASME Section XI, IWB-3610 requirements. The rejected i ndications were evaluated and determined to be acceptable for continued service without repair, as reported to the NRC. The indications were evaluated per the guidelines of ASME Section XI, IWB-3610, wh ich include acceptance criteria based on the C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-38 applied stress intensity factor s, using conservative assumpti ons in the applied stresses to determine the stress intensity factors for comparison to Code allowables. This conservative evaluation calculated a fatigue crack growth at the end of 33.1 EFPY vessel service life that is insignificant in comparison to the bounding initial crack size. It also determined that the applied stress intensity factor is well below the allowable stress intensity factor. The calculation is based on time-limited assumptions of neutron fluence and SRV blowdown cycles for 40 years. While it is not expected th at the applied stress inte nsity factor will exceed the allowable fracture toughness during the period of extended operation, cracking near the subject reactor vessel welds is managed by the Inservice Inspection (ISI) Program. Energy Northwest will re-evaluate the indicatio n based on the results of the 2015 inspection and either project this analysis through the period of extended operation or continue augmented inspections as required by the ASME code.
Disposition Cracking of the reactor vessel shell near welds BG and BM will be adequately managed through the period of extended operation by the Inservice Inspection (ISI) Program.
A.2.2.7.2 Sacrificial Shield Wall FSAR Section 3.8.3.6 provides a value of neutron fluence for the outside face of the sacrificial shield wall that is based on 40 years of pl ant operation. Projecti ons done for 60 years of operation, including increase in fluence due to power uprate, determined that the estimated neutron fluence on the sacrificia l shield wall will remain below the threshold for neutron damage of concrete a nd reinforcing steel. Therefore, the sacrificial sh ield wall can be expected to perform its radiation shielding function through the period of extended operation.
Disposition The TLAA associated with the sacrificial shield wall fluence has been projected to the end of the period of extended operation.
A.2.2.7.3 Main Steam Flow Restrictor Erosion Analyses The main steam line flow restrictors are designe d to limit coolant flow rate from the reactor vessel (before the MSIVs are clos ed) to less than 200 percent of normal flow in the event of a main steam line break outside the c ontainment. Erosion of a flow restrictor is a safety concern since it could impair the ability of the flow restrictor to limit vessel blowdown following a main steam line break. Since er osion is a time-relate d phenomenon, the analysis for the effect it has on the flow restrictors over the life of the plant is a TLAA. Cast stainless steel (SA351, C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.2-39 Type CF8) was selected for the steam flow restrictor material beca use it has excellent resistance to erosion-corrosi on from high velocity steam.
The erosion of the main steam fl ow restrictors has been project ed for the period of extended operation. The projection concludes that after 60 years of erosion on the main steam flow restrictors, the choked flow will still be less than 200 percent of normal flow. Therefore, the main steam flow restrictors w ill continue to perform their in tended function a nd the existing accident radiological release analysis will rema in valid for the period of extended operation.
Disposition The TLAA for erosion of the main steam line flow restrictors has been projected to the end of the period of extended operation.
A.2.2.7.4 Core Plate Rim Hold-Down Bolts The NRC safety evaluation report that references BWRVIP-25, "BWR Core Plate Inspection and Flaw Evaluation Guidelines," for license renewal identifies loss of preload on the core plate rim hold-down bolts as one of the TLAA th at must be addressed by applicants seeking license renewal.
Disposition At least two years prior to the period of exte nded operation, Energy Nort hwest will install core plate wedges unless: a site-specific analysis is approved by the NRC that resolves core plate bolt loss of preload due to both stress relaxation and cracking, or an NRC approved method is developed to inspect the core plate bo lts for cracking and a site-specific analysis for loss of preload due to stress relaxation of the core plate bolts is approved by the NRC. A.2.2.7.5 Crane Load Cycle Limit All in-scope cranes at Columbia were de signed to Crane Manufacturers Association of America (CMAA) Specification 70, "Specification for Electric Ov erhead Traveling Cranes
" which provides a design load cycle limit based on service class for the associated cranes. This load cycle limit for each crane wa s identified as a potential TLAA.
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 A.3-1 Disposition To address this potential TLAA a 60-year projec tion of load cycles was developed for all cranes in the scope of license renewal and co mpared to the design load cycle limits of CMAA
- 70. For all cranes the 60-year projection of load cycles is within the applicable design load cycle limit of CMAA 70. Therefore, this TLAA remains valid for the period of extended operation.
A.3 REFERENCES A.3-1 BWROG Report GE-NE-523-A71-0594-A, Rev 1, "Alternate BWR Feedwater Nozzle Inspection Requi rements," May 2000 A.3-2 EPRI Report No. 1011838, "Reco mmendations for An Effective Flow Accelerated Corrosion Program (NSAC-202L-R3)," May 2006 A.3-3 EPRI TR-103834-P1-2, "Effects of Moisture on the Life of Power Plant Cables," August 1994